ML092150636

Response to Request for Additional Information GSI-191/GL 2004-02 (TACs MC4705/4706) Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors

ML092150636
Person / Time
Site:	Point Beach
Issue date:	07/31/2009
From:	Meyer L Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-04-002, GSI-191, NRC 2009-0077, TAC MC4705, TAC MC4706
Download: ML092150636 (188)

Text

July 31,2009 NRC 2009-0077 GL 2004-02 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Res~onse to Reauest for Additional lnformation GSI-I 911GL 2004-02 (TAC NOS. MC4705147061 Potential lm~act of Debris Blockaae on Emeraencv Recirculation Durina Design Basis Accidents at Pressurized Water Reactors

References:

(1)

FPL Energy Point Beach, LLC, Letter to NRC dated February 29, 2008, Supplemental Response to GL 2004-02, Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (ML080630613)

(2)

FPL Energy Point Beach, LLC, Letter to NRC dated June 9, 2008, Supplemental Response to GL 2004-02, Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (ML081620337)

(3)

NRC Letter to FPL Energy Point Beach Nuclear Plant Units 1 and 2 GSI-1911GL 2004-02, Request for Additional lnformation (TAC NOS. MC4705/4706), dated January 7,2009 (ML083300173)

(4)

FPL Energy Point Beach, LLC, Letter to NRC dated April 7,2009, Response to Request for Additional Information, GSI-1911GL 2004-02 (TAC NOS. MC470514706) Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (ML090980523)

NextEra Energy Point Beach, LLC (NextEra), (formerly FPL Energy Point Beach, LLC),

previously submitted supplemental responses to Generic Letter (GL) 2004-02 in References (1) and (2). Reference (3) contains a request for additional information (RAI) based upon NRC Staff reviews of References (I) and (2).

Enclosure I contains the NextEra response to the RAI questions pending final debris generation and transport analyses results. To enable NRC staff review, the responses include those previously provided by Reference (4). lnformation that has been changed or appended since submittal of Reference (4) is indicated by a vertical bar in the left hand margin. Enclosures 2 through 6 are provided in support of information summarized in Enclosure 1.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.

If you have questions or require additional information, please contact Mr. James Costedio at 9201755-7427.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on July 31, 2009.

Very truly yours, NextEra Energy Point Beach, LLC Enclosures cc:

Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE I NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-1911GL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS The following information is provided by NextEra Energy Point Beach, LLC (NextEra), in response to the NRC staff's request for additional information dated January 7, 2009.

Question I Please provide the insulation material(s) for the reactor vessel. Please state whether the debris quantities generated by breaks at reactor vessel nozzles that reach the strainer are bounded by the debris that transpotts from other breaks that have already been evaluated. If the debris quantities from previously breaks are not bounding, please evaluate the effects of the reactor vessel nozzle break on strainer head loss.

NextEra Response The reactor vessels for both Point Beach Nuclear Plant (PBNP) units are insulated with reflective metal insulation (RMI). This includes the vessel circumferential insulation, the lower head and the upper reactor vessel head. The jacketing and the foils of the RMI are stainless steel.

The insulation on the piping connected to the vessel nozzles is RMI with the exception of an approximately 17" wide removable belt around each of the pipe-to-nozzle welds to facilitate inservice inspections. This belt of insulation is fibrous NUKON@ in most locations and ~emp-at@

with a sewn envelope of asbestos bearing cloth in two locations. The removable belts of fibrous insulation have a volume of approximately 5 ft3 each, for a total of approximately 20 ft3 for the four nozzles on each reactor vessel.

Breaks originating at the reactor vessel nozzles have not been explicitly modeled to determine the amount of debris that would be generated. However, the quantity of fibrous and particulate debris is reasonably bounded by other modeled breaks.

The reactor vessels are enclosed within the steel reinforced concrete primary shield wall. Jets from a break originating at the vessel nozzles would be expected to emanate in a predominantly radial direction outward from the nozzle rather than axially along the pipe wall. Axial offsetting between the reactor coolant system (RCS) piping and the reactor vessel nozzles would be substantially limited by the reactor vessel supports and the reactor primary shield wall penetration housing each of the loop pipes. As such, direct jetting axially along the pipe through an RCS loop piping penetration in the primary shield wall is expected to be minimal.

Insulation debris stripped and ejected through penetrations in the primary shield wall by a loss of coolant accident (LOCA) at the reactor vessel nozzles would be mostly RMI debris. The majority of the RMI debris generated would be expected to remain within the primary shield wall, an inactive sump. The RMI debris that may be ejected would not be subject to significant transport by the low velocity flows in the active sump. As such, the debris generated is considered to be bounded by the relatively large quantities of fibrous and particulate debris that could be generated by a break of RCS piping within the RCS loop compartments so no further detailed evaluation of the effects has been performed.

Page I of 40

Question 2 Please provide the information concerning the debris characteristics analysis that was requested in the U, S. Nuclear Regulatory Commission (NRC) staff revised content guide.

NextEra Response Revised Content Guide for Generic Letter 2004-02 Supplemental Responses, November 2007, (ML073110278), identifies the following as required specific information regarding methodology for demonstrating compliance:

"Debris Characteristics The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

Provide the assumed size distribution for each type of debris.

a Provide bulk densities (i.e., including voids between the fiberslparticles) and material densities (i.e., the density of the microscopic fiberslparticles themselves) for fibrous and particulate debris.

a Provide assumed specific surface areas for fibrous and particulate debris.

a Provide the technical basis for any debris characterization assumptions that deviate from NRC-approved guidance."

During a December 22,2008, telephone conference held between representatives of the NRC staff and FPL Energy Point Beach, it was acknowledged that much of the requested information pertaining to the characteristics of the debris would not be necessary, and that some information was not available, as a result of site efforts to qualify screen performance by testing rather than analysis. It was also acknowledged that pertinent details of the materials and surrogates used in the tests are needed, and in particular, details of how these materials and surrogates were prepared for the tests.

I Debris was added to the flumes for the screen test by weight. The weights were calculated using the volumes from the debris generation and transport calculations, multiplied by conservatively assumed as-manufactured densities of various materials. The information in the table below was excerpted from the strainer design basis loading test plan. The plan lists the debris types as determined by the debris generation analysis, the assumed densities for these debris types, the corresponding surrogate materials used in the flume tests, and the method of preparation for the surrogates. The tabulated densities were previously approved in Nuclear Energy Institute, NEI-04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, dated May 28,2004, (ML041550279, ML041550332, ML041550359 and ML041550380).

Page 2 of 40

Materials Used for Screen Testing For further details about the preparation of the surrogate materials, their size distributions (fines, smalls, larges and intact), and the technical basis for the use of the various surrogates, refer to Performance Contracting Inc. Letter to NRC, dated March 25, 2009, PCl-6016-02.01,, Sure-Flow@ Suction Strainer - Testing Debris Preparation and Surrogates, I SFSS-ID-2007, Revision 4 (Proprietary) (ML090900476).

All of the asbestos fiber (assumed to be 90% of the insulation mass) was assumed to be fine fibers and was processed as such for the screen test. The intent was to allocate 50% of the asbestos fiber as fines and 50% as smalls. The error in allocating these as fines was conservative.

Fractions are conservative from two aspects; the high percentage of assumed fiber content, and the assumption that a major fraction of all of that fiber is reduced to a fine and transportable form.

It is expected that a substantial portion of the asbestos insulation would remain in lumps or intact pieces and not be reduced to individual fibers of fine dimensions. Destructive testing was not performed.

Surrogate Processing Shredder Chunks >1"x4" Wood chipper Shredder Chunks >Ivx4 Wood chipper Shredder Wood chipper Shredder Wood chipper Shredder Shredder Powdered Powdered NIA NIA Powdered Powdered Powdered Powdered 1/64 - l/d' chips During a June 22, 2009, telephone conference held between representatives of the NRC staff and NextEra, additional information was requested detailing the size distributions for each debris type and specifying whether zones of influence (ZOls) are to be reduced from the approved guidance in NEI 04-07, Volume 11. The following information is provided based on previously performed debris analyses. No changes in approach or methodology are expected for the analyses to be performed in the future.

Surrogate Test Material Ceramic fiber NUKON@

NUKON@

Baked-out NUKON@

Temp at@

Temp Matw Temp at@

Owens Corning Fiberglass Owens Corning Fiberglass 10 pcf MW Fiber 10 pcf MW Fiber Baked-out NUKON~

Cal-Sil Cal-Sil PWR Dirt Mix Tin Powder Walnut shells Walnut shells Walnut shells Walnut shells Acrylic Debris Type Asbestos fiber (90% fines)

NUKONw (large)

NUKONw (small)

NuKON~ (fines)

Temp at@ (large)

Temp MatQ (small)

Temp Matw (fines)

Fiberglass (small)

Fiberglass (fines)

Mineral Wool (small)

Mineral Wool (fines)

Latent fibers Asbestos particulates (1 0%)

Cal-Sil Latent dirt & dust Zinc Coatings Aluminum coatings Alkyd coatings Unqualified Epoxy Qualified Epoxy in ZOI Degraded Epoxy (chips)

Page 3 of 40 Density (lb/ft3)

I 0 2.4 2.4 2.4 11.8 11.8 11.8 5.5 5.5 8.0 8.0 NIA (direct weight from analysis) 10.0 14.5 NIA (direct weight from analysis) 457 94 90 94 94 94

Debris sizing for the blast, blowdown and pool fill phases of the accident conforms to the approved guidance of NEI 04-07, Section 3.4.3.3. Two groupings are used ("large" and "small").

Forty-percent of the NUKON' and ~emp-at' were assumed to be "large" debris, with the balance, (60%), being reduced to "small" debris.

I RMI debris was assumed to be 25% large pieces with 75% reduced to "small fines".

Calcium silicate (Cal-Sil) insulation was conservatively assumed to be 100% reduced to particulates to bound erosion effects that may occur.

Other fibrous debris types (e.g., mineral wool, generic fiberglass, etc), were to be 100% reduced to "small fines".

NextEra is supporting industry efforts to demonstrate by testing andlor analysis that jacketed NUKON' will remain intact at substantially less than the 17D ZOI endorsed in NEI 04-07. Previous efforts at PBNP have credited a reduced ZOI radius of 5 pipe diameters for jacketed NUKON' insulation. Current efforts will eliminate NUKON@' to the extent necessary to ensure that it does not remain within the ZOI of large diameter, limiting pipe breaks. In particular, NUKON' is being removed from RCS piping, and from steam generator channel heads to ensure that there is no remaining NUKON' within a 50 radius of potential large bore break locations. Contingencies are in place to remove additional NUKON' to the extent necessary, including extended removal from the Unit 1 pressurizer and steam generator vertical sections, to preclude NUKON' involvement in design basis limiting ZOls.

Question 3 Please provide the information concerning the debris transport analysis that was requested in the NRC staff revised content guide.

NextEra Response The NRC staff revised content guide requests the following information:

I "e. Debris Transport The objective of the debris transport evaluation process is to estimate the fraction of debris that would be transported from debris sources within containment to the sump suction strainers.

o Describe the methodology used to analyze debris transport during the blowdown, washdown, pool-fill-up, and recirculation phases of an accident.

o Provide the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

o ldentilj any computational fluid dynamics codes used to compute debris transport fractions during recirculation and summarize the methodology, modeling assumptions, and results.

o Provide a summary of, and supporting basis for, any credit taken for debris interceptors.

o State whether fine debris was assumed to settle and provide basis for any settling credited.

Provide the calculated debris transport fractions and the total quantities of each type of debris transported to the strainers. "

Page 4 of 40

The following response reflects the methodology used in previously completed debris transport analyses. The previous analyses are being re-performed to reflect the planned elimination of fibrous insulation. It is expected that the pending analyses will use the same approach as the previous analyses without exception.

Due to variations in sump geometry, insulation types, and piping layouts, separate analyses have been performed for each unit. The analysis for each unit consists of two parts:

I 1. A distribution of debris due to the initial blowdown and subsequent pool fill, and

2. The transportation of debris toward the strainers due to containment washdown and containment sump recirculation.

I These two parts are discussed separately below.

I Blowdown I Pool-Fill The approach used derives from the approved guidance of NEI 04-07. However, the approved guidance lacks several important details and necessary assumptions. Therefore, a description of the entire process is provided below.

Debris transports due to several factors, one of the most significant being the size of the debris.

Two size groupings (small and large) are used in the analysis, consistent with NEI 04-07:

I Debris Sizing Fractions Used In Blowdown 1 Pool Fill Analysis Small Debris The small debris (defined as S 4 along its longest dimension) is expected to become suspended in the air, moving upward through openings in robust barriers surrounding the break location, as well as downward. The pressure wave will likely carry the small debris a substantial distance from the break.

Large 25%

0%

0%

0%

40%

0%

40%

Insulation Type RMI Asbestos CalSil Fiberglass

~emp-at@/ lnsulbatte Mineral Wool NUKON@

Although small debris originating in the RCS loop compartments will be widely distributed throughout containment, some of the distribution will be impeded by physical constraints within the loop compartment. Examples of such impediments include the compartment walls, the steam generators and reactor coolant pumps and the piping. In addition, the compartments contain numerous bar grate work platforms. These grates overlay the RCS components and will tend to impede small debris that is blasted up toward the refueling floor elevation. To account for these impediments, 50% of the total small debris generated is assumed to remain in the loop compartment at the end of the blast and pool fill phases. The remainder is assumed to leave the RCS loop compartment and be distributed throughout the containment building according to the sizes of the available openings.

Small Fines 75%

100%

100%

100%

60%

100%

60%

Page 5 of 40

While 50% of the "smalls" are assumed to be held up on interior structures, the testing that was performed (and is more fully described in response to Question 4) introduced a quantity of fines into the test flume that was based on a fraction of the full inventory of all fibrous debris generated within the containment. No reduction in testing fines was made as a result of the "small" debris assumed to be held up outside of the sump pool.

To determine the distribution of the small debris leaving the RCS loop compartments, the areas of the openings out of the compartments were calculated. For example, in Unit 2 on the Loop A side of containment there are seven openings. Five of the openings extend outward from the bottom of the RCS loop compartment into the sump level of the containment. The other two openings pass through the steam generator vault and reactor coolant pump cubicle. These openings provide a path through which the debris may reach the refueling floor. The following table illustrates the contribution of these openings to the fractional distribution of debris leaving the loop compartment:

Distribution of Small Debris Leaving an RCS Loop Compartment The small debris transported to the refueling floor is distributed with preference toward the side of containment closest to the pipe break. The refueling floor surface area is divided into halves by a line running through the approximate center of the containment separating the two loop compartments. For breaks originating in a loop compartment, three-quarters of the debris is assumed to remain on the half of the refueling floor closest to the break, while one-quarter is assumed to be transported and evenly distributed on the remaining half of the refueling floor.

Once small debris is distributed to the various zones as described above, it is assumed to be evenly distributed throughout the zone by the combination of blowdown and pool fill. The resulting 1 distribution logic tree for this example follows.

Opening Designation A

B C

D E

RCP Cubicle SG Vault Page 6 of 40 Passing To Zone #

1 I 1 (Sump elev.)

108 (Sump elev.)

108 (Sump elev.)

109 (Sump elev.)

I 10 (Sump elev.)

Refueling Floor Refueling Floor Area (ft2) 44.9 81.I 56.0 56.0 104.4 50.0 254.2 Total Open Area Percent of total opening area 6.9%

12.5%

8.7%

8.7%

16.1%

7.7%

39.3%

646.6 ft2

Remains in the RCS Loop Compartmenl Example Distribution Logic of Small-Sized Debris that Originate in RCS Loop Compartment; Distribution due to Blow Down and Pool Fill Transport Utstrtbuted on half of refueling floor toward pipe break 23.5% x 25% = 5.9%

50% X 47% = 23.5%

The large debris (defined as > 4 along its shortest dimension) is expected to be largely influenced by gravity and falls to the containment floor. Fifteen-percent of this debris is assumed to be large enough to remain on the floor below the break and not be transportable by blast or pool fill effects.

Eighty-five percent of the debris is assumed to be subject to pool fill transport and will be pushed toward the compartment walls and compartment openings that exist between the loop compartment and the sump elevation of containment. To determine how much of this debris leaves the loop through the openings and how much gets intercepted by loop walls, the lengths each opening and each span of wall are found. The following table illustrates how this was performed to determine the fractional distribution of large debris originating from the Unit 2 Loop A compartment:

50%

Page 7 of 40 Transports to Refuel Floor Distributed on half of refueling floor away from pipe break 50% x 53% = 26.5%

100%

LeavesLoop Compartment All Small Debris Transports through openings in the Loop compartment and is evenly distributed on sump level 50%

Example Distribution of Large Debris from an RCS Loop Compartment The resulting distribution logic tree for large debris is depicted below:

85% x 5.2% = 4.4%

iransports through openlngs to zone 11 I Percent of total opening length 5.2%

9.4%

6.5%

6.5%

12.1%

100%

Large Debris Length of Opening (ft) 6.42 1 I

.58 8.00 8.00 14.92 Opening Designation A

B C

D E

W l W2 W3 W4 W5 Total Length of Walls and Openings Passing to Zone #

1 I 1 (sump elev.)

108 (sump elev.)

108 (sump elev.)

109 (sump elev.)

I 10 (sump elev.)

Length of Wall Span 32.92 6.00 15.30 14.75 5.50 123.39 ft.

85%

I I

Intercepted by walls in cmpt.

85% x 15.9% = 13.5%

Transports through opening to zone 108 Initially on Loop Cmpt Floor; subject to pool fill transport 85% x 6.5% = 5.5%

I ransports tnrougn opening to zone IUY 85% x 12.1% = 10.3%

Transports through opening to zone 1 I 0 85% x 60.4% = 51.3%

Example Distribution Logic of Large-Sized Debris that Originate in RCS Loop Compartment; Distribution due to Blow Down and Pool Fill Transport 15%

Page 8 of 40 Retained in Loop Compartment Retained on Floor of Loop Compartment (too large to move during fill)

Washdown I Sump Recirculation The approved guidance of NEI 04-07 lacks substantial detail on how to perform washdown and recirculation analyses. The following information provides a complete narrative of the methods, assumptions, software, etc. previously used to perform the analyses. It is expected to continue to be used when revising analyses to account for the reductions in fibrous insulation.

Analytical Methodology The following outline presents the general methodology for performing the debris transport calculations for the PBNP Unit 2 containment following a Loss of Coolant Accident (LOCA). The methodology follows that outlined in NEI 04-07 and the associated NRC Safety Evaluation Report (SER).

I. Perform steady-state Computational Fluid Dynamics (CFD) simulation for a given break scenario.

2. Post-process the CFD results by plotting three-dimensional surfaces of constant velocity.

These velocities will correspond to the incipient transport velocities tabulated in NEI 04-07 for the debris generated in the LOCA scenario.

3. Project the extents of the three-dimensional surfaces of velocity onto a horizontal plane to form a flat contour. Automatically digitize a closed curve around the projected velocity contour and calculate the area within the curve.

4. Compare the area calculated in (3) to the total floor area of the zone containing the particular debris typelsize under consideration. This comparison gives the fraction of the floor area susceptible to transport.

5. Tabulate the results of each calculation to determine the total fraction of debris transported to the sump for each LOCA break scenario and each debris type.

Significant Assumptions The following general assumptions were made in the course of the debris transport calculations.

1. It was assumed that an equal amount of flow is drawn through all modules in each train. (The strainer array has flow control devices).

2. The flow from each break falls uninterrupted to the pool (i.e., the break flow does not impact any equipment, piping or structures). This is conservative for the purposes of flow analysis, and differs from the detailed evaluation of the potential for air entrainment that considers the presence of intervening structures.

3. Spray flow from the two containment spray headers was uniformly distributed across the refueling floor. The openings on and above the refueling floor received spray flow in proportion to the area of each opening.

4. No insulated piping or equipment exists in the sump that would significantly influence flow patterns in the pool during recirculation.

5. Stair treads (there are no risers) on the stairways entering the sump pool were not included in the model. These steps are effective in dissipating the spray flows running down the two stairwells, but do not provide a significant blockage to the horizontal flow patterns in the pool during recirculation.

Page 9 of 40

6. Stairwells are offset as necessary to ensure that stairwell spray flow could be projected onto the water surface separate from the spray flow arriving via the annular gap near the containment liner.

7. The floor drains at each elevation above the pool were assumed to be blocked. Spray flow impacting the refueling floor passed down to the lower levels over the edges of the openings at that elevation in proportion to the perimeter of each opening.

8. Spray flow reaching the level above the sump was directed through the two stairwells and the 3" gap around the periphery of the containment in proportion to their respective areas.

9. It was assumed that the spray flow that would normally enter the refueling cavity enters the accident sump through the two steam generator vault openings. This flow is proportioned between the two (2) steam generator vaults based on their respective areas. This maximizes the analyzed flow velocities by combining with break flow in the loop with the break. The refueling cavity drain discharges near one of the two strainer trains and bypasses the larger quantity of debris remaining on the floor of the loop compartment containing the break.

10. Details of the flow patterns through the 3" gap and through the two stairwells are not modeled.

The tapered containment wall and stair steps dissipate the momentum of these streams and the flow patterns entering the pool at these locations are assumed to be uniform over their respective areas.

1 I.

The generic fiberglass insulation debris is assumed to be a low density fiberglass with the same minimum tumbling velocity as NUKON'.

Computational Fluid Dvnamics (CFD) Software Several commercial software programs are used in performing the debris transport calculations.

Those that support or perform the computational fluid dynamics are:

I. GAMBIT Version 2.1.6 This program was used to generate three-dimensional solid models of the containment building from the floor elevation to the selected water surface elevation. GAMBIT was also used to 1 generate the computational mesh and to define boundary surfaces required to perform the CFD analysis.

1 II. FLUENT@

version 6.1.22 FLUENT@

Version 6.1.22 was used to perform the CFD simulations. FLUENT@

is a state-of-the-art general purpose commercial CFD software package for modeling problems involving fluid flow and heat transfer. It has been used to model flow processes for both government and industry and is one of the CFD software programs used by the NRC.

CFD Model and Boundarv Conditions The CFD model of the flooded portion of the containment was developed using GAMBIT. The model included the SFS strainer module modification including two (2) module trains. The free surface water elevation at the start of recirculation was 3' 2" above the basement floor. The numeric model did not include relatively small objects, such as support columns, pipes, pipe supports, equipment, instrument panels, etc., that are 6" along their longest dimension. Groups of objects with projected dimensions greater than 6 are generally included. In critical areas such as containment sumps and constricted flow paths, objects less than 6" are included.

Page 10 of 40

This meshed model was imported into the FLUENT@

CFD software program. The values for each boundary condition and the properties of the working fluid (water) were set in FLUENT@.

The two-equation realizable k-model was used to simulate the effects of turbulence on the flow field.

The results of the steady-state, isothermal flow simulations included component velocities (x, y and z directions), turbulent kinetic energy and the dissipation rate of turbulent kinetic energy for each cell in the computational mesh.

The following is a description of the boundary conditions used in modeling the PBNP Unit 2 containment sump flow patterns and velocity distributions. Each relevant physical boundary is listed followed by a discussion of the boundary condition applied at that surface.

I Solid Surfaces All of the solid surfaces in the containment building below the modeled water surface, including the walls, floors and structural supports, were treated as non-slip wall boundaries. At these surfaces the normal and tangential velocity components were set to zero.

Water Surfaces The upper boundary of the CFD model representing the water free surface was set at a water depth of 3' 2" above the floor and maintained constant throughout the CFD simulations. This water surface elevation corresponds to the minimum water level at the start of recirculation and is conservative since actual transport-flows slow as the sump level rises.

It has been postulated that as the water level rises during an actual event, the increased turbulence andlor the vertical velocity vector of the rising surface could cause a non-conservative result. The following discussion illustrates the reason that this does not occur and how the issue is addressed by the simulation.

During the first 30 minutes of recirculation, the inflow to the pool (break flow Qe plus containment spray flow Qs) is greater than the outflow through the containment sump, Qp. The excess of inflow versus outflow will cause the water surface to rise. The speed at which the water surface rises is calculated as:

V, = ' B '

' S - 'P where A is the exposed water surface area.

A In these simulations, the water surface rise velocity is very small compared to the expected pool velocities which would facilitate transport. Therefore, it is reasonable to assume that flow is steady and the water surface rise is treated as an outflow with a fixed vertical component of velocity where specified to satisfy continuity. This method allows for a quasi steady-state simulation of the flow patterns and velocity levels in the pool at a constant selected water depth.

I LOCA Break Flows Each break I strainer train combination was simulated. Future analyses may curtail the number of combinations if it is determined that one or a few breaks are dominant and limiting.

It was assumed that each break flow falls to the pool water surface without contacting any equipment or structures. The break flow jet accelerates under the influence of gravity as it falls towards the water surface. This is a conservative method to model the break flow as it produces the greatest lateral outflow velocities along the floor.

Page 1 1 of 40

The initial velocity V l of the water jet exiting the break is determined by:

where:

Qb = break flow (ft3/s)

D l = break inner pipe diameter (ft)

The velocity of the jet at the pool surface V2 is determined by:

where:

g = gravitational acceleration (ft/s2)

H = vertical difference between break location and water surface (ft)

The diameter of the jet D2 at the pool surface is determined by:

Each break was modeled by a circular velocity boundary surface on the top of the model under the given break location. These surfaces had a diameter D2 and a flow velocity V2 was applied normal to this surface. This method reproduces the correct flow and momentum of the jet without requiring the entire jet to be modeled from the origin of the break.

Spray Flows The flow from the spray header was introduced into the pool through a velocity inlet around the periphery of the containment, steam generator compartments and both of the open stairwells. The spray flow was distributed to each of these openings as appropriate.

The velocities of the sheeting flow across the refuel floor due to containment spray are calculated and any debris deposited is transported to the sump if the calculated velocity exceeds the incipient tumbling velocity.

Debris Size Classification The initial debris size distributions after pool fill, as provided by the debris generation / blowdown /

, pool fill analyses, were divided into "small" (dimensions less than or equal to 4") and "large" (dimensions greater than 4"). For the washdown and recirculation transport analyses, these size classes were further subdivided, based on guidance provided by the NEI 04-07, Volume II, and accounts for erosion effects.

A fraction of the large and small debris found in the basement (sump) debris zones is considered erodible into fines which remain suspended indefinitely. The remaining amount of debris may be susceptible to transport if the local flow velocity exceeds the incipient tumbling velocity of that debris type.

Page 12 of 40

Erosion of Debris A fraction of certain insulation types were assumed to erode into fines that are sufficiently small that the individual fibers or particles stay suspended in the water indefinitely. These suspended fines were assumed to move to the screens at any flow velocity and were therefore, assumed to be on the sump screen for determination of head loss. The remaining fraction of the insulation forms discreet particles which sink to the bottom of the pool and may be transported by the flow if the velocities equal or exceed the threshold velocity for incipient tumbling of that material. Erosion factors were obtained from the available test data found in available literature and used to quantify the amount of fines generated from the LOCA blast and later erosion that would arrive at the screens.

Data from NUREGICR-6808, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, for air jet testing of low density fiberglass (LDFG) at Colorado Engineering Experiment Station Inc. (CEESI), indicated an average of 20% of the insulation was classified as "non-recoverable" (i.e., fines). The same document summarized a single test on LDFG at the Ontario Power Generating (OPG) testing facility using heated, pressurized water. The quantity of fines was measured as 47%.

Due to the more numerous tests at the CEESI facility (and consistent with the preference toward air jet testing over water jet testing for establishing the destruction ZOI), more weight was given to the CEESI test results and an average of 30% of low density fiberglass insulation is assumed to be disintegrated into fines. These fines would remain in suspension and be filtered out by the sump screen. This fraction was applied to both the NUKON@ and fiberglass debris types.

CalSil was also tested at the OPG facility. The quantity of debris too small to be collected was termed "dust" and its mass was calculated by subtracting the collected mass from that of the initial target insulation. NUREGJCR-6808 indicates that the maximum mass of dust from seven tests was approximately 28% of the initial mass. Since some of the smaller, discrete fragments would further dissolve, it was assumed that a total of 35% of the initial amount of CalSil disintegrates into fines, remains in suspension and is filtered out by the sump screen.

A fraction of the smaller CalSil fragments, such as those blown into the containment by the initial break energy, would dissolve in the heated water of the pool. NUREG-6772, Separate-Effects Characterization of Debris Transport in Water, Section 3.3.1, summarizes tests on 10 gm (0.35 ounce) samples of CalSil placed in heated water with and without stirring. From 46% to 76% of these smaller fragments disintegrated as a suspension in the water.

It was assumed that only a small percentage of large chunks of CalSil would disintegrate, so the total fraction of the initial amount of CalSil converted into a suspension would be 35%.

NextEra also accounted for mineral wool, lnsulbatte (~emp-at@) and asbestos, and these insulation types. These may also be subject to disintegration into fines which stay suspended in the water.

NUREGJCR-6772, Table I.I, indicates that some types of mineral wool are similar to KaowoolTM and that KaowoolTM is a low density fiberglass. Therefore, it was assumed that the percentage of mineral wool fragmented into fines, suspended in the water pool and filtered out by the sump screens is the same as used for fiberglass, (30%).

lnsulbatte (~emp-at@) was tested in the CEESI facility and NUREGJCR-6808, Section 3.2.1.2, indicates a damage pressure of 17 psig was recommended by the Boiling Water Reactor Owners Group (BWROG) for unjacketed insulation. Similar testing for Knauf fiberglass and NUKON@

fiberglass indicated the recommended damage pressures were 10 psig for both insulation types.

Page 13 of 40

This indicates that the fiberglass insulation was easier to damage. To be conservative, the extent to which lnsulbatte disintegrates into fines was assumed to be the same as for fiberglass (30%).

There has been no testing with asbestos insulation, likely due to the special handling requirements associated with the hazardous material. Without explicit knowledge of material properties compared to other insulation types, it was conservatively assumed that 50% of asbestos insulation is fragmented into fines, suspended in the water pool and filtered out by the sump screens. This assumed percentage is higher than used for any other type of insulation due to the lack of test data or related information.

The flume testing described in the response to Question 4 introduced a scaled quantity of fines based on a percentage of all of the fibrous debris generated in containment, including debris calculated to be retained in other compartments and debris expected to be transported to the sump. There was no reduction in the fines inventory used in the testing due to a calculated transport fraction.

Debris Transport Characteristics Settling velocities and incipient tumbling velocities for the debris insulation types were obtained from NUREGICR-6772 and as summarized in NEI 04-07, Table 4-2. These velocities were applied to the fractions of debris insulation types that remain after erosion during the blowdown and washdown phases.

Calculation of Debris Transport Fraction Using the results of the CFD simulations, velocity isosurfaces and streamline plots were generated for use in predicting debris transport. Plots were generated corresponding to areas where velocities are equal to or greater than the velocities associated with incipient tumbling of the debris found in each zone. The velocity plots were obtained by projecting down to the containment sump floor the maximum lateral extent of a three-dimensional volume in which the velocities were equal to or greater than the selected incipient tumbling velocity. This method accounts for and bounds velocities at all elevations in the pool.

To determine the transport fraction of debris, the velocity contours were examined for isolated regions that were not contiguous with the strainer modules. Streamline and vector plots were used to identify isolated eddies that had velocities higher than the incipient tumbling velocity, but did not contribute to debris transport from the zone to the strainers. These vectors were also used to identify regions of the velocity contours that, while they may have been contiguous with the strainer, the flow was directed away from the strainers. These areas were subtracted and did not contribute to the recirculation transport fraction.

Overlays of the remaining velocity contours with the zone definition plots were used to determine the floor area which would be susceptible to transport for each break location. The fraction of the zone floor area that is susceptible to transport constitutes the recirculation transport fraction for each debris type. The total fraction of debris transported to the strainer from each zone is determined by the following equation:

Fraction of Debris Transported to Strainer per Zone

= Erodible Fraction + (I-Erodible Fraction) x Transport Fraction This process is applied for each debris type, in each zone and for each break analyzed.

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Results (Includinq Transport Fractions and Total Quantities of Debris)

Due to the pending large scale removal of fibrous debris sources from containment, the debris generation and transport analyses are being re-performed using the methodologies, assumptions and modeling described above. The results of this effort will be completed by December 18, 2009, consistent with the milestone provided in the June 12, 2009, letter from NextEra to the NRC (ML091660326).

Debris Interceptors Although previously installed in Unit 1, debris interceptors (Dl) are no longer being pursued as a credited solution for reducing debris reaching the sump strainers.

The debris interceptors installed in Unit I have three significant components:

o The Dls are vertical panels of bar grating that are covered almost entirely by l/d' perforated plate from the floor to a level above the maximum flood level of containment. These panels completely surround the strainers, separating them from the RCS compartments. These Dls feature a 4 high full width gap without perforated plate and with minimal obstructions that is located below the minimum flood level of containment. This gap forms a submerged weir and is sized to ensure that the sump screens cannot be starved of flow, even if all of the perforated plate would be completely blocked by debris.

o A pipe extension that diverts water draining from the refueling cavity (which may contain some small suspended debris) away from the vicinity of the strainers to a location upstream of the main debris interceptors.

o Metal curbing around part of the perimeter of the refueling floor (approximately 40% of the perimeter). This curbing prevents washdown water that may contain entrained fines from falling downstream of the main Dls. The water is diverted instead across the refueling floor to un-curbed locations (such as the refueling cavity, or the portion of the refueling floor perimeter that is not curbed.

NextEra is not crediting the Dls and some or all of them may be removed at a future date. In the interim, their presence is not being modeled in the various transport analyses. If it is postulated that the Dls do not retain debris and are completely ineffective, then they would also have no effect on the flow distributions through the containment sump pool. Conversely, if it is postulated that they retain debris to the point that they alter sump flow patterns, then their net effect would be beneficial in reducing the quantity of debris delivered to the strainers. Therefore, it was determined that neglecting them in the transport modeling is conservative and acceptable.

Debris Settlinq As described above, debris characterized as "fines" are assumed to remain in suspension indefinitely. "Small" and "large" debris sizes were modeled and transported if the calculated flow velocities exceeded the incipient tumbling velocities. Settling, though known to occur, was not explicitly modeled. Settling phenomena were accounted for in the design and conduct of the strainer qualification flume testing.

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Question 4 Please provide the following head loss and vortexing testing-related information.

a) Information requested by the NRC staff's revised content guide that was not previously submitted due to the testing being incomplete, or that changed during subsequent testing.

6) Flow rates in the flume c)

Scaling factors d) Debris amounts added to the testing apparatus, and debris size distributions for added fibrous debris e) Debris preparation and introduction methods which ensure prototypical debris transpott and bed formation NextEra Response a) A review of both the Staff's guidance and Reference (I) found one item not previously provided:

"3.f.4.

Provide a summary of the methodology, assumptions, and results of prototypical head loss testing for the strainer, including chemical effectsJ' Three tests were performed:

1. Differential pressure testing of the clean proto-type strainer

2. Transport testing of miscellaneous debris types

3. Design basis debris load testing The first test established a baseline differential pressure for the test strainer and associated piping and connections. The second test demonstrated the latent debris types (e.g., tie wraps, tape, labels, foreign material exclusion plugs) that were subject to transport in the sump pool flow streams. The third test was designed to establish a maximum differential pressure for a hypothetical worst case debris loading. The results of the test established a maximum upper bound for debris quantity.

By first adding the lightest, most transportable debris and subsequently introducing progressively heavier, less transportable debris, the test was designed to ensure the establishment of a "worst case" thin bed (see size distribution and sequence response to Question 4(d), Page 20. The more transportable debris was permitted to progress to the screen without the presence of heavier debris that may filter it out of suspension. The addition of the heavier debris was observed to stir up deposits of the lighter debris types that had previously settled to the floor of the test flume.

The test was also designed to demonstrate the effects of a circumscribed bed, if it was possible to form such a debris bed.

The general conduct of the test involved recirculating water through the test strainer while adding debris to the test flume and permitting the debris to transport to the strainer in a prototypical manner under the influence of the flow stream. The differential pressure across the strainer was continuously measured and the flume was permitted to recirculate to ensure quasi-equilibrium was I reached between debris additions.

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After all debris had been introduced and permitted to recirculate overnight to reach an equilibrium differential pressure, chemical surrogates were added to the flume to simulate the postulated formation of insoluble precipitants while the differential pressure continued to be monitored for trends.

I Design of the Test Flume The flume was designed to reflect a prototypical flow stream velocity. The following describes the analytical steps used to define the dimensions of the test flume:

I.

Use the CFD post-processing software to numerically seed each active module train face with mass-less tracer particles (mass-less tracer particles show the direction of the flow at every point along their path).

2.

Back-calculate the trajectory of the particles to define streamline traces to each module.

(This identifies the path the water follows to each strainer module face.)

3.

With the water path to each module identified, use the CFD post-processing software to define vertical planes at I

' increments from the module train, along the paths defined in Step (2).

4.

Trim each plane such that the velocities within that plane are those which convey water to the module.

5.

At each 1' increment from the module train, record the cross section average of the velocity magnitude across the plane. If the paths diverge around objects in the flow, follow each bifurcated path individually. Record these averages over a total of 20' from the module train.

6.

Conduct Steps (I) through (5) for each of the four trains in the array.

7.

Calculate the weighted average of the four flow streams at each 1' increment. The average at each increment is weighted by twice the fastest velocity at the increment under consideration in order to incorporate conservatism into the calculation.

8.

Create a plot of the calculated weighted average velocity defined in Step (7) vs. incremental distance from the module train.

9.

Using engineering judgment, create up to ten linear line segments which conservatively represent the velocity trends over the 20' distance.

10.

Calculate the width of the test flume at each line segment break using the following expression:

I

Where, I

Q = Total flow to test module (ft3/s)

I A = Flume cross sectional area (ft2)

I V = Weighted cross sectional average velocity (ftls)

I

and, I

A = WH Where: W = Flume width (ft)

I H = Water surface height in the test flume (ft)

11.

Create a table of flume width vs. line segment length to be used in defining the shape of the flume.

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The transition of the flume near the test strainer module is defined by the trajectory of the water as it approaches the modules in the prototype installation. These flow patterns are calculated in the CFD debris transport analysis.

The approach described above results in a test flume profile that replicates the most limiting (i.e.

most turbulent) flow configuration expected in the actual plant. Weighting by twice the highest velocity within a flow stream and constraining the performance of the test to a constant low level (actual plant sump level would continue to rise) ensures that the velocities obtained in the test flume are conservatively high, and that the turbulence induced is greater than that expected under actual plant conditions.

The resulting flume dimensions used for the test are depicted below:

Debris Introduction Zone Test Flume Dimensions The sharp turn at the end of the flume adjacent to the strainer is a result of the test facility configuration. However, it approximates the direction of debris entrained flow toward an edge of the strainer, similar to what is expected in the installed configuration. In the installed configuration, several modules are linked together such that the approaching flow progresses primarily toward the exposed sides of the modules and not from the ends.

Additionally, as evidenced by photographs of the test strainer during and after drain-down of the test flume, the debris cake was evenly formed over the entire surface of the strainer. There was no apparent disproportionate distribution due to the sharp turn in the flume.

Water Used in Flume The head loss tests were conducted with city domestic (tap) water at a temperature of approximately 1 OO°F to 1 20°F.

The pH of the water in the test flume was not intentionally controlled. The pH measured during tests prior to the addition of chemical effects surrogates ranged from 6.44 to 6.53. After the addition of the basic chemical effects surrogates, the pH increased to 8.82.

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1 Prototype Test Strainer The full strainer arrays consist of 14 modules each. The prototype test strainer was a single module that had originally been procured as a spare replacement module. As such, it was dimensionally identical to any module installed in the plant.

/ The surface area of the test module was 135.9 ft2.

I Turnover times, stabilization time The total volume of the flume and connecting piping was 2,460 gallons. At the targeted flow rate of 170.7 gpm, the turnover time was 14.4 minutes.

A minimum of five (5) flume turnovers elapsed after introduction of each of the fine, small and large fibrous debris inventories.

A minimum of two (2) flume turnovers elapsed following each batch addition of chemical surrogate.

Recirculation continued and the differential pressure (head loss) was monitored for a minimum of fifteen (1 5) flume turnovers following addition of all of the chemical surrogates.

I Chemical surroqates AluminumOxyHydroxide (AIOOH) was used as a chemical surrogate for the expected Sodium Aluminum Silicate (NaAISi3O8). The surrogates were prepared in accordance with the approved guidance of WCAP-I 6530-NP.

The maximum quantity of NaAlSiB08 that may be generated in the PBNP sump was determined to be 194.1 Kg. This was increased to 197 Kg (434.3 Ibs) as a contingency for possible future discoveries. The stoichiometric equivalent is 99.9 Ibs of AIOOH. This quantity was multiplied by the test scaling factor of 7.53% to obtain 7.52 Ibs, and an additional 1% was added to account for possible solubility effects. The target quantity of AIOOH was therefore 7.60 Ibs.

Flume testing of various postulated miscellaneous latent debris demonstrated that the debris would sink and was not transportable in the flow stream (labels, cable ties, sanding disks, paper, plastic pipe caps, gloves, etc). A small number of items were observed to float on the surface and transport (e.g. plastic FME plugs, duct tape, masking tape, nylon rope, danger tape, a polyethylene bag, a "hot spot" tag, and various pens). To conservatively bound the effects that such latent debris may have on a strainer array, it was assumed up to I00 ft2 of active screen surface would be blanketed by such debris. To account for this effect, the flume flow rates were increased accordingly. Refer to the Results section below for additional detail on scaling factors.

Results At a flow rate of 150 gpm and I 16OF, the clean strainer head loss directly observed during the test was 0.066'. When the flow was increased to 175 gpm at 1 18OF, the clean strainer head loss was 0.090'. This head loss is subtracted from the debris loaded test result to determine the head losses associated with the debris bed during debris loaded testing. Refer to Question 8 for the clean strainer head losses of the installed strainer array.

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The temperature and flow rate corrected debris loaded strainer head loss for the design basis debris load test was 6.45' at a reference temperature of 11 3.6OF at a flow rate of 170.7 gpm. The peak loss occurred after the design basis fibrous and particulate debris was placed in the test flume and prior to any chemical introduction.

The temperature and flow rate corrected debris loaded strainer head loss at test termination (1 00% of chemicals introduced and 15 flume turnovers) was 3.64' of water at 175 gprn at a temperature of 1 13.6'F.

The suspended debris in the test flume prevented imaging of the test strainer during the conduct of the test. However, photographs of the test strainer module after the recirculation pumps were secured and after drain-down of the test flume had started indicated that a relatively uniform "thin bed" of debris was formed. No significant bridging of the strainer disks (indicating a "circumscribed bed") was evident.

b) Flow rates in the flume The flow rates for the clean strainer test ranged from 125 gpm to 225 gpm to generate a head loss vs. flow curve.

The flow rates for the debris loaded test varied slightly through the course of the test as the flow was adjusted to remain constant as the debris bed developed and aged. At the highest recorded head loss, the flow was 170.4 gpm. At test termination, the flow rate was 175 gprn.

c) Scaling factors The full scale arrays have a total active strainer surface of 1,904.6 ft2. To account for potential debris blanketing by sheet-type latent debris, 100 ft2 was deducted from this active surface area for a net area of 1,804.6 ft2.

The test strainer had an area of 135.9 ft2. This resulted in a test scaling factor for debris and flows of:

The design basis flow rate for the strainer arrays is 2,200 gpm. Therefore, the minimum test flume flow rate was 2200 gprn x 0.0753 = 166 gprn.

The flume water depth was set at the minimum design submergence for the screens and not varied during the additions of particulate and fibrous debris. In order to prevent loss of chemical surrogates from the flume, it was necessary to allow the level to rise minimally (-1.4) with the addition of chemical surrogates. Actual containment sump levels would continue to rise significantly during the first approximately 30 minutes of containment sump recirculation.

Therefore, the low water levels of the test were conservative in that they maximized flume velocity and minimized screen submergence.

d) Debris amounts added to the testing apparatus and debris size distributions for added fibrous debris The amount of each type of debris, as well as the sequence added, was as follows:

Batch 1: 25% of Latent Fibrous Debris ( N u ~ ~ N ~ f i n e fiber, 0.15 Ibm)

Batch 2: 100% of Cal Sil(47.1 Ibm)

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Batch 3: 100% of Latent Particulate, Dirt and Dust (9.7 Ibrn)

Batch 4: 100% of Aluminum, Alkyds, and Epoxy Coatings, Acrylic Powder (142.6 Ibrn)

Batch 5: 100% of Zinc Coatings, Tin (85.1 Ibm)

Batch 6: 100% of Fine NUKON@ Fibers (3.2 Ibm)

Batch 7: 100% of Fine Ceramic Fibers ( I 1.7 Ibrn)

Batch 8: 100% of Owens Corning Fine Fiberglass Fibers (3.6 Ibm)

Batch 9: 100% of ~ernp-at@ Fine Fibers (6.0 Ibm)

Batch 10: 100% of Mineral Wool Fine Fibers (1 0.1 Ibrn)

Batch 11: 100% of Degraded Epoxy Coatings, Acrylic Chips (93.2 Ibm)

Batch 12: 100% of Small NUKON@ Fibers (0.7 Ibm)

Batch 13: 100% of Small Owens Corning Fiberglass Fibers (2.9 Ibm)

Batch 14: 100% of Small Temp-at@ Fibers (1.3 Ibrn)

Batch 15: 100% of Small Mineral Wool Fibers (4.1 Ibrn)

Batch 16: 100% of Large NUKON@ Fibers ( I.3 Ibrn)

Batch 17; 100% of Large ~ernp-at@ Fibers (4.8 Ibm) e) Debris preparation and introduction methods which ensure prototypical debris transport and bed formation For details about the preparation of the surrogate materials, their size distributions (fines, smalls, larges and intact), and the technical basis for the use of the various surrogates, refer to Performance Contracting Inc., letter to NRC, dated March 25, 2009, PCI-6016-02.01,, Sure-Flow@ Suction Strainer -Testing Debris Preparation and Surrogates, SFSS-TD-2007, Revision 4 (Proprietary) (ML090900476).

For the design basis test, all batches, except for Batch 1, were introduced at the far end of the flume (the "drop zone"), upstream of the strainer module. Batch I was introduced along the length of the flume prior to the start of the recirculation pump.

It has been noted that Batch I, being introduced prior to the start of the recirculation pump, may have introduced non-conservatism into the test protocol. Introduction of a portion of the debris representing latent (pre-existing) fibers into the sump prior to starting the test recirculation pumps was intended to more accurately represent the expected behavior of debris resident in the containment during the sump pool fill phase, Introducing this portion over the length of the flume, rather than at the "drop zone" was intended to more accurately simulate the expected distribution over containment.

The debris introduced prior to starting the recirculation pumps was limited to only fine fibers of NUKON@, which remain suspended "indefinitely", and which would be subject to transport in any flow stream regardless of the velocity (i.e. no threshold for incipient tumbling velocity). Therefore, introduction prior to the start of the recirculation pumps should not be a concern. In any case, the portion so distributed represented such a minor fraction of the total fibrous debris introduced (0.1 5 Ibs of the total 3.45 Ibs of fine NUKON@ fiber and an even smaller fraction of the total fines and total fiber in the test) into the flume that this deviation in the otherwise consistent introduction protocol had a negligible effect on the results.

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To minimize non-prototypical turbulence caused by the introduction of large quantities of debris into the flume, an inclined ramp was built that permitted pouring out the pre-wetted debris slurry onto the ramp, and allowing it to washdown the ramp into the flume. While this reduced the turbulence caused by pouring the debris in a free falling column of water, some disturbance of the settled debris was still observed with each introduction.

Question 5 At the beginning of recirculation for a small-break loss-of-coolant accident (SBLOCA), the strainer stacks are submerged by about two inches. The supplemental response stated that buoyant debris would not be present following a loss-of-coolant accident (LOCA) (based on the first tests performed at Alden Research Laboratory (ARL) and that, therefore, air ingestion through the debris on the strainer screens would not occur. However, NRC staff present at the ARL testing noted that the debris was added after being mixed together and then mixed with water. This test may have not been a prototypical test to determine whether buoyant debris can occur. The phenomenon of buoyant debris should be addressed.

NextEra Res~onse The design basis submergence for the PBNP strainers is independent of the size of a LOCA. The 2 minimum submergence cited is therefore applicable to the full range of postulated LOCA events.

The testing witnessed by NRC staff present at Alden Research Laboratory (ARL) was conducted in 2005 and 2006. Since that time, considerable industry and NRC efforts have been invested in developing a test protocol that is considered more conservative and appropriate. The tests being discussed in this response were performed in July 2008 and were conducted in accordance with the later protocol. This later protocol also pre-mixes the various debris types in water prior to introducing the debris to the test flume.

The practice of pre-mixing debris is considered conservative and appropriate because it enhances the distribution of the debris in the water and tends to free individual fibers that may otherwise "clump" together and provide less conservative results than individual fibers and small clusters of fiber. This increases the likelihood of forming a limiting thin bed with a high head loss during the test.

While the debris of each size and material type were pre-mixed with water, different debris material types and different debris sizes were pre-mixed and introduced into the flume separately.

The potential for floating debris causing a blanketing effect and leading to air ingestion at the top of the screen is considered very unlikely for several reasons:

While the level of the containment sump is at a minimum 2" at the start of sump recirculation, it continues to increase over the course of approximately 30 to 40 minutes as additional RWST volume is transferred to the containment sump by containment sprays. The final submergence level would be a minimum of 14.6" deeper when the RWST is depleted to 12% level, and additional submergence can be expected as water held up as spray droplets, steam and sheeting water drain to the sump when containment spray is terminated. By comparison, the sump turnover rate, based on the volume at the minimum recirculation level of 38, or approximately 154,000 gallons, and the maximum design sump outflow rate of 2,200 gpm, is 70 minutes. Therefore, the screens have substantially greater submergence well before a single turnover of the sump (i.e. transport of debris toward the screens) has occurred.

2. Insulation dislodged by a postulated energetic two phase jet would tend to be wetted by the 1

same jet.

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3. The debris would have a considerable "soak time" in the stagnant hot sump water prior to initiation of sump recirculation. The low viscosity and low surface tension of the hot water enhances rapid wet-out of the debris and ensures it occurs prior to the start of sump recirculation.

4. PBNP does not have closed-cell insulation types (e.g. microtherm, Min-K, or anti-sweat foam) in the LOCA ZOls. Miscellaneous debris types (e.g. electrical tape, tie-wraps, labels, etc.) were also tested for transport characteristics during the flume testing and found not to float.

Question 6 The supplemental response did not consider the potential effects of water from the break or from spray drainage falling near the strainer. Especially during the period of relatively small submergence, and possibly at times for which there are other sump pool levels, the falling water could entrain air near the strainer resulting in the air being drawn through the strainer and into the emergency core cooling system pump suction header. This potential post-LOCA phenomenon should be considered and addressed.

NextEra Response At PBNP, there is no potential for water cascading from upper levels to fall directly on a strainer assembly. An analysis comparing the rise velocity of an air bubble originating at the floor of containment to a height above the strainer with the horizontal velocity of water moving toward the strainers found that bubbles originating 2 or more from a strainer cannot be ingested by the strainer.

It has been verified that areas where water may cascade into the sump pool are located significantly greater than 2" from the strainers. In one case, the planned extension of the strainer array by an additional three modules could result in the array being below the reactor cavity drain located above. Modifications will include extending this specific drain away from the strainers, or installation of an impingement device between the strainers and the drain to prevent air ingestion.

In one other location, there is the potential for distributed droplets from containment spray to pass down several flights of an open stairwell and impinge on the pool surface immediately above a strainer. This distributed rain-like flow is judged to not be a source of entrained air bubbles.

Question 7 The supplemental response stated in one place that observations for vortexing will be accomplished during the head loss testing for the future. In another area, the supplemental response stated that the assessment of vortexing was based on empirical observations rather than a calculation (presumably during testing which had already been conducted). These two statements appear to be contradictory. The final vortexing assessment should provide the test conditions under which the observations occurred and discuss how these conditions are either prototypical or conservative with respect to expected plant conditions.

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NextEra Response The observations for vortexing were performed during the testing described in the responses to Questions 3 and 4.

The installed strainers are PC1 sure-low@ strainers which incorporate a flow control device that ensures even flow distribution among all of the strainers in the array. The test strainer module was dimensionally identical to one of the 14 strainer modules of the complete strainer array and the flow was conservatively higher than 1/14 of the design flow for the strainer array. Therefore, the test strainer was prototypical while the flow rate was conservative.

The submergence level of the test strainer was controlled to remain constant at 2" throughout the addition of fibrous and particulate debris and permitted to rise minimally (-1.4) with the addition of the chemical surrogates. In contrast, the actual post-accident sump levels would rise over the first approximately 30 minutes of sump recirculation and provide additional margin against vortex formation. Therefore, the submergence level of the test strainer was conservative.

Upon completion of fibrous insulation abatement, the potential debris loading in the containment will be bounded by the testing that was completed. Therefore, the debris loading of the test was conservative.

At no time during clean strainer head loss or debris loaded head loss testing was vortex formation observed.

Question 8 The clean strainer head loss (CSHL) value provided in the submittal was stated to be for hot sump conditions. A value for CSHL for the postulated minimum sump temperature should be provided.

NextEra Response The expansion of the strainer arrays to include three additional modules (for a total of 14 modules) required re-evaluation of the clean strainer head losses. Therefore, the following information supersedes the previous response in Reference 1, and was obtained from Table A-I in of this submittal.

The calculated clean strainer head loss (including losses from associated piping and fittings up to the containment outlet) at 212OF and the design flow rate is 0.41'.

The corresponding calculated head loss with 72OF sump water is marginally higher at 0.59'.

Question 9 The licensee assumed that all debris generated by a LOCA transports to the sump. However, no size distributions for the various debris types expected to arrive at the strainer was provided, Size distribution is an impottant factor in debris bed formation and is therefore required to perform and document a valid head loss test. Size distributions for debris expected to arrive at the strainer should be provided.

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NextEra Response The requested information will be evident as the end result of the final debris generation and transport analyses. While these analyses have been performed for the existing insulation configuration, they have not been completed with regard to the fibrous insulation replacement plan approved by the NRC, June 30,2009 (ML091800430), for PBNP Units I and 2. The following table, Unit 2 Steam Generator B Crossover Leg Nozzle with B Strainer Train Operation, was developed using the existing analyses and deducting fibrous insulation to be replaced with RMI.

The table reflects the single most limiting break that was identified for the existing insulation configuration.

The debris quantities projected for the final configuration are based on the estimated results of the existing analyses after they are revised to incorporate insulation replacements and the addition of three strainer modules per train. The quantities of coatings have been reduced as discussed in the response to Questions 15 and 16. The potential presence of miscellaneous debris was accounted for by the use of an assumed "sacrificial area" of I00 ft2 when scaling the flow rate of the flume for the test strainer (see the response to Question 4 for the discussion of scaling factors). The assumed latent debris total, 150 Ibs, has been conservatively maintained despite sampling data which indicates actual values are much lower.

The table below also contains the quantities of debris used in the July 2008 screen qualification flume test. For ease of comparison, all test quantities are presented as in full scale equivalents.

From this comparison, the planned insulation reductions will result in a total fibrous debris inventory that is substantially less than that which passed in the successful screen test.

Although not immediately apparent, the assumed quantity of latent debris fiber is bounded by the test. It was assumed that 22.50 Ibs of latent fibers are present in containment. However, the test plan only introduced a scaled equivalent of 6.64 Ibs of NUKON@ fines to the flume to account for the latent fibers. This reduction between the quantities assumed to exist in the containment and that which was introduced was intended to account for expected debris interceptor performance.

Since completion of the strainer testing, NextEra has elected to forego crediting debris interceptors. The difference between these quantities must now be reconciled through the removal of fibrous insulation.

Following the initiation of flume flow, 17.71 ft3 of NUKON@ fines were introduced in-stream to account for NUKON@ insulation. The quantity of generated NUKON@ insulation is planned to be significantly reduced or eliminated entirely. Considering the as-fabricated NUKON@ density, 2.4 lb/ft3, the in-stream NUKON@ introduction represents 42.5 Ibs of NUKON@ fines which offsets the difference between the assumed latent fiber quantity and the specifically tested quantity of latent fiber.

1 NextEra plans to complete the analyses reflecting the quantities of debris expected to arrive at the strainers by December 18, 2009, as previously stated in the June 12, 2009, letter from NextEra to the NRC (ML091660326). Although it is possible that the bounding break location may change, the quantities transported to the strainers are expected to remain bounded by the test results.

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I Unit 2 Steam Generator B Crossover Leg Nozzle with B Strainer Train Operation From the above table, the total the volume of fibrous insulation debris transported to the strainers in the current configuration is 391 ft3. After the projected reductions have been completed, it is estimated that this total will be 23.2 ft3. The scaled total quantity of fibrous insulation debris used in the screen qualification test was 97.2 ft3.

Page 26 of 40 Tested Debris Quantities ft3 (scaled up) nla nla 15.54 (fiber) 6.77 (particulate) 7.19 nla 3.87 17.71 5.40 nla I

.46 6.75 nla 7.00 8.69 nla 6.81 16.77 6.64 Ibs 38.47 127.5 Ibs 2.47 0.14 6.86 5.05 8.37 13.16 I00 f f Sacrificial Area Debris Types Asbestos (Ceramic Fiber surrogate)

NUKON@

Temp M ~ T @

Fiberglass Mineral Wool Latent Fibers Cal-Sil Latent Particulate Zinc Coating Aluminum Coating Alkyd Coating Unqualified Epoxy Coating Coating Degraded EPOXY Coating Misc. Debris Projected Generated Debris (estimated ft3) 0 16.5 nla nla 0

0 nla nla 0

0 nla nla 24.5 nla nla 0

nla nla 22.50 Ibs 7.2 127.5 Ibs 2.25 0

6.86 5.05 3.58 13.16 152 ft?

Size Distributions Large Small Fines Small Fine Large Small Fines Small Fine Large Small Fines small Fine Small Fines Small Fine Small Fines Small Fine Fine Small Fines Small Fine Particulate Particulate Particulate Particulate Particulate Particulate Chips Film Configuration Debris Transported to Strainer (estimated f. )

nla I

.50 8.52 0

nla 0

0 0

nla 0

0 nla 5.84 7.34 nla 0

0 22.50 Ibs nla 0.78 2.52 127.5 Ibs 2.25 0

6.86 5.05 3.58 13.16 59 ft?

Analytical Current Generated Debris ( ft3) 0 1 16.07 nla nla 93.97 140.95 nla nla 35.76 53.66 nla nla 114.70 nla nla 221.96 nla nla 22.50 Ibs 83.87 127.5 Ibs 2.47 0.14 6.86 5.05 8.37 13.16 189 ft?

Results for Configurations Debris Transported to Strainer (ft3) nla 10.53 58.04 28.19 nla 14.92 70.48 21.36 nla 5.68 26.83 nla 27.32 34.41 nla 26.80 66.59 22.50 Ibs nla 9.08 29.36 127.5 Ibs 2.47 0.14 6.86 5.05 8.37 13.16 nla

Question 10 Based on recent testing, it was reported that a debris interceptor would be installed that will prevent 75 percent of the debris from reaching the strainer. The amount of debris passing the interceptor should have been, or should be, evaluated considering the potential water levels above the interceptor, debris sizes, debris types, etc. The debris used in testing should match the characteristics of the debris that is expected to pass the interceptors. Therefore, the validity of the 75 percent efficiency value for the debris interceptor should be addressed and also stated to be reflected in debris quantities used in strainer testing, if applicable.

NextEra Response Upon further evaluation of the Dl qualification test results, and in consideration of additional development that may place a higher performance requirement on the Dls than they are capable of supporting, NextEra has elected to forego crediting Dls in the final resolution of these issues. The Dls previously installed in Unit I may eventually be removed, particularly since their presence creates additional complexity and postulated flow conditions in the transport analyses.

Question 11 The supplemental response did not provide an adequate response to the revised content guide question on the ability of the strainer to accommodate the maximum debris load. The supplemental response stated that debris beyond that collecting on the strainer would collect in the free volume in the lower level of the containment. The intent of the question is to ensure that the strainer either has a large enough area to prevent circumscribed bed formation, or that the formation of a circumscribed bed will not result in excessive head loss. Alternatively, a properly conducted test could show that a circumscribed bed will not result even from the maximum potential debris load. Please re-address this content guide question, given the above guidance.

NextEra Response The test performed and described in the responses to Questions 3 and 4 was designed to favor the formation of a circumscribed bed if one could be formed. By structuring the debris additions to progress from the smallest debris to the largest debris, using prototypical or bounding high transport flow velocities, all debris that might be transported to the strainer were transported.

Larger debris did not impede the transport of smaller debris. Conversely, the later addition of larger debris could (and based on the recorded data apparently did) disturb and re-suspend previously settled fine debris. As a result of the test protocol, the measured head loss results reflect the formation of a circumscribed bed if one could be formed.

As noted in the response to Question 4, post-test photographs of the test strainer found no indication that a circumscribed bed had formed.

Question 12 The supplemental response indicated in several places that a thin bed would not likely form on the complex Performance Contracting Incorporated (PCI) strainer. Based on several tests of PC1 strainers that have resulted in a relatively thin filtering bed, and the licensee's potentially challenging debris loads in terms of thin bed formation, the staff believes that the thin bed should be evaluated for the new Point Beach strainer configuration. Please justify the conclusion that such a bed would not form.

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NextEra Response The test performed and described in the responses to Questions 3 and 4 was designed to favor the formation of a thin bed, if one could be formed. By structuring the debris additions to progress from the smallest debris to the largest debris, using prototypical or bounding high transport flow velocities, fine fibrous debris that could be transported to the strainer were transported. Larger debris did not impede the transport of smaller debris. Conversely, the later addition of larger debris appears to have disturbed and re-suspended previously settled fine debris, based upon recorded data. As a result of the test protocol, the measured head loss results reflect the formation of a thin bed, if one can be formed.

As noted in the response to Question 4, post-test photographs of the test strainer suggest that a thin bed formed during the test.

Question 13 The submittal references 38 inches as the maximum allowable head loss. Based on recent test results described to the NRC in a phone call with the Point Beach licensee, it appears that this value may be too low. Please state the final maximum allowable head loss and reflect this value in net positive suction head calculations and structural evaluations.

NextEra Response The replacement strainer assemblies were originally designed to operate with a debris loaded head loss of 3 8 or less. This was based on the available net positive suction head (NPSH) margin under hot sump conditions at the start of containment sump recirculation. Later developments led to an understanding that while the sump cooled and the available NPSH increased, the differential pressure (AP) across the screens could also increase significantly because of higher head losses of the more viscous water passing through the debris bed.

As a result, the design differential pressure of the strainers and related piping and supports has been increased to 10'. This is believed to be the maximum differential pressure justifiable without a complete redesign and replacement of the strainer modules. provides the calculation of total head loss through a debris loaded screen assembly at various temperatures. The calculation is based on the results of prototypical screen testing performed with a debris loading that is conservative for the final anticipated configuration of the PBNP containments.

The results of this calculation demonstrate that if the sump is permitted to cool excessively with the high design flow rate, the 10' differential pressure limit could be exceeded. Therefore, one or more of the following three measures will be implemented to ensure that this does not occur: Limiting long-term containment sump cool down; requiring long-term sump flows to be reduced prior to cooling down below the high flow/low temperature limit, or re-performing the screen qualification testing with a debris mix representative of that which could exist after completion of the planned insulation abatements.

The maximum allowable head loss is determined by the most limiting of three considerations:

structural capability, prevention of flashing, and maintaining sufficient NPSH for the residual heat removal (RHR) pumps. These considerations are discussed as follows.

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Structural Evaluation Structural modifications to reinforce the limiting components (stiffening the end module cap and the anchoring of the end modules to carry the end thrust loads) have been completed on the Unit 1 strainer assemblies to accommodate an operating differential pressure of 10'. A similar modification is planned for Unit 2. Enclosure 3 provides excerpts from the revised structural analysis for the strainer modules demonstrating the acceptability of operating the Unit 1 strainers with differential pressure as high as 10'. Similar modifications to reinforce the Unit 2 strainer assemblies for this higher differential pressure are scheduled for installation during the fall 2009 refueling outage. provides excerpts from the revised structural analysis of the connecting piping and supports for Unit 1 demonstrating acceptability of operation with a differential pressure as high as 10'. This revision was possible without modification of the installed piping and supports. A similar revision for the Unit 2 connecting piping and supports is in progress, Flashins Evaluation contains a calculation demonstrating that a debris loaded strainer assembly head loss of 10' does not cause flashing within the strainer assembly. The calculation considers flashing at the screens and at the strainer assembly outlet over a range of operating temperatures.

A unique containment isolation valve is located at the discharge of the sump strainer screen assembly. This valve presents a flow restriction that could also cause localized flashing if pressure at the screen outlet is too low. Therefore, Enclosure 5 also includes an evaluation to ensure that the pressure loss through the strainer assembly remains low enough to preclude flashing in these outlet valves. The calculation concludes that flashing will not occur anywhere in the strainer assemblies or in the sump outlet isolation valves with a total strainer head loss of ?Or.

In reaching this conclusion, the calculation credits the pressure present in containment due to the sum of the partial pressures of air and water vapor (steam). The methodology used is consistent with Nuclear Energy Institute, NEI-04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, May 28, 2004, Volume II Attachment V-I (ML041550279, ML041550332, ML041550359 and ML041550380). It does not rely on a transient analysis of the post-accident containment pressure and temperature. The text of the calculation contains the details of the derivation of the solution methodology.

NPSH Evaluation The ECCS NPSH analyzed suction flow path begins at the outlet of the strainer assembly and assumes that the total pressure available at this point is equal only to that of the water vapor pressure (i.e., no submergence). Since the flashing evaluation demonstrates that the total pressure available at the outlet of the strainer assembly does not fall below the vapor pressure of the water, the ECCS NPSH evaluations are not affected by the maximum allowable strainer assembly head loss of 10' that may occur as the sump cools down.

Page 29 of 40

Question 14 Please list the quantity and debris characteristics of the unqualified coating debris in containment.

NextEra Response The quantity of unqualified coatings actually resident in the containments (including a 15%

allowance applied to coatings outside of the zone of influence [ZOI] for future contingencies) has been calculated to be bounded as follows:

Coating Tvpe Zinc coatings Alkyd coatings Degraded Epoxy coatings outside of the ZOI Volume I

.95 ft3 6.83 ft3 Density 457 lb/ft3 90 ib/ft3 Unqualified Epoxy coatings 5.05 ft3 94 lb/ft3 Total Coatings Volume Outside ZOI:

26.99 ft3 Consistent with the approved guidance of NEl-04-07, it is assumed that all unqualified coatings fail to their constituent particle sizes as fine dust. Degraded epoxy coatings (abraded, delaminating, etc.) located outside of the zone of interest (ZOI) are assumed to fail as chips or flakes.

No limiting sources of aluminum coatings were identified. While there may be residual aluminum coating still present under the RMI on the reactor vessel, a break occurring adjacent to the reactor vessel would result in minimal fibrous debris. Therefore, the effects of chemical precipitants on screen performance due to a break in that location is bounded by the combined chemical and fiber effects of breaks occurring in the RCS loop compartments.

During a teleconference on June 22, 2009, the NRC Staff requested additional information which justifies why epoxy based coatings that were originally qualified, but have degraded and are outside the ZOI, are assumed to fail as chips or flakes.

Testing performed for Comanche Peak Steam Electric Station by Keeler & Long (ML070230390) has been reviewed and found to be applicable to the degraded DBA-qualified epoxy and inorganic zinc coatings applied at PBNP. In that test, epoxy topcoat / inorganic zinc primer coating system chips, taken from the Comanche Peak Unit I containment were subjected to DBA testing in accordance with ASTM D 391 1-03, Standard Test Method for Evaluating Coatings Used in Light-Water Nuclear Power Plants at Simulated Design Basis Accident (DBA) Conditions. In addition to the standard test protocol contained in ASTM-D 391 1-03, 10 pm filters were installed in the autoclave recirculation piping to capture small, transportable particulate coating debris generated during the test.

The test confirmed that while the inorganic zinc failed to powder form, the phenolic epoxy topcoat

, remained as relatively large (>I132 diameter) pieces that were not transportable.

Page 30 of 40

Question 15 The supplemental response indicated that the quantity of coatings debris from steel structures is represented by the surface area of a 10 diameter (D) "half sphere. " This approach is not consistent the NRC safety evaluation (SE) dated December 6, 2004, on Nuclear Energy Institute (NEI) 04-07 "Pressurize Wafer Reactor Sump Performance Methodology," which calls for all of the coatings within a IOD ZOI of a pipe break to fail. Please provide the surface area of the coated steel structures in the ?OD zone of influence (ZOI). Is this surface area bounded by the surface area of a ?OD half sphere?

NextEra Res~onse The approach previously described was overly conservative and was inconsistent with the approved guidance of NEI 04-07. The calculation has been revised to more closely follow the guidance of NEI 04-07 in the subject of qualified coatings within the ZOI.

The revised calculation recognizes that a I OD ZOI would envelope a substantial portion of a reactor coolant system (RCS) loop compartment. To simplify the calculation, 100% of qualified steel coatings within the compartment are now assumed to fail. This amounts to 2,390 ft2 of surface area, and contributes 2.28 ft3 of epoxy coatings debris and 0.3 ft3 of zinc coating debris.

This is a net reduction from the previously calculated volume of 4.53 ft3 using the surface area of a half-sphere. Therefore, the surface area of the coated steel structures in the IOD ZOI was bounded by the surface area of the surface area of the half-sphere previously described.

Question 16 The supplemental response indicated that the quantity of coatings debris from concrete structures is represented by the surface area of a 40 "sphere." This approach is not consistent the NRC SE on NEI 04-07, which calls for the surface area of all coated concrete surfaces within a representative ZOI. Please provide the surface area of the coated concrete surfaces in a 40 ZOI around the limiting pipe break. Is this surface area bounded by the surface area of a 4D sphere?

NextEra Response The previous approach, while conservative, was inconsistent with the approved guidance of NEI 04-07 and was overly conservative. The calculation has been revised to follow the guidance of NEI 04-07 in the subject of qualified coatings within the ZOI.

The revised calculation evaluated the maximum surface area of coated concrete surfaces within a 4D ZOI. The resulting area is 400 ft2, contributing I

.3 ft3 of epoxy coatings debris.

This is a net reduction from the previously calculated volume of 4.36 ft3 using the surface area of a 4D sphere. Therefore, the surface area of the coated concrete structures in the 4D ZOI was bounded by the surface area of the surface area of the sphere previously described.

During a teleconference on June 22, 2009, the NRC Staff requested additional information relating to the basis of the 40 ZOI that was used for qualified coatings on concrete substrates.

To substantiate the reduction of the ZOI for qualified coatings to 4D, NextEra established the Level 1 concrete coatings inside postulated ZOls, and procured reports of QA "JOGAR testing of these coating systems on a concrete substrate.

Page 31 of 40

The tests consisted of subjecting representative samples of the coatings to a freely expanding jet of water with stagnation conditions greater than or equal to 210 psig and 300°F. Utilizing the industry accepted high energy line break jet model set forth in ANSIIANS-58.2-1988, Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, it was determined that the piping length-to-diameter ratio (LID) value associated with the bounding RCS cold leg break is less than 4.0. These stagnation conditions correspond to a coating damage pressure of approximately 52 psig.

The tested coating systems representative of the Level 1 coatings on concrete in the PBNP containments passed the test with no detectable damage, indicating that they have an effective ZOI of 4.0 or less.

Question 17 Considering your responses to the foregoing two RAls, please provide the total quantities of qualified coatings in the respective ZOls for concrete and steel surfaces, as well as the total quantities of degraded qualified coatings and unqualified coatings in containment. Are the quantities from your initial GL 2004-02 response (ML052500302) still accurate?

NextEra Response The total quantities of the coatings were provided in the responses to Questions 14, 15 and I 6 above. The quantities in the initial GL 2004-02 response are no longer correct. The reduced quantities stated above are being used.

Question 18 Please describe the debris characteristics and transporf percentage (size, shape, density, and thickness) of the qualified, degraded qualified and unqualified coating debris.

MextEra Response As stated in the response to Question 14, unqualified coatings and coatings within the ZOI are assumed to fail as fine particulates. Epoxy based coatings that were originally qualified, but have degraded (e.g. abraded or delaminating) and are outside the ZOI are assumed to fail as chips or flakes.

No attempts have been made to date to model transport of coatings debris by analysis. It is not planned that analysis will be performed because screen qualification testing has been used. In the tests, the quantity of coatings debris introduced into the test flume was scaled to the test screen surface area, modeling 100% of the coatings debris calculated for containment. The details of the surrogates used for coatings debris are provided in Performance Contracting Inc. letter to NRC, dated March 25, 2009, PCI-6016-02.01, Attachment 3, Sure-Flow@ Suction Strainer - Testing Debris Preparation and Surrogates, SFSS-TD-2007, Revision 4 (Proprietary) (ADAMS Accession Number not available).

Page 32 of 40

Question 19 Please provide the information requested under item (m) in the Revised Content Guide for GL 2004-02 Supplemental Response dated November 2007.

NextEra Response ltem (m) of Revised Content Guide for Generic Letter 2004-02, Supplemental Responses November 2007, dated November I I, 2007, (ML073110278) requests licensees to:

"...Provide the information requested in GL 04-02 Requested lnformation ltem 2(d)(v) and 2(d)(vi) regarding blockage, plugging, and wear at restrictions and close tolerance locations in the ECCS and CSS downstream of the sump.

GL 2004-02 Requested lnformation ltem 2(d)(v)

The basis for concluding that inadequate core or containment cooling would not result due to debris blockage at flow restrictions in the ECCS and CSS flow paths downstream of the sump screen, (e.g., a HPSl throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screens mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen sun'ace.

GL 2004-02 Reauested lnformation ltem 2(d)(vi)

Verification that the close-tolerance subcomponents in pumps, valves and other ECCS and CSS components are not susceptible to plugging or excessive wear due to extended post-accident operation with debris-laden fluids.

If NRC approved methods were used (e.g., WCAP-16406-P with accompanying NRC SE),

briefly summarize the application of the methods. Indicate where the approved methods were not used or exceptions were taken, and summarize the evaluation of those areas.

Provide a summary and conclusions of downstream evaluations. Provide a summary of design or operational changes made as a result of downstream evaluations."

GL 2004-02 Requested lnformation ltem 2(d)(v)

Industry resolution and accepted test data for in-corelin-vessel effects are pending. As such, FPL Energy Point Beach is deferring a response to this aspect of ltem 2(d)(v) pending NRC acceptance of a resolution approach.

A review of ex-vessel downstream components for potential flow restriction blockage was completed consistent with the NRC approved guidance of WCAP-I 6406-P, Revision 1. No deviations or exceptions were taken.

The ECCS sump screen perforation size is 0.066" diameter.

The limiting passageways in the ECCS and containment spray system (CSS) were reviewed, and the most limiting passageway was found to be larger than the largest assumed debris diameter.

Therefore, blockage of the ECCS and CSS passageways due to debris laden fluid is not a concern.

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The following paragraphs are excerpts from the evaluation:

"...all piping diameters in the sump recirculation 1 injection flow paths are greater than 1.5 in. There are no globe valves in the ECCS and CSS lines. RHR heat exchanger outlet valves, 1&2-RH-624 & 625 are throttled to prevent RHR pump run out at certain conditions.

These valves are 8 in butterfly valves..."

"...the RHR heat exchanger... tube ID is 0.652 in... Since the RHR Heat Exchanger tube ID is greater than the largest assumed debris diameter... that could penetrate the containment sumps screens, tube plugging is not expected. Also, heat exchanger tube velocity is generally between 3-1 5 feetlsec... which is much greater than the sump velocity.

Since the debris is assumed to penetrate the sump screens at a lower velocity, settling inside the heat exchanger tubes is not expected. Therefore, blockage inside PB-1 and PB-2 RHR heat exchanger tubing due to debris laden fluid is not a concern..."

"The smallest ECCS and CSS process piping ID is 1.5...which is larger than RHR heat exchanger tubing ID. This evaluation has determined that blockage due to debris laden fluid inside RHR heat exchanger tubing is not a concern. Therefore, since the ECCS and CSS process piping is larger than the RHR heat exchanger tubing, blockage of ECCS and CSS piping due to debris laden fluid is not a concern..."

"Since [debris] terminal settling velocities are small by comparison to the process fluid velocities, introduction of debris into the instrument tubing is not expected. Therefore, blockage and abrasive wear associated with ECCS or CSS instrument tubing due to debris laden fluid are not expected."

"Furthermore... all of PB-I and PB-2 RG I

.97 commitment instruments tap into the process piping from the horizontal position to the upper half of the piping... This excludes the possibility of debris settling in the subjected instrument tubing. Therefore, blockage and erosive wear to ECCS and CSS instrument tubing due to debris laden fluid are not expected..."

"The most limiting orifice size in the ECCS and CSS... is 0.375 (CS Nozzles). Since 0.375 is larger than the maximum debris diameter of 0.0726", blockage is not expected..."

Technical Specification Surveillance Requirement (TS SR) 3.5.2.6 requires that every 18 months (refueling interval):

"Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet debris screens show no evidence of structural distress or abnormal corrosion."

This provides assurance that the screens are free from adverse gaps or breaches.

GL 2004-02 Requested Information Item 2(d)(vi)

The evaluation of downstream effects was developed using a relatively large inventory of fibrous I and particulate coatings debris, both of which either will be, or have been, reduced substantially (see the responses to Questions 9, 15 and 16). Therefore, the evaluation for excessive wear considered a substantially higher suspended debris concentration than is expected once all planned insulation replacements have been completed. As such, the following information derives from a conservative assessment.

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An evaluation of ex-vessel downstream components was performed to verify that close-tolerance subcomponents are not susceptible to plugging or excessive wear due to extended post-accident operation with debris laden fluids. The evaluation was performed using the NRC-approved guidance of WCAP-16406-P, Revision I, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, dated October 27, 2005 (ML052500596). No deviations or exceptions were taken. The following paragraphs are excerpted from the evaluation.

"Erosive wear in the ECCS and CSS components due to debris laden fluid has been analyzed. The PB-1 and PB-2 ECCS and CSS valves, heat exchanger tubing, instrument tubing, piping, and orifices were found to have adequate thickness such that erosive wear due to debris laden fluid will not compromise the design functions of these components for the required mission times."

"The degradation of hydraulic performance for the designated mission times is acceptable based on the methodology provided in [WCAP-16406-PI. Therefore, the pump capabilities credited in the FSAR and license bases analysis to ensure that Peak (fuel) Cladding Temperature (PCT) limits are not exceeded during the time and flow critical transient portion of a design basis Loss of Coolant Accident (LOCA)."

"The mechanical seal arrangement in the Point Beach RHR, CSS, and SI pumps are John Crane Type I and 1 B mechanical seals. These seals are rugged in their construction and capable of operating at elevated temperatures. The arrangement of the spring/bellows mechanism will not be affected by the suspended solids used in this evaluation for the specified mission times. John Crane Type 1 and I B seals have been successfully used in debris laden fluid such as pulp and paper, petrochemical, food processing, and waste water treatment. The design of the John Crane Type 1 and I B mechanical seals uses a non-clogging single coil spring to supply the seal face closing force. Based on the design of the John Crane Mechanical Type 1 mechanical seal, a single point catastrophic seal failure due to the debris laden fluid used in this evaluation is not expected for the specified mission times."

"According to the guidance provided in WCAP-16406-P, it is recommended that if the seal bushing in the mechanical seals are made of graphite or carbon then these seal bushings should be replaced with bronze or a similar material which is more wear resistant than the current graphite or carbon bushing. Since this evaluation is not taking credit for failure of the mechanical seals, it is not necessary to replace these seal bushings."

"The drill-through diameters in the mechanical seal gland of the ECCS and CSS pumps are larger than the largest assumed debris size, 0.0726" that could penetrate the containment sump screens. Since there are no filters, cyclone separators, or other line obstructions present in the circuit, clogging of the mechanical seal flushlcooling lines is not expected."

"Based on the above discussions, the RHR, CSS, and SI pump mechanical shaft seals are expected to perform satisfactory due to the debris laden fluid following the postulated LOCA for the designated mission times."

"The RHR and CSS pumps of PBNP are single stage pumps and do not require pump vibration analysis."

"The SI pumps at PBNP are multistage pumps and are evaluated for pump vibration. Since limited information exists from Point Beach and the SI pump manufacturer related to the SI pump rotor-dynamics, it is assumed that this information is not available. Therefore, the WCAP-16406-P wear model is used for the pump vibration evaluation."

Page 35 of 40

"The wear rate model in [WCAP-16406-PI was used to assess the extent of wear on the wear components and its effect on SI pump vibration and hydraulic efficiency. It was determined that following a LOCA, debris induced wear on the pump wear components is not expected to exceed the design running clearance limit specified in Appendix R of WCAP-I 6406-P for the each of the wear components during the mission time of 30 days.

Therefore, per [the WCAP-16406-PI criterion, the SI pump meets the requirements for vibration operability following a postulated LOCA and no further rotor-dynamic analysis is required."

No operational changes were made as a result of the downstream evaluations.

Question 21 The maximum aluminum concentration in the containment sump has been revised from a former calculation. The updated calculations show that less than 20 parts per million (ppm) will be the maximum aluminum concentration. Please provide the calculations used to determine final aluminum concentration, highlighting the differences in the revised calculations that show why a less than 20 ppm aluminum concentration is more representative of the post-LOCA sump environment. Please identify any important assumptions (e.g., pH) that significantly affect the calculation.

NextEra Response The previous calculation was completed in April 2006; two months after the issuance of industry guidance contained in WCAP-I 6530, Evaluation of Post Accident Chemical Effects in Containment Sump Fluids to Support GSI-191, dated February 28,2006, (ML060890509). In the absence of NRC guidance at the time, the calculation was performed using conservative assumptions and corrosion rates. While the information in WCAP-I6530 was considered, most of the calculation development had occurred prior to issuance of WCAP-16530, and the results of the Integrated I Chemical Effect Test (ICET) series of tests formed the basis of methodology and values used in the calculation.

During the development of this early calculation, it was believed that aluminum would remain substantially in solution at concentrations below approximately 50 ppm based on observations from ICET #4. PBNP uses a sodium hydroxide (NaOH) buffer, and had a considerable inventory of fiberglass insulation contributing silica to the sump pool chemistry. Therefore, the most similar test of the ICET series was #4. No significant precipitate formation had been reported in that test.

Subsequent developments, including both the NRC acceptance of the methodology in WCAP-I6530 and industry guidance to assume that sodium aluminum silicate is completely insoluble at all concentrations, led FPL Energy Point Beach to create a new calculation (Enclosure 6). The new calculation implements the methodology of WCAP-I6530 without exception or deviation.

The differences in inputs, assumptions and methodology between the two calculations are extensive and substantial, so a concise side-by-side comparison of the two calculations is not practical. The later calculation is not a revision or an update of the earlier calculation.

Since the April 2006 calculation did not implement an NRC-approved method of analysis, FPL Energy Point Beach no longer considers the results of the April 2006 calculation relevant to the resolution of GL 2004-02. Because that calculation is not valid, only the later calculation is provided in Enclosure 6.

Page 36 of 40

The calculation contained in Enclosure 6 used the spreadsheet tool distributed with WCAP-I6530 to determine the total quantity of sump chemical species. Multiple runs for various sets of postulated conditions were run to assess the sensitivity of the results to changes in parameters; however, some of the permutations represented non-credible accident sequences. The results of the multiple runs were then consolidated into summary tables for comparison and evaluation purposes.

Table 5-1 on Page 21 of Enclosure 6 is a matrix depicting the combinations of inputs used for each of the runs. Enclosure 6 is an abridgement of the calculation with most of the detailed spreadsheets and appended supporting material omitted for brevity. The detailed spreadsheets for the limiting design basis case (Table 7-1, Case 2.5) have been included.

The pH and temperature profiles used in the analyses are shown on Appendices A.6 through A.8 of Enclosure 6. The values for pH and temperature were all intentionally biased high to maximize corrosion rates and to conservatively bound the expected response.

Sump pH was maintained at a conservatively high 9.5 for each case. Similarly, the spray pH during the injection phase was held at a high of 10, while recirculation spray was held constant at a high of 9.5 (the same as the sump water). The timing of the transition from injection to recirculation was varied however, and found to be significant. Longer periods of spray injection with the higher pH spray resulted in a greater amount of corrosion from exposed metallic aluminum.

The other variables considered in establishing the chemical effects envelope (see Table 7-1, Page 29, Enclosure 6) were sump level (higher level results in a greater total quantity of chemical precipitate; whether corrosion inhibition is credited (it is not; but cases were run to determine the potential effect); whether pool volume is assumed to be mixed; and the debris mix. For the design bases cases, a worst case debris mix, that combined the largest quantities of insulation debris from all of the cases simultaneously, was used (Table 7-1, Cases 1.1 through 1.6, and Cases 2.1 through 2.6). Cases with debris mixes specific to the PBNP analyzed break results were also run to determine whether a significant reduction might be realized.

Table 6-1 on Page 24 of Enclosure 6 summarizes the most significant results. Cautions on usage of Table 6-1 are identified in the Design Review Comment Form located at the beginning of. These cautions describe cases in Table 6-1 that are applicable design bases (Case 2.5 is the limiting credible case), cases that are not, and how to properly obtain the species concentrations using the information in the calculation. The concentrations listed in Table 6-1 were obtained using a different sump volume that is inconsistent with the derivation of the precipitate volumes and should not be used.

Question 22 Please provide a table that shows how the mass of precipitate formed varies as a function of sump pH and sump volume.

NextEra Response The analysis used a constant pH profile that was intentionally biased high to conservatively bound accident conditions. As such, the analysis does not predict precipitate mass as a function of sump pH.

Table 7-1 of the calculation contains a summary of the results of the analysis runs. Cases 1.I and 2.1 were the base cases and were performed with high sump levels and unmixed sumps. Cases 1.2 and 2.2 used the same inputs with the exception of low sump level. Therefore, comparison of Page 37 of 40

these two pairs of cases provides a reasonable correlation between sump level and total precipitant formed.

While these results demonstrate that a higher sump level results in a higher total quantity of precipitate, these results are not considered valid design inputs, because the use of the unmixed sump assumption is not realistic and is not valid.

During the preparation of the July 2009 response, it was discovered that the value for total precipitant mass reported for Case 1.I in the April 2009 response was in error. It has been corrected in the above table.

PBNP Unit I Question 23 PBNP Unit 2 Case I.I Case 1.2 Please discuss why dissolution of concrete surfaces will not contribute significantly to the precipitate loading in the sump.

Case

2. I Case 2.2 NextEra Response Max sump volume Total Precipitant Mass Min sump volume Total Precipitant Mass Sodium hydroxide (NaOH) is used as a sump pH buffer at PBNP. This strong base favors the formation of sodium aluminum silicate. There is no significant source of phosphates as there would be if trisodium phosphate (TSP) was used as a buffer. Therefore, free calcium ions that may dissolve into solution will not precipitate out as calcium phosphate. This is demonstrated by the inputs (see Enclosure 6, Appendix A.1) where 10,000' of submerged exposed concrete were modeled) and the results of the chemical analyses (see Section 7 results for a discussion of the precipitant specie formed).

Max sump volume Total Precipitant Mass Min sump volume Total Precipitant Mass 43,31 ft3 248.3 Kg 22,995 ft3 169.7 Kg Question 24 43,317 ft3 274.8 Kg 22,995 ft3 182.4 Kg Aluminum coatings are present on the reactor vessel as well as other components inside the containment. The supplemental response states that these coatings are formulated to withstand high temperatures and would therefore not be expected to fail during a LOCA. Operating experience at several US plants indicates that high-temperature aluminum coatings can disbond under normal operating conditions. These coatings are unqualified coatings and as such are expected to fail in pigment sized particles (including coatings outside of the ZOI). The aluminum would be separated, at least partially, from the silicone resin. These fine particles could then be readily exposed to either containment spray or sump fluid and would be available to contribute to chemical effects. For any aluminum coatings that are not covered with insulation materials that would remain intact and hold the coatings in place, please provide justification for not including the aluminum mass in the chemical effects evaluation.

Page 38 of 40

NextEra Response Research conducted in response to GL 2004-02 established that the coatings on the Unit 1 steam generators do not contain aluminum, and that the Unit 2 steam generators are not coated. The replacement insulation on the Unit 2 pressurizer (and that planned for the Unit 1 pressurizer) is not susceptible to removal based upon line break analyses.

Other smaller, line breaks in the vicinity of the pressurizer that may be close enough to remove some of the insulation and expose the underlying original aluminum based coating (e.g., a spray line or relief valve line break) are minimum and would not generate a substantial quantity of fibrous debris.

The remaining component within a loss-of-coolant (LOCA) 201 that may have an aluminum pigmented coating is the reactor vessel. As discussed in the response to Question 1 above, the reactor vessel is insulated entirely with reflective metal insulation (RMI), and a break adjacent to the vessel would not result in a significant quantity of fibrous debris.

While the quantity of metallic aluminum that may be present in applied coatings was not explicitly accounted for in the chemical effects analysis, the following information shows that the effects are reasonably bounded by the analysis.

In the case of a break adjacent to the reactor vessel, it was postulated that all insulation on the vessel could be dislodged, and that any remaining aluminum coating on the vessel would be released to the containment sump. The PBNP reactor vessels can be approximated as right circular cylinders 33' tall and 12' in diameter. This provides a total surface area of approximately 1,470 ft2, including both the upper and lower heads.

Heat resistant coatings are typically applied as very thin layers 0.001" to 0.002" thick.

Conservatively assuming a layer 0.001" thick of solid metallic aluminum (no binder) gives a total volume of 21 1 in3 (0.123 ft3) of metallic aluminum. With a material density of 0.0975 lb1in3 for aluminum, this represents a total quantity of 20.6 Ibs (9.4 Kg) of aluminum.

A review of the spreadsheet for the most limiting design case for chemical effects (Enclosure 6, Table 7-1, Case 2.5) finds that 7.29 Kg of the aluminum that would be released to the sump is attributable to leaching from 1,276 ft3 of fiberglass debris, and an additional 5.72 Kg attributable to leaching from 323 ft3 of mineral wool. This is a total of 13 Kg of aluminum from fibrous insulation alone.

Since a break capable of exposing an aluminum coated surface would involve little, if any, fibrous insulation, the quantity of aluminum that would be released is bounded by the existing chemical effects analysis.

Question 25 Please provide an evaluation for the potential of deaeration of the sump fluid as it passes through the debris bed on the strainer. If deaeration can occur, please evaluate the effect that this can have on required net positive suction head on pumps taking suction from the sump as described in Reg Guide 1.82, Rev. 3, Appendix A.

I NextEra Response An analysis of the potential for deareation has been completed that uses the guidance contained in ISL-NSAD-TR-05-01, Development and Implementation of an Algorithm for Void Fraction Calculation in the '6224 Correlation' Software Package (V.V. Palazov, 0112005).

Page 39 of 40

The analysis determined that under the worst-case design basis head loss conditions for the screens (i.e. limited to 10' of head loss under cold conditions and with a minimum submergence of 2 at the top of the screens), the maximum gas fraction evolved would be 0.64%. This is considerably less than the conservative allowable limit of 2% previously established in station calculations for gas entrainment.

I Under hot sump conditions, the void fraction is even lower at approximately 0.06%.

Page 40 of 40

ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-191IGL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS PERFORMANCE CONTRACTING, INC.,

CALCULATION TDI-6007-06, REVISION 5, JANUARY 8,2009, TOTAL HEAD LOSS - POINT BEACH NUCLEAR PLANT - UNIT I

& 2 23 pages follow

Total Head Loss - Paint Beach ~ucle&r Plant - Unit 1 & 2 f6chni+l Feumerit No. TD1$067-C16 Revision 5 I CALCULATION COVER SHEET

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Technical Document Rev. No.

5 Addenda No.: MIA Calculation

Title:

Total Head Loss - Point Beach Nuclear Plant - Unit t & 2 Safety Related?

YES Calculation Verification Mafhad (Check One):

Design Review CJ Alternate Calculation

[Z1 Qualification Testing Scope of Revision:

Specific Revision to address AREVA Large Flume testing results and addition of temperature range for repotfed values. Revision 5, Pages: All I

Documentation of Reviews and Approvals:

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m18001dB Rev 5 T&I Heed Loss.drx:

Total Head Loss - Paint Beach Nuclear Plant - Unit I (Pr 2 Tebhnitial Document No. TDI-6007-;08 Revision 5 1 CALCUtATION VERIFICATION CHECKLIST I

i Verified bylD Initials:

-.- -... -..,.. - ~abulation-~itler~abl Head Lass-;-- Paint Beach Nuclear Plant - Unit 1 8; 2--- "'

Revision: 5 nla cHEcKL1s-r I

Yes No 1

2

3.

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6.

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Were inputs correctly selected and incorporated?

Are assumptions adequately described and reasonable?

' Are the! appropriate quality and quality assurance requirements specifid?

Are 6 e ep#li*blP codes, standards and regulatory requirements identified and men ;.

Have applicable abnsbucbn and operating experience been considered?

Have the design interface requirements been satisfied?

7.

W& an a~~mpriite design method used?

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Is the output,reasonablv compared to inpul?

a n. 5 k y specified parts, equipment, and processes suitable for the required 1

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Has the deslgn properly considered radiation exposure?

I the acceptak wikria inwrporated in the design documents sufficient to a l l o w ~ w, ~ t i o ~.

MaJe ad$jqliaW preaperational and subsequent periodic test requirements been BpriJtlad? -,,,..

re adequate handling storage, daaning and shlpping requirsments specified?

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review, approval, &ntian, etc..

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Total Head Loss - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TlJ-6007-06 Revision 5 1 TABLE OF (;;ONTENTS CALCULATiON COVER SHEET

^_......-_- *-.... -_. - _..,....... _ _.. - -.-.-

CALCULATI~N VERiFlCATIQN CHECKLIST TABLE OF CONTENTS 1.0 Purpose and summary Results 2.0 peflnftions and Temfnology 3.0 Facts and Assumptions 4.0 d g n Inputs 7.0 Calculation(s) 7.1 Clean Stminer Head Loss 7.2 SCrainer Debris Laden Head Loss 7.3 PBNP Strainer Debris Laden Head Loss Summary 7.4 Debris Bypass 8.0 Conclusions I

9 0.D Drawings

~n&cti~EjaTs Point Bath Debris Laden Head Loss Temperature Corrected values I

TABLES Table 1 ToQl Debris Laden Head Loss - Temperature Adjusted to Range 212 "F

tb" 32 OF Tabk 2 Tcrtal Conrected Clean Strainer Head Lsss (TCCSHL)

Results, ft' Tabb 3 &an Strainer Head Lass - Regression Formula, ft Tbble 4 4RL T&ted Clean Strainer Head Loss Table 5 ARL Tested Debris Laden Head Loss Table 6, Wafer Dynamic: Viscosity Table 7,Tbtai Debris Laden Head Loss Table 8 Total Debris Laden Head Loss - ~ern~ekture Adjusted to Range 212 O F to 32 "F' I

lUl-6QD7-05 RBV 5 TW Head I.oss.d~

Total Head Loss - Point Beach Nuclear Plant Unit I & 2 Technical Document No. TD1-6007-06 Revision 5 1 4.0 Purpose and Summary Results I

The-US. Nuclear Regulatary_Cornmission (USNRC)-imgeneric safety-issue. (GSI)................

191 identified it was possible that debris in PWR containments could be transported to the emergency core cooling system (ECCS) sump(s) following a main steam line break (MSLB) and/or a loss of coolant accident (LOCA). It was further determined that the transported debris muld possibfy dog fhe sump screensfstrainers and impair the flow of water, thus directly afkcting the resultant operability of the various ECCS pumps and the containment spray (CS) system pumps, and their ability to meet their design basis function(s).

In order to address and resolve the various issues identified by the USNRC in GSI-AQ1, utilities have implemented a program of repiacing the existing ECCS sump screens or strainers with new and improved designs.

In order to address and resolve the speoifio issues associated with USNRC GSI-I 191 for the Point Beach Nuclear Plant - Unit I

& 2 (PBNP-112) entered into a contract with Performance Contracting, Inc. (PCI). The primary objective of the conhct was for PC1 to provide a qualified sure-FIO@ Suction @miner that has I been speci!ically-designed for PBNP-la in order to address and resolve the NRC GSI-191 ECCS sump clogging issue.

PC1 has prepared a Qualii~tion Report specifically for the subject sblner. The Q u a l i f i c a t i o n - R e p o r t - i s - a - c o m p i l a t i o n - m o n s.

that support the strainer qualification.

.___-,.a _.._.-_-.--.--

As part of the PBNP-?I:! Qualicatian Report, PCI has performed a number of hydraulic calculations in support of the replaiwment sure-FIO@ suction Strainer. I This calculation TDI-6007-06, Tofai Head Loss - Poinf Bt?ach Aluclear Plent -

UniL 7 & 2 is one df-a number of hydraulic calculations that specifically supports the design and qualification of the subject strainer.

i i This calculation addresses the total expected-head losses acrqss the suction I %

stminer assembly that has been designed specifidly for PBNP-IR. This expected. head loss is the combined total of the clean head loss associated with the strainer and attached piping, and ttie,debris head loss. The clean hesld.loss was.determined in calculation TDI-60071?5,. Clean Head Lass - Point Beach Nuclear Plant-Unit 7 $r 2. The debris head loss is determined based on actual test 'results for a 'PBNP-I12 strainer that has been specifically corrected %or the PBNP-112 Specification de~ign~basis p o s t - L ~ ~ ~

water temperature. The tests were perfoftned at the Afden Research Laboratory and independently verified by AREVA [Reference 9-41. The calculations are only pertinent to PCl's Sure-I low@ Suction Stminer.

Date ;b lo7

Total Head Loss - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TD1-6007-06 Re~islon 5 ]

The PBNP units each have two (2) separab recirculation strainer assemblies that individually and specifically feeds either the 'A' or 'B' train ECCS and CS system,- Each. horizontally oriented recirculation strainer assembly 'is comprised- -..-

of fourteen (94.)' modules each made up of ten ($0) strainer disks far a total strainer area of 1,904.6 It!,

or a total of 3,809.2 I'tz fbr each pair of strainers associated with one of the PBNP units. Flow leaves the strainers and enters a combination af pipe and mings before discharging into the containment outlet.

I PC1 drawings [Dlawings 10.1 - 10.11, inclusive1 provide details d the subject I configuration.

Based on actual test results perfoimed by PCI, it was determined that clean strainer head loss (CSHL) for the sure-FI&

Suction Strainers is a function of I tvvo (2) independent variabbs: (1) strainer internal ~jore tube diameter and (2) water flow rate exiting the strainer assembly.

The quotient of these two independent variables, In turn, results in one independent variable, which is exit velocity (VE).

The Clean Strainer Head Lass (CSHL) depends on Be specifc plant conditions '

for PSNP-IM. The resuits af the Total Corrected Clean Strainer Head Loss (TCCSHL) caIculatian considering these conditions, including uncertainty, was calculated fa be 0.660 feet of uirater. Full scale testing by AREVA at ARL found the aotual CSHL to be 0.408 feet of water with he plenum head loss added.

I The CSHL calculations account for the specific design of the PBNP-I12 strainers.

..-...-....--...-The,debris. laden-head loss. utilites-a-series. of W, mnduded-_with. a,.reduced-,-

sale strainer (with aocompanying reduced surface area, reduced water flow rate, and reduced quantities of simulated post-LOCA debris). Each of the test parameters is reduced by the same fraction (i.e,, a percentage of the full scale far each pardmeter). One parameter that is not changed is the approach velocity. It is kept the same a$ the full-scale design. The approach velocw~

is defined as the quotient of shiner flow ate and total surface area. The resultant value is 0:0026 1 Ws, an extremely low approach velocrty when compared to the design vdue 'for the orfglnal ECCS screens. The head loss acmes a particular debris bed is a funcition of two. hydiaulic variables: approach velocity and water dynamic viscosity.

Acoordinglly, the strainer specific test results, utilizing accurateiy simulated post-LOCA debris and the design approach velocity, will be awurate for. a given water: dynamic viscosity, a parameter that is a function of water.

- temperature. Therhfore, the test results require correction for the viswsity at the specified'post-LOCA water.'temperature, 212" F in the case of PBMP-112. The test results will then be representative of the full scale strainer under specified past-LOCA.mnditicms.

1

Total Head L o s s - Point Beach Nudear Plant - Unit I & 2 Technical Document No. TDl-6007-06 Revision 5 1 The resuits of the cafculation are provided in Table 2. This calculation utilizes the resub of clean strainer head loss testing previously canducted at the

--- Fairbanks Morse-Pump-Company and-the-Electric Power-Research Institute's-(EPRI) Charlotte NDE Center for Protofypes I and 11, respectively that is applicable to the current PC1 sure-~lowQb Suction Strainer. It also uUliis the I actual test results of the PBNP-2 strainer that were performed at the Alden Research Laboratory {ARL). The resufts of the subject two tests fonn the basis for calculating the PBNP-ID strainer total head loss, I

I Table 't - Tolal Debrls taden Head Loss - Tempsrature Adjusted to I

/

I I

l'emperakrre O F I

Mead Loss Ft:

I This calculaticm does not address the subjects of possible air ingestion, potential vortex, and void fraction'issues as they relate to the PBNP-112 strainer. These topics will be specifically addressed iii calculation TDI-60D747, Air Ingestion, V o h x & Void Fmcfion - Poini. Beach NucleaiPiant Unit 7 L 2.

It was cancluded that this calculation, an integral portion of the Qualification Report.completely supports the qualification, installation, and use of the PC1 sure-FI~W* Suction Strainer for Point beach Nuclear Plant Unit 1 & 2 without any issues or reservations.

The following D e f i r S i.& Temjinology are defined and described as they are utiliied iti &is calculation..

Originated. By; Date 1 /L,/o?

I..

T M ~ - ( ] B Rev 5 Tllfel Head -doc

Total Head Loss - Point Beach Nuclear Plant - UnB 1 & 2 Technical.Document No. TD1-6007-O6 Revision 5 1

~urel~low@

Suction Strainer - Strainer developed and designed by Performance Contracting, Inc. that employs S U ~ P F I ~

technology to reduce inlet approach.veloci-.....

I Emergency Core Cooling Systern (EGGS) - ~ h e ECCS is a wmbinaticn of pumps, piping, and heat exchangers that can be combined in various configurations to provide either safety injection or decay heat cooling to the reactor.

I I

Point Beach Nuclear Plant Unit f & 2 - also known as Point Beach, PBNP-IR, and PBl112.

I AREVA WP, Inc. -- also known as A W A. AREVA is cctntrqcted by PC1 to prepare and implement the Test; Plan through Alden Research Laboratory. ARL will implement the testing under the AREVA qualify program.

Aldm Research Laboratory - also known as ARL. ARL is contracted by AREVA to perform the testing in their facility located at Holden, MA. The testing wSII be performed by ARL under the diMion of PCI and ARWA.

Clean Strainer Head loss (CSHL) - Is the caloulated head loss for the sure-FIO@ Suction Strainer based on actual testing petformed at the Electric Power Resmrch-InstituteT's-(EPRI)-Gharlotte-.tank Gen@rT-and-Fairbanks-Pump-Company Hydraulic Laboratory. The later tes;ting did not involve any debris.

total Debris Laden Head Loss (Tampepiture - Corrected - AWL Test Reoulb) fl'DLFiL) - Is the TCCSML added to the A-DLHL for the sure-~lovu@

Suction Warner based on the PBNPIl/Z testing that was perfanned at the Alden.

Research laboratory (ARL). The PBNP-1/2 strainer testing performed at ARL is documented in peference 9.41.

Tofa1 ~ o f k $ e d Clean Strainer Head loss (TCCSHL) - Is the total head loss associated with the complete sure-FIO@ Suction Strainer installation canfiguration for PBNP-1/2 (i.e., strainer and connecting piping and fittings) including uncertainty.

ARL Test ~erc$ts - Debris Laden Head LOSS'- Temperature Corrected {A-DtSL) - Is tbe temperature carrected head lass for the PBNP-112 ~ure-~low@

S@an $trainer based on the ARL test results utilizing the desigf'tiasis debris loading [Reference 9.41.

Originated By:

TD1.6007-08 Rev 5 Tctal Head L d o c

Total Head Loss - Point Beach Nuclear Plant - Unit l & 2 TechniwI Document No. TDI-6007-06 Revision 5 1 3.0 Facts and Assumptions

... -..--... The.following Facts:.(designated as [fl). &.Assumptions. (designated as [A]) were -.-- -.--.- -.....

utiI'id in the preparation of this calculation.

3.1 For the speclfled minimum post-LOCA water temperature of 212" F, the containment air pressure is 14.7 psia [a.

3.2 A flow velocity of 0.0026 fps would be characteristic of the PBNP-I& I strainers, through a debris bed consisting of fibers and particulate is 100%

viscous flow.

Accordingly, the head loss is linearly proportional to dynamic viscosity [A].

3.3 A scale strainer, which is designed to maintain the same approach velocity as the full scale production strainer, can accurately simulate the performance d the full scale production strainer sa long as the same scaling factor is used for strainer area, water flow rat&, and debris quantities. The scaling factor is defined as ratio of the surface area afthe scale strainer and the surface area of fhe full scale production strainer IA].

3.4 The head lass resulting from flow through a fiber - particulate debris bed at the approach velacity fir the PBNP-iR strainer (0.0026 fVs) I

[Rebrence-9.3]Ii-is-1 OO%-visaus-fl~w~

as-opposed-teinertial-flm.As viscous flow,. head loss Is linearly dependent on the product of visvsity and veloc-&.' Theref~re to adjust the measured h - ~ d loss across a debris

-. -.. --,.i...

,....,-.--_-I bed with colder water, a ratio of water viscosities, between ihe warmer specified post-LOCA water temperature and the colder test temperature, can be multiplied by the measured head loss to obfain a prediction of the head fuss with water at the specified post-LOCA temperature [A].

I 3.5 The total strainer head IDS$

call be calculated by taking the sum of the calculated value of the Clean Strainer Head Loss [pefierence 9.33 and the I

temperature acijusted, tested debris h&d loss [A].

3.6 The PC1 ~ u r e - ~ l d ' ~ u c t i a n Strainer installations for PBNP-lf2 are the same. However, there are a number of differences with regard to the strainer discharge piping configuration for eakh of the four (4) strainer installations. Based on an assessment of each of the faur (4) strainer.

discharge piping configurations, the piping 'configuratigp a$sociated with PBNP-I Strainer.' "8" would result in the greatest head loss due to this specific strainer configuration having the greatest equivalent pipe length (i.e., combination of strqigM pipe length and. number and me of f~ngs).

Accordingly,,the PBVP-1. Strainer "8" piping aonfiguration will be Originated By:

Dab

Total Head Loss - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TDlSj007-06 Revision 5 1 conservatively utilized as the basis for PBNP-112 to bound both units and all str&ner discharge piping configutations with respect to strainer dean head-loss F- $.A],-.--

3.7 Utilization of the PBNP-I&! testing program performed at ARL and the subsequent test'data and results ~sfbmnce 9.41 to support PBNP-112 calculations are based on the PBNP-I12 Projecf specification [Reference 9-11 F=l*

3.8 Any and all references to or discussion of the PBNP-112 strainer testing, test results, and sirnifar related activities and discussions, actually means the PBNP-ll2 stminer tesfing at ARL and the subsequent test results

[Reference 9.41 [F].

3.9 PC1 has assumed that unknown piping, tubing, or openings added after the strainer installation are not directly connected to the PC! ~ u r e - ~ l o f l Strainer and are sealed &e., fluids andlor gases cannot enfer andlor exit through the openings) [A].

3.10 The input data used in MS, Excel spread sheets (if applicable) was verified by comparison to the design drawings and associated dim'erisions. The calculations resulting in output data are independently verified by hand ir;ralleuiatlon~Therefore~MS-&cel-spread-sh~t-is-~~nvenien~f~~~

but

.. not

,. relied upon as analytical software IF].

I-Originaterd By:

3.12 The Design Basis minimum specified post-LOCA water temperature is 222°F. The 212 "F temperature will be utilized to evaluate the total head loss. However; PBNP-112 has requested a series of head loss values for water with temperatures between 212 O F and 32"F, at 20 degree increments. The Total Debris Laden head loss will be determined at 212 O F and utilize the tested clean strainer head loss results to determine the final iota! Deljris Laden head loss. PC1 will utilize a tempemture mrrectlon correlation to.,obtain the subject head losses for the range

.. betwvjen.212-°F and.32-F temperatures [Q..

3.13 Reynolds numbers are calculated in attachment A1 for temperatures between 32 "F and 212 O F using the flow and piping details from the CSHL calculation [Reference 9.31 and shown in Attachment I, Table Ad. [F].

ml&OD;I-O6 Rev 6 T@al

~oaadoc 4.6

~es@ii fnputs The hrliwing mrnbiiation of PBNP-IR and PC1 Design inputs were ulilhed in the prepatation of this cal~ulation.

Total Head Loss - Point Beach Nuclear Plant - Unit 1 & 2 '

Technical Document No. TDI-6007-06 Revision 5 1 4.1 Point Beach Nuclear Plant Specification, Specification No, PB-681,

.... Rapiacement ofConItainment Sump Scmens, Revision 1, Augqst 25,2003-..

[Reference 9.11. This document provides design input associated with strainer flow rate, water temperature, and' the maximum allowable head loss.

4.2 Performance Confracting, lnc. (PCI) Calculation TDI-6007-02, SFS Sufkce Area, Fiow and Volume Calculation, Revision 1 fReference 9.1Q.

This document provides relevant dimensions and other infomation specifically assaciated with the PBNP-1l2 strainers.

4.3 PC1 Calculation TDI4007-03, Core Tube Design - Poinf Beach Nuclear Planf - ?/2, Revision 0 ~efewnce 9.211. This document provides relevant d a t ~

with regard to flow rate in the PBNP-IM stminer.

4.4 PC1 Calculation TDl-60076, Clean Head h s

- Point Beach Nuclear Hmt - I&, Revision 4 [ReFewnce 9.3l. This document provides the I 1

head lass associated with the "cleann PBNP-If2 strdner and attached pipe and fHngs.

PCI utilized iwo (2) distinct methodologies based on the entire strainer assembly I oo'nfiguratian to detemine the max'mrurn thin bed head loss for this calcuiation:

(1) c&ulate the Clean Head Loss for the PBNP-112 strainer fRefervbnce 9.31 and (2) determine the.peak design basis head loss based on reduced 'scale strainer testing utiliilig the FBNP-'I& specified design basis water temperature of 212' F m~~femnce 9.11 (adjust from the test water temperature to the specifid water ternpewtb're) and the' PBNP-If specific debris mixture. The individual head lass 'resub obtained are added together to obtain the t~tal design basis head loss for, the! entire strainer assembly configuration.

4.5

. ARWP Engineering lliformafian Record, Document Identification No.66-809;395?-QO~Poinf-Beach-Tesf-Repori ECGSSfrainer-Perforancle----

Testing Reference 9.41. These documents provide the method and value The quanMy of fiber and debris ur&irithe scale strainer testing is ba'sed an the illebri's load stated in [Reference 9.61 with a 75% fiber.redudion. PC1 believes that the assum~tions are conservative and are supported by the PBNP-112 test

'rents 9.43. Debris testing. is then used to determine if the (

of the.teskd-debris head,.lws anb.fbemeaha~~sm crf.~djli%fing the debris head Loss to the specified post-LOCA water temperature.

1 Originated By;

//th?.

Date TDl-96 Rev 5 Tatat Ha& Losadoo P

i '

Total Head Loss - Point Beach Nuclear PIant - Unit 1 & 2 Technical Document No. TD13j007-06 Revision 5 1 6.0 Acceptance Criteria I

strainer is adequate to meet the specified design cond'ttions. The actual scale strainer testing results are used as the basis for concluding that the strainer

.... - -....... bounds the propased size. and. design far. the. actual.. PBNP-'l/2 strainer. PC1 believes that the assumptions are conservative and are supported by the PBNP-112 strainer test resuits at ARL [Refc?rence 9.4l; PBNP-?& specified peference 9.4 ]'that the total debris laden strainer head loss be calculated at a temperature of 222 O F in order to meet fhe required design basis NPSH requirements, and further specified [Reference 9.q that a range of head loss values be determined between 212 "F and 32 O F at 20 degree increments. The head loss values for the full range of temperatures will be presented In Table 8.

The DLTHL-TC includes the strainer, strainer discharge, and addressing all possible debris loading combimations. This calculation addresses the possible I

debris loading combinations, and calcul.ates the head loss associated with the strainer and the strainer discharge flow into the sump.

PC1 has uptimized its design of the ~ure-~low@

Sudon Strainers for PSL-2 to ensure preservation of head loss margin.

1 ir, or&'fb-'~-emifie if--- ~

a

~

c*.dikti"dt"dZilci iIEitiO m-& hljdblog,es are I emplayed as described in section 5.0 Ma;Ulodofogy. One methodology is utilized b sspamtely calculate the head loss far the bare strainer, attached pipe and fittings, and the second methodology is used to determine the T~tal Debris Laden Head Loss - Temperature Corrected (TDLHL) design basis head loss based on PBNP-112 speoific reduced scale strainer testing using a full sized representative strainer module with debris generation allocation mixture (A-DLHL).

NOTE: The PC1 ~ure-Flow@

&uction Stminer installation for PBNP-412 is ]

very similar in nature with only a slight differrenee wifh' regard to ther strainer discharge piping configuration. Accbrdlngly, the discharge piping configuraticln differences.are greater far PBNP-I, and its associafed

. 'oonfigqration will be utiI@xl to iooun.tb bath unifs with-respect' to strainer clean head lass.

I

Total Head Loss - Point Beach Nuclear Plant - Unit 1 BL 2 Technical Document No. TD1-6007-06 Revision5 1 7.1 Clean Strainer Head Loss

.---.-.- -..-. - As summarized. in-Table 2 below, PC1 calculated the clean stainer head -,-.

loss,(CSHL) for the PBNP-IR strainer in @tefemnce 9.33.

The total CSHL includes the expected head losses from the strainer @are), I attached.strainer discharge piping and f ~ n g s connecting the strainer to the sump pit, and that associated with the water leaving the strainei discharge pipe as if enters the sump pit I

Table 2 - Total Co-d Clean Strainer ~ e a d Laps (TCCSHL)

Results, Qt TCCSHL @ 212 O F I

0.560 i

The TCCSHL below includes a strainer only head loss calculated using the PC1 regression.formula as presented in Beference 9AI. The strainer regression formula value for head loss and its temperature corrected value frrr 212 O F are presented in Table 3.. This value will be used later to determine the total debris laden head loss. Tperature correction is calculated.using methodology provided in Section 7.2.2.

7.2 Strainer Debris Laden Head Loss t

The PBNP-In strainer modules werd-sized based upon meference 9.lq.

The arnount.of and the make-up of the debris that is specific to PBNF?-112 was provided in [Reference 9.61.

The debris mixture specified ' h Reference 9.43 ws further analyzed by utilizing [Reference 9.121 to deyeblj. fie actual debris mix (i-e., debris quantity and type) %r the testing' of the PPBNP-4/2 specific strainer m8kmnctrs 9.5J.

I The PBNP-IE Clean Strainer Head Loss tests perfarmed at ARL are summarized in Tabla 4.

Tabfa 3 - Clean Strainer Head Loss

... - Regwaion,

Fomulla, R

smi."&f (2,* *F)- - - - -- - -

"' om*93&

The CSHL basedonthe test 'result fmm ARL was temperature corrected to the PBNP design,basis temperature (i.e., 212 OF).

See Seciion 7.22 far temperature c o d o n methodolagy.

. i I

Originated By:

Date I

7olgOW.08 Rev 5 Tdel Head t 4 ~ ~ A f f i Page 12 of 19 '1 i :.. '..

. I

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.(.c y&.,>s)jb,,$.,@$ $!$;j2?$;L$

v

a... cfik%4!<t.

L ~. ~ ~ ~ !. ~. & :, ~,, ~ ~

bv$::$:~\\,;H;;~~.4::i::,t~~k::;:!;~~>

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p..,y. ;> p:.:v %\\:.,

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• . ~:~~$;~$;~$~;kj;;<;.$2.;~~~&~~;$;#

/.:.:.'" :."..,?.,:

1-'..i*,-

.... L L ' ~ + { C ~

,,.-.., c.,,.. ~:?*:;.;<n$:;i'.::dn. "hip&-,c:::::,.:,m';:

.;;

mir..

.dh..NL.,v.+. :*>v

,..:6\\;!;

fa.'?,.,...

$ :\\. ::., k,,,?.:,!o>,.J.!:$i$?

. Total Head Loss - Point Beach Nuclear Plant - Unit 1. & 2 Technical Document No. TD1-6007-06 Revision 5 1 The PBNP-dl2 Debris Laden Strainer Head Loss test is summarized in Table 5.

-C I.

AREVA Test No. 6 [Reference 9.4 is the Design Basis test for PBNP-la.

The PBNP-I12 Design Basis test is intended ta show recircu1;rtion at 2200 gpm with a water level above the top of the PBNP-?I2 strainer, Additional information regarding both the Clean Head Loss and Debris Laden Head Loss testing that was performed. at ARL is specifically discussed in detail in FeEemnce 9.4.

I I

Table 4 - ARL Testecl Clean Sfminer Mead Loss

-- -I--...

f 7.2.11 Temperature Correction Strainer ~ e b & Laden Head Loss Corrected Glean Strainer Head Loss, ft,of water (212 "F) 0.0417 T W ~

Shiner FlowI gpm

{Sealtad)

I Tabla5=ARLl~8bd.Debris-Laden-Mead.Loss v-Oiiginaterd By:

170.7 1

111.1 0.090 I

Ave.

Temp. (OF)

The dynamic vis~sity'of the specific ARL test water temperatures and the PBNP-112,Design Basis temperature is taken from

[Reference 9.91, fable 6 pravides 'a summary.of the dynamic viscosities associatqd 4th the various test and Design Basis water temperatures that are utilized in this r=alculation, clean SWner Head Loss, ft of

'water

!-I

- -. f...--*&&-

Test No. 6 - 170.42 gprn scaled Row P~

Stiai-ner Debris Laden Head Lass HL c o m ~ a d to 212 O F (Ft of Water) 3.066 T M

Temp, DF k& - Ftof WaBr

,, ' 6.448

Total Head Lrrss - Paint Beach Nuclear Blani - Unit 4 & 2, Technical Document No. TO?-6007-06 Revision 5 1 The head loss for low velocity water in the laminar flow region.

through a debris bed of fibers plus particulate is linearfy dependent

.....-. --..... on the water's-dynamic viscosity;--The PBNP Design Basis water-.:-. -......---

temperature is 212 "F IReference 8.11.

The debris head loss requires correction to this temperssture to determine the head loss.

at the.JPBNP-l/2 Design Basis temperature of 212 "F. The strainer '

debris laden head lass for low velocity wier flow through a debris bed of fibem plus particulate is linearly dependent on the water's dynamic viscosity meferen~e S?O],

I Table 6 - Water byhamicc Viscosity

11.

A head loss correction, utilizing As~uanpticrn 3.4¶ which is based on the standard debris head lass equatian [1Refemnce 9.111 can be used to calculate a temperature adjusied debris head loss, HLtA.

The NLyA adjusted temperature can be calculated by taking a ratio I

of dynamic viscosity values at the two different temperatures being considered (i.e., the. test water temperature and the PBNP-IR specitic post-LOCA sumpw,ater temperature).

Equaticjn 1 HLTA = HLDL,~

(ps l pm) t I

Page 14of10 I

Total Head Loss - Point Mach Nuclear Plant - Unit 1 & 2 Technical Document No. TDI-6007-06 Revision 5 1 Where!

HLD~c=

Debris Loaded Head Loss, ft usy

=dynamic viscosity at the post-LOCA specified-[-.-- --..-.-.....

temperature

=dynamic viscosity at the average tested temperature HLrn =temperature adjusted debris head loss, 8 I

The HLm, as calculated above, is added to the clean strainer head loss that r e ~ u b in the DLT'HL-TC for PBNP-V2 based on the specified post-LOCA Design Basis temperature.

7.3 PBNP-112 Strainer Debris Laden Head Loss Summary Table 9 summarizes the bounding values of head loss discussad above.

All head losses are in feet of water. It was also wnservatively assumed to add 6% for ugcertainty and 10% for strainer discharge and collection head loss associatdd with the Clean Strainer Head Loss (CSHL) calculations to address any non-bonservatism inherent in the use of standard head loss correlations [Reference 8.31. The Clean Strainer fiead Loss values are based-a~ [RePe~rrec3i-9;3]~the-tested-str~lliner-debri~-ladn-h'ead-las~is based on Sest3oaa $2, and the ternperaturn corrected debris laden head lossfor: post-LOCA conditions is basedm Se~Bon..7r.2.2,..-..-......

PBNP-1/2 has requested a series of head loss values be calculated for water with temperatures beween 32 "F and 212"F, at 20 degree increments. The Total Debris Laden head loss will be determined at 212 O F and utilize the tested clean strainer head loss results to determine the final T&1 Debris Laden head, loss for this range of temperatures. PC1 will use tkrnperature conection correlation methodology presented in Attaohment li to calculate the subject head losses for the range between 32 "F and 212 OF temperaturns. See Table 8 for resufts of head losses calculated for the specitled range.

lDl.WM74$ Rw.5 Totel Head Loag.doo '

F ~ f S o f l S [

Total Head Loss - Point Beach Plucleac Plant - Unit 1 Ifr 2 Technical Document Ma. TDI-6007-06 Revision 5 1 7.4 Debris Bypass As part of the PBNP-?I2 strainer testing plan, water samples of the debris mixture (i.e., debris type and quantity) were taken of the strainer discharge water, immediately adjacent to the subject straln'kr. This was done in order to determine the sire, quanfi and weight of the various debris mixture components (i.e., fibers and suspended particulate) that were being transported through the strainer during the test. Analysis of the debrls bypass data is not part of the soope of this technical document.

The debris bypass analysis results can be found in the ARWA Test Report [Reference 9.41.

8.0 Conclusions I

This calmlatian and supporting porfions thereof, considered all of the previous testing that has been performed for the various PCI ~ u r e - ~ l o g Suctkn Strainer, including uncertainty. The temperature wnected head loss associated with the debris only on the strainer is 3.066 feet af water at 212 OF. The predicted result for tqtal debris laden head loss, the sum of the calculated clean strainer head loss including uncertainties and the stminer debris laden head loss is 3.474 feet of wafer at 212 "F.

It was concluded that this specific calculation cam letely supports the

......... qualification, installation.. and-use of the PC1 Sure-Fled suction Strainer. for I-..

Point Beach Nuclear Plant - Untt I

& 2 without any issues or reservations.

I TDl.6007-06 Rev 6 TOW Head Loq.doc

Total Head LOSS - Point %ach Nuclear Plant - Unit I & 2 Technical Document No. TDI-6007-06.

Revision 5 1 9.2 Performance Contracting, Inc. (PCI) Calculation TDI-6007-03, Infernal Core Tube Siof Design for PBNPs Sucfion Sfrainers, Revision 0 i

I

....-.... 9.1..... Point Beach-Nuclear Plant, Point. Beach Nuclear.. Plant.. Specification,. --.-.... - -....

Replacement of Containment Sump Scmns, Specifi~ti~n No. PB-681, Revision 2 I

t 9.3 Performance Contracting, Inc. (PC[) Calculatian TOf-6007-05, ban Head Loss - Poi& Beach Nuclear Plant - Unit -1i2, Revision 4 I

I 9.4 AREVA Engineering Information Record, Document Identification No. 66- '

9093957-002, Point Beach Test Reporf for ECCS Sfminer Performance

Tesfhg, 9.8 Crane Technical Paper No. 410, Flow of Fluids fhmugh Valves, Fiffings, and Pipe, 1988

,a.

9.5 AREVA NP Engineering Information Record, Document Identification No.

I 51-9021513-000, Poinf Beach Un#s f & 2 EGGS Shiner Performance Test Plan, Revision 0 8.9 Spirax Sarco USA Webpage (htt~:iiws~iraxsar1=0.~am/us/rasources) 1 G.6 PBNP Letter No. NPL 2008-0462, Design Information Transmittal (DIT) in' Support of Sump Strainer Qualifilcation T&~ng the week of July 14,2008, Ju'ly 9,2008 9.7 PBNP Letter No. 'NPL 200&0264, Design Infarmation Transmittal (D11") in support.of.Ek-G~!!~~lll~tior!

-r[D!BQ7:0!% Re!!z.5, O~~O?K.!~,-T.O.IIS-Q.10 USNRC NUREGICR.6224 "Conefation", publicly available soffware I

9. 1 NEI Cl44l7, "Oressuflzed Water Reactor Sump Performance Evaluation Methadoliigy", Rev. 0, DWembei, 2004 I

9.12 Performance Contrading, Inc. (PC!).&aloulation, ?'Dl-6007-02, SFS Suface A m, Flow and Volume C E ~ ~ C U I E ~ O ~,

Revision 2 9.49 Fluld Mechanics With Engineering Applications, Robert 1. Daugherfy and '

Joseph,B. Franzini, Seventh Editian, 1977, McGraw-Hill Bookl Company, IRC.

TmtW7.06 Rev 5Tel Eidad ~css.dw'.

Page 18 or19 'f

Total Head Loss - Point Beach Nucliear Plant - Unit I & 2 TechnicaI Document No. TD1-6007-06 Revision 5 1 10.0 Drawings Originated By:

Date

..& ---16,l--

SES-PBI-GA-00,- ~ev~sion, 9,. Point Beach.Unit-I, ~ure-~low!?,~trainer,......,

Recim Sump System 10.2 SFS-PBI-GA-02, Revision 9, Point Beach Unit I, s u r e - ~ l d miner, B miner 10.3 SFSPBI-GA-03, Revision 9, Point Beach Unit 1, sure-low@ ~ i k i n e ~,

A strainer 10.4 SF$-PB%G.A-M, Revision 6, Point Beach Unii. 1, sure-Flo@ Strainer, Piping 8 Layout 10.5 SFS-PBI-GA-05, Revlsion 9, Point Beach Unit 1, sure-FIO@ Strainer, Pipng A Layout 10.6-SFSPBI-PA-7100, Revision 4, Point Beach Unit 1. ~ure-~low@

Sfrainer, Module Assembly 10.7 SFS$B~-GA-00, Revision 3, Pdnt Beach Unit 2; sure-FIO@ Sfminer.

Recim Sump Sysfern Layout 10.8 SFS-PB2-GA-02, Revision 9, Point each Unit 1, ~ure~~low@'-~@&~~

A

-- - -Strarner.- ----.- -

..10.9 SFS-PB2-GA-03, Revision 9, Point Beach Unit I, sure-~lo@ Sfrainer, L3 Strainer 10.1 0 SFS-PB2-GA-04, Revision 5, Point Bsmh Unit 2, sure-low@, ~fminer, Piping Assembly Layout

-i 0.1 I sFs-PB~-PA-~I 00, Revision 3, Point Beach Unit 2, sure-FIOW' Strainer, Madufe.Assembly.

Total Head Loss - Point Beach Nuclear Plant - Unit 4 & 2 Technical Document No. TDI-6007-05 Revisian 5 1 Per Section 7.3, PBNP requested that total debris laden head losses be calculated br a range of temperatures. The total debris laden head loss is caIculated by adding the ARL test CSHL, the ARL test debris laden head loss, and the piping head lass calculated in the CSHL calculation [Reference 9.33.

This calculation has already provided these values at a design temperature of 212 OF. To calculate the head bsses for the PBNP specified range d temperatures, PC1 will use the following rnethodobgy:

I Attachment 'f

.. Point Beach Debris Laden Head Lass Temperature Corrected Values..

I A.

The CSHL value is calculated f,Refemilce 9.31 using the PC1 derived regression equation.

The equation uses kinematic viscosity to address the various temperaQres. The CSHL value used in this calculatian is based on the results of the CSHL testing performed at the ARL test facility. Temperature adjustment of the range of PBNP requested temperatures will be performed utilizing the' kinematic vismsSfy, as addressed in the PC1 regression equation. The CSHL value calculated at 212 O F will be adjusted for the range of tempemure values utilizing kinematic viscosity as follows:

1 Equatian I H l c s, ~ ~ = N b (VAT I v DT)

Where-Hk=AeTesfed-Glean-Stminer-Head-l;ossift VAT

=Kinematic viscosity at the post-LOCA adjusted temperature (i.e.; 32 "F to 190 O F )

I vm-. --.= Kinematic. viscosity. at the post-LOCA design.

temperature (212 O F )

I H k I T A temperature adjusted debris head !.ass, ft

, 1 1

j Kinematic viscosity values were calculated using the following equation; I

Equation.2 v = p l p Where 1.1

= djlnamic viscosity (IbFft-s) p

= density (lWftS) v

= kinematic viscosity [f IS)

The dynamic viswsity and d&i values were'taken from [Reference 9.9] for I

value above 32 OF. Dynamic viscosity and density values for 32 O F were taken I Originated By:

Tt3(b00?46 Rev S Tqt~l Haad Ltms&c

Total Head Loss - Paint Beach MuclGr Plant - Unit % lk 2 Technical Document No. TDf-6q07-06 Revision 5 1 Equation 3 HLpI, = f UD !f212g '

1

1.

from [Reference 9.131. The values of kinematic viscosity, dynamic viscosity and The specified temperature range requires determining friction factors for the temperature range. A Reynolds number is required to be calculated for the flow conditions in order to determine the friction factor used in the piping head loss equation. The friction factor can be taken from tabks on page A-25 in Crane weferencib 9.a afer calculating the Reynolds number. Reynolds numbers are calculated using the following equation; density are listed in Table A-4.

-..-.. -. ".~

B.

The piping head loss is wfcuiated using the following equation:

Equation 4 Re = V D I v I

I The piping head loss is oalculated in the CSHL calculation [Refcbrence 9.31 using values for 16 inch piping having a fluid velocity, V = 3.683 ftls and a piping diaMeter, D = 1.302 ff. Values for kinematic viscogty, v for the temperature range were taken from [Refemnce 9.9 and 9.131 and are included In Tabla.AII.

Values for Reynolds number were calculated for the temperature range, and the friction factor was read from the Crane table and input into Tabls A-I.

HL

= f x canstant 0,3665

= 0.012 x constant Constant

= 30.542 I

From the HL equation above, the terms UD V2/2g are assumed to be constant for.thfs. temperature adjustment. From Table 7. in. sectian'7.3. of this.caleulatiem,.

the piping head loss total at 212 O F is 0.3665 fk The friction factor, f, for 212 "F is.

0.012. Knowing the head loss and the friction factor, the constant term can be calculated as:

This cqndant term, along with the friction factors from Crane will be used in the HLprpe equation to calculete the various piping head losses withi the temperature range specified in Table 8-3.

6. ' The debris laden. head loss also requires temperature adjustment. As stated in ".

section 7.3 of this calculation; debris head toss can be temperature adjusted using the fallowing rnb@odolcigy;.

Originated By:

TDE6W7-06 Rev 5 T&l Hand Los&doe Atladunentl,PegeZoF4 I

Total Head Loss -Point Beach Nuclear Plant - Unit 1 $: 2 Technical Document No. TD16M37-06 Revision 5 1 Where NLIK = Debris Loaded Head Loss, ft I

dynamic viscosity at the, past-LOCA design temperature

- FAT.. = dynarnic..vis~~sify.

at the post-LOW. adjusted.

temperature IHILnLpT~ = temperature adjusted debris head loss, ft I The temperature adjusted CSHL from A is added to the temperature adjusted piping head loss from B and the temperature adjusted debris Iadsn head lass, from C above, to calculate. the DLTHL-TC for PBNP-lI2 based on the specified postrLOCA Design Basis temperatures.

Table A-I provides a summary of the final DLTHL-TC values as well as the various reference design inpufs, Originated By:

TDlBD07.46 Rev 6 Tblef Head Los8,doo Atfyhment I, Page3 of4 '1

Total Head Loss - PointBeach Nudes!. Plant'- Unlt.l&.&

I Technical Document No. TaI-600796

Revlsion 5 f lDl.8007-08 Rev 5 Tolal Head hs.dao

ENCLOSURE 3 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-1911GL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS PERFORMANCE CONTRACTING, INC.

CALCULATION PCI-5344-S04, SEPTEMBER 25,2008 STRUCTURAL EVALUATION OF CONTAINMENT EMERGENCY SUMP STRAINERS (ABRIDGED)

POINT BEACH NUCLEAR PLANT, UNITS I

& 2 30 pages follow

Calculation Number: PCI-5344404 I

I SafeQRelated Yes No I

Calculation

Title:

Structural Evaluation of Containment Emergency Sump Strainers Form3.1-10 Rev. 1

~

6.

8 CIient:

PerEormance Contracting Inc.

Project Number: PCI-5344 Station: Point Beach Unit(s): 1 & 2 Project

Title:

Point Beach Strainer Qualification

REVIEWER'S C H E ~ I S T FOR DESIGN CALCULATIONS S ~ E T 1 of 2 STATION:

Point Beach NUCLEAR SAPETYRELAm. YES NO c]

I PROJECTNO:

PCI-5344 CLTENT: Performance contract in^ Inc.

C m m T I O N ~

~

l 3

Structural Evaluation of Containment Emergencv Sump Strainers CALC. NO:

PCI-5344S04 CALC.REV.NO:

0 I INDICATE Tl3BDESIG.N INPUT DOCUMENTS USED:

I I --TYPE OF DOCUMENT I DOCUMENT ID, REV AND/OR 1 YES I N/A I COMMENT I

3. Design information package

&om related equipment vendor I. General Design Basis

4. Electrical Discipline Input

6. Control Systems Discipline Input

7. StrucWDiscipline Input DATE 3,5,9,13,22,23,27,31,46 I 8.

Specifications X

9.

Vendor Drwvings I

10. Design Standards I 1 1. Client Standards 1 12: Checked Calculations

13. Other (spec@)

PREPARER'S SIGNA DATE:

09/25/2008 REVIEWER'S SIGNATURR DATE:

09/25/2008 APPROVERfS SIGNATURE:

DATE:

09/25/2008

?

I I

Form 3.1-4 Rev. 3

2..

Have the applicable codes, ataudards and regulatory requiremenfs A. Properly IdeMied?

E. Properly Applied?

3. Were the input6 co~ectly se1ef;tedaud used?

4. A. Was Designhput Log used?

B. E4A is No, provide Manager's sisignakne in Comment w1w.u~ to si@

approval of Design Input Documents used in the c a l c W n.

5. Are necessary a~mq~p&ns adequately rrtated?

6. Are the assumptions reasonable?

7. Was the catoulaton methodology a p p r o p ~ ?

8. Are sym%ols and abbreviations adequately identified?

9. Are the calculatiom:

D. Presented in logical order7

10. Is the output. leasonable comparedto the inputs?'

11. Ifa computer program was used:

A. k the program listed on the ASL andhas the SRN been reviewed for any programuse limiWons?

B. Have existing user notices andlor error reports for the prodnotion version been reviewed as appropriate?

C. Were codes properly verified?

E. Were they colrectly used:

F. was dafainput.coneCt7 Form 3.1-4 Rev. 3

Form 3.1.3 Rev 2 TABLE OF CONTENTS I

. 0 PurposelObjective 5

2.0 Methodology.................................................................................................................................................. 5 3.0 Acceptance Criteria............................................................ ;

......................................................................... 9 4.0 Assumptions................................................................................................................................................ 14 5.0 Definitions and Design Input...................................................................................................................... 15 5.I Material Properties.........................................................................................................................

15 5.2 Strainer Geometry and Dimensions.............................................................................................. 17 6.0 Calculations 24 6.1 Weight Calculations 24 6.2 Strainer Loads 32 6.3 Calculation of Acceleration Drag Volumes and Hydrodynamic Mass 34 6.4 Calculation of Mass Distribufion on Strainer Components............................................................ 55 6.5 GTSTRUDL Model 61 6.6 GTST-RUDL Results...................................................................................................................... 93 6.7 Disk Pressure Loads.......;.............................................................................................................. 98 6.8 Core Tube Evaluation 103

. 6.9 Perforated Plate Evaluation 108 6.10 Wire Stiffener EvaluMon 123 6.1 1 Core Tube End Cover Assembly Evaluation.............................................................................. 124 6.12 Weld Evaluations 133 6.1 3 Rivet Evaluatioons 140 6.24 Mounting Evaluation.................................................................................................................... 145 6.1 5 Module-to-Modulb Sleeve and Latch Connection 169 6.16 Lifting Load Case Evaluation 175 6.17 Outage Load Case Evaluation 176 7.0 Results and Conclusions 177

. 8.0 References................................................................................................................................................. 179 Attachme?! - GTSTRUDL Run (Run time Mon JUI 10 13:48:48 2006) 1 -279 Attachment B - ANSYS Run for Gap Disk...................................................................................................... I - 1202

. Attachment C -Testing of 3/16 Blind Rivets (Reference [la])

................. 1 -13 Attachme~t. D - Piping Strike and Latch Test (Reference 1281) 1 - 5 Attachment E - Jay-Cee Sales Rivet Data (Reference [29])................................................................................ 1 - 4 Attachment F - Journal of Ship Research Paper (Reference 1331)

............................................................. I

-10 Attachment G -Journal of Engineering Mechanics Paper (Reference [34])

I -14 Attachment H - Perf Plate Thickness Data from Hendricks Book (Reference f351).........................................

I - I Attachment J - ACI Structural Journal (Reference [44])...................................................................................

1 -22 Attachment K-Lehigh Testing Laboratories Test No. G427(Reference 1481) 1 - 4 Attachment L - Lehigh Testing Laboratories Test No. F-19-32 (Reference [till).............................................. 1-3

CALCULATION SHEET 1.0 PURPOSEIOBJECTIVE The purpose of this calculation is to qualify the Performance Contracting Inc. (PCI) Suction Strainers to be installed in Florida Power and Light's Point Beach Nuclear Plant, Units I and 2. This calculation evaluates, by analysis, the strainer modules as well as the supporking structures associated with the new strainers.

2.0 METHODOLOGY The evaluations are performed using a combination of'manual calculations and finite element analyses using the GTSTRUDL Computer Program, (Reference [21]), and the ANSYS Computer Program (Reference [25]). The evaluations follow the requirements of the Strainer Design Specification PB-681 (Reference [I]). Exceptions from these requlrement, when taken, are discussed and justified within this calculation.

Seismic Loads The strainer is categorized as Seismic Class I equipment and is required to be operable during and after a safe shutdown earthquake (SSE) without exceeding normal allowable stresses as specified in Section 5.4.7 of DO-C03 Seismic Design Criteria Guideline (Reference [I 51). Strainer Design Specification PB-681 (Reference [I]), requires the strainer to be evaluated for two operating conditions. The first condition is a "dry" condition with no recirculation water inside or external water present. The second condition is a submerged "wet" condition with recirculation water. For the seismic evaluation the strainer will be considered submerged and full of water. The water level is considered to be a minimum of 10' above the 8'

.floor elevation (El. 41'- 2") per Reference 1461. The piping "dry" state with its associated mass being much less, will not be considered as it is less severe than the wet" state.

Per the specificafion, the seismic evaluation is required to-take into account any seismic slosh (analyzed at the seismic worst-case water level) of the recirculation water. Based on Reference 181, because of the negligible load magnitudes, it is determined that the seismic slosh loads in PWR containments are insignificant by comparison with other seismic loads. Therefore, seismic slosh loads are neglected from the analysis (refer to Section 6.2.3 for further explanation). Note that the sloshing calculation of Reference [8]

is done for the Prairie Island strainer project and it is representative for all PWR containments in general, and therefore, it is applicable for use in this calculation. The Wet" strainer operating condition considers the strainer assemblies submerged in still water at the seismic worst-case water level when subjected to seismic inertial loads. The inertial effects of the added hydrodynamic mass due to the submergence of the strainer is considered.

The strainer is seismically qualified using the response spectra method. The applicable seismic spectra are provided in Seismic Qualification Specification Sheet SQ-002243 (Reference [2]). These loads are applied to the strainer through base motion response spectra as detailed in the Seismic Design Criteria Guideline DO-C03 (Reference [I 51).

C ~ C.

NO.: PCI-5344-SO4 The strainer is located on the 8' floor elevation of the containment. The response spectrum chosen is for the 6.5' elevation of the containment. The containment liner plate is located at the 6.5' elevation and there is an additional 1.5' of concrete on top of the liner plate. The slab between the 6.5' elevation and the 8' elevation is very rigid. Thus it is appropriate to use the response spectrum for the 6.5' elevation. The vertical direction response spectrum is 2/3 the value of the maximum ground horizontal response spectra.

The strainer is excited in each of the three mutually perpendicular directions, two horizontal and one vertical.

Per Reference [I I], the modal combination is performed by the use of the double sum method to account for the effects of modal coupling in the response 0.e. closely spaced modes). An earthquake durafion of 30.24 seconds was used in the analysis per DG-COB, Appendix C. Appendix N of the ASME code indicates that the

, maximum accelerations generally occur in the first 10 seconds. Two analysis were run - one with 10 sec and one with 30.24 sec. Since the results were the same, the analysis with 30.24 seconds is the official documented seismic analysis. Responses from the vertical and one horizontal direction (worst case direction) are applied simultaneously and combined by absolute summation (Reference [151, paragraph 5.4.4.b). The cutoff frequency is taken at 30 Hz or a minimum of 5 modes are included. Zero Period Acceleration (ZPA) residual mass effectswill be considered. The ZPA response will be added to the response spectra loads by SRSS.

The strainer is considered a s a "bolted steel frames" sfructure and the damping values for seismic loads are taken a s 2% for the Operating Basis Earthquake '(OBE) and 5% for the Safe Shutdown Earthquake (SSE) a s required by Seismic Design Guide DG-CO3 (Reference [I 51):

Operating Loads

-Operating loads are comprised of weight and pressure loads. The weight of the strainer includes the weight of the strainer self weight and the weight of the debris, which accumulates on the strainer. The debris weight is taken from Reference [271.

Ttie pressure load acting on the strainer is the differential pressure across the strainer'perforated plates in the operating condition. Conservatively, this is taken a s the hydrostatic pressure associated with the maximum allowed head loss through the debris covered strainers. This is defined a s a minimum of 10 feet.

ofwater in DIT-008 (Reference 1461).

here are no thermal expansion loads since the strainers are basically free to expand without restraint.

Note that the piping is not rigidly attached to the strainer modules, therefore the piping is also free to expand without imposing any thermal loads on the strainers; The design temperature is'taken equal to the maximuin operational inlet temperature to the RH Exchangers of 250 O F (Reference [?I).

1 i i Ponn 3.1-3 Rev 2

MathCad software is used to generate the calculations. All MathCad cabulatians are independently verified for accuracy and correctness a s if they were manually generated. ANSYS is used for the analysis of the inner gap plate. ANSYS Version 5.7.1 is fully verified with no restrictions or limitations. OTSTRUDL Version 25 is used in the seismic response spectra analysis of the strainer modules. GTSTRUDL Version 25.0 is verified and validated under the AES QA program as documents in the AES validation and maintenance file (Reference 1211). The validation of GTSTRUDL was a partial validation and only validated certain commands. These commands are listed in the validation report. The GTSTRUDL runs utilized several commands outside the scope of this validation. A list of these commands, and their alternate validation method used for fhis particular application, is provided below:

Command Validation Method GENERATE The GENERATE and REPEAT commands are used to automatically generate REPEAT member nodes and incidences. These generated items for these models are verified manually.

Command Validation Method JOINT TIES The JOJNT TIES and SLAVE RELEASES commands are used in conjunction with SLAVE RELEASES MEMBER TEMPERATURE LOADS to account for the preload on the connecting rods. The commands also constrain the pipe spacers and connecting rods to move together in certain degrees of freedom..Their use is acceptable because the nodal displacements are manually compared for these nodes to confirm the command is working a s planned.

MEMBER This command applies a specified temperature increaseldecrease to a given TEMPERATURE member. This command is used a s a simple way.to generate preload in the rods.

LOADS Its use is acceptable because the preloads produced by this load are verified manually.

DEFINE GROUP This command groups members andtor joints together for easier specification of member properties and load placements. This command is verified by checking.

manually that the cross sections and loads are applied properly to each member.

MEMBER ADDED This command was used to apply the water weight of the system directly on to INERTIA members that would carry that water foia certain direction of motion. This' coinmand was verified manually by listing the dynamic mass summary and comparing the total dynamic mass in each direction to the calculated masses.

C Form 3.1-3 Rev 2

PIPE PIPE is a command used to specify the cross section of the core tube. It is necessary to use this command rather than referencing a pipe cross section from a table because the diameter and thickness are unique to the strainer and are not available in the provided tables. Because GTSTRUDL uses only the sedon properties when code checking, the properties are printed out for selected members defined by this command and those properties are verified manually.

TABLE 'RBARS'

'RBARS', 'BARS', 'ROUND', and 'MYCWAN' are predefined GTSTRUDL tables that TABCE 'BARS' contain steel cross sections for rectangular, round (for both 'BARS' and 'ROUND1),

TABLE 'ROUND' and channel shapes. The members that are defined by these tables are subjected TABLE 'MYCHAN' to loadings and then code checked in GTSTRUDL. These tables are veiifled in the same fashion a s for the PlPE command listed above. In addition any code checks performed by GTSTRUDL for these sections are manually verified.

The limitations and program error reports for GTSTRUDL Version 25 (Reference [21]) were reviewed for applicability to the GTSTRUDL runs made for this calculation. The limitations for the ASD9 Code check were found not to be applicable for this calculation (none of the components are subjected to significant torsion, therefore warping torsion stresses would be negligible). Also, steel cross sections that were not available in the GTSTRUDL cross section.iibraries had to be created for the face disk edge channels, the external radial stiffeners, the debris stops, the seismic stiffeners, the ends of the connecting rods to account for the threading, and the ends of the external radial stiffeners where they are welded to the seismic stiffeners. These cross sections were verified by outpuffing the computed properties ofthe cross sections and checking these values manually. All known issues, including Part 21 notifications, have been reviewed for applicability in accordance with the AES QA program. Work arounds to existing issues or errors have been utilized a s required. '

. Form3.1-3 Rev 2

3.0 ACCEPTANCE CRITERIA The strainer components shall meet the requirements of the strainer design specification P8-681 (Reference

[I]). As stated in PB-681, the detailed evaluations are to be performed using the rules, as applicable, of ANSIIASME 831.I Power Piping 1998 Edition through 1999 Addenda (Reference 151).

The stralners are classified as "other pressure-retaining components" as described in Paragraph 104.7 of the B31.1 Code (Reference 151). Under Paragraph 104.7.2, the code allows 'The pressure design of components not covered by the standards listed in Table 126.1 or for which design formulas and procedures are not given in this Code shall be based on calculation consistent with the design criteria of this Code. These calculations shall be substantiated by one or more of the means stated in (A), (B), (C), and (D) below.

Based on this paragraph, since the B31.I Code does not provide specific design rules for a pressure retaining component such as a strainer, design guidance will be taken from the ASME Boiler and Pressure Vessel Code (Reference [3]).

The ASME Code is consistent with the B31.1 Code and is a logical alternative to B31.I rules. The substantiation method described in Paragraph 104.7 of the 831.1 Code is Alternative Dl which allows for "detailed stress analysis, such as the finite element method, in accordance with the ASME Boiler and Pressure Vessel Code, Division 2, Appendix 4, except that the basic material allowable stressfrom the Allowable Stress Tables of Appendix A shall be used in place of S,."

Section Ill, Subsection NC of the

$ME Code will be used as this presents the most general criteria for the design of pressure retaining components.

The use of the ASME Code is primarily for the qualification of pressure retaining parts of the strainer which are not covered in 531.I (perforated plate, and internal wire stiffeners). Some parts of the strainers (radial stiffeners, connecting rods, edge channels, seismic stiffeners, etc.) are classified as part of the support structure. These types of components are covered under the AlSC Code (Reference 191). Additional guidance is also taken from other codes and standards where the AISC does not provide specific rules for certain aspects of the design.' For instance, the strainers are made from stainless steel materials. The AlSC Code does not specifically cover stainless steel materials. Therefo~e, ANSflAlSC N690-1994, "Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities", Reference [30] is used to supplement the AlSC in any areas related specifically to the structural qualification of stainless steel. Note that only the allowable stresses are used from this Code and load.

+

combinations and allowable stress factors for higher service level loads are not used.

.. The straineralso has several components made from thin gage sheet steel, and cold formed stainless sheet steel. Therefore, SEllASCE 8-02, "Specification for the Design' of Cold-Formed Stainless Steel Structural.

Members", (Reference [31]) is used for certain components where rules specific to thin, gage and cold form '

stainless steel should be applicable. The rules for Allowable Stress Design (ASD) as specified in Appendix D of this code are used. This is further supplemented by the AlSI Code (Reference P2]) where the ASCE Code is lacking specific guidance. Finally guidance is al'so taken from AWS D1.6, "Structural Welding Code -Stainless Steel", (Reference 1231) as it relates to the qualification of stainless steel welds. Detailed acceptance criteria for each type of strainer component is provided in the sections below.

Form 3-13 Rev 2

The core tube is evaluated as piping per 831.I Paragraph 104.8 as applicable. The effects of the core tube holes on the pipe stresses are incorporated using Stress Intensification Factors (SIF) for the localized effects and effective net cross section properties for global effects.

% For the perforated plates, the B31.1 Code does not provide any design guidelines as discussed above.

Therefore, the equations from Appendix A, Article A-8000 of the.ASME B&PV Code, Section Ill, 1998 Edition (Reference 131) is used to calculate We perforated plate stresses. Note that Article A-8000 refers to Subsection NB for allowable stresses, which are defined In terms of stress intensity limits, S,.

However, in keeping with the 631.1 maximum principal stress design philosophy, principal stresses are calculated and compared to the allowables based on the ASME allowable stress limit, S, taken from ASME Section 11, Part D (Reference 141). Specific limits for each component are described in further detail below.

The edge channel and the attached perforated plate work as a combined section to resist bending loads.

The effective width of the perforated plate that acts in combination with the edge channel is based on Section 6.2 of the ASCE Code (Reference [31]), which provides design guidelines for very thin stainless steel

. members such as the perforated plate. The effective width of the plate is limited by the width to thickness ratios such that local bucklhg of the plate will not occur for the compression face. The minimum spacing and edge distance required for the rivets is based on the AlSl (Reference 2221) requirements for screw spacing.

The seismic siiffeners, extkrna~ radial stiffeners and the mounting hardware are evaluated to AlSC 9th Edition (Reference [9]) as permitted in paragraph 120.2.4 of the 831.1 Code (Reference [Sj). The analysis of the anchorage to the containment concrete slab will be in accordance with the Hiki technical ~ u i d e (Reference

. tl.01).

Load Combinations The applicable load combinations for the strainers are those for Section 6.7.1 of DG-MI0 (Reference 1141) and 6.0 of DG-MO9 (Reference [I I]).

Load Condition Combination

( I a) Normal Operating DP + DW (I

b) Normal Operating (OutageRift Load)

DW+ LL (2) Upset DP+DW+WD+OBE (3) EmergencylFaulted DP+DW+WD+SSE

CALCULATION SHEET DW = Dead Weight Load LL = Live Load (additional loads on strainers during outages or during installation, live load is not applicable during operation)

WD = Weight of Debris DP = Differential Pressure OBE= Operating Basis Earthquake SSE = Safe Shutdown Earthquake Note that combination (3) is classified as Emergency Condition for all ASME Code evaluations and Faulted for

.all components governed by AISC and ACI Codes. Also note that wind, snow, tornado', and jet force loads are not applicable. Flood loads are considered for Load Combinations 2 and 3. Fiood loads consist of the effects due to earthquake in a submerged condition (sloshing and added mass). There is no hydrostatic pressure loads associated with flooding since 'the flood waters are present on all sides. Thermal expansion stresses are considered negligible a s described in Section 2.0.

Core tube The Core tube is evaluated as plping per B31.1 Paragraph 104.8 as applicable. Since the 831.1 does not explicitly identify how to Incorporate the Emergency SSE loads, PBNP uses ASME Section Ill as a guide a s discussed in Section 6.0 of DG-MOB (Reference [I I]).

831.I Ea. No Load Condition

. Load Combination Allowable Stress I1 Normal DW 1.0 S, 12 (OBE)

Upset DW + OBE 1.2 S, 12 (SSE)

Emergency DW + SSE 1.8 S, Strainer Pressure Retaining Plates For the pressure retaining plates, such a s fhe perforated platithe B31.j Code does not provide any design guidelines a s discussed above. For the perforated plate, the equations from Appendix A, Article A-8000'of the ASME B&PV Code, Section 111,1998 Ediion through 1999 Addenda (Reference PI) is used to calculate the stresses. Note that Article A-8000 refers to Subsection NB for allowable stresses, which are defined in terms of stress intensity limits, S,.

However, jn keeping with the B31.1 maximum principal stress design philosophy, principal stresses are calculated and compared to the allowabies based on the ASME allowable stress limit, S.

CALCULATION SHEET Stress limits for the pressure retaining plates are taken from NC-3321 (Reference [3])

Load Condition Stress Tvpe Allowable Stress Desian Level NomaIlUpset*

Primary ~embrane Stress

.l.O Sh

'Level A Primary Membrane (or Local).t Bending 1.5 Sh Emergency.

Primary Membrane Stress 1.5 Sh.

Level C Primary Membrane (or Local) + Bending 4.8 Sh

• Allowable stresses for Upset condition may be increased by 10% as permitted by NC-3321 (Reference 131)

Strainer Structural Components Based on the discussion provided earlier in this section, the allowable stresses on the strainer structural

,components is based on the AISC 9th Edition (Reference [9]). The allowable stress for the SSE Load Combinations is taken from Sectton 6.9 of DG M I 0 (Reference [I 43).

Load Condition Load Combination Allowable Stress Normal Operating la, I b 1.0 AlSC Upset 2

1.0 AISC Faulted 3

1.5 AlSC but not to exceed 0.9 SY Additional details for the various types of support components are provided below.

Compression Per Reference [30], because stainless steel does not display a single, well defined modulus of elasticity, me allowable compression stress equations from the AlSC are not applicable for stainless steels. Therefore, the allowable compression stress will be based on the lower allowables from Reference [30] as opposed to those provided in the AISC Code (Reference [9]). Per Q1.5.9.2 of ~eference [30], the allowable stresses for tension, shear, bending and bearing for stainless steel can be taken as the same allowables provided for carbon steel, therefore the AlSC 9th Edition will be used for allowables for these types of stresses.

I Porn 3.17'3 Rev 2

GTSTRUDL Code Check Most support components are qualified using the GTSTRUDL code check features. The use of the 9th Edition Code check feature of GTSTRUDL is acceptable for this application with the exception of the allowable compression stress a s described above. The effective buckling length factor, K, will be manually adjusted to account for the lower compression stress allowable. See Section 6.5.8 for additional discussion.

Edge Channels The edge channel and the attached perforated plate work a s a combined section to resist bending loads.

The effective width of the perforated plate that acts in combination with the edge channel is based on Section 2.3 of the ASCE Standard for Cold-Formed Stainless Steel Structural Members (Reference [31]), which provides design guidelines for very thin members such a s the perforated plate. The effective width of the plate is limited by the width to thickness ratios such that local buckling of the plate will not occur for the compression face. The minimum spacing and edge distance required for the rivets is based on the AlSl (Reference 1221) requirements for screw spacing..

Welds There are no provisions given in the 831.I Code for the strainer structural welds to the piping components (radial stiffener to core tube). Therefore, these welds are evaluated in accordance with paragraph NC-3356(c) of the ASME B&PV Code, Section Ill (Reference [3]). Welds for strainer support components, such a s for the seismic stiffeners to radial stiffeners, end cover connecting tabs, and those for the floor track support system, are qualified per the AISC 9th Edition (Reference [9]91)-: AVVS D1.6 (Reference [23]) was reviewed to ensure that any special qualification requirements associated with stainless steel welding were considered. Since the weld allowables provided in AWS Dl.6 are essentially fhe same a s allowed for carbon steel welds under AWS Dl.I (Reference [I 3]), no special adjustments are required to account for stainless steel.

Rivets There are three areas in the strainer module where rivets are used a s Fasteners. The disk'faces are riveted to the perforated edge channels. The gap disk is fashioned into a ring using two rivets. The sleeve that connects adjacent module core tubes together is held in place by two latches that uses four rivets each to attach to the thin gauge steel. The rivets' capacities are based on testing. From Reference [18], the capacities of the rivets are taken a s the average value from six tests (six tests for shear and six tests for tension). A factor of safety is then calculated according to the ASCE Standard (Reference [31]) a s supplemented by the AlSl Code (Reference 1221) accounting for the capacities being found experimentally via a small sample group (n = 6).

This factor of safety (FS = 2.50 p.er Section 6.13 of this calcula~on) will be used on these ultimate capacities for OBE. An increase of 1.5 is allowed for SSE, resulting in a FS11.5 for SSE.

Mounting Hardware Hilti Kwik-Bolt Ills will be used to mount the strainers to the floor. The analysis and design of expansion anchors shall be In accordance with the Hilti Technical Guide (Reference [I PI) however a Factor of Safety of 4 against ultimate will be used. Qualifications of the boltslpins used to attach the strainers to the track will be based on the ASCE Standard (~eference [311). Neither of the AISC Codes (References [B] & [30]),

provide specific bolting allowables for stainless steel bolting.

Form 3.1-3 Rev 2.

CALCULATTON SIBOEET 4.0 ASSUMPTIONS This calculation evaluates thti.Unit ?. strainers incfuding the addgonal modules and new end b e r s associated wlth Unit 1 fo be added under EC 12601 and EC 12603. It 1s also applicable for the Unit 2 strainers, including changes to be installed at a later date, provided the following assumptjons hold true:

The end cover assembly and strainers are identical to Unit I New 518" expansion anchors at 4-112" embedment maintain a minimum of 6" anchor-to-anchor spacing for an interior anchor and 3 anchor-to-anchor spacing for anchors at the end of individual fracks (coupled with a mirt. 8-112 edge distance)

New 518" expansion anchors at 4-112 embedment maintain a minimum of 5" edge distance to expansion Joints in the concrete floor (coupled with a min. 8-912" anchor-to-anchor spacing)

~ ~, ~ ~ p + t a - t &

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8 Rev 2 i

Form 3.I-3

CALCULATION SHEET Date: 09/25/2008 5.0 DEFINITIONS AND DESIGN INPUT 3

Define, ksi 3 10.psi kips = 103.1bf kPa := 1000.Pa ORIGIN E 1 5.1 Material Properties Material Types per Reference [6bJ:

Perforated Plate:

Stainless Steel ASTM A-240, Type 304 Core Tube:

Stainless Steel ASTM A-240, Type 304 Radial Stiffeners:

Stainless Steel ASTM A-240, Type 304 Wire Stiffeners:

Stainless Steel ASTM A-493, Type 304 (Drafted to 1 I 0 ksi - 130 ksi)

Rivets:

Stainless Steel ASTM A-240, Type 304 Connecfing Rods:

Stainless Steel ASTM A-276, Type 304 Nuts:

Stainless Steel ASTM A-194, Grade 8 Washers:

Stainless Steel ASTM A-240, Type 304 Spacer Sleeves:

Stainless Steel ASTM A-312, Type 304 Seismic Stiffeners:

Stainless Steel ASTM A-240, Type 304 Angle Iron:

Stainless Steel ASTM A-276, Type 304 Mounting pins:

Stainless Steel ASTM A-276, Type 304 Hitch Pins:

Stainless Steel ASTM A-580, Type 304 End Cover Assembly Stainless Steel ASTM A-240, Type 304 Latch and Strike Plate:

Stainless Steel ASTM A-240, A-580, A-31 3, Type 304 Latch Rivets:

Stainless Steel ASTM A-4931A-313;Type 304 Design Temperature TdeS= 2500 F ( Reference [I]

)

Modulus of Elasticity at 250° F (Reference [4]),

E, := 27300. ksi Yield strength at 250OF (Reference 1411, Ultimate Strength at 2500 F (Reference [4j),

S, := 68.6. ksi 831.I Allowable Stress at 2500 F (Reference [5]),

Sh

= 17.2.ksi Note these propetties are conservative for the Type 304 wire stiffeners which are draffed to a higher tensile strength than standard Type 304 stainless steels

' Wire Material The ASTM Standard (Ref. [47]) does not report a yield strength for this material as the typical application of wire Is tension only. Therefore, a test was performed (Ref. 1481) to determine the yield strength of the wires (both radial and circumferential). The reported values for the yield strength are 89-1 12ksi. However, due to the low number of tests performed, a conservative value of 65ksi is used for the yield strength of the wire material at elevated temperatures (250.F).

Yield Strength at 2500 F (Ref [48])

SF,,,

= 6Sksi Other Miscellaneous Properties Ibf pdel := 501.-

Density of stainless steel from Reference [20],

ft3 Density of carbon steel from Reference 1201, Ibf Pc.steel:= 490.7 ft Poisson's Ratio from Reference [20],

v := 0.305 Ibf Density of water at temperature of 68OF(Ref. [12])

Y~20.1 := 62.4.-

it3 Density of water at temperature of 250°F(Ref. 1381)

Ibf YH20.2 := 58.8'-

ft3 CoIMcient of Thermal Expansion (CTE) of stainless steel, CTE := 9.1.10- 6 (going from 70°F to 250°F (Ref. [4])

• .Hydrodynamic mass Is based on the density of wafer at temperature. Slnce the yield sfrengfb of stainless steel decreases with temperature faster than the density of water decreases, it is acceptable to use the lower density of water as long as the material yield strengths are also reduced for temperature.

1 Form 3.1-3 Rev 2

CAL-ATION SHEET All data are per Ref. f641 unless ofherwise noted.

Perforated Plate Dimensions Thickness of 18 gage perforated plate a s per Reference 1351 tperf := 0.048.in Hole diameter of perforated disk plate, Ddiskholes := 0.066.in Ref. L6gI Pitch distance between perforation h ~ l e s in disk plate (Center-to-center distance)

Pdiskholes := O.1zS.in Ref. I6gI Disk Dimensions Strainer disk size Lldisk:= 33.0.in L2dlsk := 36.O.h Number of disks per strainer module Ndlsk:= 10.

Strainer disk edge channel dimensions dchan := 0.5.in Ref. I6gI bchan':= 0.5.in Ref. 16gI Width of each middle disk assembly Wdisk := dchsn + 2'tperf Wdisk = 0.596 in legs

.Width of,gap spacing between consecutive disks.

Wi,, := 1.0.in Figure 6.2 Side view of Strainer Module I

i Form 3.15.

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CALCUL,ATION SHEET External Radial Stiffener and Seismic Stiffener Dimensions The disks are supported by radial stiffeners which are welded to the core tube.

Thickness of external radial external stiffeners and debris stops tsar := 0.375.in Ref. [6fl Width of external radial sfiffeners Width of debrjs stop wdastop := 0.84375.in Outer diameter of the debris stop ODdebris := 17.565.in Width of top and bottom external radial stiffener ends Length of top stiffener ends hand

= 2.5.in Length of bottom stiffener ends LBaend

= 4.5. in Length of the support legs Width of support legs and seismic stiffeners w,,,

= 1.5-in Thickness of support legs and seismic stiffeners Seismic stiffener to radial stiffener weld thickness tw.ob := 0.1875.in Seismic stiffener to radial stiffener weld length (on either side of tab) ww.,b := 1 -in Form 3.1-3

". Rev 2.,

CALCULATION SHEET Connectina Rod Dimensions Number of connecting rods Nmd := 8 Connecting rod diameter ODrnd := 0.5-in Ref; [6fl 0.9743.in Connecting Rod tensile diameter ODtens :=I ODrnd -

ODte,,

= 0.425 in Ref. 191 13 Outside diameter of spacers (112" ID, SCH 80)

ODspacer := 0.84.in Ref. I91 Thickness of spacers (112" ID, SCH 80) t,,,,,,,

= 0.147-in Ref. [91 Eccentricity between edge of disk and outer connecting rod emd := 0.9375-in Connecting rod tightening torque Tmd := 20.ft.lbf diameter of centerline of inner tension rods BCrod := 17.254-in Core Tube Dimensions Outer diameter of perforated core tube ODhbe := 15.815.in CorrosionlFabrication Allowance t,

= 0.0-in Core tube wall thickness (16 ga.)

tlesa := 0.0595.in Ref. 1351 Core tube wall thickness after allowance t t ~ b ~

= t16ga tca

$be = 0.0595 in Ref. @fl Core tube extension beyond last disk face Lsfub := 2.2S.in Outer diameter of disk gap ODgap := 18.19.in Number of rows of core tube holes Nhole := 5 Ref. [Be]

Number of holes per row Nhole.clrc :=

Ref..[6e]

Radial stiffener to core tube weld thickness tWd := 0.0625-in Radial stiffener to core tube weld length (per individual weld) w w. ~

= 1.5. in

..:.[::)

The orientation of the hole alongathe circumference

.deg

. Ref.[6e]

270 Rivet Dimensions Number of edge channel rivets per disk side (excluding comer rivets)

Nldvet:= 10 N2,ivet :,= 11 End cover, facelgap disk rivet head diameter cdisk.-

= 0.375.in Ref. 16fl (item #Is PR64FFP and PR62FFP, respecfively. See Ref. 1291)

Ref. [6h]

Sleeve Rivet diameter (118" Stainless Steel Rivets) cslv.flvet := 0.125.in

... Rev 2

CALCULATION SHEET Number of intermediate disk face rivets Number of inner gap rivets holding the hoop together Eccentricity between the edge channel rivets and the adjacent edge of disk qmt

=

Offset from line connecting center of core tube and center of outer rod eo~:= 1.25.in (Refer to subsection Internal Wire Sfiieneffi in Section 6.1 for more detail)

Internal Wire Stiffener Dimensions (All data per Ref. 1601 unless othewise noted)

Number of intermediate circumferential stiffeners Diameter of radial wire stiffeners (7 ga) dwireSrad := 0.177.in Ref. 16bI Diameter of circumferential wire spacers (8 ga) dwireecirc := 0.162.in Ref. [6bl Inner circumferential stiffener width

LClGi,

= ODbbb, -I-1.5.in hl,,i,

= 17.32 in Outer circumferential stiiener width (Side 1)

L l ciz,o,t:=

L1 disk - 2.emd L l circ.cut = 31,125 in Outer circumferential stiffener width (Side 2)

LZci,,,t:=

&disk - Peerod L2circ.out = 34.125 in Corner distance for outer circumferential Lcl,.cor

= 1.5.in End Cover Assemblv Dimensions (Dimensions per Ref. r6v1)

Thickness of end cover tbackpl := 0.5.in Diameter of back plate ODback.pl := 19.3150in Diameter of sleeve ODsIeeve.ec := 15.815in Thickness of sleeve tsleeve.ec = 0-06 in Length of base plate Lbase,pl := I a n Thickness of base plate '

tbase.pl := 0.5in Length of tube steel support Lts,ec := 28.2095in Length of sleeve Ls1eeve.e~ := 1-5in Eccentricity between edge of base plate and anchor bolt

= 1.25in Height of stiffener J

hSw := 31n Thickness of stiffener

&tiff := 0.5in Size of tube steel support wbSeC

= 4in

CALCULATION SBCIEET End Cover Assemblv Dirnenslons (continuedl Minimum distance from the expansion anchor bolt postion parallel to the base plate edges. (Note that the holes can be drilled any where between the two positions shown in section 6.1 1)

Weld thickness for all base plate connections of end cover assembly Weld thickness of tube steel to the back plate tw.,,& := 0.25in Weld thickness of sleeve to the back plate tw.ec.sleeve := 0.0625in Tolerance for the offset for connection between the back plate and the tube steel in the horizontal direction doffset := 0.5in Anchor Bolts for end cover assembly ODhkbaec := 0.5in Other Miscellaneous Dimensions

~iameter of mounting pin connecting the strainer to the angle iron track ODpln := 0.5.in Ref. [6h]

Angle iron thickness tangle := 0.25.in Ref. [6g Length of vertical leg of the angle iron track Lvertleg := 2-in Ref. [6i]

Ecceniricity from bolt connkction to bottom of angle e b a : = 1.125.in Ref. [6i]

Eccentricity from corner of angle to anchor bolt ehkbT,:=l.5.in Eccentricity from edge of angle leg to anchor bolt ehkb.2 := 1.5-in Span between two adjacent anchor bolts Lhkb := 19.9567-in Ref. 16il Eccentricity between two adjacent module supports esp,.t := 6.5.in.

Ref. [6ij

. Length of alternate angie iron segment in case of rebar interference:

bit := 4.541 Ref. [6c]

Alternate angle iron segment to angle iron track weld length (full) wWwalt

= 2.in.

Ref. [6c]

Pom3.1-3.

Rev 2

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...:,!,,..: 22.: %....>... $....,,

~ ! : ; ~ ~ < ~ 5 ~,. ~ ~ t ; ~ ~ ; ~ ~ ~ ~ ~ ; ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ i ! : ~ ~ ~ : : j, < ~ ~ ~ < $ i ~ r + x ~ ; ~. ~ z, ~, ~ $ ~ ~ i ~ ~ ~ ~ f i 4 ~ ; ; 2 ~ ~ p ~ j

$,?:,Jc2;,,reF; 2 t+:k,:lz.,,; rT.+ 8 I.-.!,

I,; i-.,; i.,c 3$9~.~fl~ty;;;;f~2;;;~qit;;i;j;ii;

!,,:,,,-! ~ ~ - a + z ; ~. : ~

F ~.., ;:!.f

L>..*:,*..... -,.,a 4?r,:'bx?2t+j:,--..,.+

• .,fs

,,*.:I,........:,,+$ ;.:,

~ --

'I.

0 '

Other Mtscellaneous Dimensions (cont'd)

Alternate angle iron segment weld thickness tWaElt

= 0.187541

'Ref. [64 Diameter of hitch pin ODhltch := 0.177. in Ref. 16bl Diameter of Hilti Kwik anchor bolt ODhkb := 0.625.in Ref. t6cI Diameter of core tube connection sleeve ODsleeve := 15.8723 in Ref. [6h]

Thickness of sleeve connecting two adjacent modules (22 ga. See Ref. P5]) tslseve := 0.0293.in Ref. [6h]

W~dth of sleeve connecting two adjacent modules

wsle,

= 3.5.in Ref. [6hl Number of latches per sleeve Nlatch := 2 Ref. [6h]

Span between two module supports for a given module krt := 13,4567.in Ref. 16il Pool Boundaries (All data per Ref. r6al unless otherwise noted).

Minimum height of the water above the floor H,:=

384n Gap between the bottom of the strainer and the floor gf:= 3.in Gap between the top of the strainer and the minimum water level surface gt := 2.in Approximate distance from containment walllfloor interface to adjacent e,:= 641-1 Ref. [6j]

strainer train (Unit I controls)

Angle of the reactor containment wall a,!,

= atan -

(2) acvall

= 73.30 deg

, Ref. [6jl Minimum average gap between the side of

. g,

= Q + o.s.Ll disk + gf g,=

11.85 in Ref. 16jl the strainer and the nearest wall (Unit 1 tan(%311) and controls)

Ref. [6a]

Strainer Trains '

The holelslot distributions along the length the core tube are given in terns of dimensions H (the width of the slot or the diameter of the hole) and L2 the length of the slot. The length of the slot (L2) is orientated along the axis of the core tube. There are four holes around the circumference of each row. There are N number of rows. M is provided in array format and L;2 and LIia are provided as constants (see Reference [eel), where the rows are the hole locatlons, the first row being the smallest hole on the end module, and the last being the largest hole on the end module. The first column represents the holes associated with the 0 and 180 degree locations of the end the module, and the second column represents the holes associated with the 90 and 270 degree locations of the end module.

k := 1.. Nhole j := 1.. 2 0

90 180 270 0 90 deg 2.34 2.39 j

i i

2.34 2.39 j

H:= 2.34 2.39.in

) riq-1 [TI r ::: :;J I

I-- H -4 i

[T] [T]

L2 := 2.49-in

[TI Lllg := 0.5.in rhole := min -

,0.25-in (i 1 Figure 5.2 Partial View of Strainer Trains (Flgure is a partial view of complete layout, see Ref. [6e])

Note the holes at 0 degrees and 180 degrees are ihe same size, and the holes at 90 degrees and 270 degrees are also the same size (see "Sure-Flow Shiner Trains1' Reference [&I).

Safety Related yes NO L.l I Date: 09/25/2008 Corner Stiffener I

lntermediate Circumferential Radial Circumferential Stiffener I StifFener stiffener I

d. rivet Figure 6.1 Intermediate Wire Stiffener Pattern and Notation

CALCIILATTON SHEET The results of this calculation indicate that the strainers meet the acceptance criteria for all applicable loadings.

A summary of the maximum stress Interaction Ratios (calculated stress divided by allowable stress) Is provided Strainer Component External Radial Stiffener (Including Debris Stops)

Tension Rods Edge Channels (Rims Disks)

Core Tube (Biggest Moles)

Perforated Plate (Seismic Case)

Perforated Plate (Rim Disks)

Perforated Plate (Gap Disk)

Wire Grill Stiffener 6.10 IRd,=

0.69 End Cover Assembly Components End Cover Assembly Anchor Bolts End Cover Assembly Welds 6.12.q IR,,~ - (0.3 r 0.23 )

Weld of ~ i d i a l Stiffener to Core Tube 6.122 T

IRweld.ct = (0.30 0.47 )

Weld of Radial Stiffener to Seismic Stiffener 6.12.3 I

R

~

~

~

~

~

= (0.51 0.50 )

CALCULATION SELEET RESULTS AND CONCLUSIONS (Cont.1 Strainer Component Ref. Section Interaction Ratio Rim ~ l s k Blind Rivets Gap Disk Blind Rivets Clevis Witch Pins Angle Iron Mounting Tracks Expansion Anchors to Floor Angle Iron-fo-Angle [ron Track Weld Module-to-module Sleeve Module-to-module Sleeve Connection (ogtional Strap and Clip included)

CALCULATION SmET

8.0 REFERENCES

[I]

Point Beach Nuclear Plant Specification PB-681, "Replacement of Containment Sump Screens", Revision 2.

121 Point Beach Nuclear Plant Seismic Qualification Specification Sheet SQ #002243, Revision 0.

[3]

ASME B&PV Code, Section Ill, ~ivision 1, Subsections NB, NC, and Appendices, 1998 Edition, through 1999 Addenda.

[4]

ASME B&PV Code, Section 11, Part D, Material Properties, 1998 Edition, through 1999 Addenda.

[5]

ANSIIASME 831.I Power Piping Code, 1998 Edition, through 1999 ~ddenda.

[6]

Performance Contracting, Inc.(PCI), Sure-Flow Sucfion Strainer Drawings.

6a. PC1 Drawing No. SFS-PB2-GA-00, "Sure-Flow Strainer Recirc Sump System Layout", Revision 2.

6b. PC1 Drawing No. SFS-PB2-GA-OI, "Sure-Floy Strainer General Notes", Revision 7.

6c. PC! Drawing No. SFS-PB2-GA-02, "Sure-Flow Strainer A Strainer", Revision 9.

6d.. PC1 Drawing No. SFS-PB2-PA-7100, "Sure-Flow Strainer Module Assembly", Revlsion 1.

6e. PC1 ~ k w i n ~

No. SFS-PB2-PA-7101, "Sure-Flow Strainer rains", Revision 1.

6f.

PC1 Drawing No. SFS-PB~-PA-71 02, "Sure-Flow Strainer Module ~ssen;bl~",

Revision 3.

6g.. PC1 Drawing No. SFS-PB2-PA-7103, "Sure-Flow Strainer ~ections'and Details", Revision 0.

6h. PC1 Drawing No. SFS-PB2-PA-7105, "Sure-Flow Strainer SleeveslCoverlSupportsIPins", Revision 4.

61.

PC1 Drawing No. SFS-PB2-PA-7150, "Sure-Flow Strainer Mounting Track A1IB1", Revision 2.

6j.

PC1 Drawing No..SFS-PBI-GA-00, "sure-low Strainer Recirc Sump System", Revision 9.

6k.

PC1 Drawing No. SFS-PB2-PA-7106, "Sure-Flow Strainer End Cover", ~ehsion

1.

61.

PC1 Drawing No: SFS-PB1-GA-04, "Sure-Flow Strainer General Notes", Revision 12.

, 6m. PC1 Drawing Noi SFS-PB1-GA-02, "sure-low Strainer A Strainer", Revision 9..

' 6n. PC1 Drawing No. SFS-PB1-PA-7100, "~ure-Flow Strainer Module Assembly", Revision 4.

60. PC1 Drawing No. SFS-PB1-PA-7101, "Sure-Flow Strainer Trains", Revision 5.

6p. PC1 Drawing No. SFS-PB1-PA-7102, "Sure-Flow Strainer Module Assembly", Revision 3.

6q. PC1 Drawing No. SFS-PB1-PA-7103, "Sure-Flow Strainer Sections and Details", Revision 3.

6r. ' PC1 Drawing No. SFS-PBl-PA-7105, "Sure-Flow Strainer SleeveslCoverlSupport~lPins~~,

Revision 12.

Pom3.1-3

, Rev 2

CALCULATION SEIEET 6t.

PC1 Drawing No. SFS-PBl-PA-7153, "Sure-Flow Strainer Monting Track A3, B3", Revision 3.

6u. PC1 Drawing No. SFS-PBI-GA-03, "Sure-Flow Strainer B Strainer", Revision 9.

6v.

PC1 Drawing No. SFS-PB2-GA-03, "Sure-Flow Strainer B Strainer", Revision 9.

6w. PC1 Drawing No. SFS-PBI-PA-7151, "Sure-Flow Strainer Mounting Track A2, B2", Revision 2.

6x.

PC1 Drawing No. SFS-PB2-PA-7151, "Sure-Flow Strainer Mounting Track A2/B2", Revision 3.

6y.

PC1 Drawing No. SFS-PB1-PA-7152, "Sure-Flow Strainer Module End Cover Assembly", Revision 3.

62.

PC1 Drawing No. SFS-PBI-GA-07, "Sure-Flow Strainer Piping A Layout", Revision 2.

[7]

"Formulas for Natural Frequency and Mode Shape," by Robert D. Blevins,l979,Van Nostrand Reinhold.

[B]

AES Calculation PCI-5343-S03, "Prairie Island Strainer Sloshing Evaluation", Revision 0.

[9]

AlSC Manual of Steel Construction, 9th Edition.

[I 01 Wilti Product iechnical Guide, 2008.

[?I]

Wisconsin Electric Guideline DG-MO9, Design Requirements for Piping Stress Analysis, Revision 2.

[I21 "Engineering Fluid Mechanics" by John A. Roberson and Clayton 7.

Crowe, 2nd Edition, Rudolf Steiner

' Press, 1969; Library of Congress Catalog Number 79-87855.

[I 31 AWS Dl.lID1.I M:2002, "Structural Welding Code - Steel".

[14]

Wisconsin Electric Gufdelioe DG-MI 0, Pipe Support Guidelines, Revision 2.

[I51 Wsconsin Electric ~uideline DG-C03, Seismic Design Criteria Guideline, Revision 0.

[16]

"Roark's Formulas for Stress & Strain" by Warren C: Young, 6th Edition, McGraw-Hill 1989.

' [17]

"Theory of Plates and Shells" by Stephen P. Timoshenko and S. Woinowsky-Krieger, 2nd Edition, McGraw-Hill, 1959.

[I 81 PC1 lntra -Company Correspondence from Greg Hunter, Dated February 20,2006, Subject, "Testing of 3/16" Blind Rivets and 3/16" Closed End Rivets" (with test reports attached). (Attachment C)

[I91 ASME Publication, "Pressure Vessel and Piping: Design and Analysis," Volume 2, 1972, Components and Structural Dynamics, Paper Title " Design of Perforated Plate," by 08Donnell& Langer Reprinted from Journal of Engineering for Industry, 1962.

Form 3.1-3. :

%

• I.

,I. 1.2.

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i;;

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1..:.

.. 6 3..

...... :I..

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li.

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nt...,.,. -1..'

I

,,< " ' I..... ; ':". '

, q i j

,...,; ;*-,..,.. :,*;,....;! ;:!t,,;,i

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.,.,p. >:t-;:

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i;,*,.,; ;;,;.> :,hi' ij.8@;;tk!,,2i..;. *...!;;;;;

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' : " - ' : 'i~:.'::-:'~""

>!:@ $j:<$,$&

,$QG.xg; c*;+:.5,x

,+

.t,;(32$.#.;; is.k ;%y.;;;;

.,{:,

,,<,,,,??,.8, $,*:,; >$.: ;;

, ;;;

;; :;$$::$~{3~!i.~i.;

,;;:,::. !.$ti!

3<$!s&:.

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• .~~~?!;~;i$,:i~.;$8,i$;$;~~~~!

%!y<:

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:~!J+l:~~~-+f?

s!J t..,,.,&..,w.,:,.,.

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!,.

....... '. :..:*, ;.....,> : f;;;;:

rxr!,l P.,q >,,., 2. il,wj ~~~:;.~~,,~$~$>!&~;

,...:A,?>.ir

..**..'--+.. ":b'5 ""

s I,-,... t.-'!:J,.,.,..

, \\ A ?.j 1."......,,.,,,,, 7," -..,... >.,,,,, :...,:;.>?~.&>;,g:,:.,:;

I Form 3.1-3 Rev 2 1201 "Marks' Standard Handbook for Mechanical Engineers", by Avallone, and fjaumeister, 9th Edition, McGraw Hill.

[21j AES Verification and Maintenance File for GTSTRUDL Version 25.

1221 AISf Specification for the Design of Cold-Formed Steel Structural Members, 1996 Edition.

1231 ANSllAWS D1.6:1999, "Structural Welding Code - Stainless Steel".

[24]

Not Used

[25]

ANSYS Verification File, Version 5.7.1, dated 9/28/2003., AAESMN File No. AES.1000.0562.

[26]

EPRI Document NP-5067, "Good Bolting Practices - A Reference Manual for Nuclear Power Plant Maintenance Personnel".

[2n PC1 Technical Document TDI-6007-04, "Module Debris Weight - Point Beach Nuclear Plant Units II2",

Revision 3.

1281 Nukon Pipe.ESD-TR-146, ;Latch and Strike Tensile Strength Test," August 11,1993. (Attachment D) 1291 Jay-Cee Sales and Rivet Inc, "Expanded Product Line", 4th Edition. (~kachment E) 1301 American National Standard ANSllAlSC N690-1994, "Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for'~uc1ear Facilities"

[31]

ASCE Standard SElfASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members".

[32]. "Theory of Elastic Stability" by Stephen P. Timoshenko and James M. Gere, 2nd Edition, McGraw-Hill, 1961.

1331 Journal of Ship Research, "Sway Added-Mass Coefficients for Rectangular Profiles in Shallow Water", by Ffagg, C.N. and J.N. Newman, December 1971. (Attachment F)

[34]

Journal bf Engineering Mechanics ASCE, "Added Masses of tenses and Parallel Plates", by Sarpkaya, T.,

1960. (Attachment G)

[35]

Stainless Steel Sheet Thickness Table from Hendrick book. (Attachment H)

A

1361 Bechtel Drawing No. 6128, Containment Structure interior Plans at El. 10'-O", EL. 21'-Ow, EL. 24'-8", and EL. 38'-O", Rev. 9. (Unit 1)

EL. 38'-O", Rev. 8. (Unit 2) 3 8 "Fundamentals of Engineering Thermodynamics, SI Version" by John R. Howell and Richard 0. Buckius, McGraw-Hill, I987.

1393 "Welding Formulas and Tables for Structural and Mechanical Engineers and Pipe Support Designers", by T.S. Hobert, 1983.

[40]

EC No. 9306 affecting drawing "SFS-PB2-GA-02, and SFS-PB2-GA-03, Revision 8", Revision 0.

[41]

EC No. 9355 affecting drawing "SFS-PB2-GA-02, Revision 8", Revision 0.

[42]

EC No. 9364 affecting drawing "SFS-PB2-GA-03, Revision 8", Revision 0.

[43]

EC No. 10627 affecting drawings "SFS-PBI-GA-03, Revision 6 and SFS-PBI-GA-04, Revision 5", Revision 0.

[44]

ACI Structural Journal, January-February 1995, VOL. 92 NO. I (Atfachrnent J)

[45]

EC No. 10581 affecting drawings "SFS-PBI-GA-00, Revision 6 and SFS-PBI-GA-02, Revision 6", Revision 0.

[46]

DlT-008 for EC 12603 and EC 12601 From Point beach 911 8108.

[4n ASTM Standard Specification A493-85, "Stainless and Heat-resisting Steel for Cold Heading and Cold Forging - Bar and Wire".

[dB]

Lehigh Testing Laboratories Test No. G-4-27, dated August 3,2007, with test reports attached. (Attachment K)

[49]

Not Used.

[50J Bechfel Drawing No. C-3181, (Unit I)

[

I Lehigh Testing Laboratories Test No. F-19-32, dated July 20,2006 (with test reports attached) (Attachment L).

ENCLOSURE 4 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-1911GL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS PERFORMANCE CONTRACTING, INC.

CALCULATION PCI-5344403, REVISION 4, SEPTEMBER 24,2008 EVALUATION OF SUMP COVER AND PIPING FOR THE CONTAINMENT SUMP STRAINERS (ABRIDGED)

POINT BEACH NUCLEAR PLANT, UNIT 1 21 pages follow

bolt directly to floor, evaluated uplift load on sole plate &om valve testing Rev. I

Revision Affected Pages Revision Description Approval Signature I Signatore I Initials of Dato Proparers & Reviewers 3

1-3,69-71,83, Changed faulted shear stress allowable Prepared by:

85,93-94, to agree with DG-MI0 and the effective 110-1 1 1 shear stress area for angle andtee sections to agree with paragraph C3 (p.

5-315) of AISC 9th Edition Added 1A

, Commentary on the Specification for Reviewed by:

Allowable Stress of SingIcAngle Members. This ~esolves CAR-07-004.

4 I& 2,3,6,13, IncorporatedDIT003 for EC 12601 &

14,26,29-31, EC 12603. Reanalyzed static analysis 37, W, 40-42, (Attachment A) to incoi'porate increase 44-47,55-60, in pressure thrust. load on piping due to 62,63,65-68, strainer pressure imbalauce.

70-73,78-80, Rwiowed by:

S%88,91-95, 97,98,100-103,105,108-112,114 Added44A Attachment A

PROJECT NO: PCI-5344 C L m Performance Contractha Ino.

CALCULATION TFT

2. System Description

3. Design information package from related equipment

4. Electrical mcipline Xnput

5. Mechanical Discipline Input

6. Control Systems Discipline

7. Structural Discipline Input

8.

Sp&fications

9.

Vendor Drawings

10. Design Standards

11. Client Standards

12. ~heekdd Cdculations PREPARER'S SIGNATU DATE:. 9/24/2008 ReVEEWER'S SIGNATURE:

DATE: 9/24/2008 I

Wal;ni Sif-6.

Rev. 3

A. Properly Identified?

13, property Applicd?

3. Wcrc tile inputs correctly selccted and used?

4, A. IYas Design Input Log uscd?

5.

Axe necessary assumptions odcquatety staid?

6. Arc the nssumptions reasonable?

7. Was the calculation methodology appropriate?

8. Are symbols and abbreviations adequately idenfificd?

9. Are tl~c calculations:

C. Easy to follow?

U. Prcscntcd in logicul ordcr?

E. PI-cparcd in propcr format?

10. Xs the oufput rcesonqblc compared to the iriputs?

1 I. En computer program was used:

A.. Ts the program listed on the ASIa and has the SRN been reviewed for any program use limitations?,

B. Have existing user notices and/or error reporfs for tho productionv~rsion been revieived as appropriate?

D. Were they appropriate for tf~e application?

E. Were tiley corrcctly uscd:

F. Was data input correct?

G. ]j: tho computer progrem and revision ideaitificd?

Form 3.1-4

TABLE OF CONTENTS 1.0 purpose/Objective.....................................

5 2.0 Methodo[ogy............................................................................................................................................... 5 3.0 Acceptance Criteria.......................

....................................................................................................... 8 4.0 Assumptions...................................................................................................................................................

5.0 Definitions and Design Input.......................................................................................................................

11 5.1 Material Properties.............................................................................................................................. 11 5.2 pipe Geometry and Dimensions...............................................................

i.........................................

12 6.0 Calculations....................',...........................................................................................................................

13 6.1 Weight Calcufations......................................................................................................................... 13 6.2 Pipe Loads.........:............................i.................................................................................................

13 6.3 Calculation of Acceleration Drag Volumes and Hydrodynamic Mass 16 6.4 Piping Eva[uatlon 25 6.5 Flange Evaluafions 32 6.6 Sole Plate Analysis........................

57 6.7 Support Eva(~&~ns 63 6.8 Integral Welded Attachment Evaluation.......................................................................................

g7 6.Q Slip Jolnt Evaluation 407 7.0 Results and Conclusjons 109 8.0 References 112 Attachments

'. Paaes A

"B"@alner Piping (Static)....................................................................................................

- A31 B

B" Strainer Piping (Seismic I)

B1 - 637 B Strainer Piping (Seismic 2)

Cl - C38 C

1.0 PURPOSEIOBJECTIVE The purpose of this calbijlatiitri is f6 qualiiji ti19 sutiitj dover, piping, and pipfnq ~upports~assdciated Performance Cqntracting lnc, (PCl) Syction Strainers fo be installed In &&aFWra~ag%@@e Polnt Beach Nuclear Plant Urilt I, This calduliitjon.9ivalua~es, by analysis, the plijing a s well as the supporting struciure$ associak9d with the new piping; The evaluations encompass sll pfpiing.from and including the.sump cover plate (sole plate) attached to fhe El. 8'.floor slab to fhe strainet corinedt!ons including intermediate support itnptures..

2.0 METHODOLOGY The evaluations are perfoned tieing a cqinbliiatlon of inapual calculaflons and computerized p$~lhg:

using the AutoPlPE Program (rj.efer++jnce [16]). The piplog is considered as an attachment' or extension to the strainers and are therefdie siibjeqk to the $qu~rem.~,pts of Strainer Design Specification P'B-68i'(Reference [I]).

.E:xcepijoris froom, these requl@inenfsi if taken, are discussed' and justified wlihin this calculation; seismic Loads Tiie.straiher piping is categodzed a s Seis;mic Class I equipment and is.required to be opBrable dudng and.

after a safe shutdbwn earthquake (SSE) witlidqt %ceed[tig normal aliowaljle stresses a s specified In Secefon 5.4.7 of DGC03 Seismrc Dgiigfi Criterjq GuideI& (Reference.[?4])., Strainer DeslQn Speolfication PEi-681 (Reference [I]),

@qirjr:$,the @pifig to.be ~valuateg for tLvo operating conditions. The first Condition

. is.a !'dr$J condltlo~ with no ~circtil.ation water i.ns1d.e oi egemal waterpresent. The second ctjndition Is a submerged wet'! condiflori %jth dcl~gla@on water? For the selsmic~evaluatlon the piping will be consfdered'sljbmerged and fgll of.wate; Thg wateyleve! is~considered to be a minimum of 3'4T above tHe 8' floor elevation (Ei:l,l'-.%),

The pi@rig."dry'f state with Its associated mass bdlng much Ies3, iliiill not be considered as It is less seV@re than the "wett1 state.,

Per the. specification, the seismic evaluation is. required to take into account.any seismic slosh (analyzed at the seismic worst-case water level) of the tec!reulation water. Based on Keference [ZOJ, 6ecauso.of the negllglble load rnagnitl\\des,.it is detemined that the seismic slosh loads. in PWR co~teinmenfs are insignificant by comparison with other selsiitlc loads. Therefore, seismic slosh loads are neglected from the pipe stress analysis. N0t.e ttiai the s!dshiii'g gatculgtian of Reference [?dj is done forthe Prairie lslancl strainer projeqt and it I s repk&eqtatiye fdi a!!

~ ~ ~. c o h f a ! ~ ~ m e h f s I? general, and th.erefore, It is appllcqi?,l!$.

for use in this calculation,. Tlie "wet? s t ~ l n a r op,e@ting Condition will considerthe strainer qsscmblies submerged jn stiil watef at tue seismic woist-cat?& water I&vel when subjected to selsmlc lnhlfial loads.. The inertial effects of the added hydri;ldynarnic mass duq fo the submergenca of the plpfng is corisidere'd, The piping, is seismically $biIlfied u$i"g the r8spoh~1 spec$ta method. ~iie:applicable ~eisihic sljectia $tie provided.in Selsrnic ~.uarifi.caf.iop S@ecification ~ljieet

~h-002243 (Reference [2]) These loads a& applied to the piping through base m0tio.n respun$e specte a s detailed in the Seismic Design Criteria Guideline DG-CO~

(Reference [I 4a.

CALCULATION SHEET All piping Is located on the 6' floor elevation of the containment. The response spectrum chosen is for the 6.5' elevation of the containment. The containment liner plate is located at the 6.5' elevation and there Is an additional 1.5' of concrete on top of the liner plate. The slab between the 6.5' elevation and the 6' elevation is very rigid, Thus it Is appropriate to use ths response spectrum for the 6.5' elevafion. The vertical direction response spectrum is 213 the value of the rnaxfmum ground horizontal response spectra.

The piplng is considered a s vital plping and the damping values for selsmic loads is taken a s 0.6% for both the Operating Basis Earthquake (OBE) and the Safe Shutdown Earthquake (SSE) a s required by Seismic Design Guide DG-CO3. The response spectrp Inputs are for the OBE environment. For evaluating stresses, displacements, loads, etc,, for the rnaxlmum credible earthquake (SSE), the values obtalned from the OBE analysis are to be increased by a factor of 2.0 (Reference [Ill).

The plping is excited in each of the three mutually perpendicular directions, two horizontal and one vertical.

Per Reference [I I], the modal combination Is pertonned bythe use of the double sum method to account for the effects of modal ooupling in the response (I.@. closely spaced modes). An earthquake duration of 30.24 seconds was used in the analysis per DG-CO3, Appendix C, Appendix N of the ASME code indicates th;?t the rnaxlmum acceleratlorts generally occur in the first TO seconds. Two analysis were run -one with ID sec and one with 30.24 sec. Since the results were the same, the analysis viiith 10 seconds is the official documented seismic analysis. Responses due to the three spatial components are combined by SRSS. (Reference PI],

paragraph 5.8.5). The cutoff frequency is taken at 30 hz or a minimum of' 5 modes are included, Zero Period Acceleration (ZPA) residual mass effeots are considered since they may sl~niflcantly affect the plping. The ZPA response is comblned with the response spectra response by SRSS.

Since all piping is supported from the same El. 8' floor slab, there are no relative seismic anchor movements.

Operating Loads operakg loads are comprised of weight, thermal expansion and pressure loads.

The thermal expansion is taken at a temperature equal to the maimurn operational inlet temperature to the RH Exchangers of 250 OF (Reference [I]). Small gaps (3132") are modeled on the U-bolt side only of the two-way restraints (Type PS3} on the "B" imin piping (Reference 1371. These gaps were modeled to reduce the high thermal loads encountered due to the several bends associated with the "B" train piping. The. '

design drawings (Ref. I6bJ) ensure ttiat these gaps will be available. Note the Autoplpe model was rerun to, account for these modified gaps; Because the attached piping is connected to the strainer with flexible joint it essentially behaves a s an open * '

ended system, this pressure differential will also create an axial thrust force on the plping. The maximum diffelenfial pressure load acting' on the piping is fhe hydrostatic pressure associated with the maximum allowed head loss through the debris covered strainers. This is defined a s 108 of 68 O F water In Reference

39..

CALCULATION S3EICEET Mathcad software is used to generate most of the calcuiatlons. All MathCad calcuiafions are Independently verified for accuracy and conectness as if they were manually generated. AutoPlPE Version 8.05 Is used for the piplng analysis. AutoPIPE Version 8.05 is vetifled and validated underthe AES QA program as documented in ihe AES validation and maintenance ffks (Reference [16;1). Because the AutoPlPE Version 8.05 only perfomls piping evaluations using the 2001 Editlon of the B31.1 Code instead of the required 1998 Edition, a reconciliation of the 2001 Code to the older 1998 Code is performed.

The only provisions of the code that could potentially affect the results of the piping analysis are changes in material properties and design equation provisions. A revlew of the codes and the material specifications shows that the only physical properties OF material that affect the design of code items are the minimum yield, the tenslb strengths and the coefficient of thermal expansion because these afe the basis for the allowable stresses and the tabulated llE" and "a" values at temperature. As long as the specified tensile properties of ths material have not changed, use of the later Edition does not affect the end result.

The material ailo&bles stresses are included manually into AutoPlPE based on the ASME B31.1 - 1998 Edition, which is the design code for plpe stress analysfs.. In addition, a review of the two the codes Was performed to identify revisions to the design equation pmvlsions and to determine if any material properties associated with "El1 and "a" had changed. There have been no design dependent revisions to ihe piping material and to the design code equaffons. The flexlbllity and stress intensification factors, and the method for combining moments are the same for both code editions. Therefore, the results between the two code editions wilt be Identical.

L

3.0 ACCEPTANCE CRITERIA The strainer suction piping shall meet the requirements of the strainer design specjficafion PB-681 (Reference [I]). As stated in PB-681, the detailed evaluatlons are to be performed uslng the rules, as appli~able, of ANSIIASNIE B3t.1 Power Piping 9998 Edition (Reference [S]).

The piping supports, baseplates and other mountlng hardware is evaluated to AISC 8th Edition as permlffed In paragraph 120.2.4 of the 831.1 Code. Additional guldance is also taken from other codes and standards where the AISC does not provide specific rules for certain aspects of the design. For instance, the cover plates, stiffeners angles, support components are made from stainless steel materials. The AISC Code does not specifically cover stainless' steel materials. Therefore, ANSllAlSC N690-1994, "Specification for the Design, Fabricafion, and Erection of Steel Safety Relafed Structures for Nuclear Facilities", Reference [2q is used to supplement the AISC In any areas related specifically to the sfructural qualification of stainless steel, Note that only the allowable stresses are used from this Code and load combinations and allowable stress factors for higher service level loads are not used, SEIlASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members",

(Reference 1241) is used for certalh'components (sfsinless steel bolts and pins) since the AlSC does not provide specific bolting aflowables for stainless steel bolting. The rules for Allowable Stress Design (ASD) as specified in Appendix D affhis code ate used. Finally guidance is also taken from AWS D1,6, "Structural Welding Code - Stainless Steel", (Reference [2q) as It relates to the qualiflcatlon of stainless steel welds.

Detailed acceptance criteria for each type of strainer component is provided in the sections below.

Load Combinations The applicable load combinations for the piping are those from Section 6.0 of DG-MO9 (Reference [TI]).

Load Condition

~ombin&loq (I) Normal P+DW (2) Upset P+DW+OBE (3) EmergencyIFaulted P + D W + SSE (4) Thermal M

where, DW = Dead Weight Load P = Differential Pressure OBE = Operating Basis Earthquake SSE = Safe Shutdown Earthquake TI = Thermal Expansion The thermal expansion stresses are based on a stress range from the amblent condition of 70 OF to the maximum operating condition of 250 O F (AT = 380 Or-).

CALCULATION SHEET The piping Is evaluated In accordance with ANSI B31.1 Paragraph 104.8 as applicable. Since the 831.I does not explicitly Identify how to incorporate the emergency SSE loads, PBNP uses ASME Section Ill as a guide as discussed in Secflon 0.0 of DG-MO9 (Reference [Ill).

831.1 Eu. No Load Condftion Stress Combination Allowable Stress Upset (Occasional)

P + DW+ OBE Emergency (Occasional)

P + DW + SSE Thermal (Displacement)

T I I,OS,

~fanges Since specific detailed guidance Is not provided in B31.1, the bolted flange connections at each end of the piping elbows will be evaluated in accordance with ASME Section III, Appendix L (Reference [B]) guidelines.

The flange bolts are quaiifled to the criteria presenfed In ASME 111, Appendix L (Reference [8]). Note that these are non-standanl flanges which do not meet the generic requirements of 631.7 (such as weld size),

As stated in the fbnnrard of of the B31;l Code (Reference t5]), "a designer who is capable of a more rigorous analysis than is specified In the Code may Justify a less consanmtive design, and still satisfy the basic intent of the Code." Use of a defailed stress evaluation of the flange and the flange weld, based on ASME analysis equations, certainly falls within this category of satisfying the basic intent of the Code.

Pipfng Support S@cn?hJraf C~mponenfs The allowable stresses on the piplng support components are based on the AlSC 9th Edition (Refemnce [9]).

Also, the allowables stresses for the sump sole plate tabs, bolts, and welds ambased on the A!SC 9th Edition. The allowable stress for the SSE Load Combinations is taken from Secflon 6.9 ot: DG MI0 (Reference

[I 31).

Load Condition Load Combinatlon Allob~rable Stress Nonal

' DW+TI 1.0 AISC Upset DW+ OBE + TI 9.0 AlSC Faulted DW + SSE + T I 9.5 AlSC but not to exceed 0.9 Sy

CALCULATION SBEET Per Reference 1251, because sfainless steel does not display a single, well defined modulus of elastlclty, the allowable compression stress equations from the AISC are not applicable for stainless steels. Therefore, the allowable compression stress will be based on the lower ailowables from Reference [25j as opposed to those provided in the AISC Code (Reference [9]). Per Q1.5~9.2 of Reference [25j, the allowable stresses for tension, shear, bending and bearing for stainless steel can be taken a s the same allowables pFovlded for carbon steel, therefore the AISC 9th Edition will be used for allowables for these types of stresses.

Welded Joints Allowable stresses for piping welds, such a s the flange fillet welds, are per ASME Sedlon Ill (Reference [B]),

Paragraph NC-3356, IWA welds are in accordance with ASME Code Case N-318-5 (Reference [I%]).

The allowable stresses for all other welds are based on the AlSC 9th Edition (Referenoe [9]). AWS D1.6 (Referance [26J1 was revfewed to ensure that any special qualiflcation requll.ernenls associated wlfh stainless steel welding were considered. Since the weld allowables provided in AWS Df.6 are essedlally the same as allowed for carbon steel welds under AWS D1.1, no special adjustments are required to account for stainless steel. The allowable shss for the SSE Load Combinations is taken a s 1.5 times the AlSC weld material allowable per Reference [13].

Integral Welded Attachment Evaluation The localized stresses developed In the pipe due to the Integral welded attachments (shear lugs) are added to the stresses calculated by AutoPlPE and compared to 831.V allowables. ASME Code Case N-318-5 (Reference [19]) is used to calculate the local stresses since this Is the latest version of the Code Case available.

Mouneing Hardware Hiltl Kwil<-Bolt Ills are used to mount the support baseplates to the floor. The analysls and deslgn of expansion anchors shall be in accordance with the Hlltl Technical Guide (Reference [IB]), however, a Factor of Safety of 4 against ultimate loads will be used. Prying factors are calculated in accordance with DG-COI (Reference [to]). Quallflcations of the stainless steel boitsipins used to attach the saddle plates to the structural angles is based on the ASCE Standard (Reference 1241). The AISC Code (References [B] ) does not provlde specific bolting allowables for stainless steel bolting.

'4.0 ASSUMPTIONS None.

'. I.....

a,:.,.

$..:: :?;:::..;

i.. ' L..,.........:..'. i;..:.:..

.'.... 9.;;,

i'.'::'

.-......... i.';

.:. i ;..,,..........,. ;;

.r.; F7.,;.:,:.o 4

>$. ?:;;!..

...;~,.: ?..:,

..$;:f/$;::?j::

)$;!?:;:

(;;;i$;

CALCULATXON SHEET 6.0 DEFINITIONS AND DESIGN INPUT 5.4 Material PropeNes The specific materials for the piping and support components ate taken from Reference 6m Stainless Steel ASTM A312, Type 304 or Type 304L (Dual Certified)

Stainless Steel ASTM A240, Type 304 or A774, Type 304L (Dual Certified)

Structural Steel:

Stainless Sfeel ASTM A276, Type 304 Flange:

Stainless Steel ASTM A-240, Type 304

'Flange Bolting:

St'ainless Steel ASME A-193, Gr. B8, Class I1 Design Tempemture TdBS=

250 OF ( Reference [I]

)

Properties for the pipe components and support,sftuctural components are talcen from ASMElANSl 837.1, power Plping Code, 1998 Editlon (Reference 151). Yield strength values for support sfnrctural components and flange bolting properties are not available in ANSI 831.1 Code and are takeq from ASME B&PV Code,Section I I, Part D (Reference f41). For Dual Certified materials only the controlling properties are used.

Yield strength valuefor stainless steel A240 Type 304 material at 250 DF:

Smo4 := 23.6.ksi (Ref. 141)

Modulus of Elasticity of stainless steel, material at 250 OF:

E := 27300.ksi (Ref. Fl)

Allowable plpe stress at deslgn temperature (250 OF),

Sh := 17.20- ksi (Ref. S5l)

~lfowable design stress for flange at deslgn temperature (250 OF),

Sf:= 17.2O.ksi (Ref-Allowable bolt stress at design tempemture (250 OF),

Sb

= 25.0-Icsi

( ~ e f -

~41)

Modulus of ~laitlcity (flange)

Ef := 27300.ksi (Ref- [q)

. Modulus of Elasticity (bolts)

. Eb := 27300. ksi (Ref. 141)

Other Miscellaneous Properties Ibf Density of stainless steel (Ref. [ZB]).

pstml := 501.-

ft3 Polsson's ratfa of stainless steeI (Ref. [28]).

u := 0.305 Density of water at temperafure of 68 O F (Ref. [12J)

Ibf 7 ~ 2 0

= 62.4.-

k3

5.2 Pioe Geometry and Dimensions Pipe Dimensions Outer diameter of pipe (Ref. [6b])

Pipe wall thickness (sch. ID) (Ref. [6b])

lnside diameter of pipe: IDitlpe:= ODplpe - 2.tplps r:= B.O.in Corrosion AIlowancelFabricatlon Tolerance

. Pool Boundaries Length from top of floor to centerfine d pipe (Ref. [6aI)

Mlnimum height of the water above the floor (Ref. [Gal)

Distance (left side) from wall to pipe centerline (see Section 6,3.1)

%I:= 13.85-in

.Distance (right side) from wall to pipe centertine (see Section 6.3.1)

%, := 24. in Flanae Dimanslons Outer diameter of flange at top of elbow (Ref. [6fl)

OD,,,,

= 25.O.in Inside dlameter of flange at top of elbow (Ref. [6fl) lDffange := 16.125.ln Flange thickness (Ref. [6fl) tnanga:= 0.2541 Outer diameter of 16 plpe in-line flanges (Ref. [Cb])

ODflgeq6 := 23.5.in Inside diameter of 16 pipe in-line flanges (Ref. [6b])

lDRg-le:=.16,125.in

Figure 6.4.1 - Model Plot of "B" Strainer Pipe

Client: Performance --..-.--.-.

I Revision: 2 I

Sole Plate Connectlon As shown in figure bdow the connection consists of two parts. The fabricated pipe Range is identical to fhe flange on the opposite side of the elbow, the 112" annular sole plate ts held down by twelve (12) 518" Hilti lfl expansion anchors (Reference 16~1).

Note that the 4" minlmum distance to the edge of the sump draln concrete opening a s shown in the sketch below has been reduced to a minimum of 3" in EC 10581 (Reference 135)). The centerline of the bottom end of the elbow and the associated base ring may be offset a maximum of 1" from the centerllne of the sump drain pine sleeve durina instalfation to avoid interferences.

All three types of flanges (in-line, top of elbow, sole plate) will be analyzed concurrently using arrays. Loads for the in-line flanges will be divided into NomallUpset and EmergencylFaulted loads, but enveloped between all flange pairs. Dlmenslonal parameters are adjusted as requlred for each type of flange.

7.0 RESULTS AND CONCLUSIONS J R B ~ I ~ ' ~

f= m a ~ ( l ~ p, ~,

~ R B ~ ~ B ~. I R ~ ~ - I ~ c. I I.R@IB]

S;'tress '$urnmaw for other Components lnteradfion Rafio

~ l a h g 6 ~ e l d to pipe 6.6 lRcl = 0.24 Mi~sinq BOI~S.:a p+

,j,g, p9b4vJ/bb.

b+~f@/'?/o@

Flange Bolts 8.5 l~boltmlSslng = 0!93 Flange ~endlng 6.5 lRflah$e.plssing = 9.00 Sole Plate Connection 0.17 NormaNUpset.

'Sole.'~late 0.6 IR?~,~!B.~u$

=

Ernergenc~Faultetl Sole plate Expansion Anchors 6.6

., ( 1 l R ~ ~ ! $ a ~ ~ h o r

= b.8P

CALCULATION SHEET Com~onent Type PSIIPS2 Restraint Angle Normal Stress Angle Shear Stress Expansion Anchors (Type PSI)

I I

I Expansion Anchprs Vype PS2)

Baseplate Weld of Angle to Baseplate Saddle Plate Bending Saddle Plate Shear Saddle Plate Welds Saddle Plate Pins Shear Lugs Ref. Section Interaction Ratio Integral Welded Attachments 6.8.1

CGILCWTION SHEET Safety Related I Date: 9/12/08 I

Tvw PS3 Restraint I

I I

IR shown are for Faulted Loads (SSE) v&sus Upset Allowables {OBE)

I I

W6x15 Normal Stress W6x15.Shear Stress I

Expansion Anchors Meld of W6x15 to Baseplafe I

Angle Normal 'Stress I

Angle Shear Stress I

Weld of Angle to W6x15 U-Bolt Normal Load Type PB1 Restraint Stanchion Plate Bolts Integral Welded Attachments Other Pipina com~onents Slip Jolnt 0.71 Upset Ernerg I I The evaluation of the piping and piping supports associated with the suction strainers has shown that the pipe stresses and support loads are acceptable, The piping stmsses, flanges, and support component stresses are wlthin their respecflve applicable limits and are therefore acceptabb

CALCULATION SEiEET

8.0 REFERENCES

[I] Point Beach Nuclear Plant Specification PB-681, "Replacement of Containment Sump Screens", Revision 2

[2] Point Beach Nuclear Plant Seismic Qualification Specification Sheet SQ #002243, Revlsion 0

[33 ASME/ANSI 831.1, Pressure Piping Code, 2001 Edition.

143 ASME B&PV Code, Section 11; Part Dl Material Propertfes, ID98 Edition, through 1999 Addenda

[S]

ASMUANSI 831.1 Pressure Piping Code, 1998 Edition, thraugh 1999 Addenda

[6] Performance Contracting, inc.(PCI), Sure-Flow Suction Strainer Drawlngs 6a.

PC1 Drawing No. SFS-PBI-GA-00, "PB Unit I Sure-Flow Strainer, Recirc Sump System", Revision 9 6b.

PC1 Drawing No. SFS-PBI-GA-04, "PB Unit 1 Sure-Flow Strainer, Piplng B Layout", Revision 6 6c.

PC1 Drawing No. SFS-PBI-GA-05, "PB Unit I Sure-Flow Strainer, Piping A Layout", Revision 9

, 6d. PC! Drawing No. SFS-PBI-PA-7105, "PB Unit 1 Sure-Flow Strainer, SleeveslCoverslSuppo~tsIPins~~,

I Revision 15 6e.

PC1 Drawing No. SFS-PBI-PA-7160, "PB Unit 1 Sure-Flow Strainer, Sump Inlet Cover", Revision 1.

I 6f.

PC! Drawlng No. SFS-PBI-PA-7161, "PB Unit 1 Sure-Flow Strainer, Sump Connection Elbow AIIBI",

Revision 0.

6g.

PC! DFawjng No. SFS-PBI-PA-7162, "PB Unlt 1 Sure-Flow Strainer, Pipe BZ", Revision 2 8h.

PC1 Drawing No. SFS-PBI-PA-7163, "PB Unit 1 SureFIow Strainer, Pipe B3It, Revision I.

6i.

PCI Drawing No. SFS-PBI-PA-7164, "PB Unit I Sure-Flow Strainer, Pipe 89, Revision 1.

61.

PC1 Dmwing No. SFS-PBI-PA-7185, "PB Unit I Sure-Flow Strainer, Pipe BY', Revision 3.

6k. PC1 Drawing No. SFS-FBI-PA-7166, "PB Unit 1 Sare-Flow Strainer, Pipe AT', Revision 2.

61.

PC1 Drawlng No. SFS-PBI-PA-7167, "PB onit 1 Sure-Flow Strainer, Pipe A3", Revision 2.

6m. PC1 brawing No. SFS-PBI-GA-OI.,

"PB Unit I Sure-Flow Strainer, General Notes", Revision 12 m "Fonulas for Natural Frequency ahd Mode shape," by Robert D. Blevins,l979,Van Nostrand Reinhold.

I

[q ASME B&PV Code, Section Ill, Division I, Subsections NB, NC, and NF, 1998 Editlon through 1999 Addenda, including Appendices.

[9]

AlSC Manual of Steel Cohsmction, 9th Edition.

1101 Wisconsin Electric Guideline ClG-CO1,"Guidelines for Design, Qualification, and Installation of Concrete Expansion Anchors at Point Beach Nuclear Plant" (with revisions per NPM 92-0428, April 27, 1982). Revision 0

[I I] Wisconsin Electric Guideline DG-MOD, Design Requirements for Piping Stress Analysis, Revision 2.

1121 "Engineering Fluid Mechanics" by John A, Roberson and Clayton 7.

Crowe, 2nd Edition, Rudolf Steiner Pmss, 1969, Library of Congress Catalog Number 79-87855.

[I31 V\\?sconsin Electric Guideline DG-MID, Pipe Support Guidelines, Revision 2.

1141 Wisconsin Electric Guideline DE-CO3, Sefsmic Design Criteria ~uidaline, Revision 0.

' [15]

AES Calculation PCI-5344SO1, "Structumt Evaluation of Containment Emergency Sump Strainers", Revision 0.

[I61 AutoPipe Version 8.05 QA Release 08.05.0D.16 Verification Report, AES File AES.IOOO.0513.

1171 Welding Formulas and Tables for Structural & Mechanical Engineers & Pipe Support Designers Published by I.V.1. Structural Design Service, Copyright 1983.

[I81 Hilti Product Technical Guide, Copyright 2005.

[I91 Cases of ASME Boiler and Pressure Vessel Code, Case N-318.5, "Procedure for Evaluation of the Design of Reefangular Cross Section Attachments on Class 2 or 3 Piping1', April 28, 1994.

[ZO]

AES Calculation PCI-5343403, "Pralrie Island Strainer Sloshing Evaluation", Revision 0.

1211 Roark's Formulas for Stress and Strain, Warren C. Young, 6th Edition.

[22] "Design of Welded Structures1' by Omer W. Blodgett, 1969, Library of Congress Catalog Number 66-23123.

'[23] Mechanical Engineering Design by Joseph Edward ShigIey and Larry D. Mitchell, McGraw Hiif, 1983.

[24]

' ASCE Standard SEIIASCE 8-02, "Specification for the Deslgn of Cold-Formed Stainless Steel Structural Members", Copyright 2002.

[25] ANSIIAISC N690, "Specification for the Design, Fabrication, and Erecflon of Steel Safety-Related Structures for Nuclear Facilities" copyri'ght 1994.

,1263 ANSllAWS D1.6:1999, "~twctukl Welding Code - Stainless Steel".

1277 Bechtel Drawing No. C-128, Containment Strucfure Interior Plans at El. 1O1~0", EL. 21'-01', EL 24'-B", and EL. 381-01'1 Rev. 9. (Unit I)

[28]

"Marks' Standard Handbook for ~ehhanlcal Engineers", Avallone & Baumeister, 9m Edition, McGww-Hill

[29]

Good Bolting Practice, Volume 1, EPRl Report NP-5067

[SO]

Rigid Frame Formulas, A. Klelnlogel, 2nd Edition, Frederick Ungar Book Publishing I ]

, PBNP Deslgn lnformatlon Transmittal (DIT) for Modification MR 05-017, Point Beach Unit 1 Sump Strainer New Base Plate Design, from T. Corbin (NMC) to C. Warchol (AES) and J. Blelgh (PC!), dated 01-12-07 1321 AlSl Specification for the Design of cold-~omed Steel Structural Members, 1996 Edition,

[33]

Lehigh Testing Laborafories Test No. F-19-32, July 20,2006 1341 Engineering Change Notice 10580 to Modification EC 1602 (MR 05-017). Revision 0, Dated 4/7/07 1351 Engineering Change Notice 10581 to Modificatlon EC 1602 (MR 05-0171, Revision 0, Dated 4/11/07 1361 Engineering Change Notice 10653 to Modification EC 1602 (MR 05-017), Revision 0, Dated 4119107

[371 Design Information Transmittal for Point Beach EC 10720, "Thermal Expansion Gap Requirements, Dated

[38]

ACI Structural Journal, January-Febnrary 1995, VOL. 92 NO. 1 (Included as Attachment J to Calculation PCI-5344-Sol) 1393 Point Beach Design Infomlation Transmittal DIT003 for Modification EC 12601 and EC 12603, "Differential Pressure for Debris Interceptors", Dated 8i7IO8

ENCLOSURE 5 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-1911GL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS PERFORMANCE CONTRACTING, INC.

CALCULATION TDI-6007-07, REVISION 4, MARCH 10,2009 VORTEX, AIR INGESTION, & VOID FRACTION POINT BEACH NUCLEAR PLANT, UNITS I

& 2 22 pages follow

'., ?

I I

, I 1

i I.....

Vortex, Air Ingestion &Void Fraction - Point Beach Nuclear Plant -

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 I

I=ALCULATION COVER SHEET 1

Galculation Number> --.-.- TDI-600797

~echnical Document Rev. No. 4 Addenda No.: NIA I

Calculation

Title:

Vortex, Air Ingestion & Void Fraction -.

Point Beach Nuclear Plant - Unit - 'I & 2 Safety Related?

YES Calculation Veriication Method (Check One):

.jX1 Design Review Scope of Revision:

0 Alteynate Calculation Qualification Testing Specific revision to address operating temperature range for voiding and updated air Ingestion and Froude calculations. Revision 4, Pages: All.

1

~ocurnentation of Reviews and Approvals:

Page I of ZZ I

Vartex, Air lngestion &Void Fr;iction - Point Beach Nuclear Plant -

1 Unit- 'l a2 Technical Document No. TDI-6007-07 Revision 4 I j

CALGULATION VERIFICATION CHECKLIST

---

........ - Gaiculation-Tiffe'-~-~V~it'ex~Air-.lngestion.&Vo~d~Frsl~ion

- Point Beach Nuclear Plant -- --.

-+. -.

U&i-1&2 I

No u n a n o n CIa Yes El El Revision: 4 nla U ' U CI'n m a m,

Kl a

cl

. 0 I.

iJ CIlZ

-n--B-CHECKLIST Were inputs correctly selected and incorporated?

Has U19 appropriate ~alculati&~

~uideline ~erifiktion Checklist been reikwed

'O*

and signed?

8 0'0.1 Note: This is Pel form 3060-3 Revision$

I..

I IZI a-

.a' j

I

6.

7.

8.

B.

~ a v e the design jntertace requirements been satisfied?

Was an appropriate design method used?

1s the output reasonable compared to input?

Are specified parts, equipment, and processes suitable for the required ap'plication?.

2.

3.

4.

,5.

U.

C ] m ( X I a

UEI Q..,a

-m-Are assumptions adequately described and reasonable?

Are the appropriate quality and quasi assurance requirements specified?

At6 the applicable codes, standards and regulatory requirements identified and met?

Have applicable construction and operating experience been considered?

14;

16.....

17.

18.

Has the design properly considered radiation exposure?

re the aaeptanoe criteria incorporated in the design documents sufficient p '

allow. verification?

Have, adequate pieoperational and subsequent periodic test requirements been

.specified?

Are adeq"& handling storage, cleaning and shipping requirements specified?

Are adequak identification require6nts speciiied7 Ark'requirements for r v r d preparation, review, approval, retention, etc.,

i o + ~ ~ t ' ~ i f i e d m ~ ~ 1 ~ ~ a t i b 1 ~ ~ ~ ~ 1 ~ r i ~ i l ~ n d i t i a BI mJ.-

adeiquately specified?

_."...I,.

. 11.

12.

13.

Have 'adequate main~~a~ature~.,a~drequirem~!~ts.ke~_n_

specified?.... -... - -

Are accessibility and other design provision adequate?

Has adequate accessibility been provided to perform the in-s=nrice inspection?

Vortex, f i r Ingestion & void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 TABLE OF CONTENTS CALCULATION VERlFiCATION CHECKLIST TABLE OF CONTENTS 9,O Purpose and Summary Results 2.0 Definitions and ~enninology 3.0 Facts and Assumptions 4.0 Design Inputs 6.0 Acceptance Criteria I

7.0 Calculation(s) n-vmex 7.2 Air ingestion

- -... - -.-7-;3.-

,-Void. Fraction------ --.-. - - -.. --.-------.

9.0 References ATTA;l$MENTS None TABLES Table 1 ~eklts Summary.

Table 2 ~lastiing Margin-For Operating Temperature Range Table 3 Calculation Results Date ':/c&.

- I Originated By:

I R31-600747 VorWAir lngestiori Voki Frer$pn - Rev 4.dbc Page 3 of22

~ e ~ e 3 d @

'1..

1.0 Purpose and Summary Results I

Vortex, Air Ingestion 8i Void Fraction - Point Beach Nuclear Plant -

Unit-1 &2 Technical Document No. TD1-6007-07 Revision 4 The US Nuclear Regulatory-Commission (USNRC) in generic safety issue (GS1)-.-..- -. -....

191 identified it' was possible that debris in PWR c6ntainments could be transported to the emergency core cooling system (ECCS) sump(s) foliowing a main steam line break (MSLB) and/or a loss of coolant accident (LOCA). It was further determined that the transported debris could possibly clog the sump screens/strainers and impair the flow of water, thus directly affecting the resultant operability of the various ECCS pumps and the containment spray (CS) system pumps, and their ability to meet their design basis %nction(s).

In order tq address and resolve the various issues identified by the USNRC in GSI-191, utilities have implemented a program of replacing the existing ECCS sump screens or strainers with new and improved designs.

1 In order to address and resolve the specific issues associated with USNRC GSI-191 for the Point Beach Nuclear Plant - Unit 1 & 2 (PBNP-IR), Point Beach I entered into a contract with Performance Contracting, Inc. (PCI). The primary I

objective of the contract was for PC1 to provide a qualified sure-FIO@ Suction I Strainer that has been specifically designed for PBNP-112 in order to address and resolve the NRC GSI-191 ECCS sump clogging issue.

PC1 has prepared a Qualification Report specifically for the subject strainer. The

' Q c i a l i i x t t i o n - R e p o r t - i s - a - w m p i l a t i o l that support ae drainer qualirfimtion. [t also provides a " ~ i n g f e - s ~ u r ~ ~

historical

...-....-.. recard-that can-be utilized to address any. PBNR-112-organizational or. NRC- --...... -

regulatory issues or questions associated with the replacement PC1 sure-low@ I Suction Strainer.

As part of the PBNP-IR Qualification Report, PC1 has perfom@ a number of hydgulic calculations in support of the replacement $upe-~lov$ Suction Strainer. 1 This calculation TDI-6007-07, Vortex, Air lngesfion & Void Fracficln - Poinf Beach Nuclear'Planf. - Unit - ? & 2 is one of a number of hydraulic calculations that specifically supports the design and qualification of the subject strainer.

his calculation addresses the various issues associaied with the separate but related issues associated with vortex, air ingestion, and void fraction as they relate to the sump and strainer.assembly that has been designed specifically for PBNP-112.

7 he P B N P ' U ~ ~

each have f.wo (2) separate iecirculatioin strainer assemblies that individually and specifically feeds either the 'A' or 'B' t d n ECCS and CS system.

Each of fhe horizontally oriented recirculation strainer assembly is C-Originated 'By TDIBOD7-07 Vortex Air IngesP~n Void Fradfon - Rev 4.doc

mPCI Vortex, Air Ingestion &Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 PE'RFOI?MA~~CE Technical Document No. TD1-6007-07 C.ONIIZACIING INC Revision 4 compcrised of fourteen (14) modules each made up of ten (ID) strainer disks for a total strainer area of 1,904.6 ff, or a total of 3,809.2 f f for each pair of strainers I

..-..-.....associated with one of the PBNP units; - Flow leaves the shiners and enters a-- --..--

combination of pipe and fittings before discharging into the containment outlet.

PC1 drawings tDrawings j0.j - j0.9, indusivej provide details of the subject configuration.

The results of the calcuiation are provided in Table I. The calculation utilizes the Acceptance Criteria established in both PBNP-lI2 and USNRC documents with resp$d to PWR sump performance b specifically evaluate the PBNP-'I& Sure-Flow Suction Strainer assembly.

1 ACCEPTABLE - Voids will not occur at the It was concluded that this c81culation, an integral portion of the Qualification Report completely supparts the qualification, installatich, and use of the PC1

~ u r e - ~ l o w ~

Sudon Strainer for Point Beach Nuclear Plant - Unit - I

& 2 without I any issues or reservations.

I Originated By:

Date'. 3/6'L7 '

Void Fradi~n - Rev Woo Page5of22 1

1.;.

.. ?.!. ',

MPCI Vortex, Air Ingestion & Void Fraction - Point.Beach Nuclear Plant -

Unit-I 6 2 PERFORMANCE Technical Document No. TO1-6007-07 CON1 lu),CIING INC Revision 4 2.0 Definitions and Termimology I,

The-following- ~efinitions

. &l~errninology are-defined-and-described. as -they arel.-

utilized in this calculation.

sure-lo@ Suction Strainer - Strainer developed and designed by Performance Contracting, Inc, that employs SUII%-FIOW@ technology to reduce inlet approach velocity.

Emergency Core Cooling System (ECCS) - The ECCS is a combination bf pumps, piping, and heat exchangers that can be combined in various configurations to provide ether safety injection or decay heat cooling to the reactor.

Clean Strainer Head Loss; (CSHC) - Is the calculated head loss for the Sure-FIO@ Suction Strainer based on actual testing performed at the Electric Power [

Research Institute's (EPRI) Charlotte NDE Center, and Fairbanks Pump Company Hydraulic Laboratory. The later testing did not involve any debris.

Point Reach Nuclear Plant Unit I & 2 - alsi known as Point Beach, PBN&1R, and PB-IR.

---.- --- Containment $pray-system-&also. known.as-~~~-

or-CS.... Systemis.utilized. to..- -----....

address either a MSLB 0r.a LOCA.

Loss-Of-Coolaht-Accident - also known as a LOCA. A LOCA is the re,sult of a pipe break or inadvertent leak that resutts in the discharge of primary reactor coolant from the normal huclear steam supply system (NSSS) boundary. A LOCA can be classified as a large break LOCA (LBLOCA) or a small break LOCA (SBLOCA). Classification is directly dependent upon the nominal size of the affected pipethat is associated with the LOCA.

I

,3.0 Facts and Assumptions I

.. - The fillowing Facts (designated as [FJ) & Assurnption~ (designated.as [~j)wire utilized in the preparation of this calculatidm.

3.1 A flow velocity of 0.00'26 fps would be characteristic of'the PBNPIIZ strainer, through a debris bed consisting of fibers and particulate, is 100%

viscous flow.

Acytdiiigly, the head loss 'is linearly proportional to dynamic viscosity [A].

$$+..;.

.... 'Originated By:

Date b

TDlaDD7-07 Vortex Air Ingestion VdM Fraction -Rev 4.dm

Vortex, Air ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. 7731-6007-07 Revision 4 3.2 A scale strainer, which is designed to maintain the same approach velocity I

.-..-..... ---as. the full -scale-production strainerr can. accurately-simulate the performance of the full scale production strainer so long as the same scaling factor is used for strainer area, water flow rate, and debris quantities. The scaling factor is defined as ratio of the surface area of the scale strainer to the surface area of the full scale production strainer [A].

I 3.3 The head loss resuiting from flow through a fiber - particulate debris bed at the approach velocity for the PBNP-I12 strainer (0.0026 Ws)

[Reference 8.61, is 100% viscous flow, as opposed to inertial flow. As viscous flow, head loss is linearly dependent on the product of viscosity I and velocity. Therefore, to adjust the measured head loss across a debris.

bed with colder water, a ratio of water viscosities, between the wamler specified post-LOCA water temperature and the colder test temperature, can be multiplied by the measured head loss to obtain a prediction of the head loss with water at the specified post-LOCA temperature [A].

I 4.0 Design lnpufs I

The following combination of Point Beach and PC1 Design lnputs were utilized in 1 the preparation of this calculation.

4.1 Point Beach Nuclear Plant Specification, Specification No, PB-681,

..... -........-.Replacement..of_Cs.otairtmenf -SurTi1!..~Sc_@_43:n_s~*

.Fie$_$i.W...

2,. Feknray. 17L...."---.- -..---.

2006 [Reference 9.11, document provides design input associated with strainer flow rate, water temperature, and the maximum allowable head loss.

4.2 Peflormance Contracting, Inc. (PCI) Calculation ~01-6007-02, SFS Surface Area, How and Volume Calculafion, Revision' 2 [Reference 9.1 21, 1 document provides relevant dimensions and other information specifically associated with the PBNP-I12 strainers.

4.3 PC1 Calculation TDI-6007-03, Core Tube Design - Point Beach Nuclear Planf - IL?, Revision 0 [Reference 9.61, document provides relevant data I with'regard to flow rate in the PBNP-If2 strainer.

4 4 PC1 Cabulatidn.TD16007-05, Clean Head Loss -*Point Beach ~uclear.

Plant - 14, Revision 4 [Reference 9.a, document provides the head loss associated with the "cleann PBNP-IE strainer and attached pipe and fMngs.

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 4.6 Point Beach Nuclear Plant, NPL 2009-0027 - Design Information Transmittal in Support of Calculation TDI-6007-07 Rev. 4, dated February 13, 2009 [Reference 9.47J, document provides pressure information for addressing voiding in the Point Beach strainer suction lines.

4.5 PC1 Calculation TDI-6007-06, Total Head Loss - Point Beach Nuclear

-..-.. -- -.. Plant.- fBi Revision 5 [Reference. 9;2[6], document provides-the total head loss associated with the PBNP-IR strainer and attached pipe and fittings.

5.0 Methodology PC1 utilized classical hydraulic calculations (conventional, calculation methodology) to address the subjed issues.

PC1 recognizes that if it is determined that one of the issues cannot occur andlor can be prevented, then one or more of the other bsues cannot occur (e.g., if a vortex is not predicted by I

calculation then there should be no air ingestion).

However, PC1 has '

conservatively assumed that each issue is separate, and each issue will be addressed on 'ts own merits.

This specific calculation

-.,-.....---. -- addresses

. three (3) separate but......

related

..... issues

vortex, air, ingestion and void fraction. ' i&o$inG&

eacti issue has ~t own separate acceptance criterion. The final ovehll acceptance criterion is that the PBNP-112 ECCS pumps have adequate NPSH margin under all postulated post-LOCA conditioris.

vorbex The USNRC in RG 1.82 Revision-3 [Refemnce 9.41 has indicated that air ingestion can lead to ECCS pump degradation andlor failure..A vortex is a potentiat source of air ingestion. A vortex can be preventeq,due to various

,combinations of sump configuration and the addition of vortex suppressors in the sump.

The Acdptance Criieria for vortex is the complete elimination of occurrence.

I

~016007~7 Vo,rtex Air Ingesffon Void Fkdion -.Rev 4.doc Page 8 of 22

'1 I

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-1&2 Technical Document No. TDI-6007-07 Revision 4 Air Ingestion RG I

.82-Revision 3 [Reference 9.41 states that air ingestion can lead to.ECCS....

pump degradation andlor failure if air ingestion is 3%. Accordingly, the USNRC has recommended a limit of 2% by volume limit on sump air ingestion. In addition, the USNRC has also recommended that even with air ingestion levels at 2% or less, NPSH can still be affected.

The USNRC has further recommended that if air ingestion is indicated, that the NPSH be corrected from the pump curves.

The Acceptance Criteria for air ingestion is 5 2%.

Void Fraction USNRC GSI-197 Safety Evaluation (SE) [Reference 9.31 has indicated that EGGS pumps can experience cavitation problems when inlet void fraction exceeds approximately 3%.

The Acceptance Criteria for void fiaction is 53% in conjunction with an acceptable sump pool temperature operating range as specified in Attachment V-I of [Reference 9.51.

1 in "order t5 'add;i&s ' arid' det6ri;iiiRe' the acc&ptdbilii-amd/or' issues potentially---- ".

associated with the three (3) separate but related issues of vortex, air ingestion and void fraction, a separate analysis of each issue was performed.

I 7.1 Vortex The PBNP-112 specification [Reference 9.11, specifically sections 3.6.12 and 4.1 address strainer vortex, but do not provide limitations on the new strainer design that specifically prohibits the formation of a vortex (i.e., n.0 vortex allowed). Accordingly, PCl.has utilized the guidance of USNRC RE 1.82, Revision 3 meference 9.41 to address the vortex issue for the PBNP-1/2 strainers.

In [Reference 9.43, the USNRC provided generic guidance with respect to

. PWR 'sump pixformandel sump design, and vorfex suppression. The subject reference can be utilized as a means of assessing sump hydraulic peiforma?ce, specifically the issues associated with a potential vortex in the sump.

r

Vorbx, Air Ingesiliion ek Wald Fmclficili - Point Beach Paucidar iPlis~~f -

Unit-i 8 2 i Technical Document No. TDI-FOO7-07 Revlsbn 4 I pi3~j-~g;ge~

$it hg~5-g ag-Uiiip ran b o l l ~ s - liodt~LOCA.vVatw B.Buppoitit - -.....-.

EGGS and CS functions.

Instead, the PBNP-112 ufilizes h o (2).

containment outlet penetrations located in the floor of the containnienf that are "dowered" by a vertical oval cross-section structurci. The striucfure consists,of an outer "coarse" screen (composed of a combination of 24" OD by %I' wall pipe and l/z" plate) witli slotted openings to facilitate post-LOCA water flow to the pumps. hiside of this structure are twc, (2) vertical "finen screen cylinders (one for each cbnfainment outlet) that are 13-1/2lJ ID. The 'Vinen screen cylinders preclude smaller particles and debris from entering the pumps [Qaefep.emtce 9.8 - %I4 incIua5we].

Since the PBNF-112 containment outlet structure is being modified by the

&difioq of the PC! sure-low@' suction strainer, the guidance 0ffei'e;tl by I the Us.wRC in [R(~9ference 9.43 is not entj'rely or ~FjecifiMlly' applimble, However, the guidance does provi$e some infonna~.on that can be utilized in. the $assessmsnt of the PBNP-112, stmiger confi@kratign with.r&gard to voFtex issues.

The "revised" PBNP-'~)~ strainer configuration will utilize a paii of.

horizontally oriented, PC1 sure-FIOW@

Su@Bh. strainers' each 'consisang of I

-elev~n-~j-j&~f~i~er~pdu!~~~~7h~flo~-f1:0m~e.W~~inet~Ai~.~

through atta=jF~~-aTd--fi'fittWtF'th3 FIcistin~-cTnt~fnm%t~fl6~--

loqatetj in the wntqtnment Roar. The subject strainer discfiaigje pipe rjyill' ta&e the g@@

af tfie e~~tiE~-cont~inmen~~utlef sfiUifire ~@ffa~i,h~s-lO;l.-----,~~

- %Q.A i%j$usives1.

Tiie PC1 ~~ref"low@' suction strainer will be analyzed a! addressed with I reslj~ct to vortex issues.

P C ~

$UIP~-F~OW~

@a,ciiion $trainer fovr*=rl I*

The PC1 ~ure-~low@

sQction strairier for.PBNP-IR is comprised of-I (IY) w. h o r i ~ n I ofiented modirles each containing ten (10) disks. Th@

disks-are. 8 riominal 518" thick sand are separated 1" from each gdja.&nt.

disk.

The interior. d the disks contain rectangular wire stiffeners tor

' support, donfigured as, a "sandwichu.made up of fhhae (3) layer$ of wires -

7 gaiige, 8 gauge, and 7 gauge..Tho disk@ are comple!$l~ covered with peifbrited plate having 0.0b6" holes.

The end disk of a rnodub is separated approximately 5" from the end disk of the adjacent module.

The 5" space between adjacent modules is covered with a solid sheet

Vortex, Air lngestron & Void Fraction'L Point Beach Nuclear Plant -

Unit-1&2.

~echni&l Document No. TD1-6007-07 Revision 4 metal "collar." Each of the modules has cross-bracing on the two exterior vertical surfaces of each module.

Based on the design configuration of the PBNP-112 strainer assembly, the largest opening for water to enter into the sump is through the perforated plate 0.066" holes. The sire of the perforated plate holes by themselves would preclude the formation of a vortex. However, in the unlikely event that a series of "mini-vortices" combined in the interior of a disk to form a vortex, the combination of the wire stiffener "sandwichw and the small openings and passages that direct the flow of water to the strainer core tube would further preclude the formation of a vortex in either the core tube or the sump.

The USNRC in I[Rebrence 9.41, specifically Table A-6 guidance is provided with regard to vortex suppressors. The table specifies that standard 1.5" or deeper floor grating or its equivalent has the capability to suppress the formation of a vortex with at least 6" of submergence.

The design configuration of the PC1 sure-low@ suction strainer for [

PBNP-112 due to the close spacing of various strainer components and the small hole size of the perforated plate meets and/or exceeds the guidance found-in Table A-6. The PBNP-IR strainer does not meet the 6" 7

submergence r e q u i ~ n t - T h ~ n ~ g ~ r a t i 0 n f 6 i ~ P B N P ; 1 ' / y ~

2" of submergence to the top of the strainer assembly. Howeveri there is a submergence, level of approximately-I O.Sn of submergence to the top of the core tube. In addition the water flow would have to pass through more than 8" of combined perforated plate, wire stiffe~er "sandwichesb, and cross-bracing which would further preclude the formation of a vortex, The USNRC carried out a number of tests regarding vortex suppressors at the Alden Research Laboratory (ARL) to arrive at the information surnmaiized in Table A-6 of [Refhence 9-41. The PC1 sure-FIOW@ I suction strainer prototype for PBNP-IR was also tested at ARL under various conditions. During the testing' of the PBNP-112 prototype strainer even when partially uncovered, did not exhibit any characteristics associated w'& a vortex or vortex development. Also, test observations bf the. minimum water level above a full size PBNP-IM strainer module showed no evidence of vortexing *during testing [Reference 9.181..

It can therefore be concluded that the configuration of the PBNP-112 Sure-lo@ suction strainer will prevent the formation of vortex development I

46/,.,

Dab.

TRl-600747 Vortex Air lngestlon old ~radlon - Rev 4. d ~ ~

Pagell'of22

'1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 Air Ingesaon The PBNP4R specification-[Reference 9.11, specifically section 3.6~17--..... -..

addresses air ingestion, but does not provide limitations on the new strainer design that limits air ingestion to a specific value (i.e., ~2%).

Accordingly, PC1 has utilized the guidance of USNRC RG I.82, Revision 3

[Reference! 9-41 to address air ingestion for-the PBNP-I12 strainers.

Appendix A and Table A-1 of [Reference 9.43 indicate that sump performance specifically related to air ingestion is a strong krnction of the Froude Mumber, Fr. By limiting the Froude Number to a maximum of 0.25, air ingestion can be maintained to ~2%.

The flow of post-LOCA water from a piping system associated with a LBLOCA or SBLFA, or a CS initiation associated with a MSLB or LOCA collects in the lower areas of the containment and eventually migrates to the ECCS sump. For the purposes of calculation, flow can be considered classified as open channel flow.

For open charinel flow, the Froude Number, Fr, is defined as the ratio between the force of inertia and the gtavitafional force [Reibrence 9.431. This can be expressed as follows:

Equation 1

~r 3 v I (g x L)"~

I I

g = gravitational constant, 32.2 ftls2.

I Where V = theVeliEiifWi3tWth-rBrgh a core t T i b T W t 7 '

For PBNP-I12 Vex = 3.478.W~ I[Wefeaence 9.61,

.?...-.-

L = the characteristic length L can be replaced by the hydraulic depth D defined as the ration of the cross-sectional area of the core tube divided by the width of the free surface (or the circuniferential slot width for the core tube hole velocity),

From the PBNP-112 Clean Head LOSS report [Reference 9.Q. A, = 1.344

.. fl? The PBNPrl12 Corq'Tube Design report [Reference 9,6] was used in Drawing 10.10 to calculated a slot width of 0.30 in, or 0.025 ft, for the hole' velocity of 3.4409.

Therefore, the ratio of the core tube area to the circumferential slot widthm.i$e calculated as follows:

i The most consenrative value that can be utilized for D is the case of the ratio of care tube cross-sectional area to the slot width for the first hole at the core tube exit, using the hole velocity calculated in Reference 9.6.

Vortex, Air Ingestion & Void Fraction - Point Beach Muclear Plant -

Unit-I 8 2 Technical Document No. TDI-6007-07 Revision 4 Accordingly, value of Fr can be calculated as follows.

I I

The calculated Froude Number for the PBNP-112 PC1 ~ure-FIO@ suction strainer is approximately 67% lower than the USNRC guidance found in

[Reference 9.43 of 0.25. The Froude Number decreases to 0.031 at the end of the strainer. Therefore due to the. combination of a low Froude Number and lack of an air entrainment mechanism (i.e., vortex formation) in conjunction with the complete submergence of the strainer, air ingestion is not expected to occur.

The PBNP-'IR specifidation mekrence gill, does not specifically address the issue of Void Fraction. It must be shown that flashing (i.e.,

voiding) does not occur anywhere within the strainer assembly throughout the operating sump temperature range. To demonstrate this, flashing will be evaluated across the screen itself and at the strainer assembly outlet.

It must also be shown that adequate pressure rem'ains available at the outlet of the strainer assembly to prevent flashing in the downstream SI-850 valve [Reference 9.1a.

Itcantheref-be ~ncliJ-d~d-t~at-th~PBNP~l~/2~strain'er~-wiII-have-air-ingestion of ~ 2 %.

....I...---.-..........................._._..,......I...

The pressure available to prevent flashing throughout the strainer assembly is.the sum,of the containment pressure and the pressure due to the sump water level less the dynamic losses. To prevent.flashing, the pressure available must exceed the vapor pressure of the sump water

[Reference 9.1111.

. Originated By:

Date *,

7.3 Void Fraction

Vortex, Air Ingesuon &Void Facb'on - Pdnt Beach Nuclear Plant -

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 Accordingly, PCI has utilked the guidance of Reference 9.2 and 9.151 to I address the void fraction issue for the PBNP-112 strainem.

. ---- --..... -......-..--.--.. --.--. I.......

.^_.._..-.

__,_"....,.._.._,,__I___^__

Although it is asserted in various regulatory documents that void formation is directly related to air ingestion, this is not correct. Void formation is fhe result of the pressure of a fluid being, reduced below the saturation pressure with the resulting voids being formed by the flashing of the liquid phase. Air does not need to be present to create significant voiding.

PC1 has evaluated the issue of Void Fraction for PBNP-I12 by the use of the following information provided by Point Beach [Reference 9.1u as input to hydraulic and fluid flow calculations to determine the PBNP-IR Void Fraction.

Calculation Methodology I

7.3.1 Evaluation of Flashing across the Strainer Screen [Reference 9d7J 1

Pressure Available at Screen

• Vapor Pressure I

i

=) P~lr f hapor + P~ubmer~ence - Pvelooity - A P ~ n e r

> Pvapor, then I

• AP~trainer < P Air

• P~ubrnetgence '

P~elocity I

PAW

= 12.7 psia (14.7 -2.0 psig) is the minimum containment air pressure allowed [PBNP TS 3.6.41, P~rrbme~ence

• 0 P S ~

Negligible since the minimum initial sump level provides 2" of submergence at the top of the strainer screens [PCI Drawings SFS-PBI-GA-00 & SFS-F?B243A-00], and Pv~t~~iiy - 0 psi Negligible since a flow velocity of 0.0026 fps is expected through the debris bed [PC1 Calculation TDI-6007-06 Rev. 51. A similarly small velocity is expected across the screens, then 3khy Date 1

I fTj16007-07 Vertex Air Ingestion Void Fhdion - R6v 4.doc Page 14 of 22,. '1

VortexI.Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-) & 2 Technical Document No. TDl-6007-07 Revision 4 PBNP-IR [Reference 9.13 defined the containment post-LOCA water temperature as being 212" F. The total debris laden head loss was 3.474 feet bf water [Reference 9.163, based on the 212" F water. Converting 3.474 feet of water equates to 1.44 psi.

APswiner

= 12.7 psid The maximum allowable pressure loss across the debris loaded screens to prevent flashing

--across the screen debris bed; The PBNP evaluation for strainer debris loaded differential pressure shows a maximum allowable of 12.7 psid. The 5.44 psi calculated by PC1 for the head loss across the strainer is less than 12% of the PBNP evaluated allowable differential pressure. Therefore no voiding across the strainer debris bed is expected.

I 7.3.2 Evaluation of Flashing at the Strainer Assembly Oratlet Pressure Available at Assembly Outlet > Vapor P~ssure I

P~ir Pvq~ar + P~ubnte~ence - Pverocity - W ~ n e r

> P~ipor., then I

Where,

.........-....-.-I---..

I P~ir

= 12.7 psia The minimum containment air pressure '

allowed (14.7-2.0 psig) 3.6.41.

I PSubmergence = I

.3 psi which is the minimum initial sump level provided by 38" of submergence at the strainer assembly outlet PC1 Drawings SFS-PBI-GA-00 & SFS-PB2-GA-OD], and l've~acity

= 0. I psia the dynamic velocity head ( ~ ~ / 2 ~ )

at 2200 gpk, velocity in the 18" elbow is less than 3.6 fps

[Crane 41 0Page B-14 and Eq. 1-31, then

.APstrainer = 13.9 psid is the maximum allowable pressure los's across the entire strainer assembly to ensure flashing does not occur at the assembly outlet.

Originated By:

i.

Vortex, Air Ingestion tk Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 Originated By:

Data 3 hL 7-Therefore, to assure that flashing does not occur at the strainer assembly outlet, the total head loss across the strainer assembly must be less than 113;gpsid-throughout.the operating-sump temperature:range; --.-.em.--.

The PBNP evaluation for strainer debris loaded differential pressure shows a maximum allowable of 13.9 psid. The 1.44 psi calculated by PC1 for the head loss across the strainer at 212'~ is less than 10.5% of the

'PBNP evaluated allowable differential pressure. Therefore no voiding across the strainer debris bed is expected.

7.3.3 Evaluation of Flashing near the Assembly Outlet with Sf-850 Valve P ~ s s u e Available at Assembly Outlet pressure Required fo prevent SI-850 Flashing

=;) p~lr + Pvapor -k p ~ u b m e r ~ c e

" P V B I ~ -

m s t r e n ~ r > ~ s I - ~ ~ o - ( ~ v ~ J ~ o ~ @ ~ I ~ F pvBWd, then, AP-iner

< PAP

• P~ubm@tgence '

p ~ e l o ~ - P ~ 1 - 8 5 0

+ P~epor@

212F

Where,

= 12.7 psia is the minimum containment air pressure PAir

. allowed (14.7-2.0 psig) PS 3.6.41,

- -. - -. -..---I__-

..I.

.----_I.- -------

--..I__

Psubmeme"ca = 1.3 psi which is the minimum initial sump level provided by 38" of submergence at the strainer assembly outlet $CI Drawings SFS-PBI-GA-00 & SFS-PBZ-GA-OO],

pve~~city

= 0.1 psia the velocity head (V2/2g) at 2200 gprn, velocity in the 'l8" elbow is less than 3.6 fps [Crane 410,

Page B-14, and Eq: 7-31, PSM~I)

= 20.9 psia which is 4.2 psig as required at the SI-850 valve 'assembly at 212.OF to prevent flashing

[PBNP Cab N-92-086 Rev. 43, and to assure no flashing, 2 psi is added to the predi'ded value [SER 2006-0003, PBNP Calc N-92-086 Rev. 41, and I

TDIGD0747 Vortex Air IngesHon Void Fraction - Rev4.doc Page 16 of22 1

HPCI Vortex, Air Ingestion 8 Void Fraction - Point Beach Meaclear Plant -

Unit-'I & 2 PERF~RMANC~

Technical Document No..TDI-6007-07 C ~ N I I ~ G I I N L,

INC Revision 4

, z Originahd By:

PVEpor@2q2F

= 14.7 psia A sump water vapor pressure of 212 "F is required to account for temperature dependent changes in-S1-850- flash suppression pressure requirements and the vapor pressure, then APstta~ner

= 7.7 psid is the maximum allowable pressure loss across the entire strainer assembly to ensure flashing does not occur in the S1-850 valve assembly.

Point: Beach fuFther adds that a downstream valve in the strainer sudion pathway [References 9.13 and 9.471: may cause additional flashing due to resistance and dynamic pressure changes in the valve. To address flashing at the valve (SI-850),

a total head loss across the strainer must not exceed 7.7 psid throughout the operating sump temperature range.

Table 2 has been generated to document the strainer head loss performance against the varying temperature dependent voiding limits.

Date

?/~/n 7 TDl-600797 Vwtex ASr Ingestion Void Fraction -Rev 4.doc Page 17of22 1

!,!.;?;,::

. j,:. : :.

. ::j....:..f;;;,i;j t..,..

.?,.i...

I

. :.'h;l;,;;.;*!,..$;

l,;;;,; !,?; ;:c

...: :;!,.2?,::+.::..

..;

:,I.,>

...... <-~'s~':.:;.4's~;:&:..'?.:!:i..:L"i.,!::t::,:y,,;.'i

..,..'. &q~".-:'.-'.l.?.'..... ':........... ". "

4. :z.;,,:;,>

2;.

5 ~..-'...:..;.,!fi.~~~

31,.!..).,:..... :;?;::,.

..,.,..,.r2'2'8.;ygc::-;.~:;$~i?~iii~~k~:

ti:;....a,.i::.;:j.;,<;:j;$,:,*;;;, :,:.. ?;.:.. :.. i!

~*~:~~:.~-:"9,,:!.%:<

. > ~ ~ ~ ~ ~, ~ ~ ; ~ : ~, :, r, ~ ; ~ ~

i.

:.:;.

,,,,+?..

,,xi

>,,,.:; 2.,!

1,. ?

I I

I

.I.-.- -...-...-

1 I I I I I

I I ' '

1 Table 2 - Flashing Margin For Clpekiting Temperature IRangcm.

Based on the temperature range data presented in Table 2, head loss in the PC1 strainers should not allow flashing anywhere within the strainer assembly or in the SI-850 valve throughout the operating range until temperature is reduced below 52 O F.

Flashing Margin (7.7 psid

-Allowable)-

6.26 -

.... 6-06' -......-.

5.83.

Total Coreded Head Density "F

21 2 psi Equivalent 1.44 t 152 132 112 92 72

. 52'

1g2- ".".

...3,g3-.

60.31...-

.. - 1.64 2.16 I

5.54

5.,7..

2.53 5.08 5.93 32 1 20.22 1

' 62.42 61.16 61.52 8.76 I.

,406 172 7.07 1

61.83 3.03 1

4.67 8.64 10.93 14.42 r

- 4143

1.

60.75.

62.09 I

3.72 1

3.98 1.87' 62.29 4.73 2.97 62.41' 4.45

MPCI Vortex, Air Ingestion & Void Fraction - Pdnt Beach Nuclear Plant -

Unit-I & 2 PERFORMANCE Technical Document No. TDI-6007-07 CONIIWCIINC INC Revision 4 8.0 Conclusions I

. - The-result of this-calculationr specifically-the acceptability of..the. issues....... --..-...- -

associated with vortex, air ingestion, and void fraction are summarized in Table

3.

I It was concluded that the subject issues have been addressed for PBNP-If2 and the results indicate that there are no vortex, air ingestion or void fraction issues with the installation of the PCi ~ure-~lod@

Suction Strainers. This specific calculatidn completely supports the qualification, installation, and use of the PC1 sure-lo@ Suction Strainer for Point Beach Nuclear Plant - Unit I

& 2 without any issues or reservations.

Originated By:

I T ~ I E B o ~ ~ T - ~ ~

VOM Air lngesfioli VO~!

imdion - Rev4.d~~

Page 1Bof22 I

. leeue vortex AIr lngestlan Vold Fractlon 8

'1 USNRC None (Ref:

I USNRC Fraction - Point Beach Nuclear Plant - U.nit - I[&

2 Technical Document No. TD1-6007-07 Revision 4 I

I 081-I81 Safaty 1 EvaluaUon I

I

~ a b ~ e 3 - tja~lcu~jation

~ e s u l t s detrimental '

effects on RHR, ll &

cs pumps Comments Results applicable to the PBNP-112 sure-FIO#

oe Crlterla PBNP-112 N~A NO I

~ e i u l t s

'ACCEPTABLE No Vortex - vortex formation Is

precluded by ihq PC!

sure-FIOW@'

Stralner design and

~~~~ratl,.,n I I

N'O 1 ACCEPTABLE !

2&%%' ' & ~ f ~ ~ t ~ O % w Y ~ ~ l o 2 ~ ~

$~~~~~

RHR, SI has been detem, e,j that vortex.

CS Pumps : tomalion not paur ien can be reasonably conolu(led thdt alr lngesllon stialner.

'1 Per RG 1.82, Revlsbn 3, If alr lnaestibn is o%,

the pump NPSH must be corrected by the relationship, NPSHmum I

-1

= NPFH

~aq~~md@iuld~X PI where P =I + 0. 5 0 ~ ~

and % 1s the a t lngertlon rate (In percent by volume) at the pump Inlet l,will also not occur. i decreasing throughout range.

1 NIA

!ACCEPTABLE b

Void FraoUon wiii not at the

' strainer - ca'cufslion lbdlcales OW'

..AddlHonally* the:. cat~ulauOn also.

I

,concludes that voids will not occur In the SI-850 valve with1 temperature Originated By:

i Gonventlonel. calculation methodology indidfes that no vold fraction will occur at the sfmiher.

The pressure Is sufiiclent to prevent voiding at the 61-850 vel"e through 52 OF.

Cavit~Uon at the v;ilve assembly may occur at colder

~empeiature~

for design flow cond,tjons, l'DlBO07-07 Voflex Alr IngesUan Vold Fracfion -Rev ldoo Page 19 (? 22 I

Vortex, Air Ingestion &Void Fraction - Point Beach Nuclear Plant -.

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 I 9.0 References I

9.1 Point-Beach Nuclear Plant Specification, Specification No, PB-681,..I -

Replacement of Containment Sump Screens, Revision I, August 25,2005 9.2 Information Systems Laboratories (ISL), lnc., ~ e ~ o i ISL-NSAD-TR-05-01, Development and ltktplernenfatidn of an Algorithm br Void Fmction Calculation in the "6224 Correl~fion" S o h a p Package, January 2005, prepared for the USNRC 9.3 U.S. Nuclear Regulatory Cammission, Safety Evaluation, Pmssuritad Water Reactor Sump Performance Evaluation Methodology, Guidance Report of the Nuclear Energy Institute (NEI), GSI-I91 SE, Revision 0, dated December 6,2004 9.4 U.S. Nuclear Regulatory Commission, Regulatory Guide 1 32, Wafer Soumes for Long-Term Recirr=ulafion Cooling Following a Loss-oFCoolant Accident, Revision 3, dated November 2003 9.5 U.S. Nuclear Regulatory Commission, GSI-191 SE, Affachment V-7, NUREG/CR-6224 Hsad Loss Tempemture Assessment. Revision 0, December 2004 9.6 Performance Contracting, Inc. (PCI), Technical Document Number, TDI-.

6007-03,- Con! Tube Desigri..-. Point Beach.Nuclear. Plant z Un'%...1/2,'----.

Revision 0 I

9.7 performance Contracting, Inc. (PC!), Technical Document Number, 'TDI-6007-05, Clean Head Loss - Point Beach Nuclear Plant - Unif 7/2, Revision 4 I

Bechfel, Job No. 6118, Drawing No. M-276, Revision 2, Point Beach Nuclear Plant Unit I & 2, Containment SaMy Injection Sump.

Requimmenfs for Screens Bechtel, Job No. 6118, Drawing No. C-126, Revision 7, Point Beach Nuclear Plant Unit 1 & 2, Uner Plate - Flaor Plan Bechtc$, Job No. 6118, Drawing No. C-128, Revision 9, Point Beach.

Nuckar plant Unit I

& 2, ~ontainment Sbucfure Interior Plans at El. 7@-0, El. 21'-0, El. 24-8 & 38'-Q I

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I &2 Technical Document No. TD1-6007-07 Revision 4 I

9.1 1 Not Used 1

.---.... 9.12 PCI; ~echnical'Document NumberrTDI-6007-02-, SFS Surface &a, Flow..... --......

and Volume Calculation, Revision 2 I

9.13 Nazeer, Ahmed, Fluid Mechanics, Engineering Press, Inc., 1987 0.14 Not Used I

9.15 USNRC, 6224 Cornlation, publicly available soRware I

9.16 PC1 Calculation TDI-6007-06, Total Head Loss - Poinf Beach Nuclear Plant - I&, Revision 5 I

9.17 Point Beach' Nuclear Plant, NPL 2009-0027 - Design Information Transmittal in Support of Calculation TDI-6007-07 Rev. 4, dated February 13,2009 9.18 EC-PC)-PB-6028-1001, AREVA Document No. 66-9093957-002, Point Beach Test Report for ECCS Strainer Performance Testing. Dated 11/26/2008 Originated By:

'TD16007-07 Vow Air Ingestion Void Fradion - Rev 4.doc

Pa~?21of22 I

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I 8 2 Technical Document No. TDI-6007-07 Revision 4 1

10.0 Drawings I

10.1.- SFS-PBI-GA-00, Revision 6, Point Beach Unit 1, sure-~low@~frainer,- I Recirc Sump Sysfem 10.2 SFS-PBI-GA-02, Revision 6, Point Beach u ~ R 1,; SUMOW@

stainer, A I Sfrainer 10.3 SFS-PBI-GA-04, Revision 5, Point Beach Unit I, SUE-low@ strainer, I

Piping A Layouf 10.4 SFS-PBI-GA-05, Revision 9, Point Beach Unit 1, S U ~ F I O W @

strainer, I

Piping B Layout 10.5 SFS-PBI-PA-7100. Revision 2, Point Beach Unit 1, ~ure-~low@

sfminer, I

Module Assembly 10.6 SFS-~82-G~-06, Revision 2. Point Beach unit 2, ' ~ u m - ~ ~ o w @ ~ f r a i n e ~

I Rech Sump Sysfem 10.7 SFS-PB2-GA-02, Revision 9, Point Beach Unit 2, ~un+~low@

~tmlner, A

]

Strzjriner 10.8 SF++B2-GA-04, Revision 5, Point Beach Unit 2, SUE-F~OW@

strainer, I

-.. -P@ing..Assembly: Layout-..-..

10.9 S F S - P B ~ - ~ I 00, Revision I, Point Beach Unit 2, S U ~ F I O @

Sfminer,

)

Module Assembly I

Originated By:

Date Tqf.600747 VoItexAir lngesbn Void Fraction -Rev #.doc Page220f22

, I,

ENCLOSURE 6 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-1911GL 2004-02 (TAC NOS. MC470514706)

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED WATER REACTORS AREVA CALCULATION 51 -9056525, REVISION 001,412812008.

CHEMICAL PRECIPITATION ANALYSIS FOR POINT BEACH NUCLEAR PLANT USING WCAP-I 6530-NP 45 pages follow

AREVA March 16,2009 AREVA-09-01215 Tom Kendall, PE Sr. Technical Advisor Design Engineering Point Beach Nuclear Plant Mr. ICendall, AREVA perfo~med GSI-191 Cheinical Effects Calculations for Point Beach Nuclear Plant. The deliverables were AREVA Engineering Information Record (EIR), document numbers:

5 1-9010780-00 1 and 51-9056525-001. These documents were incorrectly stamped as proprietary.

No AREVA intellectual rights or trade secret were found after completing the review of these documents, Therefore, these documents can be status as non-proprietary for use by Point Beach Nuclear Plant.

Please feel fiee to contact me if you have my questions or requests regarding this matter.

Sincerely, Ray Phan "

Manager 1 BOP System Engineering Oflice: 704-805-2231 Mobile: 704-5754924 AREVA NP INC.

A n AREVA nnd stamens campony 7207 IEM Drlve, Charlone. NC 28262 Tel.: 704 806 2000 - Fax: 704 806 2800 - w.arevaoom

ENGINEERING INFORMATION RECORD

. Documentldentlfler 51 - 9056525 - 001 Title Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP PREPARED BY:

REVIEWED BY:

NAME H. Dergel NAME R. Jetton Signature &-

Technical Manager Statement: Initials Reviewer is Independent.

Signature J~stbn Date 4.]~7!67

k.

kmarks:

1 The purpose of this document is to determine the type(s) and bounding quantities of chemical precipitates expected to form in the containment sump pool following a Design Basis Loss-of-Coolant-Accident (LOG?!), when generated I debris or other susceptible materials may be subject to acid.or caustic fluids. This evaluation has been performed I based upon plant-specific design parameters primarily using the guidance published within WCAP-16530-NP and the associated Chemical Model Spreadsheet. Sensitivity analyses were performed to investigate the effects of varying design input parameters, as well as applying specific reduction tactics directed within WCAP-16785-P.

This evaluation is required to understand the evolution of the chemical environment present inside the Unit 1 and 2 Point Beach Nuclear Plant (PBNP) reactor containment and containment sump pools following,a LOCA. The lresults of this evaluation may be used as inputs into the downstream effects evaluation or as chemical debris mixture inputs into sump strainer qualification testing for Point Beach, as results are used to direct the generation and subsequent introduction of chemical debris. This is a safety related evaluation.

This document, including the information contained herein, is the property of AREVA NP, Inc., an AREVA and Siemens Company (AREVA NP). It ~ontains proprietary information and, except for AREVA NP affiliated companies, may not be reproduced or copied in whole or in part nor may it be furnished to others without the prior written permission of AREVA NP. In any case, any re-exportatiori of the document shall be subject to the prior written permission of AREVA NP. It may not be used in any way that is or may be injurious to AREVA NP. This document and any copies that may have been made must be returned upon request AREVA NP Inc., an AREVA and Siemens company Page 1 of 56

ENGINEERING JRlFORlVIATlON RECORD I. Documentldentitier 51 - 9056525 - 001 I

I Title Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP f

I PREPARED BY:

REVIEWED BY:

I I NAME NAME I

Signature Date q,/; Fw Signature Oate 9/27/67 I

Technical Manager Statement: Initials Reviewer is Independent.

Remarks:

The purpose of this document is to determine the type@) and bounding quantities of chemical precipitates expected I m form in the containment sump pool following a Design Basis Loss-of-Coolant-Accident (LOCA), when generated debris or other susceptible materials may be subject to acid or caustic fluids. This evaluation has been performed based upon plant-specific design parameters prlmarily using the guidance published within WCAP-16530-NP and the associated Chemical Model Spreadsheet. Sensitivity analyses were pet-Formed.to investigate the effects of varying design input parameters, as well as applying specific reduction tactics directed within WCAP-I 6785-P.

This evaluation is required to understand the evolution of the chemical environment present inside the Unit 1 and 2 Point Beach Nuclear Plant (PBNP) reactor containment and containment sump pools following a LOCA. The results of this evaluation may be used as inputs into the downstream effects evaluation or as chemical debris mixture inptits into sump strainer qualification testing for Point Beach, as results areused to direct the generation and subsequent introduction of chemical debris. This is a safety related evaluation.

I I

AREVA NP hc., an AREVA and Siemens company A REV& C O ~ ~ S ~ P O ~ ~ ~ A C S

%)Q~VA -@?*/z IS i

~ I Z O 1144flCH 16, ~ @ @ q Page I of 56

QF-0528 (FP-E-MOD-07) Rev. 0 Design Review Comment Form Cornmiffad Lo Fleet Modification ~ k c e s s Sheet J-of 1 DOCUMENT NUMBEW TITLE:

Calc 51-9056525 REVISION:

001 DATE: 9/27/2007 I

REVIEWERS COMMENTS This calculation revision (-001) is a result of an error found during the owner's acceptance review of revision -000.

The error was a transcription problem in the fable of the final results (table 6-1). While correcting the original revision, enhancements Were also incorporated into the portrayal of the various cases considered (table 5-1 was the result), and in tabulating the equivalent concentrations in table 6-1.

( This calculation has been reviewed and found to be correct in the following respects:

I I

I ) A sampling of the results have been corroborated by an independent check by the Owner's Reviewer by running the same spread sheet model, and I

1 2) The inputs have been verified to be correct per the verified inputs provided to AREVA.

I However, the calculation results must be applied judiciously, and with a thorough understanding of their derivation and the underlying assumptlons.

I This calculation does not calculate one single credible '%orst case7'scenario for the Point Beach units. Rather, it uses a matrix approach to illustrate sensitivities, and to explore the bounding envelope of potential chemical effects outcomes. Specific cautions for future users are itemized below.

1. Cases I.

'I, 1.2,2.1, and 2.2 should not be used as design bases inputs. These cases each assume that there is no sump mixing, even after sump recirculation is initiated. This is an unrealistic assumption, and is not widely used in Industry. The utility of these cases is to establish the differences in chemical generation between maximum sump levels (cases 1.1 and 2.1) and minimum sump levels (cases 1.2 and 2.2). Based on those results, it is clear that using a maximum sump level assumption will result in the maximum (bounding) quantity of chemical precipitant generation. All subsequent cases use an assumption of maximum sump level with a mixed sump.

2. Cases 1.6 and 2.6 should not be used as design bases inputs without substantial additional work. These cases credited the inhibition of aluminum corrosion due to the presence of silica in the sump water. While this may be a valid mechanism for inhibiting aluminum corrosion, it would first be necessary to ensure that all such breaks will result in sufficient silica to effectively inhibit the corrosion. Since this has not been done, use of the results from these runs is not appropriate.

3. Cases 1.1-1.6 and 2.1-2.6 use a "worst of the worsf' method for determining chemical contributors from the various debris sources. These are unit specific, and can be considered the bounding chemical inputs.

After eliminating cases 1.I, I

.2,2.1,2.2, I

.6, and 2.6 from consideration (see above), case 2.5 can be seen as the most limiting. Therefore, this case should be considered the limiting design basis case. It is important to recognize that this is a contrived case that assumes a contrived case that assumes a less-than-maximum-sized LOCA. This is evidenced by the prolonged duration of containment spray on injection. LOCAs smaller than this would not likely result in the actuation of containment spray, or in the securing of containment spray earlier in the event due to not having severe core damage or high containment pressure.

4. When using table 6-1, care should be taken to not use the concentrations listed. These concentrations were derived using the maximum sump volume to establish the total mass, but then divided that mass of chemical precipitants by the mass in the minimum sump volume (this approach is noted at the bottom of the page). This produces an erroneous and excessively high chemical concentration. If chemical concentrations are desired, then they must be calculated from the chemical masses listed in the table and then divided by the mass of the maximum sump level. Both can be obtained from within the calculation.

I

5. Appendices N.l and N.2 are break-specific runs that were used to assess whether application of the silica inhibition of aluminum carrosion could be credited. In all cases considered, it appears that silica I

Page 1 of 2

concentrations would be sufficiently high to invoke the WCAP guidance on silica inhibition. However, the evaluation did not consider all potential break locations. Additionally, silica inhibition effects were found to be mlnimal because most of the corrosion occurs during injection spray when there is no silica in the spray water. Therefore, as noted in #2 above, these runs do not provide a significant benefit, and have not been shown to be bounding.

6.

Electronic files of the input spreadsheets used for this calculation were part of the deliverables to PBNP from ARENA. After consideration of the delivered calculation, It was determined that additional information was desired. Specifically, the site needs to be able to demonstrate that replacing existing asbestos insuiatlon with other types of Insulation is acceptable, and that the chemical effects of such replacements are known and bounded by this analysis.

I As noted in OAR comment #3 above, case 2.5 is.the most limiting credible condition. Therefore, the spreadsheet for case 2.5 was altered into 3 supplemental cases:

I 2.5.1 : Replace all asbestos with CalSll I

2.5.2: Replace all asbestos with generic fiberglass I

2.5.3: Replace all asbestos with NUKON I

These runs were independently prepared and verified by qualified site personnel (signatures at the bottom of this form), and the inputs and results are attached to the vendor prepared calculation.

The results of the runs show that while the total amount of precipitate can increase due to insulatlon replacements, the effect is very small, even if 100% of the asbestos is replaced. The following table summarizes the results of the supplemental runs, and should be used when considering appropriafe qualification testing:

Case #

2.5 2.5.1 2.5.2 2.5.3 Tofal Ppt Mass (kg) 194.119 194.119 196.599 195.964 Total Al Mass (kg) 19.97 19.98 20.23 20.17

[All at max sump Ivl

( P P ~ )

16.3 16.3 16.5 16.5

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-I 6530-NP Document No. 51-9056525-001 Multiple PreparerlReviewer Signature Block Name (printed) 1 Signature I PIR I Date I

I I

Note: PIR designates Preparer (P) or Reviewer (R).

Prepared or Reviewed I Page 2 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Record of Revisions Page 3 of 56 Brief Description Initial Issue The following changes were made in this revision:

1) Revisions to all Tables

2) Revisions to all Sections to clarifjr report text

3) Changes to pH inputs for Cases 1.4 and 2.4 (Appendices E & K).

4) Changes in Appendix A pH Profiles and Descriptions

5) Appendix N Summary Format Revision 0

1 Date 8/24/2007 912712007 PageslSections Changed All All

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Table of Contents Page MULTIPLE PREPAREWREVIEWER SIGNATURE BLOCK 2

RECORD OF REVISIONS........................................................................................................................ 3 LIST OF TABLES 6

ABBREVIATIONS.....................................................................................................................................

7 1.0 PURPOSE.....................................................................................................................................

8

2.0 BACKGROUND

............................................................................................................................. 9 3.0 ASSUMPTIONS.......................................................................................................................... 12 4.0 CHEMICAL MODEL SPREADSHEET I 8 4.1 Chemical Model spreadsheet inputs 20 5.0 SENSITIVITY ANALYSES 20 5.1 Case Set 1aEa: Bounding Debris Inputs 22 5.2 Case Set l b12b: Debris Generation Case Inputs 23 6.0 RESULT

SUMMARY

................................................................................................................... 23 6.1 Case Set la12a: Bounding Debris Inputs..................................................................................... 2 5 6.2 Case Set I b12b: Debris Generation Case Inputs 27

7.0 CONCLUSION

29

8.0 REFERENCES

............................................................................................................................ 31 APPENDIX A : GENERIC TEST INPUTS............................................................................................... 33 APPENDIX B : CASE 1.1 BASE CASE: MAX PH. MAX SUMP VOLUME. UNMIXED 41 APPENDIX C : CASE 1.2 BASE CASE: MAX PH. MIN SUMP VOLUME. UNMIXED 42 APPENDIX D : CASE I 3 BASE CASE: MAX PH, MAX SUMP VOLUME, MIXED 43 APPENDIX E : CASE I

. 4.

SUPPLEMENTAL CASE: MAX PH. MAX SUMP VOLUME. MIXED. SUMP

. RECIRC @ 60 MINUTES. SPRAY RECIRC @ 100 MINUTES 44 APPENDIX F : CASE 1.5.

SUPPLEMENTAL CASE: MAX PH, MAX SUMP VOLUME, MIXED. SUMP RECIRC @ 120 MINUTES. SPRAY RECIRC @ 123 MINUTES 45 Page 4 of 56

Chemical Precipifation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 APPENDIX G : CASE 1.6 - SUPPLEMENTAL CASE: ADDITIONAL INPUT EVALUTIONS USING WCAP-16785-P...........................................................................................................................46 APPENDIX H : CASE 2.1 - BASE CASE: MAX PH, MAX SUMP VOLUME, UNMIXED

..47 APPENDIX I : CASE 2.2 - BASE CASE: MAX PHI MIN SUMP VOLUME, UNMIXED 48 APPENDIX J : CASE 2.3 - BASE CASE: MAX PHI MAX SUMP VOLUME, MIXED 4 3 APPENDIX K : CASE 2.4 - SUPPLEMENTAL CASE: MAX PH, MAX SUMP VOLUME, MIXED, SUMP RECIRC @ 60 MINUTES, SPRAY RECIRC @ 100 MINUTES 50 APPENDIX L : CASE 2.5 - SUPPLEMENTAL CASE: MAX PHI MAX SUMP VOLUME, MIXED, SUMP RECIRC @ 120 MINUTES, SPRAY RECIRC @ 123 MINUTES 51 APPENDIX M : CASE 2.6 - SUPPLEMENTAL CASE: ADDITIONAL INPUT EVALUTIONS USING WCAP-16785-P........................................................................................................................... 52 APPENDIX N : DEBRIS GENERATION INPUTS & ADDITIONAL SUPPLEMENTAL CASES:

ADDITIONAL INPUT EVALUTIONS USING WCAP-16785-P FOR ALL DEBRIS GENERATION CASES................................................................................................................................... -53

chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 List of Tables Table 5-1: Test Parameters - Units I & 2 Table 6-1: Test Outputs for Units 1 & 2 Table 6-2: Comparison Concentrations for Units I & 2 Table 7-1: Test Summary - Units 1 & 2 Page 2 1 24 28 29 Page 6 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Abbreviations CalSil CSS DIT ECCS GSI HELB LOCA NPSH NaOH PBNP PWR RCS RMI RPV RWST ZOI Calcium Silicate (insulation)

Core Spray System Design Information Transmittal Emergency Core Cooling System Generic Safety Issue High Energy Line Break Loss of Coolant Accident Net Positive Suction Head Sodium Hydroxide Point Beach Nuclear Plant Pressurized Water Reactor Reactor Coolant System Reflective Metal Insulation Reactor Pressure Vessel Refueling Water Storage Tank Zone of Influence Page 7 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-00.1 1.0 PURPOSE This evaluation discusses the inputs required to address the Nuclear Regulatory Commission (NRC) request for licensees to confirm their compliance with 10 CFR 50.46 (b)(5), as recently communicated in the NRC Generic Letter (GL 2004-02) titled "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors," dated September 13, 2004, as well as NEI 04-07, uPressurized Water Reactor Sump Performance Evaluation Methodology,"

Volumes I (Methodology) and 2 (Safety Evaluation), dated December 2004 [8].

The generic letter requires that licensees of Pressurized Water Reactors (PWR) perform mechanistic evaluations of their Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) based on the potential susceptibility of PWR recirculation sump screens to debris blockage during design basis accidents requiring recirc~ilation operation of ECCS or CSS, as well as on the potential for additional adverse downstream effects due to blockage of ECCS and CSS components and flow paths by debris which has bypassed the strainer. Debris blockage and subsequent flow restriction in the ECCS flow path could impede or prevent reactor coolant recirculation to the core, leading to inadequate core cooling and thus failing the requirements within 10CFR50.46. Regulatory Guide 1.82 has been revised to include evaluations of the concerns raised in the generic letter [2].

The results of these evaluations may be used to perform plant-specific strainer qualification testing.

These activities involve head loss testing of a strainer module or modules to validate that the emergency systems will operate properly and within design margins following a Design Basis LOCA when the screen and sump recirculation water is fouled with resultant failed or precipitated materials.

NEI 04-07 states that licensees must evaluate the sump screen head loss consequences with an integrated approach which includes both fragmented debris (i.e. insulation) which has been generated, as well as corrosion products which may develop or precipitate following a LOCA

[Reference 8 Vol. 2 Section 7.41. Licensees must also ensure that the chemical effects test parameters applied during plantspecific strainer qualification testing (quantities and types of materials) are sufficiently bounding for their plant-specific conditions in order to ensure that the chemical effects issue has been addressed to the satisfaction of the regulator.

As a step toward addressing GL 2004-02, this evaluation specifically addresses the chemical evolutions which occur in the presence of postulated as-generated debris or other susceptible materials, including additional submerged or un-submerged 0.e. wetted) materials, as subject to acid or caustic fluids and in proximity of the containment sump following a Design Basis Loss-of-Coolant-Accident (LOCA). Note that debris generation, debris transport, downstream effects issues, and head loss calculations in the presence of a debris bed are normally addressed in separate evaluations.

Page 8 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 The purpose of this document is to determine the type@) and bounding quantities of chemical precipitates expected to form in the containment sump pool following a Design Basis LOCA. This evaluation has been performed based upon plant-specific design parameters primarily using the guidance published within WCAP-16530-NP and the associated Chemical Model Spreadsheet [I].

Sensitivity analyses were also performed to investigate the effects of applying specific reduction tactics directed within WCAP-16785-P [3].

This evaluation is required to understand the evolution of the chemical environment present inside the Unit 1 and 2 PBNP reactor containment and containment sump pools following a LOCA. The results of this evaluation may be used as inputs into the downstream effects evaluation or as chemical debris mixture inputs into sump strainer qualification testing for Point Beach, as results are used to direct the generation and subsequent introduction of chemical debris. The results of this evaluation will be compared to the concentration used as debris mixture inputs into previous Point Beach Sump Strainer Performance Testing. This is a safely related evaluation.

2.0 BACKGROUND

During a postulated LOCA inside containment, piping and equipment insulation can be fragmented by the jet forces exerted by the high pressure steam/water from a postulated break, and fall to the containment floor from the area of the break as 'generated' debris. This mixed debris, specific to the each plant, may consist of fibrous material (from the failure of insulation such as NUKON, and Temp Mat), particulates (from the failure of materials such as coatings, and microporous insulation),

Reflective Mirror Insulation (RMI), and other miscellaneous debris types. This 'generated' debris will then mix with other latent and miscellaneous'fibrous and particulate debris that has already become loose in containment as the sump pool fills with break water.

Immediately following a large break LOCA, it is also expected that the Containment Spray System (CSS) will actuate to mitigate a pressure spike in containment due to heat input from the high temperature break. The RWST (Refueling Water Storage Tank) source water will mix with concentrated sodium hydroxide (NaOH) to exit the system into containment through spray headers and nozzles as a borated alkaline spray solution. Once injected, the elevated pH spray solution will directly impinge upon and corrode any exposed containment inventory; including equipment, structural surfaces or coatings. Any ions that are dissociated by corrosion from inventory surfaces are then assumed to reach the sump pool, and subsequently be in proximity as possible reactants toward the precipitation of chemical debris.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 When the Emergency Core Cooling System (ECCS) is actuated following a LOCA, the containment sump will supply water to support core cooling. In-containment barriers (sump strainers) are installed to prevent or hinder mixed debris from entering the ECCS. However, debris bed formation will occur on the sump screens, resulting in possible increases in head loss and damage to downstream components. Damage to downstream components could result from head loss increases at the containment sump strainer, as well as strainer debris bypass, as small debris potentially penetrates the sump screens and affects downstream components.

To address this ongoing concern regarding the GSI-191 related effect of chemical debris upon head losses at the sump strainers, this evaluation has been performed to assess the current PBNP Unit 1 and 2 designs and perform a full plant-specific evaluation of the chemical evolutions expected to occur due to material precipitation when generated debris or other susceptible materials are subject to acid or caustic fluid following a LOCA.

Recent work, directed by the Westinghouse Owner's Group (WOG), has sought to provide supplemental insight into the chemical processes that may occur in post-accident containment sump fluids by concentrating on more individual chemical reactions to ensure proper experimental control [I

1. This work used the results of the Integrated Chemical Effect Test (ICET) Projects to target the chemical reactions expected to generate the most precipitate, through the application of more simplified configurations of individual insulation types, buffer solutions, and post-accident temperatures [lo]. Specific materials and test parameters were selected based on plant-specific quantities reported and known reactivity characteristics of each material (see the following sections within Reference 1 for justification of elimination of the following materials: Zinc based materials -

Section 6.2.2, lron based materials - Section 6.2.3, Nickel and Copper based materials - Section 5.1.2, and organic materials (i.e. with respect to aluminum-based coatings)- Section 3.2).

This follow-up testing by Westinghouse was performed on individual representative containment materials, such as Aluminum, Concrete, Calcium Silicate (CalSil), Nukon Fiberglass, High Density Fiberglass, Mineral Wool, Min-K, Fiber Frax, Durablanket, Interam, Galvanized Steel, and Uncoated Carbon Steel. During the process, samples were taken of dissolved solutions and analyzed for the presence of Aluminum (Al), Calcium (Ca), Silicon (Si), Magnesium (Mg), Phosphorus (P), Sulfur (S),

lron (Fe), Zinc (Zn), and Titanium (Ti). It was shown that the total mass element release for aluminum, silicon, and calcium were the largest contributors to the dissolved solution, and that any precipitates would therefore most likely form of these elements [I].

Three specific chemical compounds were noted to precipitate during this testing dependent upon the debris mixture and test parameters [Reference 1 Section 6.1 1. The results of the WOG test program indicated that the predominant chemical precipitates, dependent upon plant buffer type Page 10 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 and the pH of the sump medium, were aluminum oxyhydroxide (AIOOH), sodium aluminum silicate (NaAISia08), and calcium phosphate (Ca3(P04)2) (the latter only identified in the presence of irisodium phosphate (TSP)) [Reference 1 Section 6.1 & 6.41. Other minor silicates could be precipitated. However, their concenfration is expected to be minimal with respect to the dominant products (i.e. less than 5%) [I]. Therefore, the WCAP chemical model only considers the release rates of the principal elements or ions guiding relevant compound formation: aluminum, calcium and silicate.

Reference document WCAP-16530-NP, the "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191," and its associated chemical effects model spreadsheet, were published as guidance to enable the industry to.perform plant-specitic chemical precipitate analyses which may be used toward facilitating chemical precipitate application to sump strainer testing activities [I].

Using the guidance and resources associated with WCAP-16530-NP, plant-specific containment material concentrations and densities, buffer solution type, as well as sump and spray pH and temperature transients post-accident, it is possible to predict the types and amounts of chemical precipitates which may form from the chemical reactivity of certain materials in the presence of specific aggressive chemical and thermal post-accident conditions.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 3.0 ASSUMPTIONS The following engineering assumptions are made in the course of the calculation to introduce additional conservatism andlor simplify the evaluation. Unverified assumptions that require confirmation of applicability of this calculation and its results are specifically noted. Unverified assumptions must be verified by Point Beach prior to use of the chemical effects calculation.

Sump Pool and AtmospherelSpray Chemistry & Temperature Parameters with Time:

I.

To address the extended time period required in I 0 CFR50.46(b)(5), Reference 8 (Volume 2, Section 2.0, paragraph 2) states: "For this evaluation of PWR recirculation performance, the staff considers this extended time to be 30 days, and requires cooling by recirculation of coolant using the ECCS sump." Therefore, this evaluation assumes that the mission time for the ECCS operation is thirty (30) days, and that only the quantity precipitate which is generated up to that point must be calculated for use in head loss and downstream analyses.

2. Several base cases within this evaluation assume that there are no solubility limitations which would inhibit chemical precipitation (i.e. the sump is unmixed). This assumption applies conservatism in that all elemental materials generated in each liquid chemistry condition (sump 1 spray) will precipitate into a resulting chemical compound (described further in Section 4.0).

3. It is assumed from the information within Reference 5, as submitted by Point Beach, that a minimum recirculation initiation time of 27 minutes following the break, based on maximum attainable ECCS flow rates with a minimum RWST volume, is acceptable for use in the evaluation. Hence in each base case, a start time of 27 minutes is conservatively used. As stated in Reference 5, sensitivity studies could be performed to investigate the effect of a smaller LOCA, resulting in sump recirculation initiation at much longer times (i.e. 60 or 120 minutes, see Appendices E,F,K,L).

4. Based on the information reported within Reference 5, it is assumed that the containment spray system will be aligned to allow containment spray pumps to take suction from the containment sump following both the initiation of sump recirculation, and the point at which the RWST or NaOH injection is secured. At this point, for all cases, the spray pH would revert from the elevated initial injection pH (10) to the maximum sump pH (9.5), with the sump medium now considered as "mixed" (see Assumption 8 for more detail). Therefore, the time period following each initiation time of sump recirculation would indicate the lower 9.5 pH. In each case, the initial pH of containment spray would be assumed to be that of the maximum buffered spray solution (pH 10).

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-I 6530-NP Document No. 51-9056525-001 For clarity, the pH evolution scenarios investigated are summarized below (for graphical version see Appendices A.6 through A.8). Note that the supplemental sensitivity analyses are not intended to correspond to any realistic plant scenario. These runs are included for illustration purposes only to demonstrate the behavior of potential chemical effects as a function of the duration of spray injection vs. spray recirculation.

P Base Case Analyses (Cases I

.3 12.3) -Appendix A.6:

i. Start of Sump Recirculation at 27 Minutes ii. Start of Spray Recirculation at 77 Minutes iii. If sump recirculation initiates at 27 minutes, and single train'operation results in RWST depletion after approximately 77 minutes, the spray pH would remain elevated (pH 10) until containment spray is rerouted to take suction from the sump (initiation of spray recirculation). Accounting for delays in mixing or suction switchover, the spray and sump pH will be assumed as "mixed" (maximum sump pH of 9.5) after approximately 100 minutes.

P Supplemental Sensitivity Analyses (Cases 1.4 I 2.4) - Appendix A.7:

i. Start of Sump Recirculation at 60 Minutes ii. Start of Spray Recirculation at 100 Minutes iii. If sump recirculation initiates at 60 minutes, the spray pH would once again remain elevated (pH 10) until containment spray is rerouted to take suction from the sump. Accounting for delays in mixing or suction switchover, the spray and sump pH will be assumed as "mixed (maximum sump pH of 9.5) after approximately 123 minutes.

P Supplemental Sensitivitv Analvses (Cases 1.5 / 2.5) - Ap~endix A.8:

i. Start of Sump Recirculation at 120 Minutes ii. Start of Spray Recirculation at 123 Minutes iii. In this case, if sump recirculation initiates at 120 minutes, the spray pH would once again remain elevated (pH 10) until containment spray is rerouted to take suction from the sump. Therefore, the spray and sump pH will be assumed as "mixed" (maximum sump pH of 9.5) after approximately 147 minutes.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

5. Not Used

6. It is assumed from the information within Reference 5, as submitted by Point Beach, that containment spray should be evaluated to operate for a total spray duration of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (with the expectation that this may be viable for each unit in the future if usage of the Alternate Source Term is sought).

7. It is assumed that the temperature profile information submitted in Reference 5 is acceptable for use in this evaluation at this time. As these profiles are not yet internally supported 1 documented as Point Beach calculations, it must be assumed that these temperature values are unverified assumptions.

8. It is assumed that the containment sump medium will become mixed following the initiation of sump recirculation for spreadsheet evolutions that will credit sump mixing. For both the minimum and maximum pH range conditions, this is assumed to occur once the spray pH reverts to the sump pH. Therefore, for the corresponding case sets outlined in Table 5-1, the spreadsheet (column G) has been altered to reflect this credit (Yes = I).

See Appendices A.6 through A.8 for pH evolution at this point.

Sump Pool Volume I Density:

9. As guided in Reference 1, if plants do not know the mass of the recirculation water for which the volume was calculated, the density of water at the temperature at which the sump pool volume was determined should be used. Reference 5 states that a temperature of approximately 60°F is appropriate for the volumes provided, and hence an average density of 62.4 tb/ff3, as noted in Reference 5, is viable and conservative for use in all simulations (It is not necessary to use density corrections because this value is conservative for use in all simulations).

10. For conservatism, the maximum 'available' sump volume has been applied to most base case and supplemental test runs as the sump volume spreadsheet material input to ensure the appropriate and bounding calculation of the maximum quantity of generated precipitatelmaterial. This value, 43,317 d, was extracted from Reference 5.

14. All reported results indicate the calculation of simulation specific precipitate concentrations with respect to available sump or recirculation volume. This action is included for illustration purposes only to exhibit the most conservative (highest) concentration of generated precipitates from the final material quantities calculated. The minimum 'available' sump volume has been applied when calculating concentration. This value, 22,995 ft3, was extracted from Reference 5.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 General Volumes - Material / Insulation 1 Debris:

12. For primary base case simulations, the bounding / maximum amount of insulation generated for each insulation type for each unit was selected from the data for each break case in the Point Beach Units I and 2 Debris Generation Reports (seeAppendixA.1) [5,12,13]. Though it is possible that these numbers may be bounded by a higher insulation volume, given the method of evaluation used in the Debris Generation Reports, as well as the existence of multiple additional conservatisms applied in the process of this chemical precipitation evaluation, it is believed that the data from the debris generation reports is representative of the volume of insulation which could fail and reach the sump water volume.

13. The volume of debris reported by Point Beach Debris Generation Calculations states quantities of generated insulation in terms of its original condition prior to LOCA initiation (i.e.

as-fabricated) [12,13]. Therefore, the "as-fabricated" densities for each type of insulation from NEI 04-07 are used [5,8].

14. Generated material volumes include at a minimum any material which is generated during a LOCA. For certain materials (generic fiberglass, CalSil), generated material volumes are also assumed to include associated latent and miscellaneous debris.

15. Any insulation materials which do tiof fail during a LOCA are assumed to be unaffected by the spray. This unaffected volume includes any metal encapsulated I jacketed insulation materials (unless the jacketing is composed of an aluminum alloy).

16. All jacketed insulation materials are assumed to be composed of stainless steel, unless identified in the aluminum alloy invenfory within Reference 5.

Fibrous Debris - Fiberglass Insulation:

17. Point Beach has a variety of mineral wool insulation installed at both Units I and 2 [53. Based on Reference 22, this evaluation assumes that the variety of mineral wool installed has the material composition of 'MinWool', as listed in reference 1 Table 3.2-1 (steel slag + 5%

phenolic resin binder, i.e. 4042% calcium oxide, 10-19% silicon dioxide, 7-30% iron (11) oxide, 2-10% iron (Ill) oxide, 5% manganese oxide, and minor amounts of aluminum oxide, phosphorus pentoxide, sulfur and iron). This evaluation therefore also assumes that Point Beach mineral wool insulation has a similar degradation rate of 'MinWool'.

18. Given no alternatives from NEI 04-07 Table 4-1 for mineral wool insulation types, it is necessary to assume that the Point Beach mineral wool insulation installed at both Units I and 2 has an as-fabricated density of 40 Iblft3 [5,8]. This density is conservative within the Page 15 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 range of as-fabricated densities prescribed for generic mineral wool as reported within NEI 04-07 (4,6,8 and 10 lb/ft3 are standard) [Reference 8 Vol. I Table 4-11.

General - Miscellaneous Debris:

19. In accordance with the current Point Beach design input transmittal, as well as the Unit 1 and 2 Debris Generation Evaluation, all miscellaneous debris reported as taking the form of 'foam' or 'film' are not applicable to the WCAP-16530-NP evaluation methodology, and therefore it is assumed that these materials are not expected to affect the quantity or iype of precipitate generated in the sump following a LOCA [5,12,13]. Therefore, it will be assumed that only miscellaneous frbrous and particulate debris are acceptable as inputs into this evaluation.

Fibrous Debris - Latent & Miscellaneous:

20. For conservatism, when calculating the input volume of latent and miscellaneous fiber from material masses given in Reference 5, the as-fabricated density for Nukon will be used (density of Nukon = 2.4 Ib1ft3, the lowest NEI 04-07 reported fiberglass insulation density)

[Reference 8 Vol. 1 Table 4-11. This will help to ensure that the largest volume of latent and miscellaneous fiber is applied to the generic fiberglass material input when calculating the amount of subsequent corrosion I leaching.

21. The generic fiberglass insulation as-fabricated density will be applied in the actual chemical model for latent and miscellaneous fibrous debris using the values reported in NEI 04-07

[Reference 8 Vol. I Table 4-11.

22. Generic fiberglass has a higher leaching rate than other tested fiberglass insulation materials

[Reference I Section 5.2.31. Therefore, the volume of latent fibrous debris present in the sump will be applied to the generic fiberglass material input section.

Particulate Debris - Latent & Miscellaneous:

23. For conservatism, when calculating the input volume of latent and miscellaneous particulate from material masses given in Reference 5, the density for Asbestos will be used (density of Asbestos = 7 lb/ft3, the lowest NEI 04-07 reported particulate insulation density) [Reference 8 Vol. 1 Table 4-11. This will help to ensure that the largest volume of latent and miscellaneous particulate is applied to the CalSil material input when calculating the amount of subsequent corrosion 1 leaching.

24. The CalSil insulation as-fabricated density will be applied in the actual chemical model for latent and miscellaneous particulate debris using the values reported in NEI 04-07 [Reference 8 Vol. 1 Table 4-11.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

25. This evaluation assumes that any latent or miscellaneous particulate debris has a degradation rate similar to that of CalSil. This assumption is valid as CalSil has exhibited the most significant material release rates when compared to other insulation material sub-types

[Reference 'I Section 5.2.31. Therefore, the volume of latent and miscellaneous particulate debris present in the sump will be applied to the CalSil material input section.

Particulate Debris - Coatings:

26. In accordance with guidance from industry research and documentation, it is unlikely that commonly found plant-specific coatings materials will break down to produce precipitate-forming species under the temperature and chemistry conditions tested E l,I 01.

27. It is assumed that the presence of aluminum-containing coatings materials will not result in the dissociation of additional aluminum ions into the sump medium. In most industry documentation, aluminum is primarily considered to be present due to the degradation of aluminum metal and fiber insulation [10,11,23]. Also, in accordance with guidance from

.industry research and documentation, it is unlikely that commonly found plant-specific coatings materials will break down to produce precipitate-forming species under the temperature and chemistry conditions tested (See Reference I Section 3.2) [I

,10], and noted that the presence of some organics and inorganics can even serve to increase the solubility of aluminum [I

,10,24].

Concrete in Containment:

28. It is assumed that the surface area delineated within Reference 5 includes all susceptible concrete within containment.

WCAP Spreadsheet Input & Errata Assumptions:

29. Certain spreadsheet errors were detected during internal and external review (see References 44 through 20 for more detail). Most of these reported errors are not applicable, or have been corrected within the spreadsheet revision used for this evaluation. The first error reported within Reference 16 has not been revised within the spreadsheet, but does not affect this evaluation given the plant-specific conditions and insulation debris types determined for Point Beach (no usage of Microthem or Min-K insulation materials). The second error within Reference 16 has been corrected in the spreadsheet used for this evaluation. The error reported within Reference 20 has also not been revised within the spreadsheet, but does not affect this evaluation given fhe plant-specific conditions for Point Beach (errata is applicable to TSP only).

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

30. It is assumed that the apparent error on page 3 of LTR-SEE-1-01-14 (embedded within Reference 18) with respect to the first revised coefficient for aluminum release (51.15271 versus 51.1271), is insignificant in effect upon spreadsheet results. When the coefficient difference is iterated within the spreadsheet, no significant effect to overall total precipitate quantify by precipitate type is noted (less than 0.05% difference).

31. The spreadsheet does not determine release rates for the following materialslelements shown to be present in Table 5.1-2 of Reference I, Aluminum release from CalSil Aluminum release from MIN-K Calcium release from MIN-K

,. Aluminum release from Interam Calcium release from lnteram With the exception of Aluminum release from lnteram, the wt% of the element present in the insulation type is low or negligible (Interam and Min-K are not insulation types found at Point Beach Units I or 2). Therefore, it is viable to assume that the release of these particular elements from each associated insulation type is negligible or inapplicable given the other conservatisms applied during the process of this evaluation.

32. The values provided in the Design Information Transmittal (DIT) text will be used for all inputs, with the exception of. temperature profile [5]. In this case, the excel profile attachment to Reference 5 will be used.

4.0 CHEMICAL MODEL SPREADSHEET The chemical precipitates of primary concern identified during the WOG chemical effects testing activities are aluminum oxyhydroxide (AIOOH), sodium aluminum silicate (NaAISi30s), and calcium phosphate (Ca3(P04)2). Aluminum oxyhydroxide wjl normally precipitate for plants which contain aluminum either impacted by the spray or submerged in the containment sump pool. However, for plants with high silicon releases, sodium aluminum silicate may be formed instead. It is expected that available aluminum ions will react with silicon ions released from CalSil or fibrous insulation materials to form NaAISi300. Calcium phosphate is not a concern for PBNP as the buffer solution utilized by Point Beach is sodium hydroxide (NaOH).

As PBNP employs sodium hydroxide (NaOH) as their containment spray buffer during accident conditions, it is not surprising that the predominant chemical precipitates would therefore likely be a mixture of aluminum oxyhydroxide (AIOOH) and sodium aluminum silicate (NaAlSi300) when the plant-specific debris mixture is subjected to a borated alkaline medium (such as that contributed by Page 18 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 NaOH in this case) [Reference I Section 6.1 & 6.41. However, also noted in Section 6.4 of Reference 1 is the guidance that the preferential formation of these compounds is dependent upon concentration. Therefore, if the concentration of silicate is greater than 3.12 times the concentration of aluminum, all aluminum will likely precipitate as sodium aluminum silicate [I]. Given the presence of a significant amount of silicon-containing insulation types in this evaluation, it is viable that the generation of NaAlSis08 could preclude the degree of AlOOH compound generation.

The first stage of the chemical model predicts both the rate of dissolution and the solubility limits for select elements at certain points after LOCA has occurred. The quantity of the elements that make up the precipitates is calculated using the chemical model spreadsheet associated with WCAP-16530-NP. To determine the quantity of the key precipitates, it is assumed that sodium (Na),

hydroxyl (OH), and phosphate (if applicable) will be present in excess [Reference I Section 6.41.

From these outputs, it is possible to deterrnine precipitate quantities given the stoichiometry of expected chemical compounds.

During the second stage of the modeling process, all material that has dissolved into solution is conservatively assumed to form precipitates due to the limited solubility of the 'keyJ chemical precipitates [Reference I Section 6.41. Solution concentrations of the dissolved elements and the potential mass of the three primary precipitate compounds are calculated with respect to time. In order to effectively eliminate any influence of variations in temperature upon the degree of precipitate formation, based on the low solubility of the three 'keyJ materials, the model assumes that all ions generated 1 leached following a LOCA will be available to form chemical precipitates.

Therefore, 100 percent of dissociated aluminum ions (and calcium when in the presence of phosphate) will form chemical precipitates. However, as the solubility of calcium silicate increases at lower temperatures during constant pH conditions, it is expected that dissolved calcium will remain in solution in the absence of phosphate [I].

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 4.1 Chemical Model spreadsheet inputs Initial Material Quantities In order to calculate the quantity and concentration of chemical precipitation that will take place, the quantity of materials that would be exposed to reactor coolant and containment spray post-accident must be defined. The PBNP plant-specific inputs are reported in Appendix A.l through A.4. They represent the maximum debris load without transport reductions. It is not advisable to use debris volumes that take credit for transport reductions, as all materials subject to the sump medium are generally assumed to degrade (i.e. dissociate) with time.

Material Densities Material-specific density values are also required in order to convert insulation material inputs /

volumes to mass. For all insulation materials, the "as-fabricated" density values given in Table 4-1 of NEI 04-07 or density values dictated by plant requirements may be used [5,8]. These inputs are reported in Appendix A.5.

pH and Temperature Transient Profiles Separate time dependent pH and temperature profiles for both sump and spray conditions post-accident must also be developed. This information is applied through numeric Integration of the tested material release rate equations to determine the cumulative release and dissolved concentration of each species with time [Reference 1 Table A-21. These inputs are reported in Appendix A.6 through A.8.

5.0 SENSITIVITY ANALYSES The effect on precipitate mass of altering several input parameters was explored using the chemical effects model. The parameters that were vaned during this process include sump pool volume, time of sump recirculation initiation, mixing of sump pool medium, application of viable corrosion inhibition parameters, and debris generation insulation volumes by case [3,5]. The test parameter combinations explored for both Point Beach Units 1 and 2 during this sensitivity analysis are outlined in Table 5-1.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51 -9056525-001 Table 5-1: Test Parameters - Units I

& 2 Page 21 of 56 Corrosion Inhibition Case Set 2aT Bounding Debris Inputs X

X X

2.1 2

X 27 27 27 I

X I

X 1

x 1

I X

l x

l x

l X

I X

1 1

X 2.4 2

X 2.5 2

X 2.6 2

X X

X X

X X

X Case Set 2b: Debris Generation Case Inputs I

60 1

120 1

27 X

X X

X X

X X

2.3.1 2

x l

x l

1 27 X

I 2.3.2 2

X l

X l

2.3.3 2

X X

1 2.3.4 2

X X

1 2.3.5 2

X X

I X

1 27 X

27 X

27 X

27 X

27 X

27 2.6.1 2

X X

I 2.6.2 2

X X

27 X

X 27 X

X 2.6.3 2

X X

2.6.4 2

X X

2.6.5 2

X X

2.6.6 2

X X

2.6.7 2 '

x l

x 27 X

X 27 X

X 27 X

X 27 X

X 27 X

X

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 5.1 Case Set la12a: Bounding Debris Inputs Base Case Analyses:

Cases I.I - 9.3 and 2.1 - 2.3 For each unit, the first three runs within each case set are the base case simulations. These tests were performed while varying a combination of sump water volume and mixed sump inputs. The magnitude of the sump volume was varied bebeen the maximum and minimum recirculation water volumes reported by Point Beach and was found to significantly affect the degree of precipitation.

Base case numbers I and 3 were performed at the maximum sump volume (43,317 fi3),

and base case number 2 was performed at the minimum sump volume (22,995 ft3) 151. For each of the base cases, the appropriate transient pH and temperature profile may be found in Appendix A.6, reflecting the usage of a maximum pH profile and sump recirculation initiation at 27 minutes. Chemical model material and sump volume inputs are reported in Appendix A, and model predictions for elemental release and precipitation are reported for Unit I in Appendices B through D, and for Unit 2 in Appendices H through J. Bounding debris generation volumes were used as material inputs for all base cases (see process in Appendix A.2).

Supplemental Analyses:

Cases 1.4 - 1.5 and 2.4 - 2.5 For each unit, the following two sensitivity runs are supplemental analyses performed to investigate the effect of increasing the time to sump recirculation initiation on the degree of precipitation.

Supplemental case numbers 4 and 5 were performed for each unit at the maximum sump volume (43,317 ft3) [511. For these cases, the appropriate transient pH and temperature profile may be found in Appendix A.7 and A.8, reflecting the usage of a maximum pH profile and recirculation initiation at 60 and 120 minutes respectively 151. Chemical model material and sump volume inputs are reported in Appendix A (identical to the base cases), and model predicfions for elemental release ancj precipitation are reported for Unit I in Appendices E and F, and for Unit 2 in Appendices K and L.

Supplemental Analyses: Additional Input Evaluations Cases 1.6 and 2.6:

For each unit, the next sensitivity run is a supplemental analysis performed to investigate the effect of taking credit for WCAP-16785-P inhibition and solubility effects on the degree of precipitation. For this supplemental case, the appropriate transient pH and temperature profile may be found in Appendix A.6, reflecting the usage of a maximum pH profile and sump recirculation initiation at 27 minutes [5].

Other specific manipulations were performed within the chemical model spreadsheet, as outlined in WCAP-16785-P. Chemical model material and sump volume inputs are reported in Appendix A (as

' All supplemental analyses (Cases 4 through 6) are performed using base case 3 parameters for each unit.

Page 22 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 identical to each base case), and model predictions for elemental release and precipitation are reported for Unit 1 in Appendix GI and for Unit 2 in Appendix M.

5.2 Case Set I bl2b: Debris Generation Case Inputs Supplemental Analyses: Additional lnput Evaluations -All Debris Gen Cases Cases 1.3.1 - 1.3.5 and 2.3.1 - 2.3.7:

For each unit, the sub-cases have been performed using 1.3 and 2.3 base cases to ensure that all debris combinations are investigated in the process of this evaluation, as identified through debris generation calculations. All sub-case 1.3 and 2.3 simulations were performed at the maximum sump volume (43,317 f13), and the transient pH and temperature profile be found in Appendix A.6, reflecting the usage of a maximum pH profile and sump recirculation initiation at 27 minutes. Chemical model material and sump volume inputs are reported for Unit 1 in Appendix A.3, and for Unit 2 in Appendix A.4. A summary of model predictions for elemental release and precipitation are reported for Unit ? in Appendix N.1, and for Unit 2 and Appendix N.2.

Supplemental Analyses: Additional lnput Evaluations - All Debris Gen Cases Cases 1.6.1 - 1.6.5 and 2.6.1 -2.6.7:

For each unit, the last set of sensitivity runs are supplemental analyses performed to investigate the effect of taking credit for WCAP-16785-P inhibition and solubility effects on the degree of precipitation for each individual debris generation case. For these supplemental cases, the appropriate transient pH and temperature profile may be found in Appendix A.6, reflecting the usage of a maximum pH.profile and sump recirculation initiation at 27 minutes 151. Other specific manipulations were performed within the chemical model spreadsheet, as outlined in WCAP-16785-P, and as directed within Reference 18.

Chemical model material and sump volume inputs are reported for Unit I in Appendix A.3, and for Unit 2 in Appendix A.4. A summary of model predictions for elemental release and precipitation are reported for Unit 1 in Appendix N.1, and for Unit 2 and Appendix N.2.

6.0 RESULT SUIWIMARY A summary of resultant precipitate outputs is outlined in Table 6-1 for the combination of test parameters explored in the process of this evaluation (see Table 5-1 for Test Parameters).2a Of each unit set of base cases, case number 1 resulted in the most significant amount of material precipitation. As identified within Section 5.0, these test runs were performed at the maximum sump pH profile and other test parameters reported in Table 5.1 for each of the Point Beach Nuclear Plants, Table 6-1 states the Mass of Silicon and Aluminum Release in a 30 day simulation period.

Page 23 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No, 51-9056525-001 Table 6-1: Test Outputs for Units 1 & 23 The "Concentration" values reported in Table 6-1 are for illustration purposes only as these values are normally dependent on strainer test volume. In the above cases, "Concentration" is determined using the minimum sump volume as provided for this evaluation (22,995 ff3), which is not the value reported and utilized in previous chemical effects evaluation revisions (832700 L) [4]. These values are provided only to allow cross reference to previous calculations and supporting documentation.

Page 24 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 One common thread throughout all cases is the fact that no chemical precipitation of AIOOH or Ca3(P04)2 was observed to occur due to the degradation of debris in containment. Only sodium aluminum silicate (NaAlSi30s) was determined to precipitate in all cases. Therefore, as predicted in Section 4.0, given the presence of an adequate amount of silicon-containing insulation types for each unit (CalSil or fibrous-based), the generation of NaAISi30B has been found to preclude the degree of AIOOH compound generation. In the case of precipitate formation for either compound, it should be noted from each chemical formula that aluminum ions are the limiting component for chemical debris precipitation in all cases.

It can also be concluded that the higher level of chemical precipitation for Unit 2 is simply due to higher quantities of material inputs (primarily Nukon: aluminum and silicon containing insulation debris), as all other input parameters were identical between the two case sets (temperature and pH transient profiles, sump mixing, material density inputs).

It is important to also point out that the presence of a higher pH sump I spray atmosphere in this particulate debris and buffer configuration will also significantly affect precipitation, and any future decreases in pH will benefit each unit in terms of expected resultant head losses at the strainer (see Section 3.0 Assumptions).

6.t Case Set Ial2a: Bounding Debris Inputs Base Case Analyses:

Cases 1.1 - 1.3 and 2.1 - 2.3 It can be concluded from these cases that higher levels of chemical precipitation noted between Cases 1.112.1 and 1.212.2 are simply due to the conservative application of a significantly higher water volume (minimum vs. maximum), resulting in increased chemical precipitation due to the volume available for solution of dissociated ions. The assumption of the minimum sump volume in Case I

.2/2.2 had a marked effect on final precipitate quantity when compared to Case 1.112.1, resulting in greater than 30% reduction in precipitate overall quantity.

As evidenced in the above simulated test results, increased chemical precipitation can result from the presence of a higher sump volume. Dependence upon sump volume is normally attributable to the I

i fact that the release rate of aluminum from these materials decreases with time as the solubility limit is I

approached, and that the release rate from aluminum silicate insulation materials decreases with I

increasing concentration of dissolved aluminum from all sources due to the common ion effect I

[Reference I Section 6.11. These conclusions are made for these cases as all other input parameters between the first two cases were identical (temperature and pH transient profiles, unmixed sump medium, material density inputs).

Page 25 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 The introduction of sump "mixing" in these simulations (Case 1.3/2.3), also had a significant effect on final precipitate quantity when compared to Case 1.112.1. Applying sump mixing only after recirculation is expected to occur (see Assumption 8) resulted in greater than 40% reduction in precipitate overall quantity. Through this case, it is made obvious that this option within the spreadsheet has a substantial effect on precipitate generation as it allows the elemental mass already released into the sump solution to impact the dissolution rate from each material containing that element [I].

Chemical model predictions for elemental release and precipitation are reported for Unit 1 in Appendices B through D, and for Unit 2 in Appendices H through J.

Supplemental Analyses:

Cases 1.4 - 1.5 and 2.4 - 2.5 The manipulation of sump recirculation initiation time start had little to no effect on the final precipitate quantity determined to generate using the spreadsheet. This is likely a primary result of a greater percentage of exposed aluminum source materials as opposed to any submerged areas for both units.

Only the exposed source materials would be affected with any significant recirculation changes.

Though the pH passing through the containment spray injection header is elevated for a longer period when recirculation initiation is delayed, the time frame of pH subjection does not appearto be long enough for exposed aluminum materials to result in any significant increases in dissociation 1 degradation.

Chemical model predictions for elemental release and precipitation are reported for Unit I in i

I Appendices E and F, and for Unit 2 in Appendices K and L.

I Supplemental Analyses: Additional Input Evaluations (WCAP-16785-P)

Cases 1.6 and 2.6:

The application of solubility and inhibition limiters, as guided in References 3 and 18, had some effect on chemical precipitation. The following WCAP-16785-P simulations were applied to each unit base case (I

.3,2.3):

P Silicate Inhibition: for plants exceeding 75 ppm Silicon, Ss Silicate Inhibition: for plants with 50 to 75 ppm Silicon, 9 Aluminum Oxyhydroxide Solubility Limit.

All other inhibitionlsolubility cases were not applicable for Point Beach, and were therefore not applied during these evaluations.

These supplemental cases are labeled as I

.6 and 2.6 in the results above, and model predictions for elemental release and precipitation are reported for Unit I in Appendix G, and for Unit 2 in Appendix M.

I Page 26 of 56 I

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAPI 6530-NP Document No. 51-9056525-001 Only the Silicate lnhibition for a plant exceeding 75 ppm had any affect on the final total precipitate quantity. When comparing Cases 1.3 (mixed base case) and 1.6 (mixed additional inputs), with all other conditions identical, the change results in approximately 5% reduction in precipitate overall quantity. When comparing Cases 2.3 (mixed base case) and 2.6 (mixed additional inputs), with all other conditions identical, the change results in approximately 10% reduction in precipitate overall quantity. Though these sensitivity runs did result in some reduction of precipitate quantity, it may be advisable for the user to consider application of only the base case simulations until regulator approval is granted or expected for Reference 3.

6.2 Case Set I b/2b: Debris Generation Case Inputs Supplemental Analyses: Additional Input Evaluations - All Debris Gen Cases Cases 1.3.1 - 1.3.5 and 2.3.1 -2.3.7:

These cases are purely debris generation case specific. Each separate simulation corresponds to pure debris generation case debris output results, and has been included to allow for parametric review by the user. Cases I

.312.3 have been used as the correlating base cases for each unit. Any variability in precipitate results is directly related to the fibrous and insulation debris quantities applied (aluminum source materials, temperature and pH profiles, and sump mixing were all constant).

Chemical model material and sump volume inputs are reported for Unit I in Appendix A.3, and for Unit 2 in Appendix A.4. A summary of model predictions for elemental release and precipitation are reported for Unit 1 in Appendix N.1, and for Unit 2 and Appendix N.2.

Supplemental Analyses: Additional input Evaluations -All Debris Gen Cases Cases 1.6.1 - 1.6.5 and 2.6.1 - 2.6.7:

The application of solubility and inhibition limiters to specific debris generation simulations also had some effect on chemical precipitation. As for cases 1.612.6 above, the following WCAP-16785-P simulations were applied to each debris generation "base case" (I

.3.1-1.3.5, 2.3.1-2.3.7):

9 Silicate Inhibition: for plants exceeding 75 ppm Silicon, 9 Silicate Inhibition: for plants with 50 to 75 ppm Silicon, P Aluminum Oxyhydroxide Solubility Limit.

All other inhibitionlsolubility cases were not applicable for Point Beach, and were therefore not applied during these evaluations. These supplemental cases are labeled as 4.6.X and 2.6.X in the results above, and model predictions for elemental release and precipitation are reported for Unit 1 in Appendix N.1, and for Unit 2 in Appendix N.2.

Once again, only the Silicate Inhibition for a plant exceeding 75 ppm had any affect on the final total precipitate quantity. Though these sensitivity runs did result in some reduction of precipitate quantity, Page 27 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 it may be advisable for the user to consider application of only the base case simulations until regulator approval is granted or expected for Reference 3.

Comparison to Past Testing Activities If a direct comparison to past testing is desired (see also Reference 4) to be consistent with the previously performed chemical precipitation evaluation, this evaluation will use the 832700 L (Reference 4 Section 5.0 Min Sump Volume) value for sump volume for illustration purposes only.4 A summary of resultant precipitate concentrations is outlined in Table 6-2 for the combination of test parameters explored in the process of this evaluation (see Table 5-1 for Test Parameters, and Table 6-1 for original test results for the applicable cases). Note the significant difference in the volume at which the maximum precipitate mass was calculated (from Appendix A, Max Sump Volume = 4331 7 ft3 = 1226763L). Note: Future strainer testing activities will likely use a scaled correlation to strainer size to determine quantity to be introduced to strainer testing activities, therefore direct comparison to previous strainer testing or calculation results is not recommended other than for illustration purposes only.,

Table 6-2: Comparison Concentrations for Units 1 & 2 The concentration of chemical effect precipitate material that was used as a physical input into the debris configuration developed for the PBNP Sump Strainer Performance Testing to enable the simulation of the most representative chemical environment present inside the PBNP reactor containment water pool after a loss-of-coolant accident was 589 mg/L, much greater than even the most conservative simulations reported in this evaluation [4].

The "Concentration" values reported in Table 6-1 are included for illustration purposes only.

Page 28 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

7.0 CONCLUSION

The results of these model predictions have been briefly summarized for review in Table 7-1 below5.

Table 7-1: Test Summary - Units 1 & 2 For each unit, of the three base cases, run number 'I resulted in the most significant amount of material precipitation (unmixed assumption). Therefore, the maximum conservative mass of chemical precipitate materials for Unit 1 is equal to 248.252 kg (NaAISi3O~), and the maximum conservative mass of chemical precipitate materials for Unit 2 is equal to 274.808 kg (NaAISi30s).

However, if 'mixed' evaluation outputs are desired for final use in chemical debris calculations, the maximum mass of chemical precipitate materials for Unit I is equal to 144.434 kg (NaAISi30B), and the maximum conservative mass of chemical precipitate materials for Unit 2 is equal to 174.405 kg (NaAISi30s) for plant-specific conditions that include the initiation of sump recirculation 27 minutes after the accident has occurred6.

As identified within Section 5.0, fhese test runs were performed at the maximum sump pH profile, and other test parameters reported in Table 5.1 for each of the Point Beach Nuclear Plants.

Additional sensitivity runs have been incorporated into this report to permit chemical debris margin allowance changes should plant-specific parameters evolve after this report is finalized (i.e. time of sump recirculation initiation).

Page 29 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAPI 6530-NP Document No. 51-9056525-001 With respect to concentration, it can also be validated that the concentration of chemical effect precipitate material used as a physical input into PBNP Sump Strainer Performance Testing (589 mglL) is conservative 141. Therefore, this evaluation has substantiated the type and concentration of chemical effects material which has been conservatively evaluated in previous Chemical Effect Precipitation analyses as likely to precipitate in the event of a loss-of-coolant accident at the PBNP Power Station.

Page 30 of 56

8.0 REFERENCES

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

[I]

• WCAP-16530-NP 8 Spreadsheet, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191", February 2006.

[2j Regulatory Guide 1.82 Revision 3, "Water Sources for Long-Term Recirculation Coolng Following a Loss-Of-Coolant-Accident," November 2003.

[3]

• WCAP-16785-P, "Evaluation of Additional Inputs to the WCAP-16530-NP Chemical Model", April 2007.

141 AREVA NP Doc. No 51-9010780-001, "Chemical Effects Material Selection for Point Beach Sump Strainer Performance Test," April 2006.

[5] AREVA NP Doc. No 38-9056238-000, "Design Information Transmittal (DIT) in support of GSI-191 Chemical Effects Evaluation - Point Beach," July 2007 (NPL 2007-0135).

[6] AREVA NP Doc. No 38-9018142-000, "Design Information Transmittal (DIT) from Tom Kendall of Point Beach to Support," April 2006 (NPL 2006-0052).

171 AREVA NP Doc. No 38-9003352-000, "Point Beach GSI-191 Downstream Effects Input,"

October 2005 (NPL 2005-01 93).

[8] NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Volumes 1 (Methodology) and 2 (Safety Evaluation), December 2004.

191 NUREGICR-6913, "Chemical Effects Head-Loss Research in Support of Generic Safety Issue 19IJ', U.S. NRC, December2006.

[ I 01 NUREGICR-6914, "Integrated Chemical Effects Test Project: Consolidated Data Report," Los Alamos National Laboratory, December 2006.

[ I I]

NUREGICR-6873, "Corrosion Rate Measurements and Chemical Speciation of Corrosion Products Using Thermodynamic Modeling of Debris Components to Supporf GSI-1911J, U.S. NRC, April 2005.

[I23 AREVA NP DOC. No 32-5050092-002, "DEBRIS GENERATION EVALUATION FOR POINT BEACH NUCLEAR PLANT UNlT 1," June 2006.

[ I 31 AREVA NP DOC. NO 32-5052938-003, "DEBRIS GENERATION EVALUATION FOR POINT BEACH NUCLEAR PLANT UNlT 2," June 2006.

[I41

• WOG-06-402, Errata to WCAP-16530-NP, "Distribution of Errata to WCAP-16530-NP, "Method for Evaluating Post-Accident Chemical Effects in Containment Sump Fluids" (PA-SEE-0275):" March 2006.

[ I 51

• WOG-06-107, Errata to WCAP-16530-NP, "PWR Owners Group Letter to NRC Regarding Error Corrections to WCAP-16530-NP (PA-SEE-0275)," March 2006.

[I61

• WOG-06-232, Errata to WCAP-16530-NP, "PWR Owners Group Letter Regarding Additional Error Corrections to WCAP-16530-NP (PA-SEE-0275):" July 2006,

[ I 7j

• 0G-07-270 "New Settling Rate Criteria for Precipitates Generated in Accordance with WCAP-16530-NP (PA-SEE-0275)", June 2007.

11 81

• 06-07-282 "Instructions to Evaluate Specific Inputs for WCAP-I 6785-NP; PA-SEE-0354, "Incorporation of Additional Inputs in the Chemical Effects Spreadsheet"," June 2007.

I Page 31 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

[I 91

• 0G-06-255, Errata to WCAP-16530-NP, 'PWR Owners Group Letter Releasing Revised Chemical Model Spreadsheet From WCAP-16530-NP (PA-SEE-0275)," August 2006.

[20]

• 00-06-273, Errata to WCAP-16530-NP, "PWR Owners Group Method Description of Error Discovered August 15,2006 in Revised Chemical Model Spreadsheet (PA-SEE-0275);" August 2006.

[21] Lindeburg, Michael R., Mechanical Engineering Reference Manual for the PE Exam, Eleventh Edition, Professional Publications, Inc., 2001.

[22] AREVA NP Doc. No 38-9057297-000, "Design Information Transmittal (DIT) from Tom Kendall of Point Beach to Support," August 2006 (NPL 2007-0145).

[23] NUREGICR-6912, "GSI-191 PWR Sump Screen Blockage Chemical Effects Tests; Thermodynamic Simulations", U.S. NRC, December 2006.

[24] NUREG/CR'-691 5, "Aluminum Chemistry in a Prototypical Post-Loss-of-Coolant-Accident, Pressurized-Water-Reactor Containment Environment", U.S. NRC, December 2006.

[25] AREVA NP DOG. NO 38-9058651-000, "DIT from Point Beach in support of Chemical Analysis -

Final approval of Calculation 2000-0036 Rev 2,"August 2007 (NPL 2007-0160).

Note:

"This reference' is not retrievable from AREVA NP documerit control system, but can be retrieved through the Westinghouse Owner's Group. Per AREVA NP administrative procedure, 0402-01 Appendix 2, these PMlPE Signature:

Page 32 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 APPENDIX A: GENERIC TEST INPUTS A.1 Primary Evaluation Inputs Post-Accident - Base Cases Point Beach Llst of lnputs - Malarial Quantiw CS input Affected Metalllc Alumlnum Aluminum Slllcate Page 33 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Point Beach Llst OF Inputs -Material QuanfftyG Input Affected Latent Debris For conservalism, when calcUlatlng the volume of latent particulete, the denslty for Asbestos(lower density than CelSil) will be used.

Latent Particulate (@)

Mess of Lefenl Palhuale = 127.5 Ibm DensllyofAshesfos 70 1bh7' [Reference BVof. 1 Table 4-71

((f 27.5 Ibrn latent particulate 17 lbmld = 18.2 @)

Total Addltlonal Fiber tf -Unit I 47.6,%:b

!-I";KI.....

$~..?,

..... Mkcellaneous Fiber (R')

Unit I + Latent Abar (f?)

~

~

~

b

~

k o h l Additional Fibr :$ - Unit2 Id

, 'L"""!"

..4i,~i.,.:~J:~i:?

Miscellaneous flber (f?) - Unit2 + Latent Abar (p)

Latent)

Total AddiUonal Particulate (f?) - Unit I 18.716':BiL!$2BIP Mkcellaneous Particulate (RS) - Unlt I +Latent Particulate lRS)

Tolal Additlonel Particulate (tf) - Unit 2 21.21':@iC$$#f'E Mkoellaneous Particulate (d) - Unit 2 +Latent Particulate IR')

Page 34 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 A.2 Primary Evaluation Inputs Post-Accident - Bounding Case ~ a s e s ~

BASE CASES onal Spreadsneet Inputs This table includes Debris Generation inputs taken directly from Reference 5 (latent or miscellaneous debris volumes were not added). However, Unit 1 and 2 Debris Generation Quantities applied to spreadsheet evaluations do include applicable latent and miscellaneous debris additions, as applied to the CalSil and Fiberglass cell inputs (these additional quantities were added in Appendix A.1 for Bounding Cases and A.3IA.4 for Debris Generation Cases)..

!;;*: ! ~ ~ l V i 7 8 : ; > ] & ). ~ ; ~ ~ ~

~!~~&~i!llJ&;;:

i::i;$!ip~ Class:;.${.$

~ ~ $ ; $ # j ~ j ~ ~ ~ q t ~ ~

.ill, +..A:'.

'Caic~um Silicate Page 35 of 56 Additional Spreadsheet lnpuk

+-~y2kp~:,~~~~\\t~y;~~llbl!Ji!(,.

&~!~~~ii~~$.~i:h~~+l~~,i>j%it+~~:~t?,

+&.Material

/:Patame$%

~ & ! ~ ~ ~ ~ % 4 ? ~ ~ R @ ~ ~ $ ~ < $ ~ 9 ~ ;

.+.1.~1!.3,-+,1

%-A..

CalSiI Insulation f)*2$$t43j:

i~U$t$i LCase.~:

r;.:r::;::>:&;

111.84 116.07 Zc.-+tLs<g!21,i kJ&t;@

LCase S, r+,1;:s;:12y,6j 83.87 116.07 I,?':fi~!i~~>?~;

&.$4!kf4

-Cas,q,,4;,

\\#:dc+2:.?i+

83.87 80.72

'~l<~l{$%~?

$u?!$,@

I 6:.

~~,~;:;:j~~ji, 6.6 2.37 114.7 1046.65 89.42 311.3

~ ~ ~ ; $ ~ ~ $ $

uniq2$i

egg;'$

Tt~;~~;::;~,

88.46 116.07

~r:$;fl!$J!~,

~nik,%!

$g=$g, h:clz?g,jfi, 89.36 296.74 53.71 107.35 01 937.77 01 89.77 01 291.43 -

90.57 181.4 107.4B

,l<?!!$j;,kq<4

7~g!j#

.=stub

!:ii;,!!.;e5jfi 113.05 E-gle6s 159.7 Asbestos InsUlatIon 63.451 63.45

$D%?i$!,<(b:,

!:~,"i,gd:

dgeg %:

/3ji+:pietb~!

122.72 116.07

$;?ii:$r$r!;?

e2,bnit4.1:t cd%p$I

~!;~,>,,r+

y32, t,

59.53

r:7i;:*3!Jt,:,

i,,uPiki~t ti!Biig$a

~~.!,p~ll:A, 110.5 286.74, 275.371 160.81 422.14) 566.25 201 29.48

. 1301491 18.37

& \\ ~ ~ e ~ l : ! ' s ~

yl:@fc%:

i;Fbis;s

,,r,::i::d:;b;~:

83.87 68.63 i$~:,~&i18k2$

iIu.nit,~!,

ls&v&

+ j y ~ : ~ ~ ~ g ~, ~ ~,

63.57 Ol l O O l. l l 849.5 23.441.

89.281 BB.57 218.9Vl 267.211 3232 98.75 Fiberglass lnsulallon 0

11.82 179,381 125.87) 78.02 NUKON

-TernpMat 01 01 0

23.44) 20.61) 7.3

'~ineml Wod JMln-Wool 203.11(

01 01 6

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 A.3 Unit I Debris Generation Case Inputs - All Cases [518 A.4 Unit 2 Debris Generation Case Inputs -All Cases [519

]Mineral Wool 1 267.21) 323.2) 130.491 18.37) 311.3)

0) 291.431 Unit I Debris Generation Quantities applied to spreadsheet include applicable latent and miscellaneous debris additions, as applied to the CalSil and Fiberglass cell inputs (see Assumptions). These additions are included in the tables above.

Unit 2 Debris Generation Quantities applied to spreadsheet include applicable latent and miscellaneous debris additions, as applied to the CalSil and Fiberglass cell inputs (see Assumptions and Appendix A). These additions are included in the tables above.

Page 36 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-36530-NP Document No. 51-9056525-001 A.5 Material Specific Density Values Point Beach List of Inputs - Material I Parameter Densities Page 37 of 56 Calcium Silicate

[Reference 8 Table 4-1 Mineral Wool Aluminum Silicate Interam Min-Wool Rock Wool Cerablanket FiberFrax Durablanket Kaowool.

Mat-Ceramic Mineral Fiber PAROC Mineral Wool lnteram (General: as-fabricated)]

~~~j~~~f~~;~~;@~;g$#fij~ggg[$~;g~g~@f;~~$~

f!;$f;l3iJ4

[ ' ~ ~ ~ ? # ! + p ~ #, i

~ ~ ~ ; \\ i \\ I ~ ~ R # ! ~ T ~ ~ f $ ; @ ~ $ ~ f i ~ ~

i :I $5 :<' {i::

6r).!:

I el, ri~$j~;!~::+~;~l,~s,,7,Mj:,

+~~~$~~ji~~,;;,,,

+ + ~:~,~$,&:d;$ !:lcd 3~itggl,

,;~~jl~~p~p;~~,~~!i~~~p

,* u ~ n ' l ~ ~ ;

IL,,~T,. I.: +;1~~;~y~;p1cqt?lly~:~o~~;:\\p~~qp~g~~:!i:~~firpi:~1~

n ~ ~ ~ i, ? l ~. s ~ i r ~.. !

q,Gz.,

.?,#,I,$, jar: 4~1&i!!:<

j~~j?;jj$g~$g~g$@fjgjl;g~@$&,~,~jg$y~~~~#$$j~~g{j#~

~~~~gy~g~~~i;#;~~;jp~;~~(~jj@~~$g~~~~;;g~@j~]$~g g$v,ryF

,, k d d ~. ~ $ ~ ~ ~ ~ ~ $ ~ ~ ! ~ ~ ~ i 8 $ l t ~ ~ ~ ~ ~

. 2 7 *:,Uk~RI3~ Zrltf $k$!~c i, TI?~:$.I

& h ~ ~ ~ $ f / { ] ~ ~ ~ # ~ ~ \\ ~ ~ \\.

t~,J,,,,@~~:~~

!lit7,

?;!$!,$,,c, I".

$&*;!: #3l~i7,~ifi?~

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I$ IVY i

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ll;&?~i~i~~q!t~l;~$$hL\\.;~/~~~b~

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 A.6 Temperature & pH Transient Profile - Sump Recirculation @ 27 minutes, Spray Recirculation @ 77 minutes Polnt Beach List of Inputs - Tempemture &pH Profiles S ~ m v Reclmulatlon at 77 minutes Page 38 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 A.7 Temperature & pH Transient Profile - Sump Recirculation @ 60 minutes, Spray Recirculation @ 100 minutes Polnt Beach List of Inputs -Temperature & p H Profiles Page 39 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-46530-NP Document No. 51-9056525-001 A.8 Temperature & pH Transient Profile - Sump Recirculation @ 120 minutes, Spray Recirculation @ 123 minutes Point Beach List of Inputs - Temperature B pH Pmflles Spray Reclrculatlon at 123 minutes Page 40 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using VVCAP-16530-NP Document No. 51-9056525-001 APPENDIX L: CASE 2.5 - SUPPLENiENTAL CASE: MAX PH, MAX SUMP VOLUME, MIXED, SUMP RECIRC @ 120 MINUTES, SPRAY RECIRC @ 123 MINUTES L.1 Elemental Releases and Precipitation - Case 2.5 Page 51 of 56