ML090980523

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

ML090980523
Person / Time
Site:	Point Beach
Issue date:	04/07/2009
From:	Meyer L Florida Power & Light Energy Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-04-002, GSI-191, NRC 2009-0033, TAC MC4705, TAC MC4706
Download: ML090980523 (163)

Text

Poiam% Beach N~oclsar Plant April 7, 2009 FPL Energy Point Beach, LLC, 6590 Nuclear Road, Two Rivers, WI 54241 NRC 2009-0033 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units I and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Response to Request for Additional lnformation GSI-I 91IGL 2004-02 (TAC NOS. MC470514706)

Potential Impact of Debris Blockage on Emerqencv Recirculation During Design Basis Accidents at Pressurized Water Reactors

References:

(I)

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 I and 2 GSI-1 9IIGL 2004-02, Request for Additional lnformation (TAC NOS. MC4705/4706), dated January 7,2009 (ML083300173)

FPL Energy Point Beach, LLC, previously submitted supplemental responses to Generic Letter (GL) 2004-02 in References (I) and (2).

Reference (3) contains a request for additional information based upon NRC Staff reviews of References (I) and (2). contains the FPL Energy Point Beach response to Questions 1, 2, 3, 13 through 19, and 21 through 24 of the RAI. Enclosures 2 through 6 are provided in support of information summarized in Enclosure 1.

The responses to Questions 4 through 12 and Question 20 are being deferred in accordance with the provisions of Reference 3.

An FPL Group company

Document Control Desk Page 2 Summaw of Reaulatorv Commitments 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 April 7, 2009.

Very truly yours, FPL Energy PAnt 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 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR,PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-IQIIGL 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 FPL Energy Point Beach, LLC 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 transports 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, FPL Enerny Point Beach 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 varies with location and unit, and has previously been modeled in the debris generation analyses for pipe breaks of loop piping.

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 15

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.

FPL Enerqv Point Beach 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.

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.

Provide assumed specific surface areas for fibrous and particulate debris.

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

Debris was added to the flumes for the screen and debris interceptor tests 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, ML041550380).

Materials Used for Screen Testing Surrogate Processing Shredder Chunks ~ 1 ~ x 4 "

Wood chipper Shredder Chunks >Inx4" Wood chipper Shredder Wood chipper Shredder Wood chipper Shredder Shredder Asbestos particulates

( I 0%)

Cal-Sil Latent dirt & dust Zinc Coatings Aluminum coatings Alkyd coatings Unqualified Epoxy Qualified Epoxy in ZOI 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 Petformance 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) (ADAMS Accession Number not available).

Surrogate Test Material Ceramic fiber NUKON NUKON Baked-out NUKON Temp Mat Temp Mat Temp Mat Owens Corning Fiberglass Owens Corning Fiberglass 10 pcf MW Fiber 10 pcf MW Fiber Baked-out NUKON Debris Type Asbestos fiber (go%,

fines)

NUKON (large)

NUKON (small)

NUKON (fines)

Temp Mat (large)

Temp Mat (small)

Temp Mat (fines)

Fiberglass (small)

Fiberglass (fines)

Mineral Wool (small)

Mineral Wool (fines)

Latent fibers Degraded Epoxy (chips)

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.

Density (lblft3)

I 0 2.4 2.4 2.4 11.8 11.8 11.8 5.5 5.5 8.0 8.0 N/A (direct weight from analysis) 10.0 14.5 NIA (direct weight from analysis) 457 94 90 94 94 Page 3 of 15 94 Cal-Sil Cal-Sil PWR Dirt Mix Tin Powder Walnut shells Walnut shells Walnut shells Walnut shells Powdered Powdered NIA NIA Powdered Powdered Powdered Powdered Acrylic 1/64 - ?4" chips

Question 3:

Please provide the information concerning the debris transporf analysis that was requested in the NRC sfaff revised contenf guide.

FPL Energy Point Beach Response:

FPL Energy Point Beach is embarking on a fiber reduction effort to resolve questions and concerns regarding debris generation and transport. An end result of this effort will be a revision or the potential elimination of the debris transport analysis. An extension for completion of corrective actions for GL 2004-02 will be required beyond the current date of June 30, 2009, to complete the fiber reduction effort. The scope and schedule for the fiber reduction effort will be included in the extension request. The request will be submitted in accordance with the criteria contained in SECY-06-0078 (ML053620174).

Question 13:

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

FPL Energy Point Beach Response:

The replacement strainer assemblies were originally designed to operate with a debris loaded head loss of 38 inches 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 AP of the strainers and related piping and supports has been increased to 10 feet. This is believed to be the maximum AP 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 foot AP 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 flowllow 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. These measures will be incorporated into the responses to Questions 4 through 12 and 20 in accordance with the provisions of the Commission's January 7, 2009, letter (Reference 3).

Page 4 of 15

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 below.

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 I strainer assemblies to accommodate an operating AP of 10 feet. 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 I strainers with APs as high as 10 feet. Similar modifications to reinforce the Unit 2 strainer assemblies for this higher AP 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 feet. 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.

Flashing Evaluation contains a calculation demonstrating that a debris loaded strainer assembly head loss of 10 feet 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 10 feet.

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, 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 calculations of NPSH at PBNP have omitted the strainer assembly and the depth of water above the containment sump as a simplification. As long as the head loss across the screens is less than the depth of water above the containment sump floor, this permitted analytically decoupling the screen design analyses from the emergency core cooling system (ECCS) NPSH calculations. The decoupling also protects the design margins in both of the analyses.

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 Page 5 of 15

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 feet that may occur as the sump cools down.

Question 14:

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

FPL Energy Point Beach 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 Type Zinc coatings Alkyd coatings Degraded Epoxy coatings outside of the ZOI Volume Density 1.95 ft3 457 1blft3 6.83 ft3 90 lb/ft3 Unqualified Epoxy coatings 5.05 ft3 94 lb/ft3 Total Coatings Volume Outside ZOI:

Consistent with the approved guidance of NEI-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.

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 Water Reactor Sump Performance Methodology," which calls for all of the coatings within a 10D 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?

FPL Energy Point Beach Response:

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.

Page 6 of 15

The revised calculation recognizes that a ?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 ff2 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 4D "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 4 0 ZOI around the limiting pipe break. Is this surface area bounded by the surface area of a 4 0 sphere?

FPL Energv Point Beach 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 40 ZOI. The resulting area is 400 ff2, contributing 1.3 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.

Question 17:

Considering your responses to the foregoing two MIS, 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?

FPL Energv Point Beach Response:

The total quantities of the coatings were provided in the responses to Questions 14, 15 and 16 above. The quantities in the initial GL 2004-02 response are no longer correct. The reduced quantities stated above are being used.

Page 7 of 15

Question 18:

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

FPL Energy Point Beach Response:

As stated in response to Question 14 above, 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 of 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-6016-02.04, Attachment 3, Sure-Flow@ Suction Strainer - Testing Debris Preparation and Surrogates, SFSS-TD-2007, Revision 4 (Proprietary) (ADAMS Accession Number not available).

Question 19:

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

FPL Energy Point Beach Response:

ltem (m) of Revised Content Guide for Generic Letter 2004-02, Supplemental Responses November 2007, dated November 11,2007, (ML073110278) requests licensees to:

"...Provide the information requested in GL 04-02 Requested Information Item 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 surface.

GL 2004-02 Recruested lnformation ltem 2fd) fvi)

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.

Page 8 of 15

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 Information 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-16406-PI 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. The following paragraphs are excerpts from the evaluation:

"...all piping diameters in the sump recirculation 1 injection flow paths are greater than I

.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-I 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."

Page 9 of 15

"Furthermore... all of PB-1 and PB-2 RG 1.97 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 debris, as well as particulate coatings debris that have since been reduced (see the responses to Questions 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.

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 1, 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 I B mechanical seals. These seals are rugged in their construction and capable of operating at elevated temperatures. The arrangement of the springibellows mechanism will not be affected by the suspended solids used in this evaluation for the specified mission times. John Crane Type I and 1 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 1 B mechanical seals uses a Page 10 of 15

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 Sl pump manufacturer related to the SI pump rotor-dynamics, it is assumed that this information is not available. Therefore, the WCAP-I 6406-P wear model is used for the pump vibration evaluation."

"The wear rate model in [VVCAP-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-16406-P for the each of the wear components during the mission time of 30 days.

Therefore, per [the WCAP-16406-PI criterion, the Sl 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.

Page 11 of 15

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.

FPL Enerav Point Beach Res~onse:

The previous calculation was completed in April 2006; two months after the issuance of industry guidance contained in WCAP-16530, 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 Chemical Effect Test (ICET) series of tests were based upon the 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-16530 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.

The calculation contained in Enclosure 6 used the spreadsheet tool distributed with WCAP-16530 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.

Page 12 of 15

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.I 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.

FPL Energy Point Beach Res~onse:

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 I

.2 and 2.2 used the same inputs with the exception of low sump level. Therefore, comparison of these two pairs of cases provides a reasonable correlation between sump level and total precipitant formed.

Page 13 of 15

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.

Question 23:

1 Please discuss why dissolution of concrete surfaces will not contribute significantly to the precipitate loading in the sump.

PBNP Unit I FPL Energy Point Beach Response:

PBNP Unit 2 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 Case 2.1 Case 2.2 43,317 ft3 348.3 Kg 22,995 ft3 169.7 Kg Case I.I Case 4.2 would be if trisodium ~ h o s ~ h a t e (TSP) was used as a DuTter. I nererore, tree calclum Ions rnar may Max sump volume Total Precipitant Mass Sump volume Total Precipitant Mass dissolve-into solution hill riot precipitaie out as calcium phosphate. This is demonstrated by the inputs (see Enclosure 6, Appendix A. I ) where 10,000 feet 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 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 140f 15

FPL Energy Point Beach Response:

Research conducted in response to GL 2004-02 established that the coatings on the Unit I 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 I 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) ZOI that may have an aluminum pigmented coating is the reactor vessel. As discussed in the response to Question I 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 ft", 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 lblin3 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.

Page 15 of 15

ENCLOSURE 2 FPL 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-6007-06, REVISION 5, JANUARY 8,2009 TOTAL HEAD LOSS - POINT BEACH NUCLEAR PLANT - UNIT I

& 2 23 pages follow

Total Head Loss - Pdmf Beach Nuclear Plant - Unit I & 2 Te&nidal Document No. TDI-6007-1)6 Revision 5 CALCULATION COVER SHEET

- - Calculatidn TDlbD0708_ - -. -

Technical Document Rev. No.

5 Addenda No.: MIA Calculation

Title:

Total Head Lass - Point Beach Nuclear Plant - Unit 1 & 2 Safety Related?

YES Calculation Verification Method (Check One):

Design Review Alternate Calculation

[11 Qualification Testing Scope of Revision:

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

Documentation of Reviews and Approvals:

TDISW7-06 Rev 5 Total Heed -.doc

Total Head Loss - Point Beach Nuclear Plant - Unit I (Ilr 2 Technical Document No. TDI-6007-06 Revision 5 CALCULATION VERIFICATION CHECKLIST Calculation-TitlerTohl Head Lass-< Point Beach Nuclear Plant - Unit 1 &- 2--'-.-

Revision: 5 I

I I 5. I Have applicable construction and aperating expience been considered?

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Were inputs correctly selected and incorporated?

Are assumptions adequately described and reasonable?

3.

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Are the applicable codes, standards and regulatory requirements identified and met?

17. 1 Are adequate handling storage, cleaning and shipping requirements specified?

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H ~ C the deslgi properly considered radiation exposure?

Are the acceptance criteria incorporated in the design documents sufficient to atlawuerificattion?

Have adequate preoperational and subsequent periodic test requirements been Page 2 of 1 @ '1

18.

9.

' 20.

I a m n o Note: This is PC1 farm 3060-3 Revision 3 1 I3 81 Are adequate identifi@bn requirements specified?

Are requirements for record prepara'tion, review, approval, retention, etc.,

adequately specied?

Has the appropriate Calculation Guideline Verification Chkcklist been reviewed and signed?

Have the design interface requirements been satisfied?

Was an appropriate design method used?

Is the output reasonable compared to input?

Are specified parts, equipment, and processes suitable for the required application?

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Total Head Loss - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TD1-6007-Mi Revision 5 1 TABLE OF CONTENTS

_ CALCULATION COVER SHEET CALCULATION VERIFICATION GHECKLIST TABLE OF CONTENTS 1.0 Purpose and summary Results 2.0 Definitions and Terminology 3.0 Facts and Assumptions 4.0 Design lnputs 5.0 Methodology 6.0 Acceptance Cribria 7.0 Calculation(s) 7.1 Clean Strainer Head Loss 7.2 Strainer Debris Laden Head Loss 7.3 PBNP Strainer Debris Laden Head Loss Summary 7.4 Debris Bypass I

8.0 Conclusions 10.0 Drawings ATTACHMENTS Point Beach Debris Laden Head Loss Temperature Corrected I

6 Values TABLES Table 1 Total Debris Laden Head Loss -Temperature Adjusted to Range 212 "F to 32 O F Table 2 Total Corrected Clean Strainer Head Loss (TCCSHL) Results, ft Table 3 'Clean Strainer Head Loss - Regression Formula, ft Table 4 ARL Tested Clean Strainer Head Loss Table 5 ARL Tested Debris Laden Head Loss Table 6 Water Dynamic Viscosity Table 7 Total Debris Laden Head Loss Table 8 Total Debris Laden Head Loss -Temperature Adjusted to Range 212 "F to 32 "F Originated By:

Date

Total Head Loss - Point Beach NucIear Plant - Unit I & 2 Technical Document No. TDf-6007-06 Revision 5 1 I

.O Purpose and Summary Results 1

The-US-Nuclear Reg~lator~Commission (USNRC)-in generic safety issue (GSI) 191 identified it was possible that debris in PWR containments could be transported to the emergency core waling system (ECCS) sump(s) following a main steam line break (MSLB) andlor a loss of coolant accident (LOW). It was further debmined that the transported debris could possibly clog the sump screenslstrainers and impair the flow of water, thus directly affecting the resultant operability of the various ECCS pumps and the oontainment 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 (551-191, utilities have implemented a program of replacing the existing ECCS sump screens or strainers with new and improved designs.

In order to address and resolve the specific issues associated with USNRC GSI-191 for the Point Beach Nuclear Plant - Unit 1 & 2 (PBNP-1/2) entered into a contract with Performance Contracting, Inc. (PCI). The primary objective of the contract was for PC1 to provide a qualiKxl sure-~lod' Suction Strainer that has I been specifically.designed for PBNP-la 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 I

Qualification-Report-is-a-compilation-of-thevarious-dowment5-and-~I~ulations I

I that support the strainer qualification.

As part of the PBNP-412 Qualification "~eport, PCI has performed a number of hydraulic calculations in support of the replacement sure-FIU#

Suction Strainer. I I

This calculation TDl-6007-06, Tofal Head Loss - Point Beech Nuclear Plant -

Unit I & 2 is one of a number of hydraulic calck~lations that specifically supports the design and qualification of the subject strainer.

This calculation addresses the total expected head losses across the suction strainer assembly that has been designed specifically for PBNP-1/2.

This expected head loss is the combined total of the clean head loss associated with fhe strainer and attached piping, and the debris head loss. The clean head toss I

was determined in calculation TDI-6007435, Clean Head Loss - Poinf Beach Nuclear Plant - Unit 7 & 2. The debris head loss is determined based on actual I

test results -for a PBNP-112 strainer that has been specifically corrected for the PBNP-112 Specification design-basis post-LOCA watw temperature. The tests were performed at the Alden Research Laboratory and independently verified by AREVA (Reference 9 4. The calculations are only pertinent to PCl's Sure- (

FIO@ Suction Strainer.

I

'WEB007-06 Rev 5 Total Hesd Loss.dw Date I [L lo.I

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

The PBNP units each have two (2) separate recirculation strainer assemblies that individually and specifically feeds either the 'A' or 'B' train ECCS and CS

. -....... - system,-.- Each hori~ontally oriented-recirculation strainer assembly is comprised- - -

of fourteen (14) modules each made up of ten (10) strainer disks for a total strainer area of 1,904.6 ft2, or a total of 3,809.2 ft? for each pair of strainers associated with one of the PBNP units. Flow leaves the strainers and enters a combination of pipe and fittings before discharging into the containment outlet.

PCI drawings [Drawings 10.1 - 10.91, inciusiv~ provide details of the subject I configuration.

Based on actual test resufts performed by PCI, it was determined that clean strainer head loss (CSHL) for the ~ u r e - ~ l o f l Suction Strainers is a function of 1 two (2) independent variables: (1) strainer internal core tube diameter and (2) water flaw rate exiting the strainer assembly.

The quotient of these two independent variables, in turn, results in one independent variable, which is exit velocity 0.

The Clean Strainer Head Loss (CSHL) depends on the specific plant conditions for PBNP-112. The results of the Total Corrected Clean Strainer Head Loss (TCCSHL) calculation cansidering these conditions, including uneer@inty, was calculated to be 0.560 frset of water. Full scale testing by ARWA at ARL found the actual CSHL to be 0.408 feet of water with the plenum head loss added.

The CSHL calculations account for the specific design of the PBNP-lf2 strainers.

The debris laden-head loss utilizes a-series of tests conducted-with a redu@d-- - -- -

scale strainer (with ammpanying 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 for each parameter). One parameter that is not changed is the approach velocity. it is kept the same as the full-scale design. The approach velocity is defined as the quotient of strainer flow rate and total surface area. The resultant value is 0.0026 I Ws, an extremely low appmch velocity when compared to the design value for the original ECCS screens. The head loss across a particular debris bed is a function of two hydraulic variables: approach velocity and water dynamic viscosity.

Accordingly, the strainer specific test results, utilizing accurately simulated post-LOCA debris and the design approach velocity, will be accurate for a given water dynamic viscosity, a parameter that is a function of wate~

temperature. Therefore, the test results require correction for the viscosity at the specified post-LOCA water temperature, 212" F in the case of PBNP-II'L. The test results will then be representative of the full scale strainer under specified post-LOCA oonditiions.

I Originated By:

Date -

mlSanOe Rev 5 T&I ~ s a d

&.d&

Pege5of19 1

Total Head Lcws - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TDI-6007-06 Revision 5 1 The results of the calculation are provided in Table 2. This calculation utilizes the results of clean strainer head loss testing previously conducted at the

- - Fairbanks Morse-Pump-Company and-the Electric Power-Research Institute's--..

(EPRI) Charlotte NDE Center for Prototype I and 11, respectively that is applicable to the current PC1 sure-FIO@ Suction Strainer. It also utilizes the I actual test results of the PBNP-2 strainer that were performed at the Alden Research Laboratory (ARL). The results of the subjed two tests form the basis for calculating the PBNP-In strainer total head loss, I

Table i - Total Debris Laden Head Loss - Temperature Adjusted to Range 212 O F to 32 O F Temperature "F Head Loss ft i

I This calculation does not address the subjects of possible air ingestion, potential I

voltex, and void fraction issues as they relate to the PBNP-1R strainer. These topics will be specifically addressed in calculation TDl-6007-07. Air Ingesfion, Vo&x & Void Fracfion - Pohl Beach Nuclear P lanf Unit I & 2.

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

I 2.0 Definitions and Teminology I

The following Definfth & Terminology are defined and described as they are utilized in this calculation.

Originated. By:

TMgWI-08 Rev 5 Total Head Lossda:

To'tal Head Loss - Point Beach Nuclear PIant - Unit I & 2 Technical Document No. TD1-6007-06 Revision 5 1

~ure=Flow@ Suction Strainer - Strainer developed and designed by Performance Contrading, Inc. that employs sure-FIOW@ technology to reduce

......... -. inlet approach velocity;.

I Emergency Core Cooling System (ECCS) - The ECCS is a combination 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.

1 Point Beach Nuclear Plant Unit "t 2 - also known as Paint Beach, PBNP-112, I

and PB-1/2.

I AREVA WP, Ine. - also known as AREVA. AREVA is contracted by PC1 to prepare and implement the Test Plan through Aiden Research Laboratory. ARL will implement the testing under the AREVA quality program.

Ald~sn Research Laboratory - also known as ARL. RRL is contracted by AREVA to perform the testing in their faciMty located at Holden, MA. The testing will be performed by ARL under the direction of PCI and AREVA.

Clean Strainer Head Loss (CSHL) - Is the calculated head loss for the Sure-lo@ Suction Strainer based on actual testing performed at the Electric Power Research-Institute's- (EPRI)-Gharlotte-NOE-Genter;--and-Fairbanks-Pump-Company Hydraulic Laboratory. The later te$ing did not involve any debris.

I i

Total Debris Laden Head Loss (Temperature - Corrected - ARL Test Results) (TDlJk) - Is the TCCSHL added to the A-DLHL for the sure-~lov$

I Suction Strainer based on the PBNP-112 testing that was performed at the Alden Research Laboratory (ARL). The PBNP-IR strainer testing performed at ARL is I

documented in [Reference 9.41.

I Total ~orketed Clean Strainer Head Los. (TCCSHL) - Is the total head loss associated with the complete sure-FIOW@ Suction Strainer installation configuration for PBNP-1/2 (i.e., strainer and connecting piping and fittings) including uncertainty.

ARL Test Results - Debris Laden Head Loss - Temperature Corrected (A-DtHL) - Is the temperature corrected head loss for the PBNP-IR sure-low@'

Su&&ipn Strainer based on the ARL test results utilizing the desig~.basis debris loading [Reference 9.41-Originated By:

TDl.6007-06 Rev 5 Tat& Head Losadac Date

Total Head Loss - Point Beach Nuclear Plant - UniP: 1 8 2 Technical Document No. TDI-6007-06 Revision 5 1 3.0 Facts and Assumptions

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

utiiized in the preparation of this calculation.

3.1 For the specified minimum post-LOCA water temperature of 212" F, the containment air pressure is 44.7 psia [F].

3.2 A flow velocity of 0.0026 fps would be characteristic of the PBNP-112 I strainers, through a debris bed consisting of fibers and particulate is 100%

viscous flow.

Accordingly, the head loss is linearty 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 of the full scale production strainer so long as the same scaling factor is used for strainer area, water Row rate, and debris quantities. The scaling factor is defined as ratio of the surface area of the scale strainer and the surface area of the full scale production strainer [A].

3.4 The head loss resulting from flow through a fiber - particulate debris bed at the approach velocity for the PBNP-112 strainer (0.0026 Ws) I fRebrence-9.3],is-?

00%-viscous-flow, as-opposed-to-inertial-flow.-As viscous fiow,, head loss is linearly dependent an the product of viscosity and velocity; Therefore l -. to

.. adjust

,..-.... the measured

.... head loss across a debris..

bed with colder water, a ratio of water viscosities, between the warmer 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 foss with water at the specified post-LOCA temperature [A].

3.5 The total strainer head loss can be calculated by taking the sun of the calculated value of the Clean Strainer Head Loss [Reference 9.31 and the temperature adjusted, tested debris head loss [A].

3.6 The PC1 sure-log Suction Strainer installations for PBNP-If2 are the same. However, there are a number of differences with regard to the strainer discharge piping configuration for each of the four (4) strainer installations. Based on an assessment of each of the four (4) strainer discharge piping configurations, the piping configuration associated with PBNP-1 Strainer "Bn would result in the greatest head loss due to this specific strainer configuration having the greatest equivalent pipe length (i.e., combination of straight pipe length and number and type of fwngs).

Accordingly, the PBNP-I Strainer "0" piping configuration will be I

T D I m - 0 8 Rev 5 Tdal Head doe 7 /L/O 7 Date

PERFOHMANCL C:QNIIZAC:IlNCriNC Total Head LOSS - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TD1-6007-06 Revision 5 1 3.7 Utilization of the PBNP-IR testing pragram performed at ARL and the subsequent test data and results [Reference 9.41 to support PBNP-1M calculations are based on the PBNP-I12 Projed specification [Reference 9.11 Ifl.

conseryatively utilized as the basis for PBNP-1R to bound both units and all strainer discharge piping configumtions with respect to strainer clean head.. loss [F-8 A]..- -..

3.8 Any and all references to or discussion of the PBNP-I12 strainer testing, test resub, and similar related activities and discussions, actually means the PBNP-112 strainer testing at ARL and the subsequent test resutts fleferance 9.41 [F].

.I---.

3.9 PC! has assumed that unknown piping, tubing, or openings added after the strainer installation are not directly connected to the PC! sure-low@

Strainer and are sealed (he., fluids andlor gases cannot enter andlor exit through the openings) [A].

3.12 The Design Basis minimum specified post-LOCA water temperature is 222°F. The 212 "F tern~erature will be utilized to evaluate the tntal head 3.10 The input data used in MS Excel spread sheets (if applicable) was vetifred by comparison to the design drawings and associated dimensions. The calculations resulting in output data are independently verified by hand calcuiation~Theref~re~a-MS-~cel-spred-sht-is-nvenient but not relied upon as analytical software [F].

loss. However. PBNP-1-12 has requested a series of head loss values for water with temperatures between 212 "F and 32"F, at 20 degree increments. The Total Debris Laden head loss will be determined at 212 "F and utilize the tested clean strainer head loss results to determine the final Totai Debris Laden head loss. PC1 will utilize a temperature correction correlation to obWm the subject head losses for the range between 212 O F and 32 "F temperatures [F].

3.13 Reynolds numbers are catculated in attachment A1 for temperatures between 32 O F and 212 O F using the flow and piping details from the CSHL calculation Reference 9.31 and shown in Attachment I, Tabk A1. [GI.

The foll'owing combination of PBNP-I& and PC1 Design Inputs were utilized in I the preparati6n of this I

Originated By: -

calculation.

PERf ORMANC t C.C>NlI?ACI ING INC Total Head Loss - Point Beach Nudear Plant - Unit I

& 2 Technical Document No. TD1-6007-06 Revision 5 1 4.1 Point Beach Nuclear Plant Specification, Specification No, PB-681,

.... Replacement uf Containmenf Sump Sc~ens, Revision I, August 25, 2005.-.. _.

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

4.2 Performance Contracting, Inc. (PCI) Calculation TDI-6007-02, SFS Sudbce Ama, Flow and Volume Calculation, Revision 1 [Reference 9.18.

This document provides relevant dimensions and other information specifically associated with the PBNP-1I2 strainers.

4.3 PC1 Calculation TDI-6007-03, Cam Tube Design - Point Beach Nuclear Pknf - 712, Revision 0 [Reference 9.21. This document provides relevant data with regard to flow rate in the PBNP-I@ strainer.

4.4 PC! Calculatian TDI-6007-5, Clean Head Lws - Point Beach Nuclear Plant - ID, Revision 4 [Reference 9.31. This document provides the I head loss associated with the "cleann PBNP-1E strainer and attached pipe and fMings.

4.5 AREVA Engineering Information Record, Document Identification No. 66-9093957'-00& Point B e a c h - T e s t - R e - ECCS-Strainer-Performance---

Testing [Reference 9.41. These documents provide the method and value of the testededdebris head loss and the mechanism of adjusting the tested-- - -

debris head Loss to the specified post-LOCA water temperature.

I 5.0 Methodology I

PC1 utilized two (2) distinct methodologies based on the entire strainer assembly I configuration to determine the maximum thin bed head loss for this calculatian:

(1) calculate the Clean Head Loss for the PBNP-I12 strainer msrfemnce 9.33 and (2) determine the peak design basis head loss based on reduced scale strainer testing utilizing the PBNP-IM specified design basis water temperature of 21 2' F [92lefemence 9.11 (adjust from the test wqter ternperature to the specified I

water temperatQre) and the PBNP-1/2 specific debris mixture. The individual I

I head loss results obtained are added together to obtain the total design basis head loss for the entire strainer assembly configuration.

The quantiiy of fiber and debris used in the scale strainer testing is based on the debris load stated in {Reference 9.61 with a 75% fiber,reduction. PC1 believes that the assumptions are conservative and are supported by the PBNP-1R test

.4]1. Debris testing is then used to determine if the I

&?-

Date 1

TD160Q796 Rev 5 Total Head W d o c Page 10of 19 I

Total Mead Loss - Point Beach Nuclear Plant - Unit I

& 2 Technical Document No. TD!8007-06 Revision 5 I PBNP-I12 specified [Reference 9.11 that the total debris laden strainer head loss be calculated at a temperature of 222 O F in order to meet the required design basis NPSH requirements, and further specified [Reference 9.a that a range of head loss values be determined between 212 "F and 32 "F at 20 degree increments. The head loss values for the all range of temperatures will be presented In Table 8.

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

. baunds the proposed sire. and.. design for. the. actual... PBNP-lli! strainer.... PC1.

believes that the assumptions are conservative and are supported by the PBNP-14 strainer test results at ARL [Reference 9.41; The DLTHL-TC includes the strainer, strainer discharge, and addressing all possible debris loading combinations. This calculation addresses the possible debris loading combinations, and calculates the head loss associated with the strainer and the strainer discharge flow into the sump.

NOTE: The PC1 ~ u r e ~ l o w @

Suction Strainer installation for PBPIP-'iC2 is ]

I very similar in nature with only a slight dimrence with regard to the I

strainer discharge piping configuration. Accordingly, tfte discharge piping I

configuration differences am greater far PBNP-1, and its associated I

configuration will be utilized 4tro $aurrd[ bath units with respect to strainer I

clean head lass.

I PC1 has optimized its design of the sure-low@ Suction Strainers for PSL-2 to ensure preservation of head loss margin.

7.0 Calculation(s)

In order to determine the TDLHL; two (2j distinkt-c%lculiition methijdblog~es are.'

TDI-SW7-W Rev S Tml Head Loss,dac i

A Date employed as described in section 5.0 Mathodolagy. One methodology is utilized to separately calculate the head loss for the bare strainer, attached pipe and fittings, and the secund methodology is used to determine the Tatal Debris Laden Head Loss - Temperature Corrected (TDLHL) design basis head loss based on PBNP-112 specific reduced scale strainer testing using a full sized representative strainer module with debris generation allocation mixture (A-DLHL).

Total Head Loss - Paint Beach Nuclear Plant - Unit 1 8 2 Technical Document No. TD1-6007-06 Revision 5 1 7.1 Clean Strainer Head Loss

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

lass,(CSHL) for the PBNP-112 strainer in wefkrence 9.31.

The total CSHL includes the expected head losses from the strainer (bare),

attached strainer discharge piping and faings connecting the strainer to I

the sump pit, and that associated with the water leaving the strainer discharge pipe as it enters the sump pit.

I The TCCSHL below includes a strainer only head loss calculated using the PC1 regression.formula as presented in Reference 9.41. The strainer regression formula value for head loss and its temperature corrected value for 212 O F are presented in Table 3.

This value will be used later to determine the total debris laden head loss. Temperature correction is calculated using methodology provided in Section 7.2.2.

I f

t Table 3 - Clean Strainer Head Loss - Regression Formula, R 11 7.2 Strainer Debtis Laden Head L ~ s s I

Table 2 - Total Corrected Glean Sbainer Head Loss (TCCSHL)

Results, ff The PBNP-IR strainer modules were-sized biased upon meference 9.1.h].

The amount of and the makeap of the debris that is specific to PBNP-112 was provided in [Reference 9.61.

The debris mixture specified in mefewnce 9.11 was further analyzed by utilizing [Reference 9.121 to develop the adual debris mix (i-e., debris quantity and type) %r the testing of the PBNP-'ID specific strainer [Rtafemncrts 9.q.

I I

TCCSHL @ 212 "F The PBNP-1R Clean Strainer Head Loss tests performed at ARL are summarized in Table 4.

I 0.560 The GSHL based on the test result from ARL was temperature corrected to the PBNP design basis temperature (i.e., 212 OF). See Section 7.2.2 for temperature correction methodology.

Date I

TDI4007-08 Rev 5 Total Head tcrss.doc Page12ofiB I

Total Head Lcrss - Paint Beach Nuclear Plant - Unit 1, & 2 Technical Document No. TDI-6007-06 Revision 5 1 The PBNP-IR Debris Laden Strainer Head Loss test is summarized in Table 5.

I AREVA Test No, 6 [Reference 9.41 is the Design Basis test for PBNP-1R.

The PBNP-1I;C Design Basis test is intended to show recirculation at 2200 gpm with a water level above the top of the PBNP-112 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 w~femnce 9.41.

I -

I Table 4 - ARL Tested Clean Stminer Head Loss

_ _ _ - - - I _ -.

I 7.2.1 Temperature Correction Strainer Debris Laden Head Loss Corrected Clean Strainer Head Loss, ft of water (212 O F )

0.0417 I

The dynamic viscosity of the specific ARL test water temperatures and the PBNP-I12 Design Basis temperature is taken from

[Reference 9.91.

table 6 provides 'a summary of the dynamic viscosities associatqd wkh the various test and ~esign Basis water temperatures that are utilized in this calculation.

Clean Sbrainer Head Loss, ft of water 0.090 Test Strainer Flow, gprn (Scaled) 470.7 Date &

i Ave.

Temp. {OF) 111.1

-I Table-StARL-Tasted.Debris-Laden-Head Loss TDI-6007-06 Rw 5 T9t?ll Hesd Lms.dw Pagel3of19 I

- t--g$&*&

Test No. 6 - 170.42 gprn scaled Row d

Strainer Debris Laden Head Loss HL correc@d to 212 O F (Ft of Water) 3.066 Test

Temp, OF,,

HL - Ft of Wa@r 8.448 I

PERFORMANCL C:C)NWAC:IINCI INC Total Head Loss - Point Beach Nuclear Plant - Unit 1 & 2 Technical Document No. TDI-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 linearly dependent on the water's-dynamic viscosity: - -The PBNP Design Basis water temperature is 212 O F [Reference 9.11.

The debris head loss requires correction to this temperature to determine the head loss at the PBNP-fR Design Basis temperature of 212 "F. The strainer debris laden head loss for low velocity water flow through a debris bed of fibers plus particulate is linearly dependent on the water's dynamic viscosity [Reference 9.10l;.

I Table 6 - Water Dynamilc Viscosity 11 7.2.2 Post-LQCA Temperature Correction Shiner Debris Laden Head L -

A head loss correction, utilizing Assumption 3.q which is based an the standard debris head loss equation [ReFerrance 9.111 can be used to calculate a temperature adjusted debris head loss, HLTA.

The HLrr adjusted temperature can be calculated by taking a ratio of dynamic viscosity values at the two different temperatures being considered (i.e., the. test water temperature and the PBNP-1R specific post-LOCA sumpwater temperature).

I Equatidn 1 HLTA = HLoL,~

(VST I II TT)

Originated By; TDlBW7-06 Rev 5 Total Heed Lwadoc Page 14of 19 I

Total Head Loss - Point Beach Nuclear Plant - Unit 1 PL 2 Technical Document No. TW-6007-06 Revision 5 1 Where HLDLc= Debris Loaded Head Loss, ft ST

= dynamic viscosity at the post-LOCA specified - I- -- - - -

temperature

=dynamic viscosity at the average tested temperature HLTA = temperature adjusted debris head loss, Pt

]

The HLTA, as calculated above, is added to the clean strainer head loss that results in the DLTHL-TC for PBNP-lfZ based on the specified post-LOCA Design Basis temperature.

7.3 IPBNP-112 Strainer Debris Laden Head Loss Summary Table 7 summarizes the bounding values of head loss discussed above.

All head losses are in feet of water. It was also oonsenratively assumed to add 6% for uncertainty and 10% fur strainer discharge and collection head loss associated with the Clean Strainer Head Loss (CSHL) calculations to address any non-conservatism inherent in the use of standard head loss correlations WeFerence 9.31. The Clean Strainer ~ e a d Loss values are based-an- [R&rein'ee-9;3];th;e-tested-strainer-debri8laden-head-loss-is based on $ection 7.2, and the temperature corrected debris laden head loss for: post-LOCA conditions is based.on Saction.-7.2.2,-

PBNP-I/;! has requested a series of head loss values be calculated for water with temperatures between 32 O F and 2?2"F, at 20 degree increments. The Total Debris Laden head loss will be determined at 212 OF and utilize the tested clean strainer head loss results to determine the final Total Debris Laden head loss for this range of temperatures. PC1 will use temperature correction correlation methodology presented in Attachment I to calculate the subject head losses for the range between 32 OF and 212 OF temperatures. See Table 8 for resutts of head losses calculated for the specified range.

-iii<e3 ell?

c f...b z g.s

!kg

• ET iG

$ 4 7Zig 0

sz a,

c c m a]

1 E

CT) 0 3 i! g

.c -

0 iu le:

Q.o,

~1 Q) c

, a 6 Js I--

c Q]

3'-

a A

I II ua (II CR E?

3 4

4 S

Total Head Loss - Point Beach Nuclear Plant - Unit 1 4% 2 Technical Document No. TDI-6007-06 Revision 5 1 7.4 Debris Bypass As part of the PBNP-112 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 strainer. This was done in order to determine the sire, quantity 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 debris bypass data is not part of the scope of this technical document.

The debris bypass analysis results can be found in the ARWA Test Report EReferencca 4.41.

8.0 Conclusions I

This calculation and supporting portions thereof, considered all of the previous testing that has been performed for the various PCI sure-low@ Suction Strainer, including uncertainty. The temperature corrected head loss associated with the debris only on the strainer is 3.066 feet of waabr at 2G? OF. The predicted result for total debris laden head loss, the sum of the calculated clean strainer head loss including uncertainties and the strainer debris laden head loss is 3.474 feet of water at 212 OF.

It was concluded that this specific calculation corn letely supports the qualification, installation, and use of the PC! Sure-Flow ck' Suction Strainer for I-..

Point Beach Nuclear Plant - Unit 1 & 2 without any issues or reservations.

I TDI-6007-08 Rev 5 Total Head b. d d e

Total Head Loss - Point Beach Nuclear Plant - Unit I & 2 Technical Document No. TD1-6007-08 Revision 5 1 9.0 References

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

Replacement of Containment Sump Scmens, Specification No. PB-681, Revision 2 I

9.2 Performance Contracting, lnc. (PC!) Calculation TDI-6007-03, infernal Core Tube Slot Design for PBNFs Suction Strainers, Revision 0 I

9.3 Performance Contracting, Inc. (PCI) Calcuiation TDI-6007-05, ~iean Head '

Lass - Point Beach IVuck?ar Planf - Unit -f&, Revision 4 I

9.5 AREVA NP Engineering lnformation Record, Document Identification Mo.

51-902151 3-000, Point Beach Unifs 1 & 2 ECCS Shiner Performance Test Plan, Revision 17 9.6 PBNP Letter No. NPL 2008-01 62, Design Information Transmittal (DIT) in '

Support of Sump Strainer Qualification Testing the week of July 14, 2008, Juiy 9,2008 9.4 AREVA Engineering Information Record, Document Identification No. 66-9093957-002, Point Beach Test Report for ECCS Shiner Perfinance

Testing, 9.9 Spirax Sarco USA Webpage (h~p:llwww.s~iraxsar~.c~rn/us/resources)

I 9.7 PBNP Letter No. NPL 2008-0264, Design Information Transmittal (DIT) in

.......... support.of.F"G1-@!cu!ation TD!+W7-96 RK?. 5, Octeker St.2FI08-9.8 Crane Technical Paper No. 410, Flow of Fluids fhmugh VaIves, Fiffings, and Pipe, 4988 9.10 USNRC NUREGfCR.6224 "Correlation", publicly available software 9. 1 NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology", Rev. 0, December, 2004 9.12 Performance Contracting, Inc. (PCI) Calcu(ation, TDI-6007-02, SFS Surface Area, Flow and Volume Calculation, Revision 2 9.13 Fluid Mechanics With Engineering Applications, Robert L. Daugheiiy and Joseph B. Franini. Seventh Edition, 1977, McGraw-Hill Book! Company, lnc.

Originated.

i/b/oCi Date TDI-600746 Rev 5Totel Head Loss.doc PagelBof19 1

Tofal Head Lass - Point Beach Nuclear Plant - Unit I

& 2 Technical Document No. TDI-6007-06 Revision 5 1 10.0 Drawings

-.. - -- - -... ---.-16.-1-- SESPBI-GA-00,-

~evision 9, Point l3each.Unit-I, sure-low' Strainer, Recjm Sump System 10.2 SFS-PBI-GA-02, Revision 9, Point Beach Unit 1, sure-low'@ Stminer, B Sfminer 10.3 SFS-PBI-GA-03, Revision 9, Point Beach Unit 1, sure-low@ ~ i r a i n e ~

A Stminer 1 0.4 SFS-PB I-GA-04, Revision 6, Point Beach Unit 1. sure-low@ Strainer, Piping B Layout 10.5 SFS-PB1-GA-05, Revision 9, Point Beach Unit 1. sure-FIOW' Shiner, Piping A Layout 10.6-SFSPBI-PA-7100, Revision 4, Point Beach Unit 1. sure-FIOW@

Strainer, Module Assembly 40.7 SFS-PB2-GA-00, Revision 3, Point Beach Unit 2, sure-~lov$ Sfminer, Recirc Sump System Layout 10.8. SFS-PB2-GA-02, Revision 9, Point Beach Unit 1, ~ u r e - ~ l o w ' @,, ~ f ~ ~. ~ ~

A Slmher-10.9 SFS-PB2-GA-03, Revision 9. Point Beach Unit 1, sure-low@ Strainer, B Shiner 10.1 D SFS-PB2-GA-04, Revision 5, Point Beach Unit 2. sure-low@ Sfminer, Piping Assembly Layout 1 0.1 1 SFSPB2-PA-7100, Revision 3, Point Beach Unit 2, sure-low@ Sfminer, Module Assembly qt9,/O7 Date

PE liFORMANCL C:C)NIIZAC:IINC; INC Total Head Loss - Paint Beach Nuclear Plant - Unit 1 8 2 Technical Document No. TDI-6007-06 Revision 5 1 Point Beach Debris Laden Head Loss Temperature Corrected Values.......

Per Section 7.3, PBNP requested that total debris laden head losses be calculated for a range of temperatures. The total debris laden head loss is calculated by adding the ARL test CSHL, the ARL test debris laden head loss, and the piping head loss calculated in the CSHL calculation [Reference 9.31.

This calculation has already provided these values at a design temperature of 212 O F. To calculate the head losses for the PBNP specified range of temperatures, PC1 will use the following methodology:

A.

The CSHL value is calculated [Reference 9.31 using the PC1 derived regression equation.

The equation uses kinematic viscosity to address the various temperatures. The CSHL value used in this calculation 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 utiiizing the' kinematic viscos'ky, as addressed in the PC1 regression equation. The CSHL value calculated at 212 O F will be adjusted for the range of temperature values utilizing kinematic viscosity as follows:

VAT

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

1 VDT-

= Kinematic. viscosity. at the post-LOCA design temperature (21 2 O F )

HLcsSrA

= temperature adjusted debris head loss, ft I

Kinematic viscosity values were calculated using the following equation; I

Equation 2 v = p 1 p I

p

= dynamic viscosity (IbR-s)

P

= density (lWfts) v

= kinematic viscosity (@Is)

The dynamic viscsosity and density values were taken from peference 9-91 for values above 32 OF. Dynamic viscosity and density values for 32 "F were taken I Originated By:

I TD16W7.06 Rev 5 Total Head Loss-doc

Total Head Loss - Point Beach Nuclear Piant - Unit I & 2 Technical Document No. TDI-6007-06 Revision 5 1 Equation 3 HLpl, f UD v2f2g I

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 tables on page A-25 in Crane

[Reference 9.81 after 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 calculated using the following equation:

Equation la Re = V D I v I

The piping head loss is calculated in the GSHL calculation [Reference 9-31 using values for 16 inch piping having a fluid velocity. V = 3.683 Ws and a piping diameter, D = 1.302 R Values for kinematic viscosity, v for the temperature range were taken from [Refemce 9.9 and 9.131 and are included in TableAl.

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

HL

= f x constant 0,3665

= 0.012 x constant Constant

= 30.542 From the HL equation above, the terms UD v2/2g are assumed to be constant for this temperature adjustment. From Table b in section 7.3 of this calculation, the piping head loss total at 212 "F is 0.3665 ft. The friction factor, f, for 212 OF is 0.012. Knowing the head loss and the friction factor, the constant term can be calculated as:

This constant term, along with the friction factors from Crane will be used in the HLPl, equation to calculate the various piping head losses within the temperature range specified in Table A-I.

C.

The debris laden head loss also requires temperature adjustment. As stated in section 7.3 of this calculation, debris head bss can be temperature adjusted using the following methodoldgy; Equation, 5 HLDL,TA

= HLDL (VAT / lll DT)

Originated By:

TDWW17d6 Rev 5 Totel Head Lws.doc Atlachrnentl,Pege2oF4 1

Total. Head Loss - Point Beach Nuclear Plant - Unit 4 & 2 Technical Document No. TD1-6007-06 Revision 5 1 Where HLM, = Debris Loaded Head Loss, ft I

pm

=dynamic viscosity at the post-LQCA design temperature

~ A T

. = dynamic..vis~~sity at the post-LOCA adjusted.

temperature HLDLJTA

= 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 laden head loss, from C above, to calculate the DLTHL-TC for PBNP-112 based on the specified post-LOCA Design Basis temperatures.

Table A-I provides a summary of the final DLTHL-TC values as well as the various reference design inputs.

Originated By:

Date I / 6 / 0 s TDI-800746 Rev 5 Tab1 Head Loa8,doc, Page 3 of 4 1 i

TDIS00756 Rev 5 Total Head Loss.doc Total Head Loss - PointBeach Nuclear ~lanf - Unit.1 8-2 Technical Document No. TDI-6007-06 Revision 5

ENCLOSURE 3 FPL 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) 30 pages follow

Calculation Number: PCI-5344-SO4 Calculation

Title:

Structural Evaluation of Containment Emergency Sump Strainers Client:

Performance Contracting Inc.

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

Title:

Point Beach Strainer Qualification Safety Related Yes No Approval Signature /

Signature 1 Initials of Revision Affected Pages Revision Description Date Preparers & Reviewers I

i I

I I

I Form 3.1-10 Rev. 1

I PROJECT NO:

PCI-5344 CLIENT: Performance Contracting Inc.

I I CALCULATION TITLE:

Structural Evaluation of Containment Emergency sum^ Strainers I

I I

1 x 1

( 2. System Description 1

T ID, REV AND/OR I 1. General Design Basis 1 4. Electrical Discipline Input X

I I

1 x 1 I 5. Mechanical Discipline Input I I

3,5,9,13,22,23,27,31,46 X

3' Design from related equipment vendor 1 6. Control Systems Discipline I I

I I

X I

X 10$ 18,24,28,29,34,35,48; 51 I

I 1 8.

specifications I 1,2 I

X I

I I

9.

Vendor Drawings

10. Design Standards

11. Client Standards X

Input

7. Structural Discipline Input 1

12: Checked Calculations 8,24,27,28 X

I I

4,7, 12, 16, 17, 19,20,26,30,32, 33,34,35,38,39,44,47 1 13. Other (spec@)

1 21,25 (AES QA Files)

DATE:

09/25/2008 REVIEWER'S SIGNATURE:

DATE:

09/25/2008 APPROWR'S SIGNATURE:

PROJECT NO: PCI-5344 CALC. NO:

PCI-53444304.

Revision 0 I

I I

I I

REVIEWER TO COMPLETE THE FOLLOWING ITEMS:

I YES I NO 1 N/A I COMMENT I

8 I

I I

A. Properly Identified?

B. Properly Applied?

1. Has the purpose of the calculation been clearly stated?

2.. Hwe the applicab1e codes, standards and =gulatory =*men@

been:

3. Were the inputs correctly selected and used?

x I

I x I r,-

,~i~,g,$~~fJ:$27&::F++

,"I g$z +, ;, ; i ; ~ ~ $ $ ~ ~ ~ ~ ~ ~ ~ ; ~ ~ ~ $

,.jgg;5ggjgg~~;&~j$g~&@gggg&i~5F&~3aGTk

~+~~3~i2~7i~x7a~.b;x;i~~

&?a*,nL,%2,x k+A~~~qfi5~;,pk:~~!@$~~

4. A. Was Design Input Log used?

B. If4A is No, provide Manager's si-e in Comment column to signify approval of Design Input Documents used in the calculation.

5. Are necessary asm~ptions adequately stated?

6. Are the assumptions reasonable?

7. Was the calculation methodology appropriate?

8. Are symtjols and abbreviations adequately identified?

9. Are the calculations:

A. Neat?

B. Legible?

C. Easy to follow7 D. Presented in logical order?

E. Prepared in proper format?

10. Is the output reasonable compared to the inputs?

11. If a computer program was used:

A. Is the program listed on the ASL and has the SRN been reviewed for any program use limitations?

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

C. Were codes properly verified?

D.,Were they appropriate for the application?

X I I

E. Wee they comctly used:

F. Was data input correct?

G. Is the computer program and revision identified?

Form 3.1-4 Rev. 3

TABLE OF CONTENTS Safety Related Yes I2.d No u 1

2.0 Methodology.................................................................................................................................................. 5 I

Date: 09/25/2008 1

3.0 Acceptance Criteria 9

I 1

4.0 Assumptions................................................................................................................................................

14 I

1 5.0 Definitions and Design Input...................................................................................................................... 15 1

5.1 Material Properties 5

5.2 Strainer Geometry and Dimensions.............................................................................................. 17 Weight Calculations 24 Strainer Loads 32 Calculation of Acceleration Drag Volumes and Hydrodynamic Mass........................................... 34 Calculation of Mass Distribution on Strainer Components............................................................

55 GTSTRUDL Model 61 GTST. RUDL Res~ilts 93 Disk Pressure Loads 98 Core Tube Evaluation 103 Perforated Plate Evaluation 108 Wire Stiffener Evaluation 123 Core Tube End Cover Assembly Evaluation 124 Weld Evaluations 133 Rivet Evaluations......................................................................................................................... 140 Mounting Evaluation 145 Module-to-Module Sleeve and Latch Connection 169 Lifting Load Case Evaluation 175 Outage Load Case Evaluation 176 1

7.0 Results and Conclusions 177 I

1 8.0 References 179 I

Attachment A GTSTRUDL Run (Run time Mon Jul 10 13:48:48 2006) 1 279 Attachment B.

ANSYS Run for Gap Disk 1.

1202

. Attachment C.

Testing of 3/16 Blind Rivets (Reference [18])........................................................................... 1.

13 Attachment, D - Piping Strike and Latch Test (Reference [28])............................................................................ I.

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])

1 14 Attachment H - Perf Plate Thickness Data from Hendricks Book (Reference [35])........................................... 1 - 1 Attachment J - ACI Structural Journal (Reference [44J) 1 -22 Attachment K.- Lehigh Testing Laboratories Test No. G-4-27(Reference [48])................................................ 1 - 4 Attachment L - Lehigh Testing Laboratories Test No. F-19-32 (Reference [51]).............................................. 1 - 3 Form 3.1.3 Rev 2

CALCULATION SHEET Date: 09/25/2008 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 supporting structures associated with the new strainers.

1 2.0 METHODOLOGY 1.0 PURPOSEIOBJECTIVE I

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 requirement, 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 a s specified in Section 5.4.7 of DG-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. 11'- 2") per Reference [46]. The piping "dry" state with its associated mass being much less, will not be considered a s it is less severe than the "wet 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 recirculation water. Based on Reference [8], 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 121). These loads are applied to the strainer through base motion response spectra as detailed in the Seismic Design Criteria Guideline DG-C03 (Reference [I 51).

I Form 3.1-3 Rev 2

CaIc. No.: PCI-5344-SO4 Client: Performance Contracting Inc.

Revision: 0 1

Safety Related Yes No

/ Date: 09/25/2008 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 223 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 (i.e. closely spaced modes). An earthquake duration of 30.24 seconds was used in the analysis per DG-C03, 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 [15], 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 as a "bolted steel frames" structure and the damping values for seismic loads are taken as 2% for the Operating Basis Earthquake (OBE) and 5% for the Safe Shutdown Earthquake (SSE) as required by Seismic Design Guide DG-C03 (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 [27].

The pressure load acting on the strainer is the differential pressure across the strainer perforated plates in the operating condition. Conservatively, this is taken as the hydrostatic pressure associated with the maximum allowed head loss through the debris covered strainers. This is defined as a minimum of 10 feet of water in DIT-008 (Reference [46]).

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 maximum operational inlet temperature to the RH Exchangers of 250 O F (Reference [I]).

1 1

Form 3.1-3 Rev 2

CALCFTLATION SHEET Safetv Related Software MathCad software is used to generate the calculations. All MathCad calculations are independently verified for accuracy and correctness as 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. GTSTRUDL 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 this 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 JOINT 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 as planned.

MEMBER This command applies a specified temperature increaseldecrease to a given TEMPERATURE member. This command is used as 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 and/or 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 for'a certain direction of motion. This' command was verified manually by listing the dynamic mass summary and comparing the total dynamic mass in each direction to the calculated masses.

I I

I Fonn 3.1-3 Rev 2

TABLE 'RBARS' TABCE 'BARS' TABLE 'ROUND' TABLE 'MYCHAN' Safety Related Yes No Date: 09/25/2008 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 section properties when code checking, the properties are printed out for selected members defined by this command and those properties are verified manually.

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

and channel shapes. The members that are defined by these tables are subjected to loadings and then code checked in GTSTRUDL. These tables are verified in the same fashion a s for the PIPE 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 b e negligible). Also, steel cross sections that were not available in the GTSTRUDL cross section libraries 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 outputting the computed properties of the 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 as required.

Form 3.1-3 Rev 2

Fonn 3.1-3 Rev 2 CALCULATION SHEET Safety Related Yes N o 0 Date: 09/25/2008 3.0 ACCEPTANCE CRITERIA The strainer components shall meet the requirements of the strainer design specification PB-681 (Reference

[I]). As stated in PB-681, the detailed evaluations are to be performed using the rules, a s applicable, of ANSIIASME B31.I Power Piping 1998 Edition through 1999 Addenda (Reference [5]).

The strainers are classified a s "other pressure-retaining components" a s described in Paragraph 104.7 of the B31.I Code (Reference [5]). 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 831.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 B31.I Code is Alternative D, which allows for "detailed stress analysis, such a s the finite element method, in accordance with the ASME Boiler and Pressure Vessel Code, Division 2, Appendix 4, except that the basic material allowable stress from the Allowable Stress Tables of Appendix A shall be used in place of S,."

Section Ill, Subsection NC of the ASME 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 B31.I (perforated plate, and internal wire stiffeners). Some parts of the strainers (radial stiffeners, connecting rods, edge channels, seismic stiffeners, etc.) are classified a s part of the support structure. These types of components are covered under the AlSC Code (Reference [9]). Additional guidance is also taken from other codes and standards where the AlSC does not provide specific rules for certain aspects of the design. For instance, the strainers are made from stainless steel materials. The AISC Code does not specifically cover stainless steel materials. Therefore, ANSIIAISC 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 strainer also has several components made from thin gage sheet steel, and cold formed stainless sheet steel. Therefore, SEIIASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members", (Reference 1311) 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 AlSl Code (Reference [22]) where the ASCE Code is lacking specific guidance. Finally guidance is also taken from AWS Dl.6, "Structural Welding Code - Stainless Steel", (Reference [23]) 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.

The core tube is evaluated as piping per B31.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.I 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 [3]) is used to calculate the 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 B31.1 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 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.

The seismic stiffeners, external radial stiffeners and the mounting hardware are evaluated to AISC 9th Edition (Reference [9]) as permitted in paragraph 120.2.4 of the B31.I Code (Reference [5]). The analysis of the anchorage to the containment concrete slab will be in accordance with the Hilti technical Guide (Reference tlQ1).

Load Combinations The applicable load combinations for the strainers are those for Section 6.7.1 of DG-MI0 (Reference [14]) and 6.0 of DG-MO9 (Reference [I I]).

Load Condition Combination (I a) Normal Operating DP + DW (I b) Normal Operating (OutageILift Load)

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

I Form 3.1-3 Rev 2

CALCULATION S m E T 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 AlSC 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. Flood 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 as described in Section 2.0.

Core tube The core tube is evaluated as piping per B31.I Paragraph 104.8 as applicable. Since the 931.I does not explicitly identify how to incorporate the Emergency SSE loads, PBNP uses ASME Section Ill as a guide as discussed in Section 6.0 of DG-MO9 (Reference [I I]).

831.I Ea. No Load Condition

. Load Combination Allowable Stress 11 Normal DW 1.O Sh 12 (OBE)

Upset DW + OBE 1.2 S, 12 (SSE)

Emergency DW t-SSE 1.8 Sh Strainer Pressure Retaining Plates For the pressure retaining plates, such as the perforated plati the 831.I Code does not prbvide any design guidelines as discussed above. For the perforated plate, the equations from Appendix A, Article A-8000'of the ASME BBPV Code, Section 111, 1998 Edition through 1999 Addenda (Reference [3]) 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, in keeping with the B31.I maximum principal stress design philosophy, principal stresses are calculated and compared to the allowables 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 Design Level Primary Membrane Stress Primary Membrane (or Local) + Bending 1.5 Sh Primary Membrane Stress Primary Membrane (or Local) + Bending 1.8 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 AlSC 9th Edition (Reference 191). The allowable stress for the SSE Load Combinations is taken from Section 6.9 of DG M I 0 (Reference [I 41).

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

1.0 AlSC Faulted 3

1.5 AlSC but not to exceed 0.9 S,,

Additional details for the various types of support componenfs are provided below Compression Per Reference 1301, because stainless steel does not display a single, well defined modulus of elasticity, the 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 191). Per Q1.5.9.2 of Reference 1301, 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.

Rev 2

CALCULATION SHEET Calc. No.: PCI-5344-SO4 Station: Point Beach, Units 1 & 2 Calc.

Title:

Structural Evaluation of Containment Emer~ency Sump Strainers Safety Related Yes kd NO Revision: 0 Reviewed By:

Date: 09/25/2008 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 as 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.

I Edge Channels I

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 2.3 of the ASCE Standard for Cold-Formed Stainless Steel Structural Members (Reference [31]), which provides design guidelines for very thin members such as 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.

I Welds I

There are no provisions given in the B31.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 as for the seismic stiffeners to radial stiffeners, end cover connecting tabs, and those for the floor track support system, are qualified per the AlSC 9th Edition (Reference [9]): AWS D l.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 D l.6 are essentially the same as allowed for carbon steel welds under AWS D l.I (Reference [I 3]), no special adjustments are required to account for stainless steel.

I Rivets I

There are three areas in the strainer module where rivets are used as 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 as 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]) as supplemented by the AlSl Code (Reference [22]) 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 calculation) will be used on these ultimate capacities for OBE. An increase of 1.5 is allowed for SSE, resulting in a FSM.5 for SSE.

I Mounting Hardware I

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

01) however a Factor of Safety of 4 against ultimate will be used. Qualifications of the bolts/pins used to attach the strainers to the track will be based on the ASCE Standard (deference [31]). Neither of the AlSC Codes (References 191 & [30]),

provide specific bolting allowables for stainless steel bolting.

Form 3.1-3 Rev 2

4.0 ASSUMPTIONS This calculation evaluates the Unit i strainers including the additional modules and new end covers associated with Unit 1 to be added under EC 12601 and EC 12603. It is also applicable for the Unit 2 strainers, including changes to be installed at a later date, provided the following assumptions hold true:

The end cover assembly and strainers are identical to Unit 1 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 tracks (coupled with a min. 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. 8-112" anchor-to-anchor spacing)

,QUQ tw p 42r9-s h o w n 4 ; i I Cevr\\oLted c~ith reJi>;aq qp, C4lcuLm4;e-p,~+

EC R J Q ~ \\ -

0 7.- A G ~ ~ ~ - C ~ O M Q 1

+

wa& -k edc-ty~ r,.

m-S-'Pa E Form 3.1-3 Rev 2

CALCULATPON SHEET Calc. No.: PCI-5344-SO4 I Safety Related Yes I Date: 09/25/2008 I

5.0 DEFINITIONS AND DESIGN INPUT 3

Define, ksi = 10.psi 3

kips = 10.Ibf kPa := 1000.Pa ORIGIN = 1 Material Properties Material Types per Reference [6b]:

Perforated Plate:

Core Tube:

Radial Stiffeners:

Wire Stiffeners:

Rivets:

Connecting Rods:

Nuts:

Washers:

Spacer Sleeves:

Seismic Stiffeners:

Angle Iron:

Mounting pins:

Hitch Pins:

End Cover Assembly Latch and Strike Plate:

Latch Rivets:

Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-493, Type 304 (Drafted to I 10 ksi - 130 ksi)

Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-276, Type 304 Stainless Steel ASTM A-1 94, Grade 8 Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-312, Type 304 Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-276, Type 304 Stainless Steel ASTM A-276, Type 304 Stainless Steel ASTM A-580, Type 304 Stainless Steel ASTM A-240, Type 304 Stainless Steel ASTM A-240, A-580, A-31 3, Type 304 Stainless Steel ASTM A-493lA-313, Type 304 Design Temperature Tdes= 2500 F ( Reference [I]

)

CIient: Performance Contracting Inc.

Revision: 0 I Safetv Related Yes 1 Date: 09/25/2008.

I All Tvpe 304 Steels (Based on A-240, Tvpe 3041 Modulus of Elasticity at 2500 F (Reference 1411, E, := 27300.ksi Yield strength at 250° F (Reference [4]),

Sy := 23.6.ksi Ultimate Strength at 250° F (Reference [4]),

S, := 68.6.ksi 031.I Allowable Stress at 2500 F (Reference [5]),

Sh := 17.2-ksi Note these properties are conservative for the Type 304 wire stiffeners which are drafted 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. [48]) 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 250° F (Ref [48])

Sy.,ire

= 65ksi Other Miscellaneous Pro~erties Density of stainless steel from Reference [20],

Density of carbon steel from Reference [20],

Poisson's Ratio from Reference [20],

Density of water at temperature of 680F(Ref. 1121)

Ibf p steel := 501.-

ft3 Ibf Pc.steet := 490.-

ft3 v := 0.305 Ibf Y ~ 2 o. j := 62.4.-

ft3 Ibf Density of water at temperature of 250°F(Ref. [38])

YH20.2 := 58.8.-

ft3 Coefficient of Thermal Expansion (CTE) of stainless steel, CTE := 9.1 6 (going from 70°F to 250°F (Ref. [4])

• Hydrodynamic mass is based on the density of water at temperature. Since the yield strength 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.

I J

Form 3.1-3 Rev 2

CALCULATION SHLEET Calc. No.: PCI-5344-SO4 5.2 Strainer Geometry and Dimensions All data are per Ref. [6d] unless otherwise noted.

Perforated Plate Dimensions Thickness of 18 gage perforated plate as per Reference [35]

Hole diameter of perforated disk plate, Pitch distance between perforation holes in disk plate (Center-to-center distance)

Disk Dimensions Strainer disk size Lldisk:= 33.0.in LZdisk := 36.0.in Number of disks per strainer module Ndisk := 10 Strainer disk edge channel dimensions dchan := 0.5.in Ref. C6gl bchan := 0.5.in Ref. t6gl Width of each middle disk assembly Wdisk := dchan i. 2.tperf Wdisk = 0.596 in I

Width of-gap spacing between consecutive disks I

I I

W,,,

= 1.0-in Figure 5.2 Side view of Strainer Module

CALCULATION SKEET 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 Width of external radial stiffeners w, ~, := 1.5.in Width of debris stop w

~

~

~

~

~

= 0.84375 -in Outer diameter of the debris stop ODdebris := 17.565.h Width of top and bofbm external radial stiffener ends wend := 2.0-in Length of top stiffener ends LT.end := 2.5.in Length of bottom stiffener ends LB.end := 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.cb := 0.1875-in Seismic stiffener to radial stiffener weld length (on either side of tab) ww.,b := 1.in U

K,d Figure 5.2 End view of Strainer Module

CaCULATION SHEET Safety Related Yes IXI No [_1 I Date: 09/25/2008 Connectina Rod Dimensions Number of connecting rods Connecting rod diameter 0.9743. in Connecting Rod tensile diameter ODtens := ODrod -

13 Outside diameter of spacers (112" ID, SCH 80)

Thickness of spacers (112" ID, SCH 80)

Eccentricity between edge of disk and outer connecting rod Connecting rod tightening torque Diameter of centerline of inner tension rods Core Tube Dimensions Outer diameter of perforated core tube CorrosionlFabrication Allowance Core tube wall thickness (16 ga.)

Core tube wall thickness after allowance ttube := fl(iga - 2.ha Core tube extension beyond last disk face Outer diameter of disk gap Number of rows of core tube holes Number of holes per row Radial stiffener to core tube weld thickness Radial stiffener to core tube weld length (per individual weld)

The orientation of the hole along4he circumference Nmd := 8 ODmd := 0.5.in Ref. [6fl ODtens = 0.425 in Ref. 191 ODspace, := 0.84.in Ref. 191 tSpacer := 0.147 -in Ref. PI e,d

= 0.9375.in Tmd := 2O.ft.lbf BCmd := 17.254-in Ref. [35]

Ref. [6fl Ref. [6e]

Ref..[6e]

Ref. [6e]

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

NIrivet:= 10 N2,jVet := 11 End cover, facelgap disk rivet head diameter c

~

~

~

~

= 0.375. in Ref. [6fJ (item #'s PR64FFP and PR62FFP, respectively. See Ref. 1291)

Ref. [6h]

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

Number of intermediate disk face rivets Number of inner gap rivets holding the hoop together Number of rivets to attach latches and strikes to sleeve connector Eccentricity between the edge channel rivets and the adjacent edge of disk edvet := 0.25.in Offset from line connecting center of core tube and center of outer rod eOfl := 1.25 -in (Refer to subsection Internal Wire Stiffeners in Section 6.1 for more detail)

Internal Wire Stiffener Dimensions (All data per Ref. l6ql unless otherwise noted)

Number of intermediate circumferential stiffeners Diameter of radial wire stiffeners (7 ga) dwireerad := 0.177-in Ref. [6bl Diameter of circumferential wire spacers (8 ga) dwire.circ := 0.162.in Ref. 16bI Inner circumferential stiffener width LCi,.in

= ODtub, + 1.5.in Lcirc.in = 17.32 in Outer circumferential stiffener width (Side I )

L l circ.out := L l disk erod L1 circ,out = 31.125 in Outer circumferential stiffener width (Side 2)

L2circ.out := Lzdisk - 2.erod L2circ,out = 34.125 in Corner distance for outer circumferential Lcirc,cor := 1.5-in End Cover Assemblv Dimensions (Dimensions per Ref. T6y1)

Thickness of end cover Diameter of back plate ODback,pl := 19.315Oin Diameter of sleeve ODsleev,.ec := 15.815in Thickness of sleeve ts~eeve.ec = 0-06 in Length of base plate Lbase.pl := 14in Thickness of base plate tbase,pi := 0.5in Length of tube steel support L,,,,

= 28.2095in Length of sleeve LsIeeve.ec

= 1.5in Eccentricity between edge of base plate and anchor bolt ebaseSpl

= 1.25in Height of stiffener 2

hstiff := 3in Thickness of stiffener tStiv := 0.5in Size of tube steel support w&,,:=

4in

Minimum distance from the expansion anchor boR 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 I)

Weld thickness for all base plate connections of end cover assembly Weld thickness of tube steel to the back plate tw.ec.t,

= 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 doff,,,

= 0.5in tube steel in the horizontal direction Anchor Bolts for end cover assembly ODhkb.ec := 0.5in Other Miscellaneous Dimensions Diameter of mounting pin connecting the strainer to the angle iron track ODm := 0.5.in Ref. [6h]

Angle iron thickness tangle := 0.25.in Ref. [6i]

Length of vertical leg of the angle iron track Lvert.leg := 2.in Ref. [6i]

Eccentricity from bolt connection to bottom of angle ebOlt := 1.125.in Ref. [6i]

Eccentricity from corner of angle to anchor bolt ehkb.3 := 1.5.in Eccentricity from edge of angle leg to anchor bolt ehkb.2 := 1.5.in Span between two adjacent anchor bolts L~~~ := 19.9567-in Ref. Pi1 Eccentricity between two adjacent module supports eSprt := 6.5.in,

Ref. [6i]

Length of alternate angle iron segment in case of rebar interference:

Lalt := 4.5-in Ref. [6c]

Alternate angle iron segment to angle iron track weld length (full) w,.,~~

= 2-in Ref. [6c]

I Rev 2 I

ale. No.: PCI-5344-SO4 Client: Performance contract in^ Inc.

Revision: 0 1 Safety Related Yes No l-.l I Date: 09/25/2008 1

Other Miscellaneous Dimensions (cont'd)

Alternate angle iron segment weld thickness Diameter of hitch pin Diameter of Hilti Kwik anchor bolt Diameter of core tube connection sleeve Thickness of sleeve connecting two adjacent modules (22 ga. See Ref. 1351)

W~dth of sleeve connecting two adjacent modules Number of latches per sleeve Span between two module supports for a given module Pool Boundaries (All data per Ref. J6a1 unless otherwise noted)

Minimum height of the water above the floor Gap between the bottom of the strainer and the floor Gap between the top of the strainer and the minimum water level surface Approximate distance from containment wall/floor interface to adjacent strainer train (Unit 1 controls)

Angle of the reactor containment wall 0.5. L1 disk I-gf Minimum average gap between the side of g, := e, I-the strainer and the nearest wall (Unit I tan(awa1i) controls) tw.,lt := 0.1 875.in Ref. [6cl ODhitch := 0.177.in Ref. [6bl 0Dhkb := 0.625 -in Ref. f.6~1 oDSIeeve := 15.8723. in Ref. f6hl tSleeve := 0.0293. in Ref. [6hl wsleeve := 3.5. in Ref. [6h]

Nlatch :=

Ref. [6h]

L~~~~ := 13.4567.in Ref. 16iI H,:=

38.in gf:= 3.in gt := 2.in e,:= 6.in Ref. [6j]

a,,, = 73.30 deg Ref. [6j]

g,=

11.85 in Ref. [6j]

and Ref. [6a]

Calc. No.: PCI-5344-SO4 Strainer Trains The hoielslot distributions along the length the core tube are given in terms 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. H is provided in array format and L2 and Llis are provided as constants (see Reference [6e]), where the rows are the hole locations, 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.

0 90 deg 2.34 2.39 rhole := min -, 0.25-in

(:

1 Figure 5.2 Partial View of Strainer Trains (Figure is a partial view of complete layout, see Ref. [6e])

Note the holes at 0 degrees and 180 degrees are the same size, and the holes at 90 degrees and 270 degrees are also the same size (see "Sure-Flow Strainer Trains" Reference [6e]).

1 I

I Form 3.1-3 Rev 2

CALCULATION SHEET Corner Intermediate Circumferential d.rivet.edge.2 Figure 6.1 Intermediate Wire Stiffener Pattern and Notation

CALCULATION SHEET Safety Related Yes Date: 09/25/2008 7.0 RESULTS AND CONCLUSIONS 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 below.

Strainer Component External Radial Stiffener (Including Debris Stops)

Tension Rods Edge Channels (Rims Disks)

Seismic Stiffeners Pipe Spacers Core Tube (Biggest Holes)

Perforated Plate (DP Case)

Perforated Plate (Seismic Case)

Perforated Plate (Rim Disks)

Perforated Plate (Gap Disk)

Wire Grill Stiffener End Cover Assembly Components End Cover Assembly Anchor Bolts Ref. Section lnteraction Ratio (OBE SSE)

~ R ~ ~ ~. ~ ~ T

= (0.84 0.95 )

End Cover Assembly Welds 6.12.1 I

R

~

~

~

= (0.31 0.23 )

Weld of ~ a d i a l Stiffener to Core Tube Weld of Radial Stiffener to Seismic Stiffener 6.12.3 I

R

~

~

~

~

~

~

= (0.51 0.50 )

Form 3.1-3 Rev 2

CALCULATION SHEET Calc. No.: PCI-5344404 I Client: Performance Contracting Inc.

I Revision: 0 I

I Safety Related Yes I Date: 09/25/2008 I

RESULTS AND CONCLUSIONS (Cont.1 Strainer Component Ref. Section Interaction Ratio Rim ~ i s k Blind Rivets 6.13.1 I R,, ~ = (0.13 0.12 )

Gap Disk Blind Rivets 6.13.2 IRw.gap = (0.09 0.06 )

Mounting Pins Clevis Hitch Pins Angle lron Mounting Tracks Expansion Anchors to Floor Angle Iron-to-Angle lron Track Weld Module-to-module Sleeve Module-to-module Sleeve Connection (optional Strap and Clip included)

Lift Case Outage Case Form 3.1-3 Rev 2

CALCULATION SHEET Safety Related Yes bd NO Date: 09/25/2008

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, Revision 0.

[3]

ASME B&PV Code, Section Ill, Division 1, Subsections NB, NC, and Appendices, 1998 Edition, through 1999 Addenda.

[4]

ASME B&PV Code, Section ti, Park D, Material Properties, 1998 Edition, through 1999 Addenda.

[5]

ANSIIASME B31.I Power Piping Code, 1998 Edition, through 1999 Addenda.

[6] Performance Contracting, lnc.(PCI), Sure-Flow Suction 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-01, "Sure-Flow Strainer General Notes", Revision 7.

6c.

PC1 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", Revision 1.

6e.

PC1 Drawing No. SFS-PB2-PA-7101, "Sure-Flow Strainer Trains", Revision 1.

6f.

PC1 Drawing No. SFS-PB2-PA-7102, "Sure-Flow Strainer Module Assembly", Revision 3.

6g.

PC1 Drawing No. SFS-PB2-PA-7103, "Sure-Flow Strainer Sections and Details", Revision 0.

6h.

PC1 Drawing No. SFS-PB2-PA-7105, "Sure-Flow Strainer SleeveslCoverlSupportsIPins", Revision 4.

6i.

PC1 Drawing No. SFS-PB2-PA-7150, "Sure-Flow Strainer Mounting Track AIIBI", Revision 2.

6j.

PC1 Drawing No. SFS-PBI-GA-00, "Sure-Flow Strainer Recirc Sump System", Revision 9.

6k.

PC1 Drawing No. SFS-PB2-PA-7106, "Sure-Flow Strainer End Cover", Revision 1.

61.

PC1 Drawing No. SFS-PBI-GA-01, "Sure-Flow Strainer General Notes", Revision 12.

6rn. PC1 Drawing No. SFS-PBI-GA-02, "Sure-Flow Strainer A Strainer", Revision 9.

6n.

PC1 Drawing No. SFS-PBI-PA-7100, "Sure-Flow Strainer Module Assembly", Revision 4.

60.

PC1 Drawing No. SFS-PBI-PA-7101, "Sure-Flow Strainer Trains", Revision 5.

6p.

PC1 Drawing No. SFS-PBI-PA-7102, "Sure-Flow Strainer Module Assembly", Revision 3.

6q.

PC1 Drawing No. SFS-PBI-PA-7103, "Sure-Flow Strainer Sections and Details", Revision 3.

6r.

PC1 Drawing No. SFS-PBI-PA-7105, "Sure-Flow Strainer SleeveslCoverlSupportsIPins", Revision 12.

CALCULATION SmET 6s.

PC1 Drawing No. SFS-PBI-PA-7150, "Sure-Flow Strainer Mounting Track A?, Bl", Revision I.

6t.

PC1 Drawing No. SFS-PBI-PA-7153, "Sure-Flow Strainer Monting Track A3, BY, 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-PBI-PA-7152, "Sure-Flow Strainer Module End Cover Assembly", Revision 3.

6z.

PC1 Drawing No. SFS-PBI-GA-07, "Sure-Flow Strainer Piping A Layoutt', Revision 2.

[7]

"Formulas for Natural Frequency and Mode Shape," by Robert D. Blevins,l979,Van Nostrand Reinhold.

[8]

AES Calculation PCI-5343403, "Prairie Island Strainer Sloshing Evaluation", Revision 0.

[9]

AISC Manual of Steel Construction, 9th Edition.

[I 0]

Hilti Product Technical Guide, 2008.

[I I] Wisconsin Electric Guideline DG-MO9, Design Requirements for Piping Stress Analysis, Revision 2.

[I 21 "Engineering Fluid Mechanics" by John A. Roberson and Clayton T. Crowe, 2nd Edition, Rudolf Steiner Press, 1969, Library of Congress Catalog Number 79-87855.

[I 33 AWS Dl. IIDI.I M:2002, "Structural Welding Code - Steel".

[I41 Wisconsin Electric Guideline DG-MI 0, Pipe Support Guidelines, Revision 2.

[I51 Wisconsin Electric ~uideline DG-C03, Seismic Design Criteria Guideline, Revision 0.

[I61 "Roark's Formulas for Stress & Strain" by Warren C. Young, 6th Edition, McGraw-Hill 1989.

[I71 "Theory of Plates and Shells" by Stephen P. Timoshenko and S. Woinowsky-Krieger, 2nd Edition, McGraw-Hill, 1959.

[I 81 PC1 Intra -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)

[I 91 ASME Publication, "Pressure Vessel and Piping: Design and Analysis," Volume 2, 1972, Components and Structural Dynamics, Paper Title " Design of Perforated Plate," by O'Donnell & Langer Reprinted from Journal of Engineering for Industry, 1962.

Date: 09/25/2008 1201 "Marks' Standard Handbook for Mechanical Engineers", by Avallone, and ~aumeister, 9th Edition, McGraw Hill.

1211 AES Verification and Maintenance File for GTSTRUDL Version 25.

[22]

AlSr Specification for the Design of Cold-Formed Steel Structural Members, 1996 Edition.

1231 ANSIIAWS Dl.6:1999, "Structural Welding Code - Stainless Steel".

[24]

Not Used

[25]

ANSYS Verification File, Version 5.7.1, dated 9/28/2003, AESMN File No. AES.1000.0562.

1261 EPRl Document NP-5067, "Good Bolting Practices -A Reference Manual for Nuclear Power Plant Maintenance Personnel".

[27]

PC1 Technical Document TDI-6007-04, "Module Debris Weight - Point Beach Nuclear Plant Units 112",

Revision 3.

1281 Nukon Pipe.ESD-TR-146, "Latch and Strike Tensile Strength Test," August I I, 1993. (Attachment D) 1291 Jay-Cee Sales and Rivet Inc, "Expanded Product Line", 4th Edition. (Attachment E)

[30]

American National Standard ANSllAlSC N690-I 994, "Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for'~uclear Facilities"

[31]

ASCE Standard SEIIASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members".

[32]. "Theory of Elastic Stability" by Stephen P. Tirnoshenko and James M. Gere, 2nd Edition, McGraw-Hill, 1961.

[33]

Journal of Ship Research, "Sway Added-Mass Coefficients for Rectangular Profiles in Shallow Water", by FTagg, C.N. and J.N. Newman, December 1971. (Attachment F)

[34]

Journal of Engineering Mechanics ASCE, "Added Masses of Lenses and Parallel Plates", by Sarpkaya, T.,

1960. (Attachment G)

[35]

Stainless Steel Sheet Thickness Table from Hendrick book. (Attachment H)

Station: Point Beach, Units I & 2

[36]

Bechtel Drawing No. C-128, Containment Structure Interior Plans at El. 10'-0", EL. 21'-O", EL. 24'-8", and EL. 38'-O", Rev. 9. (Unit 1)

[37]

Bechtel Drawing No. C-2128, Containment Structure Interior Plans at El. 10'-O", EL. 21'-On, EL. 24'-8", and EL. 38'-On, Rev. 8. (Unit 2)

[38]

"Fundamentals of Engineering Thermodynamics, Sl Version" by John R. Howell and Richard 0.

Buckius, McGraw-Hill, 1987.

[391 "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.

1411 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. 1 (Attachment J)

[45]

EC No. 10581 affecting drawings "SFS-PBI-GA-00, Revision 6 and SFS-PBI-GA-02, Revision 6", Revision 0.

[46]

DIT-008 for EC 12603 and EC 12601 From Point beach 9/18/08.

[47]

ASTM Standard Specification A493-85, "Stainless and Heat-resisting Steel for Cold Heading and Cold Forging - Bar and Wire".

[48]

Lehigh Testing Laboratories Test No. G-4-27, dated August 3, 2007, with test reports attached. (Attachment K)

[49]

Not Used.

[50]

Bechtel Drawing No. C-3181, (Unit I )

[51]

Lehigh Testing Laboratories Test No. F-19-32, dated July 20, 2006 (with test reports attached) (Attachment L).

Form 3.1-3 Rev 2

ENCLOSURE 4 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-I 911GL 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 24,2009 EVALUATION OF SUMP COVER AND PIPING FOR THE CONTAINMENT SUMP STRAINERS (ABRIDGED) 21 pages follow

Calculation Number: PCI-5344403 Calculation

Title:

Evaluation of Sump Cover and Piping for the Containment Sump Strainers CIient:

Performance Contracting, Inc. (PCI)

Station: Point Beach Project. Number: PCI-5344 I ~nit(s): 1 I

Project

Title:

Point Beach Strainer Engineering

( Safety Related Yes No I

Revision Descliption 0

Approval Signature /

Date 1

Rev. l Signebl-e I Initials of Preparers & Reviewers All 2

Initial Issue 1 4 6 1 1 4 -

16,25-32,34-68,70-105, 107 Attachment A Attachment B Attachme~lt C 1-4,6,26,29-31,34,37,38, 40-42,44-47, 55-68,70-80, 82-92,95-98, 100-103,205, 106, 108-1 12, 114 &

Ii~corporated pressure thrust load on piping due to strainer pressure imbalance. Revised sole plate design to bolt directly to floor, evaluated up13 load on sole plate from valve testing and the application of lubricants during flange bolt-ups. Minor other editorial changes. Renumbered a11 pages from.p.

8 forward. This revision resolves AES CAR 06-006 Inco~porated ECN's 1 05 80, 1058 1, 10653 & DIT for EC 10720.

Reanalyzed static analysis (Attachment

~ j t o itlcorposate gaps of 3/32" at the U-bolti'pipe side of the 2-way restraints (PS3). Renumbered all pages froin page 73 forward.

Prepared by:

Revision Affected Pages r

I I

Revision Description Added 1A Changed faulted shear stress allowable to agree with DG-MI0 and the effective shear stress area for angle and tee sections to agree with paragcaph C3 (p.

5-315) of AISC 9th Edition Co~nmentary on the Specification for Allowable Stress of Single-Angle Members. This resolves CAR-07-004.

112,114 Added 44A Prepared by:

Approval Signature /

Date Reviewed by:

Signature / Initials of Prcparers & Reviewers Incorporated DIT003 for EC 1260 1 &

EC 12603. Reanalyzed static analysis (Attachment A) to incorporate increase in pressure tlu.ust load on piping due to strainer pressure imbalance.

I Revicwcd by:

Form 3.1 -1 0 Rev. 1

I REVIEWR'S CHECKLIST FOR DESIGN CALCULATIONS SHEET 1 of 2 STATION: Point Beach - Unit 7 NUCLEAR SAFETY RELATED: YES NO PROJECT NO:

PCI-5344 CLUENT: Performance contract in^ inc.

CALCULATION TITLE: Evaluation of sum^ Cover and Piping for the Containment Sumo Strainers

2. System Description

I I

TYPE OF DOCUMENT

1. General Design Basis COMMENT

?

DOCUMENT ID, REV AND/OR DATE 3,5,8,9, 19 YES X

3, Design information package from related equipment vendor

( 5. Mechanical Discipline Input 1 I

X I

I I

X 4, Electrical Discipline Input X

9.

Vendor Drawings

10. Design Standards

11. Client Standards I

I I

12. checked Calculations 15,20,29 X

6. Control Systems Discipline Input

7. Structural Discipline Input PREPARER'S SIGNATU DATE:

9/24/2008

\\

X 4,7, 12, 17, 18,21-39 FU3VIEWER1S SIGNATURE:

DATE:

9/24/2008 APPROVER'S SIGNATURE:

DATE:

9/24/2008 X

Form 3.1-4 Rev. 3

A. Properly Identified?

B. Properfy Applicd?

3. Were the inputs correctly selected and used?

4.

A. Was Design Input Log used?

5. Are necessary assumptions adcquatev stated?

6. Are the assumptions reasonable?

7. Was the ealcuIation methodology appropriate?

8. Are symbols and abbreviations adequately identified?

9. Are tltc caIculations:

D. Presented in logical ordcr?

E. Prcparcd in proper format?

10. Is the output rcasonabIe cornpard to the inputs?

11. X f a compufer program was used:

A.. Is the program Iisted on the ASI, and has the SEW been reviewed for any program use iimitations?

B. Nivc existing user noticcs andor error reports for the production version been reviex~ed as appropriate?

C. Were codes proyerly verified?

D. Were they appropriate for the application?

E. Were they correctly used:

F. Was data input correct?

G. Is thc computer pr Form 3.1-4

CUCULATION SHEET Client: Perforlnance Contracting Inc.

I Revision: 2 I

1 TABLE OF CONTENTS Methodology................................................................................................................................................

5 Assumptions.............+..................................................................................................................................?O Definitions and Design input.....................................................................................................................

11 5.1 Material Properties.................. i 11 5.2 Pipe Geometry and Dimensions 12 Caiculations 13 6.1 Weight Calculations 13 6.2 Pipe Loads......... ;...........................:.................................................................................................... 13 6.3 Calculation of Acceleration Drag Volumes and Hydrodynamic Mass 16 6.4.Piping Evaluation 25 6.5 Flange Evaluations 32 6.6 Sole Plate Analysis...........................................................................................................................

57 6.7 Support Evaluations 63 6.8 Integral Welded Attachment EvaIuation...........................................................................................

97 6.9 Slip Joint Evaluation 107 Results and Conclusions 109 References..............................

112 Attachments Pages A

"Bn Strainer Piping (Static)......................+......................................................................................A'l

- A31 I3

" B Strainer Piping (Seismic 1).......................................................................................................Bl

- B37 C

"B" Strainer Piping (Seismic 2)..................................................................................................... C1 - C38

Client: Perfoiina~~ce Contracting Inc.

I Revision: I Automated Engineering Services Corp The purpose of this calculation is to qualify the sump cover, piping, and piping supports associated with the Performance Contracting Inc. (PCI) Suction Strainers to be installed in W d e d m a g m e M ~ p e r a t i o n '

Point Beach Nuclear Plant Unit 1. This calculation evaluates, by analysis, the piping as well as the supporting structures associated with the new piping. The evaluations encompass all piping from and including the sump cover plate (sole plate) attached to the El. 8' floor slab to the strainer connections including intermediate support structures.

2.0 METHODOLOGY The evaluations are performed using a combination of manual calculations and computerized piping using the AutoPlPE Program (Reference [16]). The piping is considered as an attachment or extension to the strainers and are therefore subject to the requirements of Strainer Design Specification PB-681 (Reference [I]).

Exceptions from these requirements, if taken, are discussed and justified within this calculation.

C & ~ a % d ; r % ~ ~ ~

SHEE3' 1

seismic Loads 1

Page: 5 of 114 Calc. No.: PCI-5344-SO3 The strainer piping 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 DG-C03 Seismic Design Criteria Guideline (Reference [14]). Strainer Design Specification PB-681 (Reference [I]), requires the piping 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 piping will be considered submerged and full of water. The water level is considered to be a minimum of 3'- 2" above the 8' floor elevation (El. 94'- 2"). 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 specification, the seismic evaluation is required to talce into accoul~t any seismic slosh (analyzed at the seismic worst-case water level) of the recirculation water. Based on Reference [20], 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 pipe sIress analysis. Note that the sloshing calculation of Reference [20] 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 will consider 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 piping is considered.

The piping 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 piping through base motion response spectra as detailed in the Seismic Design Criteria Guideline DG-C03 (Reference 1141).

CALCULATION SH3UEET 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 213 the value of the maximum ground horizontal response spectra.

The piping is considered a s vital piping and the damping vatues for seismic loads is taken as 0.5% for both the Operating Basis Earthquake (OBE) and the Safe Shutdown Earthquake (SSE) a s required by Seismic Design Guide DG-CO3. The response spectra inputs are for the OBE environment. For evaluating stresses, displacements, loads, etc., for the maximum credible earthquake (SSE),

the values obtained from the OBE analysis are to be increased by a factor of 2.0 (Reference 1111).

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

Per Reference ['l 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 (i.e. closely spaced modes). An earthquake duration of 30.24 seconds was used in the analysis per DG-C03, 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 10 seconds is the official documented seismic analysis. Responses due to the three spatial components are combined by SRSS. (Reference [l1],

paragraph 5.6.5). The cutoff frequency is taken at 30 hz or a minimum of 5 modes arc included.

Zero Period Acceleration (ZPA) residual mass effects are considered since they may significantly affect the piping. The ZPA response is combined 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 Operating loads are comprised of weight, thermal expansion and pressure loads.

The thermal expansion is taken at a temperature equal to the maximum operational inlet temperature to the RH Exchangers of 250 O F (Reference [I]). Small gaps (3132") are modeled on the u-bolt side only of the tweway restraints (Type PS3) on the "5" train piping (Reference 1371. These gaps were modeled to reduce the high thermal loads encountered due to the several bends associated with the "£3" train piping. The design drawings (Ref. [6bj) ensure that these gaps will be available. Note the Autopipe 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 piping. The maximum differential pressure load acting on the piping is the hydrostatic pressure associated with the maximum allowed head loss through the debris covered strainers. This is defined a s 10R of 68 O F water in Reference

CALCULATION SHEET MathCad software is used to generate most of the calculations. All MathCad calculations an! independently verified for accuracy and correctness a s if they were manually generated. AufoPIPE Version 8.05 is used for the piping analysis. AutoPlPE Version 8.05 is verified and validated under the AES QA program as documented in the AES validation and maintenance files (Reference [16]). Because the AutoPlPE Version 8.05 only performs piping evaluations using the 2001 Edition of the 831.1 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 review 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 tensile strengths and the coefficient of thermal expansion because these are the basis for the allowable stresses and the tabuiated "E" and "a" values at temperature. As long as the specified tensile properties of the material have not changed, use of the later Edition does not affect the end result.

The material allo&bles stresses are included manually into AutoPlPE based on the ASME B31.1 - 1998 Edition, which is the design code for pipe stress analysis. In addition, a review of the fwo the codes was performed to identify revisions to the design equation prodsions and to determine if any material properties associated with "E" and "d' had changed. There have been no design dependent revisions to the piping material and to the design code equations. The flexibility 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 will be identical.

3.0 ACCEPTANCE CRITERIA The strainer suction piping shall meet the requirements of the stminer design specification PB-681 (Reference [I]). As stated in PB-681, the detailed evaluations are to be performed using the rules, as applicable, of ANSIiASME B31.1 Power Piping 1998 Edition (Reference [51).

The piping supports, baseplates and other mounting hardware is evaluated to AlSC 9th Edition a s permitted in paragraph 120.2.4 of the B31.1 Code. 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 jnstance, 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, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities". Reference [2q is used to supplement the AlSC in any areas related specifically to the structural quatification 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.

SEIiASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members",

(Reference [24]) is used for certain components (stainless steel boIfs and pins) since the AlSC does not provide specific bolting alTowables for stainless steel bolting. The rules for Allowable Stress Design (ASD) a s specified in Appendix D of this code are used. Finally guidance is also taken from AWS Dl -6, "Structural Welding Code - Stainless Steel", (Reference [261) a s it relates to the qualification 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-MOQ (Reference [?I]).

Load Condition Combination P+DW+OBE (3) Emergency/Faulted P -F DW+ S S E (4) Thermal TI

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 ambient condition of 70 O F to the maximum operating condition of 250 O F (AT = 180 OF).

CALCULATION SHEET CIien t: Perfo~mance Contractinn Inc.

Revision: 1 1

Cafe.

Title:

Evaluation of sum^ Cover and Pipine. for the Containment Sump Strainers Yes IZd NO U Safety Related I Date: 1130/07 I

Piping I

The piping Is evaluated in accordance with ANSI B31.1 Paragraph 104.8 a s 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-MO9 (Reference [I I]).

I B31.1 Eq. No Load Condltion Stress Combination Allowable Stress I

1 T N O ~ A (Sustained)

P+DW 1.0 Sh 12 (OBE)

Upset {Occasional)

P c DW+OBE 2.2 S,,

12 (SSE)

Emergency (Occasional)

P + DW + SSE 1.8 Sh 13 Thermal (Displacement)

T I 1.0 S, I

Flanges I

Since specific detailed guidance is not provided in B31.I, the bolted flange connections at each end of the piping elbows will be evaluated in accordance with ASME Section Ill, Appendix L (Reference [S]) guidelines.

The flange bolts are qualified to the criteria presented in ASME 1% Appendix L (Reference [8]). Note that these are non-standard flanges which do not meet the generic requirements of B31.1 (such as weld size),

As stated in the fdrward of of the B31. I Code (Reference [5]), "a designer who is capable of a more rigorous analysis than is specified in the Code may justify a less conservative design, and still satisfy the basic intent of the Code." Use of a detailed stress evaIuation of the flange and the flange weld, based on ASME analysis equations, cerkainly falls within this category of satisfying the basic intent of the Code.

I I

Piping Support Structural Components I

The allowable stresses on the piping support components are based on the AlSC 9th Edition (Reference [$I).

Also, the allowables stresses for the sump sole pIate tabs, bolts, and welds are-based on the AlSC 9th Edition. The allowabfe stress forthe SSE Load Combinations is taken from Section 6.9 of DG MI0 (Reference 11 31).

Load Condition Load Combination Allo~rable Stress Normal DVV+ T I 1.0 AlSC Upset DW -t. OBE + T I 1.0 AISC Faulted DW +. SSE + T I T.5 AISC but not to exceed 0.9 Sy

Per Reference [25], because stainless steel does not display a single, well defined modulus of elasticify, 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 allowables from Reference [25] a s opposed to those provided in the AISC Code (Reference 191). Per Q1.5.9.2 of Reference I25jjl, the allowable stresses for tension, shear, bending and bearing for stainless steel can be taken a s the same allowables provided for carbon steel, therefore the AISC 9th Edition will be used for allowables for these types of stresses.

Welded Joinfs Allowable stresses for piping welds, such a s the flange fillet welds, are per ASME Section Ill (Reference [8]),

Paragraph NC-3356. IWA welds are in accordance with ASME Code Case N-318-5 (Reference [19]). The allowable stresses for all other welds are based on the AlSC 9th Edition (Reference 191). AWS Dl.6 (Reference [26])

was reviewed to ensure that any special qualification requirements associated with stainless steel welding were considered. Since the weld allowables provided in AWS D4.6 are essentially the same a s allowed for carbon steel welds under AWS Dl.I, no special adjustments are required to account for stainless steel. The allowable stress for the SSE Load Combinations is taken as 1.5 times the AISC weld material allowable per Reference [I 31.

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 631.1 allowables. ASME Code Case N-3185 (Reference [19]) is used to calculate the local stresses since this is the latest version of the Code Case available.

Mounting Hardware Hilti Kwik-Bolt ills are used to mount the support baseplates to the floor. The analysis and design of expansion anchors shall be in accordance with the Hilti Technical Guide (Reference [18]), however, a Factor of Safety of 4 against ultimate loads will be used. Prying factors are calculated in accordance with DG-CO1 (Reference [lo]). Qualifications of the stainless steel boltsipins used to attach the saddle plates to the structural angles is based on the ASCE Standard (Reference [241). The AISC Code (References [9] ) does not provide specific bolting allowables for stainless steel bolting.

4.0 ASSUMPTIONS None.

CALCULATION SREET 5.0 DEFlNlTlONS AND DESIGN INPUT 5.t Material Properties The specific materials for the piping and support components are taken from Reference 6m Piping:

Stainless Steel ASTM A312, Type 304 or Type 304L (Dual Certified)

Pipe Fittings Stainless Steel ASTM A240, Type 304 or A774, Type 304L (Dual Certified)

Structural Steel:

Stainless Steel ASTM A276, Type 304 Flange:

Stainless Steel ASTM A-240, Type 304

'Flange Bolting:

Stainless Steel ASME A-193, Gr. B8, Class I1 Design Temperature Tdes= 250 OF ( Reference [I] )

Properties for the pipe components and support,structural components are taken from ASMElANSI B34.1, Power Piping Code, 1998 Edition (Reference [5]). Yield strength values for support stnrctural components and flange bolting properties are not available in ANSI 831.1 Code and are taken from ASME B&PV Code,Section II, Part D (Reference f4]). For Dual Certified materials only the controlling properties are used.

Yield strength value.for stainless steel A240 Type 304 material at 250 OF:

SySo4

= 23.G.ksi (Ref. 141)

Modulus of Elasticity of stainless steel material at 250 OF:

E := 27300-ksi (Ref. L51)

Allowable pipe stress at design temperatun! (250 OF).

S,:=l7.20-ksi (Ref.[51)

~llowable design stress for flange at design temperature (250 OF),

Sf := 17.20. ksi (Ref- [51)

Allowable bolt stress at design temperature (250 O F ),

Sb := 25.0-ksi

( ~ e f -

[41)

Modulus of Elasticity (flange)

Ef := 27300.ksi (Ref. [Sl)

Modulus of Elasticity (bolts)

Eb := 27300. tcsi (Ref. [41)

Other Miscellaneous Properties Ibf Density of stainless steel (Ref. 1281).

p,t,l:=

501.-

f13 Poisson's ratio of stainless steei (Ref. [28]).

v := 0.305 Ibf Density of water at temperature of 68 O F (Ref. [f23) 7 ~ 2 0

= 62.4.-

ft3

Client: Peifonnance contract in^ Inc.

I Revision: 1 I

Safety ReIated Yes kd No u 1 Date: 1/30/07 I

5.2 Pipe Geomebv and Dimensions I

Pipe Dimensions Outer diameter of pipe (Ref. [6b1)

Pipe wail thickness (sch.20) (Ref. [6b])

Inside diameter of pipe: IDljipe := ODplpe - 2'tpipe I

ODpipe Radius of pipe:

r:= -

2 Corrosion AllowancelFabrication Tolerance Pool Boundaries Length from top of floor to centerline of pipe (Ref. [6al) cf:= 19.5.in Minimum height of the water above the floor (Ref. [6a])

H,:=

3B.in Distance (left side) from wall to pipe centerline (see Section 6.3.4) c,]

= 13.85-in Distance (rigid side) from wall to pipe centerline (see Section 6.3.1)

G,,,,:= 24.in Flanae Dimensions Outer diameter of flange at top of elhow (Ref. [6fj)

Inside diameter of flange at top of elbow (Ref. [6q)

FIznge thickness (Ref. [6fJ)

Outer diameter of 16 pipe in-line flanges (Ref. [6b])

Inside diameter of 16 pipe in-line flanges (Ref. [6b])

Figure 6.4.1 - Model Plot of "B" Strainer Pipe

CUCULATXON SHEET Safety Related

( Date: 4/27/07 I

Sole Plats Connection I

As shown in figure below the connection consists of two parts. The fabricated pipe flange is identical to the flange on the opposite side of the elbow, the 1/2" annular sole plate is held down by twelve (12) 518" Hiiti Ill expansion anchors (Reference [6c]).

Note that the 4" minimum distance to the edge of the sump drain concrete opening as shown in the sketch below has been reduced to a minimum of 3" in EC 10581 (Reference [35j). The centerline of the bottom end of the elbow and the associated base ring may be offset a maximum of 1" from the centerline of the sump drain p i ~ e sleeve durina installation to avoid interferences.

\\u "b:i 51%" Hilti 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 NormallUpset and EmergencyiFaulted loads, blrt enveloped between all flange pairs. Dimensional parameters are adjusted as required for each type of flange.

CALCULATION SHEET Safety Related Yes LA No 0 Date: 9/12/08 A summary of the maximum calculated piping stresses is shown in Section 6.4. Calculated support component stresses are shown in Section 6.7. The interaction ratio for the pipe stresses, flanges, sole plate, and supports is shown below I

Pipe Stresses I

I B Strainer Pipe IR~pi~a

= m " ( l ~ ~ ~ t,

[RBI~B, I R ~ I ~ c,

IRBIJ)

IRsPpe = 0.14 I I I

Stress Summarv for other Components I I Component Ref. Section Interaction Ratio I I Flanges In-line Flanges Flange Bolting 6.5 IRbdtl = 0.61 [::I Top of Elbow At Sole Plate Flange Bending 6.5 lRflangel = [a,]

Flahge Weld to Pipe 6.5 IR,,

= 0.24 I

Flange Bolts 6.5 IRboltmissing = O a g 3 I I I

Flange Bending 6.5 1Rflanbe.rnissing = lsOO I I I

Sole Plate Connection I I sole Plate Sole Plate Expansion Anchors 6.6 IRspl,anchor = O.84 I.

CALCULATION SHEET

'lo of

'I4 Calc. No.: PCI-5344-SO3 Ref. Section Interaction Ratio Component T v ~ e PSIIPS2 Restraint NormaVUpset Angle Normal Stv5sS EmergencyIFaulted Angle Shear Stress Expansion Anchors Vype PSI) 1Rbolt-psl = 0-88 Expansion Anchors V Y P ~

PS2)

IRb01t-ps2 = 0.94 Baseplate IRbpr. = 0.61 0.59 Weld of Angle to Baseplate Saddle Plate Bending IRsP\\-bd =

Saddle Plate Shear 1Rspl-sh =

Saddle Plate Welds Saddle Plate Pins 0.30 0.06 Shear Lugs 6.7 Integral Welded Attachments 6.8.1 lRpszim = 0.29 i

CALCULATION SHEET Tvpe PS3 Restraint IR shown are for Faulted Loads (SSE) versus Upset Allowables {OBE)

W6x15 Normal Stress W6xl5 Shear Stress Expansion Anchors Baseplate Weld of W6x15 to Baseplate Angle Normal Stress Angle Shear Stress Weld of Angle to W6x15 U-Bolt Normal Load T v ~ e PB1 Restraint Stanchion P'late Bolts Integral Welded Attachments Other Pipinn Components Slip Joint Upset Emerg The evaluation of the piping and piping supports associated with the suction strainers has shown that the pipe stresses and support loads are acceptabk. The piping stresses, flanges, and support component stresses are wifhin their respective applicable limits and are therefore acceptable.

CALCULATION SElCiEET

[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, Revision 0

[$

ASMElANSl B31.1, Pressure Piping Code, 200"1dition.

141 ASME B&PV Code, Section 11, Part D, Material Properties, 1998 Edition, through 1999 Addenda

[q ASMUANSI 031.I Pressure Piping Code, 1998 Edition, through 1999 Addenda

[6]

Performance Contracting, Inc.(PCl), Sure-Flow Suction Strainer Drawings 6a.

PC1 Drawing No. SFS-PBIGA-OO, "PI3 Unit 1 Sure-Flow Strainer, Recirc Sump System", Revision 9 6b.

PC1 Drawing No. SFS-PBI-GA-04, "PB Unit 1 Sure-Flow Strainer, Piping B Layout", Revision 6 6c. PC1 Drawing No. SFS-PBI-GA-05, "PB Unit I Sure-Flow Strainer, Piping A Layout", Revision 9 6d.

PC1 Drawing No. SFS-PB1-PA-7105, "PB Unit 1 Sure-Flow Strainer, SleeveslCovers/Supports/Pins",

I Revision 12.

6e.

PC1 Drawing No. SFS-PBI-PA-7160, "PB Unit 2 SureFlow Strainer, Sump Inlet Cover", Revision I. I 6f.

PC1 Drawing No. SFS-PB1-PA-7161, "PB Unit 1 Sure-Flow Strainer, Sump Connection Elbow AIIBI",

Revision 0.

6g.

PC1 Drawing No. SFS-PBI-PA-7162, "PB Unit I Sure-Flow Strainer, Pipe B2", Revision 2.

6h.

PC1 Drawing No. SFS-PBI-PA-7163, "PB Unit 1 Sure-Flow Strainer, Pipe B3", Revision I.

6i.

PC1 Drawing No. SFS-FBI-PA-7164, "PB Unit I Sure-Flow Strainer, Pipe B4", Revision 1.

6j.

PC1 Drawing No. SFS-PBI-PA-7165, "PB Unit I Sure-Flow Strainer, Pipe B5", Revision 3.

6k. PC1 Drawing No. SFS-PB1-PA-7166, "PB Unit I Sure-Flow Strainer, Pipe AZ", Revision 2.

61.

PC1 Drawing No. SFS-PBI-PA-7167. "PB Unit I Sure-Flow Strainer, Pipe A3", Revision 2.

6m. PC1 Drawing No. SFS-PBI-GA-01, "PB Unit 'I Sure-Flow Strainer, General Notes", Revision 12

[7]

"Formulas for Natural Frequency and Mods Shape," by Robert D. Blevins,l979,Van Nostrand Reinhold.

I

[B]

ASME B&PV Code, Section Ill, Division I, Subsections NB, NC, and NF, 1998 Edition through 1999 Addenda, including Appendices.

[9]

AISC Manual of Steel Construction, 9th Edition.

CALCULATION SHEET

[lo]

Wisconsin Electric Guideline DG-C01,"Guidelines for Design, Qualification, and Installation of Concrete Expansion Anchors at Point Beach Nuclear Plant" (with revisions per NPM 92-0428, April 27, 1992), Revision 0

[I I] Wisconsin Electric Guideline DG-MO9, Design Requirements for Piping Stress Analysis, Revision 2.

[I21 "Engineering Fluid Mechanics" by John A. Roberson and Clayton T. Crowe, 2nd Edition, Rudolf Steiner Press, 1969, Library of Congress Catalog Number 79-87855.

[I 31 Wisconsin Electric Guideline DG-MI 0, Pipe Support Guidelines, Revision 2.

[I41 Wisconsin Electric Guideline DE-C03, Seismic Design Criteria Guideline, Revision 0.

[15] AES Calculation PCI-5344401, "Structural Evaluation of Containment Emergency Sump Strainers", Revision 0.

[I61 Autopipe Version 8.05 QA Release 08.05.00.16 Verification Report, AES File AES.1000.0513.

[I71 Welding Formulas and Tables for Structural & Mechanical Engineers & Pipe Support Designers Published by I.V.1. Structural Design Service, Copyright 1983.

[I 83 Hiiti 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 Rectangular Cross Section Attachments on Class 2 or 3 Piping", April 28, 1994.

1201 AES Calculation PCI-5343403, "Prairie Island Strainer Sloshing Evaluation", Revision 0.

[21]

Roark's FormuIas for Stress and Strain, Warren C. Young, 6th Edition.

[22]

"Design OF Welded Structures" by Omer W. Blodgeft, 1969, Library of Congress Catalog Number 66-23123.

1231 Mechanical Engineering Design by Joseph Edward Shigley and Larry D. Mitchell, McGraw Hill, 1983.

1241 ASCE Standard SEI/ASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members", Copyright 2002.

[25]

ANSllAlSC N690, "Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities" Copyright 1994.

[26]

ANSllAWS Dl.6:1999, "Structural Welding Code - Stainless Steel".

1271 Bechfel Drawing No. C-128, Containment Structure Interior Plans at El. 10'-O", EL. 21'-O", EL. 24'-B", and EL. 38'-0", Rev. 9. (Unit I) 1281 "Marks' Standard Handbook for Mechanical Engineers", Avallone & Baumeister, 9th Edition, McGraw-Hill

[29]

Good Bolting Practice, Volume I, EPRI Report NP-5067

[30]

Rigid Frame Formulas, A. Kleinlogel, 2nd Edition. Frederick Ungar Book Publishing

[31]

PBNP Design Information 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. Bleigh (PCI), dated 01-12-07 1321 AISl Specification for the Design of cold-~ormed Sfeel Structural Members, 1996 Edition.

[33]

Lehigh Testing Laboratories 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

[35]

Engineering Change Notice 10581 to Modification EC 1602 (MR 05-0171, Revision 0, Dated 4M1/07 1361 Engineering Change Notice 10653 to Modification EC 1602 (MR 05-017), Revision 0, Dated 4/19/07 1371 Design Information Transmittal for Point Beach EC 10720, "Thermal Expansion Gap Requirements, Dated 4/27/07 1381 ACI Structural Journal, January-February 1995, VOL. 92 NO. I (Included as Attachment J to Calculation PCI-5344-Sol)

[39]

Point Beach Design Information Transmittal DITO03 for Modification EC 12601 and EC 12603, "Differential Pressure for Debris Interceptors", Dated 8/7/08 I

ENCLOSURE 5 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GSI-I 91 IGL 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

Vortex, Air Ingestion &Void Fraction - Point Beach Nuclear Plant -

Unit-1 & 2 Technical Document No. TD1-6007-07 Revision 4 I

'CALCULATION COVER SHEET GalculatiOn Number:

TDI-6007Q7

~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 Verification Method (Check One):

.m Design Review 0

Alternate Calculation Qualification Testing Scope of Revision:

Specific revision to address operating temperature range for voiding and updated air Ingestion and Froude calculations. Revision 4, Pages: All I

Documentation of Reviews and Approvals:

Date a Date b h 6 - r n

' Date

~ / b 164

Vortex, Air Ingestion &Void Friction - Point Beach Nuclear Plant -

Unit-1 &2 Technical Document No. TDI-6007-07 Revision 4 I CALCULATION VERIFICATION CHECKLIST

.... -.. Galculation-Title:-.-.. V6rtex;Air-Ingestion-&-Vijid Fraction - Point Beach Nuclear Plant --

uriit-I & 2 I

Revision: 4 I

CHECKLIST Were inputs correctly selected and incorporated?

I I

I

2.

Have applicable construction and operating experience been considered?

lr~l 1 0 1 ~ 1 I

I I

Are assumptions adequately described and reasonable?

3.

4.

Are the appropriate quality and quasi assurance requirements specified?

Are the applicable codes, standards and regulatory requirements identified and

6.

7.

8.

l El Have the design interface requirements been satisfied?

Was an appropriate design method used?

Is the output reasonable compared to input?

Are specified parts, equipment, and processes suitable for the required application?

O~-A-r~th-esp~ei:if~d~m~IriaI~atib-1~~t~d~~vi~nbl-~TditianS?-

13.

14.

5.

16.

17.

I Verified bylD nitials:

I

...~-.....

. 11.

12.

18.

'19.

C I C ]

El El CI Has adequate accessibility been provided to perform the in-service inspection?

Has the design properly considered radiation exposure?

Are the acceptance criteria incorporated in the design documents sufficient to allow verification?

Have adequate pre-operational and subsequent periodic test requirements been specified?

Are adequate handling storage, cleaning and shipping requirements specified?

Have adequate maintena&features and requirements b&n specified?...........

-_^.

Are accessibility and other design provision adequate?

and signed?

8 0.1 Note: This is PC1 form 3060-3 Revision 3 Are adequate identification requirements specif ed?

Are requirements for r v r d preparation, review, approval, retention, etc.,

adequately specified?

Has the appropriate Calculation Guideline Verification Checklist been reviewed B u n n u o n nrxl

-a-la

, m,..

-m-a a @

n o C]

I nIX[

a @

a l l @

IXI a

-B-..

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-1 & 2 Technical Document No. TD1-6007-07 Revision 4 TABLE OF CONTENTS

.-.~AL~uLATION.CQVER-SHEET-CALCULATIOIN VERIFICATION CHECKLIST TABLE OF CONTENTS 1.0 Purpose and Summary Results 2.0 Definitions and Teminology 3.0 Facts and Assumptions 4.0 Design Inputs 6.0 Acceptance Criteria 7.0 Calculation(s)

T;-lVo'itex 7.2 Air Ingestion

.. -.-7-3.-

. - Void. Fradion---

8.0 Conclusions 9.0 References I

I ATTACWENTS None TABLES Table 1 Results Summary Table 2 Flashing Margin For Operating Temperature Range Table 3 Calculation Results I

I I

Originated By:

I TD16007-07 Vortex Air Ingestion Void Fmtiyn - Rev 4. d ~

Page 3 of 22 Page 3 of 22 '1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 1.0 Purpose and Summary Results The US Nuclear Regulatory-Commission (USNRC) in generic 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) andlor a loss of coolant accident (LOCA). It was further determined that the transported debris could possibly clog the sump screenslstrainers 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 function(s).

In order to address and resolve the various issues identified by the USNRC in GSI-191, utilities have implemented a program of replacing the edsting ECCS sump screens or strainers with new and improved designs.

In order to address and resolve the specific issues associated with USNRC GSI-191 for the Point Beach Nuclear Plant - Unit 1 & 2 (PBNP-ID), Point Beach I entered into a contract with Performance Contracting, Inc. (PCI). The primary objective of the contract was for PC1 to provide a qualifed sure-Flow@ Suction I Strainer that has been specifically designed for PBNP-1R in order to address and resolve the NRC GSI-I91 ECCS sump clogging issue.

PC1 has prepared a Qualification Report specifically for the subject strainer. The L~ua~cation-~eport-is-a-mmpilation-of-~e-various-documents-and-~calculation that support the strainer qualification. It also provides a "single-sour~" historical

....... -,. record.. that can be utilized t o address any PBNP--1R-organizational or NRC... -...

regulatory issues or questions associated with the replacement PC1 sure-Flow' 1

Suction Strainer.

As part of the PBNP-112 Qualification Report, PC1 has performed a number of hydraulic calculations in support of the replacement sure-FIO# Suction Strainer. 1 This calculation TD1-6007-07, Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear 'Plant - Unit - 1 & 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 associated 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.

The PBNP units each have Mo (2) separate recirculation strainer assemblies that individually and specifically feeds either the 'A' or 'B' trai~n ECCS and CS system.

Each of the horizontally oriented recirculation strainer ass6mbly is

'cc.

Date 3 /6 (a?

TDISOD7-07 Vortex Air ingestion Void Fradn - Rev 4.dOc Page4of22 I

Vortex, Air Ingestion EI Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 PERFORMANC Technical Document No. TDI-6007-07 Revision 4 compiised of fourteen (14) modules each made up of ten (10) strainer disks for a total strainer area of 1,904.6 ft?, or a total of 3,809.2 f? for each pair of strainers I

.associated with one of the PBNP units;. - Flow leaves the strainers and enters combination of pipe and fittings before discharging into the containment outlet.

PC1 drawings [Drawings 10.1 - 10.9, inclusive] provide details of the subject configuration.

The results of the calculation are provided in Table I.

The calculation utilizes the Acceptance Criteria established in both PBNP-112 and USNRC documents with resped to PWR sump performance to specifically evaluate the PBNP-112 Sure-low@ Suction Strainer assembly, It was concluded that this calculation, an integral portion of the Qualification Report completely supports the qualification, installation, and use of the PC1 sure-low@ Sudion Strainer for Point Beach Nuclear Plant - Unit - 1 & 2 without I any issues or reservations.

1 Originated By:

I 1

v - - - - -

I TD18007-07 Vortex Air lngestjon Void Fraction - Rev 4.doc Page 5 of22 I

I Table I - Results Summary Results ACCEPTABLE - Vortex formation is precluded by fhe PC1 sure-FIOW@ Suction Strainer design and configuration

-ACCEPTABLE---Air-ingestion will-not-occur-since there is no vortex brrnation associated with the PCI ~ u r e - ~ l o w ~

Suction Strainer design and configuration ACCEPTABLE - Voids will not occur at the strainser or before leaving the PC1 Sur*

Flow Suction Strainer assembly and discharge piping and entering the PBNP-112 containment outlet Issue Vortex

-Air-Ingestion Void Fraction Acceptance Criteria 1

USNRC No vortex

-0%-or-<2%--

~ 3 %

PBNP-412 No detrimental etrects on RHR, SI & CS Pumps

-No detrimental-effectsonRHR, SI 8 CS pumps NIA

>p I Vortex, Air Ingestion & Void Fraction - Point.Beach Nuclear Plant - 1 4 1

~RMANCF Technical Document No. TDI-6007-07 I

• I Revision 4 1

2.0 Definitions and Terminology I

--The-following-Definitions b~~errninology are defined and-described as - they are--.:---.,

utilized in this calculation.

sure-FIO@ Suction Strainer - Strainer developed and designed by Performance Contracting, Inc. that employs sure-FIOW@ technology to reduce inlet approach velocity.

Emergency Core Cooling System (ECCS) - The ECCS is a combination 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.

Clean Strainer Head Loss (CSHL) - Is the calculated head loss for the Sure-FIO@

Suction Strainer based on actual testing performed at the Electric Power I Research Institute's (EPRI) Charlotte NDE Center, and Fairbanks Pump Company Hydraulic Laboratory. The later testing did not involve any debris.

Point Beach Nuclear Plant Unit 1 $ 2 - also known as Point Beach, PBNP-AR, and PB-In.

i I

.. -- - Containment Spay System--.also. known-as. CSS or CS... System. is utilized. to...-

address either a MSLB or a LOCA.

Lossl0fiCoolant-Accident - also known as a LOCA. A LOCA is the result of a pipe break or inadvertent leak that results in the discharge of primary reactor coolant from the normal nuclear 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 pipe that is associated with the LOCA.

3.0 Facts and Assumptions The following Facts (designated as [FJ) & Assumptions (designated as [A]) were utilized in the preparation of this calculation.

3.1 A flow velocity of 0.0026 fps would be characteristic of the PBNP-112 strainer, through a debris bed consisting of fibers and particulate, is 100%

viscous flow.

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

r

' y-. --

.. Originated By:

Date TDI-6007-07 Vortex Air Ingestion Void Fradion - Rev 4.doc Page6of22 I

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-1 & 2 Technical Document No. TDI-6007-07 Revision 4 3.2 A scale strainer, which is designed to maintain the same approach velocity I

-...-... -- -- as. the. full - scale-. production strainer;- 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 resulting from flow through a fiber - particulate debris bed at the approach velocity for the PBNP-112 strainer (0.0026 Ws) pehrence 9.61, 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 warmer 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 4.8 Design Inputs I

The following combination of Point Beach and PC1 Design Inputs were utilized in I the preparation of this calculation.

4.2 Performance Contracting, Inc. (PCI) Calculation TDI-6007-02, SFS Surface Arm, Flow and Volume Calculation, Revision2 [Reference 9.421, 1 document provides relevant dimensions and other information specifically associated with the PBNP-112 strainers.

4.1 Point Beach Nuclear Plant Specification, Specification No, PB681,

- Replacement o~Confainment Sump Sc_@egs, Revision 2, Febpary 17,-

2006 [Reference 9.11, document provides design input associated with I

4.3 PC1 Calculation TDI-6007-03, COB Tube Design - Point Beach Nuclear Plant-112, Revision 0 [Reference 9.61, document provides relevant data (

I 1

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

4.4 PC1 Calculation.TD18007-051 Clean Head Loss - Paint Beach Nucl6ar.

Plant - I&, Revision 4 [Reference 9.91, document provides the head loss associated with the "clean" PBNP-ll2 strainer and attached pipe and fRtings.

strainer flow rate, water temperature, and the maximum allowable head loss.

Originated By:

Date 316 /d?

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 4.5 PC1 Calculation TDI-6007-06, Total Head Loss - Point Beach Nuclear

+. -- -

Plant - fB;-Revision 5 [Reference9i16], document provides-the total head loss associated with the PBNP-I12 strainer and attached pipe and fittings.

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 9.171, document provides pressure information for addressing voiding in the Point Beach strainer suction lines.

5.6)

Methodology 1

PC1 utilized classical methodology) to address determined that one of the hydraulic calculations (conventional the subject issues.

PC1 recognizes that Calculation if it is I

issues cannot occur andlor can be prevented, then one or more of the other issues cannot occur (e-g., if a vortex is not predicted by 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 its own merits.

I 6,Q-Acceptance-Criteria

.... - This

air,

- ingestion

- specific

-..- -, -. calculation

-. and

,..... - - void

- -. fraction, addresses

,. According-vi three (3) separate

.each i5sue.

but related if.o---".

issues -

separate-.

vortex,........... -- --. -.. -. - -

acceptance criterion. The final overall acceptance criterion is that the PBNP-1R EGGS pumps have adequate NPSH margin under all postulated post-LOCA conditions.

The USNRC in RG 1.82 Revision 3 [Refsrence 9-43 has indicated that air ingestion can lead to ECCS pump degradation andlor failure. A vortex is a potential source of air ingestion. A vortex can be prevented due to various combinations of sump configuration and the addition of vortex suppressors in the sump.

The Acceptance Criteria for, vortex is the complete elimination of occurrence.

I TDl-6007-07 Vortex Air Ingestion Void Fraction - Rev 4.doc 3/6b Date Page 8 of 22

'1

Vortex, Air lngesfion & Void Fraction - Point Beach Nuclear Plant -

Unit-1&2 Technical Document No. TDI-6007-07 Revision 4 Air Ingestion I

RG 1.82 Revision 3 [Reference 9-43 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 1

USNRC GSI-I91 Safety Evaluation (SE) [Refsrence 9.31 has indicated that ECCS pumps can experience cavitation problems when inlet void fraction exceeds approximateIy 3%.

The Acceptance Criteria for void fraction is 9%

in conjunction with an acceptable sump pool temperature operating range as specified in Attachment V-I of [Reference 9.51.

I In order to addras and dekinfiiiie the acceptabilii-and/or issues potentially---'-

associated with the three (3) separate but related issues of vortex, air ingestion and void fradion, a separate analysis of each issue was performed.

The PBNP-I/;! specification [Reference 9.41, 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., no vortex allowed). Accordingly, PC1 has utilized the guidance of USNRC RG 1.82, Revision 3 [Reference 9.41 to address the vortex issue for the PBNP-112 strainers.

In [Reference 9.41, the USNRC provided generic guidance with respect to PWR sump performance, sump design, and vortex suppression. The subject reference can be utilized as a means of assessing sump hydraulic performance, specifically the issues associated with a potential vortex in the sump.

I I

lD1-8007-07 Vortex Air Ingestion Void Fraction - Rev 4. d ~

Vortex, Me i[ngestifon ek Void Fraction - Point Beach Nuclear Plant -

Unit-4 8 2 Technical Document No. TDI-6007-07 Revision 4 PBNP-fIZ does not have-a sump that colleds post-LOCA water to suppBrt - -

ECCS and CS functions.

Instead, the PBNP-112 utilizes two (2) containment outlet penetrations located in the floor of the containment that are "covererf" by a vertical oval cross-section structure. The structure consists of an outer "coarse" screen (composed of a combination of 24" OD by %" wall pipe and ?4" plate) with slotted openings to facilitate post-LOCA water flow to the pumps. Inside of this structure are two (2) vertical "fine" screen cylinders (one for each containment outlet) that are 13-112" ID. The 'Wne" screen cylinders preclude smaller particles and debris from entering the pumps EIFPeferewccet 9.8 - 9.40, Incciisofve].

Since the PBMP-112 containment outlet structure is being modified by the addition of the PCI sure-low@ suction strainer, the guidance offered by I the USMRC in [Reference 9.41 is not entirely or specifically applicable.

However, the guidance does provide some information that can be utilized in the assessment of the PBNP-I12 strainer configuration with regard to vortex issues.

Ttie "revised" P B R I P - ~ ~ ~

strairier configuration will utilize a pair of horizontally oriented, PC1 sure-ROW@

suction strainers each consisting of I

-eleven.-

(-l-l.~strq~ner~odufess~he_flo~Wfrom~~eestraine~rsd[is~h~rg L---

through attac';hed, pipe and'- '%tii@--t5'tlie ~63stirii-cTntfiimerif'-Fultlet---p-loc,ated in the conljainment floor.

The subject strainer discharge pipe will' t-a-ke-e'fh-

- -.df th-e etti5ttiii -g-. cbiitainm8nt't'ci'u~Ief s*iuctai'~ IED&bflm h-g.-

- 98.9, Bomcllaasive].

The PC1 sure-low^ suction strainer will be analyzed and addressed with 1 respect to vortex issues.

PC1 $SSE-F~OW@ Su~~ition Strainer

-$mrJie~tl 19 The PC1 sure-FIOW@ suction strainer for PBNP-112 is comprised of-1 (1'4).f?3) horizontally oriented modules each containing ten (10) disks. The disks are a nominal 518 thick and are separated I" from each adjacent disk.

The interior of the disks contain rectangular wire stiffeners for support, configured as a "sandwich" made up of three (3) layers of wires -

7 gauge, 8 gauge, and 7' gauge. The disks are completely covered with perforated plate having 0.068" hdes. The end disk of a module is separated approximately 5" from the end disk of the adjacent module.

The 5" space between adjacent modules is covered with a solid sheet Oaigiwaated By:

Page I!? of,a

. I

Vortex, Air Ingestion & Void Fraeltion - Point Beach Nuclear Plant -

Unit-1 & 2 PERFORMANC Technical Document No. TDI-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-vorticesn combined in the interior of a disk to form a vortex, the combination of the wire stiffener "sandwichn 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 [Reference 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 I PBNP-1R 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-112 strainer does not meet the 6" submergence requi~ncTh~nfigur~ti~f~r-PBNP~l/2-results-in~nlyy 2" of submergence to the top of the strainer assembly. However, there is

- - a submergence level of approximately 10.5" 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 stiffener "sandwiches", 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 Laboratoty (ARL) to arrive at the information summarized in Table A-6 of [Reference 9.41.

The PC1 sure-low@ I suction strainer prototype for PBNP-IE 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 with a vortex or vortex development. Also, test observations of the minimum water level above a full size PBNP-112 strainer module showed no evidence of vortexing during testing [Reference 9.181.

lit can therefore be concluded that the configuration of the PBNP-If2 Sure-FIO@

suction strainer will prevent the formation of vortex development.

I Originated By:

1 I

TD1-600747 Vortex Air lngesfion Void Fraction - Rev 4.doc Page I1 of22 I

Vortex, Air Ingestion & Void Flraction - Point Beach Nuclear Piant -

Unit-1 112 Technical Document No. TDI-6007-07 Revision 4 7.2 Air Ingestifiion

. The PBNP-IR specification - [ReriFsmmce 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 speeific value (i-e., ~2%).

Accordingly, PC1 has utilized the guidance of USNRC RG 1.82, Revision 3

[Reference 9.41 to address air ingestion for the PBNP-I12 strainers.

Appendix A and Table A-I of [Rakrence 9.41 indicate that sump performance specifically related to air ingestion is a strong fundion of the Froude Number, 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 SBLOCA, 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 channel flow, the Froude Number, Fr, is defined as the ratio between the force of inertia and the gravitational force [Reference 9.131. This can be expressed as follows:

Equation 1

~r = v 1 (g x L)"~

I Where V=theVeIEif TiBtW ttifoig h m

i tiib~s1bt;---

For PBNP-112 V,

= 3.478 ft/s [Reference 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 circumferential slot width for the core tube hole velocity),

g = gravitational constant, 32.2 fVs2.

I The most conservative value that can be utilized for D is the case of the ratio of core 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.

I From the PBNP-112 Clean Head loss report [Reference 9.Tlj,

= 1.344 1

ft!.

The PBNP-IR Core Tube Design report [Refbrence 9.6J was used in Drawing 10.10 to calculated a slot width of 0.30 in, or 0.025 ft, for the hole I

I velocity of 3.4409.

Therefore, the ratio of the core tube area to the I

circumferential slot widtk;eim be calculated as follows:

Originated By:

Vort~x, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-? & 2 Technical Document No. TDI-6007-07 Revision 4

=- 53:76 ft-Accordingly, value of Fr can be calculated as follows.

1 The calculated Froude Number for the PBNP-1R PC1 ~ure-~low@'

suction strainer is approximately 67% lower than the USNRC guidance found in

[Reference 9.41 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.

I W t h - f o m be concl~d~d-thatth-~-PBNP~l%2-strainers-~il-have-air-ingestion of ~2%.

7.3 Void Fraction The PBNP-IR specifidation [Reference 9.11, 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 remains available at the outlet of the strainer assembly to prevent flashing in the downstream SI-850 valve [Reference 9.1711.

I The pressure available to prevent flashing throughout the strainer I

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 9.17j.

I Originated By:

Date

Vortex, Air Ingestion &Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 Accordingly, PC1 has utilized the guidance of peference 9.2 end 9.153 to I address the void fraction issue for the PBNP-I12 strainem.

Although it is asserted in various regulatory documents that void formation is directly related to air ingestion, this is not correct. Void formation is the 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-112 by the use of the following information provided by Point Beach [Reference 9.1g as input to hydraulic and fluid flow calculations to determine the PBNP-IR Void Fraction.

Calculation Methodofogy I

7.3.1 Evaluation of Flashing across the Strainer Screen [Reference 9.171 I

Pressure Available at Screen > Vapor Pressure I

• P~ir

+ f'vapor

+ P~ubmergenca - Pvelocity - APstrainer > Pvapor, then I

P~ir

= 12.7 psia (14.7 -2.0 psig) is the minimum containment air pressure allowed [PBNP TS 3.6.41, P~ubrnefgence 0 psi 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-PB2-GA-001, and pvdocir/

,= 0 psi Negligible since a flow velocity of 0.0026 fps is expected through the debris bed [PCI Calculation TDI-6007-06 Rev. 51. A similarly small velocity is expected across the screens, then Originated By:

?bhy Date I

TD1-6007-07 Vortex Air Ingestion Void Fraction - RE& 4.doc Page 14of22 1

Vortex,.Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-1 8 2 Technical Document No. TDI-6007-07 Revision 4 PBNP-?I;! [Reference 9.13 defined the containment post-LOCA water temperature as being 212' F. The total debris laden head loss was 3.474 feet of water [Reference 9.161. based on the 212" F water. Converting 3.474 feet of water equates to I

.44 psi.

APstrainer

= 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 1.44 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.

7.3.2 Evaluation of Flashing at the Strainer Assembly Outlet 1

Pmssu~

Available at Assembly Outlet > Vapor Pressure I

P~ir

= 12.7 psia The minimum containment air pressure allowed (14.7-2.0 psig) F S 3.6.41.

PSubmergena,

= 1.3 psi which is the minimum initial sump level provided by 38" of submergence at the strainer assembly outlet [PCI Drawings SFS-PBI -GA-00 & SFS-PB2-GA-001. and pv~tacity

= 0.1 psia the dynamic velocity head (v2/2g) at 2200 gpm, velocity in the 48" elbow is less than 3.6 fps

[Crane 4lOPage 8-14 and Eq. A-31. then AP~trainer

= 13.9 psid is the maximum allowable pressure loss across the entire strainer assembiy to ensure flashing does not occur at the assembly outlet.

Originated By:

TDI-6007-07 Vortex Air Ingistion Void Fraction - R%V 4.doc Page15of22 1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 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.

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 13.9 psid-throughout-the operating-sump temperaturelrange; 7.3.3 Evaluation of Flashing near the Assembly Outlet with SI-850 Valve I

j Pressurn Available at Assembly Outlet >Pressure Required to prevent SI-850 Flashing

Where, I

PSubmsrsence

= 1.3 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-PBZ-GA-001, P~ir

= 12.7 psia is the mihimijrn containment air pressure allowed (14.7-2.0 psig) VS 3.6.41, PSI-850

= 20.9 psia which is 4.2 psig as required at the 51-850 valve assembly at 212 O F to prevent flashing

[PBNP Calc N-92-086 Rev. 41, and to assure no flashing, 2 psi is added to the predicted value [SER 2006-0003, PBNP Calc N-92-086 Rev. 41, and PV~IOG~Q

= 0.1 psia the velocity head (V212g) at 2200 gpm, velocity in the 'l8" elbow is less than 3.6 fps [Crane 410 Paw B-.I 4, and Eq-1-31, TDl-6007-47 Vortex Air Ingestion Void Fraction - Rev 4.doc

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-1 & 2 Technical Document No-eTDI-6007-07 Revision 4 0 j Pvapore2r2F

= 14.7 psia A sump water vapor pressure of 212 'F is required to account for temperature dependent

- -- - changes in-St-850- flash suppression-.pressure requirements and the vapor pressure, then Apstrainer

= 7.7 psid is the maximum allowable pressure loss across the entire strainer assembly to ensure flashing does not occur in the SI-850 valve assembly.

Point Beach further adds that a downstream valve in the strainer suction pathway [References 9.13 and 9.171 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.

I Table 2 - Flashing Margin For Operating Temperature Range. I I

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 St-850 valve throughout the operating range until ternperatlre is reduced below 52 OF.

Originated By:

Date

?/L[R 9 TDI4007-07 Vortw ATr Ingestion Void Fraction - Rev 4.doc Pegellof22 1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 8.8 Conclusions I

- The-result of this-catculation, specifically, the acceptability of.the - issues...

associated with vortex, air ingestion, and void fraction are summarized in Table.

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 PC1 sure-FIO@ Suction Strainers. This specific caiculation completely supports the qualification, installation, and use of the PC1 sure-~lowbP Suction Strainer for Point Beach Nuclear Plant - Unit I

& 2 without any issues or resewations.

TDisO07-07 Vortex Air Ingestion Voq Fraction - Rev 4.doc Page 18 of22 1

Vortex, Air Fraction - Point Beach Nuclear Plant - Unit - I;& 2 Technical Document No. TDI-6007-07 Revision 4 1

I 1

I I

Table 3 - ~alculition Result.

TDi6007-07 Vortex Air Ingestion Void Fraction - Rev 4.doG I

Page18of22 1

Per RG 1.82, Revision 3, if air ingestion is > O%,

the pump NPSH must be corrected by the relationship, NPSH,u,m

(

= NPSH requlmd(llq~ld) x 8, where P =I

+ 0.50ap and u p is fie air ingestion rate (in percent by volume) at the pump inlet flange.

Conventional. calculation methodology indidfetes that no void fraction will occur at the sfminer.

The pressure is sufficient to prevent voiding at fie SI-850 valve through 52 OF.

Cavitation at the valve assembly may occur at colder temperatures for design flow conditions.

Originated By:

Comments Results applicable to the PBNP-112 sure-FIO#

strainer.

Air lngestlon Void Fraction Date.*

formation it Sure-Flow 0% or

<2%

( ~ e e RG

'l.82, Rev.

3,

~ 3 %

(~ef:

USNRC GSI-I 91 Safety Evaluation I

No detrimental RHR~ SI CS pumps NIA

~esults ACCEPTABLE NO Vortex - vortex precluded by the PC1 Issue Vortex ACCEPTABLE I

Air Ingestion

- calculation indicates > 0% but c 2qb.

However, since it has been determi@ that vortex formation will not occur then it can be reasonably concluded that air ingestion will also not occur.,

1

'ACCEPTABLE not pccur at Me strainer -

ipdicates

. Additionally, the' calculation also mat Occur in the SI-850 valve withi temperature decreasing throughout the operating range.

I RHRs 'I '

CS pumps Suction Strainer design and configuration 1

Acceptance Criteria USNRC None (Ref: RG 1.82a Rev.

3)

PBNP-I12 NIA No effects On

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -.

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 1

9.0 References I

9.1 Point-Beach Nuclear Plant Specifcation, Specification No, PB-681, I -

Replacement of Containmenf Sump Screens, Revision 1, August 25,2005 9.2 Information Systems Laboratories (ISL), Inc., Report ISL-NSAD-TR-05-01, Development and lmplemenfation of an Algorithm fbr Void Fmction Cahulafion in the "6224 Cornlation" Somare Package, January 2005, prepared for the USNRC 9.3 U.S. Nuclear Regulatory Commission, Safety Evaluation, Pmssurired Water Reacfor Sump Perforance Evaluation Mefhodology, Guidance Report of the Nuclear Energy Institute (NEI), GSI-191 SE, Revision 0, dated December 6,2004 9.4 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.82, Wafer Soumes for long-Tern Recirculafion Cooling Following a Loss-of-Coolant Accidenf, Revision 3, dated November 2003 9.5 U.S. Nuclear Regulatory Commission, GSI-191 SE, Atfachment V-7, NUREG/CR-6224 Head Loss Temperaturn Assessment. Revision 0, December 2004 9.6 Performance Contracting, Inc. (PCI), Technical Document Number, TDI-6007-03,- COE Tube Design - Poinf Beach Nuclear Plant - Unif 7/2,. -_. -..

Revision 0 9.7 Performance Contracting, Inc. (PCI), Technical Document Number, TDI-6007-05, Clean Head Loss - Point Beach Nuclear Plant - Unif 7/2, Revision 4 I

9.8 Bechtel, Job No. 6118, Drawing No. M-276, Revision 2, Point Beach Nuclear Plant Unit I & 2, Containment Safely injection Sump Requimenfs for Screens I

9.9 Bechtel, Job No. 6118, Drawing No. C-126, Revision 7, Point Beach Nuclear Plant Unit I & 2, Liner Plate - Floor Plan 9.10 Bechtel, Job No. 61 18, Drawing No. C-128, Revision 9, Point Beach Nuclear Plant Unit 4 & 2, Containment Stnrcture Interior Plans af El. 70'-0, El. 27'-0, El. 24'4 & 38'-0 I

TDl-600707 Vortex Air Ing~sbn Void Fraction - Rev 4.60~

Page 20 of 22

'1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I &2 Technical Document No. TDI-6007-07 Revision 4 I

9.1 1 Not Used 1

............. 9.12 PCI; Technical Document Number,.TDI-6007-02, SFS SurFace Area, Flow-.. -.-

and Volume Calculation, Revision 2 I

9.13 Nazeer, Ahmed, Fluid Mechanics, Engineering Press, Inc., 1987 9.14 Not Used I

9.15 USNRC, 6224 Comlafion, publicly available software I

9.16 PC1 Calculation TDI-6007-06, Total Head Loss - Point Beach Nuclear Plant - 11'2, 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-PCI-PB-6028-1001, AREVA Document No. 66-9093957-002, Point Beach Test Report for ECCS Strainer Performance Testing. Dated 1 1/26/2008 TDl-600747 Vortex Air Ingestion Void Fraction - F@v 4.doc

pagi21of22 1

Vortex, Air Ingestion & Void Fraction - Point Beach Nuclear Plant -

Unit-I & 2 Technical Document No. TDI-6007-07 Revision 4 I

40.0 Drawings 10.1-. SFS-PBI -GA-00, Revision 6, Point Beach Unit 1, sure-~lod~tminer,.

Recirc Sump System 10.2 SFS-PBI-GA-02, Revision 6, Point Beach Unit 1, ~ u r e - ~ o d strainer. A I Strainer 10.3 SFS-PBI-GA-04, Revision 5, Point Beach Unit 1, sure-Flod strainer, I

Piping A Layout 10.4 SFS-PB1 -GA-05, Revision 9, Point Beach Unit 1, sure-lo@ strainer.

(

Piping B Layout 1 0.5 SFS-PB1 -PA-71 00, Revision 2, Point Beach Unit 1, ~ u r p ~ o f l strainer. I Module Assembly 10.6 SFS-PB~GA-06, Revision 2, Point Beach Unit 2, 'sure-Flow@ strainer.

I Recirc Sump System 10.7 SFS-PB2GA-02, Revision 9, Point Beach Unit 2, ~urp-Flod Strainer. A I

Strainer 1 0.8 SFS-PB2GA-04, Revision 5, Point Beach Unit 2, sure-Flod Strainer.

I

-. - -. ~i~ing~ssembly Layout 10.9 SFS-PB2-PA-7100, Revision I, Point Beach Unit 2, S U ~ ~ F I O W @

Strainer.

1 Module Assembly I

Originated By:

TD1-6007-07 Vortex Air Ingestion Void Fraction -Rev 4.doC

ENCLOSURE 6 FPL 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 AREVA CALCULATION 51 -9056525, REVISION 001,412812008.

CHEMICAL PRECIPITATION ANALYSIS FOR POINT BEACH NUCLEAR PLANT USING WCAP-I 6530-NP 44 pages follow

AREVA March 16,2009 AREVA-09-0 12 15 Tom Kendall, PE Sr. Technical Advisor Design Engineering Point Beach Nuclear Plant Mr. ICendall, AREVA performed GSI-191 Chemical Effects Calculations for Point Beach Nuclear Plant. The deliverables were AREVA Engineering Information Record (EIR), docwent 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-proprietasy for use by Point Beach Nuclear Plant.

Please feel free to contact me if you have any questions or requests regarding this matter.

Sincerely, Ray Phan '

Manager 1 BOP System Engineering Office: 704-805-223 1 Mobile: 704-575-8924 AREVA NP INC.

An AREVA and Slemens company 7207 IBM Drlve. Charlotte. NC 28262 Tel.: 704 806 2000 - Fax: 704 805 2800 - www,areva.com

ENGINEERING INFORMATION RECORD I

. Document Identifier 51 - 9056525 - 001 I

Title Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP I

PREPARED BY:

REVIEWED BY:

( NAME H. Dergel NAME R. Jetton I

C Signature &-I&

Date 9,/7 Signature h h ~

Date 9 1 ~ ~ t m Technical Manager Statement: Initials Reviewer is Independent.

Remarks:

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 (LOCA), when generated debris or other susceptible materials may be subject to acid.or caustic fluids. This evaluation has been performed based upon pfant-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 4results of this evaluation may be used as inputs into the downstream effects evaluation or as chemical debris mixture inptjts 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.

AREVA NP INC. PROPRIETARY This document, including the information contained herein, is the property of AREVA NP, Inc., an AREVA and Siemens Company (AREVA NP). It contains 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-exportation 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 I of 56

ENGINEERING INFORMATION RECORD Document Identifier 51 - 9056525 - 001 Title Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP PREPARED BY:

REVIEWED BY:

NAME NAME Signature Date -

Signature Technical Manager Statement: Initials Reviewer is Independent.

Remarks:

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 (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 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 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. This is a safety related evaluation.

I AREVA NP INC. PROPRIETARY information contained here document shall be 1

AREVA NP Inc., an AREVA and Siemens company A RWQ c o l @ " ~ ; c ~ o d & t i ~ ~

~ F ~ Q ~ V A

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Page 1 of 56

QF-0528 (FP-E-MOD-07) Rev. 0 DOCUMENT NUMBEW TITLE:

Calc 51-9056525 RNISION:

001 DATE:

9/27/2007 r

REVIEWER'S 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 table 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:

1) 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

2) The inputs have been verified to be correct per the verified inputs provided to AREVA.

However, the calculation results must be applied judiciously, and with a thorough understanding of their derivation and the underlying assumptions.

This calculation does not calculate one single credible "worst case" 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, I

.2, 2.1, and 2.2 should nof 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.I 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.I-1.6 and 2.1-2.6 use a "worst of the worst" 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 1.1, 1.2, 2.1, 2.2, 1.6, 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.

5. Appendices N.1 and N.2 are break-specific runs that were used to assess whether application of the silica inhibition of aluminum corrosion could be credited. In all cases considered, it appears that silica I

Page I of 2 Sheet 1 of 1 Committed to Nuclear Excell Fleet Modification P;icess Design Review Comment Form

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 minimal 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 AREVA. 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 insulation with other types of insulation is acceptable, and that the chemical effects of such replacements are known and bounded by this analysis.

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:

2.5.1 : Replace all asbestos with CalSll 2.5.2: Replace all asbestos with generic fiberglass 2.5.3: Replace all asbestos with NUKON 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 insulation 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 appropriate qualification testing:

[All at max sump Ivl

( P P ~ )

16.3 16.3 16.5 16.5 Case #

2.5 2.5.1 2.5.2 2.5.3 Total Ppt Mass (kg) 194.1 19 194.119 496.599 195.964 Total Al Mass (kg) 19.97 19.98 20.23 20.17

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Multiple PreparerIReviewer Signature Block Page 2 of 56 Name (printed)

Note: PIR designates Preparer (P) or Reviewer (R).

Signature PIR Date PageslSections Prepared or Reviewed

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 clarify 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 9/27/2007 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 I a12a: Bounding Debris Inputs 22 5.2 Case Set 1 bl2b: Debris Generation Case Inputs 23 6.0 RESULT

SUMMARY

23 6.1 Case Set laI2a: Bounding Debris Inputs.......................................................................................

25 6.2 Case Set I bl2b: 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 1.3.

BASE CASE: MAX PH. MAX SUMP VOLUME. MIXED............................ 43 APPENDIX E : CASE 1.4.

SUPPLEMENTAL CASE: MAX PHI MAX SUMP VOLUME. MIXED. SUMP RECIRC @ 60 MINUTES. SPRAY RECIRC @ 100 MINUTES..................................................

44 APPENDIX F : CASE 1.5 -SUPPLEMENTAL CASE: MAX PHI MAX SUMP VOLUME, MIXED. SUMP RECIRC @ 120 MINUTES. SPRAY RECIRC @ 123 MINUTES................................................ 45 Page 4 of 56

Chemical Precipitation 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 PH, MIN SUMP VOLUME, UNMIXED.......................... 48 APPENDIX J : CASE 2.3 - BASE CASE: MAX PH, MAX SUMP VOLUME, MIXED............................. 49 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 PH, 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 Page 5 of 56

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 1 & 2 Table 7-1: Test Summary - Units 1 & 2 Page 21 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. 51-9056525-001 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, "Pressurized Water Reactor Sump Performance Evaluation Methodology,"

Volumes 1 (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 recirculation 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 121.

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 plant-specific 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 (i.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(s) 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 safety 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 steamlwater 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.

Page 9 of 56

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]. 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 6.11. The results of the WOG test program indicated that the predominant chemical precipitates, dependent upon plant buffer type Page 10 of 56

Chemical Preci~itation Analvsis for Point Beach Nuclear Plant Usinn WCAP-16530-NP Document No. 51-9056525-001 and the pH of the sump medium, were aluminum oxyhydroxide (AIOOH), sodium aluminum silicate (NaAISis08), and calcium phosphate (Ca3(P0&) (the latter only identified in the presence of trisodium phosphate (TSP)) [Reference 1 Section 6.1 & 6.41. Other minor silicates could be precipitated. However, their concentration is expected to be minimal with respect to the dominant products (i.e. less than 5%) [I]. T'herefore, 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-specific 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.

Page I 1 of 56

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 and/or 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 AtmosphereISpray Chemistry & Temperature Parameters with Time:

I. To address the extended time period required in 10 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 / 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).

Page 72 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-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 / 2.3) - 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 trainoperation 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 I00 minutes.

> Supplemental Sensitivity Analyses (Cases 1.4 / 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 Sensitivity Analyses (Cases 1.5 / 2.5) - Appendix 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.

Page 13 of 56

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 /

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 = 1). See Appendices A.6 through A.8 for pH evolution at this point.

Sump Pool Volume / 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 lb/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 ail 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 ff3, was extracted from Reference 5.

11. 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.

Page 14 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 General Volumes - Material I Insulation I Debris:

12. For primary base case simulations, the bounding I 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 1 and 2 Debris Generation Reports (see AppendixA.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 not 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 inventory within Reference 5.

Fibrous Debris - Fiberglass Insulation:

17. Point Beach has a variety of mineral wool insulation installed at both Units I and 2 [5]. 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. 4032% 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 10 lb/ft3 [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 Ib/ft3 are standard) [Reference 8 Vol. I Table 4-41.

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 VVCAP-16530-NP evaluation methodology, and therefore it is assumed that these materials are not expected to affect the quantity or type of precipitate generated in the sump following a LOCA [5,12,13]. Therefore, it will be assumed that only miscellaneous fibrous 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 1b/ft3, 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 1 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. 1 Table 4-41.

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 / 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 NEl 04-07 [Reference 8 Vol. I Table 4-11.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-I 6530-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 CalSii 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 [I,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 [I 0,11,23]. 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,I 0,241.

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 14 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 the 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-1-01-44 (embedded within Reference 18) with respect to the first revised coefficient for aluminum release (51.I5271 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 quantity 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 lnteram Calcium release from lnteram With the exception of Aluminum release from Interam, 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 1 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 151. 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 (NaAISisOe), and calcium phosphate (Ca3(P04)2). Aluminum oxyhydroxide will 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 NaAISi3O8. 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 (NaAlSi30e) 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 NaAISi30s 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 1 Section 6.41.

From these outputs, it is possible to determine 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 'key' chemical precipitates [Reference 1 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 'key' 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. I 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 varied 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 1 & 2 Page 21 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51 -9056525-001 5.1 Case Set Ia12a: Bounding Debris Inputs Base Case Analyses:

Cases 1.1 - 1.3 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 between 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 ft3), 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 1 in Appendices B through Dl 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 [5]. Chemical model material and sump volume inputs are reported in Appendix A (identical to the base cases), and model predictions for elemental release and precipitation are reported for Unit 1 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 151.

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.

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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 G, and for Unit 2 in Appendix M.

Case Set 1 bI2b: 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 I

.3 and 2.3 simulations were performed at the maximum sump volume (43,317 ft3), 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 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 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 [5]. 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 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 1 in Appendix N.1, and for Unit 2 and Appendix N.2, 6.0 RESULT

SUMMARY

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).'> Of each unit set of base cases, case number I 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.

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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 "Concentration1' 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 ft3), 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(P0& was observed to occur due to the degradation of debris in containment. Only sodium aluminum silicate (NaAISi308) 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 NaAISi308 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).

6.1 Case Set la12a: 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 fact that the release rate of aluminum from these materials decreases with time as the solubility limit is approached, and that the release rate from aluminum silicate insulation materials decreases with increasing concentration of dissolved aluminum from all sources due to the common ion effect

[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).

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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 1.1/2.1. 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 Dl 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 I degradation.

Chemical model predictions for elemental release and precipitation are reported for Unit 1 in Appendices E and F, and for Unit 2 in Appendices K and L.

Supplemental Analyses: Additional Input Evaluations (WCAP-96785-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 (1.3,2.3):

Silicate Inhibition: for plants exceeding 75 ppm Silicon, 9 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 1.6 and 2.6 in the results above, and model predictions for elemental release and precipitation are reported for Unit 1 in Appendix GI and for Unit 2 in Appendix M.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-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 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:

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

.3/2.3 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 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 1 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:

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" (1.3.1-1.3.5, 2.3.1-2.3.7):

9 Silicate Inhibition: for plants exceeding 75 ppm Silicon, 9 Silicate lnhibition: 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.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 = 43317 ff3 = 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 141.

The "Concentration" values reported in Table 6-1 are included for illustration purposes only.

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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 (NaAISi30B), 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 1 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 (NaAISi30e) for plant-specific conditions that include the initiation of sump recirculation 27 minutes after the accident has occurred6.

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.

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).

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-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 mg/L) is conservative [4]. 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.

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Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001

8.0 REFERENCES

[I]

• WCAP-16530-NP & Spreadsheet, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191", February 2006.

[2] Regulatory Guide 1.82 Revision 3, "Water Sources for Long-Term Recirculation Cooling Following a Loss-Of-Coolant-Accident," November 2003.

[3]

• WCAP-16785-P, "Evaluation of Additional lnputs to the WCAP-16530-NP Chemical Model", April 2007.

[4] 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 lnformation Transmittal (DIT) from Tom Kendall of Point Beach to Support," April 2006 (NPL 2006-0052).

[7] 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.

[9] NUREGICR-6913, "Chemical Effects Head-Loss Research in Support of Generic Safety Issue 19Iv, U.S. NRC, December 2006.

[ 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 Support GSI-191", U.S. NRC, April 2005.

[I21 AREVA NP DOC. NO 32-5050092-002, "DEBRIS GENERATION EVALUATION FOR POINT BEACH NUCLEAR PLANT UNIT 1," June 2006.

[I31 AREVA NP DOC. NO 32-5052938-003, "DEBRIS GENERATION EVALUATION FOR POINT BEACH NUCLEAR PLANT UNIT 2," June 2006.

[I41

• WOG-06-102, Errata to WCAP-16530-NP, "Distribution of Errata to WCAP-16530-NP, "Method for Evaluating Post-Accident Chemical Effects in Containment Sump FluidsJJ (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 71

• 06-07-270 "New Settling Rate Criteria for Precipitates Generated in Accordance with WCAP-16530-NP (PA-SEE-0275)", June 2007.

[ I 81

• 06-07-282 "Instructions to Evaluate Specific lnputs for WCAP-16785-NP; PA-SEE-0354, "Incorporation of Additional lnputs in the Chemical Effects Spreadsheet1'," June 2007.

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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]

• 06-06-273, 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 (~~~'2007-0145).

[23] NUREGICR-6912, "GSI-191 PWR Sump Screen Blockage Chemical Effects Tests:

Thermodynamic Simulations", U.S. NRC, December 2006.

[24] NUREGICR-6915, "Aluminum Chemistry in a Prototypical Post-Loss-of-Coolant-Accident, Pressurized-Water-Reactor Containment Environment", U.S. NRC, December 2006.

[25] AREVA NP Doc. 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 document control system, but can be retrieved through the Westinghouse Owner's Group. Per AREVA NP administrative procedure, 0402-01 Appendix 2, these PMlPE Signature:

Date:

9/*/~,>

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 PostAccident - Base Cases Point Beach List of Inputs -Material Quantify& InputAffected Metallic Aluminum Calcium Silicate Aluminum Silicate Page 33 of 56

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Point Beach List of Inputs -Material Quantity & Input Affected 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 PostAccident - Bounding Case ~ a s e s ~

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.l for Bounding Cases and A.3lA.4 for Debris Generation Cases)..

Page 35 of 56 Class Calcium Silicate E-glass M~neral Wool Additional Spreadsheet Inputs (ff3)

CalSil Insulation Asbestos Insulation F~berglasslnsulation NUKON TempMat Mln-Wool Unit 2 Case6 5.5 2.37 53.7 0

0 0

I Unit2 Case5 8 g 7 116.07 114.7 1046.65 89.42 311.3 Unit 2 Case7 111.84 116.07 107.35 937.77 89.77 291.43 Unit 2 Case3 83.87 68.63 53.45 422.14 20 130.49 Unit2 Case2 122.72 116.07 90.57 849.5 99.57 323.2 Unit2 Case4 83.87 80.72 53.45 566.25 29.48 18.37 Unit I Case I 113.05 296.74 179.38 0

23.44 203.11 Unit 1 Case4 63.57 159.7 98.75 0

11.82 0

Unit I Case2 110.5 275.37 125.87 0

20.61 0

Unit 1 Case5 89.36 296.74 181.4 0

23.44 218.99 Unit I Case3 59.53 160.81 7802 0

7.3 0

Unit 2 Case1 88.46 116.07 107.48 1001.1 89.28 267.21

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 Unit 1 Debris Generation Case Inputs - All Cases [518 Unit 2 Debris Generation Case Inputs - All Cases [519 Unit 1 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-16530-NP Document No. 51-9056525-001 A.5 Material Specific Density Values Point Beach List of Inputs - Material I Parameter Densities Calcium Silicate Page 37 of 56

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 Point Beach List of Inputs - Temperature 8 pH Profiles Spray Recirculation at 77 minutes Page 38 of 56 Start of Spray - 0 seconds (Reference 5)

I Start of Sump Recirculation to Core - 27 minutes (Reference 5)

I Start of Spray Recirculation - 77 minutes (Reference 5)

Assumed Initation of Sump Mixing End of Spray Injection - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 3

Time (set) 30 60 120 180 200 400 600 800 1000 1200 1400 1600 1800 3200 4600 6000 7400 8800 10200 11600" 13000 21600 46400 86400 172800 259200 345600 432000 864000 1296000 1728000 2160000 2592000 min EL 0.5 1.0 2

3 3

7

$10 13 170 20 23 27 30

' 53 77 100 123 I 147 1,170 193

' 217 360 773,

' 1440,

2880 4320 5760 1

7200 14400

,21600, 28800 36000 43200 Sump pH 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9 5 9.5 9.5 9.5 9 5 9.5 9.5 9.5 9.5 9.5 9.5 9 5 9.5 9.5 9.5 9.5 9.5 Steam or Spray pH 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 9.5 9.5 9 5 9 5 9 5 9.5 9 5 Sump Temp.

(OF) 286 286 286 286 286 286 286 286 286 286 284 282 281 279 273 269 266 263 261 259 255 252 249 21 3 194 182 176 171 168 157 151 147 143 141 Containm ent Temp.

(OF) 286 286 266 286 286 286 286 286 286 286 266 286 286 286 274 266 261 256 252 250 250 250 250 hr O-0 0

0 0

0 0

0 0 I 0

0 0

0 1 I 1

2 2

I 2

1 3

3 4

'6

' 13 24 48 72 96 120 240 360,

480 600 720 days 0

0 0

1 0 0

0 0

0 0

0 0

0 0

,O 0

, 0,

0 0

I 0

0 0,

0 0

, 1 1

2 3

4 5

I 10 15 20 25 30

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 Point Beach List of Inputs -Temperature &pH Profiles Spray Recirculation at 100 minutes Page 39 of 56 I

I Start of Spray - 0 seconds (Reference 5)

I I

1 1

8 I

1 1

I I

I a

I I

I Start of Sump Recirculation to Core - 60 minutes (Reference 5)

Start of Spray Recirculation - 100 minutes Assumed Initation of Sump Mixing I -

I End oflSpray injection - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> I

1 I

I I

(

)

1 1

I I

r I

I I

I /

I 1

I Containm ent Temp.

(OF) 286 286 286 286 286 286 286 286 286 286 286 286 286 286 274 266 261 256 252 250 250 250 250 Sump Temp.

(OF) 286 286 286 286 286 286 286 286 286 286 284 282 281 279 273 269 266 263 261 259 255 252 249 213 194 182 176 171 168 157 151 147 143 141 Sump pH 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 Steam or Spray pH 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 9.5 9.5 9.5 9.5 9.5 9.5 Time

!set) 6 30 60 120' 180 200 400 600 800 1000 1200 1400 1600 1800 3200 4600 6000 7400 8800 10200 11600 13000 21600 46400 86400 172800 259200 345600 432000 864000 1296000 1728000 2160000 2592000 hr 0

0 0

0 0

0 0

0 1

0 0

0 0

0 1

1 1

2 2

2,

3 3

4 6

13 24 48 72 '

96 120' 240 1

360

' 480 600 720, ',

I min 0

0.5 d.0 2

3 3

7 10 13 17 1

20 23 27 t

30 I

53 77 I 100 123 147 170

193, 217 360 773 1440 12880 4320 5760

'7200 14400 21600 28800 36000 1 43200 days 0

0 0

0 0

0 0

0 I 0 I

0 O

0 0

0

, ' O 0

0 0

0 0

0 0

0 I I

2 3

4,

5 10 15 I 20 25

, 30

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-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 & pH Profiles Spray Recirculation at 123 minutes Page 40 of 56 hr 0

0 0

I O 0 '

0 0

0 0

r 0

0 0

0 1

1

$ 1 2

1 2

2 3

3 4

6 13 24,

48 72 96 120 240 360'

, 480 600 720 Time (sec) 6 30 60 120 180 200 400 600 800 1000 1200 1400 1600 1800 3200 4600 6000 7 4 s 8800 10200 11600 13000 21600 46400 86400 172800 259200 345600 432000 864000 1296000 1728000 2160000 2592000 days 0

0 0

0 0

0 0

O 0

0 0

0 0 '

0 0

0 0

0 0

1 0 0

0 0

1 0

1 1

2 3

4 5

10 15 20 25 30

, min 0

0.5 I

1.0 2

3 3

7 ',

10 13 17 4 1

20 23 27 3 0 53 77 100 123 I

' 147 170 193 217 360 773 1440 2880 4320 5760 7200 14400 21600 28800 36000 43200 Containm ent Temp.

('F) 286 286 286 286 286 286 286 286 286 286 286 286 286 286 274 266

261, 256 252 250 250 250 250 I

Start of Spray - 0 seconds (Reference 5) i 1

1 1

1 I

I

, /

L I

I I

9 1

I '

Start of Spray Recirculation - 123 minutes '

Assumed lnitation of Sump Mixing 1 I I

I I

I I

I v

I 8

End of Spray Injection - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> I

I I,

I I

I 1

I

/

I I

Sump plj 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9 5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9.5 9?!

9.5 9.5 9.5 Sump Temp.

('F) 286 286 286 286 286 286 286 286 286 286 284 282 281 279 273 269 266 263 261 259 255 252 249 21 3 194 1 82 176 171 168 1 57 l5?

1 47 1 43 141 Steam or Spray pH 10 10 10 I 0 10 I 0 10 10 I 0 10 10 10 10 10 10 I 0 10 10 9.5 9.5 9.5 9.5 9.5

Chemical Precipitation Analysis for Point Beach Nuclear Plant Using WCAP-16530-NP Document No. 51-9056525-001 APPENDIX L: CASE 2.5 - SUPPLEMENTAL CASE: MAX PH, MAX SUMP VOLUME, MIXED, SUMP REClRC @ 120 MINUTES, SPRAY REClRC @ 123 MINUTES L.1 Elemental Releases and Precipitation - Case 2.5 Page 51 of 56