ML102870100
| ML102870100 | |
| Person / Time | |
|---|---|
| Site: | Catawba, McGuire, Mcguire  |
| Issue date: | 09/30/2010 |
| From: | Repko R T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-04-002 | |
| Download: ML102870100 (135) | |
Text
RIDuke T. REPKO T7ýýDukeVice President& c7Energy McGuire Nuclear Station Duke Energy MG01 VP /12700 Hagers Ferry Rd.Huntersviyle, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com September 30, 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Duke Energy Carolinas, LLC (Duke Energy)McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Draft Responses to NRC Request for Additional Information (RAI) related to Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" On September 13, 2004, the NRC issued GL 2004-02. The GL requested that all pressurized-water reactor licensees (1) evaluate the adequacy of the emergency sump recirculation function with respect to potentially adverse effects associated with post-accident debris, and (2) implement any plant modifications determined to be necessary.
Duke Energy has been actively engaged'in these evaluations, including completion of emergency sump strainer modifications at both McGuire and Catawba. Duke Energy has continued to communicate with the NRC both formally and informally on progress to address remaining issues related to strainer qualification.
The Supplemental Responses to GL 2004-02 were formally sent to the NRC by McGuire's submittals dated February 28 and April 30, 2008 and Catawba's submittals dated February 29 and April 30, 2008. From these submittals, the NRC developed plant-specific RAIs that were received by McGuire on November 18, 2008 and Catawba on November 21, 2008.Duke Energy discussed the McGuire and Catawba RAls with the NRC in a public teleconference on September 1, 2009 in order to clarify the path forward regarding prototype strainer testing and several industry-wide issues affecting credited analytical refinements.
www. duke-energy.
corn U. S. Nuclear Fe,.atci' Commission September 30, 2010 Page 2 In another teleconference with the NRC on June 9, 2010, Duke Energy agreed to provide plant-specific draft RAI responses by September 30, 2010 to assure clear understanding of methodology prior to formal RAI response submittal.
The purpose of this letter is to provide the draft RAI responses for both McGuire and Catawba. Per agreement with the NRC Project Manager, this letter does not need oath and affirmation since it contains only draft responses.
Regulatory commitments, if any, will be specified in the final RAI response submittal.
Attachment 1 provides a brief overview of the McGuire and Catawba GL 2004-02 resolution path and recent changes, including ongoing prototype strainer testing.Attachment 2 provides the draft RAI responses for McGuire.Attachment 3 provides the draft RAI responses for Catawba.Duke Energy will be working with the NRC Project Manager to arrange a follow-up teleconference early next month to discuss any points of clarification on these draft RAI responses and to establish a date for submitting final RAI responses.
If any questions arise or additional information is needed, please contact Tony Jackson at (704) 382-8882 or P. T. Vu at (980) 875-4302.Very truly yours, Regis T. Repko Attachments U. S. Nuciez;;
Commission September 30, 2010 Page 3 xc (with Attachments):
L. A. Reyes, NRC Regional I1 Administrator Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J. H. Thompson, NRC Project Manager 11555 Rockville Pike Mail Stop 8 G9A Rockville, MD 20852-2738 G. A. Hutto, Ill, NRC Senior Resident Inspector Catawba Nuclear Station Russell Keown, Supervisor Analytical
& Radiological Environmental Services Division 2600 Bull Street Columbia, SC 29201 Sandra Flemming, Director Analytical
& Radiological Environmental Services Division 8231 Parklane Road Columbia, SC 29223 Susan E. Jenkins, Manager Radioactive
& Infectious Waste Management Division of Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Tom Knight Contamination Mitigation Section Bureau of Land and Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 W. L. Cox, III, Section Chief Division of Environmental Health, Radiation Protection Section North Carolina Department of Environment and Natural Resources 1645 Mail Service Center Raleigh, NC 27699-1645 Attachment 1 Preface to Draft Responses NRC Request for Additional Information
- September 30, 2010 Preface to Draft Responses NRC Request for Additional Information September 30, 2010 Since the formal submittal of the McGuire and Catawba Nuclear Station Generic Letter (GL) 2004-02 Supplemental Responses in the spring of 2008 and the receipt of Requests for Additional Information (RAI) from the NRC in the fall of that year, Duke Energy has been actively engaged with the staff in resolving questions and concerns related to the Emergency Core Cooling System (ECCS)Sump Strainer design and qualification for each station. This interface has led to specific changes in the original approach addressing GL 2004-02 for these two facilities, and also to a clearer understanding of the additional actions required for final resolution.
In responding to the RAIs, Duke Energy has incorporated the following differences from the original GL 2004-02 approach: " McGuire and Catawba are no longer crediting Zone of Influence (ZOI)refinements for fiberglass insulation associated with WCAP-1 6710-P "Jet Impingement Testing to Determine the Zone of Influence (ZOI) of Min-K and NUKON Insulation for Wolf Creek and Callaway Nuclear Operating Plants."" McGuire and Catawba are replacing a significant amount of low density fiberglass (LDFG) insulation in specific areas of lower containment with reflective metal insulation (RMI) via Fiber Insulation Replacement Projects (FIRP). The Catawba FIRP scope is complete and the McGuire FIRP scope is in progress.* McGuire and Catawba have submitted license amendments to the NRC for ECCS Water Management modifications, which include revisions to post-accident response that reduce recirculation flowrates through the ECCS Sump Strainer and decrease the predicted volume of transported sump pool debris. Catawba's license amendment has been approved.* ECCS Sump Strainer performance for both stations will be confirmed in 2010 via a prototype Confirmatory Integrated Test (CIT), which will conservatively simulate conventional debris, debris transport, and sump pool temperature and chemistry profiles.
The CIT will assess head loss (including chemical effects) across a debris-laden representative strainer array and will be run for 30 days to match the ECCS mission time.The attached draft RAI responses integrate the differences identified above and represent the current Duke Energy understanding and proposed resolution of remaining open issues related to the McGuire and Catawba GL 2004-02 Supplemental Responses.
McGuire and Catawba discussed the response methodology for each NRC RAI question during a public teleconference with the technical staff on September 1, 2009 to facilitate final response development.
The NRC's comments/clarifications on the RAI questions and proposed response methodology received during the public teleconference, provided they have not been altered by the changes identified above, have also been incorporated.
Attachment 1 Page 1 of 1 Attachment 2 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Responses September 30, 2010 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M1. Please state whether the testing identified in the test report WCAP-16710-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) of Min-K and Nukon Insulation for Wolf Creek and Callaway Nuclear Operating Plants," was specific to the McGuire Nuclear Station, Units 1 and 2, (McGuire).
insulation systems. If not, please provide information that compares the McGuire encapsulation and jacketing systems structures with the systems that were used in the testing, showing that the testing conservatively or prototypically bounded potential damage to the insulation materials.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16710-P ZOI issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.In the McGuire GL 2004-02 SuppklE d 2/28/08 and the McGuire GL 2004-02 Amended Supplemental Re ns'e`a 4/30/, e quantity of fibrous debris generated from destr r insula and d in OCS sump pool was determined using t ffluen(oth the NEI 04-07 Guidance Report ( the ass ted N ty Eval ton (GR/SE)) and WCAP-16710-P "Jet Imping t Test to Dete e the Zone of Influence (ZOI) of Min-K and NUKON nsulatiol Ce and way Nuclear Operating Plants," revision lyt urýa S t m Strainers were initially designed using I 04- gui e relatid insulation ZOls, and the subsequent lntegr Prototype IPT) hemica ects used WCAP-refined ZOI values.In the time iod since th )ple al/Amended Supplemental Responses were submitted, i(re has de ined tffi reliance on the WCAP-16710-P jacketed fiber insulation ZOI hments is. longer necessary, and therefore the fibrous debris quantities genera om d yed fiber insulation will be based on the NEI 04-07 GR/SE only. For bo c and unjacketed Nukon fiber insulation types located in the postulated break z se ZOI used for quantification of debris is 17D as identified in the GR/SE, Table 3-2. or the Thermal-Wrap fiber insulation applications within the postulated break zones, the 17D ZOI is also invoked, consistent with the LDFG destruction pressures discussed in the GR/SE, Section 11.3.1.1.This change in approach is primarily due to the implementation of Fiber Insulation Replacement Project (FIRP) modifications, which are currently ongoing and will be completed within two Unit refueling outages. These modifications replace existing fiber insulation on the reactor coolant loop piping (hot legs, cold legs, and crossover legs), the steam generators, and the reactor coolant pumps with reflective metal insulation in Attachment 2 Page 1 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response both McGuire units. This scope reduces the destroyed fiber insulation load within the postulated 17D break ZOls significantly.
Attachment 2 Page 2 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M2. Considering that the McGuire debris generation analysis diverged from the approved guidance contained in NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Revision 0, please provide details on the testing conducted that justified the ZOI reductions for jacketed Nukon. The information should include the jacket materials used in the testing, geometries and sizes of the targets and jet nozzle, and materials used for jackets installed in the plant. Please provide information that compares the mechanical configuration and sizes of the test targets and jets, and the potential targets and two-phase jets in the plant. Please provide an evaluation of how any differ in jet/target sizing and jet impingement angle affect the ability of the ins system to resist damage from jet impingement.
Please state whether thefestg described in test report WCAP-16710-P was bounding for the McGq' " n systems. If not, please provide information that compares the Mc, n enca tio king systems structure with the system that ,,sed in the g, showing that the testing conservatively or prototypical nded potential e to the insulation materials.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16710-P ZOI issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.As den the McGuire GL 2004-02 Suppl ntal Respon" dteda/08 and tt-re McGuire GL 2004-02 Amended Suppler I Responsed 4 , the quantity of fibrous debris generated from destroyed insulation a, epos in the ECCS sump pool was determined using the Zones of n nce (ZOI ýescribed in both the NEI 04-07 Guidance Report (and the associated Safety Euation (GR/SE)) and WCAP-16710-P "Jet Impingement Testing to Determi e Z f Influence (ZOI) of Min-K and NUKON Insulation for Wolf Creek and Calla ear Operating Plants," revision 0. Specifically, the McGuire ECCS Sump aers were initially designed using the NEI 04-07 ZOI guidance relating to fiber insulation ZOls, and the subsequent Integrated Prototype Test (IPT) for chemical effects used WCAP-refined ZOI values.In the time period since the Supplemental/Amended Supplemental Responses were submitted, McGuire has determined that reliance on the WCAP-16710-P jacketed fiber insulation ZOI refinements is no longer necessary, and therefore the fibrous debris quantities generated from destroyed fiber insulation will be based on the NEI 04-07 GR/SE only. For both jacketed and unjacketed Nukon fiber insulation types located in Attachment 2 Page 3 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response the postulated break zones, the ZOI used for quantification of debris is 17D as identified in the GR/SE, Table 3-2. For the Thermal-Wrap fiber insulation applications within the postulated break zones, the 17D ZOI is also invoked, consistent with the LDFG destruction pressures discussed in the GR/SE, Section 11.3.1.1.This change in approach is primarily due to the implementation of Fiber Insulation Replacement Project (FIRP) modifications, which are currently ongoing and will be completed within two Unit refueling outages. These modifications replace existing fiber insulation on the reactor coolant loop piping (hot legs, cold,, legs, and crossover legs), the steam generators, and the reactor coolant pumps w. flective metal insulation in both McGuire units. This scope reduces the destroy r insulation load within the postulated 17D break ZOls significantly.
Attachment 2 Page 4 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M3. Please clarify if unjacketed Nukon is present in the McGuire containment and, if so, please state whether the 17D ZOI was used instead of the 7D ZOI. Please provide the resultant debris quantities for unjacketed Nukon. (Section 3(b)(2) of the supplemental response sent by letter dated February 28, 2008, stated that.:.unjacketed Nukon was present within the evaluated ZOls. The supplemental
%- response further stated that test report WCAP-16710-P demonstrates a refined 7D ZOI for jacketed Nukon@ insulation, but was silent with respect to how unjacketed Nukon was handled with respect to ZOI reduction from 17D to 7D.)McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.Unjacketed Nukon insulation (alorv within the McGuire containment bre As discussed in the- 40 RAI jacketed fiber insu i refin nts fiber insulation withi postuu fd have aZOIof17D pe NEI/ 07 This chapei Replo6metýF com Itd th insulati th the steaim% n both McGuire volume within t:tt r fiber insulation systems) is present que ion i longer implementing the ide C 10-P. As a result of this, all break sn (jackete or unjacketed) is assumed to th, plementation of Fiber Insulation atio *c are currently ongoing and will be iutages'.hese modifications replace existing fiber piping (hot legs, cold legs, and crossover legs), Dý)oolant pumps with reflective metal insulation in n scope reduces the destroyed fiber insulation eak ZOls significantly.
With the new insul o tYration, jacketed and unjacketed fiber insulation within the assumed 17D ZOI coh trfLtya total of 422 ft 3 of fiber to the ECCS sump pool for the limiting break location at cGuire (Unit 2 RC Loop B Hot Leg).Attachment 2 Page 5 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M4. Please state whether or not the break location selection was revisited when the ZOI for fibrous insulation was changed from 17D to 7D. If break selections were not revisited, please provide the rationale for not doing so. If the break selections were revisited, please provide the top four breaks in terms of debris generation for the 7D ZOI. (The supplemental response sent by letter dated February 28, 2008, indicates only that the break locations already identified for a 17D ZOI were reassessed for debris quantity generation and confirmed not to have changed relative ranking).McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as"acceptable" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.Asdiscussed in teresponse 1st~bi, McG no longer implementing the jacketed fiber insulation refinementsn'tifdi CAP-60-P.
As a result of this, all fiber insulation withint ulated zes ete injacketed) has a ZOl of 17D per the NEI 04ulate This change in app is pri mpleme ion of Fiber Insulation Replacement Project ) m *cations, h are currently ongoing and will be complete " " o Un es. ,se modifications replace existing fiber insular co oop h gs, cold legs, and crossover legs), the st generator th ctor co'o6 a umps with reflective metal insulation in both *re'units.
Thi ope 'es the destroyed fiber insulation load within the postulate
\ break ZO nliafic The replace f the fibe ulation systems in FIRP scope has the same effect from a break location ulation p epective as any other fiber reduction refinement would;therefore the McG reaý ations were revisited using the post-FIRP insulation configuration to ensu iting break was identified.
Post-FIRP break categor-/relative ranking did not change; these were documented in McGuire GL 2004-02 Supplemental Response dated 2/28/08, Section 3(a)1: 1. RCS breaks 2. Locations generating 2 or more types of debris 3. Locations with the most direct path to the strainer 4. Locations with the largest potential particulate/fiber ratio Attachment 2 Page 6 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response 5. Locations for thin-bed potential The top four break locations (all in category 1) identified with the post-FIRP insulation configuration in terms of debris generation are: " RC Hot Legs* RC Cold Legs* RC Crossover Legs* Pressurizer Surge Line The limiting break location at McGuire did not chN configuration was evaluated along with the as (remains the Unit 2 RC Loop B Hot Leg). No closer to the ECCS strainer and therefor p post-FIRP insulation D ZOI for fiber insulation B Hot Leg break is-debris.Attachment 2 Page 7 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M5. Provide information that compares the ability of the McGuire fibrous jacketing system and the test report WCAP-16710-P tested jacketing system to resist steam jet damage. Please provide information that demonstrates that the McGuire jacketing is at least as structurally robust as the jacketing that was subjected to the test report WCAP-1671 0-P steam jet impingement testing.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16710-P ZOI issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.McGuire is no longer implementi the jackete ation refine identified in WCAP-167 10-P. As a result of fiber insui ithin the postulat d break zones (jacketed or unjacketed) ha 17D pe EI 04-07 GR/SE.Reference the response to RAI que n 1 subm r fourther information.
Attachment 2 Page 8 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M6. Please provide information that verifies that test report WCAP-16710-P testing used to justify a ZOI reduction from 17D to 7D for jacketed fiber insulation was conducted prototypically or conservatively.
Include information on nozzle size, target size,..and the various test configurations (jet-to-target distance and relative angle, location of jacket seams, etc) conducted to show that the testing was prototypical or conservative.
McGuire Response: McGuire is no longer implementi, WCAP-16710-P.
As a result oft zones (jacketed or unjacketed) ha, Reference the response to RAI qu, rulation refinem identified in ,ithin the postulated break 1E 04-07 GR/SE.t ltf further information.
Attachment 2 Page 9 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M7. Please provide the fibrous size distribution (including debris amounts determined) for the debris generation calculation based on the 7D ZOI.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16710-P ZOI issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.As discussed in the response to RAI quest 1, jacketed fiber insulation refinements (i.e., th a result of this, all fiber insulation thin the pos unjacketed) has a ZOI of 17D p El 04-07 plementing the 1-16710-P.
AsýJed or This change in approach is primaril Replacement Project (FIRP) modifi'completed within two rueling insulation on the reaco c 1 the steam genera FS and the' ,ne ac both McGuire units. s cop c postulated 17D break With _ sltinchfgr contrih a total o locatio cGuire (U RC The size d ution of the tsytE modifications Mthis limitin reak e I" ation of Fiber Insulation e cu ly ongoing and will be oII fi ns replace existing fiber leqs, and crossover legs), is Wvt ireflective metal insulation in ,ed fiber insulation load within the hsulation within the assumed 17D ZOI sump pool for the limiting break.er insulation (before and after FIRP is hown in Table 7S-1 below.Attachment 2 Page 10 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 7S-1 McGuire Destroyed LDFG Debris Volumes -Limiting Break Size Distribution j Pre-FIRP..
Post-FIRP Nukon and Thermal-Wrap Fiber Volume Fiber Volume Low Density Fiberglass (LDFG) Generated Generated (17D ZOI) (17DZOI)Attachment 2 Page 11 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M8. Please provide details regarding the tags and labels equipment qualifications and-engineering judgments used as basis for reduction of tag and label quantities which were originally assumed to fail and reach the sump. Provide the technical basis for the conclusion that tags and labels outside the crane wall in lower containment are capable of withstanding post-loss-of-coolant accident (post-LOCA) conditions.
Justify the application of the Institute of Electrical and Electronics Engineers (IEEE)Standard 323-1974, "IEEE Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating Stations," in qualifying Electromark labels for a post-LOCA environment.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time, with noted clarifications from the staff This draft response is consistent with the one discussed on that call.The assumptions and engineering ju en in th Guire tag and label reduction evaluation or d subsea t to the ta label assessment were provided in the res s 3(d) d f Ehclosure 2 of the McGuire GL 2004-02 Suppl ntal Renspbnse dat a 2 08.The initial tag and lab ;sess t include ly one refinement in the form of a qualified tion. it assumed that metal tags hung with cor ons wo eS .e'ot fail, or would sink and not transport.
Thes st metalf re thse samea Electromark labels, which are discus scHter in this r -k)nse.The transý tag and Ia quan ions by area of Containment were reported in Table 3D3-2"Fclosure 2 9e Mcouire GL 2004-02 Supplemental Response dated 2/28/08. The info tion in t table is recreated below in Table 8S-1 for convenience.
Attachment 2 Page 12 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 8S-1 McGuire Refined Miscellaneous Latent Debris Quantities Lower Containment Lower Containment Upper Ice Type of Debris (Inside Crane Wall) (Outside Crane Wall) Containment Condenser Total Stickers & Labels (ft 2) 64.96 30.34 14.43 8.92 118.65 Plastic Tags w/Adhesive (ft 2) 33.32 8.79 11.2 8.02 61.36 Plastic Hanging Tags (ft 2) 11.70 4.33 0.17 16.20 RMI ID Stickers (ft 2) 103.14 32.81 < N/A 135.95 Ice Condenser Debris (ft 2) N/A N/A N/A 15.30 15.30 Total (ft 2) N/A N/__ N/A N/A_ 347.5 As stated in item 3(i)5, Electrom labels lc the Crane in the lower containment at McGuire have b e uated as e of withstandinghe limiting break, and thus were removed fron ntificatio ag and label debris assumed to transport to the ECCS Sump Strai r. ml ark el located inside the Crane Wall in lower containm re assum o fai For the purposes o e, "lab re iable sticker or marker that is affixed with esive. term "t rs to ai ilatively thick rigid plastic tag or placard that is hun affix 'vith adh Lower Cvith amr .In an ,ide l the o a f the lower containment outside the Crane 1 and not i the ondens I ower Plenum are located between the Ice Conner end walý In ti oms above the Pipe Chase. Plastic tags in these vicinities, a nerally out the b ZOls and are assumed to deform, but not become over kable (i.e., ewill nodeform enough to pass through an obstruction that has a smal mension n the tag).Ice Condenser Re, o Tags and labels locate n the Lower Plenum of the Ice Condenser (and outside the Crane Wall) are assumed to fail since the break energy is directed into this plenum by design.Tags and labels located within the Upper Plenum of the Ice Condenser would not be expected to fail immediately during the initial venting of air and steam; however, exposure to the post-LOCA environment and containment spray may lead to eventual detachment.
A minor portion of the tags and labels located in the Upper Plenum of the Ice Condenser are not located above the ice basket array and are located above Attachment 2 Page 13 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response horizontal surfaces.
As these tags and labels fail, they will fall straight down and are not expected to transport further due to containment spray.It is likely that some of the remaining tags and labels that fail within the Upper Plenum of the Ice Condenser will fall directly into the ice baskets themselves.
Given the ice baskets are made of perforated sheet metal with 1" x 1" holes and the bottom of the baskets are covered by a grid and wire mesh, any tags and labels that fall into the ice baskets will not be able to exit. In terms of cross-sectional area, it is conservatively assumed that ice basket openings comprise 50% of the ice basket array. Tags and labels that do not fall into the ice baskets themselves c all into the space between the baskets and the lattice frame which provide supp the baskets. While the lattice frame does create a tortuous path for the ta bels, it is not possible to conservatively estimate an appropriate quantity els "would remain within the lattice structure; therefore, no reduction was t "ior tags abels that may fall into this area.Containment Elevation 738'+3" Reduct The rooms above the Pipe Chase (i.e., at corn en ation 738 are not subject to jet impingement or coi ent spray'.tmIi) the rooms wi t be flooded, but as the accident progresses t um floo ,'ation will be reached. Access to the rooms is gained through an op floor fr e pipe chase below. Once the rooms are flooded, velocities in oor expec o bedto very low and tags and labels would not trans the pipe case O s an els directly above the floor opening are aSme-olasott i pe/he Upper Containme duc The majority of tags li el Upper tainment are located between the ends d(5 me r Return fan pit, and around the perso tcn.- and Is in thes a, 'are subjected only to containment that det re e ed to fa saight down and there are none that would 6 \ected to fa ctly the Refueling Canal.It is conser ti. ly assume at all and labels that reach the Containment Air Return fan pit i ass to th wer containment via the fan pit drain. A majority of tags and labels outsi fan pit located directly above grated platforms.
It is judged that many of these s "if easily captured by the 1" x 4" grating and thus the quantity of labels abo t ating is reduced by 75%.Although highly unlikely,i is conservatively assumed that all tags and labels that detach and fall to the concrete operating floor will be transported over the 3 inch curbing around the Refueling Canal and through the elevated canal drains to lower containment.
The quantity of tags and labels overflowing the Refueling Canal curb represents about 50% of the Upper Containment total quantification identified in Table 8S-1.Attachment 2 Page 14 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Qualification of Electromark Labels The Electromark labels located in containment were qualified for the LOCA environment via a comprehensive test program. The purpose of this program was to demonstrate the suitability of application for pressure sensitive markers in being able to remain in position (on equipment or structures) throughout a specified lifetime, including-,. background radiation followed by a simulated LOCA. The safety function demonstrated
'was that the markers would remain affixed to the equipment or structure without falling off.This test program was conducted under the general 323-1974 "IEEE Standard for Qualifying Class 1E&Generating Stations".
The various phases of the 1) Heat Aging -Simulation of long-term e .sf!oui typical ambient temperatures and a en years. On the basis of the suggest and p and IEEE 275, the 10 0 C rule was utilie to e demonstrate a qualified life period by a temperatures.
- 2) Radiation Aging -At the c u of the th were inspected for degradat a of fu cobalt-60 source of gamma ra4 ion n hour until a tol ulatedd of samples we n in ted agr 3) LOCA simulat -The pies we stalk subjected to an one expos, st r~es as suggested in IEEE!nt for Nuclear Powerýre outlined following:
re tol rnbient conditions at ba period of several mntined in IEEE 117 n a temperature to at ele aging period the samples aind then exposed to a-srate of 0.5 Mrads per a' i'en received.
The lradation.
funcinside a pressure vessel and eam and chemical spray for a gested IEEE 323-1974 profile. Atagain inspected and compared tion.Attachment 2 Page 15 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M9. Please provide the technical basis for the latent fiber and particulate total mass calculation.
Include a description of surface types sampled, the number of samples per surface type, the accuracy of the mass measurement, the method of computing the densities for specific areas, and the extrapolation to the scale of containment.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.Latent fiber (i.e., lint) quantities and latent late (i.e., dust dirt) quantities at McGuire were estimated using NEI 02-01 ai combination with the NEI 04-07 Guidance R Evaluation (SE). For McGuire, te Unit 1 laten both containments, since after iis- n/broad containment was sufficiently similu containment.
to represent the Unit 2 Unit 1the GR/SE) to quantify the susr ceperea* a susceptible to debris buildup Also in accordanceih\ , containment was segregated into four areas based on the presence of robus s and representative surfaces: " Lower Containment inside the Crane Wall" Lower Containment Pipe Chase (outside the Crane Wall)" Upper Containment" Ice Condenser Surface types within each of these areas were categorized as (a) Horizontal Floor Surfaces, (b) Horizontal Miscellaneous Surfaces, or (c) Vertical Surfaces.
With the Attachment 2 Page 16 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response sampling surfaces defined, specific areas were chosen in order to obtain their representative online condition.
Sampling media included Masolin cloth and sticky foam. All sampling media was pre-bagged and labeled. Each bag contained a single Masolin cloth and a single sheet of sticky foam (sized approximately 9" x 12"). Each bag was then pre-weighed on a Mettler Toledo PR500D scale, with a tolerance of +/- 0.04 grams.The area was wiped down with a Masolin cloth to pick up fine debris and to consolidate larger particulate debris. Vertical surfaces were wiped fr the bottom up to prevent a loss of debris. The sticky foam sheet was then used t up any remaining particulate debris. Both the Masolin cloth and the t am sheet were carefully folded to prevent loss of debris material and plac the sample bag. Excess air within the bag was gently forced out to allow th. s ple t easily transported and post-weighed.
After all sampling was comp le bags we ighed on the same scale. The difference between the pre- a st-weights were used to calculate the mass of the debris collected.
Sample mass measurements were increase n o -o account number of possible sampling and measure errors incd of sample me to the sample surface, air movement a scale du r easurement, and the tolerance of the scale. The error due to air pe nd the 'tlrnce of the scale could also act in a conservative direction for so sa e e inc e the measured sample mass), therefore the q plied to be sa le uher eases the conservatism leading to a higher s ent deb o There were 40 indi I laten bris sat i taken, wi the following itemization by surface type: Horizont ace Horizoi/ cellae Su s: sp Vertica' 3aces: 11 sa s Once the s , e debris n for ea urface type was quantified via scale measuremen e specific dris density for each sampled area was computed by dividing the vdi sample nasses by their respective sampled surface areas. The sample densities he Qped into the following sample sets based on common surface type and locat mon associated work activities in each area, and cleanup procedures (e.g., simila rk activities and cleanup would be expected for the floors in Lower Containment inside the Crane Wall and in the Pipe Chase):° Horizontal floor surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase-Horizontal miscellaneous surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase Attachment 2 Page 17 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response" Vertical surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase" Horizontal floor surfaces in Upper Containment and inside the Ice Condenser" Horizontal miscellaneous surfaces in Upper Containment and inside the Ice Condenser" Vertical surfaces in Upper Containment and inside the Ice Condenser A statistical analysis was then performed using the groupe sample densities to provide a conservative assessment of debris buildup over a givi srface type. While the GR/SE states that the average of at least three same .for each surface type should be applied to the entire surface, the McGuire analysi j tep further by determining the 95% confidence interval of the mean (aver-ebris sen ity. This approach provides margin in the calculation of total late bris insid containments.
The specific latent debris densities were then p ied by the appl te actual 'urface areas inside the McGuire containment, w were based onrefe e drawings and information developed from the walkdowns.
total ate debris calculated conservatively assume that 100% f the estim s areas in co ment are susceptible to debris Using this methodology, the extral containment was determined to be reported in the McGuir 004-0 2, item 3(d)2. I'-- L4 ss total for the McGuire Unit 1 ich bounds Unit 2 as ated 2/28/08, Enclosure d or the total latent debris 30 lb (15%) of which is For conservatism, quantity in both Mc considered -be Attachment 2 Page 18 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M10. Please provide the details of the methodology used for the tag and label refinement evaluation.
Provide details of the equipment qualifications and engineering judgments used as basis for reduction of tag and label quantities which are assumed to fail and reach the sump.McGuire Response: Please reference the response regarding the McGuire tag and Attachment 2 Page 19 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M1 1. Please provide the technical basis for the assumption of 10-percent erosion of fibrous debris in the containment pool. If testing was performed, please demonstrate the similarity of the flow conditions, chemical conditions, and fiberglass material present in the test or tests versus the conditions expected in the McGuire containment pool.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of vendor erosion testing issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.The quantity of constituent fiber fines transpor tgo t Guire stra ue to the erosion of submerged but non-tr rted piece inbsulation was "ýtermined by vendor testing in order to refine t rvaive er assumptions documented in the GL 2004-02 NRC guidance repqrt .-07 an companion SER).The objective of the ero testing w to u nt y co ment pool flow-induced erosion/deterioratio t m occur on densi, las r1LDFG) insulation.
This was accomplished ubjecn a meas sa the LDFG material type to a room-temperature water in a doedrtical tes oop (VTL) apparatus and a horizontal test flume or d] ons of u 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and quantifying any mass that may have. f of dgefm the test sample. These two apparat the different turbulence and energy effectrt u n the ins,, as as vg the effects of the orientation of the sample respect to ate .To quantify the fibrous mass loss, the dry weight of the tes rpies before afte test was measured and recorded.
The testing was conduc n both larg d sma pieces of low-density fiberglass to observe any effects that siz urface a had on the sample's erosion.Subsequently, a 3\ erop! test in the horizontal flume apparatus was also conducted.
Analysis O t data provided further insight into the nature of the LDFG sample composi nd consequently, its erosion characteristics.
Primarily, it was observed that during 30 days of flow impingement, the sample did not continuously disintegrate.
The fiber insulation samples appeared to yield loosely bound fiber fines early in the test, after which the erosion effects subsided.Test Inputs Erosion tests were conducted with LDFG insulation samples in conditions intended to mimic, or be conservative with respect to, the expected post-LOCA plant conditions.
Attachment 2 Page 20 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Debris Type and Size: Both Nukon and TPI Thermal-Wrap fiber insulation exist in the break zones of influence in the McGuire containments, and for the purposes of LDFG erosion evaluation can be considered equivalent consistent with the LDFG destruction pressures discussed in the GR/SE, Section 11.3.1.1.
The erosion testing used Nukonsamples with the same bulk density (2.4 Ib/ft 3) as LDFG insulation used in the plant.The Nukon fiber insulation sheets were cut into 6"x3"x 1" rectangles to represent the large pieces, and into 1"xl"x1" squares to represent the small pieces (an extra large piece measuring 6"x6"x 1" was also included).
Samples then boiled in tap water for ten minutes to remove the binder, in order to sime conditions the fiber insulation would undergo during the blowdown an sump pool recirculation phases of the predicted post-LOCA response.Test Environment:
The insulation samples were subjected to w erosion enviro in both a vertical test loop apparatus and a horizontal flume test c isted of Sucting insulation samples to flow erosion by filling e VTL or TF ter and thei iulating the water to bound the flow conditio occur in th"l S sump pool. The samples were always completely submerg Ttesting t u're conservative erosion. The large Nukon samples were fasten n to im unnecessary movement.The small Nukon sam were sta ed b g the a wire cage in the flowj ither.f a, wiferelease of terfode stream, such that th. wa nt i eraoo material.Flow Velocity The eroa vrepye .y that is equal to the incipient tumblin city spe size. I-, G samples tested, the flow velocities were rmined to b 7 fet secona-f 1 the large pieces and 0.12 feet per second de small pie incipient tumbling velocity is the velocity at which the debris d start mov this vedity bounds the greatest velocity that a piece of insulation lyingtin the contai nt poo would experience without being transported to the ECCS sump ainer. Th fore, it is considered the velocity that would produce the most fiber fines fro bhme but not transported, fiber insulation pieces.Water Temperature an, mistry Temperature
-As discussed previously, the LDFG samples were boiled prior to being subjected to flow impingement testing, to simulate the conditions present during blowdown and recirculation.
The actual tests were conducted in the VTL and the TF in room-temperature tap water (i.e., approximately 60°F-80°F).
The temperature of the water increased during testing due to continuous pump heating (up to 1 10°F for longer-duration tests). It was determined that viscosity effects on the erosion rates were insignificant.
Attachment 2 Page 21 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Chemistry
-The erosion tests were conducted in tap water and not the buffered or borated water predicted to be present in the containment sump pool post-LOCA. The use of tap water is considered appropriate because the lack of chemicals such as soluble aluminum, boron, or pH buffers will not affect the amount of fibers that would erode from a Nukon LDFG insulation sample.Analysis of Erosion Test Data During the erosion tests, the small Nukon samples generally eroded more than the large samples, despite the large samples undergoing a r flow velocity.
Small samples eroded more fibrous mass due to their increa surface area and their being prepared for the tests by shredding, which produco fines available for transport.
Since the small samples eroded more than the la am ,hes the test data analysis utilized only the small sample results in order nerate csrvative fiber erosion quantities (therefore these higher small pi osion rates ar lied to both size ranges of submerged, non-transported fi Msulation).
The small fiber test samples lost, on average , ut 30/, 7% of i eight for any given test duration, with the raw a range sp o 20%. Bec' the LDFG erosion data was not consistent, a was an and results compared using several different approaches:
- 1. Assume the 30-day small saf e est res epresented the most accurate erosiný-t_
- 2. Determine t1ý ivght losper ho\U Mt ý111: al lsample erosion tests, and then extrap that we ,t loss p r value to 0 days to properly account for the ECCS tbso i~3. Av e esio values regardless of test duration ingrcror,-
assu ue applies for the ECCS mission time.Applic of approac an ove yie ed 30-day fiber erosion estimates that appeared oe non-conser tive c01ompared to the majority of the small sample data points.o\1 Since the fiber on test r Its showed wide scatter across all test durations, the assumption was that f erosion is not directly time-dependent, and therefore could be conservativ ,ed by averaging all of the small sample erosion test results to reach an ove osion value (i.e., approach 3 above). Additionally, as noted previously it was observe that during the 30-day erosion test the samples did not continuously disintegrate.
The fiber insulation samples appeared to yield (transport) loosely bound fiber fines early in the test, after which the erosion effects subsided.
As such, the overall erosion value calculated from the small sample average is considered applicable to a 30-day mission time.The calculated average of the small sample fibrous erosion test results was approximately 6% of initial weight, with an error of +/-4% as determined by RMS error Attachment 2 Page 22 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response analysis versus the calculated average. Approach 3, then, determined a conservative estimate of the overall fiber erosion value to be 10% of the initial fiber weight (6% + 4%).Therefore, the attrition/erosion mechanism that strips away the loose pieces of LDFG via water impingement is conservatively estimated to reduce an initial weight of submerged, non-transported fiber insulation by 10% over the 30-day ECCS mission time.Subsequent to the preceding series of vendor erosion tests, extensive discussions were held between NRC technical staff, the vendor, and Licensees (including the Duke plants) regarding the testing configuration and analysi sfet odology. This led NRC to request a confirmatory LDFG erosion test from the to address issues identified with the test flume and erosion sample configura
.l 0-day confirmatory erosion test series was completed by the vendor in 20 confi a 30-day erosion value of 10% for both large and small piece LDFG, Attachment 2 Page 23 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M12. Please provide the results of the array testing conducted at the Alion Science and Technology Corporation and the Integrated Prototype Test (IPT) testing conducted at Wyle Laboratories.
For the IPT testing, in addition to head loss values, please provide the results as a function of time. Provide a thorough description of the methodology used to combine the two test results to determine the final head loss for the strainer debris bed. If a correlation was developed to determine head loss, provide the correlation along with the assumptions and bases used in the development of the correlation.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.NRC technical staff concs with th ray d the rated Prototype Test (debris preparation, 4 i duction, bris ftio low fields, bare strainer area and chemica ') liscuss pub eting convened on 11/24/08 in Washington, D. .as de mined th olution of e issues raised by the staff at that meeting would ire f r testing evaluation.
In June
ýs tes;nres was performed at Wyle bo y n-g wly des i e and conventional debris (i.e., boun iber and pai ate ) to addis the debris preparation, debris introduc ý ebris aggl atoio ling and flow field issues identified by the staff regarding teegrated P type -i.t(IPT).
NRC staff guidance from March of 2008 as well as inp om discus with the staff were used for the test protocol development fo Confirm ory Head Loss test series. Upon completion of the June 2009 testing, the p ol, s bed formation and head loss results were discussed with the staff in July g with photos and videos obtained during the testing.From these conversatio e staff concluded that the methods and protocol being utilized for conventional debris preparation and introduction during the tests met NRC expectations.
In parallel with the Confirmatory Head Loss test series in June 2009, Duke provided a draft white paper to the technical staff entitled "Duke Energy Chemical Effects Testing in Support of GSI-191".
This document served as a guide for discussions with the staff on the battery of testing performed by Duke to date in the area of chemical effects, and to address NRC concerns with the effect of potential chemical precipitates.
As a result of Attachment 2 Page 24 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response these discussions, Duke elected to continue head loss testing in the fall of 2009, using the Confirmatory Head Loss test series protocol and test flume along with pre-mixed chemical precipitates in accordance with staff guidance.
The fall 2009 testing was identified as the Chemical Precipitates Head Loss test series, which continued into the spring of 2010. The results of this pre-mixed precipitate testing were discussed with the NRC staff in April and May of 2010.Subsequent to the Chemical Precipitates Head Loss test series, Duke determined that a one-time, long-term test approach utilizing soluble alumin Lm injection (as opposed to introducing pre-mixed precipitates) would be more reprenative of the post-LOCA environments predicted in Ice Condenser containmeI entified as the Confirmatory Integrated Test (CIT), the development of the test followed discussions in June 2010 with the NRC staff. The CIT utilizes previo sti erience with conventional debris preparation, introduction, and transpo^t facilitate th rmation of a uniform debris bed on a prototype strainer array. Ainally, the CIl will be thermally and chemically representative (pH, boron, cal cu, silicon, and alum per guidance set forth in WCAP-16530 utilizing a Duke-modi lgorith as descri reviously in the McGuire Generic Letter 2004-02 Supplement sp , dated 2/2 The 30-day CIT, scheduled to b (ýe~rformed in fal -,is designed to p ovide a conservative and specific head loc the McG- strainer array as a function of the ECCS mission time, with minim sos ssing.Attachment 2 Page 25 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M13. The conditions under which vortex testing was conducted for McGuire, and the plant conditions for which the testing was being conducted, are not clear from the available documentation.
Based on the information provided to date, the NRC staff has been unable to determine what conditions resulted in vortex formation and whether the modifications made to eliminate vortices were tested under conditions that conservatively represented those expected in the plant post-LOCA. Vortex testing was conducted at 3-inch submergence (as stated in the Duke Energy Carolinas (Duke) response to RAI question 39 in Enclosure 1 to the supplemental response dated February 28, 2008), ch is greater than the expected 2-inch minimum submergence for a sall beak loss-of-coolant accident (SBLOCA) (as stated in Section 3(f)(2) of En 2 of the supplemental response).
Note that Duke further states in* se to RAI question 39 in Enclosure 1 that the minimum submerg ,6efor the \ er is expected to be "at least" 2 inches and separately that it i s ut" 4 inche lclosure 1, pages 35-36). Enclosure 2, Section 3(f)(2), sjtes that the stra *s submerged by at least 2 inches while Enclosure 2, Se 3(f)(3), states tha rating is submerged by at least 2 inches. Enclo 2, Se 3(f)(3), \ tates that the testing was performed wit "few inches rgence. This, f disparate strainer submergence val not prov coherent description of the test conditions.
io Enclosure 2, Sect 3(f)(3), st tha sting conducted at velocities between 0.01 fresp'0.09 ft/s wm pproach velocity for the strainer i spo not e a basis for the 0.052 ft/sec, other \he exped maxi approach elocity is greater than nominal by abo 'facto (Enclo 1, pages 35-36), and does not clearly stat-ha-etn Mo aoeQ.ý2ft/s(-ýcdd not result in vortices.se provi maat des the conditions expected in the plant and ths resent dun~~etn, incuding the following information-
- a. clarify wh- e act minimum submergence for the strainer is exp to be in tplant.b. If differ\ valuat in for vortexing were conducted for SBLOCAs and large break loss- accidents (LBLOCAs), please provide details for each evaluation.
- c. Please provide the basis for the maximum approach velocity.d. Please provide a quantitative value for the approach velocity during which vortices were observed to form when no vortex suppressors were installed.
- e. Please provide a quantitative value for the submergence level at which the testing was conducted with no vortex suppressors installed.
If the level changed (e.g., between SBLOCA and LBLOCA tests), please provide the test conditions for each test.Attachment 2 Page 26 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response f. Please provide information for testing that was conducted with the vortex suppression grating in place, including the minimum submergence and maximum approach velocities that were present when vortices did not occur.g. Provide a quantitative value for the vortex suppressor submergence in the reactor plant. If some suppressors are installed at different elevations than others, provide the submergence level for each location.McGuire Response: (a) through (g).Attachment 2 Page 27 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Vortex Suppression Grating (11/2" X 5/64")Vortex Test Pool Level --7_ -3"'77 11%" gap (nominal)Figure 13S-1: Top Hat Strainer Module Submergence Vortex Test Condition-4" 11/2" gap (+11/4", -0")Figure 13S-2: Top Hat Strainer Module Submergence Plant Condition a. Actual minimum submergence level of the McGuire ECCS Sump Strainer Top Hat modules (pool surface to module perforated plate topmost surface) is >_ 4 inches, as Attachment 2 Page 28 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response depicted in Figure 13S-2. The minimum submergence level occurs as a result of the SBLOCA scenario.b. Vortex evaluations were performed for the McGuire ECCS Sump Strainer design for the limiting submergence level (SBLOCA scenario) only. The testing was performed at a submergence level of three inches, which is less than the minimum predicted SBLOCA submergence level of > 4 inches, and therefore conservative.
The LBLOCA scenarios generate more pool volume and a higher submergence level.Details regarding the SBLOCA vortex evaluation are located in the McGuire GL 2004-02 Supplemental Response dated 2/28/08, Enc re 2, item 3(f)3.c. For vortex testing purposes, the as-built maximu otoach velocity for the top hats closest to the ECCS suction lines (assuming f aln'oeration) was determined to be 0.051 feet per second in McGuire Unit 2,, Wihc boup Unit 1. This approach velocity does not use the normalized flow- ibution app h. Instead, the flow is distributed among the top hat module that the interna es within the strainer top hat assemblies and plenu e pressure balance This results in a non-uniform flow distribution, which is us deter e the appla velocities.
With an initially clean ECCS sup strainer roach veloc for the top-hat modules closest to the p tion line e epcted to be higher than the predicted four-train McGuire n roach v ty (i.e., about 0.028 feet per second) by approximately a fact uf t d. Vortex suppression was p med in arts; er portion used walls to model the adjac or basea te s could artificially still vortex formation.
Wat vel wa t to 3 in ove th most surface of the perforated plate straie (reference Figure 13S-1). Initially, testing was performed starting a veloci 0.01 feet per second and then increasg u 0in i ments of 0.01 feet per second.Subently, onal ig wa ird starting at an approach velocity .of 0. et per seco d th icreasing u to 0.09 feet per second, with the same incre ts. At appr velo s at and above 0.04 feet per second, an air-entrain ortex was p nt. the vortex formed, the vortex suppressor grating w italled in t esignated location (i.e., leaving a 11/2-inch gap between topmost sur f the st einr module and the bottom of the vortex suppression grating).
Effica the ressor was demonstrated by elimination of the vortex at all approach veloc , 0.09 feet per second, with only minor surface dimpling remaining.
- e. Referencing Figure 13S-1, the submergence level of the strainer top hat module topmost surface (perforated plate) during the vortex suppression testing was fixed at 3 inches. The vortex testing did not vary the submergence for LBLOCA/SBLOCA; the submergence level modeled (3 inches) was a conservative approximation of the minimum expected submergence level (4 inches) following a SBLOCA at McGuire.f. During the vortex suppression testing (without the vortex suppressor in place), an air-entraining vortex was present at the base of the topmost surface of the strainer Attachment 2 Page 29 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response module at and above approach velocities of 0.04 feet per second. At approach velocities less than that, only surface depressions existed (no fully-formed vortices were present that could entrain air). As stated previously, the submergence level of the topmost surface of the top hat strainer module during these tests was 3 inches.Once the air-entraining vortex formed at each of the higher test increments, the vortex suppression grating was placed in the designated position, and the vortex successfully eliminated.
In this testing sequence, the top of the vortex suppression grating (when installed) was even with the top surface of the test pool, as shown in Figure 13S-1.g. Referencing Figure 13S-2, the submergence leve o te vortex suppression gratings in the McGuire plant condition is at least 2 inche locations (inside and outside the Crane Wall, as measured from the top of tie rigrat )uring a SBLOCA event.The LBLOCA event produces more pool vle and th n re more submergence.
Attachment 2 Page 30 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M14. Please provide a response to the question from the revised NRC Content Guide sent by letter dated November 21, 2007, relating to Enclosure 2 of the supplemental response dated February 28,2008, Section 3(0(5), regarding the ability of the strainer to accommodate the maximum potential debris volume. This response should apply specifically to the McGuire strainer and not be a general answer (as is found in Enclosure 2, Section 3(f)(5)).
The McGuire response to Enclosure 1, RAI question 40, sends the reader to Enclosure 2, Sections 3(f) and 3(o) to find this information.
The information is contained in neither location.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as"acceptable" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.Guidancet sue by t rtae eentofh.ib s a200 d the intent of the Content Guide question regarding the ability Staner to accommodate the maximum potentiae nd i enRecume.As identified in theG 004-' 02uLpplerfi~ta Repnsýte~d 4/30/2008, McGuire predicted a total e of densithinlass (LDF) insulation (fines and small piees toa SUes uii after a limiting break. Subsequent piees tha bunitaý-cGuecetrie flalre[ ],a'ement of a large portion of the fiber insulation ie oI~l 6,te-;S oops, Steam Generators, and RC Pump~)wsnecess-ay
ýýc iitd the F }Ibensulation Replacement Project (FIRP).This prbject. schedulep, bc within two Unit refueling outages, results in a bounding trported LD sulai olume of approximately 72 ft 3 assuming four-train ECCS culation flo , -_CCS Water Management flowrate will transport less fiber insulation a imnificant reuction in fibrous debris at the ECCS strainers.
The non-LDFG de uani) e expected to transport to the strainer (and their characteristics) weree in the McGuire GL 2004-02 Supplemental Response dated 2/28/2008, Enclo 2, items 3(c)2, 3(d)3, and 3(h)6. Additional margin was also added to the unqualified epoxy coatings debris quantity for chemical effects testing in 2009-2010 as discussed in the response to RAI 26. The appropriate values are provided in Table 14S-1 below for convenience, along with the equivalent volume conversions for each debris type.Attachment 2 Page 31 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for AdditionalI nformation Draft Response Table 14S-1 Non-LDFG Debris Quantities Transported to McGuire ECCS Sump Strainer*:,,D bri Ty e De ... : " Equ~ivalent Debris Type6,,, Debris Quantity at E v ebrisiType SDebris Volume Density Strainer at,-Strainer Qualified Epoxy 118 Ib/ft3 167.6 Ib 1.42 ft Coatings (5D ZOI)Unqualified Epoxy Coatingsc an Infi stitia4co7tri Unqualified Alkyd ( ott Coatings in bkse9 s of th Latent Dirt/Dust 169 e b/ft3 a LD\ G1.01 ft3 Latent Fiber (lint) 2.4 b/ft3 1 E 30 Ib 12.5 ft 3 Miscellaneous Latent Debris** NA a a (tags, labels, etc.) Note: Destroyed stainless steel RMI vr a int the iipviuao the ECCS Sump Strainer."*ltn ersiNote:
Miscellaneous laetdbrsiceun-"fz the flo area for maximizing the approach velocity, but has an insignificant of te top hats reltiveto rns Thus, the total v o f debri dFwl axl othe to transport t to the ECCS sump strainer afte iung bre is apl.owe te r, i fncethe sum of the debris volumes in the aboved ab in ot s bounfci n at nsported LDFG insulation volume total).The total in hi oklc retinitsg Mple she ECCS strainer (Unit 1) is 346 cth During path otde a sre cylisber) nally accumulate non-uniformly on the strainle arahe apprc varyacr s the individual strainer top hats and across~t ý,rray basedx\%,e of the top hats relative to the recirculation suction piping. ,the debrisbd axially from the top hat base plate out to the free end, upl~ h maximum load~. However, since the total transported debris volume (as dejhtrated aov)is not sufficient to completely fill the strainer interstitial volume, the strain \rsrface w retain its complex shape (multiple top hat cylinders with flow paths outside e cylinder) and flow area. With no decrease in top hat module flow area, te fdapproach velocities remain bounding.Attachment 2 Page 32 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M1 5. Please provide information that verifies that the debris preparation and introduction methods used during the array test and IPT were prototypical or conservative with respect to the transport evaluation for the plant. In general, protocols for fibrous debris preparation result in debris that is coarser than predicted by the plant-specific transport calculation.
In addition, the NRC staff has noted that debris introduction frequently results in agglomeration of debris such that it may not transport to the strainer prototypically or create a prototypical debris bed. Both of these issues can result in non-conservative head loss values during testing.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Proto T IPT) an rray ne the question are being replaced by the C atory Igrated .response to RAI question 12 of this submittal detail iscussi arding the associated Duke/NRC interfaces and the cor ndi evelopmr of the CIT.The 30-eadul:
d in 2010, is designed in accordance with the Mar/ 068stthe reC'l' if Review Gu i Loss garding Generic Letter 2004-02 Closu the Area o iner d Loss ad ortexing" and further conventional debris t exlerienc ebris aration and introduction will be consistent with a thin-bed p\ ol as descr in th iff guidance.
Conventional debris transport will be optimize erforming stem verification test prior to formal testing. Debris settling in the te.t k will bý inimized by using mechanical agitators, which in the previous " -ti re convent esti une/November 2009) have proven to be effective in this regard.Attachment 2 Page 33 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M16. Please provide information on the flowfields in the array test. The NRC staff is concerned that non-prototypical debris distribution may have occurred during testing caused by stirring of the tank. The stirring can result in the transport of debris that would otherwise not transport, or result in washing debris from the strainer screen surfaces.
Either of these phenomena can result in reduced (non-conservative) head loss values during testing.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.Wý__The Integrated Prototype Test (IPTýreplaced by the Confirmatory Integr 12 of this submittal for tailed di,, interfaces and the cog d(nced in the question are being e response to RAI question ociated Duke/NRC The 30-day CIT, sc the March 2008 "NI Closure in the Area of describe-r bed prutoclLs optim ' by performi in the te k will be testing (Ju Novembi strainer array e p positioned betw h erforne ill 201, i esigned in accordance with Guidarn Regarding Generic Letter 2004-02 oss a Vortexing" and the testing experience an lduction will be consistent with a thin-,staff gLiat .Conventional debris transport will be eprificatio 1est prior to formal testing. Debris settling d b ing mechanical agitators, which in previous nave pven to be effective in this regard. The test:rom non-prototypical flow fields by a baffle plate t tank volume and the array.Attachment 2 Page 34 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M17. Please provide debris preparation and introduction information similar to that requested in this enclosure, RAI question 15, for the testing that was used to justify that a thin bed would not form on a top hat strainer.
Note that for thin bed testing, the NRC staff considers it prototypical or conservative for fine fiber to arrive at the strainer prior to less transportable debris. Overly coarse debris preparation or non-prototypical introduction to the flume may non-conservatively affect the potential for thin bed formation.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (I replaced by the Confirmatory Int 12 of this submittal fo led interfaces and the The 30-day CIT, the March 2008 "NRC Closure in o bed pi ol as des optimize d'b performi in the tes ý will be testing (Jun, lvem strainer array positioned betwe PgT) Test o Ily documented are being egr Te .See response to RAI question disc *on re the I ciated Duke/NRC erforme iall 2010,1 designed in accordance with Guidan egarding Generic Letter 2004-02 s an ortexing" and the testing experience
ýparati a ii bduction will be consistent with a thin-aff gui .Conventional debris transport will be-rification test prior to formal testing. Debris settling b g mechanical agitators, which in previous av yen to be effective in this regard. The test tom non-prototypical flow fields by a baffle plate ,t tank volume and the array.Attachment 2 Page 35 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M18. Please provide the criteria used to judge that differential.
pressure-induced effects (e.g., boreholes) did not occur during testing.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IPT) and ,r Test originally doc nted are being replaced by the Confirmatory Integrated Te IT). Seehe respo o RAI question 12 of this submittal for a detailed discussion r ding ssociated e/NRC interfaces and the corresponding opment o The 30-day CIT, scheduled to be pin fall 2 is designed to assess the post-test condition of the debris bed via an eries perature and flow sweeps.By varying the temperaturi of the tes k po onstan epwrate first, less potential exists for debris bed evelop t \ a ss sity correlation data is collected.
Subseitol tihno peratu p t,.f eps will be performed to provide additional c ation d After te st tank is 'ained, any noted differential pressure related anomlies in t ebris be i be addressed in accordance with the March 20S, Staff ýe Ris eR ing Generic Letter 2004-02 Closure in the Are H ad s and Attachment 2 Page 36 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M19. Please provide the scaling parameters used for calculation of debris quantities and strainer approach velocities used during testing. State whether the scaling accounted for strainer areas blocked by miscellaneous debris such as labels and tape.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the sctl1dtuled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (I1 replaced by the Confirmatory Intel 12 of this submittal for a detailed interfaces and the corresponding The 30-day CIT, sch e tbe the March 2008 "1 aff Closure in the Are trainer described previously.
The test 30- I Confirnit elbdLs tes~t, 1 serie cribed pre top-hat les, a redut fro tests. The e, the scalI ara velocities wil Lased on t no appropriately fo area coverc CIT, the predictediebris bed ic ially document t re being the response to RAI question associated Duke/NRC is degned in accordance with' eneric Letter 2004-02 nd the testing experience i odi version of the test tank used for the s and ical Precipitates Head Loss test iodificatfibi ill result in a test array containing fewer ,2 x 3 test array (six top-hat modules) in the earlier meters for the CIT debris loads and approach dified test array gross surface area, adjusted by tags and labels. For scaling debris loads for the I fraction will not be included.Attachment 2 Page 37 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M20. Please provide information on whether the amount of coatings surrogate was adjusted for the volume difference created by the difference in density between the surrogate material and the potential debris in the plant.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IPT) and Arry Sil replaced by the Confirmatory Integated Test (12 of this submittal for a detaile ion rega interfaces and the corresponding e o of tt The 30-day CIT, scheduled to be pe ime fll the March 2008 "NRC eview Gi ne Closure in the Areabf Stra in -lead Lo s n described previou Regarding the use of ' ed cTIantgs partiete s parameter ctngtei mix. Tp pqtý-loenateia o(cupiesl'c Ia 'ij results creasin s tn. lown the su te material in t T (800 grit sili to match t olume of th eicte iled coatina are being!AI question Y/NRC 2010. signed in accordance with ig nric Letter 2004-02 exn"an the testing experience urrogate in the CIT, the critical e volume of the material in the debris Flume in the fibrous debris space that re higher head loss. The volume of ,a carbide) will be scaled as required s particulate.
Attachment 2 Page 38 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M21. Please discuss the NRC staff's observation that in the IPT the flow was non-prototypically directed at the top hat strainer in a direction parallel to the strainer long axis. Please address whether this non-prototypical flow direction could result in a non-prototypical formation of debris on the top hat strainer.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Dupe public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IF replaced by the Confirmatory lntý12 of this submittal for a detailed'interfaces and the corresponding The 30-day CIT, sched to be the March 2008 "N Stavf.Revic Closure in the Are train described previous ebris p bed protocol as descr in th optimized ming in the te np testing ne/Nove 00 straine ray will be pro ed f tween the mnest-ced in the tion are being-the response t ýAl question ie associated Duke/NRC igned in accordance with;'ric Letter 2004-02 I Los a rtein d the testing experience nation witroduction will be consistent with a thin-f guidafl e Conventional debris transport will be ui ation", t prior to formal testing. Debris settling u-s ic al agitators, which in previous proven e effective in this regard. The testýon-prototypical flow fields by a baffle plate iolume and the array.Attachment 2 Page 39 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M22. Please provide the clean strainer head loss for McGuire Unit 2 (only the clean strainer head loss for McGuire Unit 1 was provided).
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.For completeness, the clean strainer-head provided: The McGuire Unit 1 clean strainer head Io configuration at 60'F, is calculated as 5.71 flow condition (four trains), and 3.79 feet o operation.
The McGuire Unit 2 clean straine configuration at 60'F, is calculateda§5.
flow condition (four trains), and 3.91 et o operation.
It should be noted cG as sub ECCS Water Mana ent (dat May 28, the Commission.
In re n to CS reci the Conta Spray decrea h the phas erefore, stat n strai by the v s stated ab ire Units 1 and 2 are ier area and recirculation
,-train CS he installed strainer area and fr the maximum recirculation
~lettrain RHR/two-train CS ,Licen endment Request for 2 ), which I currently being reviewed by rculion flow, this license amendment allows wit ly one pump. This configuration willýs s strainer during the recirculation head loss will be conservatively bounded Attachment 2 Page 40 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M23. The supplemental response stated that the total head loss across the McGuire Emergency Core Cooling System Sump strainer (clean strainer head loss plus debris bed head loss) was conservatively predicted to be 9.8 ft at switchover to sump recirculation.
No explanation was provided as to how this value was derived. It appears that the licensee is taking credit for time-dependency in head loss, since the 30-day value is 15.7 ft. Please provide the time-dependent results and calculation methodology for determining net positive suction head margin throughout the 30-day mission time.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Proto replaced by the Coni 12 of this submittal interfaces and the The 30-day CIT, sch, conservaieaaýr the E issio will th ýýe converte ensure ' quirem(ocumented are being sponse to RAI questionýociated Duke/NRC the Cl.led ) ll 2010, is designed to provide a shs Guire strainer array as a function of o)csing. The strainer head loss results I at te ECCS sump recirculation pump inlets to-e met throughout the ECCS mission time.Attachment 2 Page 41 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M24. Please provide the types and amounts of debris added to each test (Array and IPT). Include information on introduction sequence.
Please provide relevant test parameters such as temperature, debris introduction times, and flow rate for the Array and IPT tests.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IP and Array ced in the q ion are being replaced by the Confirmatory nt Test (CI e the response to RAl question 12 of this submittal for a detailed s regardig associated Duke/NRC interfaces and the corresponding d op f the The 30-day CIT, schea -be perf ed i[ f 0, is *gned in accordance with the March 2008 'NR S ' ew Gui' ce Gerr ic Letter 2004-02 Closure in the Are Straine ad Los, rtexi1d the testing experience described previously Debris se for t the -bed" protocol from the staff guidance, using b *ties trd I o particulates first, followed by the fiber f n severa me ddition The chl I additions ke" after CIT pool heat-up and conventional debris bed forma as boric ac dium r borate and soluble aluminum, calcium, and silicon in acc ce with P-16530 guidance as modified by the Duke algorithm.
Bounding chemra uantitie WIll be used.The CIT pool temp are pý will follow the bounding post-accident plant sump pool profile, from a maximL"` 0-F to a 30-day ECCS mission time minimum of 90'F.CIT flowrate will be based on previous testing experience facilitating the formation of a bounding conventional debris bed. Flow/temperature sweeps will be performed at the end of the 30-day testing interval to provide head loss correlations for extrapolation to other flow conditions.
Attachment 2 Page 42 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M25. Please provide information on the amounts of debris that settled during testing for each test (IPT, Array, and Thin Bed). Note that Enclosure 1, response to RAI question 37, states that near-field settling was not credited during testing.However, the NRC staff observed significant settling during the IPT. Please provide a quantitative evaluation of how this settling affected -head losses for each test. Please state whether this settling is prototypical of plant conditions and provide a basis for the conclusion.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (11 replaced by the Confirm r Int, 12 of this submittal f d interfaces and the span n The 30-day CIT, sclhie the March 2008 "NRC Closure of describ evio mech l agitators, be effec *n this reQa est re ed in the question are being ( See tesponse to RAI question re e aissbciated Duke/NRC fall 2010, is designed in accordance with eRegarding Generic Letter 2004-02 rtexing" and the testing experience 9t nk will be minimized by using ( June/November 2009) have proven to Attachment 2 Page 43 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M26. Please provide verification that the unqualified epoxy coatings at McGuire are similar to the coatings used in the Electric Power Research Institute's analysis of original equipment manufacturer coatings.
Also, are plant records maintained for the unqualified coatings in order to track quantities and composition?
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.The unqualified coating surface area exp ,su (o post-accidentý" McGuire containments was inspected, qu d, and documen 02-01 walkdown efforts in support of GSI-191 As noted in the McGuire GL 200 Suppleme e onse d 2, item 3(h)(5), post-accident un oxy co debris based on testing and analysis of 0 c s perfo and In order to take credit for this refiner , th alifie y containments at McG e judge e Si r the e&y EPRI testing.itions inside the te artof the NEI ated 2/ 8, Enclosure quantities were refined Jocumented by EPRI.coatings inside the coatings used in the The EPRI program'c from industry, gathei and creatioA t in the ujire conk Themaj of unqL applied to Mcor described in teEIof unTqualified coatings directly onents and equipment/structural parts gs sampled included OEM and non-re representative of the type existing ings inside the McGuire containments are vendor-ent, applied via a controlled method as There are also s *c (limit situations wherein McGuire maintenance personnel apply remedial coati a rsuit of primary containment coating assessments, as detailed in the McGuir 04-02 Supplemental Response dated 2/28/08, Enclosure 1, RAI #25. These reme I coatings are manually applied and inspected in accordance with an approved specification and procedure, but remain unqualified coatings for the purposes of the EPRI comparison.
The epoxy-coated test coupons described by the EPRI test report and analysis are therefore appropriately representative of unqualified epoxy coatings in the McGuire containments.
Attachment 2 Page 44 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Plant records are maintained for all unqualified coatings in containment in a calculation file. This calculation is periodically reviewed and updated as needed to track any additions of equipment or components that do not have qualified coatings.
A cumulative total (in square feet) of the unqualified coatings is kept in this document, which in concert with the GSI-191 walkdown report, serves as a partial basis for the ECCS Strainer performance analysis coatings debris input. The unqualified coatings quantification documented in this calculation file does not reflect cumulative reductions in unqualified coating area (e.g., when an area in containment is stripped and re-coated with a qualified system), and also contains other conserv e assumptions regarding surface area and inclusion of components with unkno ting characteristics.
Also, for the chemical effects testing performed in 2009-201 dditional unqualified epoxy coatings debris margin was included.
Therefore, inherent margin in the unqualified coatings value used for GSI-191 e iuon.Attachment 2 Page 45 of 68 McGuire Nuclear Station Generic Letter 2004-02 .Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M27. Please clarify the discrepancy in quantitative values for unqualified epoxy coatings debris in Enclosure 2 to the supplemental response dated February 28, 2008, response to Section 3(e)(6), Tables 3E6-1 and 3H6-2.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time, with noted clarifications from the staff This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.In the McGuire GL 2004-02 Supplement sponse dated 2/2F8Table 3E6-1 ("Initial Debris Transport to McGuire ECCS Sump St"oers")
re esents th ial unqualified epoxy coatings quantity predicted to transpolt e FC Sump po00 r a LBLOCA event. This particulate quantity sed to initi the strainer forsign and installation.
The unqualified epo s quan6i his table contains no refinements.
In the McGuire GL 200 2 Supplen Rep dated 8/08, Table 3H6-2 ("McGuire Unqualifi Charal ei nstic' ets the refined unqualified epoxy coatings qu pre to tran e Sump pool after a LBLOCA event. This refine, tity w sed in t rformance evaluation of the strainer during the Integrated P type t for che aal effects and the Confirmatory Head Loss Test, une eQ m , described in the Table 3H6-2 notes, an ut)dsrb -d nfiýrsons'e .oserIeus~tion 26 of this submittal.
For th emical Pre tes ILoss Te series in fall 2009/spring 2010, and the Confirm lntegrated in fa\ 0, an additional quantity of unqualified coatings particulate', added for ign be the originally reported value in Table 3E6-1.This addition'
\iculate lo subsequently refined the same way as noted above, increased the to1 ass of un alified epoxy coatings debris to 392.2 Ibm (from 357.1 ibm).Attachment 2 Page 46 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M28. Please identify and describe the main features of any procedures that comprise containment cleanliness practices.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.As identified in the McGuire GL 2004-02 Suppl tal Re se dated 2/28/08, item 3(i)(1), McGuire has implemented programpat ontrols to e that potential sources of debris that may be introduced o ontainment will sessed for adverse effects on the ECCS and Containment Sp circulation functio he programmatic controls and practices relating to containmen clnlines clude: Containment Housekeeping/Ma Condition Extensive containment cleaning isd during e refueling outage using water spray, vacuuming and hand wiping c i ash d are performed as needed and visual inspections erformed the r n aa f containment.
Foreign material is removed a n ry. Up es e fori material control procedures requir er'ial, a untabili Jbe mined in Modes 1 through 4 for items carried int d out o ntainm6 hese con ls are implemented using administrative proced McGuire cEdqulrement (SR)McG e chnical S ica urveilla .equirement 3.5.2.8 requires that the E Snli? be visuall ec o verify there are no restrictions as a result of debris, and no e cýno~e of struct distr r abnormal corrosion present prior to declaring the EGGS operable isual iiection of containment is performed to ensure no loose materiall esent whi ould be transported to the Containment Sump and cause restriction
~ECC ump suction during accident conditions prior to the transition from Mod 4 operations.
When these inspections are performed, major outage work is e, and any remaining loose material in containment must be logged and tracked in ccordance with station procedures for control and accountability.
If any debris, damage or deficiency were to be discovered during the inspection, station processes require entry into the corrective action program, with the requisite investigation and implementation of appropriate corrective action prior to the transition from Mode 5 to Mode 4.Additionally, McGuire Selected Licensee Commitment 16.6.1 ensures that a visual inspection is performed to identify any loose debris inside containment and ensure it is Attachment 2 Page 47 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response removed prior to establishing containment integrity and following entries made after containment integrity is established.
Attachment 2 Page 48 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M29. Please provide the technical basis for the conclusion that all labels are capable of withstanding post-LOCA conditions in containment except inside the crane wall in lower containment.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time, with noted clarifications from the staff This draft response is consistent with the one discussed on that call.Please reference the response to RAI questni 8 of this submitta f~ldetails regarding the McGuire tag and label evaluation.
Attachment 2 Page 49 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M30. The revised "Content Guide for Generic Letter 2004-02 Supplemental Responses," sent by letter dated November 21, 2007, Section 3K, requests a summary of structural qualification design margins for the various components of the sump strainer structure assembly.
This summary should include interaction ratios and/or design margins for structural members, welds, concrete anchorages, and connection bolts as applicable.
Please provide this information.
McGuire Response: E The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.As described in the McGuire GL 2004-02 Sup en sponse da /28/08, the ECCS Sump Strainer is constru f robust m ld is positione oth inside and outside the Crane Wall (see S-1) %Attachment 2 Page 50 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Attachment 2 Page 51 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-1 below shows the design inputs for the McGuire ECCS sump strainer structural calculations, including the Top-hats, the strainer sections Inside the Crane Wall (ICW) and in the Pipechase, the ICW Strainer Enclosure, and the Pipechase Vortex Suppressor.
Unit 1 and Unit 2 design input values are identical.
Table 30S-1: Design Inputs/Loads for McGuire ECCS Sump Strainer Misc. Load (Cable Tray/Conduit)
Attachment 2 Page 52 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-2 through Table 30S-6 summarize worst-case interaction ratios (IR) and/or design margins for the various components of the McGuire Unit 1 ECCS Sump Strainer structural assembly including structural members, welds, concrete anchorages, and connection bolts.The Unit 1strainer is represented in these tables as five distinct sections:
the top-hat strainer modules, the portion of the strainer inside the Crane Wall (ICW), the portion of the strainer outside the Crane Wall (i.e., in the Pipechase), the ICW Enclosure, and the Pipechase Vortex Suppressor.
The McGuire Unit 2 ECCS Unit 1.Table 30S-2: McGuire Strainer structural qualification "summaries follow those of Unit I ECCS Sum ,P-Strdi er Top --,hat Module Structural Qualificatio.n Sýrnary 7ý_Component Description Measurement' Actual AIlowable Comments Bending Stress Q'ps- 40pi Top Hat Loading Axialtress N -gligible stressP 53ý,pi 4509 si Bendin 7O3 Sý 2 4086 psi Top Hat Buckling Axial L n6ps 311 s SCircumfer~n'ial
.8ps , 53 24 s 3/8" Diameter M ' ax IR '4 1.0 Top Cover Plate, Be'n'6 ýg Stress1,68 psi 16875 ps i Base Plate Mx-tes ',427 psi 16875 psio .Base metalbi 6 bi shear; Fillet weld r allowable
-928Ibs/in Attachment 2.Page 53 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-3: McGuire Unit 1 ECCS Sump Strainer ICW Structural Qualification Summary ICW Strainer Structural Frame Evaluation Component Description Measurement Actual Allowable Comments Strainer 3/8" Plate of Tube Module Max Stress 15302 psi 16875 psi Connector Plenum Frame Extension Member Stress Max IR 0.90 1.0 Strainer Module (End Module)Main Strainer Plate Element 3/8" Thick Max Stress 47i 75 psi End Flow Plenum Frame Main Strainer 3/8" Diameter Studs Max IR 0.82 1 .0 Plenum Structural Frame SExtension Anchor Bolts IV -I9I456 1.0 Strainer_ Structure~Extension Max 095 1.0 Strainer Module-. (End Module)3/8" Wing Plate } Using normal allowables (0.83 XdI .with faulted I~s allowables) e~r Box Evaluation Com pon ntDsr pt n,'esrmn Actual Allowable Comments 3/8" Coprýlt Nax Stre*ý' 14212 psi 16875 psi 3/8" EDia-m-tr`,1u "Max IR 0.78 1.0Full penetration Welds welds -acceptable.
Anchor Bolt Max IR 0.72 1.0 Attachment 2 Page 54 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-4: McGuire Unit 1 ECCS Sump Strainer Pipechase Structural Qualification Summary Attachment 2 Page 55 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-5: McGuire Unit I ECCS Sump Strainer ICW Enclosure Structural Qualification Summary"_ICW GratinQ Evaluation Component Description Measurement Actual Allowable Comments 3/8" Stud Max IR 0.22 1.0 Overhang Tension on 3/8" Max Tension 2714 lbs 3954 lbs Studs_ICW Sides Perforated Plate & Top Splid te Evaluation Component Description Measurement A Alowable Comments Perf. Plate; solid Plate Stress Max Stress 7 psi psi plate ok by comparison Perf. Plate; solid Horizontal Differential M Se0s9 plate okib Pressure toward Grating Max Stress 5949 psi\, plate ok by h,- ýBa s Evalpltate-i.
Component Description Mesrmnt Ata[,' loale Cmet Member Stress Max' '&9 .0 3/4" Diameter Bolts/Studs MaxlR 0.78-,1.0a Weld Faulted Weld I -Base 1.165 1.5 allowables used in evaluation Sot.ýVf C-ne~tb_\4ecIngps 17813 psi MaxR-We2d Base plate____r____
_ Bolt __92 1 Ibs 7439 Sbshe (continued)n 3" Diam'~ý n pe ýC presstýin
+ 0.41 1.0 connecptate
\' : , mding IR cneto"' ICW Frame 2 Evaluation Component Descriptibn,'
rvfasurement Actual Allowable Comments Member Stress" Max IR 0.857 1.0 Faulted 3/4" Diameter Bolts/Studs Max IR 0.78 1.0 allowables used in evaluation Max I R -W eld0.710 WeldsMetal Stress Max IR -Base0.910 Metal ShearIII (continued)
Attachment 2 Page 56 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-5 (continued)
Faulted 1" Diameter Bolt Max IR 0.45 1.0 allowables used in evaluation Shear IR 0.23 1.0 Axial Tension +3.5" PieBending IR 0.26 1.0 Frame 3."Pipe Bedn Rcomponent Axial Comp + 0.35onen Bending IR Bending Stress Base Plate 0_R 1/2" Plate Bending Stress --_si L\\,\440 psi Component Description Measuremen Actual Allowa Commentsl Faulted Member Stress Ma IR, 4 _ 1.36 )I wables usedevlainale 3/4" Diameter Bolts/Studs Ma 0.94 1.36 allowables used_ _hear [41bin evaluation 3/8" Diameter Bolts Shao Bon t 2 71--fv x R- Weid S.499 1.0 W elds x , -,B ase ,\ .81 0 oa -z .51.0 Ma I -1 1.0\Faulted 1 er Bolt 0.62 1.0 allowables used in evaluation Base"RIt \Max 1R 0.61 1.0 hear IR 0.08 1.0 3.5" Pipe + 0.20 1.0 Frame xal Comp +component Bending 0.28 1.0 ICW Frame 4 Evaluation Component Description Measurement Actual Allowable Comments Member Stress Max IR 0.376 1.0 3/4" Diameter Bolts/Studs Max IR 0.83 1.0 (continued)
Attachment 2 Page 57 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-5 (continued)
Max IR -Weld 0.731 1.0 Metal Stress Max IR -Base 0.861 1.0 Metal Shear Faulted 1" Diameter Bolt Max IR 0.97 1.0 alo ed allowables Shear IR 041 1.0 Axial Tension + 026 10 Frame 3.5" Pipe Bending IR component Axial Comp + 2 1.0 Bending IR Base Plate Bending Stress M0.99 Base Plates on IRson Wall 1/2" Plate at Connection Bending Stress 1584 psi 14257813p Faulted 3/8" Bolt IR (No tension) 1.0 l 3/8" Clip Plate 1.0 IC :in e ivaluato Component Description Measuri'n Ata Aloable Comments Member StesMxI 5,, 1.0 1/2" Diameter Piin Max Shear 4336 psi 4961 pbs 3/8" Connection lat(n u 7rake Stress ý/0109 psi 14250 psi 1/"Co re~n- ctAcceptable using/Bakt--/3 psi 21375 psi faulted 1/2 \ allowables e tes 0.96 1.0~Acceptable using Nýx IR -e 1.49 1.5 faulted Aetal Shear " allowables Beding Stress 05 .3/4"Bas Plae _Max IR3/4"Bs0laeco Bolt Max i 0.99 1.0 Shear governing_IRL on bolts1/4"Plate Bending Stress 15360 psi 17813 psi 3/8" Diameter Studs Alwbe 480 Ibs 1318 Ibs Tension Faulted 1" Diameter Bolt Total Shear 6516 Ibs 7439 Ibs allowables used in evaluation 3" Diameter Piping Max Shear 4336 psi 9500 psi (continued)
Attachment 2 Page 58 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response T Max IR- Weld Metal Stress 1AInlr~Base 0.915 1.0 Metal Shear (continued), Attachment 2 Page 59 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-6 (continued) 18" Diameter Cross Connector Pipes and Supports Evaluation Component Description Measurement Actual Allowable Comments Tube Steel 2"x2"x1/4" Max IR 0.94 1.0 Acceptable per AnhrBlsMax IR 0.98 1.0 engineering AnchorBoltsjudgment Plate Stress IR 0.86 1.0 MaxI W eld 0.8 'n.WlsMetal Stress 01.Max IR -Base connection bolts.The Unit 2 strainer is represente tables a distinct sections:
the top-hat strainer modules, the portion of themapiide the e Wall (ICW), the portion of the strainer outside the Crane Wall (e in t chas e ICW Enclosure and the Pipechase Vortex Suncrete a The Unit 2 straine built o stage ten st ainer section inside the Crane Wall and its ciated losure sr ure were a ded separately in a refueling outage subsequent tosnsturn of the of the Main Strainer and its Ta OS-7: Mc U CCS S Strainer Top-hat Module Structural Qulfication Summary Component iption V6asurement Actual Allowable Commentsýe ding Stress 703 psi 4509 psi Top-hat Loadin ýial Stress 86 psi Negligible stress Hoop Stress 533 psi 4509 psi\ ending Moment 703 psi 24086 psi Top-hat Buckling Axial Loading 86 psi 33011 psi Circum. Loading 533 psi 1248 psi 3/8" Diameter Studs Max IR 0.04 1.0 Top Cover Plate Bending Stress 3168 psi 16875 psi Base Plate Max Stress 4287 psi 16875 psi Base metal 1/16" Fillet Weld Max Force 83.95 Ib/in 563 lb/in shear; weld allow.= 928 lb/in Attachment 2 Page 60 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-8: McGuire Unit 2 ECCS Sump Strainer ICW Structural Qualification Summary IOW Strainer Wing Plenums Structural Frame Evaluation Component Description Measurement Actual Allowable Comments 3/8" Plate of Tube Module Max Stress 8320 psi 17250 psi Member Stresses Max IR 0.334 1.0 Plate Element 3/8" Thick Ancho Boltss based on top7 hat Joint SýW.Faulted loads 3/8" Diameter Studs Max IR 016/utilized w/normal allowable1Conservative based on top hat Anchor Bolts Max IR 0.97 1 0 t izes utilized (40" vs. Sha actual)1/2"x3"x3.5" Plate " dicStress
ý/67psi 1\ICW Strainer Extension Plenum Strld Or m -~a e ýOf Oc.Connector and Waterbox)Component Description
!Measur ...._,, -, ctuali Allowble Comments 3/8" Plate of Tube Module Max Str "as" ",0 si 1 '075 psi Middle plenum Member Stresses/ý Max IR 0861 End plenumýml Stes 1C ,3 s 05 s Faulted Plate Element 3/8' " 020 ps allowable Joint R .9 1.0 Connectorpeu 3/8" DimaxlSdsKý R 1.0 WB 127')a1trSusaXR&661.0 End plenum hor Bolts aqR0.97 1.0Concr plenum 1/2"x5i"X Plate I 30 s 67 s End plenum\Max Ilk 0.857 1.0 Ma edEnd plenum;IR-Wl 0.619 1.0 Faulted load;Wed tal Stress normal allowable Weld End plenum;_/Meax IR-Bshea 0.968 1.0 Faulted load;I -Mtal hearnormal allowable Attachment 2 Page 61 of 68 McGuire Nuclear Station Ge-neric Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-9: McGuire Unit 2 ECCS Sump Strainer Pipechase Structural Qualification Summary Pipechase Waterbox (WB) & Plenums Evaluation Component Description Measurement Actual Allowable Comments 3/8" Plate on WB Max Stress 17040 psi 17813 psi Plate at top of water box Bending Stress 14615 psi 17813 psi Plate at end of 3/8" Cover Plate Stress on 3/8" 4 angle 4784 psi/ ,;7813 psi water box Member Stresses Max IR 0 7 1.0 Plenum Plate Element 3/8" Thick MaxStress 1Max MaxNr \189813 psi 3/8" Diameter Studs Max IR 1/2" Diameter Studs Joint Max IR -0.91 1.0 Max IR -Weld .An Blts M e 1 54 p 1.0 WlsMMNP Faulted Meal,1 1ý 1.5 allowables Anchor Bolts Me:\i .0 1/2" DiameterWeg Bolt M Max lR 1.0 1/4" Diameter WedgeB olt Max IR, ..eo0 3/4" Plate StreWss iý tes 174pi,/17 psiFate_ allowables Clip Angles ýCr .,1Ibs/in 2784 Ibs/ in Allowable forvalution3/16" weld cmoettDescri ptin Me;ument Actual Allowable Comments Men~býSre M~a-ý \ 0.135 1.0 Plae lehik lx tlasý,556 psi 16875. psi\3Z8, \_ Ma IR 0.135 1.0 Plate Eiemete "tus ,Max IR 0.135 1.0 Welds xx IR -ed 0.452 1.0/Metal Stress Max IR -Base 0.510 Metal Shear 0.510 AnhrBlsMax IR 0.86 1.0 AncorBotsMax Plate Stress 12480 psi 16875 psi Attachment 2 Page 62 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-10: McGuire Unit 2 ECCS Sump Strainer ICW Enclosure Structural Qualification Summary ICW Grating Evaluation Component Measurement Actual Allowable Comments Description 1/4" Diameter U-Bolt Horizontal Load 184 lbs 5472 lbs Per panel basis 1/4" Tap Screw Capacity Horizontal Load 108 lbs 2792 lbs Per panel basis 3/8" diameter stud Tension 413 lb 1604 lbs Shear 22/b 578 Ibs Angle 3"xl-1/4"xl/4" Bending Stress 1 s 17813 psi Angle 3"x3"x1/4" Bending Stress 6 pi 1375 psi ICW Perforated PI id Plate Evaluatio Component Measurement Actual Allow _ 5p Comments Descriptionx Hydrodynamic Pressure O i Plateg Max 0 tress 1 si 2139 psi Plouratd PlateiMan26 719o ICW Frame 3 Evaluationin this evaluation)
Component Measurement Actual Allowable Comments Desc~ription-
__________
___________
_______Faulted Member stre~sses ,Max IR >0.566 1.0 alwbe1.0allowables
'W~ ~,e~ Stresse 0.589 1.0 Bs ý_tIV~ hear 0.921 1.0 Anchor Me \ -ax R11 0.56 1.0 Base P tx stress 12028 psi 21375 psi Tube steel to Baq'fPat Conetinear Stress 4321 psi 8494 psi Bearing of 1" Bolt jdaring Stress 13568 psi 21375 psi ICW Frame 3 Evaluation Component Measurement Actual Allowable Comments Description Member stress Max IR 0.566 1.0Fale allowables Welds Max IR 0.78 1.0 (continued)
Attachment 2 Page 63 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Attachment 2 Page 64 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-10 (continued)
ICW Enclosure Miscellaneous Evaluations Component Measurement Actual Allowable Comments Description Faulted Frame 1(2) IR 0.424 1.0 alo e s allowables Horizontal Beams Frame 3 IR 0.768 1.0 Faulted Horizontaa Bem Ilowa bles Supporting Grating Frame 4 IR_0848 1.0 Normal h_ _allowables used Frame 5(6) IR 1.0 Faulted____________________allowables Reactions MxI Welds Weld stress 14233 psi 1 si Fltd allowable Diagonal Bracing of Shear Stress\ 5869 psi 1274 _______Frames Bearing Stress 885 ý. 213751 Plate and Tube Steel to Bend Stress 20250 psi Wel F rce22/- 3375 Ibs/in-Support Grating 3/4" Diar~Tý Bl Horizontal Field Splices Ma 1.0ram platex ' s 2i250 psi Vertical posts ICW Extnsion Straneytructural'fm Member stress Ma IR 0.811 1.0 M ress 0.691 1.0 1Faulted 1 .289 1.5 allowables used in evaluation 3"x3 angle Shea ss 13970 psi 14250 psi Torsion Anchb M 0.78 1.0 Base~'\F tax Stress 5525 psi 14250 psi 1" Bolt -Conne6 f Tube steel to Basý ear force 5858 lb 7852 lb allowable Plates Shear Stress 3214 psi 9500 psi 3" Piping Compression/Bending 0.54 1.0 L6_ IR Attachment 2 Page 65 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response Table 30S-11: McGuire Unit 2 ECCS Sump Strainer Pipechase Vortex Suppressor Structural Qualification Summary Attachment 2 Page 66 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for AdditionalI nformation Draft Response M31. The NRC staff considers in-vessel downstream effects to not be fully addressed at McGuire as well as at other pressurized-water reactors.
The supplemental response for McGuire refers to the evaluation methods of Section 9 of Topical Report (TR) WCAP-16406-P, Revision 1, "Evaluation of Downstream Sump Debris Effects in Support of GS-1 91" for in-vessel downstream evaluations and makes reference to a comparison of plant-specific parameters to those evaluated in TR WCAP-16793-NP, Revision 0, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." The NRC staff has not issued a final Safety Evalua (SE) for TR WCAP-16793-NP. The licensee may demonstrate that in-vess nstream effects issues are resolved for McGuire by showing that the lice e. plant conditions are bounded by the final TR WCAP-16793-NP and the Or rg final NRC staff's SE, and by addressing the conditions and limitat the fi E. The licensee may also resolve this item by demonstrating wi eference t6 WCAP-16793 or the NRC staff's SE that in-vessel downs effects have b ddressed at McGuire. In any event, the licensee s>, Id report how it ha ressed the in-vessel downstream effects issue withi- ays uance o inal NRC staff's SE on TR WCAP-16 3. The NR veloping a R tory Issue Summary to inform the in the NRC expectations anc plans regarding resolution of this -aspect .1-191.McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16793-NP in-vessel downstream effects issues" at that time.This draft response is consistent with the one discussed on that call.McGuire N ar Station ddre e in-vessel downstream effects issue within 90 days of issu of the sta nal Evaluation on TR WCAP-16793.
Attachment 2 Page 67 of 68 McGuire Nuclear Station Generic Letter 2004-02 Supplemental Response 11/18/08 NRC Request for Additional Information Draft Response M32. Please discuss why the IPT provided a representative debris bed on the top-hat strainer module for filtering chemical precipitates.
The NRC staff observed the debris addition video and concluded that the fibrous debris introduced into the test tank was more agglomerated than what may arrive at the strainer under post-LOCA flow conditions in the plant. Please discuss whether the amount of bare strainer area observed in the test is representative of what is expected to occur with the plant strainer array if a LBLOCA were to occur. The use of chemical effects test results derived from a test which formed a non-prototypically partially clean screen fiber bed would not be appropriate.
McGuire Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.m The Integrated Prot replaced by the C 12 of this submittal interfaces and the c The 30--~l Q1;S the M ,r Y20 N Closuri he Area describe viousl mechanical\
tor be effective inthS 2009 test series an a nce in the question are being rated e response to RAI question iscussio garding the associated Duke/NRC I eeomeo CIT.p)e int 2010, is designed in accordance with wGuid8 eRegarding Generic Letter 2004-02 Loss and Vortexing" and the testing experience ttlin the test tank will be minimized by using evio esting (June/November 2009) have proven to Kpected, conventional debris bed formation during the e more uniform.Attachment 2 Page 68 of 68 Attachment 3 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Responses September 30, 2010 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C1. Please state whether or not the break location selection was revisited when the Zone of Influence (ZOI) for fibrous insulation was changed from 17D to 7D. If break selections were not revisited, please provide the rationale for not doing so. If the break selections were revisited, please provide the top four breaks in terms of debris generation for the 7D ZOI (The supplemental response sent by letter dated February 29, 2008, indicates only that the break locations already identified for a 17D ZOI were reassessed for debris quantity generation and confirmed not to have changed relative ranking.)Catawba Response-The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable".
This draft response is consistent with the one discussed on that call.In the time period since the S p1ental/Ar Supplemental ose were submitted in spring 2008, lo has dere r:ev ed that reliance on the WCAP-16710-P jacketed fiber ion antd a efine1 tos is no longer necessary, primarily due to larg, eplaefiber in eachti-p enmit 1 Unit cooat has already taken placed amut o fib usbakinsulat ion on sctionso than that of Unit 1cs even afte gs cloet. s position to the staff in a the steam genera~~~~~~~tors withinate iD Oeanenrpaedwt M nuain letter dated Julyanke i itionallyi ous debris quantities generated from destroyed fjb-[JsulN e e NEI 04-07 GR/SE only (i.e., using The f tement of o T e 1 B lo o n hot teg s has the same fbuiltionof a the hocatio le ation per b t ivhe Dloay other fiber reduction refinement would.Therefore, tleaatawba g e rns were revisited in Unit 1 using the post-insulation rep c ent co fiuration and a 17D ZOI to ensure the limiting break was Following fiber inu placement, each Unit 1 reactor coolant loop still contains limited amounts of fibrous blanket insulation on sections of the hot legs and the crossover legs closest to the steam generators.
Fibrous insulation on the steam generators within the 17D ZOI has been replaced with RMI insulation.
The amount of blanket insulation remaining on the crossover legs is identical for each of the loops. The 1 B loop hot leg has the most fibrous insulation of all the hot legs, followed by the 1 D loop, the 1 A loop, and finally the 1 C loop.In general, a hot leg break at Catawba generates more debris than a cold leg or crossover leg break. This is due to the fact that the hot legs for the B and C Attachment 3 Page 1 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response steam generators and the hot legs for the A and C steam generators are adjacent to each other. Thus, a hot leg break on one loop has a larger impact on the adjacent loop than a cold leg or crossover leg break would.The limiting break from a debris generation perspective is thus on the 1 B loop hot leg with a break location adjacent to the steam generator.
The next limiting breaks would then be on the D loop hot leg; followed by the A loop hot leg and finally the C loop hot leg. The next limiting series of breaks would be on the crossover legs, in the same order.In terms of debris generation, breaks in Unit 2 are since Unit 2 contains significantly less fibrous ir/containment.
led by the Unit 1 breaks, Attachment 3 Page 2 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C2. Please state whether the testing identified in the test report WCAP-16710-P,"Jet Impingement Testing to Determine the Zone of Influence of Min-K and Nukon Insulation for Wolf Creek and Callaway Nuclear Operating Plants," was specific to the Catawba Nuclear Station, Units 1 and 2, (Catawba)insulation systems. If not, please provide information that compares the Catawba encapsulation and jacketing systems structures with the systems that were used in the testing, showing that the testing conservatively or prototypically bounded potential damage to the insulation materials.
Catawba Response: The response methodology for the oe dis eon a cussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable".'
This draft response is consistent with the one discussed on that call.Wý-In the Catawba GL 2004-02 Sti Catawba GL 2004-02 Amend d quantity of fibrous debris genera in the ECCS sump pool was deb described in both the 04-07 Safety Evaluation -Determine theZ of lnflu de Creek and Callawa uclear Catawba E S sump guida n ibe l a Protc est (IF" u11 ch'1z 1"' fi (dated 2/2 and the a) -ei~ponse dated 43/2008, the J10 *Jber insulation and deposited the s of Influence (ZOIs)P (an associated NRC I ar ýý nIm ngement Testing to-K a~n KON Insulation for Wolf ts," Revision 0. Specifically, the esigned using the NEI 04-07 ZOI the subsequent Integrated
ýs WCAP-refined ZOI values.In the ti nriod were subm C jacketed fiber this position to t generated from de only.3up ental/Amended Supplemental Responses s dete'Y ined that reliance on the WCAP-16710-P iefinements is no longer necessary.
Catawba statedýtter dated July 28, 2009. Fibrous debris quantities)er insulation will be based on the NEI 04-07 GR/SE Attachment 3 Page 3 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C3. Considering that the Catawba debris generation analysis diverged from the approved guidance in NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Revision 0, please provide details on the testing conducted that justified the ZOI reductions for jacketed Nukon.The information should include the jacket materials used in the testing, geometries and sizes of the targets and jet nozzle, and materials used for jackets installed in the plant. Please provide information that compares the mechanical configuration and sizes of the test targets and jets versus the potential targets and two phase jets in the plant. Please evaluate how any differences in jet/target sizing and jet impingeme angle affect the ability of the insulation system to resist damage from je ngement. Please state whether the testing described in test report C-16710-P was bounding for the Catawba insulation systems. If not, ps'I) *de information that compares the Catawba encapsulation acketi stems structure with the system that was used in the tes owing tha testing conservatively or prototypically bo d potential dam to the insulation materials.
Catawba Response:
C The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable".
This draft response is consistent with the one discussed on that call.In the ta L00 uLIPP al ponse dated 2/29/2008 and the Cata4 L 200 me Supper tal Response dated 4/30/2008, the quanti t fibrous de e"d from estroyed fiber insulation and deposited in the s ump poo de ned using the Zones of Influence (ZOls)described I h the NEI 07 G lance Report (and the associated NRC Safety Evalua i(GR/SE nd WCAP-16710-P "Jet Impingement Testing to Determine the o f Infl ece (ZOI) of Min-K and NUKON Insulation for Wolf Creek and Callawa Operating Plants," Revision 0. Specifically, the Catawba ECCS sum p iners were initially designed using the NEI 04-07 ZOI guidance relating to fiber insulation ZOls, and the subsequent Integrated Prototype Test (IPT) for chemical effects uses WCAP-refined ZOI values.In the time period since the Supplemental/Amended Supplemental Responses were submitted, Catawba has determined that reliance on the WCAP-1 6710-P jacketed fiber insulation ZOI refinements is no longer necessary.
Catawba stated this position to the staff in a letter dated July 28, 2009. Fibrous debris quantities generated from destroyed fiber insulation will be based on the NEI 04-07 GR/SE only.Attachment 3 Page 4 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C4. The NRC staff is not convinced that Catawba's currently postulated limiting break, that results in no fine fibrous debris, but does result in 195 ft3 of small pieces and 130 ft3 of large pieces, is truly the limiting break from a final head loss perspective.
Please provide the fibrous size distribution (including debris amounts determined) for the debris generation calculation based on the 7D ZOI. Please provide the basis for the determination that no fine fibrous debris would be generated by the limiting break. (The NRC staff considers the assumption of no fine fibrous debris to be non-conservative and inconsistent with previous industry and NRC insulation destruction test data that indicates that a fraction of the debris formed within a 7D Z would be destroyed into fines. The NRC staff guidance for break selec E Guidance Report and NRC staff Safety Evaluation) requires that" reaks shall be postulated combination of isri typsntdhesur the wos casic with the goal of creating the largest quanl is and/or the worst case combination of debris types at the sup een. fiber is a basic constituent of a limiting debris bed.) erent bre ation would result in the generation of fine fibrous debris en if the total d amount is less than the currently postulated Cata i mting b ak, that reent break may actually be the limiting break. The licen slu evaluate potential break location from debris ration to rt (including er ion and ensuing transport) to head etermin ich break is actually limiting.)
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response differs in part from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010 in relation to incorporating benefits of ECCS water management modifications into the debris transport analysis.As discussed in t sp s to RAI questions 2 and 3 and in a letter to the NRC staff dated Ju 9, Catawba is no longer implementing the jacketed fiber insulation refiner s (i.e. the 7D ZOI) identified in WCAP-16710-P.
As a result of this change, a 17D ZOI for postulated breaks for both jacketed and unjacketed fibrous insulation will be used in accordance with NEI 04-07 Guidance Review/Safety Evaluation (GR/SE). Consistent with this approach, a more industry standard four size distribution model consisting of individual fines, small pieces, large pieces and intact blankets is being imposed on the fibrous debris that-is generated.
Within the overall 17D ZOI, the size distribution of the debris that is generated varies depending on the distance of the insulation from the break. (i.e. insulation debris generated near the break location would consist of more fines and small pieces than insulation debris generated near the edge of Attachment 3 Page 5 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response the ZOI). Three sub-zones were determined within the 17 D ZOI for quantifying the debris size distribution.
The breakdown of fiber sizes within each sub-zone is provided in Table 4S-1 below: Table 4S-1 Low Density Fiberglass (LDFG) Debris Distribution Within Each Sub-Zone Size 0D 11.~A9-7.0 D 17. 0-111. 9 D'ZOI o ZOI Fines (Individual Fibers) 20% 13% 8%Small Pieces (<6" on a side) 80% 4% 7%Large Pieces(>6" on a side) 0% ý% 41%Intact (covered)
Blankets 0% 17% 44%This change in approach is primarily due e remov fibrous insulation from the steam generator barrels and reduce ris transpoi tions as a result of approval of the ECCS water managem License Amend Request (LAR).The ECCS water management LAR an odifica (scheduled for implementation in Fall 2010 for Unit 2 and " for Unit a1ys the start time for initiation of containmi ay after a nergy pipe bre inside containment, allows the syste t6i!jp qte with one Containment Spray train, increases the minimum req e of th fueling Water Storage Tank (RWST), and lo the R -b v The anges result in higher containment es (d in e t and increased useable RWST e) an wer su o turb e (due to lower spray return flows throuvo riefu .g cavityd P ins, and an overall lower flow/approach veloci t oLh EGG m p Strainer).
The tatio id ics del previously utilized to calculate debri sport frac r t) rious si es and types of debris was utilized again to tify the be *ts o S water management.
There were no changes to model its ,y ti tnput parameters of ice melt flow rate, steady state \inment up pool volume, spray return flow rate, and overall!The previous debris qlaies reported in the response to RAI 12 in Catawba GL 2004-02 Amended Supplemental Response dated April 30, 2008 as well as the current debris quantities (modified via fiber reduction, size distribution, and revised CFD model) are provided below in Table 4S-2: Attachment 3 Page 6 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 4S-2 Catawba Fiber Insulation Debris Loads -Limitinq Break DebrisQuant Transpýort Fraction Q1 Nukona and Thermal- !'WrapD Lo'w Density,~Reported on'4/30/2008 (7D ZOI)'Current Debris zQuantity (I 7D ZOI)Water FManagement
<Debris Transport.
Fraction Reported~4/30/2008 QuIantity at'Strainer, 36.75 100%0 ft 3 36.75 131.6 ft 3 10.4%88.3 ft 3 13.68 ft 3 42.84 ft 3 p0%. 13.0 ft 3 4.28 45.87"0%0 ft 3 0 257.1 ft 3 Total LDFG Transported 54.71 ft 3 Attachment 3 Page 7 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C5. Industry debris destruction testing was used as a basis to revise assumptions concerning the ZOls and debris size distributions for Nukon,,Knauf, and Thermal Wrap low-density fiberglass insulations.
Please describe the jacketing, banding and latchinq mechanisms, and cloth covers of these three types of insulation installed at Catawba and compare them to the insulation for which destruction testing was performed in order to demonstrate the,.applicability of the industry destruction tests results to Catawba.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.In the Catawba GL 2004-02 Supplemer Catawba GL 2004-02 Amende ,SupplE quantity of fibrous debris genb:at[d-foi in the ECCS sump pool was de described in both the NEI 04-07 da Safety Evaluation (G a Determine the Zo1ce (Z Creek and Calla uce peratin Catawba ECCS su M train were in guidance pr g to fi n 01 Prototvy for ica espon dated 2/2aOr08 and the ase dated, 008, the Qtý fiber insulation and deposited ig th nes of Influence (ZOIs)port cthe associated NRC"Jet ngement Testing to tn NU N Insulation for Wolf Re-=n 0. Specifically, the designed using the NEI 04-07 ZOI the subsequent Integrated WCAP-refined ZOI values.In theWJRperi(were 3ui d jacketed fibr this position t generated from only.N>ementalTAmended Supplemental Responses; d!imned that reliance on the WCAP-16710-P efinehi nts is no longer necessary.
Catawba stated tter dated July, 28, 2009. Fibrous debris quantities r insulation will be based on the NEI 04-07 GR/SE Attachment 3 Page 8 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C6. Please specify whether latent debris samples were collected as part of the containment walkdowns performed described in the supplemental response sent by letter dated April 30, 2008, and describe how these samples were used to estimate the latent debris quantities for both units. In addition, if samples were not collected, please justify how the use of photographs and walkdown notes of the Catawba containments, as described in the response, provide assurance that the 200 Ibm of latent debris assumed for the supporting calculations is bounding.Catawba Response:
___The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.m Latent fiber (i.e., lint) quantitie quantities at Catawba were es guidance, in combination with th companion Safety Evaluation (SE, estimates were us esent equivalent to Uni nce inr maintenance and ning p tici assumption was yen saS viav and the r,ý +/-ismls e I a (i.e., dust an 6Art)SND -01 containment walkdown 7 GuiL Report (GR) and its t a th it 2 latent debrisý sI s 1 is expected to be ar ely identical and intical bet een Units. This lpfing comparisons between Units Iotals.Methkcgpgy The folg quantify tf*Estir, I ied (in accordance with the GR/SE) to containment:
vertical surface area" Evaluatek en ris buildup* Define spec eis densities* Determine frac nal surface area susceptible to debris buildup-Calculate total quantity and composition of debris Also in accordance with the GR, containment was segregated into four areas based on the presence of robust barriers and representative surfaces:* Lower Containment inside the Crane Wall" Lower Containment Pipe Chase (outside the Crane Wall)" Upper Containment" Ice Condenser Attachment 3 Page 9 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Surface types within each of these areas were categorized as (a) Horizontal Floor Surfaces, (b) Horizontal Miscellaneous Surfaces, or (c) Vertical Surfaces.With the sampling surfaces defined, specific areas were chosen in order to obtain their representative online condition.
Sampling media included Masolin cloth and sticky foam. All sampling media was pre-bagged and labeled. Each bag contained a single Masolin cloth and a single sheet of sticky foam (sized approximately 9" x 12"). Each bag was then pre-weighed on a Mettler Toledo PR5002 scale, with a tolerance of +/- 0.04 grams.The area was wiped down with a Masolin cloth to p fine debris and to consolidate larger particulate debris. Vertical s s were wiped from the bottom up to prevent a loss of debris. The sti heet was then used to pick up any remaining particulate debris. % i e M cloth and the tacky foam sheet were carefully folded to prevn ss of debri terial and placed back in the sample bag. Excess air w i he bag was ge frced out to allow the sample to be easily transported an t-weigh .After mpling was complete the bags were weighed on the sai scfeThe differ between the pre- and post-weights were th ed to calI e mass of the bris collected.
Sample mass measurements wer re an t to account for a number of possible/
and m .e e rs uding loss of sample media to the sa r1 rfa' ir mov, o scale during measurement, an toler e of the. ce. The error due to air movement and the tolerance of ca Id aso\ in a conservative direction for some samplesmse Wmass), therefore the offset applied to thes p e er i ses t i st vatism leading to a higher estimtd latent de adi There wer -individua nt d samples taken, with the following itemization b ace typ , Horizontal Floor e samples Horizontal Miscellan urfaces: 25 samples Vertical Surfaces:
16 s pies Once the sample debris mass for each surface type was quantified via scale measurements, the specific debris density for each sampled area was computed by dividing the individual sample masses by their respective sampled surface areas. The sample densities were then grouped into the following sample sets based on common surface type and location, common associated work activities in each area, and cleanup procedures (e.g., similar work activities and cleanup Attachment 3 Page 10 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response would be expected for the floors in Lower Containment inside the Crane Wall and in the Pipe Chase): " Horizontal floor surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase" Horizontal miscellaneous surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase" Vertical surfaces in Lower Containment inside the Crane Wall and in the Pipe Chase-Horizontal floor surfaces in Upper Contain and inside the Ice Condenser" Horizontal miscellaneous surfaces i e tainment and inside the Ice Condenser Note. The mean density for th o ntal miscella is surfaces measured through the above p s was less than er areas of containment.
Instead of using thhe s fsure 'jlue, a m ensity of twice that of lower cont ent horiz. elaneous s es (maximum debris dens , used in alculation to more conservatively account ff of uppy tainment.Vertical surfaces in Uppe nta and "e the Ice Condenser A statistical anal as t erfor sgth ped sample densities to provide a conservai asse ent of e is buildup over a given surface type.While the GR/SE stas hat verage at east three samples for each surfacenir b .rface, the Catawba analysis goes a ste 4,-er lm e 95n ce interval of the mean (average)debri sity. This' roaý ovides Mrgin in the calculation of total latent debris i s'the conta s. e maximum predicted mean latent debris densities v then multi iby y appropriate actual surface areas inside the Catawba con ent, wh were based on reference drawings and information developed from walkd s. The total latent debris loads calculated conservatively ass tt 00% of the estimated surface areas in containment are susceptible to de cumulation.
Using this methodology, the extrapolated latent debris mass total for the Catawba Unit 2 containment was determined to be approximately 113 Ib, which bounds Unit 1 as reported in the Catawba GL 2004-02 Supplemental Response dated 2/29/08, Enclosure 2, item 3(d)2. It should be noted that applying the methodology described in the SE of using a simple sample mean, the total estimated mass would be approximately 70 Ibm.Attachment 3 Page 11 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response For conservatism, an overall value of 200 lb was assumed for the total latent debris quantity in both Catawba Unit 1 and Unit 2 containments, 30 lb (15%) of which is considered to be latent fibers per the GR/SE.Attachment 3 Page 12 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C7. Please describe the analytical method used to extrapolate the total amount of latent debris in containment.
If a statistical method was used, please provide the confidence level of the results.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable".
This draft response is consistent with the one discussed on that call.Please reference the response to RAI #6 of regarding the methodology used to calcu containment at Catawba.within Attachment 3 Page 13 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C8. Please provide the details of the methodology used for the tag and label refinement evaluation.
Please provide details of the equipment qualifications and engineering judgments used as basis for reduction of tag and label quantities assumed to fail and reach the sump.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable".
This draft response is consistent with the one discussed on that call.Please reference the response to RAI #1 regarding the Catawba tag and label r Attachment 3 Page 14 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C9. Please provide the technical basis for the assumption of 10-percent erosion of 'fibrous debris in the containment pool. If testing was performed to support this assumption, please demonstrate the similarity of the flow conditions, chemical conditions, and fiberglass material present in the test versus the conditions expected in the Catawba containment pool.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of vendor erosion testing issues" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.The quantity of constituent fiber fines tr rting t e Cata trainer due to the erosion of submerged but Peof fiber was determined by testing in orde the con erive erosion assumptions documented in the GIL 2004-02\_'f Jance r 04-07 and the companion SER). The objective of t sting t y ntainment pool flow-induced erosion/t riorati at ma w d byon lo nsity fiberglass insulation (LDFG). This was mplishd by su ting a measured test sample of the LDFG mat t pe t water flow in a closed vertical test rt aal t flume (TF) for durations of up to 72 hý(r a ýga1 ia ss MWt:ýy have eroded off of or otherwise dislod rom the te mp hese two apparatuses (VTL and TF) were used to comp e differen ulei nd energy effects upon the insulation, as th e of theZientation of the sample with respect to the water flow. ntify th frous mass loss, the dry weight of the test samples before and after est w smeasured and recorded.
The testing was conducted on both small pieces of low-density fiberglass to observe any effects that size o ace area had on the sample's erosion.Subsequently, a 30-day erosion test in the horizontal flume apparatus was also conducted.
Analysis of this test data provided further insight into the nature of the LDFG sample composition and consequently, its erosion characteristics.
Primarily, it was observed that during 30 days of flow impingement, the sample did not continuously disintegrate.
The fiber insulation samples appeared to yield loosely bound fiber fines early in the test, after which the erosion effects subsided.Attachment 3 Page 15 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Test Inputs Erosion tests were conducted with LDFG insulation samples in conditions intended to mimic, or be conservative with respect to, the expected post-LOCA plant conditions.
Debris Type and Size: Both Nukon and Thermal-Wrap fiber insulation exist in the break zones of influence in the Catawba containments, and for the purposes of LDFG evaluation can be considered equivalent consistent with the LDFG destruction pressures discussed in the GR/SE, Section 11.3.1.1.
The erosio testing used Nukonsamples with the same bulk density (2.4 Ib/ft 3) as t'sed in the plant. The Nukon fiber insulation sheets were cut into 6"x rectangles to represent the large pieces, and into 1"x1"x1" squares to rels small pieces (an extra large piece measuring 6"x6"x1" was also i ed). S les were then boiled in tap water for ten minutes to remove the r, in or ut th conditions the fiber insulation would undergo durii ge blowdown an CS sump pool recirculation phases of the predicted po -CA res onse.Test Environment:
S The insulation samples were s to a flo sion environment in both a vertical test loop apparatus and 1 flume, ch test consisted of subjecting insulation samples to filile " i r fi.in9 VTL or TF with tap water and then cir e wate bou w c,.itions that occur in the ECCS sump POO. e saN es wer y co ly submerged during testing to ensure rervatiV osion. large Nu kon samples were fastened to a screen p neces.a, movement.
The small Nukonsamples Im i awire cage in the flow stream, such that t as er with e o or the release of eroded material.Flow V ' st The erosi sts were p rme a flow velocity that is equal to the incipient tumbling vel for the s ific size. For the LDFG samples tested, the flow velocities were rmineo be 0.37 feet per second for the large pieces and 0.12 feet per seco r mall pieces. Since the incipient tumbling velocity is the velocity at which e ris would start moving, this velocity bounds the greatest velocit at a iece of insulation lying in the containment pool would experience without being transported to the ECCS sump strainer.
Therefore, it is considered the velocity that would produce the most fiber fines from submerged, but not transported, fiber insulation pieces.Water Temperature and Chemistry:
Temperature
-As discussed previously, the LDFG samples were boiled prior to being subjected to flow impingement testing, to simulate the conditions present during blowdown and recirculation.
The actual tests Attachment 3 Page 16 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response were conducted in the VTL and the TF in room-temperature tap water (i.e., approximately 60°F-80°F).
The temperature of the water increased during testing due to continuous pump heating (up to 11 0°F for longer-duration tests). It was determined that viscosity effects on the erosion rates were insignificant., Chemistry
-The erosion tests were conducted in tap water and not the buffered or borated water predicted to be present in the containment sump pool post-LOCA.
The use of tap water is considered appropriate because the lack of chemicals such as soluble aluminum, boron, or pH buffers will not affect the amount of fibers that would ero om a Nukon LDFG insulation sample.Analysis of Erosion Test Data During the erosion tests, the small Nuko s ge Ily eroded more than the large samples, despite the large s s undergoing her flow velocity.Small samples eroded more fibrous m e to their increa esurface area and their being prepared for the tests byI ddin ich pro more fines available for transport.
Since th, small sa ed more t large samples, the test data analysi d only th all sample results in order to generate conservative fiber ero tities, (th ore these higher small piece erosion rates are applied to both- r ofsb1 ed, non-transported fiber insulation).
The small fiber t ample tona aq, abou :/o to 7% of initial weight for any given test dura" h ý raw da nge spanning 0% to 20%. Because the LDFG n da t isten Ne data was analyzed and results compa etrap if n rc 1. sume the y s sample sion test results represented the I onre 2. Det e the wei h loss p hour rate for all small sample erosion tests, and th trapolat at weight loss per hour value to 30 days to properlyECSmssion time.3. Average all o all sample erosion values regardless of test duration (including RMS or), and assume this value applies for the ECCS mission time.Application of approaches 1 and 2 above yielded 30-day fiber erosion estimates that appeared to be non-conservative when compared to the majority of the small sample data points.Since the fiber erosion test results showed wide scatter across all test durations, the assumption was made that fiber erosion is not directly time-dependent, and therefore could be conservatively described by averaging all of the small sample Attachment 3 Page 17 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response erosion test results to reach an overall erosion value (i.e., approach 3 above).Additionally, as noted previously it was observed that during the 30-day erosion test the samples did not continuously disintegrate.
The fiber insulation samples appeared to yield (transport) loosely bound fiber fines early in the test, after whic[N-he erosion effects subsided.
As such, the overall erosion value calculated from the small sample average is considered applicable to a 30-day mission time.The calculated average of the small sample fibrous erosion test results was approximately 6% of initial weight, with an error of +/-+ as determined by RMS error analysis versus the calculated average. Ap a R 3, then, determined a conservative estimate of the overall fiber erosio e to be 10% of the initial fiber weight (6% + 4%).Therefore, the attrition/erosion mechani LDFG via water impingement is conse weight of submerged, non-transported I ECCS mission time.Subsequent to the preceding sW discussions were held between C (including the Duke plants) regar th methodology.
Thi/ ' to requ u entifie the vendor to ad r ssu iss nt1 f configuration.
Th -dlay coirmator, vendor in 2010 and _11 -day small pie ilati by 10' "1 Ne 30-day n.tests, extensivee vendor, and Licensees Con ation and analysis ry LG erosion test from te e and erosion sample test series was completed by the value of 10% for both large and Attachment 3 Page 18 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C10. Please provide details of the tags and labels equipment qualifications and engineering judgments used as the basis for reduction of tags and label quantities which are assumed to fail and reach the sump. Specifically, please justify the application of Institute of Electrical and Electronics Engineers#lEEE)
Standard.323-1974, "IEEE Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating Stations," in qualifying Electromark labels for a post-loss-of-coolant-accident (post-LOCA) environment with respect to nondebris transport to the sump strainer.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time, with noted clarifications from the staff This draft response is consistent with the one discussed on that call.The assumptions and engine reduction evaluation performe provided in the responses to Catawba GL 2004- p le The initial tag e ass qualified tag reducti Fo- t with braidless not tr ed .ro Elect rk labe ich The tra ed tag an e reported i e 3D3-2 r Supplementa sponse t below in Table 11 for c n n ments in the Catawba tag and label ad t nt to th I ial assessment were item d) and ojf Enclosure 2 of the ment eso ,ted 08.i only o refinement in the form of a reductio t was assumed that metal tags hung i.ns ' td either not fail, or would sink and netal .e the same as qualified iscuss dter in this response.I ifications by area of Containment were IO 2 of the Catawba GL 2004-02 d 2/29/08. The information in this table is recreated enience.Attachment 3 Page 19 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 1OS-1 Catawba Refined Miscellaneous Latent Debris Quantities Debris Lower Containment Lower Containment Upper (Inside Crane Wall) (Outside Crane Wall) Containment stickLabels ltt) 116.611 58.669 21.380 2.850 199.51 Plastic Tags wlAdhnsive (f1.469 2.774 4.099 4.600 12.942 Plastic Hanging 3438 5.003 4.450 1.500 14.388 Tags (ft 2) __4___.____.45
_ ._5_0_14__
8 RMI ID Stckers[ftel 277.597 66,234 0,000 0.000 343.831 Ice Condenser Debris Ift?) N/_A NIA N/tA 15.3 15.3 Total (ft 2) 399a115 132,677 29,929 24 250 585,971 For the purposes of this response, "Iake marker that is affixed with adhesive.
TI rigid plastic tag or placard that is hung or As stated in item 3(i)5, Electroie lower containment at Catawba h b the limiting break, an s were r vec label debris assu r ttrms" % sport tof E)utside the Crane Wall in the capable of withstanding n,) quantification of tag andý) Sýi n er.As stated in item percentage of tac labels ir'conservatively estimate the lis in the ZOI; therefore, all tags and ,,nt will be assumed to fail.In an Ic -denser co *nme e only areas of the lower containment outside th ne Wall a \ o 't in-it the Ice Condenser Lower Plenum are located betw he Ice C r-enser end walls and in the rooms above the Pipe Chase. Plastic in icinities are generally outside the break ZOls and are assumed to de t lot become overly pliable (i.e., they will not deform enough to pass throu obstruction that has a smaller dimension than the tag).Attachment 3 Page 20 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Ice Condenser Reduction Tags and labels located within the Lower Plenum of the Ice Condenser (and outside the Crane Wall) are assumed to fail since the break energy is directed into this plenum by design.Tags and labels located within the Upper Ple'num of the Ice Condenser would not be expected to fail immediately during the initial venting of air and steam;however, exposure to the post-LOCA environment and containment spray may lead to eventual detachment.
A minor portion of the tags and labels located in the Upper Plenum of the Ice Condenser are not locatabove the ice basket array and are located above horizontal surfaces.
tse tags and labels fail, they will fall straight down and are not expected b tr nsport further due to containment spray.It is likely that some of the remaining ta labels tha within the Upper Plenum of the Ice Condenser will fall y into the ice b s thems ice s themseves.
Given the ice baskets are made of perfb d sheet etal wi 1" holes and the bottom of the baskets are covered by I r mes tags and s rm r cros labels that fall into the ice base ill not be exit. In term oss-sectional area, it is conservativ med tha basket openings comprise 50% of the ice basket array. Ta n Is that ot fall into the ice baskets themselves could fall into the spa etw e ba and the lattice frame which provide sup baske h tc me does create a tortuous path for ags abels, to conservatively estimate an appropriate qu2. of Ia that w remain w~ihin the lattice structure; therefore, no reduction as ki;for tag d labels that may fall into this area.Elev io Red u The s above th ec at elevýn 565'+3" are not subject to jet impinge t or contai t sI Initially, the rooms will not be flooded, but as the accide t ogresses floor e ation of these rooms may be reached.Access to th ms is ga through an opening in the floor from the pipe chase below. the ro s are flooded, velocities in the rooms are expected to be very low an a bels would not transport to the pipe chase below.Only tags and labels above the floor opening are assumed to transport to the pipe chase below.Upper Containment Reductions The majority of tags and labels within Upper Containment are located between the ends of the Ice Condenser walls, in the fan pit, and around the personnel hatch. Tags and labels that detach are expected to fall straight down and there are none that would be expected to fall directly into the refueling canal.Attachment 3 Page 21 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response It is conservatively assumed that all tags and labels that reach the fan pit will pass to the refueling canal. A majority of tags and labels outside the fan pit are located directly above grated platforms.
It is judged that a majority of these labels will .bepasily captured by the grating and thus the quantity of labels above grating is reduced by 75%. Although highly unlikely, it is conservatively assumed-that all tags and labels that detach and fall to the concrete operating floor will be transported over the 3 inch curbing around the refueling canal and through the elevated refueling canal drains to lower containment.
Qualification of Electromark labels The Electromark labels located in containment w e alified for the LOCA environment'via a comprehensive test program purpose of this program was to demonstrate the suitability of applicatin sure sensitive markers in being able to remain in position (on equip r stru s) throughout a specified lifetime, including background ion follow a simulated LOCA.The safety function demonstrated wa ' the markers wo emain affixed to the equipment or structure without fallin, This test program was conduqf iLnder the guidelines as gested in IEEE 323-1974 "IEEE Standar lifying Equipment for Nuclear Power Generating Stations".
Th a hases e program are outlined following:
- 1) Heat Agin Ima of Ion, p o plant ambient conditions attypical a nt tem ratures atmospheric pressure for a period of several years. th s of th4 ggestions and procedures contained S10 le was utilized to extrapolate an;mon0s life period by accelerated
' nai g at dev e e w s.2) R ion Aging the lusion of the thermal aging period the sam were insp cd for gradation and loss of function, and then expose a cobalt source of gamma radiation at a nominal dose rate of 0.5 Mr er ntil a total accumulated dose of 200 Mrads had been receiv e amples were then inspected again for wear and degradation.
- 3) LOCA simulation
-The samples were installed inside a pressure vessel and subjected to an environmental exposure of steam and chemical spray for a period of 30 days in accordance with the suggested IEEE 323-1974 profile. At the conclusion of the exposure the samples were again inspected and compared with the control samples for suitability of function.Attachment 3 Page 22 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Cll. Please provide the results of the array testing conducted at Alion Science and Technology Corporation and the Integrated Prototype Test (IPT)testing conducted at Wyle Laboratories.
For the IPT testing, in addition to head loss values, please provide the results as a function of time. Please provide a thorough description of the methodology used to combine the two test results to determine the final head loss for the strainer debris bed.If a correlation was developed to determine head loss, please provide the correlation along with the assumptions and bases used in the development of the correlation.
Catawba Response-The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.NRC Technical staff mnens with tegrated Prototype Test (debris prep , bis intr gtigfr neration, flow fields, bare strainer aread che I effec discb d at the public meeting convened on 11/240m *nangton, I It was determined that resolution of the issues d by th smeet would require further testing and evaluafiýIn Jute(, f 2009, a ma Head test series was performed at Wyleby Duke6 iane designed flume and conventional debris (i.e., boundingd parti to address the debris preparation, debris introduction, is agglo ation/settling and flow field issues identified by the staff regardinag ,ntegrat Prototype Test (IPT). NRC staff guidance from March of 2008 as ia t from discussions with the staff were used for the test protocol deVE r the Confirmatory Head Loss test series. Upon completion of the June k09 testing, the protocol, debris bed formation and head loss results were discussed with the staff in July 2009, along with photos and videos obtained during the testing. From these conversations, the staff concluded that the methods and protocol being utilized-for conventional debris preparation and introduction during the tests met NRC expectations.
In parallel with the Confirmatory Head Loss test series in June 2009, Duke provided a draft white paper to the technical staff entitled "Duke Energy Chemical Effects Testing in Support of GSI-191".
This document served as a guide for discussions with the staff on the battery of testing performed by Duke to date in Attachment 3 Page 23 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response the area of chemical effects, and to address NRC concerns with the effect of potential chemical precipitates.
As a result of these discussions, Duke elected to continue head loss testing in the fall of 2009, using the Confirmatory Head Loss test series protocol and test flume along with pre-mixed chemical precipitates in accordance with staff guidance.
The fall 2009 testing was identified as the Chemical Precipitates Head Loss test series, which continued into the spring of 2010. The results of this pre-mixed precipitate testing were discussed with the NRC staff in April and May of 2010.Subsequent to the Chemical Precipitates Head Loss test series, Duke determined that a one-time, long-term test approac /izing soluble aluminum injection (as opposed to introducing pre-mixed pr ates) would be more representative of the post-LOCA environment ed in Ice Condenser containments.
Identified as the Confirmato
'gr Test (CIT), the development of the test protocol followed ussions i e 2010 with the NRC Paýstaff. The CIT utilizes previous testing rience with c tional debris preparation, introduction, and transpor acilitate the form of a uniform debris bed on a prototype strainer array. ition the CIT will be thermally and chemically repre tative (P alcium, ilic nd aluminum) per guidance set f VCAP-I3 tilizing a Duke-modified algorithm as described previou atawb neric Letter 2004-02 Supplemental Response, dated /0 The 30-day CIT, sc to bep me 01 designed to provide a conservative an E ific loss a strainer array as a function of the E issiO e, wi, imal pos -processing.
Attachment 3 Page 24 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C12. Please provide information that establishes that vortex testing was conducted at less than or equal to the expected 3.75-inch minimum strainer submergence.
The licensee's response to RAI question 38 in Enclosure 1 to the supplemental response sent by letter dated February 28, 2008, and Enclosure 2 of this supplemental response, Section 3(f)(2), state that the strainer modules are submerged by 3.75 inches under limiting sump level conditions.
The licensee's response to RAI question 38 states that testing was conducted at a submergence of 3 inches.Enclosure 2, Section 3(f)(3), states that the te**Qg was conducted with a"few inches" of water coverage above the s r modules. Separately, Enclosure 2, Section 3(f)(3), states that p& ch velocities for testing were between 0.01 ft/sec and 0.09 ft/s e expected maximum approach velocity for the plant strai 0.04 c. In order to clarify the conditions under which vortey ng was co ted, please provide the following information:
- a. Please provide the basis for the imapproach ity value of 0.045 ft/sec.b. Please discuss how fl .t was rolled during vortex testing.c. Please pr quantit val e a pach velocity during which fy01 r bl~e 1 d. Please pi e a cntitative e for the vortex suppressor grating eýlease Itha vortex ttwas conducted at less than or equal o 3.75 inc f st r submergence, with or without a vortex prsor gi ig f. Pe tate wh er vortex formation occurred during testing and what ctnionss re present at such times (submergence level, approac v i and grating installation).
Catawba Response.The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time, with noted clarifications from the staff This draft response is consistent with the one discussed on that call.Attachment 3 Page 25 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response a. For vortex testing purposes, the as-built maximum approach velocity for the top hats closest to the ECCS suction lines (assuming operation with 2 RHR pumps and 2 Containment Spray pumps) was determined to be 0.048 feet per second in Catawba Unit 2, which bounds Unit 1.This approach velocity does not use the normalized flow distribution approach.
Instead, the flow is distributed among the top hat modules such that the internal losses within the strainer top hat assemblies and plenums are pressure balanced.
This results in a non-uniform flow distribution, which is used to determine the approach velocities.
With an initially clean ECCS sump strainer surf , approach velocities for the top hat modules closest to the pump on lines are expected to be higher than the predicted Catawb oinal approach velocity (i.e.about 0.021 feet per second) by a y a factor of two.It should be noted that CNS s mitted a se Amendment Request for ECCS water m ement, which ha en approved by the Commission.
In relation CS flo this licei mendment allows the Containment Spray s. rate with one pump.This will decrease t erall EGC and thus approa h velocity)through the sump st I pleme, n of the License Amendment is scheduled for fall 2 f 2 and g 2011 for Unit 1.b. During fng, flu elo cs on r ed via a throttle valve on the se ide of c l ump. Vortex testing was perform Hitia lestabli, flow at 0. 1 feet per second for a time of 0 e t a x suppressor installed.
Flow was Seas) e ond increments until a vortex was mse m of I u s was allowed at each flow rate to allow time vo to form. nce an air-entraining vortex was served, t ppr was installed.
Flow was incrementally as-ed in 0. oot p~e econd increments up to a maximum test ap ch veloci f 0.09 feet per second, which is approximately twice the m x um e pted approach velocity for the Catawba Units.c. Without vort suppression installed, at approach velocities at and above 0.04 feet per second an air-entraining vortex was present. With the vortex suppressor installed, vortices were eliminated and only minor surface dimpling remained up to the maximum test approach velocity of 0.09 feet per second (approximately twice the current maximum expected Catawba approach velocity, with even more margin post ECCS water management implementation).
Attachment 3 Page 26 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response d. The in-plant configuration has the top of the vortex suppressor grating at an elevation of 554 feet, 9 inches, which is also the same as the minimum flood level in containment.
It should be noted that this minimum flood level is based on a number of significant conservatisms.
It is calculated based on the break being small enough that no ice melt has occurs, the Reactor Coolant system remaining water solid, the Refueling Water Storage Tank being at the minimum volume allowed by plant Technical Specifications, and the incore room beneath the reactor is completely flooded. In addition, with a break resulting in these conditions, the plant EC.flow would be significantly less than the full ECCS flo eled in vortex testing.With implementation of the ap modifications, the minimum T(Refueling Water Storage ta Refueling Water Storage Tah+to the recirculation mode of E(of these changes in ,ST se sump pool volumes.e.' During vo. ing, the t three ý theTo volume for the the I-l re transferred e ~uced. Both ute t creased below for clarification of vortex ncnd the as-built plant configuration.
wnFovided in the CNS GL 04-02 Supplemental 29/08Vere based on the as designed configuration the minimum clearance allowed between the top of V the bottom of the vortex suppression grating. The*d below refer to the as-built configuration.
Attachment 3 Page 27 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Vortex Suppression Grating (1%" X 5/64")HHHHHHH HHHHM'Vortex Test Pool Level ---_¥_ 13" 1%" gap Vortex Suppression Grating (19SW-4, 11/4" x 3/16")554' 61/2" (Unit 1)554' 7" (Unit 2)Strainer Module Submergence Plant Conditions
- f. Vortex testing was performed with a submergence level consistent with Figure 12S-1. Vortices were observed at flows of 0.04 feet per second and above without the vortex suppressor installed.
With the vortex suppressor installed, no vortices were observed up to flows of 0.09 feet per second (approximately 2 times the maximum approach velocity).
Attachment 3 Page 28 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C13. Please provide a response to the question from the NRC Content Guide sent by letter dated November 21, 2007, relating to Enclosure 2 of the supplemental response sent by letter dated February 29, 2008, Section 3(f)(5), regarding the ability of the strainer to accommodate the maximum potential debris volume. This response should apply specifically to the Catawba strainer and not be a generic answer.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time. This draft response differs from the one discussed on that call, but is aligned with more recent discussions with the technical staff in spring 2010.As sttdin Catawba GIL 2004-02 Supl tal Ie onstated/3/2 08 Catawba predicted 101.3 Wt of Low i (LDFG) 11SI tion to be transported to the ECCS sum sie.Su-,qt t ta subm~it~l, Catawba has removed significant amoursýb i us insl ti from containment, discontinued reliance on WCAP- e.t Imp ent Testing to Determine the Zone of Influence I) of Min nd In tion for Wolf Creek and Subsequelnt N submit Clantsawb d ud eoycputational Fluid Dynamics analy ýtheip uaroved erS quantity foragement modifications.
These changes retshoding tra scirted LDFG insulation volume of approximatel 55 ft3 f this submittal for details regad, DF ne xpths i The n DFG deli' inldexecte to transport to the strainer (and their charact iss) were pi v le itCatawba GIL 2004-02 Supplemental Response a d2/29/20 -nclo 2, items 3(c)2, 3(d)3, and 3(h)6.Subsequent t ' um ,;i atawba adutdthe uqaiedepoxy coatings failure rate and ti the Ulate debris quantity for conventional debris head loss testing, using iq nethodology described in the McGuire Nuclear Station GL 2004-02 mental Response dated 2/28/08, Enclosure 2, item 3(h)(5), and further described/justified in the McGuire Nuclear Station response to RAI #26 from NRC RAI letter dated 11/18/2008.
The appropriate values are provided in Table 13S-1 below for convenience, along with the equivalent volume conversions for each debris type.Attachment 3 Page 29 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 13S-1 Non-LDFG Debris Quantities Transported to Catawba ECCS Sump Strainer*Debis ypeDebris Type y Debris Quantity at Debri TypeDensity~
Strainer Qualified Epoxy 118 Ib/ft3 155.8 lb Coatings (5D ZOI)Unqualified Epoxy 94 lb/ft 3 216.7 lb Coatings I Unqualified Alkyd Coatings Latent Dirt/Dust Latent Fiber (lint)169 lb/ft 3 Miscellaneous Latent Debris** NA (tags, labels, etc.)* Note: Destroyed stainless steel RMI** Note: Miscellaneous latent debris the approach velocity, but has an insig ar to no(jýsport to the ECCS Sump Strainer.redu shtrainer flow area for maximizing
ý al ,-d umu p ribution.Thus, the total voW n O is (LD a eeced to be transported to the ECCS sump ner afe limitin'g, is .p imately 72.25 ft 3 (the sum of the debris volum the ve tabl ard the bounding transported LDFG insulatio~etota eis volume of the limiting (Unit 2)Catawha"'SvL eb!Durn ident.con s, th bris be will initially accumulate non-uniformly on the s r. The app ch v " will vary across the individual strainer top hats and a the array sed o ,e location of the top hats relative to the recirculation s n piping ocally, the debris bed will build axially from the top hat base plate o the f end, up to the maximum debris load. However, since the total tran ,bris volume (as demonstrated above) is not sufficient to complete lyfltthe strainer interstitial volume, the strainer surface will retain its complex shape (multiple top hat cylinders with flow paths outside and inside the cylinder) and flow area. With no decrease in top hat module flow area, the evaluated approach velocities remain bounding.Attachment 3 Page 30 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C14. Please provide information that verifies that the debris preparation and introduction methods used during the array test and IPT were prototypical or conservative with respect to the transport evaluation for the plant. In general, protocols for fibrous debris preparation result in debris that is coarser than predicted by the plant-specific transport calculation.
In addition, the NRC staff has noted that debris introduction frequently results in agglomeration of debris such that it may not transport to the strainer prototypically or create a prototypical debris bed. Both of these issues can result in non-conservative head loss values during testing.The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Protol Test (IP d ar e ced in the question are being replaced b atory r-I See the response to RAI #11 of this s ittal fo etailec sion r-ding the associated Duke/NRC interfac d th rrespo g development of the CIT.The 30_d checIW n rm fall 2010, is designed in accor A0 1eview Guidance Regarding Gen etter 200 CIO in the of Strainer Head Loss and Vortexing" and the experic des d previously.
Debris preparation and introducti
'1 be consi .t wita hin-bed protocol as described in the staff guidance.
nrýntional is transport will be optimized by performing a system verifica .,est pri o formal testing. Debris settling in the test tank will be minimized by i nical agitators, which in previous testing (June/November 20 e proven to be effective in this regard.Attachment 3 Page 31 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C15. Please provide information on the flow fields in the array test. The NRC staff is concerned that non-prototypical debris distribution may have occurred during testing as a result of stirring of the tank. Stirring can result in the transport of debris that would otherwise not transport, or result in debris being washed from the strainer screen surfaces.
Either of these phenomena can result in reduced (non-conservative) head loss values during testing.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test array te ferenced in the question are being replaced by the Confirmato\[nt d Tes .. See the response to RAI #11 of thssubrnt-if)r a deta e -cthis egq, g the associated Duke/NRC interfa,,es d orel ettohteCt The 30-day CITS-ndI ý e in fall 01"0, is designed in accordance with th= ýI ý.d "NC eview Guidance Regarding Generic-0.CAre Strainer Head Loss and Vortexing" and theuse- hscribIe~l r 0 1ly. Debris preparation and intro c on will be *ste *th thin -d protocol as described in the staff guidan ,onvention51te)r"r will be optimized by performing a system ve ' tion test p to fo I testing. Debris settling in the test tank will be minimize using me nical agitators, which in previous testing (June/Novemb
- 09) ha proven to be effective in this regard. The test strainer array will rol d1 from non-prototypical flow fields by a baffle plate positioned between e m n test tank volume and the array.Attachment 3 Page 32 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C16. Please provide information that verifies that the debris preparation and introduction methods used during the thin bed testing for the top hat strainer design were prototypical with respect to the plant-specific debris generation and transport evaluation for Catawba. Note that for thin bed testing, the NRC staff considers it prototypical or conservative for fine fiber to arrive at the strainer prior to less transportable debris. Overly coarse debris preparation or nonprototypical introduction to the flume may non-conservatively affect the potential for thin bed formation.
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated p ren , te in the question are being replaced by3the ror RAI #Pagef33hof 60 RI#1othss!ýi a11rqdetailýý( 0 rQ-gdig the associated Duke/NRC interrff and tht oresp 1 "d ceve bae'nt of the CIT.The 30-day CIT, sct ed erforr(T iin fall 2010, is designed in a e 'Ndk 2St eview Guidance Regarding Gene.x6 L 0sr nt t Strainer Head Loss and Vortexing" and 2 Stpf an ting epT Csbdprt usly.-.Debris preparation and e sib ebJ c o introd wl n thin-bed protocol as described in the staff guidance., entional\
- tr ort will be optimized by performing a system verf test pri ) ormal testing. Debris settling in the test tank will be minimizec bIng nical agitators, which in previous testing (June/November l.4 proven to be effective in this regard.Attachment 3 Page 33 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C17. Please provide the criteria used to judge that differential pressure-induced effects (e.g., boreholes) did not occur during testing. The existence of pressure-induced effects could invalidate the application of temperature scaling. Please state whether pressure-induced effects were identified and, if so, the resultant effect on the application of temperature scaling.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.mI The Integrated Prototype Te st being replaced by the RAI #11 of this submittal for a c Duke/NRC interfaces and the c The 30-day CIT, sc1 the post-test cond and flow sweeps.ýflowrate first, less p(head los " cc sweeI After ýtest tank the debNsý ed will b Review GU ce R Strainer Head ss) and a3 erenced inýuestion are.tegrate (CIT). See the response to l cussio0 6arding the associated o'j devel nt of the CIT.rme 01 designed to assess oI series of temperature nmp re of the t6st tank pool at constant)r del eibed borehole development while olle ,e. Subsequent to the temperature ide additional correlation data.oted -erential pressure related anomalies in accordance with the March 2008 "NRC Staff Letter 2004-02 Closure in the Area of Attachment 3 Page 34 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C18. Please provide the scaling parameters used for calculation of debris quantities and strainer approach velocities used during testing. Please state whether the scaling accounted for strainer areas blocked by miscellaneous debris such as labels and tape.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IPT) and ýtest renced question are being replaced by the Confirmry Integrae,-lT IT). See th sponse to RAI #11 of this submittal for a~e~dM the associated Duke/NRC interfaces and thes' e lng de heCT The 30-day CIT, scheduled to be" 2 is designed accordance with. thl _a 008 ' t'f ",tv G lnce Regarding Generic Letter 2 d Cal L e in th- S nHead Loss and Vortexing" and the testing e ")in c ;sribed I': V_ sly.The test tan for the be a dified version of the test tank used for the sHe the Chemical Precipitates Head Loss ý eries ed iously. modification will result in an array with fewer at mdul red n from t e 2 x 3 test array (six top-hat modules)in the eae itests. Sin e C a combined test for Catawba and McGuire, the most c rvative de loa d strainer dimensions (adjusted appropriately e area oered by tags and labels) will be used to calculate debris loads an roac vlocities for the CIT and will be scaled to the modified test array Sace area. For scaling debris loads for the CIT, the predicted debris bed raction will not be included.Since the quantities of fibrous and particulate debris at Catawba are bounded by the quantities at McGuire and the McGuire sump strainers have less surface area than the Catawba sump strainers, there is additional conservatism/margin in the CIT for Catawba. Flow/temperature sweeps will be performed at the end of the 30-day testing interval to provide head loss correlations of this conservative debris bed for extrapolation to other flow conditions.
Attachment 3 Page 35 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C19. Please discuss the NRC staff's observation that in the IPT the flow was non-prototypically directed at the top hat.strainer in a direction parallel to the top hat long axis. Please address whether this non-prototypical flow direction could result in a non-prototypical formation of debris on the top hat strainer.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to. this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.m The Integrated Prototype TesW being replaced by the Confirm RAI #11 of this submittal for a d(Duke/NRC interfaces and the cc.ferenced inth'uestion are'CIT). See the response to rding the associated)Tt of the CIT.The 30-day CIT, accordance with Generic Letter 200 and the tes exr introdu b'guidan ...on,.syste ification be mini by u!(June/Nove r 2 strainer array positioned betw be pe I10,'ý designed in"NR( vie uidance Regarding in the of Strainer Head Loss and Vortexing" d pre isly. Debris preparation and 4i rotocol as described in the staff trans 11 be optimized by performing a mal testing. Debris settling in the test tank will i itators, which in previous testing rove to be effective in this regard. The test I from non-prototypical flow fields by a baffle plate test tank volume and the array.Attachment 3 Page 36 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C20. Please provide the clean strainer head loss for Catawba Unit 1 (only the clean strainer head loss for Catawba Unit 2 was provided).
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.\and 2 are provided below.The Catawba Unit 1 clean strainer head and configuration at 60'F, is calculateP recirculation flow condition.
The Catawba Unit 2 clean stra r head and configuration at 60'F, is cd recirculation flow condition.
It should be noted thas sub ECCS water
[vich ha relation to ECCS(0, this Ii se am system to operate v 1nly PUmP.through týd stara conse ,J e Lice mendme sch d for 1.installed strainer area'for the maximum finer area L'eseA' ndment Request for by he Commission.
In nt allo he Containment Spray will decrease the overall ECCS flowýclean strainer head loss will be Perein. Implementation of the 10 for Unit 2 and spring 2011 for Unit Attachment 3 Page 37 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C21. Please provide the time-dependent results and calculation methodology for determining net positive suction head (NPSH) margin throughout the 30-day mission time.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.Reference the response to information.
Attachment 3 Page 38 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C22. Please provide the basis for the debris introduction information that indicates that no 'fine fibrous debris would be generated during a loss-of-coolant accident (LOCA). If the assumption of zero fibrous debris generation is in error please provide the amount of fibrous debris generated by the limiting break and justify why, in such a case, the head loss test results would remain valid.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.Please reference the response to RAI regarding Zones of Influence (ZOI) for L fiber size distribution for destro d LdEFG of the destroyed insulation.
Insulation, quantities Attachment 3 Page 39 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C23. Please provide the types and amounts of debris added to each test (Array and IPT) and include information on introduction sequence.
Please provide relevant test parameters such as temperature, debris introduction times, and flow rate for the Array and IPT tests.Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (IPT) and test renced question are being replaced by the Confirmaory Integra tIT). Seeth sonse to RAI #11 of this submittal for a* 4d dliscuJsi4 garding the associated Duke/NRC interfaces and the ( gdev 11,ment of the CIT.The 30-day CIT, scheduled to be ýe 2-,.l designed in accordance with theM4ý2008 "NR pta nceRearin Generic Letter 2 e in th S Head Loss and Vortexing" and the testing eý ýeec lesribed ) iusly.Debris sequnm fo~ ýiWllo tin-bed" protocol from the staffI In ucing all of the particulates first, follo,~vra d h i esi ýe ntal additions.
,5ll v i i'The c al addition I talK 1ce after CIT pool heat-up and conventional debris be ation, as i ac ..dium tetraborate and soluble aluminum, calcium, an o with WCAP-16530 guidance as modified by S1ron in ac, ancewihWA163gudneamofedb the Duke algor Boung chemical quantities will be used.The CIT pool temp ofile will follow the bounding post-accident plant sump pool profile, fro aaximum of 190'F to a 30-day ECCS mission time minimum of 90'F.CIT flowrate will be based on previous testing experience facilitating the formation of a bounding conventional debris bed. Flow/temperature sweeps will be performed at the end of the 30-day testing interval to provide head loss correlations for extrapolation to other flow conditions.
Attachment 3 Page 40 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C24. Please provide information on the amounts of debris that settled during testing for each test (IPT, Array, and Thin Bed). Note that Enclosure 1 of supplemental response dated February 29, 2008, stated that near-field settling was not credited during testing. However, the NRC staff observed significant settling during the IPT. Please provide a quantitative evaluation of how this settling affected head losses for each test. Please state whether this settling is prototypical of plant conditions and provide a basis for the conclusion.
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The IntegratedPrototype Test ( pr uytes.Deb setn in the question are being replaced by3tll tto RAI #11 of this S60f- \rdrg the associated Duke/NRC &ý n horresp n indevel ent of the CIT.The 30-day CIT, sct d prfr in fall 2010, is designed in accorda e C St eview Guidance Regarding anGene Lst ex 1b a ed 1 p Strainer Head Loss and Vortexing" an ~ ~pting ex ec dsq r usly. Debris settling in the test tank will be i J ized by usi 111ck cal agitators, which in previous testing (June/No e 2009) 1 prýnto be effective in this regard.Attachment 3 Page 41 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C25. The supplemental response stated that the head loss across the Catawba Emergency Core Cooling System Sump strainer (clean strainer head loss plus debris bed head loss) is conservatively predicted to be 5.4 ft at switchover to sump recirculation.
However, no explanation was provided as to how this value was derived. It appears that credit was taken for time-dependency in head loss, since the 30-day value is 8.2 ft. Please provide the time dependent results and calculation methodology for determining NPSH margin throughout the 30-day mission time.Catawba Response:
/ />The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Prototype Test (I being replaced by th irmat RAI #11 of this SU 'rdE Duke/NRC interf arend t 'o The 30-day CIT, scl conservt[functi e head o results w recircu ipumpC throughot CC in d test efnced in the question are*eg s (T See the response to S ar aTg the associated gdevel ent of the CIT.for I in fall 2010, is designed to provide a 1cos e Catawba strainer array as afin al post-processing.
The strainer eý d vailable NPSH at the ECCS sump e requirements of the pumps are met tirii Attachment 3 Page 42 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C26. Please state whether the containment cleaning actions described in Duke's response to Bulletin 2003-01, sent by letter dated August 7, 2003, will remain in effect at Catawba (in order to assure that debris source assumptions made as part of the GL 2004-02 resolution remain valid).Specifically, please identify the procedures which control the cleanliness actions for containment and any commitments regarding the long-term applicability of these procedures.
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.As identified in the Catawba GL 2004-02 le I Respo ated 2/29/08, item 3(i)(1), Catawba has i mented p atic controls ensure that potential sources of debris " be in dced into containment will be assessed for adverse effects o S and ainment Spray recirculation functions.
The programmatic prac relating to containment cleanliness include,* Containme eaning i\conductd or to M e4. Extensive containment cleaning is co ted g water Sy. In general, washdowns are limited to the i on th uld be submerged under large break ns. sib i 11 surfaces and mechanical equipment wvashe ý L ed was p is are performed as directed by Radiation tection. Con me anliness currently verified prior to entry into Mode 4 rocedure 0 )A/A 001 -Controlling Procedure For Unit Startup." Visua inections o cont 'flment sump screen area are performed during a refuelin d tage in Ides 5 and 6 in order to evaluate sump availability.
Catawba 0 Co Directive
3.5 currently
describes the process and expectations s inspections, which are performed in accordance with PT/1 (2)/A/4400 Unit 1(2) Containment Building Civil Structures Inspection.
PT/O/AI4200/002
-Containment Cleanliness Inspection is performed prior to Mode 4 to fulfill Catawba Technical Specification (TS) Surveillance Requirement (SR) 3.5.2.8. Catawba TS SR 3.5.2.8 requires that the ECCS sump be visually inspected every refueling outage in order to verify there are no restrictions as a result of debris, and no evidence of structural distress or abnormal corrosion present prior to declaring the ECCS sump operable.* Refueling Canal cleanliness is currently verified during Mode 6 or No-Mode by an inspection controlled by procedure MP/O/B/7150/012
-Refueling Canal Cleanliness.
The procedure is not required by TS or Selected Licensee Attachment 3 Page 43 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Commitment (SLC), but is performed prior to the transition from Mode 6 to Mode 5 operations." In order to satisfy TS SR 3.6.15.1, the Refueling Canal Drain Valves are currently verified as locked open and unobstructed prior to entry into Mode 4 by procedure PT/1 (2)/A/4600/016
-Surveillance Requirements for Unit 1(2) Startup.* In order to satisfy TS SR 3.6.15.2, visual verification that no debris is present in the Refueling Canal or Upper Containment that could obstruct the Refueling Canal Drains is currently performed once every 92 days per procedure PT/1 (2)/A/4600/003B
-Quarterly Surveillance Items.* Upgrades to existing foreign material coni accountability logs to be maintained in Mode and out of containment.
These contro Catawba Site Directive 3.1.2 -Access High Pressure Steam Relief Devices.* Prior to establishing containme grity a made after containment integrity i tablishe ensures that a visual inspection is p med debris inside containment As identified in the Catawba GL Supple item 3(i)(3), Duke's modification es procedure that direc) ei a mp the plant. This prd lcts the system interactio As pia i this e[ tn, thE consideration of an tentia verse e t with debris tran-s&od Daths o ith th cntain*ocedures require material iugh 4 for items carried into irrently implemented using b9uilding and Areas Having containment entries 16.6.1 currently ove any loose il Response dated 2/29/08, Kies an administrative tion gineering changes in' nges be evaluated fordirection to include regard to debris sources and/or ment sump.Attachment 3 Page 44 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C27. The revised "Content Guide for Generic Letter 2004-02 Supplemental Responses," sent by letter dated November 21, 2007, Section 3k, requests a summary of structural qualification design margins for the various components of the sump strainer structural assembly.
This summary should include interaction ratios and/or design margins for structural members, welds, concrete anchorages, and connection bolts as applicable.
Please provide this information.
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and the approach as outlined was dispositioned as"acceptable".
The following information is an expan of the data provi the February 29, 2008 Catawba response to the Secti infor tion req Figure 27S-1 and Table 27S-1 are repeated ,m the pre, es onse to aid eader. As can be noted in Figure 27S-1, dified st assembly is constructed of robust materials and is located e crane I in both Unit 1 and Unit 2.East Vortex Suppression Rack is not shown for clarity Plenums Flow* Plenums-. .C rossflow" ; iPlenum Top-hats ECCS/C'sN-Recirculation Piping West, w Water Boxes Figure 27S-1: Catawba Modified ECCS Sump Strainer Attachment 3 Page 45 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 27S-1 shows the design inputs for the Catawba modified ECCS sump strainer structural calculations, including the Top-hats, the Main Structure, the structure including the Wing Walls/Water Boxes, and the Vortex Suppression Rack.Table 27S-1: Design Inputs/Loads for Catawba ECCS Sump Strainer I Max SSE Vertical Acc.*Bounding top-hat lei Attachment 3 Page 46 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Tables 27S-2 through 27S-5 summarize interaction ratios and/or design margins for the various Unit 1 components of the sump strainer structural assembly including structural members, welds, concrete anchorages, and connection bolts.Attachment 3 Page 47 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 27S-~ 3Anj~ysis of S m~ra-Iner futr WnPemsaA grBkS Water Box Evaluation Component Description Measurement Actual Allowable Comments Top Horizontal 3/8" Plate Max Stress 5624 psi 16875 psi 3/8" Diameter Studs Max Interaction 0.15 1.0 Ration (IR)Front Vertical 3/8" Plate Max Stress 12448 psi 16875 psi Acceptable as 3/8" Diameter Studs Max IR bounded by Top Horizontal Stud_values Back Vertical 3/8" Plate Max Stress 13 1 16875 psi Acceptable as 3/8" Diameter Studs Max IR bounded by Top Horizontal Stud___values All Members MFaulted allowable used 1/2" Diameter Studs Max IR IR_- 010 1/2" Anchor Bolts M iMa/ I P 1.03 s Base Plate \ax Steý,1 173ps dlen Wing Mdules Evaluation Componenrtrpe
/easurment Actual Allowable Comments 3/8" PlateIof u Mlaxe S/ J 6869 psi 17813 psi All Memb es es 0.087 1.0 3/8" Diam\ete Studs Max IR 0.47 1.0 1/2" Diamete "Sd M R 0.61 1.0 Anchor Bol Max IR 0.42 1.0 Component DescP r i/easurement Actual Allowable Comments 3/8" Plate of Tube Moe Max Stress_ 6869 psi 17813 psi All Members Max Stress IR 0.246 1.0 3/8" Diameter Studs Max IR 0.22 1.0 1/2" Diameter Studs Max IR 0.79 1.0 Anchor Bolts Max IR 0.42 1.0 Attachment 3 Page 48 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 27S-4:~ Analysis oftlSump Strainer Structure Excludinq Wing Plenums and Water Boxes Plenum Segment Component Description Measurement Actual Allowable Comments Max Interaction 0.575 1.0 All Members Ratio (IR).-.., Max Stress 11830 psi 16875 psi -1/2" Diameter Studs Max IR 0.51 1.0 -1/2" Diameter Hex Bolts Max IR 0.63 1.0 -1/2x10x15" Plate Bending Stress 16262 psi 16875 psi -Bending Stress 9096 psi 13500 psi -Tube Steel 3x3x1/4" 0 Axial stress is Shear Stress 9000 psnigiicn/ ý, insignificant Weld Metal 1.0 Weld at Base Plate Stress 17250apiBase Metal ,Faulted Shear 1.057 1.3 allowable used 1/2" Diameter Anchor Bolts Max I1{ ," 0.41 " !.0 Flow Plen aluti Component Description earement ct al Allowale Comments 3/8" PlateS 2-016 psi 16875 psi All Members l , 1.0 1/2" Diameter Studs Max R 079 1 0-J 1/2" Diameter Hexito Max Acua 1.0 1 P t Stress 15440 i 16875 psi 1/2"Plat etaq 0.107 1.0 Weld atF se aBase e 0.177 1.0 Max Sheairý ýess 17250 psi 20250 psi Faulted Base Pla allowable used R 0.62 1.0 Cross Flow Plenum Evaluation Component Desc t /easurement Actual Allowable Comments 17.75x45" Plate Stress 20164 psi 21375 psi Faulloaleued 3/8" Diameter Bolt Max IR 0.79 1.0-Miscellaneous Evaluation Component Description Measurement Actual Allowable Comments Stress 7281 psi 16875 psi-Plenum End Cover Plate MxTnin 375 Ibs 1546 Ibs Shalodi Load insignificant Max IR 0.25 1.0-Attachment 3 Page 49 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Tabl-e 2ýS-53 Analysis of Vortex Suppr~essio~n Rack ~Grating Evaluation Component Description Measurement Actual Allowable Comments Grating Max Load 79,5 psf 318 psf 1/4" Diameter Stud Shear Load 145 lbs 229 lbs Cut Grating (Bearing Bars) Max-Interaction 0.91 1.0 Normal allowable Around Interference Ratio (IR) used Plate Max IR 0.65 1.0Faulted allowable used Supporting Grating Panels Bending Stress 1800 17813 psi Shear stress is_____ ___________
insignificant Weld Allowable Load Welds Base Metal -ShearL0oad.91231 0baute P3/"e nBcum ket ament EvaluatiR7 Component Description Measurem t AAc .IA Comments 1/2" Diameter Bolt Bolts MaxI 154601bs Component Desc Measx l La Allowable CommentsIntera t Faulted and All Members ato 11.0: 1.3 normal ation(1f _allowables used" alowbleFaulted 5/8" Dteela 2Se 0.55 1.0 isgfcant 5/8(c )0.79 1.0 M~R -0.758 1.0 WeldsM ýrs Mlaxl I Base 09210Faulted Metal Shear _ 09210allowable used 3/4" Bracket Plat Max IR 0.77 1.0-1BaePaeMax Bending 0.76 1.0-1" Bas PlateStress I R_3/4" Diameter Anchor BoMax IR 0.88 1.0-1 " Diameter Anchor Bolts Max IR 0.85 1.0-TbSte xx/ Max Bending 0.610Shear stress is Stress IR insignificant (continued)
Attachment 3 Page 50 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11 /21/08 NRC Request for Additional Information Draft Response Table 27S-5 (continued)
East Wing Rack Evaluation Component Description Measurement Actual Allowable Comments sMax Interaction Faulted and All Members 0:853 .1.0 normal Ratio (IR) allowables used 5/8" Diameter Bolts Max IR 0.32 1.0 Faulted allowable used 5/8" Bracket Plate Max IR 0.72 1.0 Max IR -Weld Welds Metal Stress 0.753 1.0 Max IR -Base 1.0 Faulted Metal Shear allowable used 1" Base Plate Max Bending 1a"Bo eStress IRu 3/4" Diameter Anchor Bolts Max IR/ 0.91. 1.0 Vortex Supprk /aluation\
Component Description Measuremnenl',ýAta AloW CommentslnFaulted and All Members M0. 1.0 normal allowables used WeldsMea _b M!ax IR. 1.01ý(5/8" t ltFaulted 5/8"DiaeterBols R _ .421.0allowable used Ma2Tc 0.9 Faulted 0.91 1.0allowable used 6/8" BfN.k t Plate N-, ' a xlIR 0.99 1.0 Angle 2x2 " 1s 12955 psi 14250 psi 1" Base Plate Max Bending 0.57 1 0 Dameter AhStress IR 0.6 1.0 3/4" Diam;neter AnctMxI 0.96 1.0 Attachment 3 Page 51 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Tables 27S-6 through 27S-9 summarize interaction ratios and/or design margins for the various Unit 2 components of the sump strainer structural assembly including structural members, welds, concrete anchorages, and connection bolts.Attachment 3 Page 52 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response'Table 27S-7. Analysis of Sump Strainer Structure Wing ,Pjnum~s and Wt oes Water Box Evaluation Component Description Measurement Actual Allowable Comments Top Horizontal 3/8" Plate Max Stress 5624 psi 16875 psi -3/8" Diameter Studs Max Interaction 0.15 1.0 Ration (IR)Front Vertical 3/8" Plate Max Stress 12448 psi 16875 psi Acceptable as 3/8" Diameter Studs Max IR bounded by Top Horizontal stud values Back Vertical 3/8" Plate Max Stress si 16875 psi Acceptable as 3/8" Diameter Studs Max IR -bounded by Top Horizontal stud values All Members Max Stressl 0.76 1/2" Diameter Studs at aIR P Bottom PlateduRes ate Bx El Welds eas 6869 s 171 psi -trssR 0-8 1.s0 e sýa0.861 1.0 1/2" Anchor Bolts Max' 1.0-Base Plate ax s 17813 psi-in Modules ofNater Boxes Evaluation Compone hDRpeiont Actual Allowable Comments 3/8" Pate f Tube Mdl Max Stress 6869 psi 17813 psi Memb s Max Stress IR 0.087 1.0 3/8" Diariet -Studs Ma-xIR 0.47 1.0 1/2" Diamee Sts MaII)R 0.61 1.0 -Anchor Bolts Max IR 0.42 1.0Modules of Water Boxes Evaluation Component Descrip 1 n Measurement Actual Allowable Comments 3/8" Plate of Tube Module' Max Stress 6869 psi 17813 psi Members Max Stress IR 0.246 1.0 3/8" Diameter Studs Max IR 0.22 1.0 1/2" Diameter Studs Max IR 0.79 1.0 Anchor Bolts Max IR 0.42 1.0 Attachment 3 Page 53 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response fable 27S-8: Analysis of Sump Strainer Structure Excluding Wing-Plenums and iWater Boxes~Plenum Segment Component Description Measurement Actual Allowable Comments Max Interaction 0.575 1.0 Members Ration (IR)Max Stress 11830 psi 16875 psi -1/2" Diameter Studs Max IR 0.51 1.0 1/2" Diameter Hex Bolts Max IR 0.63 1.0 Other stresses 1/2x10x15" Plate Bending Stress 16262 16875 psi .are acceptable since loads are small.Bending Stress r 1 13500 psi f all Tube Steel 3xx/4" I Axial stress is Shear~ Stes -4600 psi ps'Copnn Derer M es 90t as insignificant W eld M tr .4 \
W eld at Base Plate Ba e M tl.'Since IR > 1.0, Baehetar 1, ý faulted allowable--used 1/2" Diameter Anchor Bolts IR0.81.0 CrossaFlow Plenumn Evaluat-ion Componi Measurc Allowable Comments 3/84" Plate Max Stres 26b4 pýs 16875 psi Members " Max IF 0.z10 1/2" Diameter Blt !Max IR 0.17 1.0 1(onDiameeed 15440 psi 167ps t xaeF 0.107 1.0a 0.177 1.0 Base Plata Max Stress 18864 psi 20250 psi allowable used SMax IR 0.931 1.0 .5/8" D iam eter A nchor 3oltM ax IR 0 .8 1 .0 Cross Flow Plenum Evaluation Component Description Measurement Actual Allowable Comments 17.75x45" Plate Max Stress 20164 psi 21375 psi Faulloateued 3/8" Diameter Bolt Max IR 0.79 1.0-(continued)
Attachment 3 Page 54 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Table 27S-8 (continued)
Miscellaneous Evaluation Component Description Measurement Actual Allowable Comments Max Stress 7281 psi 16875 psi Plenum End Cover Plate1.0 Attachment 3 Page 55 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Attachment 3 Page 56 of 60 Catawba. Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response Attachment 3 Page 57 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C28. Please describe the basis for concluding that there is no potential of debris blockage at the Ice Condenser drains and refueling canal drains for accident scenarios where containment spray is necessary.
Catawba Response: The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "acceptable" at that time.This draft response is consistent with the one discussed on that call.In addition to the programmatic controls canal cleanliness described in the Ca submittal, Catawba has implemente regarding Ice Condenser cleanliness: " The Ice Condenser is i nected for from Mode 5 to Mode debri, by the Containment Sur r" In order to fujTTchnical ecifi (SR) 3.6. 15dense o every 18 rwýisdcung~shutdo verified for t Co nser floe I)ntainment and refueling I tRAI #26 of this rammatic controls he transition is evaluated Mode 4.no ci .TS)'uveilance Requirement d vehrified as operable once lfour he following items must be rains to be considered operable: paired by ice, frost, or debris vidence of damage ioes not exceed a prescribed value nser floor to the lower compartment is force c)ndE The programmat'frbntrps revent debris blockage at the Ice Condenser and refueling canal drai n' o an accident.
In order to reach and potentially block the refueling canal or Condenser floor drains, any debris created during an accident in lower containment would have to be carried by steam into the Ice Condenser and navigate a torturous path through the Ice Condenser baskets;this is not a credible scenario.Attachment 3 Page 58 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C29. The NRC staff considers in-vessel downstream effects to not be fully addressed at Catawba, as well as at other pressurized-water reactors.
The supplemental response for Catawba refers to the evaluation methods of Section 9 of Topical Report (TR) WCAP-16406-P, Revision 1,"Evaluation of Downstream Sump Debris Effects in Support of GS-191,"for invessel downstream evaluations and makes reference to a comparison of plant-specific parameters to those evaluated in TR WCAP-16793-NP, Revision 0, "Evaluation of Long Term Cooling Considering -Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." The NRC staff has not issued a final Safety Evaluation (SE) for TR WCAP-1T3NP.
The licensee may demonstrate that in-vessel downstream effe ues are resolved for Catawba by showing that the licensee's p onditions are bounded by the final TR WCAP-16793-NP and the c/ and limitations identified in the final NRC staffs SE. The license "nay a solve this item by demonstrating without reference " AP167 the NRC staffs SE that in-vessel downstream effec ye been addres t Catawba. In any event, the licensee should re low it has addres the in-vessel downstream effects issue within 90 i t of* snce of th l NRC staff's SE on TR WCAP The N is developing, egulatory Issue Summary to infor1 s' iStry of IRC staff's expectations and plans regarding resolutionthihainig ctof GS1-191.Catawba Repose The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "dependent on resolution of WCAP-16793-NP in-vessel downstream effects issues" at that time.This draft response is consistent with, the one discussed on that call.Catawba Nuqc tation address the in-vessel downstream effects issue within 90 days o 4sqanc the staff's final Safety Evaluation on TR WCAP-16793.Attachment 3 Page 59 of 60 Catawba Nuclear Station Generic Letter 2004-02 Supplemental Response 11/21/08 NRC Request for Additional Information Draft Response C30. Please discuss why the Integrated Prototype Test (IPT) provided a representative debris bed on the top-hat strainer module for filtering chemical precipitates.
The NRC staff observed the debris addition video and concluded that the fibrous debris introduced into the test tank was more agglomerated than what may arrive at the strainer under post-LOCA flow conditions in the plant. Is the amount of bare strainer area observed in the test representative of what is expected to occur with the plant strainer array if a large break LOCA were to occur? The use of chemical effects test results derived from a test which formed a non-prototypically partially clean screen fiber bed would not be appropria Catawba Response-The response methodology for this RAI question was discussed with the NRC technical review staff during the scheduled September 1, 2009 Duke public telecon, and was dispositioned as "related to future planned confirmatory integrated head loss testing" at that time. As discussed in the introductory remarks to this draft RAI submittal, confirmatory testing is still in progress.
The specific RAI question response is aligned with more recent discussions with the technical staff in spring 2010.The Integrated Pr tO n (IPT) a e in the question are being replaced b Confi tory In d Tes'T). See the response to RAI #11 of this subl I fo etailed ussion regarding the associated Duke/NRC*
ateces," tjý con&ondl development of the CIT.TheDUe/R 3C.ce pIDe° ,eveopmen of the Cfall 2010, is designed in acco ce with th ch 'NRC Review Guidance Regarding Generil ter 2004-0 sur he Area of Strainer Head Loss and Vortexing" and the te .experien escr previously.
Debris settling in the test tank will be minin i by using rchanical agitators, which in previous testing (June/Novemb
- 09) ha' proven to be effective in this regard. As expected, debris bed forma nur e 2009 test series also tended to be more uniform.Attachment 3 Page 60 of 60