LIC-15-0100, Updated Status of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance

Updated Status of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance

ML16006A048
Person / Time
Site:	Fort Calhoun
Issue date:	01/04/2016
From:	Cortopassi L Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-2004-02, GSI-191, LIC-15-0100
Download: ML16006A048 (43)

Text

Omaha Public Power District LIC-1 5-0100 January 4, 2016 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 1 Fort Calhoun Station (FCS), Unit 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

Updated Status of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. I

References:

See Enclosure 1 On June 29, 2015, a teleconference between NRC staff and representatives of the Omaha Public Power District (OPPD) was held to discuss closeout of GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" for Fort Calhoun Station, Unit No. 1. OPPD was requested to review and update information that OPPD informally transmitted to the NRC Project Manager (PM) in April 2010 (Reference 4) and updated in October 2010 (Reference 6).

The background for this request is as follows. The information provided in April 2010 was in response to a February 2010 NRC draft request for additional information (RAI) (Reference 3) regarding OPPD's supplemental responses to GL 2004-02 submitted February 29 and October 16, 2008 (References 1 and 2 respectively). OPPD's response to the NRC RAI was informally transmitted in April 2010 (Reference 4) and again in October 2010 (Reference 6) following a June 2010 teleconference (Reference 5).

As requested during the June 29, 2015 teleconference, Enclosure 2 is submitted to formally docket the information emailed to the NRC PM in April and October 2010. The information contained in Enclosure 2 has been updated due to the length of time that has transpired since this information was provided to the NRC in 2010.

OPPD's last docketed correspondence concerning GL 2004-02 was submitted in May 2013 (Reference 7). Enclosure 3 provides a brief update of information provided in Reference 7 and Enclosure 4 is an update to the list of commitments made in Reference 7.

Any other actions discussed in this submittal represent intended or planned actions by OPPD, and are provided for information only; they are not regulatory commitments.

444 SOUTH 16TH STREET MALL ° OMAHA, NE 68102-2247

U. S. Nuclear Regulatory Commission LIC-1 5-0100 Page 2 If you should have any questions regarding this submittal or require additional information, please contact Mr. Bill R. Hansher at 402-533-6894.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 4, 2016.

Louis P. Cortopassi Site Vice President and CNO LPC/JKG/mle

Enclosures:

1. Reference List

2. Fort Calhoun Station, Unit No. 1 Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Issue (GSI) 191, Pressurized Water Reactor Sump Performance

3. Update of OPPD Letter LIC-1 3-0058 and Revised Regulatory Commitments

4. Regulatory Commitments c: M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector

LIC-1 5-0100 Page 1 Reference List

1. Letter from OPPD (Richard P. Clemens) to NRC (Document Control Desk), "Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated February 29, 2008 (LIC-08-0021) (ML080650369)

2. Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk), "Fort Calhoun, Unit 1 - Notice of Completion of Corrective Actions taken in Response to Generic Letter 2004-02 and Response to Request for Additional Information (RAI) for Fort Calhoun Station (ECS) Unit No. 1," dated October 16, 2008 (LIC-08-0107) (ML082960244)

3. Letter from NRC (Lynnea Wilkins) to OPPD (D. J. Bannister), "Fort Calhoun Station, Unit No. 1 -Request for Additional Information RE: Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors (TAC No. MC4686)," dated February 12, 2010 (NRC 0011 ) (ML100150072)

4. Email from OPPD (Donna L. Lippy) to NRC (Lynnea Wilkins), "OPPD Draft GL 2004-02 RAI Responses for FCS," dated April 30, 2010

5. Letter from NRC (L. Wilkins) to OPPD, "Summary of June 30, 2010 Category 1 Teleconference Meeting with Omaha Public Power District on Generic Letter 2004-02 (TAC No. MC4686)," dated July 29, 2010 (NRC-I10-0059) (ML101940301 )

6. Email from OPPD (Donna L. Lippy) to NRC (Lynnea Wilkins), "OPPD's Revised Draft Responses to the Generic Letter 2004-02 Requests for Additional Information (TAC No.

MC4686)," dated October 1, 2010

7. Letter from OPPD (L. P. Cortopassi) to NRC (Document Control Desk), "Proposed Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, "Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. 1," dated May 15, 2013 (LIC-13-0058) (ML13136A047)

LIC-1 5-0100 Page 1 Fort Calhoun Station (FCS) Revised Responses to Requests for Additional Information (RAIs) from Generic Letter (GL) 2004-02 Supplemental Response dated October 16, 2008 The Omaha Public Power District's (OPPD's) response to the Nuclear Regulatory Commission's (NRC's) draft request for additional information (RAI) transmitted on February 12, 2010 (Enclosure 1, Reference 3) is shown below. OPPD's response to the NRC RAI was informally transmitted in April 2010 (Enclosure 1, Reference 4) and again in October 2010 (Enclosure 1, Reference 6) following a June 2010 teleconference (Enclosure 1, Reference 5). As requested during a June 29, 2015 teleconference with NRC staff, this letter is provided to formally docket and update the information emailed to the NRC in April and October 2010.

RAI #3 Question:

The NRC staff requested that the licensee identify the source of the test data used to support the debris size distribution assumed for calcium silicate insulation and compare the banding, jacketing, and manufacturing process for the calcium silicate installed at FCS to the material used for the destruction testing. The licensee's response dated October 16, 2008, identified that the assumed debris size distribution was based on testing conducted by Ontario Power Generation, Inc. (OPG) and an analysis of this data contained in Appendix II to the December 6, 2004, safety evaluation (SE) on Nuclear Energy Institute (NEI) 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology" (the safety evaluation and NEI 04-07 are non-publicly available). The response also discussed the insulation jacketing and the manufacturing process for the calcium silicate base material. The staff did not consider the response to have fully addressed the RAI because the bandingllatching design of the FCS calcium silicate and the jacketing thickness were not described sufficiently and compared to the corresponding design values for the OPG test debris. Please provide this additional design information and comparison with the OPG test.

NOTE: Despite the discussion in Appendix II of the NRC's SE on NEI 04-07, based on scaling issues similar to those raised in the review of recent Pressurized Water Reactor Owners Group (PWROG) zone of influence (ZOI) testing, it is not apparent that the scaling of the OPG testing (i.e., primarily the scaling of the relatively small jet from a 2.86-inch-diameter nozzle to the large 48-inch-long target at the distances tested) provides prototypical or conservative results with respect to the characteristic debris size distribution that would be expected under plant conditions. The NRC staff has not yet determined the significance of this issue. Should it be determined significant, the NRC may raise this as a future issue with FCS testing in the resolution of GL 2004-02.

OPPD RAI #3 Response:

The FCS calcium silicate piping insulation is currently not installed in a manner that is equivalent to the configuration of the insulation that was used in the OPG testing. To demonstrate the equivalency, and to continue to use the approach discussed in the previous RAI response (i.e., , Reference 2), OPPD will be installing additional banding to calcium silicate insulation within the zones of influence which support the ECS debris generation analysis. The following

LIC-15-0 100 Page 2 table compares final FCS calcium silicate piping insulation jacketing and proposed banding to what was used in the OPG testing:

FCS { OPG Testing Jacketing Material Aluminum Aluminum Jacketing Thickness 0.020" 0.016" Banding Material Stainless Steel Stainless Steel Banding Thickness 0.020" 0.020" Banding Fastener Type Crimped Crimped Spacing of Banding < 8.25" > 6.5" & < 8.25" OPG testing of calcium silicate insulation used aluminum jacketing 0.016" thick. ECS aluminum jacketing is 0.020" thick, which is bounded by jacketing thickness used for OPG testing. OPG testing used stainless steel bands with a thickness of 0.020" secured by standard crimp connectors with a spacing ranging from 6.5" to 8.25". FCS will install additional banding material in a manner similar to the configuration in the OPG Testing with 0.020" bands spaced less than 8.25 inches which will bring the configuration at ECS to a similar configuration as that used in the OPG Testing. A significant amount of banding will need to be added in many locations that are in inaccessible, high radiation areas.

As noted in Reference 3, OPPD was requested to describe the bandwidth spacing selected for the modification. The modification (i.e., EC 49722) will:

1. install stainless steel bands on jacketed calcium silicate insulation in containment (within a prescribed zone of influence inside the SG bays),

2. replace existing fiberglass insulation on the 3" pressurizer spray line near emergency core cooling system (ECCS) strainers with reflective metal insulation (RM I), and

3. replace existing calcium silicate insulation containing asbestos on reactor coolant pump RC-3D with RMI.

Use of stainless steel banding on jacketed calcium silicate insulation allows the existing insulation and jacket to remain in place and eliminates the need to dispose of it. The banding will be placed at nominal 6-inch spacing, which is similar to the spacing used in OPG testing.

RAI #5 Question:

The NRC staff requested that the licensee address the potential for operation of a low pressure safety injection (LPSI) pump following the switchover to recirculation (either due to the potential single failure of an LPSI pump to trip or through procedurally permitted operator actions) and any consequent effects of this increased flow on post loss-of-coolant accident (LOCA) debris transport. By letter dated October 16, 2008, the licensee stated that a "turbulent jet analysis" had been performed, which showed no additional debris entrainment in the flow, but showed the potential for additional tumbling transport.

The licensee also indicated that the only scenario where the LPSl pump would be procedurally operated post-LOCA would be for alternate hot-leg recirculation, under which

LIC-1 5-0100 Page 3 conditions the sump flow rate would be less than the currently bounding analyzed flow rate. The NRC staff questioned the adequacy of this response because, for the case of a LPSl pump single failure to trip, (1) it was not apparent to the staff that a turbulent jet analysis is appropriate for modeling the flow in question, (2) sufficient detail concerning the turbulent jet analysis was not presented, (3) sufficient information was not presented to justify the unexpected conclusion that the additional flow from a LPSl pump would not lead to additional debris remaining in suspension in the flow to the strainers, (4) it is not apparent that the potential for additional debris to transport to the strainer via tumbling with the additional flow from a LPSl pump can be neglected, and (5) it is apparent that a break in the steam generator (SG) A compartment could not become more limiting if the additional pool turbulence created by the flow from an LPSI pump were to result in additional debris remaining in suspension. Please provide additional information that sufficiently addresses these remaining items.

OPPD RAI #5 Response:

During the June 30, 2010 teleconference, the NRC staff suggested that OPPD identify the expected time needed to secure the LPSI pump and provide a basis for the time chosen. Pending review by the NRC, OPPD will revise its position regarding the potential for increased debris transport due to an LPSI pump failure to trip. OPPD will no longer credit the turbulent jet analysis referenced in the PAI. As a result, the detail requested for this RAl is no longer applicable. The following discussion provides information for appropriate closure of this issue.

The Fort Calhoun Station (FCS) Emergency Operating Procedures (EOPs) are written in a two-column format, with instructions in the left hand column and contingency actions in the right hand column. If the instruction cannot be successfully accomplished, the operator immediately carries out the contingency action. FCS EOP-03's mitigation actions associated with the failure of a LPSI pump to trip post-RAS are in compliance with CEN-152, "Combustion Engineering Emergency Procedure Guidelines."

EOP-03, "Loss of Coolant Accident," R38 currently prescribes the following actions in response to initiation of recirculation:

EOP-03 Page 40 of 74 INSTRUCTIONS CONTINGENCY ACTIONS Sc32. IF SIRWT level falls to 16 inches, 32.1 IF RAS is NOT actuated by STLS, THEN verify that STLS initiates RAS by THEN perform step a or b:

performing the following:

a. IF any of the STLS relays have

a. Verify ALL of the following STLS NOT tripped, relays have tripped: THEN manually initiate using the AI-30A/B key on the STLS Test S 86A/STLS S

86B1/STLS Switches.

0 86B/STLS 86A1 /STLS S

• 86A/STLS Test Switch

• 86B/STLS Test Switch

b. Verify ALL. of the following RAS relays have tripped: b. Manually establish RAS flow path by performing the following:

0 86A/RAS S

86B1/RAS 0

86B/RAS 1) Open both SI Pump Suction S

86A1 /RAS Containment Isolation Valves.

• HCV-383-3

c. Verify BOTH of the following

• HCV-383-4 valves are open:

• HCV-383-3, SI Pump Suction 2) Clos.__e both SI Pump Suction Containment Isolation Valve

• HCV-383-4, SI Pump Suction S IRWT Isolation Valves.

Containment Isolation Valve

• LCV-383-1

• LCV-383-2 (continue) 3) Ensure both LPSI Pumps stop.

• SI-1A

• SI-lB (continue)

SContinuously Applicable or Non-Sequential StepR3 R38

LIC-15-0 100 Page 4 EOP/AOP Floating Steps, R7 contains the following guidance for LPSI stop and throttle criteria:

EOP/AOP FLOATING STEPS Page 25 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. IF RAS has initiated, 1.1 IF EITHER LPSI pump is still running, THEN ensure both LPSI pumps, THEN perform any or all of the SI-1A/B, stop. following steps a, b, or c:

a. Throttle LPSI Injection Valves to approximately 250 gpm flow.

b. (LOCAL) OQpen running LPSI pump breaker by performing the following (Switchgear Rooms):

1) Place the 69 permissive switch to "PULL TO LOCK":

• 69/SI-IA, "PERMISSIVE CONTROL SWITCH SI-IA BREAKER" (1A3-7)

• 69/S1-1IB, "PERMISSIVE CONTROL SWITCH SI-lB BREAKER" (1A4-14)

(continue)

2) IF SI-lA was running, THEN press the "OPEN" push button.

(continue)

R7

EOP/AOP FLOATING STEPS Page 26 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 .b. (continued)

3) IF SI-I1B was running, THEN press the "MANUAL TRIP" push button.

c. IF SI-IA breaker is to be racked down, THEN perform the following:

1) De-energize Bus 1A3.

2) (LOCAL) Remove dust cap and install jumper into breaker door receptacle.

3) (LOCAL) Install motor plug (continue) adapter and limitilt shorting plug into the junction box.

4) (LOCAL) Install elevator motor into cubicle.

5) (LOCAL) Place the elevator motor select switch to lower.

(continue)

R7

EOP/AOP FLOATING STEPS Page 27 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 .0. (continued)

6) (LOCAL) Pull clutch handle until elevator motor limits out.

d. IF SI-I1B breaker is to be racked down, THEN perform the following:

1) De-enerqize Bus 1A4.

2) (LOCAL) Install elevator motor in SI-lB cubical.

3) (LOCAL) Place elevator motor selector switch to "LOWER".

4) (LOCAL) Pull clutch handle until elevator motor limits out.

e. IF both Vital buses are deenergized, THEN GO TO EOP-20, Functional Recovery Procedure, MVA-AC.

R7

EOP/AOP FLOATING STEPS Page 28 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

2. IF RCS pressure is greater than 200 psia and controlled, THEN storp both LPSI Pumps, SI-IA/B.

3. IF RCS pressure is less than 200 psia, AND RAS has NOT occurred, THEN initiate LPSI flow by performing the following:

a. Start both LPSI Pumps, SI-IA/B.

b. Ensure all of the LPSI Loop Injection Valves are open.

R7

LIC-15-0 100 Page 5 EOP Floating Step B was revised to first throttle the LPSI valves (1.1 .a above) and then locally open the LPSI pump supply breaker by manually tripping the breaker and/or racking down the breaker (1.1 .b, 1.1 .c respectively). By throttling the LPSI valves, the total emergency core cooling system (ECCS) flow through the ECCS suction strainer is reduced to less than the strainer design maximum flow.

The recirculation actuation signal (RAS) is automatically generated by the engineered safeguards actuation system when indicated safety injection and refueling water tank (SIRWT) level drops to 16 inches. In addition to the operator monitoring SIRWT level following the initiation of the event, when level in the SIRWT drops to 16 inches, an annunciator on the control room panel will alarm, alerting the operator to this condition. Since throttling the LPSI valves is the first action the operator would take following the SIRWT level dropping to 16 inches, and these actions can be accomplished from inside the control room, these actions would be completed within a few minutes of an RAS.

During preparation of this response it was identified that EOP floating step 1.1 .a, throttling the LPSI loop injection valves, is an optional step. As it appears that this action should be mandatory, OPPD will disposition this issue in its corrective action program.

RAI #7 Question:

The NRC staff requested that the licensee provide further justification to support the assumption that only 5 percent of fine fibrous and particulate debris blown to upper containment would be washed down to the containment pool (e.g., by condensate drainage). The RAI stated that the staff considered 10 percent to be a more appropriate number and cited NUREG/CR-6762, "GS1-191 Technical Assessment Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance," dated August 2002 (ADAMS Accession No. ML022470077), which found that 5 percent was a low estimate for washdown without spray and 10 percent was a high estimate. In its response dated October 16, 2008, the licensee stated that the assumption of 5 percent washdown from condensate flow was based on assumed condensate flow from the volunteer plant study in Appendix VI, "Detailed Blowdownl~ashdown Transport Analysis for Pressurized Water Reactor Volunteer Plant," to the NRC SE on NEI 04-07. The response further discussed the lack of post LOCA containment spray operation and the effect of fan cooler operation. The licensee also stated that a washdown percentage of 10 percent for the debris loading in SG B compartment would be bounded by the more-limiting debris loading from SG A compartment (with a washdown percentage of 5 percent). Based on the factors below and engineering judgment, the NRC staff considers it prudent to assume a 10 percent washdown, consistent with the technical guidance discussed above.

a. The condensate flow from a plant without containment sprays would be significantly greater and longer-lasting than the volunteer plant, which would presumably lead to an increase in the transport fraction due to condensate drainage versus the volunteer plant.

b. Appendix VI to the safety evaluation, as well as the Drywell Debris Transport Study (NUREG/CR-6369) results on which many of its assumptions are based, indicate that substantial uncertainties are associated with the transport estimates, which applies

LIC-15-0 100 Page 6 particularly to blowdown and washdown. The washdown percentages in Appendix VI did not factor in uncertainties typically accounted for in licensing basis calculations and were based on the assumption that most containment surfaces would be sprayed, and few would be subject to condensate drainage only.

c. To support a washdown transport fraction that may not contain significant margin to account for uncertainties, it is important that the blowdown transport of fine debris to the upper containment be performed in a sufficiently conservative manner. Given that the FCS containment contains significant solid flooring, which could potentially impede its transport to upper containment, it is not apparent that the assumed quantity of fine debris (80 percent) would actually be capable of reaching the upper containment during the blowdown phase to subsequently be retained there.

d. Some of the fine debris assumed to be inertially held up on vertical surfaces or the underside of horizontal surfaces during blowdown will lose adherence over time and may gravitationally float down to the containment pool. It was not apparent that this phenomenon was accounted for in the 5 percent washdown assumption made by the licensee.

e. Although, in response to the NRC staff's concerns about washdown uncertainties, the licensee stated that acceptable strainer performance could be assured when assuming 10 percent washdown for a non-limiting break scenario, it was not apparent to the staff whether 10 percent debris washdown could be tolerated for the limiting strainer debris loading case.

The impact on the debris source term of doubling (for example) the assumed washdown percentage would be significant for FCS. Based on this consideration and the questions above, please provide additional information that addresses these remaining items or demonstrate that the strainer will perform acceptably if a washdown transport fraction for fine fibrous and particulate debris that incorporates adequate allowance for uncertainties is assumed.

OPPD's RAI #7 Response:

In the Enclosure to OPPD letter LIC-08-0107 dated October 16, 2008 (Enclosure 1, Reference 2), OPPD supplied a response to RAl #7. The response to RAl #7 described OPPD's basis for the conservatism of using a 5% washdown transport fraction for OPPD's postulated large break post-LOCA operation. OPPD's post-LOCA operation does not employ containment spray.

In its reply to the OPPD letter (Enclosure 1, Reference 2), the NRC staff provided additional clarification of their RAl and stated that the staff continues to consider it prudent to assume a 10%

washdown transport fraction, and requested that OPPO provide additional information to adequately address the staff's remaining items, or use a washdown transport fraction for fine fibrous and particulate debris that incorporates adequate allowance for uncertainties (i.e., 10%).

In lieu of development of additional analyses in support of the 5% washdown transport fraction, OPPD will incorporate a 10% washdown transport fraction into the currently in progress revision of the debris transport analysis for FCS. OPPD considers this item to be closed with no additional actions required.

LIC-1 5-0100 Page 7 RAI #10 Question:

The NRC staff requested that the licensee provide further justification to resolve the staff's long-standing concerns (including those raised during the staff's pilot audit of FCS) associated with the use of a Stokes' Law approach and turbulent kinetic energy (TKE) metrics to determine the settling behavior of fine debris. In its October 16, 2008, response to this RAI, the licensee provided an extensive discussion that responded to each part of the staff's question. However, after reviewing the response, the staff continues to have concerns that the licensee's approach has not been adequately justified on a technical basis. The NRC staff's concerns are summarized below. Please provide information to address these concerns

a. The licensee's TKE metrics do not appear to have been experimentally benchmarked.

Benchmarking would likely show that a single metric for what is in reality a distribution of particulate sizes to be insufficient. Acoustic measurements of velocity made at low flow rates are not directly related to the fundamental phenomena governing suspension and settling (phenomena which are associated primarily with turbulence).

Therefore, the NRC staff did not agree that such measurements would be sufficient to resolve this concern.

b. The licensee did not sufficiently address uncertainties in the TKE model(s) in the computational fluid dynamics (CFD) code used to provide confidence that the TKE model used was adequate for the purpose of determining the settlement of fine debris.

The licensee's October 16, 2008, response dealt primarily with uncertainty in velocity (both computational and experimental), not TKE.

c. The licensee did not adequately justify the specification of shape factors and drag coefficients that are applicable for modeling the settling behavior of spherical objects in quiescent fluid to model the settling behavior of irregularly shaped debris, including fibers, in the plant containment pool. Although a detailed discussion was provided in the licensee's October 16, 2008, letter in response to this point, the NRC staff does not agree with certain arguments being made. Specifically, the NRC staff does not agree that shape factors and drag coefficients are not relevant to the settlement of irregularly shaped debris. Sufficient information was not presented to justify this conclusion. It is not apparent that shapes formed by strands of fibers would behave similarly to a spherical particle. In addition, sufficient justification was not presented to show that the presence of turbulence would not affect the TKE metrics that were calculated for quiescent conditions. The staff also did not agree with the licensee's explanation for the drag force being insignificant for slowly settling fine debris because it is the significance of drag (or the drag coefficient) for the fine debris (relative to the other forces acting on the fine debris) that leads to the very small settling velocity.

d. The licensee did not sufficiently justify the correlation of terminal settling velocity to TKE. In response to this item, the licensee indicated that the "TKE" used by Alion Science and Technology (Alion) was based on a bulk velocity, rather than the fluctuating components of the velocity. Based on its review of the licensee's October 16, 2008, submittal, it was not apparent how this quasi-TKE quantity was implemented in the CFD code and analysis, how it is consistent with CFD plots that were included in

LIC-1 5-0100 Page 8 the supplemental responses to GL 2004-02, and whether several of the conclusions following from the use of this quasi-TKE quantity are valid. The NRC staff requests further discussion regarding the use of 0.01 feet per second (ft/s) as a transport metric for fines, the isotropy of turbulence, and the correlation of terminal settling with TKE.

OPPD RAI #10 Response:

OPPD's implementation of the Pressurized Water Reactor (PWR) Water Management Initiative is targeted at both increasing the water available for initial safety injection and reducing the amount of debris transported to the ECOS sump strainers. Reduction in the amount of debris transported to the strainer reduces the reliance on the strainer for maintaining recirculation during the 30-day mission time. Relative to OPPD's initial response to this RAl and ongoing dialogue with the NRC, there is reasonable assurance that the combination of a high recirculation pool water level, low recirculation flow rates, and containment geometry provide a condition where flow in the recirculation pool is not sufficient to transport fine debris from all areas of the pool.

As a result of discussions between the NRC and OPPD that occurred on June 30, 2010, OPPD is addressing the issues raised in the February 2010 NRC letter (Enclosure 1, Reference 5) by providing reasonable assurance that portions of the containment post-LOCA pool will not communicate with the ECCS sump strainer for a large break (LB) LOCA. This is accomplished with CFD simulation data that shows that the recirculation pool geometry (represented in Figures 1 and 2 below) 1 . The two SG compartments are separated from each other and flow from one compartment (the break compartment) to the strainer results in a condition where flow from the compartment opposite of the break does not communicate with the strainer. Hence, there is no driving force to transport debris from the compartment opposite of the break to the strainer.

Following an LBLOCA, it is conservatively assumed that insulation debris (RMI, low density fiberglass (LDFG), and TempMat) are uniformly distributed between the locations where it would have been destroyed and the sump strainers. This is a conservative assumption since the blowdown and pool fill-up phases involve multi-directional flows that would tend to disburse debris around containment (including areas with lower transport potential). This assumption is also valid for no spray flow conditions since the initial distribution of debris is driven by the blowdown and pool fill-up transport rather than the washdown transport and is established prior to recirculation.

1 Note all CFD figures illustrate only the old sump strainers. For conservatism, the actual new strainer geometry and surface area was not modeled. The old strainers were used in the CFD model since more localized flow around the strainers in the original design potentially increased debris transport fraction than what would be expected with the current installed design. Modeling the original strainer design represented a conservative transport fraction estimate and with the installation of the new larger surface area strainers did not jeopardize the conservatism of the transport calculation. As such, all figures illustrate a model with the original strainers for conservatism only. The new strainers have considerably larger strainer areas and thus the flow velocities around the new strainers will be lower than they were in the old design.

LIC-1 5-0100 Page 9 Figure 1: Location of Sump Strainers and Break Inside SG Compartment A The recirculation occurs in a pool with a depth of approximately 5 feet and a recirculation pool flow of less than 1000 gpm. Figure 2 presents a visual representation of the turbulence data present in the recirculation pool where the yellow represents the regions of the pool where the turbulence value is only half the amount required to keep fiber fines in suspension. The flow velocities, combined with the turbulence data, support the fact that the recirculation pool is essentially stagnant in many regions of the pool, especially in the SG compartment opposite of the break.

LIC-1 5-0100 Page 10 Figure 2: Half of the TKE Metric Required to Suspend Individual Fiber The dominating velocities and TKE for the analysis are within Compartment A where the break occurs and throughout the containment pool in the path to the sump strainers, as shown in Figure

3. The pool within Compartment B and around its doorway has very little flow and can be considered stagnant.

Turbulence is required to suspend and transport fines but as shown in Figure 2, because TKE in Compartment B is less than half required to keep fiber in suspension, it would therefore settle to the floor.

LIC-1 5-0100 Page 11 Figure 3: Velocity Magnitude and TKE for Break in SG Compartment A Vel. Mag.

5 0.120 I:00 TKE Ž 0.034 ft2 fs2 To further substantiate the lack of flow associated with the SG compartment opposite of the break compartment, an additional evaluation of the CFD simulation data at the entrance to the SG compartment of the break compartment was performed.

For the case of a line break in SG Compartment A, the flux of the containment pool at several locations inside the containment building was analyzed to determine debris transport areas. Flux is defined as the amount of fluid flowing through a unit area per unit time.

To analyze the pool flow patterns, six different surface planes were created throughout the containment building as shown in Figure 4, to determine fluid flux magnitudes of the containment pool at various doorways and corridors resulting from a break inside SG Compartment A. The flux planes are hypothetical surfaces set perpendicular to containment pool circulation located at the two compartment doorways, near the sump strainers, and in the flow path from Compartment A to the sump strainers. For example, Flux Plane 1 was analyzed to determine flux metrics through the doorway of Compartment A and Flux Plane 6 was analyzed to determine flux metrics through the doorway of Compartment B.

LIC-1 5-0100 Page 12 Figure 4: Location of Surface Planes where Flux of the Containment Pool was Analyzed

LIC-15-0 100 Page 13 Figure 5 shows results of the analysis at the location of flux plane 6 in which velocity magnitude was integrated through the SG Compartment B doorway. The figure shows the velocity magnitude at a specific point in time in order to illustrate very low flow metrics at the doorway, with an average velocity magnitude of 0.005 ft/s (into the compartment) at 1,850 seconds.

Figure 5: Integration on Flux Surface 6 for Evaluation of Velocity Magnitude and Turbulence through Doorway to SG Compartment B at 1850 Seconds

LIC-1 5-0100 Page 14 The plot in Figure 6 shows the flux magnitude over time for the six flux surfaces illustrated in Figure 4. Positive and negative flux values correspond to direction of flow through the doorways and corridors. A positive flux represents the containment pool flowing in the positive x and y directions towards the break and negative flux represents the containment pool flowing in the negative x and y directions away from the break. Figure 6 shows flow consistently moves from the break to the strainer as demonstrated by the flux values for surfaces 1, 2, and 3. Figure 6 also shows that little or no flow occurs in areas of the pool that are not in the expected path of flow from the break location to the strainer. This is demonstrated by the fact that the fluxes in the containment pool at or near the doorway to Compartment B through surfaces 4, 5, and 6 have the lowest magnitude and fluctuate about zero with a maximum of 0.5 and -0.5 ft3/s. This shows that water does slosh in and out of Compartment B over time, but at a low rate which is not likely to transport debris.

The CFD model demonstrates that since there are no sprays or other flow sources in Compartment B to push debris out of the compartment, there is only minor "sloshing" of water in and out of the compartment and this "sloshing" would result in no significant transport of debris from Compartment B to the ECCS strainers. Therefore, use of a reduced transport fraction in the FCS debris transport analysis which accounts for the lack of transport from SG compartment opposite of the break compartment is acceptable.

LIC-1 5-0100 Page 15 Figure 6: Comparison of Flux at Six Surface Planes in the Containment Pool Resulting from a Break in Compartment A Fluid Flow At Flux Planes Time (s) 1100 1300 1500 1700 1900 2100 2300 2500 1.00E+O00i 5.00E-01

-5.00E-01

-4e-fflux surface 1

-- fflux surface 2 ffux surface 3 E.-1.00E+00 x =X-*-fflux surface 4

• -**fflux surface 5

--*-fflux surface 6

-1.50E+00

-2.00E+O0

-2.50E+00 --

-3&00E+O0

LIC-1 5-01 00 Enclosure 2 Page 16 RAI #15 Question:

The NRC staff requested additional information on the temperature extrapolation methodology because it is not apparent that the bore holes that occurred at the low test temperatures would occur at the higher temperatures in the actual sump pool. The licensee provided additional information in its October 16, 2008, response regarding the methodology used for temperature extrapolation. The methodology used is considered valid for the pressure range in which the temperature measurements were taken. However, the temperatures which the results are being extrapolated to are much higher than those tested. The higher LOCA temperatures result in significantly lower head losses than those measured at the test temperature. The licensee provided graphs of head loss, flow, and temperature for some of the tests. The licensee stated that, based on short-term head-loss variation, the formation and filling of bore holes appeared to be about the same between the high-temperature testing, about 95 degrees Fahrenheit (0 F) and the low-temperature test, about 65°F. The same head-loss curves appear to indicate, based on short-term head-loss variation from the average, that bore holes form at higher head losses and may not be present or have as large an effect at lower head losses. Two of the examples appeared to result in larger short-term variations in head loss at about 40 to 50 inches of head loss, and one example appeared to result in larger variations at about 20 inches of head loss. Therefore, the staff cannot conclude that bore holes would form at lower differential pressures. The licensee should justify that the reduction in head loss due to bore hole formation would be expected to occur at lower differential pressures. It is likely that the debris bed morphology has a significant impact on bore hole formation.

This could explain some differences in the short-term head-loss variation between tests.

The licensee should provide additional information that justifies that the temperature compensation applied to the test results does not affect the evaluation non-conservatively.

OPPD's RAI #15 Response:

The design of the FCS strainers includes two distinct design bases, one for a large break (LB)

LOCA condition and one for a small break (SB) LOCA condition.

LBLOCA:

Extension of LBLOCA test results to the plant strainer does not include temperature-based scaling or extrapolation of head loss; flow through the strainer is assumed to be completely turbulent due to observed bore hole formation during testing. The differential pressures that lead to bore hole formation in the LBLOCA condition are the same in the plant as they are in the test. This issue has been resolved.

SBLOCA:

The debris load for the FCS strainers in an SBLOCA condition is very low in fiber; the maximum nominal fiber bed thickness is less than 1 / 1 6 th of an inch. Since fiber does not accumulate uniformly, there were areas of the strainer 6 inches x 1 inch in size that were left uncovered, as well as, evenly covered areas.

Per the March 2008 NRC Guidance:

LIC-1 5-01 00 Enclosure 2 Page 17 If [boreholes havel occurred, the maximum head loss value determined during testing (at lower than accident temperatures) may be assumed to be the head loss value for the strainer. If indications of channeling were observed during testing, any correction for temperature should be justified from both the perspective of laminarand turbulent flow regimes, and from the perspective that the head loss could have been higher if debris bed morphology changes due to differential pressure had not occurred.

The following paragraphs address RAI #15 within the bounds of the March 2008 guidance:

After head loss had reached steady state in the limiting test, General Electric (GE) reduced the water temperature while maintaining flow. This temperature reduction served two purposes, it indicated that bore hole formation was relatively independent of temperature, and it measured the laminar/turbulent distribution of head loss by adjusting the fluid viscosity while maintaining the approach velocity constant.

Bore hole formation is indicated by the high-frequency oscillations of the measured head loss in the limiting SBLOCA test case, 11 M-SBLOCA-Jacketed. The head loss oscillations are indicative of the formations and closing of bore holes, and are usually associated with a high-calcium silicate, low fiber debris load like that at FCS. Bore holes are clearly formed at head losses below the maximum allowable plant head loss of 4.79 feet (57.5 inches) of water, and this is particularly apparent when the repeat tests, 8M-RPT and 9M-RPT, are considered (see below).

Although the test was performed at a lower temperature, i.e., higher viscosity and differential pressure than the limiting conditions at the plant (196.6°F), bore hole formation is predicted to occur at both temperatures and pressures. In the limiting test case, 11 M-SBLOCA-Jacketed, bore hole formation begins to occur as soon as head loss begins to become measurable, (as shown in the figure below). The amplitude of the effect of bore holes on head loss increases with measured head loss until reaching a maximum at a measured head loss of 50-60 inches of water, where the bore hole head loss oscillation amplitude remains approximately constant at 8 inches of water for the duration of the test.

The debris bed shift that occurs at approximately 1100 minutes is typical of large-scale head loss testing, and is not indicative of bore hole formation.

Although the SE of NEI 04-07 indicates that temperature scaling should not be used directly in cases where bore hole formation is apparent, by using viscosity scaling with consideration of the laminar/turbulent head loss distribution, the measured head loss can be extrapolated to the plant sump temperature.

Plots of head loss during repeat tests 8M-RPT and 9M-RPT are included below to show that bore hole behavior occurs even at lower head losses than in test 11 M-SBLOCA. Note that tests 8M and 9M were exact replicates of test 11 M-SBLOCA, and stabilized at lower head losses with significant bore hole activity. Tests 8M and 9M were performed after test 11 M.

LIC-1 5-0100 Page 18 Figure 15.1 - Head Loss Plot for Test 11IM-SBLOCA-Jacketed 1 I I I I 1 11 M-,SBLOCA-Jacketed 0

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Figure 15.2 - Head Loss Plot for Test 8M-RPT

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0 200 400 600 800 1000 1200 1400 1600 1800 Time (mai)

LIC-15-01 00 Page 19 RAI #20 Question:

The NRC staff requested additional information regarding how the velocities and turbulence in the test flume compare to similar variables predicted in the plant sump pool (i.e., for tests that allowed settling (no stirring), provide a comparison of the flows predicted around the strainer in the plant and the flows present in the test flume during the testing). In its RAI response, dated October 16, 2008, the licensee provided additional descriptions of how the velocities in the area of the strainer were determined. A CFD analysis of the sump pool, run at a higher flow rate than would be expected using the current assumptions, indicates that the average flow in the area of the strainers is about 0.075 feet per second (ft/sec). Adjusting this for the currently assumed flow rate, the velocity in the area of the strainer is calculated to be about 0.05 ft/sec. The RAI response stated that the flow in the test apparatus was set to be equal to the predicted flow rate by adjusting the width of the test flume walls. The basic concepts applied in the RAI response are accepted by the staff. However, it appeared that the CFD analysis referenced by the response had some non-uniform flow areas near the strainer. Areas of higher flow would promote transport. Additionally, the RAI response did not address the turbulence available to maintain debris in suspension. The licensee should provide a more detailed CFD analysis in the area of the strainer so that flow velocities would be more fully defined. In addition, the licensee should provide a comparison of turbulence around the strainers in the plant and the turbulence in the test flume. Alternately, the licensee could provide alternate information that shows that the test was not non-conservatively biased due to these factors.

OPPD's RAI #20 Response:

OPPD must defer a response to RAI #20 until staff acceptance of the testing protocol discussed in RAI #10. The outcome may determine whether additional testing is required to be conducted, which is tracked by Commitment #2 in Enclosure 4.

RAI #25 Question:

The NRC staff requested additional information on the justification for treating unqualified alkyd original equipment manufacturer (OEM) coatings as chips at FCS despite the contradictory data presented in Electric Power Research Institute (EPRI) report #1011753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings" (OEM report), dated September 2005. The licensee stated that the alkyd OEM coatings fail in 5 mil chips. The staff does not accept alkyds failing as chips since in the EPRI OEM report, alkyds do not fail as chips. According to staff guidance ["NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing," dated March 2008 (ADAMS Accession No. ML080230038).], the alkyd OEM coatings failing as fine particulate would be more conservative since a thin bed has been observed to form. The impact of the alkyd OEM coatings could add an additional 40 pounds mass (Ibm) to the debris. Please provide additional justification for treating alkyd OEM coatings debris as chips rather than treating it as particulate.

LIC-1 5-0100 Enclosure 2 Page 20 OPPD's RAI #25 Response:

The original FOS Debris Generation Post-LOCA calculation (FC06985) identified 0.23 ft3 of alkyds with another alkyd coated component being installed (HE-44 telescoping crane - 0.25 ft3 estimate). Thus, a total of 0.48 ft3 of alkyd coatings were identified as potential alkyd coating sources in containment for the original debris generation calculations.

Since the original debris generation calculation was performed, the additional removal of some alkyd coated components in containment has been completed. Twenty (20) out of the twenty-five (25) lead blanket storage drums (alkyd coated) were permanently removed. The reactor head lifting device (alkyd coated) was replaced with the rapid refueling package component. The yellow jib crane, noted in the original calculation, is installed at elevation 1045 feet, outside of any jet ZOI. It is significantly higher than the containment sump pool elevation and is not subject to containment spray.

Also, since the original debris generation calculation was performed, telescoping cranes HE-44 and HE-48 have been added to containment. The two cranes are at elevation 1057 feet and thus would not be immersed or subject to containment spray. They are completely outside of the bioshield areas containing the direct jet impingement blasts. Due to interfering structures, components, and grating, HE-44 would not be subjected to a direct impingement blast. The coating system from the telescoping cranes was included as part of the original OEM EPRI testing program.

The only way that un-impinged alkyd coatings affect head loss at FCS (containment spray is not employed post-LOCA) is if they are submerged and subject to dissolution of the coating matrix.

OPPD recognizes that pigments in alkyd coating systems could potentially dissolve out of the coating matrix over time due to buffer or temperature effects. Thus, from the original debris generation calculation, the only sources of alkyds that could be subject to immersion are electrical junction boxes, Limitorque valve operators, detector well cooling fans, reactor coolant drain tank (RCDT) pumps, five lead blanket storage drums. This amounts to approximately 0.17 ft 3 of plant alkyd material (16 Ibm).

As described in the draft responses to RAIs #34 and #35, OPPD modified the nuclear detector well cooling units during the spring 2011 RFO; which reduced the post-LOCA aluminum inventory in the FOS containment sump and the resulting amounts of chemical precipitates. The resolutions for the LBLOCA cases are described below:

LBLQCA: The existing FCS strainer qualification test data is based on a plant equivalent of 647.3 lb of sodium aluminum silicate, and after the spring 2011 REO, there is a preliminarily calculated maximum of 109.8 lbs of sodium aluminum silicate in the FCS containment sump. This results in a 537-pound sodium aluminum silicate surplus in the LBLOCA strainer design basis, and this surplus more than compensates for an additional 16 lb of particulates added to the debris load to account for unqualified coatings.

SBLOCA: The existing FCS strainer qualification test data was based on a plant equivalent of 10.6 lbs of sodium aluminum silicate, and a plant equivalent of 6.0 lbs of aluminum oxyhydroxide.

With the completion of the spring 2011 outage, a significant source of aluminum has been removed. However, the impact of that aluminum removal related to chemical debris formation is still being evaluated related to the previous strainer testing. In conclusion, OPPD will treat potentially submerged alkyd coatings as a source of fine particulates in the manner stated above.

The unqualified and unsubmerged alkyd coatings in containment will be assumed to fail as chips;

LOIC-5-0100 Page 21 this is considered conservative based on bench-top test observations as well.

RAI #34 Question:

The NRC staff requested additional information regarding a number of issues related to silicate inhibition of aluminum corrosion, including among other information the type and amount of plant debris assumed as the source of silicates. In its October 16, 2008, letter, the licensee provided a table which listed the debris generated for various breaks including a small break. The table indicates that there is only a small amount of calcium silicate destroyed by the small break. Since silicate inhibition of aluminum corrosion is credited for all potential breaks, please provide greater detail on the source of the silicate for the small break. Specifically, please explain whether the silica source for a small-break LOCA is strictly calcium-silicate or whether fiberglass or other materials are considered a source of silicate. This point is important because the inhibition of aluminum corrosion may not occur if insufficient silicate is present for the small-break case.

OPPD's RAI #34 Response:

OPPD identified the need to modify the nuclear detector well cooling units at FCS. These units provided the largest contribution of aluminum to the analysis of chemical precipitate formation.

As a result, a revised WCAP-16530 assessment indicates that a substantial reduction in the formation of sodium aluminum silicate and aluminum oxyhydroxide is realized through removal of the cooling fins. FCS modified nuclear detector well cooling units, and with this modification, the aluminum source term within containment was significantly reduced. After the spring 2011 RFO, no credit will be taken for silicon inhibition, and thus this RAl will no longer apply to FCS. Although the NRC found OPPD's original response to this RAI acceptable, an interim draft response is included below.

The debris quantities that were outlined in the previous response (i.e., Enclosure 1, Reference 2) to the NRC were used to calculate chemical precipitate formation. The silicon-contributing debris sources for the small break cases consisted of low-density fiberglass, CaI-Sil, and a Cal-Sil/asbestos composite. The silicon release rate correlation from WCAP-16530 was applied to the SBLOCA debris loads to determine the resultant silicon concentration. No additional quantities of silicon were assumed to be released other than what was considered formed based upon strict utilization of WCAP-1 6530 as applied to the SBLOCA case, and OPPD utilized the methodology described in WCAP-1 6530 and WCAP-1 6785 to calculate the formation of chemical precipitates at FCS. The small break LOCA scenario from preliminary calculations is now dominated by the dissolution of aluminum in containment. During an upcoming refueling outage, OPPD plans (EC 49722) on replacing the pressurizer spray line fiberglass insulation with RMI, which will decrease the quantity of potential chemical debris.

RAI #35 Question:

The NRC staff requested additional information concerning how aluminum solubility was credited. This information was needed to determine if the long-term solubility credit is based on the pool temperature never dropping below 140°F. Based on review of the licensee's October 16, 2008, RAI response, it is unclear to the staff whether the licensee's

LIC-1 5-0100 Enclosure 2 Page 22 analysis applies the aluminum hydroxide precipitate at a delayed time or if the aluminum is assumed to remain in solution for the duration of the post-LOCA mission time. The licensee did not provide justification for the credit taken as discussed in the staff's March 2008 chemical effects review guidance ["NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant Specific Chemical Effect Evaluations," dated March 2008 (ADAMS Accession No. ML080380214).]. Please provide additional detail regarding the solubility of aluminum and how aluminum-based precipitates are accounted for, or discounted, in the final analysis.

OPPD's RAI #35 Response:

Only the small break (SB) LOCA portion of the previous response (Enclosure 1, Reference 2) is being revised for this submittal based on the discussion with the NRC during the June 30, 2010 teleconference.

OPPD modified the nuclear detector well cooling units at FCS during the 2011 refueling outage (RFO). These units provided the largest contribution of aluminum to the analysis of chemical precipitate formation. As a result, a revised WCAP-16530 assessment indicates that a substantial reduction in the formation of sodium aluminum silicate and aluminum oxyhydroxide has been realized through removal of the cooling fins for the LBLOCA bounding scenario. The modification to the nuclear detector well cooling units made during the spring 2011 REQ has significantly reduced the aluminum source within containment.

Table 35.1 -Predicted FCS Aluminum Precipitate Quantities Existing Aluminum Inventory' Tested Debris Quantities Revised InventoryAluminum (2Q11)

WCAP1653 & WAP-1530Preliminary WCAP-WCP150&WCAP-16530 WCP15016530 WCAP-1 6785 Surrogates (oslct niiin LBLOCA- SBLOCA LBLOCA LBLOCA SBLOCA RC2A RC2A SBLOCA LBLOCA SBLOCA RC2A Scenario Scenario Scenario Sodium aluminum 4.3 kg 286.1 kg 4.3 kg 327 kg 4.8 kg 293.6 kg 6.0 kg 47.3 kg silicate _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _ _ _ _ _ _ _

Almnm0 kg 0 kg 108.0 kg 33 kg 2.7 kg 0 kg 30.3 kg 0 kg oxyhydroxide _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _

To address the differences between the two limiting break scenarios, the response is provided in two sections:

LIC-1 5-0100 Enclosure 2 Page 23 LBLOCA:

The aluminum source term calculated by WCAP-1 6530 will generate zero aluminum oxyhydroxide and less sodium aluminum silicate than was tested with the FCS strainers and will provide sufficient margin such that the consideration of aluminum solubility is not required and the subject of this RAI is no longer applicable to the LBLOCA case. No credit for aluminum solubility is taken for the LBLOCA case.

SBLOCA:

From the summary of the June 30, 2010, teleconference, OPPD agreed to reevaluate OPPD's RAI response and either show the pH is greater than 7.5 or reevaluate solubility. OPPD has determined that the current Technical Specification sodium tetraborate (NaTB) Volume Required for RCS Critical Boron Concentration (ARO, HZP, No Xenon) curve supports a pH of 7.05.

Therefore, OPPD will reevaluate solubility.

With regard to possible increased predicted amounts of aluminum oxyhydroxide (SBLOCA),

OPPD plans to remove the fiberglass insulation from the pressurizer spray control line piping (see Figure 4 of Reference 1 ) and replace it with RMI (EC 49722) during an upcoming refueling outage.

This will eliminate the source of any transportable fibrous debris. Without transportable fibrous debris, the chemical pr'ecipitate formed is anticipated to pass through the strainer without forming a debris bed.

The RAl #35 response is based on Enclosure 1, Reference 7.

RAI #37 The latent fiber quantity assumed by FCS is only 2.79 percent, rather than the default suggested 15 percent from the NRC staff's SE on NEI-04-07. The NRC staff believes the apparent difference is due to fiber collection efficiency differences between the debris collection method used by FCS (scraping with a metal scraper) and vacuuming or wiping with masolin cloths used by the other plants in the NUREG-6877 survey. Although the collection efficiency of fiber is not discussed specifically, the NUREG notes that differences in collection method have a large impact and specifically notes that the metal scraper method resulted in a much lower fraction of fine particulate for FCS as compared to the other plants surveyed. Additionally, the fiber percentage of 2.79 percent is based on only eight samples (total mass: 27 grams), which is not enough for statistical accuracy for scaling a mass distribution up by a factor of over 2500. For properly collected debris samples, a fiber-mass proportion of 15 percent should be applied to the total inventory estimate in the absence of site-specific supporting evidence. Therefore, the NRC staff believes the licensee should have used the standard 15 percent value for the latent fiber mass percentage as opposed to relying upon the Plant C data in NUREG/CR-6877, "Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment Buildings," dated July 2005 (ADAMS Accession No. ML052430751). Please justify the use of 2.79 percent latent debris distribution fiber for FCS.

LIC-1 5-0100 Page 24 OPPD's RAI #37 Response:

The latent debris sampling procedure was enhanced by increasing the sample size from 8 to 25 samples and better defining sample locations. Samples are to be collected on both horizontal and vertical services as follows:

• 4 horizontal collection areas on the 1060-foot elevation

• 4 horizontal collection areas on the 1045-foot elevation

• 5 horizontal collection areas on the 1013-foot elevation

• 6 horizontal collection areas on the 994-foot elevation

• I vertical collection area on the concrete wall

• I vertical location area on the liner

• I vertical location area on piping/equipment

• 3 horizontal collection areas in large equipment collection areas The exact locations were chosen based on engineering experience, judgment, and knowledge.

The procedure cautions to ensure that areas are representative of debris type and accumulation.

Locations considered are those near robust barriers (i.e., each steam generator bay, annulus bay and the operating floor). Specifically, the LOCA impingement area inside the bioshield will be sampled. Latent debris path areas and areas that are in the surnp flow path will also be sampled.

Sample weights are averaged for each representative area and linearly extrapolated to the corresponding surface areas in containment. One hundred percent (100%) of the 994-foot elevation horizontal surface latent debris is included in the source term, since the 994-foot elevation is flooded after a LOCA. Only the 994-foot elevation is exposed to coolant, due to the lack of containment sprays during a LOCA at FCS, but additional latent debris sources are included in the source term for conservatism. One hundred percent (100%) of the vertical surface latent debris, 50% of the 1013-foot elevation horizontal surface, 50% of the 1045-foot elevation horizontal surface, and 25% of the 1060-foot elevation horizontal surface latent debris are included in the latent debris source term, even though these areas will generally not be exposed to post-LOCA coolant flow.

Latent debris sampling at FCS is conducted near the end of the outage but prior to Radiation Protection (RP) personnel commencing containment cleaning, so latent debris accumulation on the containment surfaces is at a maximum during sampling due to the latent debris generated by outage activities. Since RP cleaning does not occur prior to latent debris sampling, the timing of ECS latent debris sampling results in conservative latent debris quantities and conservatively bounds the latent debris quantities that remain in containment after RP cleaning has been completed. To confirm the conservatism of this collection timing, OPPD has taken samples near the beginning and end of the outage since 2011.

Although NEI 04-07 recommended an assumed fibrous debris fraction for latent debris of 15%,

this value is directed toward containments utilizing primarily fibrous insulation. The FCS containment contains almost exclusively calcium silicate insulation and reflective metallic insulation (RMI), so it is reasonable to expect that there is a reduced latent fibrous debris fraction as compared to a primarily fiberglass insulation containment. The fraction of latent fiber at ECS is determined using measurements of existing containment latent debris, rather than through the application of an assumed fiber fraction that is not applicable to ECS considering the absence of fibrous insulation in containment.

It is OPPD's position that the original survey information was properly obtained, is accurate, and

LIC-15-0 100 Page 25 is conservatively representative of actual conditions. Therefore, 2.79% is used in the analysis rather than a default of 15%. However, based on discussions with the NRC, OPPD will adopt a fibrous fraction for latent debris of 15%.

As noted in Reference 3, OPPD's April 2010 draft response addressed the NRC staffs concerns with regard to a 15 percent assumption and OPPD agreed to add a description of the debris collection method, which is as follows. Procedure SE-PM-AE-1005, "Latent Debris Collection Inspection" provides instructions to quantify debris inside containment to monitor the effectiveness of cleanliness programs and confirm assumptions made in the Debris Generation Calculation.

The procedure ensures that the amount of latent debris inside containment does not present a challenge to ECCS Sump operability following a LOCA.

Section 7.1 of SE-PM-AE-1 005 addresses determination of collection areas.

Four (4) horizontal collection areas are identified on the 1060-foot and 1045-foot elevation of containment. Five (5) horizontal collection areas are identified on the 1013-foot elevation of containment and six (6) horizontal collection areas are identified on the 994-foot' elevation of containment.

Section 7.3 of SE-PM-AE-1 005 addresses sample collection.

Debris samples are collected from the areas described above using mechanical means such as a nylon brush or putty knife to loosen debris for collection. Once the debris has been loosened, the Masslin cloth is wiped over the entire area to ensure that all debris in the sample area is collected. The sample debris and Masslin cloth are deposited in a collection bag where it is weighed with a highly accurate scale. Data analysis is then performed to ensure total containment latent debris load is less than 159 Ibm.

RAIs #39 through #47 The following set of issues relate to ZOI issues that were not specifically identified in the 2008 RAIs for FCS. These issues were developed as a result of NRC staff review of certain documents developed by the PWROG that are used as a basis for certain assumed ZOI reductions for FCS.

The PWROG is planning to respond to some of these generically, but it is unknown which of the issues below will be generically answered and which will be site-specific.

Questions:

39. Although the American National Standards Institute/American Nuclear Society (ANSlIANS) standard predicts higher jet centerline stagnation pressures associated with higher levels of subcooling, it is not intuitive that this would necessarily correspond to a generally conservative debris generation result. Please justify the initial debris generation test temperature and pressure with respect to the plant-specific reactor coolant system conditions in the plant hot-and cold-leg operating conditions. If ZOI reductions are also being applied to lines connecting to the pressurizer, please discuss the temperature and pressure conditions in these lines.

Were any tests conducted at alternate temperatures and pressures to assess the variance in the destructiveness of the test jet to the initial test condition specifications? If so, please provide that assessment.

40. Please describe the jacketing/insulation systems used in the plant for which the

LIC-1 5-0100 Page 26 testing was conducted and compare those systems to the jacketinglinsulation systems tested and demonstrate that the conditions and materials adequately represented the plant jacketinglinsulation system. Please describe differences in the jacketing and banding systems used for piping and other components for which the test results are applied, potentially including valves and other fittings. At a minimum, the following areas should be addressed:

a. Please describe how the characteristic failure dimensions of the tested jacketing/insulation compare with the effective diameter of the jet at the axial placement of the target. The characteristic failure dimensions are based on the primary failure mechanisms of the jacketing system, e.g., for a stainless steel jacket held in place by three latches where all three latches must fail for the jacket to fail, then all three latches must be effectively impacted by the pressure for which the ZOI is calculated. Applying test results to a ZOl based on a centerline pressure for relatively low nozzle-to-target spacing would be nonconservative with respect to impacting the entire target with the calculated pressure.

b. Please describe if the insulation and jacketing system used in the testing was of the same general manufacture and manufacturing process as the insulation used in the plant. If not, please describe what steps were taken to ensure that the general strength of the insulation system tested was conservative with respect to the plant insulation. It is known that there were generally two very different processes used to manufacture calcium silicate whereby one type readily dissolved in water but the other type dissolves much more slowly. Such manufacturing differences could also become apparent in debris generation testing, as well.

c. Please provide an evaluation of scaling the strength of the jacketing or encapsulation systems to the tests. For example, a latching system on a 30-inch pipe within a ZOI could be stressed much more than a latching system on a 10 inch pipe in a scaled ZOI test. If the latches used in the testing and the plants are the same, the latches in the testing could be significantly under-stressed. If a prototypically sized target were impacted by an undersized jet, it would similarly be under-stressed. Evaluations of banding, jacketing, rivets, screws, etc., should be made. For example, scaling the strength of the jacketing was discussed in the OPG report on calcium silicate debris generation testing.

41. There are relatively large uncertainties associated with calculating jet stagnation pressures and ZOls for both the test and the plant conditions based on the models used in the WCAP reports. Please explain the steps taken to ensure that the calculations resulted in conservative estimates of these values. Please provide the inputs for these calculations and the sources of the inputs.

42. Please describe the procedure and assumptions for using the ANSI/ANS-58-2-1988 standard, "Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture," to calculate the test jet stagnation pressures at specific locations downrange from the test nozzle.

a. In 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

LIC-1 5-0100 Page 27 Operating Plants," was the analysis based on the same initial condition as the initial test temperature, specified as 550°F? If not, please provide an evaluation of the significance of the difference.

b. Please explain whether the water subcooling used in the analysis was that of the initial tank temperature or the temperature of the water in the pipe next to the rupture disk. Test data indicated that the water in the piping had cooled below that of the test tank.

c. The break mass flow rate is a key input to the ANS/IANS-58-2-1988 standard.

Please explain how the associated debris generation test mass flow rate was determined. If the experimental volumetric flow was used, please explain how the mass flow was calculated from the volumetric flow given the considerations of potential two-phase flow and temperature dependent water and vapor densities. If the mass flow was analytically determined, then describe the analytical method used to calculate the mass flow rate.

d. Noting the extremely rapid decrease in nozzle pressure and flow rate illustrated in the test plots in the first tenths of a second, please explain how the transient behavior was considered in the application of the ANSI/ANS-58-2-1988 standard.

Specifically, please explain whether the inputs to the standard represent the initial conditions or the conditions after the first extremely rapid transient (e.g.,

say at one tenth of a second).

e. Given the extreme initial transient behavior of the jet, justify the use of the steady state ANSI/ANS-58-2-1 988 standard jet expansion model to determine the jet centerline stagnation pressures rather than experimentally measuring the pressures.

43. Please describe the procedure used to calculate the isobar volumes used in determining the equivalent spherical ZOI radii using the ANSI/ANS-58-2-1988 standard.

a. What were the assumed plant-specific reactor coolant system temperatures and pressures and break sizes used in the calculation? Note that the isobar volumes would be different for a hot leg break than for a cold leg break since the degrees of subcooling is a direct input to the ANSI/ANS-58-2-1988 standard and which affects the diameter of the jet. Note that an under-calculated isobar volume would result in an under-calculated ZOI radius.

b. What was the calculational method used to estimate the plant-specific and break-specific mass flow rate for the postulated plant LOCA, which was used as input to the standard for calculating isobar volumes?

c. Given that the degree of subcooling is an input parameter to the ANSI/ANS 2-1988 standard and that this parameter affects the pressure isobar volumes, what steps were taken to ensure that the isobar volumes conservatively match the plant-specific postulated LOCA degree of subcooling for the plant debris generation break selections? Were multiple break conditions calculated to ensure a conservative specification of the ZOI radii?

LIC-15-0 100 Page 28

44. Please provide a detailed description of the test apparatus specifically including the piping from the pressurized test tank to the exit nozzle including the rupture disk system.

a. Based on the temperature traces in the test reports, it is apparent that the fluid near the nozzle was colder than the bulk test temperature. How was the fact that the fluid near the nozzle was colder than the bulk fluid accounted for in the evaluations?

b. How was the hydraulic resistance of the test piping which affected the test flow characteristics evaluated with respect to a postulated plant-specific LOCA break flow where such piping flow resistance would not be present?

c. What was the specified rupture differential pressure of the rupture disks?

45. WCAP-1 671 0-P discusses the shock wave resulting from the instantaneous rupture of piping.

a. Was any analysis or parametric testing conducted to get an idea of the sensitivity of the potential to form a shock wave at different thermal-hydraulic conditions? Were temperatures and pressures prototypical of pressurized-water reactor hot legs considered?

b. Was the initial lower temperature of the fluid near the test nozzle taken into consideration in the evaluation? Specifically, was the damage potential assessed as a function of the degree of subcooling in the test initial conditions?

c. What is the basis for scaling a shock wave from the reduced-scale nozzle opening area tested to the break opening area for a limiting rupture in the actual plant piping?

d. How is the effect of a shock wave scaled with distance for both the test nozzle and plant condition?

46. Some piping oriented axially with respect to the break location (including the ruptured pipe itself) could have insulation stripped off near the break. Once this insulation is stripped away, succeeding segments of insulation will have one open end exposed directly to the LOCA jet, which appears to be a more vulnerable configuration than the configuration tested by Westinghouse. As a result, damage would appear to be capable of propagating along an axially-oriented pipe significantly beyond the distances calculated by Westinghouse. Please provide a technical basis to demonstrate that the reduced ZOls calculated for the piping configuration tested are prototypical or conservative of the degree of damage that would occur to insulation on piping lines oriented axially with respect to the break location.

47. WCAP-16710-P noted damage to the cloth blankets that cover the fiberglass insulation in some cases resulting in the release of fiberglass. The tears in the cloth covering were attributed to the steel jacket or the test fixture and not the steam jet.

It seems that any damage that occurs to the target during the test would be likely to occur in the plant. Was the potential for damage to plant insulation from similar

LIC-1 5-0100 Page 29 conditions considered? For example, the test fixture could represent a piping component or support, or other nearby structural member. The insulation jacketing is obviously representative of itself. Please explain the basis for the statement in the WCAP that damage similar to that which occurred to the end pieces in not expected to occur in the plant. It is likely that a break in the plant will result in a much more chaotic condition than that which occurred in testing. Therefore, it would be more likely for the insulation to be damaged by either the jacketing or other objects nearby.

OPPD's RAI #39-47 Response:

OPPD will not credit the reduced ZOI resulting from the testing activities associated with the subject of RAIs #39 through #47. However, there is one insulating system installed at FCS where an alternate ZOI is based on similar materials and configuration as those approved in the NRC's SE on NEI 04-07.

A reinforced PCI stainless steel jacketed NUKON insulation system is installed on a 3-inch pressurizer spray line and its proximity to the strainer results in this being a debris source for the SBLOCA event at FCS. The banding used to reinforce this NUKON insulation is similar to the banding used in the OPG testing of CaI-Sil insulating material. The banding is 0.5 inches wide by 0.03 inches thick stainless steel secured with standard banding crimp connectors. The bands are spaced 3 inches + 0.25 inch. The bands used for the OPG CaI-Sil tests were 0.020-inch thick stainless steel bands with standard crimp connectors with an average spacing of 6.5 inches.

Latches, as in the original PCI NUKON insulation system, were found to be the failure point at the Colorado Experiment Engineering Station, Inc. (CEESI) Air Jet Testing. The main function of the bands is to ensure that the jacketing does not peel off following the blast due to a hypothetical double ended guillotine break. If the jacketing remains in place following the overpressure pulse caused by the instantaneous break, then the jacketing will provide the necessary protection of the NUKON from the effects of jet impingement.

The SE to NEI-04-07 states that "'Sure Hold" bands installed over the PCI NUKON insulation system would have a destruction pressure of 90 psig, which translates to a 2.4D ZOI. The stainless steel banding for the reinforced NUKON insulation system at FCS is more closely spaced than the "Sure Hold" bands which provides additional restraining capability to preclude the stainless steel jacketing from peeling off due to the initial blast and subsequent jet impingement.

The jacketing of the PCI NUKON insulation system is stainless steel as opposed to the Al jacketing used in the OPG tests. The main function of the jacketing is to protect the underlying NUKON jackets from damage caused by normal operational challenges and to protect the NUKON blankets from direct jet impingement. There is no significant difference between aluminum and stainless steel jacketing thickness (on the order of 0.01 inch) - both jacketing types will provide considerable protection of the underlying insulation system from the effects of direct jet impingement. Note that the yield strength of stainless steel is approximately 30 to 35 ksi, and the yield strength of aluminum ranges from 10 to 30 ksi depending on the alloy. Therefore, it is likely that the actual strength of the stainless steel jacket is equivalent or better than the strength of the aluminum jacket used in the OPG tests.

The debris generation analysis applies a 3D ZOI to quantify the amount of fiber material that would be generated in the bounding SBLOCA event for this 3-inch pressurizer spray line. This

LIC-1 5-0100 Page 30 3D ZOI is more conservative than the 2.4D ZOI which is applicable to the installed reinforced NUKON insulation system.

However, subsequent discussions with the NRC on the subject of these RAIs concluded that for OPPD to continue to apply the reduced ZOI for the current insulation configuration in this location, banding must be installed.

LIC-1 5-0100 Page 1 Update of OPPD Letter LIC-13-0058 Proposed Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. 1

LIC-1 5-0100 Page 2 QPPD provides this update regarding the status of bypass testing, strainer head loss testing, and characterization of in-vessel effects mentioned in OPPD letter LIC-13-0058 dated May 15, 2013 (Enclosure 1, Reference 7). OPPD is also providing an update to the resolution schedule and commitments made in that letter.

Characterization of Current Containment Fiber Status As discussed in Enclosure 1, Reference 1, bypass testing was conducted with TempMat material and found a bypass fraction of 2.2% (Page 93 of Enclosure to OPPD letter LIC-08-0021). OPPD has utilized the PWRQG methodology for evaluating cold leg break bypass at hot leg switchover, as documented in WCAP-17788-NP Volume 3 (ML15210A668). Preliminary calculations for the cold leg break with maximum fiber load utilizing the 2.2% (documented bypass test fraction), yield in-vessel results of less than 3 grams/fuel assembly fiber loading.

Characterization of Strainer Head Loss Status NRC staff acceptance of testing protocol discussed in RAI #10 (Enclosure 2, Page 7), may determine if additional testing is required to be conducted. See Commitment #2 in Enclosure 4.

Characterization of In-Vessel Effects On July 17, 2015 (ML15210A668), the PWROG submitted WCAP-1 7788 "Comprehensive Analysis and Test Program for GSI 191 Closure (PA-SEE-1090)," which is under review by the NRC. QPPD intends to follow the resolution strategy proposed by the PWRQG in WCAP-1 7788 following NRC review.

Updated Resolution Schedule Changes to the Resolution Schedule provided in QPPD letter LIC-13-0058 dated May 15, 2013 (Enclosure 1, Reference 7) are provided below and are reflected as commitments in Enclosure 4:

• In letter LIC-13-0058, QPPD committed to complete measurements for insulation replacement and remediation by the end of the first refueling outage (RFQ) following January 1, 2013. At the time, due to the 2011 Missouri River flood and the station's entry into Inspection Manual Chapter 0350, ECS had not restarted from the 2011 REQ. In May 2013, anticipating re-start in the near future, the next RFO was expected to begin during the fall of 2014. However, as the 2011 REQ did not end until December 21, 2013, the next REQ occurred in the spring of 2015. This commitment is complete as the measurements for insulation replacement and remediation was performed during the spring 2015 REQ. This is Commitment #1 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to complete strainer head loss and fiber bypass testing by the end of 2015. QPPD is revising this commitment such that any required additional testing will be completed within 18 months of staff acceptance of the testing and debris transport protocol discussed in RAI #10 (Enclosure 2, Page 7). This is Commitment #2 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to complete the necessary insulation replacements, remediation, or model refinements by the completion of the third REQ following January 1,2013, which at the time was expected to be in the fall of 2017. However, as stated above, due to the length of the 2011 RFO, the third REQ following January 1, 2013 will take place in the spring of 2018. Therefore, OPPD is extending the timeframe for completing this commitment from the fall 2017 REQ to the spring 2018 REQ. This is Commitment #3 in Enclosure 4.

LIC-1 5-0100 Page 3

• In letter LIC-13-0058, OPPD committed to submit a final updated supplemental response to support closure of GL 2004-02 for FCS within six months of establishing a final determination of the scope of insulation replacement or remediation. OPPO specified that this would be completed within six months after the fall 2014 RFO. OPPD is revising the due date to be within 18 months of NRC staff acceptance of the testing protocol discussed in RA! #10 (Enclosure 2, Page 7). This is Commitment #4 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to update the current licensing basis (CLB) (i.e., USAR, etc.) following NRC acceptance of the updated supplemental response for FCS and completion of the identified removal or modification of insulation debris sources in containment. OPPD committed to complete this by the end of the fall 2017 RFO. As stated above, the RFO originally anticipated to occur in the fall of 2017 will now occur in the spring of 2018. To allow sufficient time to process the necessary CLB changes, OPPD is revising the due date for updating the CLB to be within 90 days of the end of the spring 2018 REQ. This is Commitment #5 in Enclosure 4.

LIC-1 5-01 00 Page 1 Regulatory Commitments The following table is updated from the Regulatory Commitments Table provided by OPPD in of OPPD letter LIC-1 3-0058 (Enclosure 1, Reference 7). The original commitment is shown in the left column with the revised commitment and/or due date in the right column.

1Complete measurements for insulation replacement and remediation by the end of the first REQ following January 1, 2013; currently expected to be completed during the fall of 2014. [AR 59106]

Due: By the end of the fall 2014 RFO. Due: Completed during the spring 2015 RFO.

2 Strainer head loss and fiber bypass testing Any required additional strainer head loss is expected to be completed by the end of and fiber bypass testing will be completed 2015. [AR 59106] within 12 months of NRC staff acceptance of the testing protocol discussed in RAI

1. 10. [AR 59106]

Due: December 31, 2015. Due: Within 18 months of NRC staff acceptance of testing protocol discussed in RAI #10.

3 Complete the necessary insulation Complete the necessary insulation replacements, remediation, or model replacements, remediation (i.e., EC refinements by the completion of the third 49722), or model refinements by the refueling outage (RFO) following January completion of the third refueling outage 1, 2013 (fall 2017). [AR 59106] (RFO) following January 1, 2013 (spring 2018). [AR 59106]

Due: By the end of fall 2017 RFO.

Due: By the end of spring 2018 RFO.

4 Within six months of establishing a final Within six months of establishing a final determination of the scope of insulation determination of the scope of insulation replacement or remediation, OPPD will replacement or remediation, OPPD will submit a final updated supplemental submit a final updated supplemental response to support closure of GL 2004-02 response to support closure of GL 2004-02 for FCS. for FCS.

[AR 59106] [AR 59106]

Due: Within six months following fall 2014 Due: Within 18 months of NRC staff REQ. acceptance of testing protocol discussed in RAI #10.

LIC-1 5-0100 Page 2 5 OPPD will update the current licensing OPPD will update the current licensing basis (USAR, etc.) following NRC basis (USAR, etc.) following NRC acceptance of the updated supplemental acceptance of the updated supplemental response for FCS and completion of the response for FCS and completion of the identified removal or modification of identified removal or modification of insulation debris sources in Containment, insulation debris sources in Containment.

[AR 59106] [AR 59106]

Due: By the end of fall 2017 RFO. Due: Within 90 days of completing the spring 2018 REQ.

Omaha Public Power District LIC-1 5-0100 January 4, 2016 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 1 Fort Calhoun Station (FCS), Unit 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

Updated Status of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. I

References:

See Enclosure 1 On June 29, 2015, a teleconference between NRC staff and representatives of the Omaha Public Power District (OPPD) was held to discuss closeout of GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" for Fort Calhoun Station, Unit No. 1. OPPD was requested to review and update information that OPPD informally transmitted to the NRC Project Manager (PM) in April 2010 (Reference 4) and updated in October 2010 (Reference 6).

The background for this request is as follows. The information provided in April 2010 was in response to a February 2010 NRC draft request for additional information (RAI) (Reference 3) regarding OPPD's supplemental responses to GL 2004-02 submitted February 29 and October 16, 2008 (References 1 and 2 respectively). OPPD's response to the NRC RAI was informally transmitted in April 2010 (Reference 4) and again in October 2010 (Reference 6) following a June 2010 teleconference (Reference 5).

As requested during the June 29, 2015 teleconference, Enclosure 2 is submitted to formally docket the information emailed to the NRC PM in April and October 2010. The information contained in Enclosure 2 has been updated due to the length of time that has transpired since this information was provided to the NRC in 2010.

OPPD's last docketed correspondence concerning GL 2004-02 was submitted in May 2013 (Reference 7). Enclosure 3 provides a brief update of information provided in Reference 7 and Enclosure 4 is an update to the list of commitments made in Reference 7.

Any other actions discussed in this submittal represent intended or planned actions by OPPD, and are provided for information only; they are not regulatory commitments.

444 SOUTH 16TH STREET MALL ° OMAHA, NE 68102-2247

U. S. Nuclear Regulatory Commission LIC-1 5-0100 Page 2 If you should have any questions regarding this submittal or require additional information, please contact Mr. Bill R. Hansher at 402-533-6894.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 4, 2016.

Louis P. Cortopassi Site Vice President and CNO LPC/JKG/mle

Enclosures:

1. Reference List

2. Fort Calhoun Station, Unit No. 1 Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Issue (GSI) 191, Pressurized Water Reactor Sump Performance

3. Update of OPPD Letter LIC-1 3-0058 and Revised Regulatory Commitments

4. Regulatory Commitments c: M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector

LIC-1 5-0100 Page 1 Reference List

1. Letter from OPPD (Richard P. Clemens) to NRC (Document Control Desk), "Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated February 29, 2008 (LIC-08-0021) (ML080650369)

2. Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk), "Fort Calhoun, Unit 1 - Notice of Completion of Corrective Actions taken in Response to Generic Letter 2004-02 and Response to Request for Additional Information (RAI) for Fort Calhoun Station (ECS) Unit No. 1," dated October 16, 2008 (LIC-08-0107) (ML082960244)

3. Letter from NRC (Lynnea Wilkins) to OPPD (D. J. Bannister), "Fort Calhoun Station, Unit No. 1 -Request for Additional Information RE: Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors (TAC No. MC4686)," dated February 12, 2010 (NRC 0011 ) (ML100150072)

4. Email from OPPD (Donna L. Lippy) to NRC (Lynnea Wilkins), "OPPD Draft GL 2004-02 RAI Responses for FCS," dated April 30, 2010

5. Letter from NRC (L. Wilkins) to OPPD, "Summary of June 30, 2010 Category 1 Teleconference Meeting with Omaha Public Power District on Generic Letter 2004-02 (TAC No. MC4686)," dated July 29, 2010 (NRC-I10-0059) (ML101940301 )

6. Email from OPPD (Donna L. Lippy) to NRC (Lynnea Wilkins), "OPPD's Revised Draft Responses to the Generic Letter 2004-02 Requests for Additional Information (TAC No.

MC4686)," dated October 1, 2010

7. Letter from OPPD (L. P. Cortopassi) to NRC (Document Control Desk), "Proposed Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, "Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. 1," dated May 15, 2013 (LIC-13-0058) (ML13136A047)

LIC-1 5-0100 Page 1 Fort Calhoun Station (FCS) Revised Responses to Requests for Additional Information (RAIs) from Generic Letter (GL) 2004-02 Supplemental Response dated October 16, 2008 The Omaha Public Power District's (OPPD's) response to the Nuclear Regulatory Commission's (NRC's) draft request for additional information (RAI) transmitted on February 12, 2010 (Enclosure 1, Reference 3) is shown below. OPPD's response to the NRC RAI was informally transmitted in April 2010 (Enclosure 1, Reference 4) and again in October 2010 (Enclosure 1, Reference 6) following a June 2010 teleconference (Enclosure 1, Reference 5). As requested during a June 29, 2015 teleconference with NRC staff, this letter is provided to formally docket and update the information emailed to the NRC in April and October 2010.

RAI #3 Question:

The NRC staff requested that the licensee identify the source of the test data used to support the debris size distribution assumed for calcium silicate insulation and compare the banding, jacketing, and manufacturing process for the calcium silicate installed at FCS to the material used for the destruction testing. The licensee's response dated October 16, 2008, identified that the assumed debris size distribution was based on testing conducted by Ontario Power Generation, Inc. (OPG) and an analysis of this data contained in Appendix II to the December 6, 2004, safety evaluation (SE) on Nuclear Energy Institute (NEI) 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology" (the safety evaluation and NEI 04-07 are non-publicly available). The response also discussed the insulation jacketing and the manufacturing process for the calcium silicate base material. The staff did not consider the response to have fully addressed the RAI because the bandingllatching design of the FCS calcium silicate and the jacketing thickness were not described sufficiently and compared to the corresponding design values for the OPG test debris. Please provide this additional design information and comparison with the OPG test.

NOTE: Despite the discussion in Appendix II of the NRC's SE on NEI 04-07, based on scaling issues similar to those raised in the review of recent Pressurized Water Reactor Owners Group (PWROG) zone of influence (ZOI) testing, it is not apparent that the scaling of the OPG testing (i.e., primarily the scaling of the relatively small jet from a 2.86-inch-diameter nozzle to the large 48-inch-long target at the distances tested) provides prototypical or conservative results with respect to the characteristic debris size distribution that would be expected under plant conditions. The NRC staff has not yet determined the significance of this issue. Should it be determined significant, the NRC may raise this as a future issue with FCS testing in the resolution of GL 2004-02.

OPPD RAI #3 Response:

The FCS calcium silicate piping insulation is currently not installed in a manner that is equivalent to the configuration of the insulation that was used in the OPG testing. To demonstrate the equivalency, and to continue to use the approach discussed in the previous RAI response (i.e., , Reference 2), OPPD will be installing additional banding to calcium silicate insulation within the zones of influence which support the ECS debris generation analysis. The following

LIC-15-0 100 Page 2 table compares final FCS calcium silicate piping insulation jacketing and proposed banding to what was used in the OPG testing:

FCS { OPG Testing Jacketing Material Aluminum Aluminum Jacketing Thickness 0.020" 0.016" Banding Material Stainless Steel Stainless Steel Banding Thickness 0.020" 0.020" Banding Fastener Type Crimped Crimped Spacing of Banding < 8.25" > 6.5" & < 8.25" OPG testing of calcium silicate insulation used aluminum jacketing 0.016" thick. ECS aluminum jacketing is 0.020" thick, which is bounded by jacketing thickness used for OPG testing. OPG testing used stainless steel bands with a thickness of 0.020" secured by standard crimp connectors with a spacing ranging from 6.5" to 8.25". FCS will install additional banding material in a manner similar to the configuration in the OPG Testing with 0.020" bands spaced less than 8.25 inches which will bring the configuration at ECS to a similar configuration as that used in the OPG Testing. A significant amount of banding will need to be added in many locations that are in inaccessible, high radiation areas.

As noted in Reference 3, OPPD was requested to describe the bandwidth spacing selected for the modification. The modification (i.e., EC 49722) will:

1. install stainless steel bands on jacketed calcium silicate insulation in containment (within a prescribed zone of influence inside the SG bays),

2. replace existing fiberglass insulation on the 3" pressurizer spray line near emergency core cooling system (ECCS) strainers with reflective metal insulation (RM I), and

3. replace existing calcium silicate insulation containing asbestos on reactor coolant pump RC-3D with RMI.

Use of stainless steel banding on jacketed calcium silicate insulation allows the existing insulation and jacket to remain in place and eliminates the need to dispose of it. The banding will be placed at nominal 6-inch spacing, which is similar to the spacing used in OPG testing.

RAI #5 Question:

The NRC staff requested that the licensee address the potential for operation of a low pressure safety injection (LPSI) pump following the switchover to recirculation (either due to the potential single failure of an LPSI pump to trip or through procedurally permitted operator actions) and any consequent effects of this increased flow on post loss-of-coolant accident (LOCA) debris transport. By letter dated October 16, 2008, the licensee stated that a "turbulent jet analysis" had been performed, which showed no additional debris entrainment in the flow, but showed the potential for additional tumbling transport.

The licensee also indicated that the only scenario where the LPSl pump would be procedurally operated post-LOCA would be for alternate hot-leg recirculation, under which

LIC-1 5-0100 Page 3 conditions the sump flow rate would be less than the currently bounding analyzed flow rate. The NRC staff questioned the adequacy of this response because, for the case of a LPSl pump single failure to trip, (1) it was not apparent to the staff that a turbulent jet analysis is appropriate for modeling the flow in question, (2) sufficient detail concerning the turbulent jet analysis was not presented, (3) sufficient information was not presented to justify the unexpected conclusion that the additional flow from a LPSl pump would not lead to additional debris remaining in suspension in the flow to the strainers, (4) it is not apparent that the potential for additional debris to transport to the strainer via tumbling with the additional flow from a LPSl pump can be neglected, and (5) it is apparent that a break in the steam generator (SG) A compartment could not become more limiting if the additional pool turbulence created by the flow from an LPSI pump were to result in additional debris remaining in suspension. Please provide additional information that sufficiently addresses these remaining items.

OPPD RAI #5 Response:

During the June 30, 2010 teleconference, the NRC staff suggested that OPPD identify the expected time needed to secure the LPSI pump and provide a basis for the time chosen. Pending review by the NRC, OPPD will revise its position regarding the potential for increased debris transport due to an LPSI pump failure to trip. OPPD will no longer credit the turbulent jet analysis referenced in the PAI. As a result, the detail requested for this RAl is no longer applicable. The following discussion provides information for appropriate closure of this issue.

The Fort Calhoun Station (FCS) Emergency Operating Procedures (EOPs) are written in a two-column format, with instructions in the left hand column and contingency actions in the right hand column. If the instruction cannot be successfully accomplished, the operator immediately carries out the contingency action. FCS EOP-03's mitigation actions associated with the failure of a LPSI pump to trip post-RAS are in compliance with CEN-152, "Combustion Engineering Emergency Procedure Guidelines."

EOP-03, "Loss of Coolant Accident," R38 currently prescribes the following actions in response to initiation of recirculation:

EOP-03 Page 40 of 74 INSTRUCTIONS CONTINGENCY ACTIONS Sc32. IF SIRWT level falls to 16 inches, 32.1 IF RAS is NOT actuated by STLS, THEN verify that STLS initiates RAS by THEN perform step a or b:

performing the following:

a. IF any of the STLS relays have

a. Verify ALL of the following STLS NOT tripped, relays have tripped: THEN manually initiate using the AI-30A/B key on the STLS Test S 86A/STLS S

86B1/STLS Switches.

0 86B/STLS 86A1 /STLS S

• 86A/STLS Test Switch

• 86B/STLS Test Switch

b. Verify ALL. of the following RAS relays have tripped: b. Manually establish RAS flow path by performing the following:

0 86A/RAS S

86B1/RAS 0

86B/RAS 1) Open both SI Pump Suction S

86A1 /RAS Containment Isolation Valves.

• HCV-383-3

c. Verify BOTH of the following

• HCV-383-4 valves are open:

• HCV-383-3, SI Pump Suction 2) Clos.__e both SI Pump Suction Containment Isolation Valve

• HCV-383-4, SI Pump Suction S IRWT Isolation Valves.

Containment Isolation Valve

• LCV-383-1

• LCV-383-2 (continue) 3) Ensure both LPSI Pumps stop.

• SI-1A

• SI-lB (continue)

SContinuously Applicable or Non-Sequential StepR3 R38

LIC-15-0 100 Page 4 EOP/AOP Floating Steps, R7 contains the following guidance for LPSI stop and throttle criteria:

EOP/AOP FLOATING STEPS Page 25 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. IF RAS has initiated, 1.1 IF EITHER LPSI pump is still running, THEN ensure both LPSI pumps, THEN perform any or all of the SI-1A/B, stop. following steps a, b, or c:

a. Throttle LPSI Injection Valves to approximately 250 gpm flow.

b. (LOCAL) OQpen running LPSI pump breaker by performing the following (Switchgear Rooms):

1) Place the 69 permissive switch to "PULL TO LOCK":

• 69/SI-IA, "PERMISSIVE CONTROL SWITCH SI-IA BREAKER" (1A3-7)

• 69/S1-1IB, "PERMISSIVE CONTROL SWITCH SI-lB BREAKER" (1A4-14)

(continue)

2) IF SI-lA was running, THEN press the "OPEN" push button.

(continue)

R7

EOP/AOP FLOATING STEPS Page 26 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 .b. (continued)

3) IF SI-I1B was running, THEN press the "MANUAL TRIP" push button.

c. IF SI-IA breaker is to be racked down, THEN perform the following:

1) De-energize Bus 1A3.

2) (LOCAL) Remove dust cap and install jumper into breaker door receptacle.

3) (LOCAL) Install motor plug (continue) adapter and limitilt shorting plug into the junction box.

4) (LOCAL) Install elevator motor into cubicle.

5) (LOCAL) Place the elevator motor select switch to lower.

(continue)

R7

EOP/AOP FLOATING STEPS Page 27 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 .0. (continued)

6) (LOCAL) Pull clutch handle until elevator motor limits out.

d. IF SI-I1B breaker is to be racked down, THEN perform the following:

1) De-enerqize Bus 1A4.

2) (LOCAL) Install elevator motor in SI-lB cubical.

3) (LOCAL) Place elevator motor selector switch to "LOWER".

4) (LOCAL) Pull clutch handle until elevator motor limits out.

e. IF both Vital buses are deenergized, THEN GO TO EOP-20, Functional Recovery Procedure, MVA-AC.

R7

EOP/AOP FLOATING STEPS Page 28 of 121 2.0 FLOATING STEPS B. LPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS

2. IF RCS pressure is greater than 200 psia and controlled, THEN storp both LPSI Pumps, SI-IA/B.

3. IF RCS pressure is less than 200 psia, AND RAS has NOT occurred, THEN initiate LPSI flow by performing the following:

a. Start both LPSI Pumps, SI-IA/B.

b. Ensure all of the LPSI Loop Injection Valves are open.

R7

LIC-15-0 100 Page 5 EOP Floating Step B was revised to first throttle the LPSI valves (1.1 .a above) and then locally open the LPSI pump supply breaker by manually tripping the breaker and/or racking down the breaker (1.1 .b, 1.1 .c respectively). By throttling the LPSI valves, the total emergency core cooling system (ECCS) flow through the ECCS suction strainer is reduced to less than the strainer design maximum flow.

The recirculation actuation signal (RAS) is automatically generated by the engineered safeguards actuation system when indicated safety injection and refueling water tank (SIRWT) level drops to 16 inches. In addition to the operator monitoring SIRWT level following the initiation of the event, when level in the SIRWT drops to 16 inches, an annunciator on the control room panel will alarm, alerting the operator to this condition. Since throttling the LPSI valves is the first action the operator would take following the SIRWT level dropping to 16 inches, and these actions can be accomplished from inside the control room, these actions would be completed within a few minutes of an RAS.

During preparation of this response it was identified that EOP floating step 1.1 .a, throttling the LPSI loop injection valves, is an optional step. As it appears that this action should be mandatory, OPPD will disposition this issue in its corrective action program.

RAI #7 Question:

The NRC staff requested that the licensee provide further justification to support the assumption that only 5 percent of fine fibrous and particulate debris blown to upper containment would be washed down to the containment pool (e.g., by condensate drainage). The RAI stated that the staff considered 10 percent to be a more appropriate number and cited NUREG/CR-6762, "GS1-191 Technical Assessment Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance," dated August 2002 (ADAMS Accession No. ML022470077), which found that 5 percent was a low estimate for washdown without spray and 10 percent was a high estimate. In its response dated October 16, 2008, the licensee stated that the assumption of 5 percent washdown from condensate flow was based on assumed condensate flow from the volunteer plant study in Appendix VI, "Detailed Blowdownl~ashdown Transport Analysis for Pressurized Water Reactor Volunteer Plant," to the NRC SE on NEI 04-07. The response further discussed the lack of post LOCA containment spray operation and the effect of fan cooler operation. The licensee also stated that a washdown percentage of 10 percent for the debris loading in SG B compartment would be bounded by the more-limiting debris loading from SG A compartment (with a washdown percentage of 5 percent). Based on the factors below and engineering judgment, the NRC staff considers it prudent to assume a 10 percent washdown, consistent with the technical guidance discussed above.

a. The condensate flow from a plant without containment sprays would be significantly greater and longer-lasting than the volunteer plant, which would presumably lead to an increase in the transport fraction due to condensate drainage versus the volunteer plant.

b. Appendix VI to the safety evaluation, as well as the Drywell Debris Transport Study (NUREG/CR-6369) results on which many of its assumptions are based, indicate that substantial uncertainties are associated with the transport estimates, which applies

LIC-15-0 100 Page 6 particularly to blowdown and washdown. The washdown percentages in Appendix VI did not factor in uncertainties typically accounted for in licensing basis calculations and were based on the assumption that most containment surfaces would be sprayed, and few would be subject to condensate drainage only.

c. To support a washdown transport fraction that may not contain significant margin to account for uncertainties, it is important that the blowdown transport of fine debris to the upper containment be performed in a sufficiently conservative manner. Given that the FCS containment contains significant solid flooring, which could potentially impede its transport to upper containment, it is not apparent that the assumed quantity of fine debris (80 percent) would actually be capable of reaching the upper containment during the blowdown phase to subsequently be retained there.

d. Some of the fine debris assumed to be inertially held up on vertical surfaces or the underside of horizontal surfaces during blowdown will lose adherence over time and may gravitationally float down to the containment pool. It was not apparent that this phenomenon was accounted for in the 5 percent washdown assumption made by the licensee.

e. Although, in response to the NRC staff's concerns about washdown uncertainties, the licensee stated that acceptable strainer performance could be assured when assuming 10 percent washdown for a non-limiting break scenario, it was not apparent to the staff whether 10 percent debris washdown could be tolerated for the limiting strainer debris loading case.

The impact on the debris source term of doubling (for example) the assumed washdown percentage would be significant for FCS. Based on this consideration and the questions above, please provide additional information that addresses these remaining items or demonstrate that the strainer will perform acceptably if a washdown transport fraction for fine fibrous and particulate debris that incorporates adequate allowance for uncertainties is assumed.

OPPD's RAI #7 Response:

In the Enclosure to OPPD letter LIC-08-0107 dated October 16, 2008 (Enclosure 1, Reference 2), OPPD supplied a response to RAl #7. The response to RAl #7 described OPPD's basis for the conservatism of using a 5% washdown transport fraction for OPPD's postulated large break post-LOCA operation. OPPD's post-LOCA operation does not employ containment spray.

In its reply to the OPPD letter (Enclosure 1, Reference 2), the NRC staff provided additional clarification of their RAl and stated that the staff continues to consider it prudent to assume a 10%

washdown transport fraction, and requested that OPPO provide additional information to adequately address the staff's remaining items, or use a washdown transport fraction for fine fibrous and particulate debris that incorporates adequate allowance for uncertainties (i.e., 10%).

In lieu of development of additional analyses in support of the 5% washdown transport fraction, OPPD will incorporate a 10% washdown transport fraction into the currently in progress revision of the debris transport analysis for FCS. OPPD considers this item to be closed with no additional actions required.

LIC-1 5-0100 Page 7 RAI #10 Question:

The NRC staff requested that the licensee provide further justification to resolve the staff's long-standing concerns (including those raised during the staff's pilot audit of FCS) associated with the use of a Stokes' Law approach and turbulent kinetic energy (TKE) metrics to determine the settling behavior of fine debris. In its October 16, 2008, response to this RAI, the licensee provided an extensive discussion that responded to each part of the staff's question. However, after reviewing the response, the staff continues to have concerns that the licensee's approach has not been adequately justified on a technical basis. The NRC staff's concerns are summarized below. Please provide information to address these concerns

a. The licensee's TKE metrics do not appear to have been experimentally benchmarked.

Benchmarking would likely show that a single metric for what is in reality a distribution of particulate sizes to be insufficient. Acoustic measurements of velocity made at low flow rates are not directly related to the fundamental phenomena governing suspension and settling (phenomena which are associated primarily with turbulence).

Therefore, the NRC staff did not agree that such measurements would be sufficient to resolve this concern.

b. The licensee did not sufficiently address uncertainties in the TKE model(s) in the computational fluid dynamics (CFD) code used to provide confidence that the TKE model used was adequate for the purpose of determining the settlement of fine debris.

The licensee's October 16, 2008, response dealt primarily with uncertainty in velocity (both computational and experimental), not TKE.

c. The licensee did not adequately justify the specification of shape factors and drag coefficients that are applicable for modeling the settling behavior of spherical objects in quiescent fluid to model the settling behavior of irregularly shaped debris, including fibers, in the plant containment pool. Although a detailed discussion was provided in the licensee's October 16, 2008, letter in response to this point, the NRC staff does not agree with certain arguments being made. Specifically, the NRC staff does not agree that shape factors and drag coefficients are not relevant to the settlement of irregularly shaped debris. Sufficient information was not presented to justify this conclusion. It is not apparent that shapes formed by strands of fibers would behave similarly to a spherical particle. In addition, sufficient justification was not presented to show that the presence of turbulence would not affect the TKE metrics that were calculated for quiescent conditions. The staff also did not agree with the licensee's explanation for the drag force being insignificant for slowly settling fine debris because it is the significance of drag (or the drag coefficient) for the fine debris (relative to the other forces acting on the fine debris) that leads to the very small settling velocity.

d. The licensee did not sufficiently justify the correlation of terminal settling velocity to TKE. In response to this item, the licensee indicated that the "TKE" used by Alion Science and Technology (Alion) was based on a bulk velocity, rather than the fluctuating components of the velocity. Based on its review of the licensee's October 16, 2008, submittal, it was not apparent how this quasi-TKE quantity was implemented in the CFD code and analysis, how it is consistent with CFD plots that were included in

LIC-1 5-0100 Page 8 the supplemental responses to GL 2004-02, and whether several of the conclusions following from the use of this quasi-TKE quantity are valid. The NRC staff requests further discussion regarding the use of 0.01 feet per second (ft/s) as a transport metric for fines, the isotropy of turbulence, and the correlation of terminal settling with TKE.

OPPD RAI #10 Response:

OPPD's implementation of the Pressurized Water Reactor (PWR) Water Management Initiative is targeted at both increasing the water available for initial safety injection and reducing the amount of debris transported to the ECOS sump strainers. Reduction in the amount of debris transported to the strainer reduces the reliance on the strainer for maintaining recirculation during the 30-day mission time. Relative to OPPD's initial response to this RAl and ongoing dialogue with the NRC, there is reasonable assurance that the combination of a high recirculation pool water level, low recirculation flow rates, and containment geometry provide a condition where flow in the recirculation pool is not sufficient to transport fine debris from all areas of the pool.

As a result of discussions between the NRC and OPPD that occurred on June 30, 2010, OPPD is addressing the issues raised in the February 2010 NRC letter (Enclosure 1, Reference 5) by providing reasonable assurance that portions of the containment post-LOCA pool will not communicate with the ECCS sump strainer for a large break (LB) LOCA. This is accomplished with CFD simulation data that shows that the recirculation pool geometry (represented in Figures 1 and 2 below) 1 . The two SG compartments are separated from each other and flow from one compartment (the break compartment) to the strainer results in a condition where flow from the compartment opposite of the break does not communicate with the strainer. Hence, there is no driving force to transport debris from the compartment opposite of the break to the strainer.

Following an LBLOCA, it is conservatively assumed that insulation debris (RMI, low density fiberglass (LDFG), and TempMat) are uniformly distributed between the locations where it would have been destroyed and the sump strainers. This is a conservative assumption since the blowdown and pool fill-up phases involve multi-directional flows that would tend to disburse debris around containment (including areas with lower transport potential). This assumption is also valid for no spray flow conditions since the initial distribution of debris is driven by the blowdown and pool fill-up transport rather than the washdown transport and is established prior to recirculation.

1 Note all CFD figures illustrate only the old sump strainers. For conservatism, the actual new strainer geometry and surface area was not modeled. The old strainers were used in the CFD model since more localized flow around the strainers in the original design potentially increased debris transport fraction than what would be expected with the current installed design. Modeling the original strainer design represented a conservative transport fraction estimate and with the installation of the new larger surface area strainers did not jeopardize the conservatism of the transport calculation. As such, all figures illustrate a model with the original strainers for conservatism only. The new strainers have considerably larger strainer areas and thus the flow velocities around the new strainers will be lower than they were in the old design.

LIC-1 5-0100 Page 9 Figure 1: Location of Sump Strainers and Break Inside SG Compartment A The recirculation occurs in a pool with a depth of approximately 5 feet and a recirculation pool flow of less than 1000 gpm. Figure 2 presents a visual representation of the turbulence data present in the recirculation pool where the yellow represents the regions of the pool where the turbulence value is only half the amount required to keep fiber fines in suspension. The flow velocities, combined with the turbulence data, support the fact that the recirculation pool is essentially stagnant in many regions of the pool, especially in the SG compartment opposite of the break.

LIC-1 5-0100 Page 10 Figure 2: Half of the TKE Metric Required to Suspend Individual Fiber The dominating velocities and TKE for the analysis are within Compartment A where the break occurs and throughout the containment pool in the path to the sump strainers, as shown in Figure

3. The pool within Compartment B and around its doorway has very little flow and can be considered stagnant.

Turbulence is required to suspend and transport fines but as shown in Figure 2, because TKE in Compartment B is less than half required to keep fiber in suspension, it would therefore settle to the floor.

LIC-1 5-0100 Page 11 Figure 3: Velocity Magnitude and TKE for Break in SG Compartment A Vel. Mag.

5 0.120 I:00 TKE Ž 0.034 ft2 fs2 To further substantiate the lack of flow associated with the SG compartment opposite of the break compartment, an additional evaluation of the CFD simulation data at the entrance to the SG compartment of the break compartment was performed.

For the case of a line break in SG Compartment A, the flux of the containment pool at several locations inside the containment building was analyzed to determine debris transport areas. Flux is defined as the amount of fluid flowing through a unit area per unit time.

To analyze the pool flow patterns, six different surface planes were created throughout the containment building as shown in Figure 4, to determine fluid flux magnitudes of the containment pool at various doorways and corridors resulting from a break inside SG Compartment A. The flux planes are hypothetical surfaces set perpendicular to containment pool circulation located at the two compartment doorways, near the sump strainers, and in the flow path from Compartment A to the sump strainers. For example, Flux Plane 1 was analyzed to determine flux metrics through the doorway of Compartment A and Flux Plane 6 was analyzed to determine flux metrics through the doorway of Compartment B.

LIC-1 5-0100 Page 12 Figure 4: Location of Surface Planes where Flux of the Containment Pool was Analyzed

LIC-15-0 100 Page 13 Figure 5 shows results of the analysis at the location of flux plane 6 in which velocity magnitude was integrated through the SG Compartment B doorway. The figure shows the velocity magnitude at a specific point in time in order to illustrate very low flow metrics at the doorway, with an average velocity magnitude of 0.005 ft/s (into the compartment) at 1,850 seconds.

Figure 5: Integration on Flux Surface 6 for Evaluation of Velocity Magnitude and Turbulence through Doorway to SG Compartment B at 1850 Seconds

LIC-1 5-0100 Page 14 The plot in Figure 6 shows the flux magnitude over time for the six flux surfaces illustrated in Figure 4. Positive and negative flux values correspond to direction of flow through the doorways and corridors. A positive flux represents the containment pool flowing in the positive x and y directions towards the break and negative flux represents the containment pool flowing in the negative x and y directions away from the break. Figure 6 shows flow consistently moves from the break to the strainer as demonstrated by the flux values for surfaces 1, 2, and 3. Figure 6 also shows that little or no flow occurs in areas of the pool that are not in the expected path of flow from the break location to the strainer. This is demonstrated by the fact that the fluxes in the containment pool at or near the doorway to Compartment B through surfaces 4, 5, and 6 have the lowest magnitude and fluctuate about zero with a maximum of 0.5 and -0.5 ft3/s. This shows that water does slosh in and out of Compartment B over time, but at a low rate which is not likely to transport debris.

The CFD model demonstrates that since there are no sprays or other flow sources in Compartment B to push debris out of the compartment, there is only minor "sloshing" of water in and out of the compartment and this "sloshing" would result in no significant transport of debris from Compartment B to the ECCS strainers. Therefore, use of a reduced transport fraction in the FCS debris transport analysis which accounts for the lack of transport from SG compartment opposite of the break compartment is acceptable.

LIC-1 5-0100 Page 15 Figure 6: Comparison of Flux at Six Surface Planes in the Containment Pool Resulting from a Break in Compartment A Fluid Flow At Flux Planes Time (s) 1100 1300 1500 1700 1900 2100 2300 2500 1.00E+O00i 5.00E-01

-5.00E-01

-4e-fflux surface 1

-- fflux surface 2 ffux surface 3 E.-1.00E+00 x =X-*-fflux surface 4

• -**fflux surface 5

--*-fflux surface 6

-1.50E+00

-2.00E+O0

-2.50E+00 --

-3&00E+O0

LIC-1 5-01 00 Enclosure 2 Page 16 RAI #15 Question:

The NRC staff requested additional information on the temperature extrapolation methodology because it is not apparent that the bore holes that occurred at the low test temperatures would occur at the higher temperatures in the actual sump pool. The licensee provided additional information in its October 16, 2008, response regarding the methodology used for temperature extrapolation. The methodology used is considered valid for the pressure range in which the temperature measurements were taken. However, the temperatures which the results are being extrapolated to are much higher than those tested. The higher LOCA temperatures result in significantly lower head losses than those measured at the test temperature. The licensee provided graphs of head loss, flow, and temperature for some of the tests. The licensee stated that, based on short-term head-loss variation, the formation and filling of bore holes appeared to be about the same between the high-temperature testing, about 95 degrees Fahrenheit (0 F) and the low-temperature test, about 65°F. The same head-loss curves appear to indicate, based on short-term head-loss variation from the average, that bore holes form at higher head losses and may not be present or have as large an effect at lower head losses. Two of the examples appeared to result in larger short-term variations in head loss at about 40 to 50 inches of head loss, and one example appeared to result in larger variations at about 20 inches of head loss. Therefore, the staff cannot conclude that bore holes would form at lower differential pressures. The licensee should justify that the reduction in head loss due to bore hole formation would be expected to occur at lower differential pressures. It is likely that the debris bed morphology has a significant impact on bore hole formation.

This could explain some differences in the short-term head-loss variation between tests.

The licensee should provide additional information that justifies that the temperature compensation applied to the test results does not affect the evaluation non-conservatively.

OPPD's RAI #15 Response:

The design of the FCS strainers includes two distinct design bases, one for a large break (LB)

LOCA condition and one for a small break (SB) LOCA condition.

LBLOCA:

Extension of LBLOCA test results to the plant strainer does not include temperature-based scaling or extrapolation of head loss; flow through the strainer is assumed to be completely turbulent due to observed bore hole formation during testing. The differential pressures that lead to bore hole formation in the LBLOCA condition are the same in the plant as they are in the test. This issue has been resolved.

SBLOCA:

The debris load for the FCS strainers in an SBLOCA condition is very low in fiber; the maximum nominal fiber bed thickness is less than 1 / 1 6 th of an inch. Since fiber does not accumulate uniformly, there were areas of the strainer 6 inches x 1 inch in size that were left uncovered, as well as, evenly covered areas.

Per the March 2008 NRC Guidance:

LIC-1 5-01 00 Enclosure 2 Page 17 If [boreholes havel occurred, the maximum head loss value determined during testing (at lower than accident temperatures) may be assumed to be the head loss value for the strainer. If indications of channeling were observed during testing, any correction for temperature should be justified from both the perspective of laminarand turbulent flow regimes, and from the perspective that the head loss could have been higher if debris bed morphology changes due to differential pressure had not occurred.

The following paragraphs address RAI #15 within the bounds of the March 2008 guidance:

After head loss had reached steady state in the limiting test, General Electric (GE) reduced the water temperature while maintaining flow. This temperature reduction served two purposes, it indicated that bore hole formation was relatively independent of temperature, and it measured the laminar/turbulent distribution of head loss by adjusting the fluid viscosity while maintaining the approach velocity constant.

Bore hole formation is indicated by the high-frequency oscillations of the measured head loss in the limiting SBLOCA test case, 11 M-SBLOCA-Jacketed. The head loss oscillations are indicative of the formations and closing of bore holes, and are usually associated with a high-calcium silicate, low fiber debris load like that at FCS. Bore holes are clearly formed at head losses below the maximum allowable plant head loss of 4.79 feet (57.5 inches) of water, and this is particularly apparent when the repeat tests, 8M-RPT and 9M-RPT, are considered (see below).

Although the test was performed at a lower temperature, i.e., higher viscosity and differential pressure than the limiting conditions at the plant (196.6°F), bore hole formation is predicted to occur at both temperatures and pressures. In the limiting test case, 11 M-SBLOCA-Jacketed, bore hole formation begins to occur as soon as head loss begins to become measurable, (as shown in the figure below). The amplitude of the effect of bore holes on head loss increases with measured head loss until reaching a maximum at a measured head loss of 50-60 inches of water, where the bore hole head loss oscillation amplitude remains approximately constant at 8 inches of water for the duration of the test.

The debris bed shift that occurs at approximately 1100 minutes is typical of large-scale head loss testing, and is not indicative of bore hole formation.

Although the SE of NEI 04-07 indicates that temperature scaling should not be used directly in cases where bore hole formation is apparent, by using viscosity scaling with consideration of the laminar/turbulent head loss distribution, the measured head loss can be extrapolated to the plant sump temperature.

Plots of head loss during repeat tests 8M-RPT and 9M-RPT are included below to show that bore hole behavior occurs even at lower head losses than in test 11 M-SBLOCA. Note that tests 8M and 9M were exact replicates of test 11 M-SBLOCA, and stabilized at lower head losses with significant bore hole activity. Tests 8M and 9M were performed after test 11 M.

LIC-1 5-0100 Page 18 Figure 15.1 - Head Loss Plot for Test 11IM-SBLOCA-Jacketed 1 I I I I 1 11 M-,SBLOCA-Jacketed 0

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Figure 15.2 - Head Loss Plot for Test 8M-RPT

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35 Final chem 251st Fiber ctem Final fiber *

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Figure 15.3 - Head Loss Plot for Test 9M-RPT 35, Terminate ------

20,25 'Finnal rhem o Fininllfiber

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0 200 400 600 800 1000 1200 1400 1600 1800 Time (mai)

LIC-15-01 00 Page 19 RAI #20 Question:

The NRC staff requested additional information regarding how the velocities and turbulence in the test flume compare to similar variables predicted in the plant sump pool (i.e., for tests that allowed settling (no stirring), provide a comparison of the flows predicted around the strainer in the plant and the flows present in the test flume during the testing). In its RAI response, dated October 16, 2008, the licensee provided additional descriptions of how the velocities in the area of the strainer were determined. A CFD analysis of the sump pool, run at a higher flow rate than would be expected using the current assumptions, indicates that the average flow in the area of the strainers is about 0.075 feet per second (ft/sec). Adjusting this for the currently assumed flow rate, the velocity in the area of the strainer is calculated to be about 0.05 ft/sec. The RAI response stated that the flow in the test apparatus was set to be equal to the predicted flow rate by adjusting the width of the test flume walls. The basic concepts applied in the RAI response are accepted by the staff. However, it appeared that the CFD analysis referenced by the response had some non-uniform flow areas near the strainer. Areas of higher flow would promote transport. Additionally, the RAI response did not address the turbulence available to maintain debris in suspension. The licensee should provide a more detailed CFD analysis in the area of the strainer so that flow velocities would be more fully defined. In addition, the licensee should provide a comparison of turbulence around the strainers in the plant and the turbulence in the test flume. Alternately, the licensee could provide alternate information that shows that the test was not non-conservatively biased due to these factors.

OPPD's RAI #20 Response:

OPPD must defer a response to RAI #20 until staff acceptance of the testing protocol discussed in RAI #10. The outcome may determine whether additional testing is required to be conducted, which is tracked by Commitment #2 in Enclosure 4.

RAI #25 Question:

The NRC staff requested additional information on the justification for treating unqualified alkyd original equipment manufacturer (OEM) coatings as chips at FCS despite the contradictory data presented in Electric Power Research Institute (EPRI) report #1011753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings" (OEM report), dated September 2005. The licensee stated that the alkyd OEM coatings fail in 5 mil chips. The staff does not accept alkyds failing as chips since in the EPRI OEM report, alkyds do not fail as chips. According to staff guidance ["NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing," dated March 2008 (ADAMS Accession No. ML080230038).], the alkyd OEM coatings failing as fine particulate would be more conservative since a thin bed has been observed to form. The impact of the alkyd OEM coatings could add an additional 40 pounds mass (Ibm) to the debris. Please provide additional justification for treating alkyd OEM coatings debris as chips rather than treating it as particulate.

LIC-1 5-0100 Enclosure 2 Page 20 OPPD's RAI #25 Response:

The original FOS Debris Generation Post-LOCA calculation (FC06985) identified 0.23 ft3 of alkyds with another alkyd coated component being installed (HE-44 telescoping crane - 0.25 ft3 estimate). Thus, a total of 0.48 ft3 of alkyd coatings were identified as potential alkyd coating sources in containment for the original debris generation calculations.

Since the original debris generation calculation was performed, the additional removal of some alkyd coated components in containment has been completed. Twenty (20) out of the twenty-five (25) lead blanket storage drums (alkyd coated) were permanently removed. The reactor head lifting device (alkyd coated) was replaced with the rapid refueling package component. The yellow jib crane, noted in the original calculation, is installed at elevation 1045 feet, outside of any jet ZOI. It is significantly higher than the containment sump pool elevation and is not subject to containment spray.

Also, since the original debris generation calculation was performed, telescoping cranes HE-44 and HE-48 have been added to containment. The two cranes are at elevation 1057 feet and thus would not be immersed or subject to containment spray. They are completely outside of the bioshield areas containing the direct jet impingement blasts. Due to interfering structures, components, and grating, HE-44 would not be subjected to a direct impingement blast. The coating system from the telescoping cranes was included as part of the original OEM EPRI testing program.

The only way that un-impinged alkyd coatings affect head loss at FCS (containment spray is not employed post-LOCA) is if they are submerged and subject to dissolution of the coating matrix.

OPPD recognizes that pigments in alkyd coating systems could potentially dissolve out of the coating matrix over time due to buffer or temperature effects. Thus, from the original debris generation calculation, the only sources of alkyds that could be subject to immersion are electrical junction boxes, Limitorque valve operators, detector well cooling fans, reactor coolant drain tank (RCDT) pumps, five lead blanket storage drums. This amounts to approximately 0.17 ft 3 of plant alkyd material (16 Ibm).

As described in the draft responses to RAIs #34 and #35, OPPD modified the nuclear detector well cooling units during the spring 2011 RFO; which reduced the post-LOCA aluminum inventory in the FOS containment sump and the resulting amounts of chemical precipitates. The resolutions for the LBLOCA cases are described below:

LBLQCA: The existing FCS strainer qualification test data is based on a plant equivalent of 647.3 lb of sodium aluminum silicate, and after the spring 2011 REO, there is a preliminarily calculated maximum of 109.8 lbs of sodium aluminum silicate in the FCS containment sump. This results in a 537-pound sodium aluminum silicate surplus in the LBLOCA strainer design basis, and this surplus more than compensates for an additional 16 lb of particulates added to the debris load to account for unqualified coatings.

SBLOCA: The existing FCS strainer qualification test data was based on a plant equivalent of 10.6 lbs of sodium aluminum silicate, and a plant equivalent of 6.0 lbs of aluminum oxyhydroxide.

With the completion of the spring 2011 outage, a significant source of aluminum has been removed. However, the impact of that aluminum removal related to chemical debris formation is still being evaluated related to the previous strainer testing. In conclusion, OPPD will treat potentially submerged alkyd coatings as a source of fine particulates in the manner stated above.

The unqualified and unsubmerged alkyd coatings in containment will be assumed to fail as chips;

LOIC-5-0100 Page 21 this is considered conservative based on bench-top test observations as well.

RAI #34 Question:

The NRC staff requested additional information regarding a number of issues related to silicate inhibition of aluminum corrosion, including among other information the type and amount of plant debris assumed as the source of silicates. In its October 16, 2008, letter, the licensee provided a table which listed the debris generated for various breaks including a small break. The table indicates that there is only a small amount of calcium silicate destroyed by the small break. Since silicate inhibition of aluminum corrosion is credited for all potential breaks, please provide greater detail on the source of the silicate for the small break. Specifically, please explain whether the silica source for a small-break LOCA is strictly calcium-silicate or whether fiberglass or other materials are considered a source of silicate. This point is important because the inhibition of aluminum corrosion may not occur if insufficient silicate is present for the small-break case.

OPPD's RAI #34 Response:

OPPD identified the need to modify the nuclear detector well cooling units at FCS. These units provided the largest contribution of aluminum to the analysis of chemical precipitate formation.

As a result, a revised WCAP-16530 assessment indicates that a substantial reduction in the formation of sodium aluminum silicate and aluminum oxyhydroxide is realized through removal of the cooling fins. FCS modified nuclear detector well cooling units, and with this modification, the aluminum source term within containment was significantly reduced. After the spring 2011 RFO, no credit will be taken for silicon inhibition, and thus this RAl will no longer apply to FCS. Although the NRC found OPPD's original response to this RAI acceptable, an interim draft response is included below.

The debris quantities that were outlined in the previous response (i.e., Enclosure 1, Reference 2) to the NRC were used to calculate chemical precipitate formation. The silicon-contributing debris sources for the small break cases consisted of low-density fiberglass, CaI-Sil, and a Cal-Sil/asbestos composite. The silicon release rate correlation from WCAP-16530 was applied to the SBLOCA debris loads to determine the resultant silicon concentration. No additional quantities of silicon were assumed to be released other than what was considered formed based upon strict utilization of WCAP-1 6530 as applied to the SBLOCA case, and OPPD utilized the methodology described in WCAP-1 6530 and WCAP-1 6785 to calculate the formation of chemical precipitates at FCS. The small break LOCA scenario from preliminary calculations is now dominated by the dissolution of aluminum in containment. During an upcoming refueling outage, OPPD plans (EC 49722) on replacing the pressurizer spray line fiberglass insulation with RMI, which will decrease the quantity of potential chemical debris.

RAI #35 Question:

The NRC staff requested additional information concerning how aluminum solubility was credited. This information was needed to determine if the long-term solubility credit is based on the pool temperature never dropping below 140°F. Based on review of the licensee's October 16, 2008, RAI response, it is unclear to the staff whether the licensee's

LIC-1 5-0100 Enclosure 2 Page 22 analysis applies the aluminum hydroxide precipitate at a delayed time or if the aluminum is assumed to remain in solution for the duration of the post-LOCA mission time. The licensee did not provide justification for the credit taken as discussed in the staff's March 2008 chemical effects review guidance ["NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant Specific Chemical Effect Evaluations," dated March 2008 (ADAMS Accession No. ML080380214).]. Please provide additional detail regarding the solubility of aluminum and how aluminum-based precipitates are accounted for, or discounted, in the final analysis.

OPPD's RAI #35 Response:

Only the small break (SB) LOCA portion of the previous response (Enclosure 1, Reference 2) is being revised for this submittal based on the discussion with the NRC during the June 30, 2010 teleconference.

OPPD modified the nuclear detector well cooling units at FCS during the 2011 refueling outage (RFO). These units provided the largest contribution of aluminum to the analysis of chemical precipitate formation. As a result, a revised WCAP-16530 assessment indicates that a substantial reduction in the formation of sodium aluminum silicate and aluminum oxyhydroxide has been realized through removal of the cooling fins for the LBLOCA bounding scenario. The modification to the nuclear detector well cooling units made during the spring 2011 REQ has significantly reduced the aluminum source within containment.

Table 35.1 -Predicted FCS Aluminum Precipitate Quantities Existing Aluminum Inventory' Tested Debris Quantities Revised InventoryAluminum (2Q11)

WCAP1653 & WAP-1530Preliminary WCAP-WCP150&WCAP-16530 WCP15016530 WCAP-1 6785 Surrogates (oslct niiin LBLOCA- SBLOCA LBLOCA LBLOCA SBLOCA RC2A RC2A SBLOCA LBLOCA SBLOCA RC2A Scenario Scenario Scenario Sodium aluminum 4.3 kg 286.1 kg 4.3 kg 327 kg 4.8 kg 293.6 kg 6.0 kg 47.3 kg silicate _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _ _ _ _ _ _ _

Almnm0 kg 0 kg 108.0 kg 33 kg 2.7 kg 0 kg 30.3 kg 0 kg oxyhydroxide _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _

To address the differences between the two limiting break scenarios, the response is provided in two sections:

LIC-1 5-0100 Enclosure 2 Page 23 LBLOCA:

The aluminum source term calculated by WCAP-1 6530 will generate zero aluminum oxyhydroxide and less sodium aluminum silicate than was tested with the FCS strainers and will provide sufficient margin such that the consideration of aluminum solubility is not required and the subject of this RAI is no longer applicable to the LBLOCA case. No credit for aluminum solubility is taken for the LBLOCA case.

SBLOCA:

From the summary of the June 30, 2010, teleconference, OPPD agreed to reevaluate OPPD's RAI response and either show the pH is greater than 7.5 or reevaluate solubility. OPPD has determined that the current Technical Specification sodium tetraborate (NaTB) Volume Required for RCS Critical Boron Concentration (ARO, HZP, No Xenon) curve supports a pH of 7.05.

Therefore, OPPD will reevaluate solubility.

With regard to possible increased predicted amounts of aluminum oxyhydroxide (SBLOCA),

OPPD plans to remove the fiberglass insulation from the pressurizer spray control line piping (see Figure 4 of Reference 1 ) and replace it with RMI (EC 49722) during an upcoming refueling outage.

This will eliminate the source of any transportable fibrous debris. Without transportable fibrous debris, the chemical pr'ecipitate formed is anticipated to pass through the strainer without forming a debris bed.

The RAl #35 response is based on Enclosure 1, Reference 7.

RAI #37 The latent fiber quantity assumed by FCS is only 2.79 percent, rather than the default suggested 15 percent from the NRC staff's SE on NEI-04-07. The NRC staff believes the apparent difference is due to fiber collection efficiency differences between the debris collection method used by FCS (scraping with a metal scraper) and vacuuming or wiping with masolin cloths used by the other plants in the NUREG-6877 survey. Although the collection efficiency of fiber is not discussed specifically, the NUREG notes that differences in collection method have a large impact and specifically notes that the metal scraper method resulted in a much lower fraction of fine particulate for FCS as compared to the other plants surveyed. Additionally, the fiber percentage of 2.79 percent is based on only eight samples (total mass: 27 grams), which is not enough for statistical accuracy for scaling a mass distribution up by a factor of over 2500. For properly collected debris samples, a fiber-mass proportion of 15 percent should be applied to the total inventory estimate in the absence of site-specific supporting evidence. Therefore, the NRC staff believes the licensee should have used the standard 15 percent value for the latent fiber mass percentage as opposed to relying upon the Plant C data in NUREG/CR-6877, "Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment Buildings," dated July 2005 (ADAMS Accession No. ML052430751). Please justify the use of 2.79 percent latent debris distribution fiber for FCS.

LIC-1 5-0100 Page 24 OPPD's RAI #37 Response:

The latent debris sampling procedure was enhanced by increasing the sample size from 8 to 25 samples and better defining sample locations. Samples are to be collected on both horizontal and vertical services as follows:

• 4 horizontal collection areas on the 1060-foot elevation

• 4 horizontal collection areas on the 1045-foot elevation

• 5 horizontal collection areas on the 1013-foot elevation

• 6 horizontal collection areas on the 994-foot elevation

• I vertical collection area on the concrete wall

• I vertical location area on the liner

• I vertical location area on piping/equipment

• 3 horizontal collection areas in large equipment collection areas The exact locations were chosen based on engineering experience, judgment, and knowledge.

The procedure cautions to ensure that areas are representative of debris type and accumulation.

Locations considered are those near robust barriers (i.e., each steam generator bay, annulus bay and the operating floor). Specifically, the LOCA impingement area inside the bioshield will be sampled. Latent debris path areas and areas that are in the surnp flow path will also be sampled.

Sample weights are averaged for each representative area and linearly extrapolated to the corresponding surface areas in containment. One hundred percent (100%) of the 994-foot elevation horizontal surface latent debris is included in the source term, since the 994-foot elevation is flooded after a LOCA. Only the 994-foot elevation is exposed to coolant, due to the lack of containment sprays during a LOCA at FCS, but additional latent debris sources are included in the source term for conservatism. One hundred percent (100%) of the vertical surface latent debris, 50% of the 1013-foot elevation horizontal surface, 50% of the 1045-foot elevation horizontal surface, and 25% of the 1060-foot elevation horizontal surface latent debris are included in the latent debris source term, even though these areas will generally not be exposed to post-LOCA coolant flow.

Latent debris sampling at FCS is conducted near the end of the outage but prior to Radiation Protection (RP) personnel commencing containment cleaning, so latent debris accumulation on the containment surfaces is at a maximum during sampling due to the latent debris generated by outage activities. Since RP cleaning does not occur prior to latent debris sampling, the timing of ECS latent debris sampling results in conservative latent debris quantities and conservatively bounds the latent debris quantities that remain in containment after RP cleaning has been completed. To confirm the conservatism of this collection timing, OPPD has taken samples near the beginning and end of the outage since 2011.

Although NEI 04-07 recommended an assumed fibrous debris fraction for latent debris of 15%,

this value is directed toward containments utilizing primarily fibrous insulation. The FCS containment contains almost exclusively calcium silicate insulation and reflective metallic insulation (RMI), so it is reasonable to expect that there is a reduced latent fibrous debris fraction as compared to a primarily fiberglass insulation containment. The fraction of latent fiber at ECS is determined using measurements of existing containment latent debris, rather than through the application of an assumed fiber fraction that is not applicable to ECS considering the absence of fibrous insulation in containment.

It is OPPD's position that the original survey information was properly obtained, is accurate, and

LIC-15-0 100 Page 25 is conservatively representative of actual conditions. Therefore, 2.79% is used in the analysis rather than a default of 15%. However, based on discussions with the NRC, OPPD will adopt a fibrous fraction for latent debris of 15%.

As noted in Reference 3, OPPD's April 2010 draft response addressed the NRC staffs concerns with regard to a 15 percent assumption and OPPD agreed to add a description of the debris collection method, which is as follows. Procedure SE-PM-AE-1005, "Latent Debris Collection Inspection" provides instructions to quantify debris inside containment to monitor the effectiveness of cleanliness programs and confirm assumptions made in the Debris Generation Calculation.

The procedure ensures that the amount of latent debris inside containment does not present a challenge to ECCS Sump operability following a LOCA.

Section 7.1 of SE-PM-AE-1 005 addresses determination of collection areas.

Four (4) horizontal collection areas are identified on the 1060-foot and 1045-foot elevation of containment. Five (5) horizontal collection areas are identified on the 1013-foot elevation of containment and six (6) horizontal collection areas are identified on the 994-foot' elevation of containment.

Section 7.3 of SE-PM-AE-1 005 addresses sample collection.

Debris samples are collected from the areas described above using mechanical means such as a nylon brush or putty knife to loosen debris for collection. Once the debris has been loosened, the Masslin cloth is wiped over the entire area to ensure that all debris in the sample area is collected. The sample debris and Masslin cloth are deposited in a collection bag where it is weighed with a highly accurate scale. Data analysis is then performed to ensure total containment latent debris load is less than 159 Ibm.

RAIs #39 through #47 The following set of issues relate to ZOI issues that were not specifically identified in the 2008 RAIs for FCS. These issues were developed as a result of NRC staff review of certain documents developed by the PWROG that are used as a basis for certain assumed ZOI reductions for FCS.

The PWROG is planning to respond to some of these generically, but it is unknown which of the issues below will be generically answered and which will be site-specific.

Questions:

39. Although the American National Standards Institute/American Nuclear Society (ANSlIANS) standard predicts higher jet centerline stagnation pressures associated with higher levels of subcooling, it is not intuitive that this would necessarily correspond to a generally conservative debris generation result. Please justify the initial debris generation test temperature and pressure with respect to the plant-specific reactor coolant system conditions in the plant hot-and cold-leg operating conditions. If ZOI reductions are also being applied to lines connecting to the pressurizer, please discuss the temperature and pressure conditions in these lines.

Were any tests conducted at alternate temperatures and pressures to assess the variance in the destructiveness of the test jet to the initial test condition specifications? If so, please provide that assessment.

40. Please describe the jacketing/insulation systems used in the plant for which the

LIC-1 5-0100 Page 26 testing was conducted and compare those systems to the jacketinglinsulation systems tested and demonstrate that the conditions and materials adequately represented the plant jacketinglinsulation system. Please describe differences in the jacketing and banding systems used for piping and other components for which the test results are applied, potentially including valves and other fittings. At a minimum, the following areas should be addressed:

a. Please describe how the characteristic failure dimensions of the tested jacketing/insulation compare with the effective diameter of the jet at the axial placement of the target. The characteristic failure dimensions are based on the primary failure mechanisms of the jacketing system, e.g., for a stainless steel jacket held in place by three latches where all three latches must fail for the jacket to fail, then all three latches must be effectively impacted by the pressure for which the ZOI is calculated. Applying test results to a ZOl based on a centerline pressure for relatively low nozzle-to-target spacing would be nonconservative with respect to impacting the entire target with the calculated pressure.

b. Please describe if the insulation and jacketing system used in the testing was of the same general manufacture and manufacturing process as the insulation used in the plant. If not, please describe what steps were taken to ensure that the general strength of the insulation system tested was conservative with respect to the plant insulation. It is known that there were generally two very different processes used to manufacture calcium silicate whereby one type readily dissolved in water but the other type dissolves much more slowly. Such manufacturing differences could also become apparent in debris generation testing, as well.

c. Please provide an evaluation of scaling the strength of the jacketing or encapsulation systems to the tests. For example, a latching system on a 30-inch pipe within a ZOI could be stressed much more than a latching system on a 10 inch pipe in a scaled ZOI test. If the latches used in the testing and the plants are the same, the latches in the testing could be significantly under-stressed. If a prototypically sized target were impacted by an undersized jet, it would similarly be under-stressed. Evaluations of banding, jacketing, rivets, screws, etc., should be made. For example, scaling the strength of the jacketing was discussed in the OPG report on calcium silicate debris generation testing.

41. There are relatively large uncertainties associated with calculating jet stagnation pressures and ZOls for both the test and the plant conditions based on the models used in the WCAP reports. Please explain the steps taken to ensure that the calculations resulted in conservative estimates of these values. Please provide the inputs for these calculations and the sources of the inputs.

42. Please describe the procedure and assumptions for using the ANSI/ANS-58-2-1988 standard, "Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture," to calculate the test jet stagnation pressures at specific locations downrange from the test nozzle.

a. In 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

LIC-1 5-0100 Page 27 Operating Plants," was the analysis based on the same initial condition as the initial test temperature, specified as 550°F? If not, please provide an evaluation of the significance of the difference.

b. Please explain whether the water subcooling used in the analysis was that of the initial tank temperature or the temperature of the water in the pipe next to the rupture disk. Test data indicated that the water in the piping had cooled below that of the test tank.

c. The break mass flow rate is a key input to the ANS/IANS-58-2-1988 standard.

Please explain how the associated debris generation test mass flow rate was determined. If the experimental volumetric flow was used, please explain how the mass flow was calculated from the volumetric flow given the considerations of potential two-phase flow and temperature dependent water and vapor densities. If the mass flow was analytically determined, then describe the analytical method used to calculate the mass flow rate.

d. Noting the extremely rapid decrease in nozzle pressure and flow rate illustrated in the test plots in the first tenths of a second, please explain how the transient behavior was considered in the application of the ANSI/ANS-58-2-1988 standard.

Specifically, please explain whether the inputs to the standard represent the initial conditions or the conditions after the first extremely rapid transient (e.g.,

say at one tenth of a second).

e. Given the extreme initial transient behavior of the jet, justify the use of the steady state ANSI/ANS-58-2-1 988 standard jet expansion model to determine the jet centerline stagnation pressures rather than experimentally measuring the pressures.

43. Please describe the procedure used to calculate the isobar volumes used in determining the equivalent spherical ZOI radii using the ANSI/ANS-58-2-1988 standard.

a. What were the assumed plant-specific reactor coolant system temperatures and pressures and break sizes used in the calculation? Note that the isobar volumes would be different for a hot leg break than for a cold leg break since the degrees of subcooling is a direct input to the ANSI/ANS-58-2-1988 standard and which affects the diameter of the jet. Note that an under-calculated isobar volume would result in an under-calculated ZOI radius.

b. What was the calculational method used to estimate the plant-specific and break-specific mass flow rate for the postulated plant LOCA, which was used as input to the standard for calculating isobar volumes?

c. Given that the degree of subcooling is an input parameter to the ANSI/ANS 2-1988 standard and that this parameter affects the pressure isobar volumes, what steps were taken to ensure that the isobar volumes conservatively match the plant-specific postulated LOCA degree of subcooling for the plant debris generation break selections? Were multiple break conditions calculated to ensure a conservative specification of the ZOI radii?

LIC-15-0 100 Page 28

44. Please provide a detailed description of the test apparatus specifically including the piping from the pressurized test tank to the exit nozzle including the rupture disk system.

a. Based on the temperature traces in the test reports, it is apparent that the fluid near the nozzle was colder than the bulk test temperature. How was the fact that the fluid near the nozzle was colder than the bulk fluid accounted for in the evaluations?

b. How was the hydraulic resistance of the test piping which affected the test flow characteristics evaluated with respect to a postulated plant-specific LOCA break flow where such piping flow resistance would not be present?

c. What was the specified rupture differential pressure of the rupture disks?

45. WCAP-1 671 0-P discusses the shock wave resulting from the instantaneous rupture of piping.

a. Was any analysis or parametric testing conducted to get an idea of the sensitivity of the potential to form a shock wave at different thermal-hydraulic conditions? Were temperatures and pressures prototypical of pressurized-water reactor hot legs considered?

b. Was the initial lower temperature of the fluid near the test nozzle taken into consideration in the evaluation? Specifically, was the damage potential assessed as a function of the degree of subcooling in the test initial conditions?

c. What is the basis for scaling a shock wave from the reduced-scale nozzle opening area tested to the break opening area for a limiting rupture in the actual plant piping?

d. How is the effect of a shock wave scaled with distance for both the test nozzle and plant condition?

46. Some piping oriented axially with respect to the break location (including the ruptured pipe itself) could have insulation stripped off near the break. Once this insulation is stripped away, succeeding segments of insulation will have one open end exposed directly to the LOCA jet, which appears to be a more vulnerable configuration than the configuration tested by Westinghouse. As a result, damage would appear to be capable of propagating along an axially-oriented pipe significantly beyond the distances calculated by Westinghouse. Please provide a technical basis to demonstrate that the reduced ZOls calculated for the piping configuration tested are prototypical or conservative of the degree of damage that would occur to insulation on piping lines oriented axially with respect to the break location.

47. WCAP-16710-P noted damage to the cloth blankets that cover the fiberglass insulation in some cases resulting in the release of fiberglass. The tears in the cloth covering were attributed to the steel jacket or the test fixture and not the steam jet.

It seems that any damage that occurs to the target during the test would be likely to occur in the plant. Was the potential for damage to plant insulation from similar

LIC-1 5-0100 Page 29 conditions considered? For example, the test fixture could represent a piping component or support, or other nearby structural member. The insulation jacketing is obviously representative of itself. Please explain the basis for the statement in the WCAP that damage similar to that which occurred to the end pieces in not expected to occur in the plant. It is likely that a break in the plant will result in a much more chaotic condition than that which occurred in testing. Therefore, it would be more likely for the insulation to be damaged by either the jacketing or other objects nearby.

OPPD's RAI #39-47 Response:

OPPD will not credit the reduced ZOI resulting from the testing activities associated with the subject of RAIs #39 through #47. However, there is one insulating system installed at FCS where an alternate ZOI is based on similar materials and configuration as those approved in the NRC's SE on NEI 04-07.

A reinforced PCI stainless steel jacketed NUKON insulation system is installed on a 3-inch pressurizer spray line and its proximity to the strainer results in this being a debris source for the SBLOCA event at FCS. The banding used to reinforce this NUKON insulation is similar to the banding used in the OPG testing of CaI-Sil insulating material. The banding is 0.5 inches wide by 0.03 inches thick stainless steel secured with standard banding crimp connectors. The bands are spaced 3 inches + 0.25 inch. The bands used for the OPG CaI-Sil tests were 0.020-inch thick stainless steel bands with standard crimp connectors with an average spacing of 6.5 inches.

Latches, as in the original PCI NUKON insulation system, were found to be the failure point at the Colorado Experiment Engineering Station, Inc. (CEESI) Air Jet Testing. The main function of the bands is to ensure that the jacketing does not peel off following the blast due to a hypothetical double ended guillotine break. If the jacketing remains in place following the overpressure pulse caused by the instantaneous break, then the jacketing will provide the necessary protection of the NUKON from the effects of jet impingement.

The SE to NEI-04-07 states that "'Sure Hold" bands installed over the PCI NUKON insulation system would have a destruction pressure of 90 psig, which translates to a 2.4D ZOI. The stainless steel banding for the reinforced NUKON insulation system at FCS is more closely spaced than the "Sure Hold" bands which provides additional restraining capability to preclude the stainless steel jacketing from peeling off due to the initial blast and subsequent jet impingement.

The jacketing of the PCI NUKON insulation system is stainless steel as opposed to the Al jacketing used in the OPG tests. The main function of the jacketing is to protect the underlying NUKON jackets from damage caused by normal operational challenges and to protect the NUKON blankets from direct jet impingement. There is no significant difference between aluminum and stainless steel jacketing thickness (on the order of 0.01 inch) - both jacketing types will provide considerable protection of the underlying insulation system from the effects of direct jet impingement. Note that the yield strength of stainless steel is approximately 30 to 35 ksi, and the yield strength of aluminum ranges from 10 to 30 ksi depending on the alloy. Therefore, it is likely that the actual strength of the stainless steel jacket is equivalent or better than the strength of the aluminum jacket used in the OPG tests.

The debris generation analysis applies a 3D ZOI to quantify the amount of fiber material that would be generated in the bounding SBLOCA event for this 3-inch pressurizer spray line. This

LIC-1 5-0100 Page 30 3D ZOI is more conservative than the 2.4D ZOI which is applicable to the installed reinforced NUKON insulation system.

However, subsequent discussions with the NRC on the subject of these RAIs concluded that for OPPD to continue to apply the reduced ZOI for the current insulation configuration in this location, banding must be installed.

LIC-1 5-0100 Page 1 Update of OPPD Letter LIC-13-0058 Proposed Resolution Path for Closure of Generic Letter (GL) 2004-02 and Generic Safety Issue (GSI) 191, Pressurized Water Reactor Sump Performance at Fort Calhoun Station Unit No. 1

LIC-1 5-0100 Page 2 QPPD provides this update regarding the status of bypass testing, strainer head loss testing, and characterization of in-vessel effects mentioned in OPPD letter LIC-13-0058 dated May 15, 2013 (Enclosure 1, Reference 7). OPPD is also providing an update to the resolution schedule and commitments made in that letter.

Characterization of Current Containment Fiber Status As discussed in Enclosure 1, Reference 1, bypass testing was conducted with TempMat material and found a bypass fraction of 2.2% (Page 93 of Enclosure to OPPD letter LIC-08-0021). OPPD has utilized the PWRQG methodology for evaluating cold leg break bypass at hot leg switchover, as documented in WCAP-17788-NP Volume 3 (ML15210A668). Preliminary calculations for the cold leg break with maximum fiber load utilizing the 2.2% (documented bypass test fraction), yield in-vessel results of less than 3 grams/fuel assembly fiber loading.

Characterization of Strainer Head Loss Status NRC staff acceptance of testing protocol discussed in RAI #10 (Enclosure 2, Page 7), may determine if additional testing is required to be conducted. See Commitment #2 in Enclosure 4.

Characterization of In-Vessel Effects On July 17, 2015 (ML15210A668), the PWROG submitted WCAP-1 7788 "Comprehensive Analysis and Test Program for GSI 191 Closure (PA-SEE-1090)," which is under review by the NRC. QPPD intends to follow the resolution strategy proposed by the PWRQG in WCAP-1 7788 following NRC review.

Updated Resolution Schedule Changes to the Resolution Schedule provided in QPPD letter LIC-13-0058 dated May 15, 2013 (Enclosure 1, Reference 7) are provided below and are reflected as commitments in Enclosure 4:

• In letter LIC-13-0058, QPPD committed to complete measurements for insulation replacement and remediation by the end of the first refueling outage (RFQ) following January 1, 2013. At the time, due to the 2011 Missouri River flood and the station's entry into Inspection Manual Chapter 0350, ECS had not restarted from the 2011 REQ. In May 2013, anticipating re-start in the near future, the next RFO was expected to begin during the fall of 2014. However, as the 2011 REQ did not end until December 21, 2013, the next REQ occurred in the spring of 2015. This commitment is complete as the measurements for insulation replacement and remediation was performed during the spring 2015 REQ. This is Commitment #1 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to complete strainer head loss and fiber bypass testing by the end of 2015. QPPD is revising this commitment such that any required additional testing will be completed within 18 months of staff acceptance of the testing and debris transport protocol discussed in RAI #10 (Enclosure 2, Page 7). This is Commitment #2 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to complete the necessary insulation replacements, remediation, or model refinements by the completion of the third REQ following January 1,2013, which at the time was expected to be in the fall of 2017. However, as stated above, due to the length of the 2011 RFO, the third REQ following January 1, 2013 will take place in the spring of 2018. Therefore, OPPD is extending the timeframe for completing this commitment from the fall 2017 REQ to the spring 2018 REQ. This is Commitment #3 in Enclosure 4.

LIC-1 5-0100 Page 3

• In letter LIC-13-0058, OPPD committed to submit a final updated supplemental response to support closure of GL 2004-02 for FCS within six months of establishing a final determination of the scope of insulation replacement or remediation. OPPO specified that this would be completed within six months after the fall 2014 RFO. OPPD is revising the due date to be within 18 months of NRC staff acceptance of the testing protocol discussed in RA! #10 (Enclosure 2, Page 7). This is Commitment #4 in Enclosure 4.

• In letter LIC-13-0058, OPPD committed to update the current licensing basis (CLB) (i.e., USAR, etc.) following NRC acceptance of the updated supplemental response for FCS and completion of the identified removal or modification of insulation debris sources in containment. OPPD committed to complete this by the end of the fall 2017 RFO. As stated above, the RFO originally anticipated to occur in the fall of 2017 will now occur in the spring of 2018. To allow sufficient time to process the necessary CLB changes, OPPD is revising the due date for updating the CLB to be within 90 days of the end of the spring 2018 REQ. This is Commitment #5 in Enclosure 4.

LIC-1 5-01 00 Page 1 Regulatory Commitments The following table is updated from the Regulatory Commitments Table provided by OPPD in of OPPD letter LIC-1 3-0058 (Enclosure 1, Reference 7). The original commitment is shown in the left column with the revised commitment and/or due date in the right column.

1Complete measurements for insulation replacement and remediation by the end of the first REQ following January 1, 2013; currently expected to be completed during the fall of 2014. [AR 59106]

Due: By the end of the fall 2014 RFO. Due: Completed during the spring 2015 RFO.

2 Strainer head loss and fiber bypass testing Any required additional strainer head loss is expected to be completed by the end of and fiber bypass testing will be completed 2015. [AR 59106] within 12 months of NRC staff acceptance of the testing protocol discussed in RAI

1. 10. [AR 59106]

Due: December 31, 2015. Due: Within 18 months of NRC staff acceptance of testing protocol discussed in RAI #10.

3 Complete the necessary insulation Complete the necessary insulation replacements, remediation, or model replacements, remediation (i.e., EC refinements by the completion of the third 49722), or model refinements by the refueling outage (RFO) following January completion of the third refueling outage 1, 2013 (fall 2017). [AR 59106] (RFO) following January 1, 2013 (spring 2018). [AR 59106]

Due: By the end of fall 2017 RFO.

Due: By the end of spring 2018 RFO.

4 Within six months of establishing a final Within six months of establishing a final determination of the scope of insulation determination of the scope of insulation replacement or remediation, OPPD will replacement or remediation, OPPD will submit a final updated supplemental submit a final updated supplemental response to support closure of GL 2004-02 response to support closure of GL 2004-02 for FCS. for FCS.

[AR 59106] [AR 59106]

Due: Within six months following fall 2014 Due: Within 18 months of NRC staff REQ. acceptance of testing protocol discussed in RAI #10.

LIC-1 5-0100 Page 2 5 OPPD will update the current licensing OPPD will update the current licensing basis (USAR, etc.) following NRC basis (USAR, etc.) following NRC acceptance of the updated supplemental acceptance of the updated supplemental response for FCS and completion of the response for FCS and completion of the identified removal or modification of identified removal or modification of insulation debris sources in Containment, insulation debris sources in Containment.

[AR 59106] [AR 59106]

Due: By the end of fall 2017 RFO. Due: Within 90 days of completing the spring 2018 REQ.