L-13-176, Generic Safety Issue 191 Resolution Plan

Generic Safety Issue 191 Resolution Plan

ML13136A144
Person / Time
Site:	Beaver Valley
Issue date:	05/16/2013
From:	Emily Larson FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GSI- 191, L-13-176, TAC MC4665, TAC MC4666
Download: ML13136A144 (23)

Text

FENOC

_E FirstEnergy Nuclear Operating Company Beave r V alley Powe r Sfafion P.O. Box 4 Shippingport, PA 15077 Eric A. Larson S[e Vice President 724-682-5234 Fax: 724-643-8069 May 16, 2013 L-1 3-1 76 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 10 cFR 50.54(f)

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BVPS-1 Docket No. 50-334, License No. DPR-66 BVPS-2 Docket No. 50-4'12, License No. NPF-73 Generic Safety lssue 191 Resolution Plan (TAC Nos. MC4665 and MC4666)

This letter fonryards information regarding resolution of Generic Safety lssue 191, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance,"

for BeaverValley Power Station, Unit Nos. 1 (BVPS-1) and 2 (BVPS-2).

Nuclear Regulatory Commission (NRC) staff has interacted with the industry and stakeholders to develop options forthe resolution of Generic Safety lssue 191" The NRC staff paper SECY-12-0093 presents closure options for Generic Safety lssue 191.

Attachments 1 and 2 provide information regarding the current status of efforts to address Generic Letter 20A4-02, "Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

and describe the Generic Safety lssue 191 closure option, resolution plan, and associated implementation schedule for BVPS-1 and BVPS-2, respectively.

These attachments also describe mitigation measures appropriate to support the implementation schedule. provides references for information cited in Attachments 1 and 2, and related correspondence.

Attachment 4 provides a list of regulatory commitments included in this submittal. lf there are any questions or if additional information is

required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.

Beaver Valley Power Station, Unit Nos. 1 and 2 L-13-176 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on May I b,2a8.

Sincerely, ilf,/^-

Eric A. Larson Attachments:

1. Beaver Valley Power Station, Unit No. 1, Generic Safety lssue 191, In-Vessel Effects Resolution Plan

2. Beaver Valley Power Station, Unit No. 2, Generic Safety lssue 191, In-Vessel Effects Resolution Plan

3. References

4. Regulatory Commitment List cc: NRC Region lAdministrator NRC Resident lnspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

ATTACHMENT 1

L-1 3-1 76 Beaver Valley Power Station, Unit No. 1, Generic Safety lssue 191, In-Vessel Effects Resolution Plan Page 1 of 9

==

Introduction:==

FirstEnergy Nuclear Operating Company (FENOC) has selected Option 2, deterministic path, of Nuclear Regulatory Commission (NRC) staff paper SECY-12-0093, "Closure Options for Generic Safety lssue 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance,"

for Beaver Valley Power Station, Unit No. 1 (BVPS-1) and intends to pursue refinements to evaluation methods and acceptance criteria. ln addition, the resolution schedule defined in SECY-12-0093 will also be adhered to as described herein. The Nuclear Energy Institute (NEl) closure option template dated November 9, 2012 was used to develop this response. This submittal provides a resolution plan that follows the deterministic path of Option 2 (referred to as NEI template option 2a).

To support use of this path, for the period required to complete the necessary analysis and testing, FENOC has evaluated the design and procedural capabilities that exist to identify defense in depth measures to detect and mitigate in-vessel blockage. A description of these measures to detect and mitigate in-vessel blockage is provided later in this document. A summary of the existing margins and conservatisms that exist for BVPS-1 are also included in this document. References cited in this attachment are listed in Attachment 3.

Current Gontainment Fiber Status:

A bypass test was performed with an objective to collect and record the fibrous debris bypass fraction for the BVPS-1 prototypical sump strainer. Sump strainer bypass testing was completed for BVPS-1 in the spring of 2008, at the Alion Hydraulics Laboratory.

This testing established the quantity of fiber that passed through the strainer over a range of approach velocities and head losses. Scaled debris quantities used for these tests were derived from the primary line break within containment that yields the most fibrous debris and the highest approach velocity.

The strainer bypass testing was credited for determining the amount of fiber that reaches the vessel. In addition, the bypass testing was reviewed and generally aligns with the concepts outlined by the NRC at the NEI Workshop held on October 18, 2012.

For example, the debris was added to the tank in increments of approximately one sixteenth inch of bed thickness, and a bag capture method was used.

Sump strainer head loss testing was conducted in the spring of 2008 in accordance with the March 2008 protocol (Reference

1) prepared by the NRC. In the summer of 2010 an additional strainer head loss test was performed for a 6 inch pressurizer safety relief valve line break. The total quantity of fibrous insulation was 1 1.8 pounds (lbs) for 2008 L-13-176 Page 2 of I strainer head loss Test 6 (Reference 2). This quantity (1 1.8 lbs) bounds breaks other than the pressurizer safety relief valve line break, including the debris quantities for the BVPS-1 loop break, pressurizer surge line break, safety injection line break, residual heat removal line break, and reactor vessel nozzle break. The total quantity of fibrous insulation was 1A7.4 lbs for 2010 strainer head loss Test 7 (Reference 3). This quantity (107.4 lbs) bounds the debris quantities for the BVPS-1 pressurizer safety relief valve line break. Thirty pounds of latent fiber must be added to these quantities (the quantity of 30 lbs is based on testing and bounds the calculated quantity).

This makes the total debris quantity that could be transported to the strainer, 41.8 lbs (1 1.8 lbs plus 30 lbs) for Test 6 and 137.4lbs (1 07.4 lbs plus 30 lbs) for Test 7.

The fiber bypass percentage was 8.0 percent, based on the results of plant specific strainer bypass testing. The bypass fraction can be applied over the range of breaks since the testing showed that there was open screen area with the bounding fiber load.

The amount of fiber bypass that may reach each fuel assembly was calculated on a per assembly basis. The amount of fiber per assembly in grams (g) was calculated by multiplying the total fiber (41.8 lbs or 137.4 lbs) by the fiber bypass percentage (8.0 percent) and converting the result from pounds to grams and then dividing that by the number of fuel assemblies (157). Thus, the total fiber bypass load was'9.7 grams per fuel assembly (g/FA) for Test 6 and 31.8 grams per fuel assembly for Test 7, as shown in the following equations.

41.8 lbs x0.080 x 453.6 g lb=

Test 6:

Test 7:

157 FA s.7 3-FA

= 31.8 g FA 137

.4lbs x0.080 x 453.6 g

tb 157 FA Therefore, for BVPS-1, the only break that exceeds the acceptance criteria of 15 g/FA (Reference

4) was a break associated with a pressurizer safety relief valve line. Other breaks fall within the acceptance criteria. Actions to address the pressurizer safety relief valve line are described under the heading Resolution Schedule below.

Strainer Head Loss Status:

FirstEnergy Nuclear Operating Company has previously provided the results of strainer head loss testing, including chemical effects, in references 2 and 3. These tests demonstrated acceptable results with regard to allowable strainer head loss. Concerns of the NRC staff associated with Generic Safety lssue 191, "Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance,"

Generic Letter L-1 3-1 76 Page 3 of 9 2004-0l "Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

and BVPS-1 strainer head loss have been addressed as described in references 2, 3, and 5. There are no outstanding issues with respect to head loss testing.

Characterization of I n-Vessel Effects :

FirstEnergy Nuclear Operating Company intends to follow the resolution strategy proposed by the Pressurized Water Reactor Owners Group (PWROG) to establish in-vessel debris fimits for the BVPS-1 type plant design. This approach is expected to establish in-vessel debris limits in excess of that currently established by Westinghouse reportWCAP-16793, Revision 2 (Reference 6).

Licensing Basis Commitments:

There is currently only one open FENOC commitment for BVPS-1 regarding Generic Letter 2004-02. The commitment to the NRC is contained in a June 30, 2009 FENOC letter (Reference

2) that states:

It is recognized thatthe NRC review of WCAP-16793-NP, Revision 1, has not been completed. Any additional actions required to address NRC questions will be addressed.

The due date is, within 90 days after issuance of the final NRC safety evaluation on WCAP-16793-NP, Revision 1.

The NRC has not issued a safety evaluation forWCAP-16793-NP, Revision 1. In the interim WCAP-16793-NP, Revision 2, October zAfi (Reference 6), was issued by Westinghouse for review and approval by the NRC. A safety evaluation for Revision 2 of WCAP-16793-NP was made available by the NRC on April 16, 2013 (Reference 7).

The above commitment, listed as Commitment 1 in Attachment 2 of Reference 2, is hereby replaced with the commitments described under the Resolution Schedule heading below and listed in Attachment 4.

Resolution Schedule:

FirstEnergy Nuclear Operating Company will achieve closure of Generic Safety lssue 191 and address Generic Letter 2004-02 per the schedule provided below. The results of the PWROG in-vessel effects testing effort will be utilized to close the remaining issue for BVPS-1 if it provides a conclusion that supports the current fibrous fuel assembly loading forthe BVPS-1 limiting break (pressurizer safety relief valve line).

The PWROG program results are expected to be available and provided to the NRC staff for review in the fall of 2014 (Reference 8).

Measurements will be taken in preparation for insulation modifications associated with the BVPS-1 pressurizer safety relief valve inlet lines, if it is determined that the PWROG in-vessel effects testing effort does not support closure of Generic Safety lssue 191 and L-13-176 Page 4 of 9 Generic Letter 2AA4-02 for BVPS-1. This action will be accomplished during the BVPS-1 refueling outage in the fall of 2013.

Insulation for the BVPS-1 pressurizer safety relief valve inlet lines will be replaced or modified as appropriate if it is determined that the PWROG in-vessel effects testing effort does not support closure of Generic Safety lssue 191 and Generic Letter 2004-02 for BVPS-1. This action will be accomplished by the end of the refueling outage in the fafl of 2A16, if needed.

The final Generic Letter 2004-02 supplemental response for BVPS-1 will be provided to the NRC within 6 months after the NRC approves, by issuance of a safety evaluation, the new PWROG topical report addressing additional in-vessel effects testing efforts that are currently being pursued.

Updated Final Safety Analysis Report changes will be completed to update the current licensing basis for BVPS-1 as appropriate, following NRC acceptance of the final docketed Generic Safety lssue 191 response for BVPS-1 and completion of any identified insulation modifications for BVPS-1 that may be required.

Summary of Actions Gompleted:

A strainer replacement was installed at BVPS-1 during the fall 2007 refueling outage (1R18). The new replacement strainer is of Control Components Incorporated (CCl) design, and increased the available surface area from approximately 130 square feet to 3400 square feet.

The BVPS-1 start signal for the recirculation spray system pumps was changed from a fixed time delay to an engineered safety features actuation system signal based on a refueling water storage tank level low coincident with a containment pressure high-high signal. This change was completed during the fall 2007 refueling outage (1R18), and will allow sufficient pool depth to cover the sump strainer before initiating recirculation flow.

Beaver Valley Power Station, Unit No. 1, is considered a low to medium fiber plant based on approved industry and NRC standards associated with Generic Safety f ssue 191 and Generic Letter 2004-02 This has been accomplished through extensive replacement of fibrous and particulate insulation with reflective metal insulation.

The sodium hydroxide containment sump buffer was replaced in the spring 2012 refueling outage (1R21). The replacement buffer is sodium tetraborate.

This lowers the chemical loading for BVPS-1 as discussed in references 9 and 10.

A containment coatings inspection and assessment program and containment cleaning program became effective for BVPS in April 2008 and applies to BVPS-1 refueling

Attachment 1

L-1 3-1 76 Page 5 of 9 outages beginning with the spring 2009 refueling outage (1 R19). This reduces the quantity of unqualified coatings and latent debris that may be transported to the sump.

lodine filters, containing a significant amount of thin aluminum that would have been submerged, were removed from the BVPS-1 containment.

This supported a significant reduction in the generation of chemical precipitates.

Details of additional actions and modifications completed to address Generic Safety lssue 191 for BVPS-1 are provided in FENOC letters dated June 30, 2009 and September 28, 2010 (references 2 and 3).

Summary of Margins and Gonseruatisms:

Margins and conservatisms with respect to debris generation, debris transpott, strainer head loss and chemical effects have been summarized in previous Generic Letter 2004-02 submittals (references 2 and 3). Additional information is ofiered here.

Reactor In-vessel Fiber Loadinq The limiting break is a pressurizer safety relief valve line break, which generates 31.8 g/FA of fiber. Significant fiber reduction has been conducted in containment such that no other evaluated breaks result in excess of 15 g/FA of fiber.

The fibrous debris limits were generated from testing conducted at limiting reactor coofant system flows (that is, 44.7 and 15.5 gallons-per-minute per fuel assembly

[gpm/FA]). Pressurized Water Reactor Owners Group fuel assembly testing has demonstrated that maximum head loss occurs at high flow conditions (44.7 gpm/FA).

A small break loss of coolant accident at a 6 inch pipe near the top of the pressurizer (such as the pressurizer safety relief valve line break) will experience a greatly reduced flow due to the significantly smaller break size and large elevation difference between the reactor core and the break. An analysis of the pressurizer safety relief valve line break using the containment transient analysis code MAAP-DBA, determined the emergency core cooling system flow rate to be approximately 23.7 gpm/FA. Fuel assembly testing performed per Westinghouse fuel assembly test report, WCAP 17057-P, Revision 1 (Reference 11), used flow rates of 44.7 gpm and 15.5 gpm. These tests consistently demonstrated that maximum head loss occurs at high flow conditions; therefore, the smaller flow rate associated with the pressurizer safety relief valve line break would translate into a higher acceptable in-vessel fiber load.

Hot Les Drivinq Head According to WCAP-16793-NP, Rev. 2, the PWROG testing demonstrated that since the hot leg break is limiting with respect to allowable fiber loading, the calculation of the hot leg available driving head is the relevant value when determining if a reduction in core flow occurs. For the pressurizer safety relief valve line break with 31.8 g/FA of fiber, the RCS'loops will remain pressurized due to the size and elevation of the break; L-1 3-1 76 Page 6 of 9 therefore, the hot leg driving head methodology provided by WCAP-16793 does not apply to this break.

The available driving head value for a hot leg break can be compared to the differential pressure value recorded from the test conducted with 15 grams of fiber to demonstrate that significant margin exists between the expected pressure loss due to a debris bed and the expected driving head available to support core flow. The available driving head for a hot leg break at BVPS-1 is 16.37 psi; the limiting AREVA Enterprises Inc. (AREVA) fuel assembly test that uses 15 grams of fiber (12-FG-FPC) resulted in a total fuel assembly pressure drop of only 2.7 psid. This confirms that significant margin exists between the hot leg break available driving head and the expected pressure loss through the debris bed, and there will be no significant reduction in core flow.

Several tests were performed at Westinghouse, which show debris head loss results that are comparable to the available driving head at BVPS-1. Tests ClB49, ClB50, and Cf B51 each use 50 g/FA of fiber and produced debris head losses of 14.94,17.80, and 15.35 psid, respectively (following addition of chemical precipitates).

Since a BVPS-1 hot leg break will produce less than 15 g/FA of fiber, significant margin exists between the actual fiber quantity at BVPS-1 and that which was used in these Westinghouse FA tests.

Reactor Vessel Desiqn Reactor vessel designs were considered (Reference

6) and a limiting design was chosen based upon the vessel design that would be the most limiting with respect to core inlet flow blockage. Three Westinghouse vessel designs were considered; designed barrel/baffle (B/B) upflow, converted B/B upflow and B/B downflow. For Westinghouse designed plants, the most limiting vessel design is the B/B downflow, since the only means for the flow to enter the core is through the lower core plate. A Westinghouse B/B downflow plant was evaluated using the WCOBRA/TRAC thermal-hydraulic computer code. For this evaluation, it was concluded that sufficient liquid can enter the core to remove core decay heat once the plant has switched to sump recirculation with up to 99.4 percent core blockage. Beaver Valley Power Station, Unit No. 1, is a converted barrel/baffle upflow plant, which is less limiting for fuel assembly blockage as stated above and provides an alternate path to provide cooling to the core.

Boron Precipitation Boron precipitation is an issue that becomes problematic with cold leg breaks. As stated in WCAP-16793-NP, Revision 2 (Reference 6):

The limiting scenario for boric acid precipitation is a cold leg break where the core flow is stagnant with only enough core inlet flow to replace core boil-off.

Therefore, boron precipitation is not an issue for any other breaks including the limiting BVPS-1 pressurizer safety relief valve line break.

L-1 3-1 76 Page 7 of 9 Summary of Defense-in-Depth Measures BVPS-1 has a low concentration of precipitates formed during a loss of coolant

accident, due to the reduction of available submerged aluminum in containment and the change of the containment sump buffering agent from sodium hydroxide to sodium tetraborate. The existing emergency operating procedure guidance for transfer to hot leg recirculation is 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after initiation of a loss of coolant accident. For this time frame, considering injection time and transfer to recirculation time, it is expected that a reduced concentration of precipitates would be formed. However, to enhance the capability to address postulated core inlet blockage, the BVPS-1 emergency operating procedures will be revised using recent guidance from the PWROG to implement early switchover to hot leg recirculation should plant parameters indicate that core blockage is occurring.

This action will be taken prior to transfer to the existing "Response to Degraded Core Cooling" procedure.

Appropriate operator training will be completed to address this emergency operating procedure revision prior to implementation.

These actions will be completed within six months of the submittal date of this Generic Safety lssue 191 resolution plan.

Existing defense-in-depth mitigative measures are described below.

Containment Sump Screen Actions taken in response to NRC Bulletin 2003-01, "Potential lmpact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors,"

are described in Referen ce 12. These actions continue to remain in effect at BVPS-1.

ln-Vessel The following describes the plant specific design features and procedural capabilities that exist for detecting and mitigating fuel blockage.

Detection of Inadequate Reactor Core Flow

- Increasing core exit thermocouple temperature indication Core exit thermocouples are monitored as part of emergency operating procedure monitoring of status trees and the safety parameter display system. As part of operator training, the operating crew must demonstrate the ability to detect increases in core exit thermocouple temperature indication and transition to the appropriate emergency operating procedure for dealing with this condition. This guidance is provided in the BVPS emergency operating procedure "User's Guide" section titled "Control Room Usage of Status Trees," and the BVPS-1 "Core Cooling Status Trees" procedure.

- Decreasing reactor water level indication The reactor vessel level indication system is monitored throughout the emergency operating procedures.

Through continuing training, operators demonstrate the ability to monitor and understand the implications of a decreasing reactor vessel water level and L-13-176 Page 8 of 9 appropriately transition within the emergency operating procedure framework to mitigate this condition.

- Increasing containment or auxiliary building radiation levels Increasing radiation levels will be indicated by alarms in the control room with specific procedural steps in both alarm response procedures and the emergency operating procedures for addressing the condition.

Mitigation of Inadequate Reactor Core Flow

- Start a reactor coolant pump Beaver Valley Power Station, Unit No. 1, procedure titled, "Response to Inadequate Core Cooling,"

provides direction to start a reactor coolant pump if core exit thermocouple temperature indication is greater than 12A} degrees Fahrenheit.

This action would aid in the removal of the established blockage to the core to once again allow normal recirculation injection flow paths to become effective at maintaining adequate core cooling.

- lmplementation of Severe Accident Management Guidelines Severe Accident Management Guidelines (SAMG) provide additional guidance and actions for addressing inadequate core flow conditions. At BVPS-1 the SAMGs are entered from the following function restoration procedure for inadequate core cooling.

Procedure titled, "Response to Inadequate Core Cooling,"

when core exit thermocouple temperatures are greater than 1200 degrees Fahrenheit and actions to cool the core are not successful.

The SAMGs provide for alternate injection paths into the reactor coolant system (RCS).

Specific SAMGs that provide guidance in this area are listed below.

"lnject lnto the RCS" guideline "RCS Injection to Recover Core" guideline The SAMGs provide for flooding containment to provide for convective circulation cooling of the reactor. Specific SAMGs that provide guidance in this area are listed below.

"lnject Into Containment" guideline "Flood Containment" guideline "Containment Water Level And Volume" guideline L-13-176 Page 9 of 9 Although these measures are not expected to be required based on the very low probability of an event that would result in significant quantities of debris being transported to the reactor vessel that would inhibit the necessary cooling of the fuel, they do provide additional assurance that the health and safety of the public would be protected. These measures provide support for the extension of time required to completely address Generic Letter 2004-02fior BVPS-1.

Conclusion The Generic Safety lssue 191 resolution path for BVPS-1 is acceptable, based on the information provided in this document. The execution of the actions identified in this document will result in successful resolution of Generic Safety lssue 191 and closure of Generic Letter 20A4-02.

ATTACHMENT 2

L-13-176 Beaver Valley Power Station, Unit No. 2, Generic Safety lssue 191, In-Vessel Effects Resolution Plan Page 1 of 8

==

Introduction:==

FirstEnergy Nuclear Operating Company (FENOC) has selected Option 2, deterministic path, of Nuclear Regulatory Commission (NRC) staff paper SECY-12-0093, "Closure Options for Generic Safety lssue 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance,"

for Beaver Valley Power Station, Unit No. 2 (BVPS-2) and intends to pursue refinements to evaluation methods and acceptance criteria. In addition, the resolution schedule defined in SECY-12-0093 will also be adhered to as described herein. The Nuclear Energy Institute (NEI) closure option template dated November 9, 2012 was used to develop this response. This submittal provides a resolution plan that follows the deterministic path of Option 2 (referred to as NEI template option 2a).

To support use of this path, for the period required to complete the necessary analysis and testing, FENOC has evaluated the design and procedural capabilities that exist to identify defense in depth measures to detect and mitigate in-vessel blockage. A description of these measures to detect and mitigate in-vessel blockage is provided later in this document. A summary of the existing margins and conservatisms that exist for BVPS-2 are also included in this document. References cited in this attachment are listed in Attachment 3.

Current Gontainment Fiber Status:

A bypass test was performed with an objective to collect and record the fibrous debris bypass fraction for the BVPS-2 prototypical sump strainer. Sump strainer bypass testing was completed for BVPS-2 in the fall of 2008, at the Alion Hydraulics Laboratory.

This testing established the quantity of fiber that passed through the strainer over a range of approach velocities and head losses. Scaled debris quantities used for these tests were derived from the primary line break within containment that yields the most fibrous debris and the highest approach velocity The strainer bypass testing was credited for determining the amount of fiber that reaches the vessel. In addition, the bypass testing was reviewed and generally aligns with the concepts outlined by the NRC at the NEI Workshop held on October 18,2012.

For example, the debris was added to the tank in increments of approximately one sixteenth inch of bed thickness, and a bag capture method was used.

Sump strainer head loss testing was conducted in the fall of 2008 in accordance with the March 2008 protocol (Reference

1) prepared bythe NRC. The total quantity of fibrous insulation was 36.4 pounds (lbs) for BVPS-2 strainer head loss Test 1A (references 2 and 3). This quantity bounds the debris quantities for BVPS-2 breaks.

L-13-176 Page 2 of 8 Thirty pounds of latent fiber must be added to this quantity (the quantity of 30 lbs is based on testing and bounds the calculated quantity). This makes the total debris quantity that could be transported to the strainer, 66.4 lbs (36.4 lbs plus 30 lbs). This total value was conservative based on the conservatisms utilized for the analyses.

The fiber bypass percentage was 4.2 percent, based on the results of plant specific strainer bypass testing. The amount of fiber bypass that may reach each fuel assembly was calculated on a per assembly basis. The amount of fiber per fuel assembly in grams (g) was calculated by multiplying the total fiber (66.4 lbs) by the fiber bypass percentage (4.2 percent) and converting the result from pounds to grams and then dividing that by the number of fuel assemblies (1 57). Thus, the total fiber bypass load for BVPS-2 was 8.1 grams per fuel assembly (g/FA), as shown in the following equation.

66.41bsx0.042x453.6 g

Based on this fiber bypass load (8.1 g/FA), bounding breaks for BVPS-2 met the fibrous limit per fuel assembly of 15 grams per fuel assembly (Reference 4 and SECY-12-0093).

However, FENOC plans to continue to support the Pressurized Water Reactor Owners Group (PWROG) effort to establish improved margin for in-vessel effects that apply to BVPS-2.

Strainer Head Loss Status:

FirstEnergy Nuclear Operating Company has previously provided the results of strainer head loss testing, including chemical effects in references 2, and 3. These tests demonstrated acceptable results with regard to allowable strainer head loss. Concerns of the NRC staff associated with Generic Safety lssue 191, "Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance,"

Generic Letter 2004 -02, "Potential lmpact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

and BVPS-2 strainer head loss have been addressed as described in references 2, 3, and 5. There are no outstanding issues with respect to head loss testing.

Characterization of I n-Vessel Effects :

FirstEnergy Nuclear Operating Company intends to follow the resolution strategy proposed by the PWROG to establish in-vessel debris limits for the BVPS-2 type plant design. This approach is expected to establish in-vessel debris limits in excess of that currently established by Westinghouse report WCAP-16793, Revision 2 (Reference 6).

= 8.1 8-FA

Attachment 2

L-1 3-1 76 Page 3 of 8 Licensing Basis Commitments:

There is currently only one open FENOC commitment for BVPS-2 regarding Generic Letter 2004-02. The commitment to the NRC is contained in a June 30, 2009 FENOC letter (Reference

2) that states:

It is recognized that the NRC review of WCAP-16793-NP, Revision 1, has not been completed. Any additional actions required to address NRC questions will be addressed.

The due date is, within 90 days after issuance of the final NRC safety evaluation on WCAP-16793-NP, Revision 1.

The NRC has not issued a safetyevaluation forWCAP-16793-NP, Revision

1. In the interim WCAP-16793-NP, Revision 2, October 2011 (Reference 6), was issued by Westinghouse for review and approval by the NRC. A safety evaluation for Revision 2 of WCAP-16793-NP was made available by the NRC on April 16,2013 (Reference 7).

The above commitment, listed as Commitment 1 in Attachment 2 of Referenc,e 2,

hereby replaced with the commitments described under the Resolution Schedule heading below and listed in Attachment 4.

Resolution Schedule:

FirstEnergy Nuclear Operating Company will achieve closure of Generic Safety lssue 191 and address Generic Letter 2004-02 perthe schedule provided below. The fuel assembly fibrous debris loading for BVPS-2 is within the limits established by Westinghouse evaluation report WCAP-16793, Revision

2. The results of the ongoing PWROG in-vessel effects testing effort will be utilized to close this remaining issue for BVPS-2, if it provides a conclusion that supports additional margin for the current BVPS-2 fuel assembly fibrous debris loading. The PWROG program results are expected to be available and provided to the NRC staff for review in the fall of 2Q14 (Reference 8).

The final Generic Letter 2004-02 supplemental response for BVPS-2 will be provided to the NRC within 6 months after the NRC approves, by issuance of a safety evaluation, the new PWROG topical report addressing additional in-vessel effects testing efforts that are currently being pursued.

Updated Final Safety Analysis Report changes will be completed to update the current licensing basis for BVPS-2 as appropriate, following NRC acceptance of the final docketed Generic Safety lssue 191 response for BVPS-2.

Summary of Actions Gompleted:

A strainer replacement was installed at BVPS-2 during the fall 2006 refueling outage (2R12). The new replacement strainer is of Enercon design with bypass eliminators, L-1 3-1 76 Page 4 of 8 which increased the available surface area from approximately 150 square feet to 3300 square feet.

The BVPS-2 start signal for the recirculation spray system pumps was changed from a fixed time delay to an engineered safety features actuation system signal based on a refueling water storage tank level low coincident with a containment pressure high-high signal. This change was completed during the spring 2008 refueling outage (2R13),

and will allow sufficient pool depth to cover the sump strainer before initiating recirculation flow.

Beaver Valley Power Station, Unit No. 2, is considered a low fiber plant based on approved industry and NRC standards associated with Generic Safety lssue 191 and Generic Letter 2004-02. This has been accomplished through extensive replacement of fibrous and particulate insulation with reflective metal insulation.

The sodium hydroxide containment sump buffer was replaced in the fall 2009 refueling outage (2R14). The replacement buffer is sodium tetraborate.

This lowers the chemical foading for BVPS-2 as discussed in references 13,14 and 15.

A containment coatings inspection and assessment program and containment cleaning program became effective for BVPS in April 2008 and applies to BVPS-2 refueling outages beginning with the spring 2008 refueling outage (2R13). This reduces the quantity of unqualified coatings and latent debris that may be transported to the sump.

lodine filters, containing a significant amount of thin aluminum that would have been submerged, were removed from the BVPS-2 containment.

This supported a significant reduction in the generation of chemical precipitates.

Details of additional actions and modifications completed to address Generic Safety lssue 191 for BVPS-2 are provided in FENOC letters dated June 30,2009 and September 28, 2010 (references 2 and 3).

Summary of Margins and Conservatisms:

Margins and conservatisms with respect to debris generation, debris transport, strainer head loss and chemical effects have been summarized in previous Generic Letter 2004-A2 submittals (references 2 and 3). Additional information is offered here.

Reactor In-vessel Fiber Loadinq The BVPS-2 strainer design includes Enercon cylindrical top-hat style strainer assemblies with debris eliminators.

This limits the amount of debris that can bypass the strainer and make it into the vessel. In addition, a significant debris reduction effortwas undertaken.

This, in conjunction with the debris eliminators and the associated bypass fraction, yields an in-vessel loading of less than 15 g/FAfor all evaluated breaks.

L-1 3-1 76 Page 5 of I Hot Leq Drivino Head According to WCAP-16793-NP, Rev. 2, the PWROG testing demonstrated that since the hot leg break is limiting with respect to allowable fiber loading, the calculation of the hot leg available driving head is the relevant value when determining if a reduction in core flow occurs. The available driving head value for a hot leg break can be compared to the differential pressure value recorded from the test conducted with 15 grams of fiber to demonstrate that significant margin exists between the expected pressure loss due to a debris bed and the expected driving head available to support core flow. The available driving head for a hot leg break at BVPS-2 is 16.51 psi; the limiting AREVA fuel assembly testthat uses 15 grams of fiber (12-FG-FPC) resulted in a total fuel assembly pressure drop of only 2.7 psid. This confirms that significant margin exists between the hot leg break available driving head and the expected pressure loss through the debris bed, and there will be no significant reduction in core flow.

Several tests were performed at Westinghouse, which show debris head loss results that are comparable to the available driving head at BVPS-2. Tests ClB49, ClB50, and CfB51 each use 50 g/FA of fiber and produced debris head losses of 14.94, 17.80, and 15.35 psid, respectively (following addition of chemical precipitates).

Since a BVPS-2 hot leg breakwill produce less than 15 g/FA of fiber, significant margin exists between the actual fiber quantity at BVPS-2 and that which was used in these Westinghouse FA tests.

ReactolVessel Desiqn Reactor vessel designs were considered (Reference

6) and a limiting design was chosen based upon the vessel design that would be the most limiting with respect to core inlet flow bfockage. Three Westinghouse vessel designs were considered; designed barrel/baffle (B/B) upflow, converted B/B upflow and B/B downflow. For Westinghouse designed plants, the most limiting vessel design is the B/B downflow, since the only means for the flow to enter the core is through the lower core plate. A Westinghouse B/B downflow plant was evaluated using the WCOBRA/TRAC thermal-hydraulic computer code. For this evaluation, it was concluded that sufficient liquid can enter the core to remove core decay heat once the plant has switched to sump recirculation with up to 99.4 percent core blockage. Beaver Valley Power Station, Unit No. 2, is a designed barrel/baffle upflow plant, which is less limiting for fuel assembly blockage as stated above and provides an alternate path to provide cooling to the core.

BVPS-2 Chemical Precipitates The BVPS-2 containment sump buffer was changed from sodium hydroxide to sodium tetraborate during the fall 2009 refueling outage (2R14). Additionally, iodine filters that contain a significant amount of thin aluminum that would have been submerged were removed from the BVPS-2 containment. These modifications have resulted in a decrease in the quantity of post-LOCA chemical precipitates aluminum oxyhydroxide and sodium aluminum silicate. For the loop break, no aluminum oxyhydroxide is

Attachment 2

L-13-176 Page 6 of 8 predicted to be formed, and the other chemical precipitates of concern are predicted to be in the form of sodium aluminum silicate.

Tests have been performed by the Argonne National Laboratory that compare the head loss properties of atuminum oxyhydroxide, sodium aluminum silicate, and other potential chemical precipitates.

These tests have shown that much more sodium aluminum silicate than aluminum oxyhydroxide is needed to cause significant head loss. In the Westinghouse fuel assembly test program, chemical precipitates were represented by al um inum o4yhyd roxide.

Therefore, post-LOCA chemical precipitates of concern have been reduced, and FA testing is based on a more limiting chemical precipitate (aluminum oxyhydroxide) than the form predicted to be present.

Summary of Defense-in-Depth Measures BVPS-2 has a low level concentration of precipitates formed during a loss of coolant accident due to the reduction of available submerged aluminum in containment and the change of the containment sump buffering agent from sodium hydroxide to sodium tetraborate.

The existing emergency operating procedure guidance for transfer to hot leg recirculation is 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after initiation of a loss of coolant accident. For this time frame, considering injection time and transfer to recirculation time, it is expected that a reduced concentration of precipitates would be formed. However, to enhance the capability to address possible core inlet blockage, the BVPS-2 emergency operating procedures will be revised using recent guidance from the PWROG to implement early switchover to hot leg recirculation should plant parameters indicate that core blockage is occurring.

This action will be taken prior to transfer to the existing "Response to Degraded Core Cooling" procedure.

Appropriate operator training will be completed to address this emergency operating procedure revision prior to implementation.

These actions will be completed within six months of the submittal date of this Generic Safety lssue 191 resolution plan.

Existing defense-in-depth mitigative measures are described below.

Containment Sump Screen Actions taken in response to NRC Bulletin 2003-01, "Potential lmpact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors,n are described in Referen ce 12. These actions continue to remain in effect at BVPS-2.

In-Vessel The following describes the plant specific design features and procedural capabilities that exist for detecting and mitigating fuel blockage.

L-1 3-1 76 Page 7 of 8 Detection of Inadequate Reactor Core Flow

- Increasing core exit thermocouple temperature indication Core exit thermocouples are monitored as part of emergency operating procedure monitoring of status trees and the safety parameter display system. As part of operator

training, the operating crew must demonstrate the ability to detect increases in core exit thermocouple temperature indication and transition to the appropriate emergency operating procedure for dealing with this condition. This guidance is provided in the BVPS emergency operating procedure

.User's Guide" section titled "Control Room Usage of Status Trees," and the BVPS-2 "Core Cooling Status Trees" procedure.

- Decreasing reactor water level indication The reactor vessel level indication system is monitored throughout the emergency operating procedures.

Through continuing training, operators demonstrate the ability to monitor and understand the implications of a decreasing reactor vessel water level and appropriately transition within the emergency operating procedure framework to mitigate this condition.

- Increasing containment or auxiliary building radiation levels Increasing radiation levels will be indicated by alarms in the control room with specific procedural steps in both alarm response procedures and the emergency operating procedures for addressing the condition.

Mitigation of Inadequate Reactor Core Flow

- Start a reactor coolant pump Beaver Valley Power Station, Unit No. 2, procedure titled, "Response to Inadequate Core Cooling,"

provides direction to start a reactor coolant pump if core exit thermocouple temperature indication is greater than 12A0 degrees Fahrenheit.

This action would aid in the removal of the established blockage to the core to once again allow normal recirculation injection flow paths to become effective at maintaining adequate core cooling.

- lmplementation of Severe Accident Management Guidelines Severe Accident Management Guidelines (SAMG) provide additional guidance and actions for addressing inadequate core flow conditions.

At BVPS-2 the SAMGs are entered from the following function restoration procedure for inadequate core cooling.

Procedure titled, "Response to Inadequate Core Cooling,"

when core exit thermocouple temperatures are greater than 120A degrees Fahrenheit and actions to cool the core are not successful.

L-13-176 Page 8 of I The SAMGs provide for alternate injection paths into the reactor coolant system (RCS).

Specific SAMGs that provide guidance in this area are listed below.

"lnject Into the RCS' guideline "RCS Injection to Recover Core" guideline The SAMGs provide for flooding containment to provide for convective circulation cooling of the reactor. Specific SAMGs that provide guidance in this area are listed below.

"lnject Into Containment" guideline "Flood Containment" guideline "Containment Water Level And Volume" guideline Although these measures are not expected to be required based on the very low probability of an event that would result in significant quantities of debris being transported to the reactor vessel that would inhibit the necessary cooling of the fuel, they do provide additional assurance that the health and safety of the public would be protected. These measures provide support for the extension of time required to compfetely address Generic Letter 2004-02 for BVPS-2.

Conclusion The Generic Safety lssue 191 resolution path for BVPS-2 is acceptable, based on the information provided in this document. The execution of the actions identified in this document will result in successful resolution of Generic Safety lssue 191 and closure of Generic Letter 2044-02.

1.

2.

ATTACHMENT 3

L-13-176 References Page 1 of 2 NRC Staff Review Guidance Regarding Generic Letter 2A04-A2 Closure in the Area of Strainer Head Loss and Vortexing, dated March 2008, Accession No.

M1080230038.

FENOC Letter L-09-152,

Subject:

Supplemental

Response

to Generic Letter 2AA4-02 (TAC Nos. MC4665 and MC4666),

dated June 30, 2009, Accession No. ML091830390.

FENOC Letter L-10-1 15,

Subject:

Response

to Request for Additional Information Related to Generic Letter 2004-02 (TAC Nos. MC4665 and MC4666), dated September 28, 2010, Accession No. ML102770023.

NRC Letter,

Subject:

NRC Review of Nuclear Energy lnstitute Clean Plant Acceptance Criteria for Emergency Core Cooling Systems, dated May 2,2A12, Accession No. ML120730181 NRC Letter,

Subject:

Summary of April 21,2010 Category 1 Teleconference with FirstEnergy Nuclear Operating Company on Generic Letter 2004-02 (TAC Nos.

MC4665 and MC4666),

dated May 18,2010, Accession No. ML101320665.

Westinghouse Report WCAP-16793-NP, Revision 2, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid," dated October 2011, Accession No. ML112924421.

NRC Letter,

Subject:

Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2, Evaluation of Long-Term Cooling Considering Particulate Fibrous and Ghemical Debris in the Recirculating Fluid" (TAC No. ME1234),

dated April 8,2013, Accession Nos. ML13084A152 and M113084A154.

Pressurized Water Reactor Owners Group Letter OG-12-395,

Subject:

PWR Owners Group GSI-191 In-Vessel Debris Program (PA-SEE-0312 Rev. 4 and PA-SEE-4872, Revision 0), dated September 20, 2012, Accession No. ML122900033.

FENOC Letter L-1 1 -141,

Subject:

License Amendment Request 10-021, Replacement of Beaver Valley Power Station Unit No. 1 Spray Additive System by Containment Sump pH Control System, dated May 27,2A11

, Accession No.

ML111510646.

10. NRC Letter,

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 - lssuance of Amendments Regarding the Spray Additive System by Containment Sump pH Control System (TAC Nos. ME6352 and ME6353),

dated March 14,2012, Accession No. ML120530591 3.

4.

5.

6.

7.

8.

9.

L-1 3-1 76 Page 2 of 2

11. Pressurized Water Reactor Owners Group Letter OG-1 1-291, Subject PWR Owners Group For Information Only - WCAP-17057-P/NP, Revision 1, "GSl-191 Fuel Assembly Test Report for PWROG," (PA-SEE-0312, Revision

2) dated October 12, 2011, Accession No. ML11293A098 12.NRC Letter,

Subject:

BeaverValley PowerStation Unit Nos. 1 and 2 (BVPS-1 and 2)

Response

to NRC Bulletin 2003-01, Potential lmpact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors (TAC Nos. MB9554 and M89555), dated September 6, 2005, Accession No. ML052410375.

13. FENOC Letter L-08-236,

Subject:

License Amendment Request No.08-006, Replacement of Beaver Valley Power Station Unit No. 2 Spray Additive System by Containment Sump pH Control System, dated September24,2008, Accession No. M1082730716.

14. FENOC Letter L-08-350, Subject. Response to Request for Supplemental Information Regarding Containment Spray Additive System License Amendment Request (TAC Nos. MD9734 and MD9735),

dated November 10, 2008, Accession No. M1083180133.

15.NRC Letter,

Subject:

BeaverValley Power Station, Unit Nos. 1 and 2 - lssuance of Amendments Re: Spray Additive System by Containment Sump pH Control (TAC Nos. MD9734 and MD9735),

dated April 16, 2009, Accession No. M1090780352.

ATTACHMENT 4

L-13-176 Regulatory Commitment List Page 1 of 2 The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not Regulatory Commitments.

Please notify Mr. Thomas A. Lentz, Manager - Fleet Licensing, at 330-315-6810 of any questions regarding this document or associated Regulatory Commitments.

Requlatory Commitment Due Date

1. Measurements will be taken in This action will be accomplished preparation for insulation modifications during the BVPS-1 refueling outage associated with the BVPS-1 pressurizer in the fall of 2013.

safety relief valve inlet lines, if it is determined that the PWROG in-vessel effects testing effort does not support closure of Generic Safety lssue 191 and Generic Letter 2004-02 for BVPS-1.

2. lnsulation for the BVPS-1 pressurizer This action will be accomplished by safety relief valve inlet lines will be the end of the refueling outage in replaced or modified as appropriate if it is the fall af 2A16, if needed.

determined that the PWROG in-vessel effects testing effort does not support closure of Generic Safety lssue 191 and Generic Letter 2004-02 for BVPS-1.

3. The final Generic Letter 2OO4-02 Within 6 months after the NRC supplemental response for BVPS-1 will

approves, by issuance of a safety be provided to the NRC.

evaluation, the new PWROG topical report addressing additional in-vessel effects testing efforts that are currently being pursued.

4. Updated Final Safety Analysis Report Following NRC acceptance of the changes will be completed to update the final docketed Generic Safety current licensing basis for BVPS-1 as lssue 191 response for BVPS-1 and appropriate.

completion of any identified_

insulation modifications for BVPS-1 that may be required.

Attachment 4

L-13-176 Page2 ot 2 Requlatorv Commitment Due Date

5. The BVPS-I emergency operating These actions will be completed procedures will be revised using recent within six months of the submittal guidance from the PWROG to implement date of this Generic Safety early switchover to hot leg recirculation lssue 191 resolution plan.

should plant parameters indicate that core blockage is occuning.

This action will be taken prior to transfer to the existing "Response to Degraded Core Cooling' procedure.

Appropriate operator training will be completed to address this emergency operating procedure revision prior to implementation

6. The final Generic Letter 2004-02 Within 6 months after the NRC supplemental response for BVPS-2 will

approves, by issuance of a safety be provided to the NRC.

evaluation, the new PWROG topical report addressing additional in-

1,:J'",iJ":T1"3,"J

7. Updated Final Safety Analysis Report Following NRC acceptance of the changes will be completed to update the final docketed Generic Safety current licensing basis for BVPS-2 as lssue 191 response for BVPS-2.

appropriate.

8. The BVPS-2 emergency operating These actions will be completed procedures will be revised using recent within six months of the submittal guidance from the PWROG to implement date of this Generic Safety early switchover to hot leg recirculation lssue 191 resolution plan.

should plant parameters indicate that core blockage is occurring. This action will be taken prior to transfer to the existing 'Response to Degraded Core Cooling" procedure. Appropriate operator training will be completed to address this emergency operating procedure revision prior to implementation.