L-08-269, Supplement Reply to Request for Additional Information for the Review of License Application and License Renewal Application Amendment No. 22

Supplement Reply to Request for Additional Information for the Review of License Application and License Renewal Application Amendment No. 22

ML082390815
Person / Time
Site:	Beaver Valley
Issue date:	08/22/2008
From:	Sena P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-269, TAC MD6593, TAC MD6594
Download: ML082390815 (12)

Text

FENOC Beaver Valley Power Station P.O.Box 4 Shippingport,PA 15077 FirstEnergyNuclearOperatingCompany PeterP. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 August 22, 2008 L-08-269 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Supplement to Reply to Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos. MD6593 and MD6594) and License Renewal Application Amendment No. 22 Reference 1 provided the FirstEnergy Nuclear Operating Company (FENOC) License Renewal Application (LRA) for the Beaver Valley Power Station (BVPS). References 2 and 3 requested additional information from FENOC regarding the BVPS license renewal integrated plant assessment in Sections 2.4 and B.2.6 of the BVPS LRA.

During a conference call between FENOC and the U.S. Nuclear Regulatory Commission (NRC) on August 11, 2008, regarding the FENOC reply in References 2 and 3, the NRC staff asked for supplemental information to clarify ten of the FENOC responses. The Attachment provides the ten FENOC supplemental responses to the NRC request for additional information. The Enclosure provides Amendment No. 22 to the BVPS LRA.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Clifford I. Custer, Fleet License Renewal Project Manager, at 724-682-7139.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August A,2008.

Pete Pely, Peter P. Sena Ill

Beaver Valley Power Station, Unit Nos. 1 and 2 L-08-269 Page 2

References:

1. FENOC Letter L-07-113, "License Renewal Application," August 27, *2007.

2. FENOC Letter L-07-21 1, "Reply to Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos. MD6593 and MD6594), License Renewal Application Amendment No. 18, and Revised License Renewal Boundary Drawing,"

July 24, 2007.

3. FENOC Letter L-08-190, "Reply to Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos. MD6593 and MD6594) and License Renewal Application Amendment No. 12," June 9, 2008.

Attachment:

Supplement to Reply to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Sections 2.4 and B.2.6

Enclosure:

Amendment No. 22 to the BVPS License Renewal Application cc: Mr. K. L. Howard, NRC DLR Project Manager Mr. S. J. Collins, NRC Region I Administrator cc: w/o Attachment or Enclosure Mr. B. E. Holian, NRC DLR Director Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRC DORL Project Manager Mr. D. J. Allard, PA BRP/DEP Director Mr. L. E. Ryan, PA BRP/DEP

ATTACHMENT L-08-269 Supplement to Reply to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Sections 2.4 and B.2.6 Page 1 of 6 Section 2.4 Question RAI 2.4.12-1 (Follow up)

The RAI response states that the connection between retaining wall and the building is minor but it did not make a conclusion relative to potential adverse interaction between the wall and the building.

RESPONSE RAI 2.4.12-1. (Follow up)

The Beaver Valley Power Station (BVPS) Emergency Response Facility Substation Building and the neighboring retaining wall are nonsafety-related, non-seismic structures. The Emergency Response Facility Substation Building is in-scope for 10 CFR 54.4(a)(3) support of Fire Protection functions. NRC scoping guidance

("License Renewal Issue 98-0082," contained in Nuclear Energy Institute (NEI) document NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule," Appendix C, "References,"

Reference 12, page C-78) excludes the need to include nonsafety-related structures that could interact with structures that meet the 10 CFR 54.4(a)(3) scoping criteria, unless the current licensing basis (CLB) specifically includes such a consideration or operating experience illustrates that an adverse interaction could exist. The retaining wall is not identified in the BVPS CLB, nor does plant or industry operating experience indicate that an adverse interaction problem could exist. Therefore, there is no basis for potential adverse interaction between the wall and the Emergency Response Facility Substation Building, and the wall is not in-scope.

Question RAI 2.4.15-1 (Follow up)

Need to discuss the response. For example, the interior walls for control building (Pg 2.4-20) in Table 2.4-7 have "SHD" intended function. How is this case different from the response to RAI 2.4.15-1?

RESPONSE RAI 2.4.15-1 (Follow up)

The Structure Description in BVPS License Renewal Application (LRA) Section 2.4.7, "Control Building (Unit 2 only)," identifies that a function of the Control Building is to

Attachment L-08-269 Page 2 of 6 provide radiological shielding for the control room operating crew during and following an accident. However, the Gaseous Waste Storage Vault, which is the subject of RAI 2.4.15-1, is not credited with providing radiological shielding to plant personnel during or following an accident or for providing shielding in support of any 10 CFR 54.4(a)(1),

(a)(2) or (a)(3) functions, since no plant personnel need to access the structure for plant safe shutdown or accident mitigation actions. Radiological shielding, therefore, is not an intended function of the Gaseous Waste Storage Vault walls.

Question RAI 2.4.22-3 (Follow up)

Need to further discuss the function of blow-out panels and in general understand their configuration, location, etc.

RESPONSE RAI 2.4.22-3 (Follow up)

The BVPS Unit 2 blowout panels are located on the "roof' of the Reactor Cavity contained within the Primary Shield wall and located below the Reactor Cavity Water Seal (this area is also known as the Incore Instrument Tunnel "keyway" roof). There are two disks; one mounted on either side of a vertical, conical metal housing. The 24-inch diameter stainless steel disks are designed to "fail" at 1.5 - 2.0 psig pressure to vent the area below the reactor.

The Leak-before-break analyses have eliminated the need for these disks at Unit 1, and they are not credited for pressure relief. Therefore, for Unit 1, the blowout panels are not in-scope for license renewal.

Question RAI 2.4.22-4 (Follow up)

Vortex baffles and sump modifications - open item for Sections 2.4.22 and Table 3.5.2-22.

RESPONSE RAI 2.4.22-4 (Follow up)

Vortex baffles have been removed from the Unit 1 containment sump, and vortex devices have been added to the Unit 2 containment sump as a result of recent modifications associated with Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR Sump Performance," and NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors." Component function "DF" (Direct flow) has

Attachment L-08-269 Page 3 of 6 been added to the "Sump screen assembly and liner" component at both units. LRA Tables 2.4-22 and 3.5.2-22 are revised to reflect these changes.

Based on a review of the Unit 1 and Unit 2 containment sump modifications, other than the vortex baffles, there are no additional component types, materials or environments that have not previously been addressed in the BVPS LRA. Therefore, no other LRA changes are necessary.

See the Enclosure to this letter for the revision to the BVPS LRA.

Question RAI 2.4.22-5 (Follow up)

Does BVPS have high alumina cement? (Ref: IN 97-11 and 98-26)

RESPONSE RAI 2.4.22-5 (Follow up)

Review of the BVPS Unit 1 porous concrete mix design indicates that no alumina cement was used. The Unit 2 porous concrete mix design, however, does specify calcium aluminate (high alumina) cement. However, the Containment structures at both units are founded well above the site's normal groundwater level.

FirstEnergy Nuclear Operating Company (FENOC) reviewed NRC Information Notice (IN) 98-26, "Settlement Monitoring and Inspection of Plant Structures Affected by Degradation of Porous Concrete Subfoundations," dated July 24, 1998, and IN 97-11, "Cement Erosion From Containment Subfoundations at Nuclear Power Plants," dated March 21, 1997. It was noted that IN 97-11 identified that North Anna and Surry had identified groundwater in their relief sumps. North Anna and Surry used the same Architect/Engineer/Constructor, Stone & Webster, and are of similar design and construction, as BVPS Unit 1. Neither North Anna or Surry used calcium aluminate cement under the Containment structures, which is consistent with the design of BVPS Unit 1 (constructed following Surry's construction and prior to the construction of the North Anna plants).

The North Anna and Surry findings mirror the history of BVPS, in that water has been found in the Unit I relief sumps. However, the BVPS Unit I water was concluded to be rainwater, and not groundwater. The Unit 1 Containment instrument pit sumps' access shafts are located outside the Containment. Their entrances are above grade, and part of their length abuts the Containment wall above the waterproof membrane. Rainwater running off of the Containment enters the ground and then the access shaft -

accumulating in the sump at the bottom of the shaft. Conversely, the Unit 2 Containment instrument pit sumps' access shafts are located inside the Safeguards Building, and those sumps have remained dry.

Attachment L-08-269 Page 4 of 6 Question RAI 2.4.28-1 (Follow up)

The response to the RAI only addresses floors 4 thru 7. Need to discuss floors 1 thru 3?

RESPONSE RAI 2.4.28-1 (Follow up)

Floors 4 through 7 of the South Office and Shops Building were referenced in the FENOC response to RAI 2.4.28-1 because the structure's record design calculations focused on those floors for diaphragm action. The lower floors were also evaluated, but primarily for special deadloads (shop use).

The existing floor slab design configuration is essentially constant throughout the structure, and the license renewal Structures Monitoring Program requires inspection of all floor level slabs.

Question RAI 2.4.31-2 (Follow up)

The turbine building foundation elevation is 715.3 per Fig 2.5.4-41. If the grade elevation is at 730'-6", we have approximately 15 feet difference between the foundation elevation and grade elevation. Need more clarification/discussion.

RESPONSE RAI 2.4.31-2 (Follow up)

The Unit 2 Turbine Building foundation elevation of 715.3' per Updated Final Safety Analysis Report (UFSAR) Figure 2.5.4-41 is the lowest elevation of numerous founding elevations for the structure, since its base mat varies in thickness in different areas.

The 715.3' elevation portion houses the 108" circulating water lines. Most upper surfaces of the basement slabs are at 730'-6" or within five (5) feet of it (725'-6"). A five foot deep grade beam exists at the perimeter of the building where the slab upper surfaces are lower. The grade beam spans between piers that are founded on the base mat sections, which support the building's perimeter steel columns. The perimeter grade beam is part of the Turbine Building foundation mat, which is included in the BVPS LRA as in-scope and subject to aging management review.

Attachment L-08-269 Page 5 of 6 Question RAI 2.4.33-1 (Follow up)

Same question as RAI 2.4.15-1... Need to discuss the response. For example, the interior walls for control building (Pg 2.4-20) in Table 2.4-7 have "SHD" intended function. How is this case different from the response to RAI 2.4.33-1?

RESPONSE RAI 2.4.33-1 (Follow up)

As presented for the RAI 2.4.15-1 (Follow up) response, the Structure Description in BVPS License Renewal Application (LRA) Section 2.4.7, "Control Building (Unit 2 only),"

identifies that a function of the Control Building is to provide radiological shielding for the control room operating crew during and following an accident. However, the Waste Handling Building, which is the subject of RAI 2.4.33-1, is not credited with providing radiological shielding to plant personnel during or following an accident or for providing shielding in support of any 10 CFR 54.4(a)(1), (a)(2) or (a)(3) functions, since no plant personnel need to access the structure for plant safe shutdown or accident mitigation actions. Radiological shielding, therefore, is not an intended function of the Waste Handling Building walls.

Question RAI 2.4.36-1 (Follow up)

Need to clarify that the current licensing basis does not require consideration of interaction.

RESPONSE RAI 2.4.36-1 (Follow up)

1. Seismic and tornado interaction - The superstructures of the Unit 1 and Unit 2 Turbine Buildings, which are in-scope for license renewal and structurally support the cranes that are the subject of the follow up question, are designed to not collapse onto adjacent structures due to an earthquake or tornado-generated loading. Therefore, no Turbine Building crane-related seismic or tornado interaction needs to be considered.

2. Missile generation - The possible generation of missiles by collapse of the cranes is evaluated in the UFSAR as a general statement for all plant equipment. Specifically, any missile generated by the "breakup" of a "nontornado" structure is not as severe as the most critical design basis tornado missile. Therefore, no Turbine Building crane collapse-generated missile interaction needs to be considered.

3. Interaction with in-scope systems, structures or components - The Unit 1 UFSAR states that loss of the safety-related Unit 1 River Water piping where it enters the Turbine Building will not result in loss of system function. This issue was addressed

Attachment L-08-269 Page 6 of 6 in the response to RAI 2.1-5, provided in FENOC letter L-08-123, dated April 3, 2008. Nonsafety-related system, structure or component interactions with quality class "Q" instruments in the Turbine Buildings was addressed in the response to RAI 2.1-4, also in FENOC letter L-08-123. The Unit 1 and Unit 2 Turbine Building cranes are not discussed elsewhere in the CLB as having a possible adverse interaction with in-scope equipment systems, structures or components. Therefore, no Turbine Building crane-related interaction with in-scope equipment needs to be considered.

Based on the above, the CLB for Unit 1 and Unit 2 Turbine Building cranes does not require consideration of interaction.

Section B.2.6 Question RAI B.2.6-3 (Follow up)

Please clarify the scope of the BVPS Bolting Integrity Program.

RESPONSE RAI B.2.6-3 (Follow up)

The BVPS Bolting Integrity Program applies to safety-related and nonsafety-related bolting for pressure retaining components, bolting for Nuclear Steam Supply System component supports, and structural joint bolting, except for the Reactor Vessel closure studs and the Reactor Vessel Internals bolting. The program applies to all bolting within the scope of license renewal, regardless of size, with the exceptions noted above. The Reactor Vessel closure studs are addressed by the Reactor Vessel Closure Studs Program. The Reactor Vessel Internals bolting is addressed by the Water Chemistry Program and the license renewal future commitments listed in LRA Appendix A, Tables A.4-1 (Unit 1) and A.5-1 (Unit 2), related to management of Reactor Vessel Internals.

ENCLOSURE Beaver Valley Power Station (BVPS), Unit Nos. 1 and 2 Letter L-08-269 Amendment No. 22 to the BVPS License Renewal Application Page 1 of 4 License Renewal Application Sections Affected Table 2.4-22 Table 3.5.2-22 The Enclosure identifies the correction by Affected License Renewal Application (LRA)

Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text ined-out and added text underlined.

Enclosure L-08-269 Page 2 of 4 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Table 2.4-22 Page 2.4-69 Unit 1, 2 2 nd Row - New Intended Function, and 2 3 rd Row (Delete)

Page 2.4-71 Unit 2, 1 1 th Row - New Intended Function, and New Row LRA Table 2.4-22, "Reactor Containment Building Components Subject to Aging Management Review," requires revision because vortex baffles were removed from the Unit 1 containment sump, and vortex devices were added to the Unit 2 containment sump, as a result of recent modifications associated with Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on PWR Sump Performance," and NRC Generic Letter (GL) 2004 02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors." LRA Table 2.4-22 is revised, affecting rows 22 and 23 on page 2.4-69, row 11 on page 2.4-71, and a new added row, and now reads:

Component Type Intended FunctionW Unit I Reactor Containment Building Sump screen assembly and liner DF, SSR Voede*bafflea Unit 2 Reactor Containment Building Sump screen assembly and liner DF, SSR Vortex devices Dr,SS

6 Enclosure L-08-269 Page 3 of 4 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Table 3.5.2-22 Page 3.5-152 Unit 1, Row 44 - New Intended Function, and Rows 45-47 (Delete)

Page 3.5-160 Unit 2, Row 102 - New Intended Function, and New Row LRA Table 3.5.2-22, "Containments, Structures, and Component Supports - Reactor Containment Building -

Summary of Aging Management Evaluation," requires revision because vortex baffles were removed from the Unit 1 containment sump, and vortex devices were added to the Unit 2 containment sump, as a result of recent modifications associated with Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR Sump Performance," and NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors." LRA Table 3.5.2-22 is revised, affecting rows 44-47 on page 3.5-152, row 102 on page 3.5-160, and a new added row, and now reads:

Aging Effect Aging M NUREG-Row Component Intended Material Environment Requiring anagement 1801 Table 1 Notes No. Type Function Management Program Volume 2 Item Item Unit 1 44 Sump screen DF, SSR Stainless Protected None None 1l.B5-6 3.5.1-5 C assembly and steel from weather (TP-4) liner 45 Veebaffles SSR Galvanized Preteeted Less of matoria! Boric Acid Corroson 115-4 3-.5155 steel ftem -weatheF (8..7) fp-3) 517

Enclosure L-08-269 Page 4 of 4 Neie None Arem- weather Protected None None assembly and from weather liner