ML093000505

Response to Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant License Renewal Application (TAC No. ME0274) - and Amendment #6

ML093000505
Person / Time
Site:	Crystal River
Issue date:	10/22/2009
From:	Franke J Florida Power Corp, Progress Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1009-08, TAC ME0274
Download: ML093000505 (18)

Text

iProgress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 54 October 22, 2009 3F1009-08 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant License Renewal Application (TAC NO. ME0274) - and Amendment #6

References:

(1) CR-3 to NRC letter, 3F1208-01, dated December 16, 2008, "Crystal River Unit 3 - Application for Renewal of Operating License" (2) NRC to CR-3 letter, dated September 22, 2009, "Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant License Renewal Application (TAC NO. ME0274)"

Dear Sir:

On December 16, 2008, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc. (PEF), requested renewal of the operating license for Crystal River Unit 3 (CR-3) to extend the term of its operating license an additional 20 years beyond the current expiration date (Reference 1). Subsequently, the Nuclear Regulatory Commission (NRC), by letter dated September 22, 2009, provided a request for additional information (RAI) concerning the CR-3 License Renewal Application (Reference 2). Enclosure 1 to this letter provides the response to Reference 2. Enclosure 2 provides Amendment #6 to the License Renewal Application.

No new regulatory commitments are contained in this submittal.

If you have any questions regarding this submittal, please contact Mr. Mike Heath, Supervisor, License Renewal, at (910) 457-3487, e-mail at mike.heath@pgnmail.com.

Si Franke Vice President Crystal River Unit 3 JAF/dwh

Enclosure:

1. Response to Request for Additional Information

2. Amendment #6 - Changes to the License Renewal Application xc: NRC CR-3 Project Manager NRC License Renewal Project Manager NRC Regional Administrator, Region II Senior Resident Inspector Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Power Line Street rJCL (L2 Crystal River, FL 34428

U. S. Nuclear Regulatory Commission Page 2 of 2 3F1009-08 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

(Jon A. Franke Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this c____ day of Oc)zbe , 2009, by Jon A. Franke.

Signature of Notary Public State of Florida (Print, type, or stamp Commissioned Name of Notary Public)

Personally R Produced Known / -OR- Identification

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

U. S. Nuclear Regulatory Commission Enclosure 1 3F1009-08 Page 1 of 5 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Request for Additional Information (RAI) 2.4.2.1-1 The watertight sleeves around the raw water sump vents that protect the Auxiliary Building (AB) at an elevation of 95 ft against flood levels up to an elevation of 129 ft are listed as additional components required for local protection in the Crystal River Unit 3 Nuclear Generating Plant (CR-3) AB final safety analysis report (FSAR) Section 2.4.2.4. This commodity/component is not listed in Table 2.4.2-1 of the CR-3 license renewal application (LRA) for the AB. Please justify its exclusion from the scope of license renewal.

Response

The watertight sleeves around the raw water sump vents that protect the AB are included in the scope of License Renewal. The watertight sleeves are included with the "Other Miscellaneous Structures" in the component/commodity group "Platforms, Pipe Whip Restraints, Jet Impingement Shields, Masonry Wall Supports, and Other Miscellaneous Structures"in the CR-3 License Renewal basis document. The watertight sleeves are discussed in LRA Subsection 2.4.2.1. The watertight sealant is included with the component/commodity group "Seals and Gaskets." The grout is included with the component/commodity group "Concrete: Above Grade."

However, LRA Table 3.5.2-2 should have included an aging management review (AMR) line item for stainless steel material in the Air - Outdoor environment for component/commodity group "Platforms,Pipe Whip Restraints, Jet Impingement Shields, Masonry Wall Supports, and Other Miscellaneous Structures." This LRA change is shown in Enclosure 2 to this submittal.

RAI 2.4.2.3-1 As mentioned in the CR-3 LRA, the borated water storage tank (BWST) and shield wall structure include an attached reinforced concrete structure that has been abandoned but contains several components that are included in scope of license renewal, as stated in the CR-3 LRA on page 2.4-15. Clarify if this abandoned structure has been included in scope or justify its exclusion as having no impact, especially on the flood barrier capabilities on the adjacent structures that are included in scope of license renewal and therefore subject to an aging management review (AMR).

Response

The attached reinforced concrete structure is part of the BWST Foundation and Shield Wall structure and is included in the scope of License Renewal. Some of the equipment inside the attached reinforced concrete structure was abandoned,but not the structure. (The LRA states that the structure contains abandoned tanks, but the LRA does not state that the structure itself is abandoned.) The specific structural component/commodities for the attached structure include the reinforced concrete structure, the anchorage/embedmentsfor support steel and pipe supports, a platform, supports for the ventilation fan and duct work, pipe supports, and a door.

These are identified in the License Renewal basis document.

U. S. Nuclear Regulatory Commission Enclosure 1 3F1009-08 Page 2 of 5 During the preparationof this RAI response, it was determined the door to the structure is not a flood door. The flood door to the BWST access area is located in the concrete flood barrierwall described in LRA Subsection 2.4.2.15. This response requires changes to the LRA that are described in Enclosure 2 to this submittal.

RAI 2.4.2.3-2 STYROFOAM TM covers a full range of extruded polystyrene building products used primarily for M T

insulation of floors, walls and roof systems. Please clarify if the one inch thick STYROFOAM filler, located in the gap between the concrete missile barrier and side of the borated water storage tank, is included in scope of license renewal or justify its exclusion from the scope of license renewal.

Response

The Styrofoam material attached to the BWST is not in the scope of License Renewal, because it performs no License Renewal intended functions. The Styrofoam was used as a filler between the tank liner plate and the wall during placement of the concrete wall around the BWST. The Styrofoam does not provide a support or protection function for the BWST or the concrete structure.

During operation, the maximum temperature of the BWST is 100'F per the applicable design basis document. This temperature is below the temperaturethat could cause degradation to the concrete structure due to elevated temperatures. The non-safety related Styrofoam was not credited for freeze protection in the License Renewal basis calculation because CR-3 utilizes proceduralizedcold weather monitoring which ensures equipment availabilityand function are not impacted. Design requirements are met without reliance on the Styrofoam insulation to maintain safety related functions.

RAI 2.4.2.3-3 The BWST is listed as a Class 1 structure per CR-3 FSAR, Section 5.1.1.1. Its foundation is made of reinforced concrete and its primary use is to support and provide missile protection to the BWST which is a mechanical component. Please clarify if this Class 1 reinforced concrete foundation is completely above grade or else justify why the below grade concrete is not listed in Table 2.4.2-3 of the CR-3 LRA, as in scope of license renewal and subject to an AMR.

Response

The reinforced concrete BWST Foundation and Shield Wall structure described in LRA Subsection 2.4.2.3 was placed directly on the AB reinforced concrete slab at 119 ft. elevation.

A reinforced concrete foundation is not used. The reinforced concrete shield wall attaches directly to the AB reinforced concrete slab at 119 ft. elevation. Plant grade is at elevation 118.5 ft., therefore the structure is completely above grade.

U. S. Nuclear Regulatory Commission Enclosure 1 3F1009-08 Page 3 of 5 RAI 2.4.2.4-1 Both bridges forming the cable bridge span the discharge canal and provide support for electrical circuits required to mitigate a postulated station blackout (SBO) event. The SBO conduits are considered to be within the cable bridge structure from where they exit the ground on one side to cross the bridge to where they re-enter the ground on the other side of the bridge. Due to the proximity to a body of water and the entrance/exit of electrical cables required to mitigate an SBO event, from the ground and their supports, please indicate if there are any seals, gaskets or any other applicable flood barriers or insulation that should be included in Table 2.4.2-4 of the CR-3 LRA, and therefore subject to an AMR.

Response

There are no seals, gaskets, flood barriers, or insulation associated with either non-safety related cable bridge.

The east cable bridge has SBO conduits which enter and exit the structure in open manholes which are part of the cable bridge abutment. The in scope cables exit the conduit to cable trays which are supported on the cable bridge.

The west cable bridge structure has SBO conduits which exit the ground on the north and south sides of the bridge and are supported by the bridge. These conduits are continuous and do not have any seals. The west cable bridge is also an enclosed concrete tunnel which internally supports DC power cables entering from a common Unit I and 2 concrete tunnel. These cables terminate in an electrical panel in the 230 KV Terminal House before entering a cable trench located in the 230 KV Switchyard.

In addition, FSAR Section 2.4.2.4 does not identify any flood protection features associatedwith

,the cable bridges at CR-3.

RAI 2.4.2.4-2 In Table 2.4.2-4 for the cable bridge section of the CR-3 LRA, the cable tray, conduit, heating, ventilation, and air conditioning (HVAC) ducts and tube tracks are listed as being in scope of license renewal and therefore subject to an AMR. Please clarify if the HVAC ducts system component supports are needed and included in scope of license renewal or justify its exclusion from Table 2.4.2-4 of the CR-3 LRA.

Response

The cable bridge structure described in LRA Subsection 2.4.2.4 does not include heating, ventilation, and air conditioning (HVAC) ducts and tube track. The methodology employed by CR-3 used a generic component/commodity group for "Cable Tray, Conduit, HVAC Ducts, Tube Track" throughout LRA Sections 2.4 and 3.5. Refer to the discussion of "Component!

Commodity" in LRA Section 3.0, page 3.0-3. In the cable bridge structure, only Cable Tray and Conduit are applicable.

U. S. Nuclear Regulatory Commission Enclosure 1 3F1009-08 Page 4 of 5 RAI 2.4.2.8-1 Protection of the intake structure during a postulated probable maximum hurricane peak tide is provided by a cut-off wall extending downward into the competent caprock at the entrance to the structure (See CR-3 FSAR 2.4.2.4). Please confirm that this wall is considered part of the intake structure and thus included in scope of license renewal or else justify its exclusion from the scope of license renewal.

Response

This cut-off wall is considered part of the Circulating Water Intake Structure and is included in the scope of License Renewal. Intake Structure design drawings show the face of the bottom mat of the Circulating Water Intake Structure extends down from elevation 67 ft. 0 in. for seven feet at the entrance to the Circulating Water Intake Structure from the Intake Canal. The Intake Canal bottom is shown on plant documents as elevation 67 ft. 0 in. at the Circulating Water Intake Structure.

RAI 4.5-1 The fourth column of the first row in CR-3 LRA, Table 4.5-1, "Summary of Tendon Data," lists the tendon force value extrapolated to the end of the period of extended operation for dome tendons as 1255 kips. However, in CR-3 LRA, Figure 4.5-1, "Projected Force in Dome Tendons," the trend line based on individual lift-off forces from surveillance data indicates that the projected lift-off force in the dome tendons at the end of the period of extended operation (i.e., 63 years after initial tensioning) would be approximately 1330 kips. Please explain the discrepancy between the projected tendon force values at the end of the period of extended operation in the dome tendons indicated in CR-3 LRA, Table 4.5-1, and in CR-3 LRA, Figure 4.5-1 and identify the correct value.

Response

In response to the RAI, it was determined that the correct extrapolated value for the Dome tendons is shown on LRA Table 4.5-2 as 1321 kips. The value reportedin the LRA summary Table 4.5-1 is incorrect and will be updated to agree with the correct value shown in LRA Table 4.5-2. In addition to the RAI response, additional information regarding containment tendon surveillance data has been preparedby the tendon Surveillance Contractorused at CR-3 and is provided below. The revised information affects the data presented in the LRA as discussed in the following paragraphs.

The Final Report for the 3 0th Year Tendon Surveillance has been revised by the Surveillance Contractorto include data from the I"t and 2 fd interval tests in the tendon prestress regression analysis. The revised values have been incorporatedinto the basis document and they affect LRA Tables 4.5-1, 4.5-2, 4.5-3, and 4.5-4. However, the changes in the tables resulted in negligible changes to Figures 4.5-1 through 4.5-6; therefore, the figures were not revised. Refer to the specific LRA changes identified in Enclosure 2 to this submittal.

Table 4.5-1 shows that the computed values demonstrated that prestress in all three groups of tendons should remain above the applicable minimum required values for the period of

U. S. Nuclear Regulatory Commission Enclosure 1 3F1009-08 Page 5 of 5 extended operation ending on December 3, 2036, and that the tendons should maintain their design basis function.

The column of "computedvalues" shown on the original LRA tables has been removed. These values were computed using the regression coefficients for the tendon group applied to each tendon at the time of the surveillance test of the tendon. They would have changed slightly as did the values projected for each group at 63 years because the adjustments to the data resulted in slightly different regression coefficients. The computed values using the regression coefficients for each tendon at the time of each surveillance test are not the same as the value used to judge acceptability of the measured force for the tendon. The values used to judge acceptability are calculated using Regulatory Guide 1.35.1 which incorporates several factors specific to the tendon itself, such as, the order in which it was initially prestressed. Including these computed values based on the regression coefficients on the tables was confusing so they were eliminated.

In considerationof the recent discovery of a gap in the concrete of the outer radius of the CR-3 containment structure (subject of Event Notification 45416, dated October 7, 2009, and NRC Special Inspection Team Press Release No. 11-09-055, dated October 9, 2009), CR-3 will evaluate the need to revise the technical response to this RAI at a later date. This evaluation will be complete following the root cause determination that is currently in progress and subsequent assessment of any impact on the technical and aging management programs discussed in this response.

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE 2 AMENDMENT #6 CHANGES TO THE LICENSE RENEWAL APPLICATION

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 1 of 9 Amendment #6 - Changes to the License Renewal Application Source of License Renewal Application Amendment #6 Changes Change RAI 2.4.2.1-1 For Auxiliary Building stainless steel "Platforms, Pipe Whip Restraints, Jet Impingement Shields, Masonry Wall Supports, and Other Miscellaneous Structures" in Table 3.5.2-2, on LRA page 3.5-76, add the following line item to address Air - Outdoor:

ir - Outdoor Loss of Material Structures Monitoring 35I154-70 (TP-6) 532 RAI 2.4.2.3-1 To update the status of the Door (Non-Fire) in the BWST Foundation and Shield Wall structure, delete the last sentence of the first paragraph of Subsection 2.4.2.3 on page 2.4-15. Also, on page 2.4-15, revise the second sentence under FSAR and Drawing References to read:

FSAR Section 5.1.1.1 identifies the structure itself as Class I.

Delete Intended Function C-8 from the Component/Commodity "Doors (Non-Fire)" on both Table 2.4.2-3 on page LRA 2.4-16 and Table 3.5.2-4 on page 3.5-83.

On LRA Figure 2.2-1, revise the Flood Barrier Wall between the BWST and the Reactor Building to agree with the following sketch:

Reactor Building d'.n jPlant VE Flood Barrier Building RAI 4.5-1 Replace Tables 4.5-1, 4.5-2, 4.5-3, and 4.5-4 on LRA pages 4.5-3 through 4.5-10 with the corresponding following tables.

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 2 of 9 TABLE 4.5-1

SUMMARY

OF TENDON DATA Minimum Required Average Value Extrapolated to End of Tendon Values Period of Extended Conclusion (Kips/Tendon) Operation (Kips/Tendon)

Dome 123 1215 1321 Note 1 Vertical 144 1149 1484 Note 1 Hoop 282 1252 1328 Note 1 Note:

1. The value at the end of the period of extended operation is greater than the minimum required value.

U. S. Nuclear Regulatory Commission Enclosure 2 3F17009-08 Page 3 of 9 TABLE 4.5-2 DOME TENDON DATA Surveillance No. Tendon Years Since Initial Measured Force/Tendon Tensioning D139 3.22 1590 D215 3.25 1644 D221 3.14 1511 D228 3.11 1524 D234 3.1 1513 D340 3.11 1562 D122 5.42 1647 D140 5.41 1587 2 D208 5.4 1593.5 D323 5.47 1525.5 D331 5.5 1460.5 D123 6.99 1304 3 D212* 6.9 1338 D322 6.91 1494 D329 6.92 1506 D105 12.92 1453 4 D212* 12.9 1276 D328 12.89 1619 5 D224 18.96 1425 D231 19.06 1335 D113 22.99 1427 D115 22.7 1380 6 D212* 23.04 1335 D304 23 1598 D311 23.03 1408

U. S. Nuclear Regulatory Commission Enclosure 2 3F17009-08 Page 4 of 9 TABLE 4.5-2 (continued) DOME TENDON DATA Surveillance Tendon Years Since Initial Measured No. TForce/Tendon Tensioning D126 26.81 1376.9 7 D212* 26.82 1292 D339 26.47 1507 D129 32.96 1289.64 8 D212* 32.92 1277 D238 32.89 1511.53 Extrapolated 63 (Note 1) 1321

• Indicates Control Tendon Notes:

1. The extended period of operation will end in the 63rd year from the date of initial tensioning.

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 5 of 9 TABLE 4.5-3 VERTICAL TENDON DATA Surveillance Years Since Measured Tendon Initial Force/Tendon No. FTensioning 12V19 3.26 1590 12V20 2.87 1785 12V21 3.26 1633 1 23V15 3.22 1590 34V06 3.16 1590 45V03 3.19 1678 56V01 3.28 1719 12V12 5.47 1718 12V20 5.2 1740 23V05 5.57 1580 2 34V01 5.54 1569 45V06 5.53 1685 56V01 5.59 1707 56V20 5.48 1630 34V1 9 7.02 1640 45V16 7.01 1575 3

56V11 7.04 1565 61V05 7.07 1519 12V01" 13.07 1535 4 34V04 13.07 1623 56V02 13.04 1648 61V14 19.08 1587 5

56V15 19.18 1541 12V01* 23.15 1471 6 23V02 23.15 1609 61V21 23.25 1525

U. S. Nuclear Regulatory Commission Enclosure 2 3F17009-08 Page 6 of 9 TABLE 4.5-3 (continued) VERTICAL TENDON DATA Surveillance ears Since Measured No. Initial Force/Tendon Tensioning_

12V01* 27.028 1446 12V02 27.12 1546 7 23V24 27.14 1522 45V14 26.94 1552 61V08* 26.91 1476 45V20 33.03 1456.8 8 61V08* 33.05 1505.98 61V17 33.11 1580.18 Extrapolated 63 (Note 1) 1484

• Indicates control tendon -The original control tendon, 12V01, required retensioning in the 3rd and 7th interval surveillance. Tendon 61V08 was selected as the new control tendon for surveillances after the 7 th interval.

Notes:

1. The extended period of operation will end in the 63rd year from the date of initial tensioning.

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 7 of 9 TABLE 4.5-4 HOOP TENDON DATA Surveillance Years Since Measured No. Tendon Initial Force/Tendon Tensioning 13H10 2.733 1524 13H19 2.992 1485 13H37 3.081 1605.5 13H47 3.100 1606 62H9 3.006 1573.5 46H21" 2.925 1502 46H29 2.911 1463 46H31 2.886 1457 461H46 2.861 1464 51H11 2.903 1474 13H22 5.094 1572 13H32 5.097 1611 13H43 5.300 1583 35H24 5.086 1533 35H28 5.086 1430 35H44 5.111 1622 46H42 5.161 1548 51H10 5.111 1572 51H23 5.275 1528 51H37 5.300 1567 13H46 6.592 1546 35H35 6.786 1328 35H40 6.608 1458 42H20 6.614 1544 3 42H40 6.636 1466 46H110 6.678 1478 51H26* 6.569 1424 51H45 6.792 1492 62H34 6.617 1546 13H20 12.575 1456 13H40 12.606 1471 4 51H26* 12.608 1411 51H41 12.817 1362 64H19 12.728 1470

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 8 of 9 TABLE 4.5-4 (continued) HOOP TENDON DATA Surveillance since Measured No.Suvilac Initial Force/Tendon Tensioning 35H01 18.942 1572 42H01 18.928 1560 46H21" 18.833 1425 5 46H28 18.728 1375 46H30 18.781 1382 46H47 18.817 1468 62H08 18.814 1435 42H18 22.692 1476 42H29 22.786 1448 42H30 22.733 1389 42H31 22.775 1338 42H32 22.703 1355.5 42H33 22.772 1361 42H34 22.733 1377.5 42H35 22.781 1296.5 42H36 22.744 1408 6 42H37 22.775 1401.5 42H44 22.711 1471.5 51H25 22.822 1363 51H26* 22.628 1320 51 H27 22.836 1265.5 51H28 22.647 1450.5 53H02 22.797 1611 53H46 22.667 1459.5 62H41 22.797 1426 62H46 22.736 1485 46H21* 26.656 1388 46H30 26.694 1355.6 46H31 26.667 1343.3 46H32 26.628 1366.8 7 46H33 26.664 1357.8 46H34 26.619 1424.7 46H35 26.653 1376.8 46H36 26.608 1343.5 46H37 26.644 1293.5

U. S. Nuclear Regulatory Commission Enclosure 2 3F1009-08 Page 9 of 9 TABLE 4.5-4 (continued) HOOP TENDON DATA Surveillance Years since Measured No. Initial Force/Tendon Tensioning_

46H38 26.625 1353.4 46H39 26.647 1356.2 53H16 26.628 1475.4 63H02 26.672 1551.6 63H09 26.814 1431.8 13H33 32.817 1306.46 13H34 32.636 1368.61 13H35 32.814 1244.25 13H36 32.622 1385.23 13H37 32.825 1289.87 13H38 32.639 1395.05 42H46 32.692 1558.63 46H19 32.711 1358.61 46H20 32.619 1298.13 46H21* 32.694 1330 46H22 32.622 1311.48 46H23 32.708 1329.97 46H24 32.636 1425.85 51 H34 32.644 1464.70 62H29 32.736 1369.63 62H30 32.681 1290.84 62H33 32.733 1313.39 62H34 32.686 1378.71 Extrapolated 1 63 (Note 1) 1328

• Control Tendon - Tendon 51 H26 was used as the control tendon when testing was performed during outages for surveillances 3, 4 and 6. Tendon 46H21 was used during online testing for surveillances 5, 7 and 8.

Notes:

1. The extended period of operation will end in the 63rd year from the date of initial tensioning.