ULNRC-05178, Technical Specification Revisions Associated with the Steam Generator Replacement Project

Technical Specification Revisions Associated with the Steam Generator Replacement Project

ML052220138
Person / Time
Site:	Callaway
Issue date:	07/29/2005
From:	Herrmann T AmerenUE
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05178
Download: ML052220138 (8)

Text

AmerenUE PO Box 620 Callaway Plant Fulton, MIO65251 July 29, 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555 ULNRC-05178 Ladies and Gentlemen:

DOCKET NUMBER 50-483 UNION ELECTRIC COMPANY CALLAWAY PLANT WAmeren TECHNICAL SPECIFICATION REVISIONS ASSOCIATED WITH THE STEAM GENERATOR REPLACEMENT PROJECT UE

References:

1. ULNRC-05056 dated September 17, 2004

2. ULNRC-05117 dated February 11, 2005

3. ULNRC-05145 dated May 26, 2005

4. ULNRC-05157 dated June 17, 2005

5. ULNRC-05159 dated June 17, 2005

6. ULNRC-05169 dated July 15, 2005

7. ULNRC-04158 dated December 3, 1999 In Reference 1 above AmerenUE transmitted an application for amendment to Facility Operating License Number NPF-30 for the Callaway Plant in support of the replacement steam generators to be installed during Refuel 14 (fall 2005). This letter provides additional information on three issues recently raised by the NRC staff during the review of that application.

In Reference 5 above AmerenUE responded to several requests for additional information (RAIs) from the NRC Reactor Systems Branch. The following discussion provides additional clarification and a new commitment regarding the responses to RAI Questions 11, 12, and 45.

RAI #11 In the response to Question 11 in Reference 5 it is stated that "the calculated values bound the maximum allowable PORV setpoints specified in the current Pressure Temperature Limits Report (PTLR); therefore, the PTLR is not being revised for the RSG project." The NRC staff would like to know the calculated values and maximum allowable PORV setpoints.

-(V(5 a subsidiary ofAmeren Corporation

ULNRC-05 178 July 29, 2005 Page 2 The statements made in Section 4.3.5 of Appendix A to Reference 1 and in the response to Question 11 in Reference 5 were predicated on an assumption by Westinghouse that AmerenUE would limit the operation of reactor coolant pumps (RCPs) below a reactor coolant system (RCS) temperature of 200'F such that a maximum of two (2) RCPs would be allowed to operate whenever the RCS temperature was

< 2000 F. AmerenUE has decided to implement Cold Overpressure Mitigation System (COMS) setpoints that have no RCP operation restrictions. These COMS setpoints are more limiting than their counterparts that have RCP operation restrictions, but will not present any undue burdens on plant operation. Since these COMS setpoints are more limiting than those currently presented in the PTLR, the PTLR will be revised and the COMS setpoints will be changed to implement the RSG project. The attachment to this letter shows the draft PTLR changes, which provide the new calculated values for RSG as well as the current maximum allowable PORV setpoints. The current COMS setpoints have not been revised since the initial PTLR submittal to NRC in Reference 7.

RAI #12 The NRC staff would like to know what computer methodology was used to perform the sensitivity discussed in the last paragraph of the response to Question 12 in Reference 5.

At the request of the NRC reviewer during the meeting held on May 18, 2005, Westinghouse performed an additional calculation based on the Loss of Load/Turbine Trip analysis in Section 6.3.4 of Appendix A to Reference 1 (which is the FSAR Chapter 15 deterministic-style analysis) crediting the second safety grade signal to demonstrate the adequacy of the installed primary safety valve (PSV) capacity in preventing over-pressurization of the RCS. The methodology, including the computer code used -

RETRAN, is identical to that of Section 6.3.4 of Appendix A. The difference in these analyses involves the high pressurizer pressure reactor trip function, credited as the first safety grade signal in the FSAR Chapter 15 deterministic-style analysis; that trip was conservatively ignored in the additional sensitivity discussed in the last paragraph of the response to Question 12 in Reference 5.

RAI #45 Regarding question 45, a telecon was held on July 26, 2005 between Westinghouse and the NRC to discuss NRC's concern about the impact of non-integer size breaks on the limiting small break loss of coolant (SBLOCA) peak clad temperature (PCT) analyses. Callaway was a participant on that telephone call. The NRC staff recently identified a potential issue related to the impact of non-integer size breaks on the limiting SBLOCA PCT analysis for another plant which is considering a power up-rate.

NRC approached Westinghouse to determine the extent of this issue on the analyses they have performed for other licensees. With respect to Callaway's amendment request for SG replacement (Reference 1 above), NRC has theorized that a break between the 3-inch analyzed small break case (which did not result in accumulator injection) and the 4-inch analyzed small break case (which did result in accumulator injection to stem the cladding

ULNRC-05178 July 29, 2005 Page 3 temperature rise) would result in a higher PCT value than reported for those analyzed cases. AmerenUE and Westinghouse believe this to be a generic industry issue.

New large break and small break LOCA analyses were performed in support of the Callaway steam generator replacement project. Callaway's reanalyses were performed by Westinghouse using NRC-approved methodologies. Large break LOCA analysis results demonstrate a limiting PCT of 1939'F (per the responses to questions 16 and 17 in Reference 5) and SBLOCA analysis results demonstrate a limiting PCT of 1043IF (for the 4-inch case in Table 6.2.2-5 of Appendix A to Reference 1). These results demonstrate that Callaway is large break limited with respect to PCT and has greater than 11 50'F margin to the 2200'F regulatory limit for SBLOCA. AmerenUE and Westinghouse have a high degree of confidence that this issue would not result in the PCT for SBLOCA becoming limiting or encroaching upon 2200'F.

The following commitment is submitted with the intent that its formal transmittal would allow NRC to proceed with completing the Safety Evaluation and issue the license amendment for the Callaway steam generator replacement project:

"In support of AmerenUE's license amendment request (Reference 1 above),

associated with the steam generator replacement project, AmerenUE has initiated a corrective action program item (CAR 200406948 Action 27) that will track this issue. Callaway staff will follow this generic industry issue and will assess non-integer intermediate breaks consistent with the generic resolution. AmerenUE will provide a status update and discuss the method of how this issue will be assessed in the next required 10 CFR 50.46 report."

References 2, 3, 4, and 6 above provided various RAI responses, supplemental Technical Specification changes, and additional RAI response clarification in support of this amendment request. Nothing in the information provided above invalidates the findings of the licensing evaluations contained in Attachment 1 of Reference 1. The requested approval date and implementation plans for this amendment application remain unchanged from Reference 1. If you have any further questions on this amendment application, please contact us.

Very truly yours, Timothy E. Herrmann Manager-Nuclear Engineering Services GGY/

Attachment - Draft PTLR Changes

ULNRC-05178 July 29, 2005 Page 4 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy)

Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 7E1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street PO Box 360 Jefferson City, MO 65102-0360 Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102

STATE OF MISSOURI )

) SSs COUNTY OF CALLAWAY)

Timothy E. Herrmann, of lawful age, being first duly sworn upon oath says that he is Manager, Nuclear Engineering Services, for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

Timothy E. Herrmann Manager, Nuclear Engineering Services SUBSCRIBED and sworn to before me this Pith day of ,

9 2005.

I PAULkAA. JOHNSON Notary Public - Notary Seal STATE OF MISSOURI Callaway County My Commission Expires: July 31, 2007

51-5065107-00 Page 3 of 9 Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT CALLAWAY COMS MadmumAllowable PORV S61pclfls a 1ao 200 300 4 RCS Temperature (degree F)

. - High PORV $e~pnt - Low PO1W SetpoltI FIGURE2.2-1 Maxim AlowedPORVSetpointfor theColdOverpressure Mitigao System Callaway Plant 7 Revision3

51-5065107-00 Page 4 of 9 FIgure 14.9 PRESSURE AND TEMPERATURE LIMiTS REPORT TABLE 2.2-1 CALLAWAY PLANT COMS MA)XMUM ALLOWABLE PORV SETPOINTS AT 20 EFPY Maximum Allowable Function Generator Setpoo ts (Breakpoontsl Breakpoint Number mperature - High Setpoint Low Se m\

2 l _ 76 501 _ _ 471 l 3 _ _130- _ 525 495_

4 _ 170 -. M 505 5 _ 220 _ 520

_550 6 _ 245 __' 650 600 7 _ 270,,-_ 750 700 a __2f5_ 750 700 9 __420 2350 _ 2350D NOTE: Seipoints assurne fha( recor p\\ rrunigorTc0F and \ha t rprtor250 f Tmperatur - F 70 Setpoint High . Low 97Setpo nt /

RCS (°F) (psig) (psig) 70 496 466 80 496 466 90 496 466 ISO 527 497 230 758 512 250 750 597 270 750 597 280 750 597 350 2335 2185 Caliaway Plant Revision 3

I .

51-5065107-00 Page 6 of 9 CALLAWAY COMS Maximum Allowable PORV Setpolnts 2500 T. j i i7T7i f I 2000

&n cL

_0. 1500 +*+~-~+/-f&* I: 1!1, A.11l iI a

u) 1000 0

CL 500 . -1 i  ;'

0 0 50 10O 15D 200 250 300 350 400 RCS Temperature (degree F)

-'High PORV Setpolnt - - - Low PORV Setpoint INSERT 2 FIGURE 14.9