ML041450502

Response to Request for Additional Information Re Proposed Amendment to Increase the Setpoint of the Main Steam Safety Valves

ML041450502
Person / Time
Site:	Calvert Cliffs
Issue date:	05/19/2004
From:	Vanderheyden G Constellation Energy Group, Constellation Generation Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC1578, TAC MC1579
Download: ML041450502 (6)

Text

George Vanderheyden Vice President Calvert Cliffs Nuclear Power Plant Constellation Generation Group, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.4455 410.495.3500 Fax I3 Constellation Energy May 19,2004 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. I & 2; Docket Nos. 50-317 & 50-318 Request for Additional Information:

Proposed Amendment to Increase the Setpoint of the Main Steam Safety Valves (TAC Nos. MC 1578 and MCI 579)

REFERENCES:

(a)

Letter from Mr. G. Vanderheyden (CCNPP) to Document Control Desk (NRC), dated December 9, 2003, "Increase of the Lift Setpoint of the First Bank of Main Steam Safety Valves and Increase in the Completion Time to Reset the Power Level-High Trip Setpoint" (b)

Letter from Mr. G. S. Vissing (NRC) to Mr. G. Vanderheyden (CCNPP),

dated March 26, 2004, "Request for Additional Information Re: Proposed Amendment to Increase the Setpoint of the Main Steam Safety Valves (TAC Nos. MC 1578 and MC 1579)

Reference (a) proposed to increase the lift setpoint of the first bank of main steam safety valves and increase the completion time to reset the power level-high trip setpoint. This letter is in response to the questions posed in Reference (b).

The responses in Attachment (1) do not change the significant hazards discussion in Reference (a).

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Document Control Desk May 19, 2004 Page 2 Should you have questions regarding this matter, we will be pleased to discuss them with you.

STATE OF MARYLAND COUNTY OF CALVERT

TO WIT:

I, George Vanderheyden, being duly sworn, state that I am Vice President - Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to b reliable.

Subscribed and sworn before Maea Notary Pub]c in and for the State of Maryland and County of

/-

/this

/

h' day of 2004.

WiTt4ESS mry I-land and Notarial Seal:

My Commission Expires:

aNotery Public t

Date GV/EMT/bjd

Attachment:

(1)

Request for Additional Information for the Review of Calvert Cliffs Technical Specification Changes to Increase Main Steam Safety Valve Setpoints cc:

J. Petro, Esquire J. E. Silberg, Esquire Director, Project Directorate 1-1, NRC G. S. Vissing, NRC H. J. Miller, NRC Resident Inspector, NRC R. 1. McLean, DNR

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF CALVERT CLIFFS TECHNICAL SPECIFICATION CHANGES TO INCREASE MAIN STEAM SAFETY VALVE SETPOINTS Calvert Cliffs Nuclear Power Plant, Inc.

May 19, 2004

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF CALVERT CLIFFS TECHNICAL SPECIFICATION CHANGES TO INCREASE MAIN STEAM SAFETY VALVE SETPOINTS QUESTION I In support of the Technical Specification (TS) changes, the licensee reanalyzed several events to determine the effects of the revised main steam safety valve (MSS19 setpoints on the analysis of record Please provide the following information for the staff to review:

Question (1)

Discuss the methods and computer codes used in the reanalysis and address their acceptability for the licensing application.

CCNPP Response For non loss-of-coolant accident (LOCA) transients, CESEC-Il1 (approved for use in Combustion Engineering plants by Reference 1) was used as the general plant transient code to calculate the time dependent core power, flow, coolant temperatures, and pressure.

The application methodologies used in the non-LOCA analyses are consistent with those previously used in the most recent Calvert Cliffs analyses reviewed and approved by the Nuclear Regulatory Commission.

The small break LOCA Emergency Core Cooling System performance analysis was performed with the CENPD-137 Evaluation Model (S2M), Supplement 2, version of the Westinghouse small break LOCA evaluation model for Combustion Engineering designed pressurized water reactors (Reference 2).

Question (2)

Identify for the reanalyzed events the values of input parameters that are different from those assumed in the analysis of record, andjustify the adequacy of the values.

CCNPP Response There are no differences in the input parameters for the reanalyzed events in this license amendment request from those in the analysis of record [Updated Final Safety Analysis Report (UFSAR)

Chapter 14].

Question (3)

Provide a table listing for each reanalyzed case, the calculated peak pressurizer pressure, peak steam generator pressure, and peak pressurizer water level, and demonstrate that the applicable acceptable criteria are met.

1

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF CALVERT CLIFFS TECHNICAL SPECIFICATION CHANGES TO INCREASE MAIN STEAM SAFETY VALVE SETPOINTS CCNPP Response Peak Reactor Coolant System (RCS) pressure, steam generator pressure, and pressurizer level were tabulated for the limiting UFSAR Chapter 14 peak pressure events - Loss of Load and Feedwater Line Break. Other reanalyzed events are non-limiting with respect to peak pressure.

Summary of Limiting Event Parameters Loss of Load Feed Line Break Peak RCS Pressure 2686 psia (1,2) 2749.9 psia(6. 2)

Peak Steam Generator Pressure 1106 psia (3, 4) 1057.7 psia(6 4)

Peak Pressurizer Level

< 1100 ft3 (5)

< 1200 ft3 ()

()

UFSAR Table 14.5-2, "Sequence of Events for Loss of Load Event to Maximize Calculated RCS Peak Pressure" (2)

RCS Pressure Includes Elevation Head (3)

UFSAR Table 14.5-4, "Sequence of Events for Loss of Load Event to Maximize Calculated Secondary Peak Pressure" (4)

Steam Generator Pressure Includes Downcomer Liquid Head (5) UFSAR Figure 14.5-6, "Loss of Load Event - Pressurizer Water Volume vs Time" (6)

UFSAR Table 14.26-3, "Sequence of Events for Feedwater Line Break with LOAC Following Reactor Trip"

()

UFSAR Figure 14.26-11, "Feedline Break Event with LOAC Following Reactor Trip - Pressurizer Water Volume vs Time" As can be seen from the above table, the peak RCS pressure remains below the bounding value of 2750 psia, the peak steam generator pressure remains below the bounding value of 1115 psia, and the pressurizer does not fill. Therefore, all acceptance criteria have been satisfied.

QUESTION 2 The purpose of the high power level trip listed in TS Table 3.7.1-1 is to provide overpressure protection with various numbers of operable MSSVs. Justify that the high power trip setpoints remain valid in light of the proposed changes to the MSSVsetpoints.

CCNPP Response As discussed in the section in Reference 3 describing the analysis of loss of load event (limiting case), the loss of load analysis does not credit the Power Level-High trip. Analysis of operation with inoperable MSSVs demonstrated that peak primary and secondary pressure criteria are met.

We have reviewed Technical Specification Table 3.7.1-1 and have confirmed that the values in the table remain valid for this requested change.

2

ATTACHMENT (1)

REQUEST FOR ADDITIONAL INFORMATION FOR TILE REVIEW OF CALVERT CLIFFS TECHNICAL SPECIFICATION CHANGES TO INCREASE MAIN STEAM SAFETY VALVE SETPOINTS QUESTION 3 Verify the edition of the Code that is being referenced for the Section LII repairs and non-destructive examination.

CCNPP Response This proposed license amendment does not include repairs or non-destructive examinations. The mention of the American Society of Mechanical Engineers Code in Reference 3 is a discussion of the design capacity of the main steam safety valves. This capacity is not being changed.

REFERENCES

1.

Letter from C. 0. Thomas (NRC) to A. E. Scherer (CE), dated April 3, 1984, "Combustion Engineering Thermal-Hydraulic Computer Program CESEC III"

2.

CENPD-137, Supplement 2-P-A, April 1998, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model"

3.

Letter from G. Vanderheyden (CCNPP) to Document Control Desk (NRC), dated December 9, 2003, "Increase of the Lift Setpoint of the First Bank of Main Steam Safety Valves and Increase in the Completion Time to Reset the Power Level-High Trip Setpoint" 3