ML080560007

Request for Additional Information for License Amendment Request to Modify Plant Design and Licensing Basis and Change to USAR, Re Uprate of Shutdown Cooling System Entry Conditions

ML080560007
Person / Time
Site:	Fort Calhoun
Issue date:	02/27/2008
From:	Markley M NRC/NRR/ADRO/DORL/LPLIV
To:	Bannister D Omaha Public Power District
Markley, M T, NRR/DORL/LP4, 301-415-5723
References
TAC MD6993
Download: ML080560007 (10)

Text

February 27, 2008 Mr. David J. Bannister Vice President and CNO Omaha Public Power District Fort Calhoun Station FC-2-4 Post Office Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - REQUEST FOR ADDITIONAL INFORMATION RE: LICENSE AMENDMENT REQUEST, UPRATE OF SHUTDOWN COOLING SYSTEM ENTRY CONDITIONS (TAC NO. MD6993)

Dear Mr. Bannister:

By letter dated October 12, 2007 (Agencywide Documents Access and Management System Accession No. ML072890192), the Omaha Public Power District (OPPD) submitted a proposed license amendment request to modify the plant design and licensing basis to increase the shutdown cooling (SDC) system entry temperature from 300 degrees Fahrenheit (°F) to 350 °F (cold leg), and the SDC entry pressure from 250 pounds per square inch absolute (psia) to 300 psia (indicated at the pressurizer). OPPD also requested changes to the Updated Safety Analysis Report described design methodology applied to the SDC heat exchangers.

The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided and determined that additional information is required in order to complete the evaluation. The enclosed request for additional information was discussed with Mr. Thomas Matthews and others of your staff on January 23 and 31, 2008. As agreed upon, the NRC staff is expecting a response within 30 days of the date of this letter.

If you have any questions, please contact me at 301-415-5723.

Sincerely,

/RA/

Michael T. Markley, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

Request for Additional Information cc w/encl: See next page

(s).: ML080560007 OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/LPL4/BC NAME MMarkley JBurkhardt THiltz DATE 2/27/08 2/27/08 2/27/08

Ft. Calhoun Station, Unit 1 (11/26/2007) cc:

Winston & Strawn ATTN: James R. Curtiss, Esq.

1700 K Street, N.W.

Washington, DC 20006-3817 Chairman Washington County Board of Supervisors P.O. Box 466 Blair, NE 68008 Mr. John Hanna, Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 310 Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. Joe L. McManis Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550 Ms. Melanie Rasmussen Radiation Control Program Officer Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319

REQUEST FOR ADDITIONAL INFORMATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285 By letter dated October 12, 2007 (Agencywide Documents Access and Management System Accession No. ML072890192), the Omaha Public Power District (OPPD) submitted a proposed license amendment request (LAR) to modify the plant design and licensing basis to increase the shutdown cooling (SDC) system entry temperature from 300 degrees Fahrenheit (°F) to 350 °F (cold leg), and the SDC entry pressure from 250 pounds per square inch absolute (psia) to 300 psia (indicated at the pressurizer). OPPD also requested changes to the Updated Safety Analysis Report (USAR) described design methodology applied to the SDC heat exchangers (HX).

The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided and determined that additional information is required in order to complete the evaluation. The following request for additional information was discussed with Mr. Thomas Matthews and others of your staff on January 23 and 31, 2008.

A.

SDC Pumps [Low Pressure Safety Injection (LPSI) Pumps] Design Section 3.2.2 of the amendment request states that [t]he SDC pumps, SI-1A and SI-1B, also referred to as the LPSI pumps, have been analyzed for the new design pressure and temperature of 550 psig [pounds per square inch gauge] and 350 °F from a Code-stress standpoint. In order to meet Code allowable stress limits, the existing pump hold-down bolting must be replaced with bolting composed of a stronger material.

1.

Describe the pump analysis performed and provide a summary of the analysis results which determined that the existing pump hold-down bolting needs to be modified to meet Code-allowable stress limits. Provide current, revised, and allowable values.

2.

Provide the Code of reference for evaluating the SDC pump anchorage. If different than the design basis Code of record, provide justification.

3.

Provide the schedule of completion for the hold-down bolting modification of the SDC pumps.

B.

SDC Piping Re-Analysis

1.

Provide a quantitative summary of the SDC piping analysis results at the proposed increased design conditions (temperature and pressure) that shows conformance with the criteria of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section III, and any USAR commitments as applicable. The evaluation should include maximum calculated stresses, fatigue usage factors, and Code-allowable values. At critical locations, ENCLOSURE

such as nozzles and penetrations, show that the allowable loads and movements have been satisfied.

2.

Confirm whether a review of postulated pipe-break criteria has been performed and provide justification that existing locations still meet the pipe-break criteria for the increased design conditions. In addition, verify whether new postulated pipe-break locations were identified and provide justification.

3.

Identify any pipe support modifications required due to the increased design loads.

C.

Review of TS Changes to Increase the SDC Entry Temperature and Pressure and the Associated Low Temperature Overpressure Protection (LTOP) Analysis

1.

Sections 3.2.2 and 3.3 indicated that the heat-transfer capacity of the SDC HX is adequate for the proposed range of the SDC temperature and pressure conditions, because a calculation verified that when a component cooling water (CCW) inlet temperature to the SDC HX is less than 110 °F, the SDC/CCW system has the capability to cool down from the new initiation reactor coolant temperature of 350 °F to 130 °F at nominal full-power of 1500 megaWatts thermal (MWt) and normal service fouling level in the original design basis time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based on the results, Section 3.3 indicated that for a loss-of-coolant accident (LOCA) during plant shutdown, the period during which automatic initiation of the emergency core coolant system is not available during the shutdown is bounded by the analysis of record (AOR).

a.

Please discuss the computer codes and/or methods used in the SDC HX capacity calculation, and reference the associated NRC safety evaluation (SE) that approved the codes and/or methods. Address if there have been any changes to the NRC-approved codes and/or methods used in the SDC HX capacity calculation, and justify that the changes are acceptable. Also, discuss the operating procedures that are used to control the CCW inlet temperature to the SDC HX within 110 °F.

2.

Sections 3.4 and 4.1.3 indicated that the RELAP5 Mod 3.2d model was used to reanalyze mass addition and heat addition cases in support of the existing LTOP setpoint curve.

a.

Please list the NRC SE that approved the use of the RELAP5 Mod 3.2d model for the LTOP reanalysis and show how the restrictions or conditions in the NRC SE approving the use of the model have been met.

Address if there have been any model changes including the nodal scheme in the NRC-approved code used in the reanalysis, and justify the changes. If the RELAP5 Mod 3.2d model was not previously approved by the NRC, provide a discussion of the model with the code verification applicable to the SDC conditions for the NRC staff to review and approve.

3.

Section 3.7 indicated that the boron dilution event was reanalyzed based on the revised SDC conditions. The results of the reanalysis showed that the available operator time to terminate the event was reduced by 0.55 minutes as compared to the results in the AOR.

a.

Please identify any model and assumptions used in the reanalysis that are different from those used in the AOR, and justify the differences.

Discuss the assumptions and the associated effects used in the analysis that result in a reduction of the operator time by 0.55 minutes.

D.

Review of the LTOP Analysis (OPPD Calculation FC07187, Revision 0)

1.

The operational restrictions assumed in the LTOP reanalysis are provided in Table 2.

a.

Please list the corresponding TS sections that include the operational restrictions specified in Table 2 relating to reactor coolant pumps (RCPs),

high-pressure safety injection pumps (HPSIs), SDC, pressurizer steam void, and reactor coolant system (RCS) pressure. If the operational procedures were used to implement the operational restrictions, discuss the operator actions in the procedures and address the compliance with the requirements of paragraph 50.36.c(2)(ii) of Title 10 of the Code of Federal Regulations (10 CFR) that specify criteria for each item to be included in the Technical Specifications (TS).

Also, the last item (RCS pressure) in Table 2 requires that when starting the first RCP, the RCS pressure be at least 61 pounds per square inch (psi) below the LTOP setpoint pressure at a given RCS temperature, in order to prevent a power-operated relief valve (PORV) lift.

b.

Please discuss how the value of 61 psi was determined.

2.

Page 8 indicated that the LTOP setpoints were established to limit pressure transients to below the pressure-temperature (P/T) limits as shown in Figure 1 and Table 6.

a.

Please list the NRC SE that approved the P/T limits. If the limits were not previously reviewed and approved by the NRC, please provide a derivation of the limits, and justify the acceptance of the P/T limits for the licensing application.

3.

Page 9 indicated the pressure correction factor (PCF) ranging from 61 psi below 210 °F to 67 psi above 210 °F was used to account for the elevation and flow effects on the P/T limits that are based on the pressurizer pressure.

a.

Please provide a derivation of the PCF values of 61 psi and 67 psi based on the pressure loss and elevation difference between the reactor vessel beltline and pressurizer, and show that the PCF values are conservative

and applicable to the replacement steam generators and pressurizer in determination of the P/T limits.

4.

Table 7 listed the actual and analytical values for the LTOP setpoints. For example, at 220 °F, the analytical LTOP pressure is 690 psia.

a.

Please provide a derivation of the actual setpoint of 587.75 psia and the analytical value of 690 psia, and show that the PORV actuation system uncertainties of 16.3 °F and 66.9 psi are adequately considered in determining the LTOP actual and analytical setpoints.

Also, page 10 (last paragraph) indicated that the pressure uncertainty used for the PORV actuation pressure adjustment is 66.7 psi, which is different from the value of 66.9 psi specified in Tables 3 (page 14) and 7 (page 17).

b.

Please clarify the inconsistencies.

5.

Page 22 (last paragraph) indicated that the decay heat used in the LTOP reanalysis is 20 percent greater than the 1.4 percent value shown in the CESEC code based on a cooldown time of 2.18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

a.

Please discuss the decay heat model (decay heat level versus time) used in the CESEC code and justify the acceptance of the model used for the LTOP reanalysis.

6.

Table 10 (page 28) listed the cases used to support the LTOP setpoints.

Cases 1 through 8 are not changed from Reference 1, Low Temperature Overpressure Protection (LTOP) Analysis in Support of Steam Generator Replacement, OPP006-REPT-001. Page 29 indicated that Reference 1 provided a detailed description of each case.

a.

Discuss the analysis, test data, and/or procedures that are used to assure that the PORVs will close in the steam, two-phase or liquid conditions applicable to the assumptions used in the LTOP analysis for the mass-addition events. Closure of the PORVs will avoid occurrence of a small break LOCA resulting from a stuck-open PORV.

b.

Reference the NRC SE that approved the use of the void in the pressurizer for consequence mitigation of the heat-addition events considered in the LTOP analysis.

c.

Please list the NRC SE that approved Reference 1. If the reference was not previously approved by the NRC, provide the reference for the NRC to review and approve.

7.

Section 3.2.1 of Enclosure to an October 12, 2007, letter indicated that the SDC suction-to-RCS valves (HCV-347 and HCV-348) interlock setpoint was changed

from 250 psia to 300 psia at pressurizer. We found that USAR-9.3 (page 6) discussed the valve interlock functions. Specifically, it stated that

... if the breaker is closed and the operator attempts to open either of these valves when pressure in the RCS is above 250 psia [300 psia for the new SDC entry pressure], an inhibit will prevent opening the valve, an alarm will sound and both valves will shut automatically...

It is not clear from the above statement whether this SDC interlock feature will automatically shut the valves (HCV-347 and HCV-348) or not if the RCS pressure increases above the interlock setpoint (300 psia) when a mass or heat-addition event occurs during SDC operating conditions.

a.

If the valves are not automatically closed when the RCS increases above the interlock setpoint, discuss the design features and procedures used to prevent the SDC from over-pressurization for a mass or heat-addition event.

8.

Section 1.0 of Enclosure to an October 12, 2007, letter indicated that the SDC entry pressure would increase from 250 psia to 300 psia (indicated at the pressurizer). In support of the SDC entry pressure change, the related TS and the SDC design pressure were changed.

Section 2.4 discussed the proposed TS 3.16(1)a, which stated that the portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 300 psig...

Table 1 of Section 3.2.1 indicated that the new design pressure for the SDC pumps suction piping is 350 psig.

Since the SDC entry pressure of 300 psia (indicated at the pressurizer) is based on the pressure measurement at a high elevation of the pressurizer, the corresponding pressure at the SDC system would be the pressurizer pressure plus the gravitational head and pressure loss between the pressurizer and the SDC system, and the pressure measurement uncertainty of +50 psi indicated (Table 3 of Attachment 3 to an October 12, 2007, letter). Therefore, it is likely that the maximum SDC operating would be greater than the SDC entry pressure of 300 psia.

a.

Please justify that the proposed leakage testing pressure of 300 psig in TS 3.16(1)a and design pressure of 350 psig for the SDC pumps suction piping are high enough and adequate to support the proposed SDC entry pressure of 300 psia with consideration of the pressure measurement uncertainty, the pressure difference due to the gravitational head and pressure loss between the pressurizer where the pressurizer pressure is

measured and the SDC system where the leakage tests are performed, and the SDC piping design pressure is established.

9.

The proposed TS 2.1.1(11)(b) specified that the LTOP cannot be placed on SDC until the RCS has cooled to an indicated RCS temperature of less than or equal to 350 °F. Table 3 of Attachment stated that the RCS temperature measurement uncertainty is +14 °F. With inclusion of the temperature measurement uncertainty, the LTOP may not be put in service until the actual RCS temperature is equal to 364 °F.

a.

Please justify that the proposed design temperature of 350 °F (Table 1 of Section 3.2.1) is adequate to support the LTOP operation at 364 °F.

E.

In the LAR, the licensee states in order to ensure adequate pump seal and bearing cooling, the CCW inlet temperature at the seal cooler must not exceed the design value of 100 °F when the RCS temperature is between 300 °F and 350 °F. The evaluation refers to the CCW inlet temperature to the SDC HX of 110 °F would result in an approximately 26 percent increase temperature difference and an increased decay heat load. The evaluation refers to a calculation that verifies the SDC/CCW system has the capability to cool down the RCS from 350 °F to 130 °F at full-nominal power of 1500 MWt in the original design basis time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Assuming the CCW supplied to the pump seal and bearing cooling is the same CCW being supplied to the SDC HX, then CCW cannot be supplied to the SDC at 110 °F when the RCS is between 300 °F and 350 °F (i.e., limited to 100 °F). The application refers to procedure controls to ensure the limit is not exceeded.

1.

Considering the design basis maximum for river water temperature is 90 °F (with a limit of the outlet of the CCW HX to less than 100 °F, not 110 °F) and future power uprate to 1765 MWt, what would be the effect on the SDC HXs ability at these limits and constraints to remove the required design heat load at 350 °F and achieve a cooldown of the RCS in the time required?

2.

The current design temperature limit for the SDC pump discharge piping and SDC HX is 350 °F. The design temperature limit for the SDC pump suction pumping and the SDC pump seals is only 300 °F. The licensee proposes to increase the system operating temperature to 350 °F, which would equal the design temperature for SDC LPSI pumps, suction piping, discharge piping, and the SDC HX, and would exceed the design temperature for the SDC pump seals.

This increase would result in no safety margin between the design and operating parameters. Additionally, in the LTOP analysis, the licensee states there is a 14 °F uncertainty in RCS temperature, which could result in the operating SDC with the RCS higher than operating/design limits. General design criteria under 10 CFR 50, Appendix A, requires the RCS and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the RCP boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

a.

The licensee is requested to explain how operations with no margin and an expected uncertainty that would exceed design limits would be acceptable.

3.

The current design temperature limit for the LPSI pumps seals is 300 °F (reference USAR, page 13 of 31 in Section 6.2). The LAR implies that the LPSI pumps are currently rated for 350 °F (Table 1 on page 5 and discussion for LPSI pumps on page 7).

a.

The licensee is requested to explain the difference in ratings when the seals are usually encompassed with the pump ratings. Additionally, the licensee is requested to provide an evaluation of whether the pump seals can safely operate at the new proposed operating limits with an adequate safety margin, considering the higher temperature and pressure, including uncertainties.