ML081420824

Stretch Power Uprate License Amendment Request Additional Information in Connection with the NRC Audit Held on May 13, 2008 in Rockville, Maryland

ML081420824
Person / Time
Site:	Millstone
Issue date:	05/21/2008
From:	Gerald Bichof Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-04501
Download: ML081420824 (11)

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Dominion Nuclear Connecticut, Inc.

',')00 Dominion Boulevard, Glen Allen, Virginia 2'060 - Dominion

\\ 'ebi\ddress: www.dom.com May 21,2008 U. S. Nuclear Regulatory Commission Serial No.: 07-04501 Attention: Document Control Desk NLOS/MAE: R1 One White Flint North Docket No.: 50-423 11555 Rockville Pike License No.: NPF-49 Rockville, MD 20852-2378 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST ADDITIONAL INFORMATION IN CONNECTION WITH THE NRC AUDIT HELD ON MAY 13, 2008 IN ROCKVILLE, MARYLAND Dominion Nuclear Connecticut, Inc. (DNC) submitted a stretch power uprate license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3) in letters dated July 13, 2007 (Serial Nos. 07-0450 and 07-0450A), and supplemented the submittal by letters dated September 12, 2007 (Serial No.

07-04508), December 13,2007 (Serial No. 07-0450C), March 5, 2008 (SE~rial No.

07-04500), March 27, 2008 (Serial No. 07-0450E) and April 24,2008 (Serial No.

07-0450F). The NRC staff forwarded requests for additional information (RAls) in October 29,2007, November 26,2007, December 14,2007, December 20,2007 and April 23, 2008 letters. DNC responded to the RAls in letters dated November 19, 2007 (Serial No. 07-0751), December 17, 2007 (Serial No.

07-0799), January 10, 2008 (Serial Nos. 07-0834, 07-0834A, 07-0834C, and 07-0834F), January 11, 2008 (Serial Nos. 07-08348, 07-0834E, 07-0834G, and 07-0834H), January 14, 2008 (Serial No. 07-08340), January 18, 2008 (Serial Nos. 07-0846, 07-0846A, 07-08468, 07-0846C, and 07-08460), January 31, 2008 (Serial No. 07-08341), FElbruary 25, 2008 (Serial Nos. 07-0799A and 07-0834J), March 10, 2008 (Serial Nos. 07-0846E and 07-0846F), March 25, 2008 (Serial No. 07-0834K), April 4, 2008 (Serial No. 07-0834L), April 29, 2008 (Serial No. 08-0248) and May 15, 2008 (Serial No. 08-0248A).

Please find attached Westinghouse Electric Company's letter NEU-08-31, "Millstone Unit 3 (NEU) Stretch Power Uprate (SPU) Program Rod Withdrawal at Power (RWAP) Peak Reactor Coolant System Pressure", dated May H~, 2008.

This letter is provided to the NRC in connection with the NRC's recent audit held on May 13, 2008 in Rockville, Maryland.

The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the December 13, 2007 DNC letter (Serial No. 07-0450C).

Serial No. 07-04501 Docket No. 50-423 Supporting Information Page 2 of 3 Should you have any questions in regard to this submittal, please contact Ms.

Margaret Earle at 804-273-2768.

Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, ancl that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this ~ I sr day of m~ ,2008.

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Serial No. 07-04501 Docket No. 50-423 Supporting Information Page 3 of 3 Commitments made in this letter: None Attachment cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. G. Lamb Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-881A Rockville, MD 20852-2738 Mr. J. D. Hughey Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-883 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No. 07-04501 Docket No. 50-423 ATTACHMENT STRETCH POWER UPRATE LICENSE AMENDMENT REQUESl:

WESTINGHOUSE ELECTRIC COMPANY'S LETTER NEU-08-31, MILLSTONE UNIT 3 (NEU) STRETCH POWER UPRATE (SPU) PROGRAM ROD WITHDRAWAL AT POWER (RWAP) PEAK REACTOR COOLANT SYSTEM PRESSURE MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

• Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box355 Pittsburgh, Pennsylvania 15230-0355 USA Mr. Ron Thomas Directtel: 412-374-6345 Dominion Nuclear Connecticut Direct fax: 412-374-3257 Rope Ferry Road, Route 156 e-mail: rogos1d1@westinghouse.com Waterford, CT 06385 Customer P.O.: 70155283 YY Sales Order: 38944 Our ref: NEU-08-31 Schedule WBS: NIA Schedule Activity: NIA May 19,2008 DOMINION NUCLEAR CONNECTICUT MILLSTONE POWER STATION - MILLSTONE UNIT 3 Millstone Unit 3 (NEW Stretch Power Uprate (SPW Program Rod Withdrawal at Powler (RWAP) Peak Reactor Coolant System Pressure

Dear Mr. Thomas:

The purpose of this letter is to transmit the Millstone Unit 3 (NEU) Stretch Power Uprate (SPU) Program Rod Withdrawal at Power (RWAP) Peak Reactor Coolant System Pressure. This is contained in Attachment 1.

If you have any questions concerning this matter, please contact me at 412-374-6345.

Very truly yours, WESTINGHOUSE ELECTRIC COMPANY

~(~/I.,.

Donna Rogosky Customer Project Manager lam Attachment

Page 2 of7 Our ref: NEU-08-31 May 19,2008 cc: M.Kai Dominion M. O'Connor Dominion J. A. Lewis Dominion J. Murray Dominion B. S. Kaufman Dominion M. Elmahrabi Dominion N. Richardson Dominion R. C. Grendys Westinghouse D. Rogosky Westinghouse D. P. Dominicis Westinghouse D. C. Kovacic Westinghouse

y. Stetson Westinghouse A. Marshall Westinghouse NED Project Letter File Internal

Reference:

LTR-TA-08-103

Page 3 of7 Our ref: NEU-08-31 May 19,2008 Attachment 1 Introduction This letter provides supplemental information regarding the potential for reactor coolant system (RCS) pressure exceeding the limit ('110 percent of the design value) as a result of a rod withdrawal at power (RWAP) event for the Millstone Unit 3 Stretch Power Uprate (SPU). A conservative generic analysis, which bounds most Westinghouse plants, demonstrated that the existing plant protection features are adequate for preventing the RCS pressure from exceeding the limit. However, it has been noted that the proposed SPU NSSS power (3666 MWt) exceeds the NSSS power modeled in the generic analysis (3608 MWt). Therefore, sensitivity calculations were performed to quantitatively demonstrate that, given other consE~rvatisms in the generic analysis, it remains applicable for Millstone Unit 3 at the SPU conditions.

Generic Analysis Description The generic RWAP RCS pressure analysis was performed with the LOFTRAN computer code (Reference 1). As in other peak RCS pressure analyses (e.g., loss of load/turbine trip),

conservative initial condition uncertainties and modeling features are applied in the generic RWAP analysis so as to maximize the resultant peak RCS pressure. In order to obtain conservative RCS pressure results that can be applied to multiple plants, the following assumptions were made in the generic RWAP analysis:

(1) The initial NSSS power level is 8 percent of 3608 MWt. Eight percent corresponds to the minimum power level at which the high neutron flux low setting reactor trip can be blocked (10 percent) minus 2 percent uncertainty.

(2) Minimum reactivity feedback, including a +7 pcmrF moderator temperature coefficient, was assumed in the generic study to allow the core power to increase more rapidly, which results in a greater power mismatch between the primary and secondary systems.

(3) The range of positive reactivity insertion rates considered is consistent with the bounding range that was examined in the RWAP DNB analysis. A sensitivity study showed that insertion rates less than 20 pcm/sec are non-limiting with respect to ReS prE~ssure. The maximum reactivity insertion rat<e analyzed was 110 pcm/sec, which exceeds the maximum possible reactivity insertion rate associated with the simultaneous withdrawal of the two control rod banks having the maximum combined worth at the maximum speed.

(4) The initial reactor vessel average temperature (Tavg) is 586.5°F, which is very conservative (high) for an initial power level of 8 percent; a high initial Tavg is conservative because the rate of liquid expansion becomes more severe with increased temperature.

(5) The initial pressurizer water level, which corresponds to 10% power plus uncertainty, is 35.1 % of span. Maximizing the initial pressurizer water level minimizes the available pressurizer vapor volume space and maximizes the net pressurization effect for a given pressurizer liquid insurge.

(6) Accounting for an uncertainty of +/-50 psi, cases were evaluated at initial pressurizer pressure values of 2200 psia and 2300 psia. A sensitivity study showed that the direction

Page 4 of7 Our ref: NEU-08-31 May 19, 2008 of conservatism is dependent on the reactivity insertion rate, and thus a ranne of initial pressurizer pressure values was considered.

(7) There was no credit taken for the pressurizer power-operated relief valves' (PORVs) relief capacity.

(8) There was no credit taken for the pressurizer spray system to control RCS pressure.

(9) There was no credit taken for the steam dump control system.

(10) The pressurizer safety valve (PSV) lift setpoints were assumed to be at a maximum value of 2600 psia, which accounts for 3 percent setpoint tolerance plus 1 percent setpoint shift.

The setpoint shift is modeled along with a purge delay time of 1.5 seconds to account for water-filled PSV loop seals as discussed in WCAP-12910 Rev. 1-A (Reference 2).

(11) A maximum (bounding for all 4-loop plants) pressurizer surge line friction factor was applied to maximize the pressure drop between the RCS and pressurizer, and thereby maximize the peak RCS pressure during PSV relief conditions.

(12) Maximum (bounding for all4-loop plants) main steam safety valve setpoints were applied to delay the secondary-side steam relief.

(13) The generic RWAP analysis showed that the following two reactor trip functions were sufficient in helping (along with the PSVs) provide the protection required to limit the peak RCS pressure to an acceptable level: high pressurizer pressure (HPPT = High Pressurizer Pressure Trip) and high positive neutron flux rate (PFRT = Positive Flux Rate Trip). For the HPPT, a setpoint of 2440 psia and a signal delay time of 2 seconds were applied. For the PFRT, a setpoint of 9 percent with a time constant of 2 seconds and a signal delay time of 3 seconds were applied.

(14) The RCCA trip insertion characteristics were based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.

The generic analysis for 4-loop Westinghouse plants resulted in peak RCS pressures for RWAP events of 2708 psia and 2704 psia with plus and minus initial pressure uncertainties, respectively.

Comparison with Millstone Unit 3 SPU Table 1 provides a comparison of thB critical input parameters between the generic analysis input and the Millstone Unit 3 SPU configuration. As noted in Table 1, the only Millstone Unit 3 SPU critical input parameters not bounded by the generic analysis are the nominal (100%)

NSSS power and the initial pressurizer water level. On the other hand, there are several parameters in the generic analysis which are overly conservative with respect to the SPU. Of particular note, Millstone Unit 3 does not have water-filled loop seals, so there is no PSV loop seal purge delay.

Page 5 of7 Our ref: NEU-08-31 May 19, 2008 Table 1 I

Comparison of Westinghouse Generic RWAP RCS Pressure Analysis Critical Parameters to Millstone Unit 3 SPU Parameters Is Generic Analysis Generic Millstone Unit 3 Parameter Critical Parameter Analysis SPU Bounding?

Nominal (100%) NSSS Power :3608 MWt 3666 MWt No Power Uncertainty -2% -2% Yes Moderator Temperature

+7 pcmrF +5 pcmrF Yes Coefficient Maximum Reactivity Insertion 1 '1 0 pcm/sec 110 pcm/sec Yes Rate Initial Vessel Average Temperature (at 10% power), 586.5°F 565.25°F Yes Including Uncertainty Initial Pressurizer Water Level 35" 1 % of span 39.2 % of span No Nominal RCS Pressure 2250 psia 2250 psia Yes RCS Pressure Uncertainty +/-50 psi +/-50 psi Yes PSV Setpoint, Including 2600 psia 2575 psia Yes Tolerance and Setpoint Shift PSV Loop Seal Purge Delay 1.50 sec 0.0 sec (1) Yes HPPT Setpoint 2440 psia 2425 psia Yes HPPT Delay 2.0 sec 2.0 sec Yes PFRT Setpoint / Rate Time 6.08%(2) /2 sec 9.0% /2 sec Yes Constant PFRT Delay 3.0 sec 0.5 sec Yes (1) Millstone Unit 3 does not haVl3 water-filled loop seals.

(2) Based on the nominal trip setpoint of 5.0% plus the uncertainty of 1.08%.

Page 60f7 Our ref: NEU-08-31 May 19,2008 In order to quantify the cumulative impact of the higher NSSS power, the higher initial pressurizer level, and the fact that there are no water-filled loop seals at Millstone Unit 3, the following sensitivity cases were run using the LOFTRAN code.

• The generic analysis was modifiEld to model the SPU NSSS power of 3666 MWt with initial pressurizer pressures of 2300 psia and 2200 psia (2250 psia +/- 50 psi).

• The generic analysis was modifiEld to model the SPU NSSS power of 3666 MWt and the SPU initial pressurizer water level of 39.2% span with initial pressurizer pressures of 2300 psia and 2200 psia.

• The generic analysis was modified to model the SPU NSSS power of 3666 MWt, the SPU initial pressurizer water level of 39.2% span, and a PSV loop seal purge delay time of 0.0 second with initial pressurizer pressures of 2300 psia and 2200 psia.

Table 2 summarizes the results of these cases in comparison to the generic analysis results. As noted previously, the generic analysils resulted in a peak RCS pressure of 2708 psia. For the two cases with the SPU power modeled, the peak RCS pressures reached is 2716 psia. For the cases where the initial pressurizer water level was increased to correspond to the Millstone Unit 3 SPU program, the peak RCS pressure increased to 2730 psia. Note that these cases also modeled the increased power. Therefore, when the generic analysis is run with all critical parameters bounding the Millstone Unit 3 SPU values, the peak RCS pressure remained below the limit of 2750 psia. For the cases with the PSV loop seal purge delay time removed, the peak RCS pressure decreased to 2703 psia. This is bounded by the results of the generic analysis.

Note that these cases are still very conservative since the Tavg, moderator temperature coefficient, HPPT setpoint, PFRT setpoint, PFRT delay, and the PSV setpoint have not been modified. Therefore, the generic analysis remains applicable as a bounding analysis for Millstone Unit 3 at SPU conditions.

Table 2 Peak RCS Pressure Results Peak RCS Pressure

(+1- Initial Pressure Case Descripltion Uncertainty)

Generic Case 2708 psia 1 2704 psia NSSS Power =3666 MWt 2716 psia 12700 psia NSSS Power =3666 MWt, and 2730 psia 1 2709 psia Initial Pressurizer Water Level =3SI.2%

NSSS Power =3666 MWt, Initial Pressurizer Water Level =39.2%, and 2703 psia 1 2E;66 psia PSV Loop Seal Purge Delay =0.0 second

Page 7 of7 Our ref: NEU-08-31 May 19,2008 Conclusion The results of the analysis demonstrate that a RWAP event will not result in RCS overpressurization for Millstone Unit 3 at the SPU conditions. When the critical parameters in the generic analysis are compared with the Millstone Unit 3 parameters at the SPU conditions, only the NSSS power and the pressurizer water level are not bounded. The sensitivities documented above show that when the Millstone Unit 3 SPU power level and pmssurizer water level are accounted for, the peak RCS pressure remains below the limit value. In addition, when credit is taken for the fact that the plant does not have water-filled loop seals which would cause a delay in PSV relief, the peak RCS pressure is less limiting than that reached in the generic analysis. Therefore, the generic analysis is applicable to Millstone Unit 3 at the SPU conditions.

References

1. WCAP-7907-A, "LOFTRAN CodE~ Description," 1. W. 1. Burnett, April 1984.

2. WCAP-12910 Revision 1-A, "Pressurizer Safety Valve Set Pressure Shift, WOG Project MUHP 2351/2352," G.O. Barrett, June 1993 (Westinghouse Proprietary Class 2).