ML18142B948

R. E. Ginna - Letter Transmitting 3 Signed Originals and 19 Copies of Document Entitled, Application for Amendment to Operating License.

ML18142B948
Person / Time
Site:	Ginna  (DPR-018)
Issue date:	12/22/1977
From:	LeBoeuf, Lamb, Leiby & MacRae, Rochester Gas & Electric Corp
To:	Case E Office of Nuclear Reactor Regulation
References
Download: ML18142B948 (24)

Text

DISTRIBUTION AFTER ISSUANCE OF OP TING LICENSE U.S. NUCI.EAR REGULATORY COMMISS, OOCKET NUMBER

~NRC EORM 195 I2 7BI PILE NUMBER NRC DISTRIBUTION FoR PART 50 DOCKET MATERIAL FROM: OATE OF OOCUMENT TO:

LeBoeuf, Lamb. Leiby & MacRae 12/22/77 Mri Edson Gi Case Washington~ Di C OATS RECEIVEO LeBoeuf, Lamb. Leiby & MacRae 12/27/77 CINOTOR IZE O PROP INPUT FORM NUMBER OF COPIES RECEIVEO I LG R ORIGINAL KKINCLASSIFIEO Cicor v OGSCRIPTION ENCLOSURE License No+ DPR-18 Appl for Amend: tech specs proposed change concerning addition of requirements for undervo1tage requirements

~ inotorized 12/16/77 of Serviceo ~ ~ ~

'/att Certificate (1-P) (2<

rr JAMES F. JOHNSON, EUGENC B. THOMASr JR 44 LEONARD M. TROSTENr4 @4rr j4rr'rrrrc CASLC ADDRESS HARRY H. VOIOT r 4 . RONALD D. JONES LEX K. LARSONS 4 ORANT S. LEWIS H. RICHARD WACHTKL OCRARO P. WAT9ON 'o '. LCSWIHr HEW YORK TCLCX: 4234IS 4 RCSIDENT PARTNERS WASHINGTON OFFICE 4 ADMITTED TO THC DISTRICT OF COLUMBIA BAR December 22, l977 Mr. Edson G. Case Acting Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Rochester Gas and Electric Corporation R. E. Ginna Nuclear Power Station, Unit. No. 1 Docket No. 50-244

Dear Mr. Case:

As counsel for Rochester Gas and Electric Corporation, we hereby transmit three (3) signed originals and nineteen (l9) copies of a document entitled, "Application for Amend-ment to Operating License," together with forty (40) copies of proposed changes to Technical Specifications and accompany-ing safety evaluation. Thi's Application seeks to amend the Technical Specifications to add requirements for under-voltage protection.

A Certificate of Service showing service of these documents upon the persons listed therein is also enclosed.

Very truly yours, WTE/bg Enclosures

III BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of )

)

ROCHESTER GAS AND ELECTRIC, ) Docket No. '50-244 CORPORATION '

. )

(R. E. Ginna Nuclear Power )

Station, Unit No. '1) )

CERTIFICATE OF SERVICE The undersigned hereby certifies that the "Application for Amendment to Operating License" together with a copy of the proposed changes to Technical Specifications and accom-panying safety evaluation were this twenty-second day of December, 1977, served, by first class mail, postage prepaid, on the following persons:

Chairman, Atomic Safety L. Dow Davis, IV, Esquire and Licensing Board Panel Office of the Executive Legal U.S. Nuclear Regulatory Director Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Edward Luton, Esquire U.S. Nuclear Regulatory Atomic Safety and Licensing Commission Board Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Mr. Michael L. Slade Washington, D.C. 20555 1250 Crown Point Drive Webster, New York 14580 Dr. Emmeth A. Luebke Atomic Safety and Licensing Warren B. Rosenbaum, Esquire Board One Main Street East U.S. Nuclear Regulatory 707 Wilder Building Commission Rochester, New York 14614 Washington, D.C. 20555

)

Dr. A. Dixon Callihan Jeffrey L. Cohen, Esquire Union Carbide Corporation ~ New York State Energy P. O. Box Y Office Oak Ridge, Tennessee 37830 Swan Street Building Core I, Second Floor Mr. Robert N. Pinkney Empire State Plaza Supervisor Albany, New York 12223 Town of Ontario 107 Ridge Road West Ontario, New York 14519 h

Walter T. Eccard LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas and Electric Corporation

UNITED STATES OF AMERICA NUCLEAR, REGULATORY COMMISSION In the Matter of )

)

Rochester Gas and Electric Corporation ) Docket No. 50-244 (R.E. Ginna Nuclear Power Plant, )

Unit No. 1) )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the U.S.

Nuclear Regulatory Commission (the "Commission" ), Rochester Gas and Electric Corporation ("RGSE"), holder of Provisional Operating License No. DPR-18, hereby requests that the Technical Specifications set forth in Appendix A to that License be amended to add require-ments for undervoltage protection. This request for a change in the Technical Specifications is submitted in accordance .with a letter from A. Schwencer, Chief, Operating Reactors Branch 51, dated June 3, 1977.

The proposed technical specification change is set forth in Attachment A to this Application. A safety evaluation is set forth in Attachment B. This evaluation also demonstrates that the proposed change does not involve a significant change in the types or a significant increase in the amounts of effluents- or any change in the authorized power level of the facility.

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WHEREFORE, Applicant respectfully requests that Appendix A to Provisional Operating License No. DPR-18 be amended in the form attached hereto as Attachment A.

Rochester Gas and Electric Corporation By L. D. White, Jr.

Vice President, Electric and Steam Production Subscribed and sworn to before me on this (Q day of December 1977.

STC: II CN K trIDA IIOTARY PUBLIC, State of II.Y, I tonrce County gy Commission Expires Inarch 30, 197+

Attachment A

1. Remove pages 2.3-4, 2.3-8, 3.5-4, 3;5-4a, and 4.1-7.

2. Insert the enclosed revised pages 2.3-4, 2.3-8, 2.3-9, 3.5-4a, and 4.1-7.

f. Low reactor coolant flow > 90% of normal indicated flow.

g. Low reactor coolant pump frequency > 57.5 Hz.

2.3.1.3 Other reactor t'ri s

a. High pressurizer water level < 92% of span

b. Low-low steam generator water level > 5% of narrow range instrument span 2.3.2 Protective instrumentation settings for reactor trip interlocks shall be as follows:

2.3.2.1 Remove bypass of "at power" reactor trips at high power (low pressurizer pressure and low reactor coolant flow) for both loops:

Power range nuclear flux -' 8.5% of rated power (1) (Note: During cold rod drop tests, the r

pressurizer high level trip may be bypassed.)

2.3.2.2 Remove bypass of single loss of flow trip at high power:

Power range nuclear flux -' 50% of rated power 2.3.3 Relay settings for 480 volt safeguards bus protection shall be as follows:

2.3.3.1 Loss of voltage relay operating time' 8.5 seconds for 480 volt safeguards bus voltages < 368 volts 2.3.3.2 Degraded voltage relay operating times and setpoints, for 480 volt safeguards bus voltages < 414 volts and 368 volts. See Figure 2.3-1.

Basis:

The high flux reactor trip (low set point) provides redundant protection in the power range for a power excursion beginning from low power. This trip value was used in the safety analysis.

2.3-4 PROPOSED

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the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase. For this reason the single pump loss of flow trip can be bypassed below 50%

power.

The loss of voltage and degraded voltage trips ensure opera-bility of safeguards equipment during a postulated design basis event concurrent with a degraded bus voltage condition. (9) (10)

References:

(1) FSAR 14.1.1 (2) FSAR, Page 14-3 (3) FSAR 14.3.1 (4) FSAR 14.1.2 (5) FSAR 7.2, 7.3 (6) FSAR'.2.1 (7) FSAR 14.1.6 (8) FSAR 14.1.9 (9) Letter from L.D. White, Jr. to A. Schwencer, NRC, dated July 21, 1977 (10) Letter from L.D. White, Jr. to A. Schwencer, NRC, dated September 30, 1977 PROPOSED 2.3-8

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40 1600 0

I0 Ul pearaded Voltage Protection "8 1400 E' Figure 2. 3-1

'I 1200 ~

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S4 0 1000 l5 I

8 ~ ~ ~ ~ I C

800 I 8

l5 0 600 4

8 l5 400 8 I Q I (Allowable Rela 0s crating 200 I 4 ~

0 Secondary Volts (120 Volts) 70 80 90 92 103. 5. 120 Primary Volts (480 Volts) 240 320 368 414 480 Percent Volts (460 Volt Base) 52% 70% 80% 90% 104%

Safeguards Bus Voltage 2.3-9

Page 3 of 3 1 2 3 4 5 6 NO. of MIN. MIN. ~

PERMISSIBLE OPERATOR ACTION NO. of CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS COLUMN 3 or 5 CANNOT BE MET

17. Circulating Water Flood Protection a.Screenhouse Power operation may be con-tinued for a period of up to 7 days with 1 channel in-operable or for a period ~

24 hrs. with two channels-~

inoperable.

b.Condenser Power operation may be con-tinued for a period of up to 7 days with 1 channel in-operable or for a period of 24 hrs. with two channels inoperable.

18. Loss of Voltage Maintain hot shutdown or 480 Volt Safe- 2/bus 1/bus 1/bus

• place bus on diesel generator guards Bus 19- Degraded Voltage Maintain hot shutdown or 480 Volt Safe- 2/bus 1/bus 1/bus

• place bus on diesel generator guards Bus NOTE 1: When block condition exists, maintain normal operation.

F.P. = Full Power

.* Not Applicable

• • If both rod misalignment monitors (a and b) are inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, the nuclear overpower trip shall be reset to 93% of rated power in addition to the increased surveillance noted.

• • • If a functional unit is operating with the minimum operable channels, the number of channels to trip the reactor will be column 3 less column 4.

+ A channel is considered operable with 1 out of 2 logic or 2 out of 3 logic.

PROPOSED

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TABLE 4.1-1 (CONTINUED Channel Check Calibrate Test Remarks

25. Containment Pressure M Narrow range containment pressure

(-3.0, +3 psig excluded)

26. Steam Generator Pressure

27. Turbine First Stage Pressure-

28. Emergency Plan Radiation Instruments

29. Environmental Monitors N.A. N.'A.

30. Loss of Voltage 480 N.A.

Volt Bus

31. Degraded Voltage 400 Volt N.A.

Safeguards Bus Each Shift M Monthly D Daily P Prior to each startup if not done previous week B/N Biweekly Each Refueling Shutdown Quarterly N.A. Not applicable PROPOSED

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Attachment B By letter dated June 3, 1977, the NRC requested that RG&E assess the susceptibility of safety related electrical equipment with regard to (1) sustained degraded voltage conditions at the offsite power sources and (2) interaction between the offsite and onsite emergency power sources.

An analysis of undervoltage protection at Ginna Station was performed and submitted to the NRC in a letter dated July 21, 1977.

This analysis reviewed the current undervoltage protection and described a modification which would reduce the Station's su cep-tibility to a sustained degraded voltage. The basis for the modification was described; however, due to lack of specific in-formation regarding certain equipment, specific setpoints for the additional undervoltage protection could not be established.

Following receipt of additional information, RGsE submitted suggested setpoints in a letter to the NRC dated September 30, 1977. The proposed specifications are identical to those pre-sented in the September 30 letter.

The proposed Specifications will provide protection both for complete loss of 480 volt bus voltage and for degraded bus voltage.

The loss of voltage protection, provided by existing equipment, will assure that assumptions of all safety analyses are met for a complete loss of voltage to safeguards equipment. Specifically, equipment will be loaded onto the diesel generators within the time assumed in the safety analyses. The undervoltage relay

~1 protection, which was identified as "second-level protection" in the June 3 NRC letter and in our two subsequent letters, will protect equipment against a bus voltage which is greater than the loss of voltage relay setpoint but less than the voltage guaranteed by equipment manufacturers for continuous duty for Ginna safeguards equipment. The proposed undervoltage setpoints will provide the required protection while also assuring that spurious trips will-not occur while equipment is being sequenced onto the diesel generators.

Loss of voltage and degrade voltage conditions simulated during the refueling shutdown are performed to verify both system per-formance and relay calibration. The safeguards 480 volt loss of voltage and degraded voltage protection systems are not part of the reactor protection system. It is not practical to simulate these conditions during plant operation to satisfy a monthly test requirement.

Based on the analyses provided in our letters of July 21, 1977 and September 30, 1977, the proposed Technical Specification will provide protection against 480 volt bus undervoltage.

As stated in our letter of July 21, 1977, we expect that, following receipt of approval of the modification design by the NRC, detailed engineering and procurement of equipment will re-quire approximately 9 months. Installation must be performed at a cold or refueling shutdown and would require approximately two weeks. The effective date of a Technical Specification should be set consistent with this schedule.

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