ML17244A742

Application for Amend to License DPR-18 Re Undervoltage Protection

ML17244A742
Person / Time
Site:	Ginna
Issue date:	07/31/1979
From:	White L ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17244A741	List: ML17244A740 ML17244A742
References
NUDOCS 7908080446
Download: ML17244A742 (18)

Text

BEFORE THE UNXTED STATES NUCLEAR REGULATORY COMMISSSXON In the Matter of ROCHESTER GAS AND ELECTRIC CORPORATION Docket No.52-244 (R.E. Ginna Nuclear Power Station, Unit No. 1)

CERTIFICATE OF SERVICE I hereby certify that I have served a document entitled "Application for Amendment to Operating License" with three (3) documents, Attachments A, B, and C, attached thereto, by mailing copies thereof first class, postage pre-paid, to each, of the following persons this 3rd day of August, 1979:

Mr. Michael L. Slade 12 Trailwood Circle Rochester, New York 14618 Warren B. Rosenbaum, Esq.

One Main Street 707 Wilder Building Rochester, New York 14614 Edward G. Ketchen, Esp.

Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. Robert N. Pickney Supervisor Town of Ontario 107 Ridge Road West Ontario, New York 14519 V 908080 Q/(p '-

Jeffrey L. Cohen, Esq.

New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza Albany, New York l2223 Edward Luton, Esq.

Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commis sion Washington, D.C. 20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commis sion Washington, D.C. 20555 Dr. Dixon Callihan Union Carbide Corporation P. O. Box Y Oak Ridge, Tennessee 37830 L x K. Larson LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas and Electric Corporation

II 'i f

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. l) )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE

90% of normal indicated flow ~

g. Low reactor coolant pump frequency - > 57.5 Hz.

2.3.1.3 Other reactor tri 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 inter-locks 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 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 Acceptable degraded voltage relay operating times and setpoints, for 480 volt safeguards bus voltages < 414 volts'and > 368 volts are defined by the safeguard equip-ment thermal capability curve shown in Figure 2.3-1.

Basis:

TTTe Eigh flux reactor trip (low set point) provides redundant pro-tection in the power range for a power excursion beginning fop low power. This trip value was used in the safety analysis.

2. 3-4 PROPOSED

e 0 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)(11)

References:

(1) FSAR 14.1.1 (2) FSAR, Page 14-3 (3) FSAR 14.3.1 (4) FSAR 14.1.2 (5)FSAR72I 7~3 (6) FSAR 3. 2. 1 (7) FSAR 14.1.6 (8) FSAR 14.1.9 (9) Letter from L.D. White, Jr. to A. Schwencer, NRC, dated September 30, 1977 (10) Letter from L.D. White, Jr. to A. Schwencer, NRC, dated September 30 1977 (11) Letter from L.D. White, Jr. to D. Ziemann, NRC, dated July 24, 1978 2.3-8 PROPOSED

I 1600 I

o. Second Level 0 I

~

I.

'Under Voltage Protection L I

1400 Figure 2.3-1 1200 ~ l

-+-'

1000 td I I'I "I I

~

800 ~ 1 I ~

8 8 I ~ ~

d 600 0

0 400 AllI+

I" owa ble I

. Re la I'Iy Oper ating I l

200 ~ iRe ion 0

Secondary Volts (120 Volts). 70 80 90 92 120 Primary Volts (480 Volts) t 240 320 368 103.5'14

=480 Percent Volts (460 Volt Base) 528 70% 80% 90% 104%

ISafeguards Bus Voltage'.3-9 PROP OSED s ~

I

Page 3 of 3 1 2 3 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 2 Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 h~

with two channels in~

operable.

b.Condenser Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 hrs.

with two channels inoperable.

18. Loss of Voltage/

Degraded Voltage 480 Volt Safe-guards Bus 4/bus 2/bus 2/bus Maintain hot shut-down or place bus on diesel generator.

NOTE 1: When block condition exists, maintain normal operation.

F.P. Full Power

• Not Applicable

• • If both rod misalignment monitors (a and b) 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 in-creased 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.

TABLE 4.1-1 (CONTINUED)

Channel Check Calibrate Test Remarks

25. Containment Pressure Narrow range containment pressure

(-3.0, +3 psig excluded)

26. Steam Generator Pressure

27. Turbine First Stage Pressure S

28. Emergency Plan Radiation Instruments

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

30. Loss of Voltage/Degraded .N.A.

Voltage 480 Volt Safe-guards Bus S - Each Shift M Monthly D Daily P Prior to each startup if not done previous week B/N Biweekly R .- Each Refueling Shutdown Q Quarterly N.A. - Not applicable

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 the basis for a modification which would reduce the Station's susceptibility to a sustained degraded voltage.

However, due to lack of specific information regarding equipment, setpoints for the additional undervoltage protection were not established at that time.

Following receipt of additional information, RG&E submitted setpoints in a letter to the NRC dated September 30, 1977. The proposed specifications are identical to those presented in the September 30 letter.

These Specifications will provide protection for a complete loss of 480 volt.bus voltage as well as for degraded voltage conditions. Both relaying systems will assure that assumptions of all safety analysis are met. Specifically, equipment will be loaded onto the diesel generators within the time assumed in the safety analyses. The undervoltage relay protection, which was identified as "second-level protection" in the June 3, 1977 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.

The NRC Staff requested that the proposed system be modified to provide "coincident logic" as described in position 1, part b of their June 3, 1977 letter. Provision for "coincident, logic" does not. effect the setpoint information contained in the December 16, 1977 Application for Change to Operating License and re-submitted herewith. However, it has changed the required number of relays necessary to trip. The coincident logic scheme is described in our July 24, 1978 letter (ref. 11 to the Technical Specification Basis).

Loss of voltage and degraded voltage conditions simulated during the refueling shutdown are performed to verify both system performance 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, September 30, 1977, and July 24, 1978 the proposed Technical Specification will provide protection against 480 volt bus under-voltage.

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 require approximately 9 months. Installation must be performed at. a cold or refueling shutdown and would require approximately two weeks. .The effective data of a Technical Specification should be set consistent with this schedule.

Attachment C The Technical Specification change proposal revises a proposal which was filed prior to originally submitted December 22, 1977 The the effective date of 10 CPR 170.22. revision is necessary to incorporate a change in design which was requested by the NRC Staff. The design change was described in our submittal of July 24, 1978. Because it is a revision which the Staff requested and is now reviewing, and is not a new request, no fee is required.

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