ML20033A217
| ML20033A217 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 09/30/1981 |
| From: | Weber D EG&G, INC. |
| To: | Bajwa S Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6429 EGG-EA-5598, NUDOCS 8111240924 | |
| Download: ML20033A217 (15) | |
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DEGRADED GRID PROTECTION FOR CLASS lE PnWER SYSTEMS,
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....ns1. did This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 t
FIN No. A6429 8111240924 810930 PDR RES o
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FORM EG8G 3M (Rev ii19p INTERIM REPORT Accer: ion No.
Report No. _ EGG-EA-5598 Contract Program or Project
Title:
Selected Operating Reactor Issues Program (III)
Cubject of this Document:
Degraded Grid Protection for Class lE Power Syttems, Indian Point Nuclear Station, Unit 3, Docket No. 50-286 Type of Document:
Informal Report Author (s):
D. A. Weber Date of Document:
September 1981 Responsible NRC Individual and NRC Office or Division:
S. S. Bajwa, Division of Licensing This document was prepared primarily for preliminary or internal use. it has not received full review and approval. Since there may be substantise changes,this document shoulo not be considered final.
EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE-AC07 761D01570 NRC FIN No.
A6429 INTERIM REPOAT e
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t DEGRADED GRID PROTECTION FOR CLASS lE POWER SYSTEMS INDIAN POINT NUCLEAR STATI0ft UNIT 3 Docket No. 50-286 4
D. A. Weber Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.
September 1981 T
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ABSTRACT This report contains the EG&G Idaho, Inc. evaluation of the Indian Point duclear Station, Unit 3, degraded grid protection for Class IE systems.
FOREWORD This report is supplied as part of the " Selected Operating Reactor Issues Program (III)" being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing, by EG&G Idaho, Inc., Reliability and Statistics Branch.
The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20 19 01 06, FIN No. A6429.
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CONTENTS
1.0 INTRODUCTION
1 2.0 DESIGN BASE CRITERIA............................................
2 3.0 EVALVATION......................................................
2 3.1 Existing Undervoltage Protection..........................
3 3.2 Modifications..............................-..............
3 3.3 Discussion................................................
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4.0 CONCLUSION
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5.0 REFERENCES
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TECHNICAL EVALUATION REPORT DEGRADED GRID PROTECTION FOR CLASS lE POWER SYSTEMS INDIAN POINT NUCLEAR STATION UNIT 2
1.0 INTRODUCTION
On June 3, 1977, the NRC requested the Consolidated Edison Company (Con-Ed) to assess the susceptibility of the safety-related electrical equipment at the Indian Point Nuclear Station Unit No. 3 (IP-3) to a sus-tained voltage degradation of the offsite source and interaction of the offsite and onsite emergency power systems.I The letter contained three positions with which the current design of the plant was to be compared.
After comparing the current design to the staff positions, Con-Ed was required to either propose modifications to satisfy the positions and cri-teria or furnish an analysis to substantiate that the existing facility design has equivalent capabilities. Since the NRC's original letter of June 3, 1977, the Power Authority of the State of New York (PASNY) has assumed ownership of the Indian Point Unit 3 facilities.
Con-Ed responded to the NRC letter with two submittals dated August 29, 1977.0'7 These submittals and the submittals of September 20, 1976,2 September 24, 1976,3 December 17, 1976,4 March 31, 1977,5 September 19, 1977,0 February 11, 1980 (PASNY),9 May 30, 1980 (PASNY),10 and the Indian Point Unit No. 3 Final Safety Analysis Report (FSAR)II complete the information reviewed for this report.
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2.0 DESIGN BASE CRITERIA The design base criteria that were applied in determining the accep-tability of the system modifications to protect the safety-related equip-ment from a sustained degradation of the offsite grid are:
1.
General Design Criterion 17 (GDC 17), " Electrical Power Systems," of Appendix A, " General Design Criteria for 16 Nuclear Power Plants," of 10 CFR 50 2.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations"I7 3.
IEEE Standard 308-1974, " Class lE Power Systems for Nuclear Power Generating Stations"18 4.
Staff positions as detailed in a letter sent to the 3
licensee, dated June 2, 1977 l
S.
ANSI Standard C84.1-1977, " Voltage Ratings for Electri-cal Power Systems and Equipment (60 Hz)."I9 3.0 EVALUATION This section provides, in Subsection 3.1, a brief description of the
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existing undervoltage protection at IP-3; in Subsection 3.2, a description of the licensee's proposed modifications for the second-level undervoltage l
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protection; and in Subsection 3.3, a discussion of how the proposed modifi-cations meet the design base criteria.
3.1 Existing Undervoltage Protection. There are four 480V class lE buses (2A, 3A, SA, and 6A) for Indian Point 3.
Each of the buses is equip-ped with CV-7 inverse-time relays set at 46% (220V) which automatically strip their associated loads (except safeguard MCC36A, 36B and 36C) after 2 seconds. These buses are also equipped with additional CV-7 relays which will initiate load shedding, start the emergency diesel generators, and energize the emergency buses through load sequencing operation.
3.2 Modifications. The licensee has proposed to install two second-level undervoltage relays on each 480 volt safety-related bus in a two-out-of-two logic. The set point for each relay is 403 volts (84%) with a time delay of 180 seconds. The existing time delay on the loss-of-voltage relays has been extended from 120 cycles (2 seconds) to 3 seconds.
In addition the licensee has added undervoltage relays on each of the safety-related buses which will provide annunciation to the operator when the bus voltage drops to 93.3%.9 Proposed changes to the plant's technical specifications were also furnished by the licensee.
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3.3 Discussion. The first position of the NRC staff letter required that a second level of undervoltage protection for the onsite power system be provided. The letter stipulates other criteria that the undervoltage protection must meet. Each criterion is restated below, followed by a dis-cussion regarding the licensee's compliance with that criterion.
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1.
"The selection of voltage and time setpoints shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite system dis-tribution levels."
The licensee has provided an analysis of the voltage requirements of the safety-related loads at all onsite system distribution levels and have concluded that the 460V motors are the most limiting safety-related equip-ment. The analysis was performed for the continously running safety-related motors, all of which have ser-vice factors of 1.15 and running load less than the nameplate rating of the motor.10 PASNY's proposed Technical Specifications require that the 480V Emergency Bus Undervoltage (Degraded Voltage) relays have a setpoint of 403V + SV. This setpoint and tolerance will provide adequate protection for the safety-related loads at all onsite system distribution levels.
2.
"The voltage protection shall include coincident logic to preclude spurious trips of the offsite power P
sources."
The proposed modification incorporates a two-out-of-two logic scheme, thereby satisfying this criterion.10 4
3.
"The time delay selected shall be based on the follow-ing conditions:
a.
"The allowable time delay, including margin, shall not exceed the maximum time delay that is assumed in the FSAR accident analysis."
The proposed maximum time delay of 3 seconds +
1 second for the loss-of-voltage relays does not exceed this maximum time delay.
b.
"The time delay shall minimize the effect of short-duration disturbances from reducing the unavaila-bility of the offsite power source (s)."
The licensee's proposed minimum time delay of 180 seconds is long enough to override any short, inconsequential grid disturbances and the starting of large motors.
c.
"The allowable time duration of a degraded voltage condition at all distribution system levels shall not result in failure of safety systems or compon-ents."
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The proposed time delay of 180 seconds +
30 seconds will not result in failure of the safety-related equipment.
4.
"The voltage monitors shall automatically initiate the disconnection of of fsite power sources whenever the voltage setpoint and time-delay limits have been exceeded."
A review of the licensee's proposal substantiates that this criterion is met.
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5.
The voltage monitors shall be designed to satisfy the requirements of IEEE Standard 279-1971."
The licensee has stated in his proposal that the modi-fications are designed to meet or exceed IEEE Stan-dard 279. O 6.
"The technical specifications shall include limiting conditions for operation, surveillance requirements, trip setpoints with minimum and maximum limits, and allowable values for the second-level voltage protec-tion monitors."
The licensee has provided surveillance requirements but the requirement to " test" every 18 months (noted as "R" 6
for refueling in the proposed Technical Specification) is not acceptable. Testing (Channel Functional Test) frequency should agree with the NRC model Technical Specifications (at least once per 31 days).
The second NRC staff position requires that the system design automat-ically prevent load-shedding of the emergency buses once the onsite sources are supplying power to all sequenced loads. The load-shedding must also be reinstated if the onsite breakers are tripped.
The existing undervoltage relaying scheme for all safety-related buses already has these features incorporated. Only the time delay will be extended, from 2 seconds to 4 seconds when the system is modified for second-level undervoltage protection.
The third NRC staff position requires that certain test requireuents he added to the technical specifications. These tests were to demonstrate the full-functional operability and independence of the onsite power sources, and are to be performed at least once per 18 months during shut-down. The tests are to simulate loss of offsite power in conjunction with a safety-injection actuation signal, and to simulate interruption and sub-sequent reconnection of onsite power sources. These tests verify the proper operation of the load-shed system, the load-shed bypass when the emergency diesel generators are supplying power to their respective buses, and that there ir no adverse interaction between the onsite and offsite power sources.
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The position is satisfied as the Indian Point 3 Technical Specifica-tions describe tests to demonstrate the full-functional operatility and independence of the onsite systems.
4.0 CONCLUSION
S Based on the information provided by Con-Ed and PASNY it has been determined that the proposed modifications comply, with one exception. to the NRC staff positions as described in their letter of June 3, 1977.I To comply with this letter the licensee should:
1.
Change the unit technical specification surveillance requirements for second-level and loss-of-voltage Channel Functual Test to agree with the NRC requirements (at least once per 31 days).
5.0 REFERENCES
1.
NRC letter (R. W. Reid) to Con-Ed, " Staff Positions Relative to the Emergency Power Systems for Operating Reactors," dated June 3, 1977.
2.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated September 20, 1976.
3.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated I
September 24, 1976.
(A partial response to the NRC's generic letter of August 12, 1976 (Effects of Degraded Grid Voltage) and updating the i
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Con-Ed letter of September 20,1976.)
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l 4.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated December 17, 1976.
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5.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated March 31, 1977.
6.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated August 29, 1977.
(A complete response to the NRC's generic letter of August 12, 1976, and updating Con-Ed's letter of September 24,1976.)
7.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated August 29, 1977. (Responding to the NRC letter of June 3,1971.)
8.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (W. Gamill), dated September 19, 1977.
9.
PASNY letter (Paul J. Early) to NRC (W. Gammill), dated February 11, 1980.
- 10. PASNY letter (Paul J. Early) to NRC (S. A. Varga), dated May 30, 1980.
- 11. Final Safety Analysis Report (FSAR) for the Indian Point Nuclear Sta-tion Unit 3.
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- 12. General Design Criterion 17, " Electric Power Systems," of Appendix A,'
" General Design Criteria f or Nuclear Power Plants," to 10 CFR Part 50,
" Domestic Licensing of Production and Utilization Facilities."
13.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
14.
IEEE Standard 308-1974, " Standard Criteria for Class lE Power Systems for Nuclear Power Generating Stations."
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- 15. ANSI C84.1-1977, " Voltage Ratings for Electric Power Systems and Equip-ment (60 Hz)."
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16.
IEEE Standard 141-1976, "IEEL flecor.nended Practice f or Electric Power Distribution for Industrial Plants."
17.
NEMA Standard,liEMA MGl-1972, " Motors and Generators."
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