ML20031F025
| ML20031F025 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 09/30/1981 |
| From: | Weber D ENERGY ENGINEERING GROUP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6429 EGG-EA-5477, NUDOCS 8110190070 | |
| Download: ML20031F025 (11) | |
Text
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So tM7 EGG-EA-5477 SEPTEMBER 1981 i
FoK AMk DECRADED GRID PROTECTION FOR CLASS lE POWER SYSTEMS, pris INDIAN POINT NUCLEAR STATION UNIT 2, DOCKET NO. 50-247 A)Sif (f
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This is an informal report intended for use as a preilminary or working document l
Prepared for the U.S. Nuclear Regulatory Comission g
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- INTERIM REPORT Accession No. _____
Report No _EGGdA-5477 Contract Program or Project
Title:
Selected Operating Reactor Issues Program (III)
Subject of this Document:
Degraded Grid Protection for Class lE Power Systems, Indian Point Nuclear Station Unit 2, Docket No. 50-247 Type of Document:
Technical Evaluation Report E lesearci arx "ecinical
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Assistance Report Date of Document:
September 1981 Responsible NRC Individual and NRC Office or Division:
P. C. Shemanski, 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 substantive changes, this document should 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 REPORT
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i DEGRADED GRID PRCTECTION FOR CLASS lE POWER SYSTEMS INDIAN POINT NUCLEAR STATION UNIT 2 Docicet No. 50-247 D. A. Weber Reliability and Statistics Branch l
Engineering Analysis Division EG&^ Idaho, Inc.
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ABSTRACT
,i In June 1977, the NRC sent all operating reactors a letter outlining three positions the staff had taken in regard to the onsite emergency power systems. Consolidated Edison Company (Con-Ed) was to assess the suscepti-bility of the safety-related electrical equipment at the Indian Point Nuclear Station Unit 2, to a sustaine'd. voltage degradation of the offsite source and interaction of the offsite and onsite emergency power systems.
This report contains an evaluation of Con-Ed's analyses, modifications, and technical specification changes to comply with these NRC positions. The evaluation has" determined that Con-Ed does not comply with one of the NRC positions.
FOREWORD I
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 i
Jauthorization, B&R 20 19 01 06, FIN No. A6429.
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CONTENTS I
1.0 INTRODUCTION
1 2.0 DESIGN BASE CRITERIA...........................................
3.0 EVALUATION......................................................
1 3.1 Existing Undervoltage Protection..........................
2 3.2 Modifications.............................................
2 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 GRIO PROTECTION FOR CLASS lE POWER SYSTEMS INDIAN POINT NUCLEAR STATICN UNIT 2
1.0 INTRODUCTION
On June 2,1977, 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. 2 (IP-2) to a sus-taineo voltage degradation of the offsite source and interaction of the offsite and onsite emergency power systems.1 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 rrdifications to satisfy the positions and cri-teria or furnish an analysis to substantiate that the existing facility design has equivalent capabilities.
1977.gog-EdrespondedtotheNRCletterwithtwosubmittalsdatedAupst29, ThesesubmittalsandthesubmjttalsofSeptembpr 20, 1976, 24, 1976,5 Decem ScptegberSeptembegl5,1977,ger17,1976,31, 1980,g, Marc 1979,g0 14,1981,g' April 27, April
, 1980,
- 1977, October 1
- 1980, December Augusp5,nd the Indian Point Unit No. 2 Final Safety Analysis Report 1981, l6 a (FSAR) complete the information reviewed for this report.
2.0 DESIGN BASE CRITERIA The design base critaria 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 D ign Criteria for NuclearPowerPlants,"of10CFR50pD i
for Nuclear Power Generating Stations"gotection Systems IEEE Standard 279-1971, " Criteria for 2.
3.
IEEE Standard 308-1974, " Class 1 Power Systems for Nuclear Power Generating Stations"18 4.
Staff positions as detailed in a letter sent to the 1
licensee, dated June 2, 1977 5.
ANSI Standard C84.1-1977, " Voltage Rating {gh Mech cal Power Systems and Equipment (60 Hz)."
3.0 EVALUATION This section provides, in Subsection 3.1,'a-brief description of the existing undervoltage protection at IP-2; in Subsection 3.2, a description 1
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of the licensee's proposed modifications for the second-level undervoltage 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 1E buses (2A, 3A, SA, and 6A) for Indian Point 2.
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 MCC26A and 268) after 2 sec-onds. 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 l
relayshasbeenextendedfrom120 cycles (2 seconds)to3 seconds.gotage In addition the licensee has added undervoltage relays on each of the safety-relatedbuseswhichwil{provideannunciationtotheoperatorwhenthebus voltage drops to 93.3%.
Proposed changes to the plant's technical specifications were also furnished by the licensee.
I required 3.3 Discussion. The first position of the NRC staff letter 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.
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 systern 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.
Con-Ed'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
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levels.
2.
"The voltage protection shall include coincident logic to preclude spurious trips of the offsite power sources."
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The proposed modification incorporates a two-out-qf-two logic scheme, thereby satisfying this criterion.l3 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 f ailure of safety systems or compon-ents."
The proresed time delay of 180 seconds +
30 secor.ds will not result in failure of the safety-related equipment.
4.
"The voltage monitors shall automatically initiate the disconnection of offsite 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.
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 279.11 6.
"The technical specifications shall include limiting conditions for operation, surveillance requirements, trip setpoints with minimum and maximum limits, and allowable valces for the second-level voltage protec-ti monitors."
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The licensee has provided surveillance requirements but the requirement to " test" every 18 months (noted as "R" for refueling in the proposed Technical Specification) is not acceptable. Testing (Channel Functional Test) frequencyshouldagreewiththeNRCmodelTghnical Specifications (at least once per 31 days)
The second NRC staff position requires that the system design automat-ically prevent load-shrdding 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 seconu-level undervoltage protection.
The third NRC staff position requires that certain test requirements be added to the technical specificatior These tests were to demonstrate the full-functional operability and in.apendence of the onsite power sources, and are to be performed at 1 cast once per 18 months during shut-down. The tests are to simulate loss of offsite power in conjunction with a safety-injection actuation sigr.al, 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 is no adverse interaction between the onsite and offsite power sources.
The position is satisfied as the Indian Point 2 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, it has been determined NRCstaffpositionsasdescribedintheNRCletterofJune2,1977.githe that the proposed modifications, generally, do not comply with one 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
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1.
NRC letter (R. W. Reid) to Con-Ed, " Staff Positions Relative'to the Emergency Power Systems for Operating Reactors," dated June 2, 1977.
2.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated August 29, 1971.
(Responding to the NRC's generic letter of August 12, 1976 (Effects of Degraded Grid Voltage) and updating the Con-Ed letter of September 24,1976.)
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3.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated August 19, 1977.
(Responding to the NRC letter of June 2, 1977 regarding emer-gency power systems.)
4.
Con-Ed letter (W. J. Cahill, Jr.) to NRC-(R. W. Reid), dated September 20, 1976.
5.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated September 24, 1976.
6.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated December 17, 1976.
7.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated March 31, 1977.
8.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated June 17, 1977.
9.
Con-Ed letter (W. J. Cahill, Jr.) to NRC (R. W. Reid), dated September 15, 1977.
- 10. Con-Ed letter (W. J. Cahill, Jr.) to NRC (W. Gamill), dated October 16, 1979.
- 11. Con-Ed letter (W. J. Cahill, Jr.) to NRC (A. Schwencer), dated April 28, 1980.
- 12. Con-Ed letter (P. Zarakas) to NRC (S. A. Varga), dated-August 1, 1980.
- 13. Con-Ed letter (J. D. O'Toole) to NRC (S. A. Varga), dated December 31, 1980.
- 14. Con-Ed letter (J. D. O'Toole) to NRC (S. A. Varga), dated April 14, 1981.
- 15. Con-Ed letter (J. D. O'Toole) to NRC (S. A. Varga), dated April 27, 1981.
- 16. Telecon, J. Toma, S. Mascell, NRC, D. Weber, EG&G Idaho, Inc.,
M. Scott and P. Szabados, Con Ed, August 27, 1981.
- 17. Final Safety Analysis Report (FSAR) for the Indian Point Nuclear n a-tion Unit 2.
- 18. General Design Criterion 17, " Electric Power Systems," of Appendix A,
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50,
" Domestic Licensing of Production and Utilization Facilities."
19.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
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20.
IEEE Standard 308-1974, " Standard Criteria for Class lE Power Systems for Nuclear Power Generating Stations."
- 21. ANSI C84.1-1977, " Voltage Ratings for Electric Power Systems and Equip-ment (60 Hz)."
- 22. IEEE Standard 141-1976,. IEEE Recommended Practice for Electric Power Distribution for Industrial Plants."
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