ML19347E912
| ML19347E912 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/30/1981 |
| From: | Cleveland C, Roberts E EG&G IDAHO, INC., EG&G, INC. |
| To: | Shemanski P Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6429 8105140225, EGG-EA-5374, NUDOCS 8105140255 | |
| Download: ML19347E912 (10) | |
Text
4 hEGsS.....e FORM EG44308
- Nv 1119)
INTERIM REPORT Accession No.
Report No.
EGG-EA-5374 Contract Program or Project
Title:
Selected Operating Reactor Issues Program (III)
'q Subject of ihls Document Degraded Grid Protection for Class lE Power Systems, Big Rock Point Nuclear Plant Type of Document:
Technical Evaluation Report Author (s):
C. J. Cleveland /E. W. Roberts
- 1C Researci anc~ecinica o ~ o' o ac - "':
April 1981 N Assistance Report RIsponsible 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.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE AC07 761D01570 NRC FIN No. A6429 INTERIM REPORT 1
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i DEGRADED GRID PROTECTION FOR CLASS 1E POWER SYSTEMS BIG ROCK POINT NUCLEAR PLANT Docket No. 50-MS IS$
J C. J. Cleveland /E. W. Roberts Reliability and Statistics Branch Engineering Analysis Division EG4G Idaho, Inc.
j April 1981 i
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4 TAC No.'10008
ABSTRACT 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
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systems. Consumers Power Company (CPC) sas to assess the suseptibility of me safety-related electrical equipment at the Big Rock Point Nuclear Plant to a sustained voltage degradation of the offsite source and interaction of the offsite and onsite emergency power systems.
This report contains an evaluation of CPC's analyses, modifications, and technical specification changes submitted in response to these NRC positions.
The evaluation has determined that CPC does not comply with all of the NRC positions.
FORWORD This report is supplied as part of the " Selected Operating Reactor Issues Program (III)" being conducted for the U.S. Regulatory Commission, Office of Nucleer Reactor Regulation, Division of Licensing, by EGSG Idaho, Inc., Reliability and Statistics Branch.
Tne U.S. Nuclear Regulatory Comission funded the work under the authorization, B&R 20 19 01 16, FIN No. A6429.
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CONTENTS
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1.0 I N TR OD UC T I O N.......................................................
I 2.0 D E SI GN BASE C RI T E RI A...............................................
1 3.0 EVALUATION.........................................................
2 3.1 Exi sti ng Undervol tage Protection..............................
2 3.2 Modifications.................................................
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3.3 Di s c u s s i o n.................................................... 2 4.0 CO NC L U S I ON S........................................................ 5 S.0 R E F E R E NC E S......................................................... 5 W
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TECHNICAL EVALUATION REPORT DEGRADED GRIIFPRUltLilDN PUR CLA55 lE POWER SYSTEMS BIG ROCK POINT NUCLEAR PLANT
1.0 INTRODUCTION
On June 3,1977, the NRC requested the Consumers rower Company (CPC) to assess the susceptibility of the safety-related electrical eqtipment at the Big Rock Point Nuclear Plant (BRP) to a sustained voltage degradation of the offsite) source and interaction of the offsite and onsite emergency
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power systems.
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, CPC was required to either propose modifica-tions to satisfy the positions and criteria or furnish an analysis to sub-stantiate that the existing facility design has equivalent capabilities.
By letter dated July 20, 1977, CPC acknowledged receipt of the NRC letter and stated that by early October 1977, an a pleted and a thorough response would be submitted.galysis would be com-4 On February 7,1978, CPC wrote the NRC explaining that, due to unforeseen manpower and equip-ment problems, the gesponse would be delayed until May 1978.3 By letter dated June 4,1978, CPC proposed certain design modifications and anal-yses in response to the June 1977 NRC letter.
On September 1, submitted a schedule of implementation of these modifications.5 On April 2,1979, upon ampletion of my initial review, several areas Nere in need of clarification by the licensee and a request for additional infor-mation was sent to the NRC.
By letter dated July 9,1979, the NRC reques-ted CPC to furnish the needed infomation.
By letter dated August 23, 1979, CPC stated that, because of other comittments, a response to th CPC responded to the request.p,1980 request for infomation would not be completed until January l By letter dated August 20, 1980, In October 1980, a second request for additional infomation was sent to the licensee by the NRg.
The licensee responded to the request by letter dated December 12, 1980.
These submittals contained the analyses and modiff-cations the NRC requested for second-level undervoltage (UV) protection.
By letter dated February 3 1981,9 CPC submitted a request for techni-cal specification changes as req,uested by the NRC letter of June 3,1977.
The NRC required that UV relay setpoint and time delay, with maximum and minimum allowable limits, surveillance requirements, and certain test requirement:: be included in the technical specification changes.
2.0 DESIGN BASE CRITERIA The design base criteria that were applied i.n determining the accept-ability of the system modifications to protect the safety-related equipment 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 Nuclear Power Plants," of 10 CFR 50.10 1
lear Power Generating Stations."gfer Protection Systems for Nuc-IEEE Standard 279-1971, "Criteri 2.
3.
IEEE Standard 308-1974 " Class lE Power Systems for Nuclear Power Generating Stations."I5 4.
dated June 3,1977.jetailed in a letter sent to the licensee, Staff positions as Systems and Equipment (60 Hz)."Ip Ratings for Electrical Power 5.
ANSI Standard C84.1-1977, " Volta 3.0 EVALUATION
)i This section provides, in Subsection 3.1, a brief description of the existing undervoltage protection at BRP; in Subsection 3.2, a description of the licensee's proposed modifications for the second-level undervoltage nrotection; and in Subsection 3.3, a discussion of how the proposed modifi-cations meet the design base criteria.
3.1 Existing Undervoltage Protection.
480V safet'y-related bus 2B has two UV relays set at less than or equal to 50%.
These relays are arranged in a two-out-of-twr c. incident logic scheme and are instantaneous.
Upon actuation, the dhesel generator is started. When the diesel generator reaches more than 31% voltage, an overvoltage (OV) relay in coincidence with the two UV relays trip the feed breakers to 28 and close the diesel-generator (DG) breaker.
The 28 safety-related bus is not load shed prior to closing tne DG breaker and, consequently, the bus is block loaded.
The UV relays do not annunciate; however, the feed breakers to the 2B bus are annunciated as they open.
3.2 Modifications.
The modification proposed by the licensee for second-level undervoltage protection will consist of three UV relays arranged in a three-out-of-three coincident logic scheme.
These relays will have a Jetpoint of 89% (+2%, -0%) and will monitor the bus voltage of the 2400V non-class 1E bus that feeds the 480V safety-related bus. These relays have a time delay of 0.5 s (+0.1 s) whose coincident signal is fed through a single time delay relay set at 10 s (+0.5 s).
This logic will then trip the feed breaker (1136) to the 2400Y bus.
This will, in turn, trip the UV relays on the 480V bus 28 and initiate the sequence of events as described above.
This plant does not load shed or sequence load safety-related bus 28.
The diesel generator is block loaded as it reaches at least 91% voltage.
Changes to the plant's technical specifications were also proposed by the licensee, adding a requirement to test and calibrate the new UV relays and adding a limiting condition of operation stating that any one of these relays may ba taken out of operation as long as the output from it is in the tripped condition.
l 3.3 Discusst on.
The first position of the NRC staff letter required that a second level of undervoltage protection for the onsite 2
power system be provided.
The letter stipuletes other criteria that the undervoltage protection must meet. Each criterion is restated below fol-lowed by a discussion reg.arding the licensee's compliance with that cri terion.
1.
"The selection of voltage and time s'etpoints shall be determined from an analysis of_ the voltage requirements of the safety-related loeds at all onsits system distributien levels."
The ifcensee's proposed setpoint of 89%.(-2, -0%) reflected down to the 480V bus 2B corresponds to a voltage of 402.5V to 412.1V.
This value was a 22,1980,grived at using the licensee's wbmittal of d
August and the worst case shown. This value was also arrived at using the voltage drops through the transfor-mers. The licensee did not state if these were full-load values as well as minimum values of the sources, however.
I find this setpoint as raflected to bus 2B to be too low as the licensee has stated that his MCCs are. rated for a low voltage of 408V.
This being the case, there is a possibility of one or more contactors not piping up.
k This setpoint reflected from the 2B bus to the 100 hp electric fire pump would be 396.5V to 406V using the supplied voltage drop of 1.5%.O As this is the worst case condition and the motors are qualified for 396V (440 + 10%), I find the setpoint accept-able at this level.
2.
"The voltage protection shall include coincidence logic to pre-clude spurious trips of the offsite power sources."
The proposed modification inconorates a three-out-of-three logic scheme, thereby satisfying this criterion.
3.
"The time delay selected shall be based on the following conditions:"
"The allowable time delay, including margin, shall not a..
exceed the maximum time delay' that is assumed in the FSAR accident analysis."
The proposed maximum time delay of the UV relays and time delay relay of 11.1 s does not exceed this maximum time delay.
b.
"The time delay shall minimize the effect of short-duration disturbances from reducing the unavailability of the offsita j
power source (s)."
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The licensee's proposed minimum time delay of 10 sec. is long enough to override any. short, inconsequential grid disturbances. Thq licensee has analyzed for this condition in his subaittal.9 1
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c.
"The allowable time duration of a degraded voltage condition at all distribution system voltaqe levels shall not result in failure of safety systems or components."
A review of the licensce's submittals reveals that the time delay in combination with the setpoint will not cause ther-mal damage to the safety-related motors.
4.
"The voltage monitors shall automatically initiate the discon-nection 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 cri-h L
terion is met.
5.
"The voltage monitors shall be designed to satisfy the require-ments of IEEE Standard 279-1971."
The licensee has stated in his proposal that the modifications are designed to meet or exceed IEEE Standard 279.4 However, upon review of his submittals and logic diagrams, I conclude that the modifications do not meet IEEE Standard 279.
The one time-delay relay in series with the second-level UV relays could be the cause. of a single failure incident negating the trip of the offsite source, thereby subjecting the safety-related bus to a degraded voltage that would cause thermal damage to the safety-related motors.
It was also the staff's intention that this relay scheme for second-level UV protection be a part of the class 1E power system and be designed as such.
Inasmuch as the licensee has proposed to install the relays on a non-class 1E bus, to trip a nonsafety-related breaker, with no provisior that these relays directly trip the safety-related bus feed breakers, I find this unacceptable.
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 protection monitors."
The limiting conditions for operation (LCO) proposed by the licensee does not meet the intent of this NRC position.
There is no time limit that a channel must be placed in the tripped post-tion when it is removed.
Furthennore, there is no requirement for channel functional tests called out in the techrical specifi-cations, only calibration tests once per operating cycle.
This could result in a failed relay negating a safety function if a degraded grid condition came about.
For this same reason, the surveillance requirements do not meet the intent of this NRC position.
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In addition, the licensee has failed to include proposed trip setpoints and allowable limits in the technical specifications.
The failure to include the second-level UV protection setpoints, time delays, and allowable limits in the technical s disagrees with the NRC criteria and is unacceptable.pecifications 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.
i I find this position does not apply to this plant as it does not use a load-shed scheme nor does it sequence on the safety-related loads.
a The third NRC staff position requires that certain test requirements be added to the technical specifications.
These tests were to demonstrate the full-funcCenal 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 is no adverse interaction between the onsite and offsite power sourtes.
The testing procedures used by the licensee at present adequately test the diesel generator as far as this position is concerned.
Since the plant does not load shed or sequence safety-related loads, the third NRC position is not applicable to Big Rock Point.
4.0 CONCLUSION
Based on the infonnation provided by CPC, I find that the proposed modifications do not comply fully with the criteria or the intent of the NRC in meeting position 1.
The licensee also fails to meet the NRC tech-nical specification requirements of position 1 (criteria 6) since the proposed surveillance requirements do not satisfy the NRC criteria and setpoints, allowable limits, and time delays are not included in the licensee's proposal.
I also find that staff positions 2 and 3 do not pertain to this plant.
5.0 REFERENCES
1.
NRC letter to CPC dated June 3,1977.
2.
CPC letter (D. A. Bixel) to NRC (Director) dated July 20, 1977.
3.
CPC letter (W. S. Skibitsky) to NRC (D. L. Ziemann) dated February 7, 1978.
4.
CPC letter (W. S. Skibitsky) to NRC (D. L. Ziemann) dated June 14, 1978.
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j9g8}etter (D. A. Bixel) to NRC (D. L. Ziemman) dated September 1, 5.
6.
CPC letter (D. A. Bixel) to NRC (D. L. Ziemann) dated August 23, 1979.
7.
CPC letter (D. P. Hoffman) to NRC (D. M. Crutchfield) dated August 22, 1980.
8.
CPC letter (D. P. Iloffman) to NRC (D. M. Crutchfield) dated December 12, 1980.
i 9.
CPC letter (D. P. Hoffman) to NRC (D. M. Crutchfield) dated February 3,1981.
10.
General Design Criterion 17, " Electric Power Systems," of Appendix A,
" General Design Criteria for Nuclear Power Plants," to 10 CFR 50,
" Domestic Licensing of Production and Utilization Facilities."
11.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
12.-
IEEE Standard 308-1974, " Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations."
13.- ANSI C84.1-1977, " Voltage Ratings for Electric Power Systems and Equipment (60 Hz)."
- 14. CPC letter (R. B. Sewell) to NRC (Director) dated December 7,1976.
- 15. Final Hazard Safety Report, Appendix A, Big Rock Point Plant Technical Speci fications.
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