ML20031F056
| ML20031F056 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/31/1981 |
| From: | Weber D EG&G, INC. |
| To: | Shemanski P Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6429 EGG-EA-5476, EGG-EA-5476-01, EGG-EA-5476-1, NUDOCS 8110190107 | |
| Download: ML20031F056 (11) | |
Text
_ _.
[0 -A / f 1
l EGG-EA-5476 AUGUST 1981 f0k DEGRADED GRID PROTECTION FOR CLASS lE POWER SYSTEMS, ggg 0YSTER CREEK NUCLEAR POWER STATION UNIT 1, alt /8 h/SI&
Docket No. 50-219 g
e l
I
'g f7g hgs l
D. A. weber 007 u
Q7C 7esearch and TechnicaWY~$W 3"t30Ce Report 5
m U.S. Departrr. nt of Energy Idaho Operations Office = idaho National Engineering Labo.atory
-v f
\\
~,
x
,M.M Mmummq
. -I
~.,
. Of,,
' 888Mut Mammuun v]
- n? >
n9
'9 r-g _g_ __: g, gg+,
. ), #
g,_
g r7- -M gy f-? eb m.
This is an informal report intended for use as a preliminary or working document a
Prepared for the U.S. Nuclear Regulatory Comission U
E 6 E 6 ldaho Under DOE Contract No. DE-AC07-76ID01570 Q
FIN No. A6429 8110190107 810831 PDR RES
.o P DR
E idaho inc
=-
INTERIM REPORT Accession No.
Report No. EGG-EA-5476 e
Contract Program or Project
Title:
Selected Operating Reactor Issues Program (III)
Subject of this Document:
Degraded Grid Protection for Class lE Power Systems, Oyster Creek Nuclear Power Station Unit 1, Docket No. 50-219
[
Type of Document:
Technical Evaluation Report Author (s):
D. A. Weber YlC lesearci anc Tecinica' Drte of Document:
A nssistance Report August 1981 Responsible NRC Individual and NRC Office or Division:
P. C. Shemanski, Division of Licensing This document was prepared primanly 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
]
0430J J
DEGRADED GRID PROTECTION FOR CLASS lE POWER SYSTEMS OYSTER CREEK NUCLEAR STATION UNIT 1 Docket No. 50-219 D. A. Weber Reliability and Statistics Branch
~
Engineering Analysis Division EG&G Idaho, Inc.
August 1981 NRC Researci anc Tec1nica Assistance Repor':
r-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 pow 3r systems. Jersey Central Power & Light Company (JCP&L) was to assess the susceptibility of the safety-related electrical equipment at the Oytt?.r Creek Nuclear Station, Unit 1, to a sustained voltage degradation of tne offsite source and interaction of the offsite and onsite emergency power systems. - This report contains an evaluation of JCP&L's analysis, modifica-tions, and technical specification changes to. comply with these NRC posi-tions. The evaluation has determined that JCP&L does not comply with one of the NRC positions.
FOREWORD This report is supplied as part of the " Selected Operating Reactor Issues Program (III)" being conducted for the U.S. Nuclear Regulatory Com-mission, 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.
G 9
ii
r s
CONTENTS
!.0 INTRODUCTION....................................................
I 2.0 DESIGN BASE CRITERIA............................................
1 3.0 tVALUATION......................................................
1 3.1 Existing Und_rvoltage Protection..........................
2 3.2 Modifications.............................................
2 3.3 Discussion................................................
2
4.0 CONCLUSION
S.....................................................
5
5.0 REFERENCES
5 4
9 k
iii
TECHNICAL EVALUATION REPORT DEGRADED GRID PROTECTION FOR CLASS 1E POWER SYSTEMS OYSTER CREEK NUCLEAR STATION UNIT 1
1.0 INTRODUCTION
On June 2, 1977,I the NRC requested the Jersey Central Power & Light Company (JCP&L) to assess the susceptibility of the safety-related electri-cal equipment at the Oyster Creek Nuclear S:ation to a sustained voltage degradation of the offsi emergency power systems.je source and interaction of the offsite and onsite The letter contained three positions with which the current design of the plant was to be compared. Af ter comparing the current design to the staff positions, JCP&L was required to either propose modifications to satisfy the positions and criteria or furnish an-analysis to substantiate that the existing facility design has eq ivalent capabilities.
l JCP&L responded to the NRC letter of June 2, 1977 with a submittal i
dated September 25, 1979.2 This submittal and submittals of September 16, 1976,3 October 14 1976,4 November 5, 1976,5 February 1 1977,6 April 18,1977,7,ugust lg, April 30 l
1,1979,9,anuary 18, A
1977,8, November J
1980,10 August 11, 1980,1 1981 12 and the Oyster Creek Final Safety Analysis Report (FSAR)f3 comp,lete the information reviewed for this report.
2.0 DESIGN BASE CRITERIA e
The design base criteria that were applied in determining the accepta-bility of the system modifications to protect the safety-related equipment from a sustained degradation of the offsite grid are:
1.
Gereral Design Criterion 17 (GDC 17), " Electrical Power Systems," of Appendix A, " General De Nuclear Power Plants," of 10 CFR 50 gign Criteria for 1
2.
IEEE Standard 279-1971, " Criteria for forNuclearPowerGeneratingStations"gotectionSystems 3.
IEEE Standard 308-1974, " Class lE Nuclear Power Generating Stations"gwer Systems for 4.
Staffoositionsasdetailedjnalettersenttothe licenoee, dated June 2, 1977 5.
ANSIStandardC84.1-1977,"VoltageRatinggforElectri-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 Oyster Creek; in Subsection 3.2, a 1
description of the licensee's proposed modifications for the second-level undervoltage protection; and in Subsection 3.3, a discussion of how the proposed modifications meet the design base criteria.
3.1 Existing Undervoltage Protection.
For loss-of-voltage protection, each of the safety-related 4160V buses 1C and 1D has a set of General Elec-tric type IAV53K under/overvoltage indication relays. The undervoltage trip setpoint for each relay is 68.8% (2864V).
Each relay will operate in 3 seconds on total loss of po er. The 68.8% on the 4160V buses will result in voltage of 317 (66%) and 297 (51.8%) for the 480V substations and motor control centers (MCC's), respectively. Operation of either relay will initiate isolation of the 4160V buses and loads, initiate load-shedding and start of the emergency diesel generators (DC's), energize the emergency buses with permanently connected loads and energize the automatically con-nected emergency loads through a load segaencer.
3.2 Modifications. As a result of the NRC request, JCP&L has installed a second-level undervoltage scheme to protect safety-related equipment from a sustained degraded grid. The schene consists of the addi-tion of independent undervoltage relays for buses 1C and 10. The three relays on each bus are connected in a two-out-of-three coincident logic, with a setpoint of 3671V +1% (36.7V) and a time delay of 10 seconds +1%
(0.1 sec). Either bus reTay logic will initiate disconnection of the off-site power source whenever the voltage setpoint and time limits have been exceeded. With the offsite power disconnected, the existing loss-of-voltage
~
relays on the emergency buses will operate as described in Section 3.1.
The licensee has proposed changes to the plant's technical specifica-tions including:
relay surveillance requirements, setpoints and limits, and limiting conditions for operation.
3.3 Discussion. The first position of the NRC staff letterI 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 fol-lowed by a discussion regarding the licensee's compliance,with that criterion.
1.
"The selection of voltage and time setpoints shall be determined from an analysis c,f the voltage reqairements of the safety-related loads at all onsite system distri-buti)n levels."
The licensee's proposed setpoint of 3671V (88.5% of 4160V) results in voltages of 88.5% at the 460V rated motor starters. The motor starters will pickup at 85%
voltage and the control circuitry can withstands a lower voltage.
This setpoint allows worst case terminal volt-ages of 91.6%, 85%, 87.5% and 90.5% for the correspond-ing safety-related 4000V, 480V, 460V, and 440 motors.
The minimum rating is 99% for the 4000V motor, and 86.6% for the worst cast 480V, 460V, and 440V motors 2
L (which consider a 1.15 service factor). At the pro-posed setpoint all 4000V, 460V, and 440V safety-related equipment will operate at voltages above the minimum required. However the setpoint allows the 480V motors and some 460V motor staters to be operated continuously at voltages below thefv ainimum rating. Therefore the proposed setpoint is not satisfactory.
The licenses subraittal of April 30, 1981 12 points out that the analysis does not consider the automatic operation of newly installed voltage regulators which will maintain the 4160V bus at 4100V when the grid is at its minimum analyzed valve. However, credit cannot be given for the regulators since they have a limited voltage regulation (+10%) and there are no Technical Specifications Limiting Conditions for Operation (LCOs) regarding plant operation should the regulators be bypassed or out of operation.
2.
"The voltage protection shall include coincident logic to preclude spurious trips of the offsite power sources."
The proposed modification incorporates a two-out-of-three coincident logic scheme, thereby satisfying this
~
criterion.
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 10 seconds
(+0.1 seconds) does not exceed this maximum time j
delay.2 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 10 seconds is long enough to override any short, inconsequential grid disturbances and voltage dips caused from 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."
3
7 A review of the licensee's voltage analysis indi-cates that the time delay will not cause any fail-ures of the safety-related equipment.2 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 lic:nsee's submittals confirms 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.
6.
"The technical specifications shall include limiting conditions for operation (LCOs), surveillance require-ments, trip setpoints with minimum and maximum limits, and allowable values for the second-level voltage pro-tection monitors."
The licensee's proposal for technical specification changes' includes all the required items for the second-level protection monitors. However, there are no LCOs governing plant operations should the regulators be bypassed or out of service.
The second NRC staff position requires that the system design auto-matically 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 the emergency buses already has these features incorporated. The second-level undervoltage protection will be blocked automatically when the emergency buses 7.re_being fed from the onsite sources.
The third NRC staff position requires that certain test requirements be 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 bases, and that there is no adverse interaction between the onsite and offsite power sources.
4
Thetestingprocedugsproposedbythelicenseecomplywiththefull intent of this position.
Load-sheoding on offsite power trip is. tested.
Load-sequencing, once the diesel generator is supplying the safety buses, is tested. The time duration of the tests (equal to or greater than 5 min-utes) will verify that the time delay is sufficient to avoid spurious trips and that the-load-shed bypass circuit is functioning properly.
4.0 CONCLUSION
S Based on the information provided by JCP&L, it has been determined that the installed modifications do not comply with NRC staff position 1.
Certain 480V motors may operate at voltages below their minimum ratings at the present second-level undervoltage relay setpoint, when the offsite grid is at its minimum analyzed valve. Credit cannot be given for the installed voltage regulators as the regulators provide limited regulation (+10%) and there are no LCOs governing plant operations should the regulators be bypassed,r out of service.
The existing load-shed circuitry complies with staff position 2 and will prevent adverse interaction of the offsite and onsite emergency power systems.
The proposed changes to the technical specifications adequately test the system modifications and comply with staff position 3.
The surveillance requirements, limiting conditions for operation, minimum and maximum limits for the trip setpoint, and allowable values meet the intent of staff posi-tion 1.
It is therefore concluded that the setpoint of the installed secord-e level undervoltage relays is not acceptable. The proposed changes to the technical specifications are acceptable, except for the seccnd-level under-voltage relay setpoint.
5.0 REFERENCES
1.
NRC letter (R. W. Reid) to JCP&L, dated June 2, 1977.
2.
JCP&L letter (I. R. Finfrock) to the Director, Nuclear Reactor Regu-lation, dated September 25, 1979.
3.
JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated September 16, 1976.
4.
JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated October 14, 1976.
5.
JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated November 5, 1976.
5
6.
JCP&L' letter (I. R. Finfrock) to.c. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated February 1,-1977.
7.
JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated April 18, 1977.
6 8.
JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating Reactors Branch No. 3, Division of Reactor Licensing, dated August 15, 1971.
9.
JCP&L letter (I. R. Finfrock) to the Director of Nuclear Reactor Regu-
-lation, dated November 1, 1979.
- 10. JCP&L letter (I. R. Finfrock) to the Director of Nuclear Reactor Regu-lation, dated 'anuary 18, 1980.
- 11. JCP&L letter (I. R. Finfrock) to the Director of Nuclear Reactor Regu-lation, dated August 11, 1980.
- 12. JCP&L letter (I. R. Finfrock) to the Director of Nuclear Reactor Regu-lation, dated April 30, 1981.
- 13. Final Safety Analysis Report (FSAR) for the Oyster Creek Nuclear Station.
- 14. General Design Criterion 17, " Electric Power Systems," of Appendix A, i
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, e
" Domestic Li:ensing of Production and Utilization Facilities."
15.
IEEE Standa"d 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
16.
IEEE Standard-308-1974, " Standard Criteria for Class lE Power Systems for Nuclear Power Generating Stations."
- 17. ' ANSI C84.1-1977, " Voltage Ratings for Electric Pcwer Systems and Equip-ment (60 Hz)."
18.
IEEE Standard 141-1976, "IEEE Recommended Practice for Electric Power Distribution for Industrial Plants."
d 1
6
_ _ - -