ML19348A054

Degraded Grid Protection for Class IE Power Sys,Oyster Creek Nuclear Power Station,Unit 1.

ML19348A054
Person / Time
Site:	Oyster Creek
Issue date:	08/31/1981
From:	Weber D EG&G, INC.
To:	Shemanski P Office of Nuclear Reactor Regulation
Shared Package
ML19348A053	List: ML19348A054
References
CON-FIN-A-6429 EGG-EA-5476, EGG-EA-5476-02, EGG-EA-5476-2, NUDOCS 8111020600
Download: ML19348A054 (10)

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INTERIM REPORT Accession No.

~ Report No. EGr;-FA-5476 Contract Program or Proje:t

Title:

Selected Operating Reactor Issues Program (III)

Subjec1of this Document

• Degraded Grid Protection for Class lE Power Systems, Oyster Creek i Nuclear Power Station Unit 1. Docket No. 50-219 -

Type of, Document:

Technical Evaluation Report "

Author (s):

D. A. Weber D:t]cf Document: ~

August 1981 Responsible NRC Individual and NRC Office or Division:

P. C. Shemanski, Division of Licensing This document was prepared primarily for preliminary orinternal use. it has not received full review and approval. Since there may be substantive 2.hanges, this document should not be considered final.

. . EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 i

' Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

Under DOE Contract No. DE-AC07-76 f D01570 NRC FIN No. A6429

' INTERIM REPORT 8111020600 810831 "f f i PDR RES -

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1 DEGRADEDGRIDPROTECTION'FORCLASS1EPOWERSYSTENS OYSTER CREEK NUCLEAR STATION UNIT 1 Docket No. 50-219

' D. A. Weber -

Reliability and Statistics Bra ~nch -

Engineering Analysis Division  :

EG&G Idaho, Inc. .

August 1981 TAC No. 10038 O

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I 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 systems. Jersey Central Power & Light Company (JCP&L-) was to assess the susceptibility of the safety-related electrical equipment at the Oyster ,

Creek Nuclear Station, Unit 1, to a sustained voltage degradat#qn of the offsite source and interaction of the offsite and onsite emergency power

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systems. This report contains an evaluation of JCP&L's analysis, modifica-tions, and technical specification changes to complv 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.

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CONTENTS ,

1.0 INTRODUCTION

.................................................... 1 2.0 DESIGN BASE CRITERIA ....'.'....................................... 1 3.0 EVALUATION ...................................................... 1 3.1 Exi sting Undervoltage P rotection . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2 Modifications ............................................. 2 3.3 Discus 31)n ................................................ 2 l 4.0 CONCLUSI*.S ..................................................... 5 5.0 REFERE'CES...................................................... 5 1

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i TECHNICAL EVALUATION REPORT .

, DEGRADED GRID PROTECTION FOR CLASS 1E POWER SYSTEMS OYSTER CREEK NUCLEAR STATION UNIT 1

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1.0 INTRODUCTION

On June 2, 1977,1 the NRC requested the Jersey Central Power & Light Company (JCP&L) to assess the susceptibility of the safety-'related electri-cal equipment et the Oyster Creek Nuclear Station to a sustained voltage 4

degradation of the offsite source and interaction of the offsite and onsite f I

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, JCP&L was required to either 3 propose modifications to satisfy the positions and criteria or furnish an # '

i analysis to substantiate that the existing facility design has

• equivalent 1

capabilities.

JCP&L responded to the NR,C letter of June 2,1977 with a submittal dated September 25, 1979 2 This submittal pnd submittals of Sgptember 16, 1976,3 October 14 1976,4 November 5, 1976,3 February 1 1977 April 18,1977,7 ,ugust A 15,1977,8, November 1,1979,9 J,anuary,0 18, 1980,10 August 11, 1980,1 i April 30 1981,12 and the Oyster Creek .

Final Safety Analysis Report (FSAR)I3 coinplete the information reviewed .

for this report.

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. 2.0 DESIGN BASE CRITERIA

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The design base criteria that were applied in determining the accepta-bility of the system modifications to protect the safety-related equipment l

from a sustained degradation of the offsite grid are:

1. General Design Criterion 17 (GDC 17), " Electrical Power Systems," of Appendix A, " General Dg Nuclear Power Plants," of 10 CFR 50'gign Criteria for i

2. IEEE Standard 279-1971, " Criteria for rotection Systems

- - for Nuclear Power Generating Stations"{D .

3. IEEE Standard 308-1974, " Class 1E t Nuclear Power Generating Stations"gwer Systems for l 4. Staff positions as detailed ; n a letter sent to the licensee, dated J:::s 2,1977-

5. ANSI Standard C84.1-1977, " Voltage Ratingfor Electri-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 Prosection. For loss-of-voltage protection, each of the safety-related 4160V buses 1C and 10 has a set of General Elec- 3 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 power. The 68.8% on the 4160V buses will result in voltage of 317 (66%) and 297 (61.8%) 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 (DG's), energize the emergency buses with permanently connected loads and energize the automatically con-nected emergency loads through a load sequencer. ,

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 scheme 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 letterl required that a second level of undervoltage protection for the onsite power system be provided. The letter stipulates other criteria that the uncervoltage protection must meet. Each criterion is restated below fol-lowed by a discussior regarding the licensee's compliance with that j c r.i ter i on.

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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 distri- ,

bution levels."

The licensee's proposed setpoint of 3671V (88.5% of l

4160V) results in voltages of 88.5% at the 460V rated mctor starters. The motor starters will pickup at 85%

! voltage and the control circuitry can withstand a lower voltage. This setpoint allows worst case terminal volt- e 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 90% for the 4000V motor, and

• 86.6% for the worst case 480V, 460V, and 440V motors l

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9 i (which c nsi6er a 1.15 s rvics facter). At the. pro- '

posed setpoirt all 4000V, 460V, and 440V safety-related equipment wili' operate at voltages above the minimum required. However the setpoint allows the 480V motors and some 460v motor starters to be operated continuously at voltages below their minimum rating. Therefore the proposed setpoint is not satisfactory. -

The licenses submittal 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

+10%) and there are no Technical voltage regulation Specifications Limit(ina Conditions for Operation (LCOs)  :

regarding plant operation should the regulators be .

bypassed or out of operation.

2. "The voltage protection sh'all 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 follcw-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 delay.geconds) does not exceed this maximum time

- - b. "The time delay shall minimize the effect of short-duration disturDances 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.-

incon:equential grid disturbances and voltage dips caused from the starting of large motors.

c. "The allowable time duration of a degraded voltage condition at ali 'istribution system levels shall not result, in failure of safety systems or compon-ents."

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A review of the licensee's voltage analysis indi-cates that the time delay will not cayse any fail-ures of the safety-related equipment.4

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'4. "The voltage monitors shall automatically initiate the  ;

disconnection of offsite power sources whenever the voltage setpoint and time-delay limits have been T l exceeded /

A review of the licensee's submittals confirms that this >

criterion is p -

5. The voltage monitors shall be designed to sati>fy the  !

requirements of IEEE Standard 279-1971."

I The licensee has stated in his proposal that the modi- .

fications are designed to meet or exceed IEEE Stan-dard 279.

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6. "The technical specifications shall include limiting conditions for operation (LCOs), surveillance require-

ments, trip setpoints with minimum and maximum limits, 1 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 are being fed from the onsite sources. ,

l 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- t

- sequent reconnection of onsite power sources. These tests verify the proper g 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.

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I The testing procedu intent of this position.qqs proposed by the licensee comply with the full Load-shedding on offsite power trip is. tested.

Load-sequencing, once the diesel generator is s~upplying 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 r there are no LCOs governing plant operations should the regulators be bypassed or out of ser~vice.

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 second-level undervoltage relays is not acceptsole. The proposed changes to the i technical specifications are acceptable, except for the second-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 l 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.

c 5. JCP81 letter (I. R. Finfrock) to Mr. George Lear, Chief, Operating l Reactors Branch No. 3 Division of Reactor Licensing, dated ~

l November 5, 1976.

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6. JCP&L letter (I. R. Finfrock) to Mr. George Lear, Chief,' Operating .

Reactors Branch No. 3. Division of Reactor Licensing, dated February 1, 1977.

7. JCP&L letter (I. R. Finfrock) t'o Mr. George Lear, Chief, Operating Reactors Branch No. 3. Division of Reactor Licensing, dated April 18, ,

1977.

'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 January 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 ,'

Stati.on. ,,

14. General Design Criterion 17. " Electric Power Systems," of Appe'ndix A.

" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, ,

" Domestic Licensing of Production and Utilization Facilities."

15. IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."

16. IEEE Standard 308-1974, " Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations."

17. ANSI C84.1-1977, " Voltage Ratings for Electric Power Systems and Equip-ment (60 Hz)."

18. IEEE Standard 141-1976, "IEEE Recommended Practice for Electric Power Distribution for Industrial Plants." '

'19. NEMA Standard, NEMA MGl-1972, " Motors and Generators."

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