ML18139B491

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Auxiliary Feedwater Sys Automatic Initiation & Flow Indication.
ML18139B491
Person / Time
Site: Surry Dominion icon.png
Issue date: 04/27/1981
From: Pandey S
FRANKLIN INSTITUTE
To:
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ML18139B489 List:
References
TER-C5257-277-3, NUDOCS 8108240335
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e TECHNICAL EVALUATION REPORT AUXILIARY FEEDWATER SYSTEM AUTOMATl'C INITIATION.AND FLOW INDICATION VIRGINIA ELECTRIC AND POWER COMPANY SURRY UNITS 1 AND 2 NRC DOCKET NO. 50-280 NRCTACNO. 12639 FAG PROJECTC5257 NRC CONTRACT NO. NRC-OJ:.79-118 FRC TASKS 2 77 31_5 Prepared by Franklin Research Center Author: S. Pandey-The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader: K, s. Fertner Prepared for Nuclear Regulatory Commission --,

Washington, D.C. 20555 Lead NRC Engineer: Rick Kendall

. April 27, 1981 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or Implied, or assumes any legal llablllty or responsibility for any third party's use, or the results of such use, of any Information, apparatus, produc~ cir process disclosed In this report, or represents that its use by such third party would not Infringe privately owned rights. *

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e e TER-C5257-277/315 CONTENTS Section Title l INTRO DUCT ION

  • l 1.1 Purpose of..Rev iew l 1.2 Generic Issue Background l 1.3 Plant-Specific Backgroun~. *l 2 REVIEW CRITERIA 3 3 TECHNICAL EVALUATION. 5 3.1 General Description of A.SW Syste~ 5 3.2 Automatic Initiation. 6 3.2.l Evaluation 6 3.2.2 Conclusion 8
3. 3 Flow Indication * - .._ .. 8 3.3.l Evaluation 8 3.3.2 Conclusion 9
3. 4 Steam C-nerator Level Indication Description. 9 4 CONCLUSIONS 11 5 REFERENCES 12 iii

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TER-C5257-277/315 I

1. INTRODUCTION 1.1 PURPOSE OF REVIEW The purpose of* this review is to provide a technical evaluation of the emergency feedwater system desig~ to verify that both safety-grade automatic initiation circuitry and flow indication are provided at Surry Units 1 and 2.

In addition, the steam generator level indication available at Surry is

  • described to assist subsequeot NRC staff review.

' 1.2 GENERIC ISSUE BACKGROUND A post-accident design review by the Nuclear Regulatory Cornmision (NRC) after the March 28, 1979 incident at Three Mile Island ('lMI) Unit 2 has established that the auxiliary feedwater (AEW) system should be treated as a safety system in a pressurized water reactor (PWR)- plant. The designs of

_ safety systems in a nuclear power plant are required to meet general design criteria (GDC) specified in Appendix A of the 10 CFR Part 50 [1] **

The relevant design er i ter ia for the AJ:W s:z*stem design are GDC 13, GDC 20, and GDC 34. GDC 13 sets forth the requirement for instrumentation to

..~niter variables and systems* (over their anticipated ranges of operation) that can affect reactor ~afety. GDC 20 requires that a protection system be designed to initiate automatically in order to assure that acceptable.fuel design limits are not exceeded as a result of anticipated operational occurrences. GDC 34 requires that the safety function of the designed system, that is, the residual heat removal by the Al-W system, be ~ccomplished even in the case of a single failure.

On September 13, *1979, the NRC issued a letter [2] to each PHR licensee that defined a set of short-term requirements specified in NUREG-0578 [3]. It required that the AJ:W system have automatic initiation and single failure-:

proof design consistent with the requirements of GDC 20 and GDC 34. In addition, auxiliary feedwater flow indication in the control room shall be*

provided to satisfy the requirements set forth in GDC 13.

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- TER~cs2s7-277/315 During the week of September 24, 1979, seminars were held* in four regions of the country to discuss the impact of the short-term requirements. On *.,

Cx:tober 30, 1979, another letter was issued to each F'r'iR licensee providing additional clarification of the NRC staff short-term requirements without altering their intent [4-].

Post-'Il1I analyses of primary system response to feedwater transients and reliability of installed AFW systems also .established that, in the long term, the AfW system should be UI?~ra,de.d in accordance with safety~rade require-ments. These long-term requirements were clarified in the letter of September

~. 1980 [SJ. 'Ihis letter incorporated in one document, NUREG-0737 [6], all

'!MI-related items approved by the commission for implementation at this time.

Section II. E. l. 2 of 1'.'UREG-07 37 clarifies the requirements for the AfW system automatic initiation and flow indication.

1. 3 PL.?..NT-SPECIFIC BACKGROUND

'Ihe Virg"inia ~ectr ic Power Company responded to NRC requirements thr_ough letters [7-9], with supporting documents and logic diagrams, describing the AFl'i systems at Surry Nuclear Power Plant Units l and 2 .

The FRC staff started a review of the AFW systems at Surry Nuclear Power Plant on September 10, 1980, based on the critiria described in Section 2 of this report. In a conference call among staff of the Licensee, FRC, and NRC on September 30, 1980, FRC requested more information, and the Licensee documented *the additional information requested in a letter to the NRC dated December 12, 1980 [10]".

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2. REVIEW CRITERIA To improve the reliability of the AFW system, the NRC required licensees to upgrade the system, where necessary, to ensure timely automatic initiation when required. The syst.em upgrade was to proceed in two phases. In the short term, as a minimum, controi grade signals and circuits were to be used to auto-matically initiate the AFW system. This control grade system was to meet the following requirements of NUREG-0578, Section 2.1.7.a [3]:
2. The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.
3. Testability of the initiating signals and circuits shall be a feature of the design.
4. The initiating signals and circuits shall be powered from the emergency buses.
5. Manual capability to initiate the auxiliary feedwater sys-tem from the control room ~hall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
6. The ac motor-driven pumps and valves in the auxiliary feed-water system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emer-gency buses. *
7. The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFW system from the control room.*

In the long term, these signals and circuits were to be upgraded in accor-dance with safety-grade requirements. Specifically, in addition to the above requirements, the.automatic initiation signals and circuits must have indepen-dent channels, use environmentally qualified components, have system bypa~sed/

inoperable status features, and confor~ to control system interaction criteria, as stipulated in IEEE Std 279-1971 [11].

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- e TER-CS257-277/315 The capability to ascertain the AEW system performance from the control room must also be provided. In the short term, steam generator level indica~ ,'

tion and flow measurement were to be used to assist the operator in maintaining the required steam generator level during AEW system operation. This system was to meet the following requirements from NUREG-0578, Section 2.1.7.b:

2. The auxiliary feedwa ter.
  • flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency po~er diversity requirements of the auxiliary feedwater system set forth in Auxiliary System Branch Technical Position 10-1 of the St'andard Review Plan, Section 10.4.9 [Ref. 12 in this report].*

The NRC staff has determined that, in the long term, the overall flowrate indication system for Westinghouse plants should include at least one au~iliary feedwater flowrate indicator for each steam gene~ator. The safety-grade flow-

.rate indication system must satisfy the single failure criterion, be environ-mentally qualified, have as a desig~ feature the capability to test the indi-cating channels, and conform to the control system interaction criteria, as stipulated in -IEEE Std 279-1971. --~

Tne operator relies on steam generator level instrumentation, in addition to auxiliary feedwater flow indication, to determine AEW system performance.

The requirements fo*r this steam generator level instrumentation are specified in Regulatory Guide 1.97, Revision 2, *rnstrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident* (13}.

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3. TECHNICAL EVALUATION 3 .1 GENERAL DESCRIPTION OF AfW SYSTD1 Surry Uni ts l and 2 are Westinghouse-designed "three loop* nuclear power generating plants. The AFW systems for the two units are essentially identical. For each uriit, the AFW system consists of one steam-driven pump and two motor-driven pumps.

The AFW system is not considered.an engineered safety feature (ESF) at this plant. The AFW system is d~scribed in Section 10 of the Surry Final Safety Analysis Report (FSAR), "Steam and Power Conversion_System," under the subhead-ing of the "Condensate and Feedwater Systems.* The AFW system is not identi-fied in FSAR Section 6, "Engineered Safeguards," and is not designed as an ESF system, particularly with respect to testability and bypass alarm functions.

The automatic initiation circuitry for the auxiliary feedwater pumps orig-instes in the engineered safeguards and reactor protection system, which is designed in accordance with IEEE Std 279-1971 [11].

Each of the three pumps discharge into two headers, aligned by manual valves. There are three lines from ;ach header, and each line contains a motor-operated valve (MOV) located inside containment. The lines join down-stream of the MOVs and fo~rn a common discharge *to supply each steam generator via the associated main feed line. In the event of failure of one heaoer, the supplies from the pumps may be isolated by manually operated valves to assure steam generator water flow from the other header. The MOVs required to estab-lish a flow path from the discharge of these pumps to the steam generators are left in the open position and also receive automatic signals. The discharge valves fail as-is. Steam generator level is controlled manually from either the main control room or the auxiliary shutdown panel by operating the appro-priate HOV in the auxiliary feedwater line.

Steam to the turbine-driven pump is obtained from each main steam line upstream of the main steam isolation valves and routed through a parallel combination of one motor and one air-operated valve. The air-operated valve.

is normally closed, and fails open~ the parallel MOV is normally clos~d and fails as-is.

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e TER-C5257-277/315 The AF~ system discharge lines of both units are cross-connected but are isolated by normally closed MOVs. Operator a.i=tion will permit the A.Yr'1 system**'

of one unit to supply water to the steam generators of the other unit. These HOVs receive automatic signals to open during certain steam line rupture conditions. They are powered from a vital bus and are controlled manually from the control room.

3.2 AUTOMATIC INITIATION 3.2.l EVALUATION The automatic initiation signals and circuitry for the AIW system*at Surry ~nits 1 and 2 comply with the general functional requirements of IEEE Std 279-1971. The following signals are used for automatic initiation of the AIW system:

A. Turbine-Driven Auxiliary Feed Pump

1. low-low steam generator level (2 out of 3)
2. undervoltage on the reactor coolant pump (RCP) buses (2 out of 3)

B. Motor-Driven Auxiliary Feed Pumps

1. low-low level from any one steam generator
2. loss of reserve station power {station blackout)
3. trip of both main feedwater pumps
4. safety injection.

In addition, both the turbine- and motor-driven pumps can be manually activated.

The automatic initiation signals and circuits for the A.r-W system at Surry Units 1 and 2 comply with the single failure criterion of IEEE Std 279-1971.

The scope of the single failure analysis in the present task is limited to redundancy of power feed configuration, diversity of actuating signals, and independent and redundant automatic initiation circuits.

The steam generator level and undervoltage initiation signals (on transfer buses - station blackout) are redundant and independent. The input signals from the safety injection system (SIS) and the main feed pump breaker signals are both redundant and independent. Undervoltage signals on the RCP are not redundant.

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  • The AFW system and components are tested in accordance with technical specification requirements. The sensors used for the steam generator water . ,

level indicators and the 4-V bus loss of voltage and undervoltage detectors are tested each shift for operational availability by cross-checking between channel. The channel functional tests for logic trip circuits and trip setpoints are performed once a month. The calibration of the channels is performed during refueling.

The AFW system is tes~:d ~v~ry refueling outage for initiation by safety injection and loss of offsite power. The auxiliary feedwater pumps start between 50 and 60 seconds after a signal is given. In addition, the initiation of auxiliary feedwater is tested for actuation on a low-low steam generator level on any one steam generator. The low-low steam generator level tr.ip is tested monthly up to the relay that starts the pump. Pump control is placed in pull-to-lock position so the pumps do not start; therefore; relay operation cannot be verified. The relay are tested.at every refueling.

Since the AFW system.is not considered an ESF system, there is no direct indication that the pump has been placed in the pull-to-lock position. The Licensee has stated that th automatic initiation due to loss of both main feedwater pumps has not been tested~-~but a test procedure will be written *

. The determination of adequate environmental qualification of the circuits and components is being *reviewed separately by NRC staff and is beyond the scope of present FRC work. No modifications have been proposed that would result in interaction between control and safety circuits of the Surry Units l and 2 ~-W system.

The operating brpasses associated with the automatic initiation logic circuitry (including sensors used for auto~atic initiation) during start-up or operation of the reactor are as follows:

1. The steam generator low-low level (SGLLL) initiation circuitry is always active and can be bypassed by placing a particular channel in the test position. This action is restricted by the Technical Specifications. Since these channels are always active, a bypass removal mechanism is not needed.
2. Safety injection (SI) initiation circuitry is provided with a bypass for startup purposes and is separately alarmed in the control room. This bypass (block) is. automatically unblocked and requires no operator action.

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3. Automatic initiation of t h e ~ system due to los~ of the main feedwater pumps is bypassed during start-up by placing two of the.

four pumps' breakers in the test position. This prevents AFW sys~em initiati6n until th~ breakers aie procedurally taken out of the test position when the main feed system is placed in operation.

4. The auxiliary feedwater pumps may be prevented from starting by placing the 'pump controls in the "pull-to-lock* position. The auxiliary feedwatet control is procedurally returned to the *auto*

position when the main feed system is placed in operation.

5. Reactor coolant syst_em loop isolation valves provide a signal, when closed, which prevents automatic start of auxiliary feedwater pumps from a steam generator low-low water level signal in the affected loop. This signal is automatically reinstated upon reopening of the valves. In the event this block is initiated, a permissive status light is lit in the control room to alert the operator of the condition. This is, however, not considered an operating bypass .

since the plant operation is restricted to three-loop operation and at no time would it be operated with a loop isolated.

6. The reactor coolant pump (RCP) undervoltage *channels which sense the voltage on the station service buses A, B, and Care not provided with bypass capability during start-up or operation.

No bypass capability is provided for the station blackout signal, which senses the voltage on the station transfer buses.

3.

2.2 CONCLUSION

The initiation signals, logic, and associated circuitry of the Surry Units land 2 AFW system automatic initiation system comply with the long-term safety-grade requirements of NUREG-0578, Section 2.1.7.a and the subsequent.

clarification issued by the NRC, except that during the monthly testing of the low-low steam generator level trip, the auxiliary feedwater pumps are locked out. This is equivalent to bypassing the automatic initiation circuitry.

This action is not annunciated in the control room.

3.3 FLOW INDICATION 3.3.l EVALUATION The capability to ascertain the performance of the AFW system at Surry Units land 2 is provided by flow indication. The flow indication system con-

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e .e TER-C~257-277/315 sists of individual sensors, current loop circuitry, and individual meters in the control room for each of the three auxiliary feedwater lines, one to each' steam generator. The flow channels are powered from redundant 120-V ac vital power buses.

The AFW system flow indication by itself does not satisfy the single failure criterion; however, e~ch flow channel is backed up by steam generator level indicators. The flow channels have no control function, and no interac-tion between protection and concrol circuits is noted. The total accuracy of the flow channels is within+ 10%.

The determination of environmental qualification of the auxiliary feedwater flow channels will be performed separately by NRC staff and is beyond the scope of the present FRC task.

The Licensee has stated that the auxiliary feedwater flow indication to each steam generator that is displayed in the control room is safety-grade

  • equipment. The transmitters, however, are not pre~ently safety-grade.

The testability of all the channels is a feature of the design, but the test frequency has not been specified by the Licensee. The* Licensee states that periodic tests to cover the auxiliary feedwater indication system are currently being written.

3.

3.2 CONCLUSION

The auxiliary feedwater flow instrumentation at. Surry Units 1 and 2 complies with the long-term safety-grade requirement of NUREG-0578, Section 2.l. 7 .b and the subsequent clarification issued by the NRC, except for the control*-grade flow transmitters.

3. 4 STEAM GENERATOR LEVEL INDICATION DESCRIPTION Each of the three steam generators has three narrow-range level transmit-ters and one wide-range level transmitter. All four read in 0-100 percent of range. The narrow-range* transmitters_ read from just below the U-bends of the tube bundle up to the steam dryer section. The wide-range channels read from

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- - TER~C5257-277/315 above the tube sheet up to the steam dryer section above the tube bundle. The level transmitters and their power sources are as follows: .,

Steam C-enerator "A" Wide-Range* Level Transmitter LT-14 77 - Vital Bus II Steam C-enerator "B" Wide-Range Level Transmitter LT-1487 - Vital Bus III Steam Generator "c* Wide-Ra.nge Level Transmitter LT-1497 - Vital Bus IV Steam Generator "A" Narrow-Range Level*

Transmitter LT-1474 Vital Bus I LT-1484 Vital Bus II LT-1494 Vital Bus III Steam C-enerator "B" Narrow-Range Level Transmitter LT-1475 - Vital Bus I LT-1485 Vital Bus II LT-1495 - Vital Bus III

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Steam Generator "C" Narrow-Range Level Transmitter LT-1476 - Vital Bus I LT-1486 - Vital Bus II LT-1496 - Vital -Bus III The narrow-range level channels are tested by channel checks each shift and channel functional tests every 30 days. Each 2 out of 3 logic for start of the motor-driven auxiliary feedwater pumps is tested monthly with the pump breakers in the pull-to-lock position.

Each narrow-range indicator channel is proviqed with an edgewise analog indicator in the control room. The three wide-range channels are indicated on a single strip-chart recorder in the control room. Also, the operator has

"'on-demand" ~*apability to call up any of these narrow- or wide-range channels on the plant computer.

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4. CONCLUSIONS The initiation signals, logic, and associated circuitry of the Surry Units land 2 AFW system automatic initiation system comply with the long-term
  • safety-grad~ reqU:iremen.ts of NUREG-QS78, Section 2.1. 7 .a and the subsequent clarification iss*ued by the NRC, with the exception that, during the monthly testing of the low-low steam generator level trip, the auxiliary feedwater pumps are locked out. This is equivalent to bypassing the automatic initiation circuitry. This pul1.-to-lock position is not.annunciated in the control room.

The auxiliary feedwater flow instrumentation at Surry Units land 2 complies with the long-term safety-9rade requirements of NUREG-0578, Section

~~1.7.b and the subsequen~ clarification issued by the NRC, except for the control-grade flow transmitters.

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s. REfERENCES
1. Code of Federal Regulations, Title 10, Office of the Federal Register, National Archives and Records Service, General Services Administration, Revised January 1, 1980.
2. NRC generic lett*er to all PWR licensees regarding short-term requirements resulting from Three Mile Island Accident, September 13, 1979.
3. NUR.EG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommenea~ions,~ USNRC, July 1979 *.
4. NRC generic letter to all PWR licensees clarifying lessons learned short-term requirements, October 3.0, 1979.
5. NRC generic letter to all PWR licensees regarding short-term requirement resulting from Three Mile Island Accident, September 5, 1980 *

. 6. NUR.EG-0737, "Clarification of 'n1I Action.Plan Requirements," USNRC, November 1980.

7. Virginia Electric and *Power Company, Letter to H. R. Denton (NRC),

October 24, 1979.

8. Virginia Electric and Powe~_~ompany, Letter to H. R. Denton (NRC),

February l, 1980.

9. Virginia Electric and Power Company, Le,tter to H. R. Denton (NRC), '

July 7, 1980.

10. Virginia Electric and Power ~ompan*y, Letter to H. R. Denton (NRC) ,

December 12, 1980.

11. IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Inc., New York, New York.
12. NUR.EG-75/087, Standard Review Plan, Section 10.4.9, Rev. 1, USNRC, no date.
13. Regulatory Guide 1.97 (Task RS 917-4}, "Instrumentation for Light-Water- Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,~ Rev. 2, USNRC, December 1980.
14. IEEE.Std 323-1974, Qualifying Class lE Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Inc., New York, New York.

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