ML20246J319

Evaluation of Operator Roles in Mitigating High Energy Line Break at Fort St Vrain Nuclear Generating Station

ML20246J319
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/31/1987
From:	Baxter D, Bruske S, Valenti L EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20246J287	List: ML20246J279 ML20246J313 ML20246J319
References
CON-FIN-D-6037 EGG-NTA-7794, TAC-61894, NUDOCS 8907170335
Download: ML20246J319 (16)

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9 EVALUATION OF OPERATOR ROLES IN MITIGATIfG HIGH ENERGY LINE BREAKS (HELBs) AT THE t FORT ST. VRAIN NUCLEAR GENERATING STATION f

D. E. Baxter S. J. Bruske L. N. Valenti Published July 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the '

U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. D6037

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. i ABSTRACT l

This document was prepared for the Nuclear Regulatory Commission to l ' assist them in evaluating the operator roles required in the mitigation of l High Energy Line Breaks (HELBs) at the Fort St. Vrain Nuclear Generating l Station (FSV). A comprehensive review of the utility submittals related to this issue was performed. The reviewers determined that the licensee's responses were consistent with current NRC Guidelines.

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Docket No. 50-267 TAC No. 61894 ii R;

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FOREWORD This report is supplied as part of the Operating Reactors Electrical, Instrumentation and Control System reviews being. conducted for the U.S.

Nuclear Regulatory Comission (NRC), Office of Nuclear Reactor Regulation (NRR) by EG&G Idaho', Inc.

The U.S. Nuclear Regulatory Comission funded the work under

- authorization B&R 20 19 10 11 2, FIN No. D6037, Project-9, " Role of Operators in Mitigating High Energy Line Breaks at Fort St. Vrain."

4 Docket No. 50-267 TAC No. 61894 j iii

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.,ji CONTENTS.

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,,- ABSTRACT .............................................................. .ii l~

4 FOREkORD .............................................................. i ii

. 1. INTRODUCTION....................................................... 1

- 2. BACKGROUND ....................................................... 2

.3. QUESTIONS / RESPONSES .............................................. 3

'A. ' EVALUATION ........................................................

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5. CONCLUSION'.................................... .................. 10-

6. REFERENCES ....................................................... 11' We iv

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EVALUATION OF OPERATOR ROLES IN MITIGATING HIGH ENERGY LINE BREAKS (HELBs) AT THE FORT ST. VRAIN NUCLEAR GENERATING STATION

1. INTRODUCTION As a result of the Nuclear Regulatory Commission (NRC.) staff review of the Fort St. Vrain Nuclear Generating Station (FSV) Equipment Qualification Program, questions arose regarding the role of operators arid certain operator actions required for mitigation of high energy line breaks (HELBs). The purpose of this report is to provide an evaluation of the responses of Public Service Company of Colorado (PSC), the licensee for FSV, to NRC question:: concerning operator roles and actions.

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2. BACKGROUND On February 22, 1986, at a meeting between the NRC staff and PSC on the Fort St. Vrain electrical equipment environmental qualification (EQ) program, the role of operators in mitigating the effects of-an HELB was discussed. As a result of these discussions, it was determined that the operators' role required clarification.

The NRC requested additional information regarding this issue on April 6, 1986 (Reference 1). PSC responded to this request on June 26, 1986 (Reference 2). PSC also submittef revised responses to the NRC questions on January 15, 1987 (Reference 3) and on March 31, 1987 (Reference 4).

This report is based on the latest licensee responses, as contained in References 3 and 4, and evaluates the PSC responses to NRC staff questions relating to the role of operators and certain operator actions required for mitigation of certain assumed HELBS.

The evaluation performed concentrated on the complexity of the actions required, the environment in which the actions must be performed, and the timeframe available for performance of the actions.

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3. QUESTIONS / RESP 0NSES On February 22, 1986, a meeting was held with the licensee to address the role of the operators in mitigating the effects HELB accidents. Several items that were addressed required the licensee to supply additional information the NRC staff. The following information was requested by the staff.,

1. List and briefly describe local manual actions that would' be taken by the operators following an HELB accident or equivalent primary coolant leak.

2. provide an evaluation of the ability of the operators to perform these actions in potentially.high temperature environments. This evaluation should address the same factors covered in approving FSV operation at 35 percent power.

3. If future modifications to the plant are contemplated that affect this issue, please describe these modifications and your proposed schedule for their completion.

l The licensee responded to each of these requests as follows.

NRC Ouestion No. 1 List and briefly describe local manual actions that would be taken by the operators following an HELB accident or equivalent primary coolant leak. _

pSC Response The equipment qualification program flow path for the resumption of forced cooling following an HELB involves the use of firewater to drive one circulator and provide cooling water to the economizer, evaporator, 3

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and superheater (EES) sections of one steam generator. The following a

. local manual actions must be performed by the operators following an HELB:

o Alignment of the firewater system to the emergency condensate system-is required to provide a firewater flow path to the steam generator and circulator pelton wheel. Two manual valves (HV-4518, HV-4519) must be opened to align the flow path,:and two manual . valves (HV-4520, HV-45201) must be closed to isolate

. portions of the firewater system not required for safe shutdown to preclude firewater from being directed to any sprinklers activated by the elevated ambient temperatures. These four valves are.

located outside the harsh environment.

o' The discharge valve (HV-22821 for Loop 1 or HV-22822 for Loop'2) in the flow path for firewater cooling must be manually opened.

The valves are located in the Reactor Building and the operators rely on the manual override handwheels.

o Local manual actions would be required to line up secondary cooling for certain pipe break locations in combination with specific single active failures. The valves are environmentally qualified and therefore should be operable from the control room.

However, if one did fail, the following actions would be performed.

Break Location Sincle Active Failure Manual Action i

l Feedwater or Main HV-2237 Note 1 I Steam Loop 1 or

_ Piping FV-2205 Note 2 Feedwater or Main HV-2238 Note 1  !

Steam Loop 2 or Piping FV-2206 Note 2 Note 1 Local operator manual action would consist of turning a handwheel to manually override the valve.

Note 2 Local operator manual action would consist of opening two small valves to admit hydraulic fluid to the valve actuator from a local accumulator. ]

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f o The operator must verify.the steam generator outlet pressure via a-local indicator (PI-22129-2 for Loop 1 or PI-22130-2 for Loop-2).

The operator must then monitor outlet. temperature on a local j temperature gauge (TI-22823 for Loop 1 and TI-22824 for Loop 2). J The temperature gauges are located in the Reactor Building, and the pressure gauges are located in the Turbine Building.

Following a primary coolant leak ~ (i.e., Design Basis. Accident No. 2, . Rapid Depressurization Blowdown), primary coolant circulation and secondary coolant heat removal could be interrupted-by actuation of the Steam Line Rupture Detection Isolation. System (SLRDIS). Although there are actions which the operator would perform from the Control Room to ensure safe shutdown of the plant, there are no immediate or necessary actions

outside the Control Room which are required of him.

Evaluation PSC's response to NRC Question No. 1 is acceptable. Opening two manual valves (HV-4518 and HV-4519) and closing two manual valves (HV-4520 and V-45201) are relatively simple operator actions that take place on equipment located outside a harsh environment. Manual opening of the discharge valve (HV-22821 for Loop 1 or HV-22822 for Loop 2) .is an operator role requiring a relatively simple action _ using a manual override handwheel.

Manual action on the single active failure valves (HV-2237 or FV-2205 and HV-2238 or FV-2206) would consist either of turning a handwheel to override the valve or opening two small valves to admit hydraulic fluid to the valve actuator from a local accumulator. The manual action on the single active failure valves is a relatively simple act.

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f" Verification.of steam generator outlet pressure and temperatures is l also a simple act.

All of the required operator actions are relatively simple acts, which an operator should be able to perform with a minimal amount of training.

l NRC Ouestion No. 2 Provide an evaluation of the ability of the operators to perform these actions in potentially high temperature environments. This evaluation should address the.same factors covered in approving FSV operation at 35 percent power.

pSC Response Opening of the discharge valve, reading the gauges, and overriding the single failure require entrance into an area with a potential harsh environment. The approximate time that the above actions would be required would be 1/2 hour following break detection or inteh uption of forced circulation. The composite EQ program temperature prufile shows that temperatures would be 140 degrees Fahrenheit or less. Access with or without cool suits would be possible.

All other manual actions identified in Response No. 1 do' not require access to the harsh environment, including the access routes from the l

Control Room to the action locations.  !

Evaluation -

PSC's response to NRC Question No. 2 is acceptable. A review of the ecmposite EQ Program temperature profile (as seen in Reference 5) verified that the temperatures would be approximately 140 degrees Fahrenheit or less. This potentially harsh environment should not 6

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hinder:the operator, with or without a cool suit, from mitigating an HELB. In addition, as stated in the response, 30 minutes are available for completion of the required actions.

I Temperatures in the reactor building appear to remain at approximately 1250F for several hours after the accident. Temperatures in the-turbine building appear to remain approximately 1350F for several hours after the accident. Therefore, ice vests and air packs must be available for operator use in the event that any complications arise while performing the required operator actions. A letter from PSC will adequately close out this item.

NRC Ouestion No. 3 If future modifications to the plant are contemplated that affect this issue, please describe these modifications and your proposed schedule for their completion.

pSC Response .

During the outage at FSV, several system modifications have been implemented to reduce the operator access requirements in areas of high temperature environment following an HELB and to facilitate the initiation of safe shutdown cooling. The modifications listed in Reference 1 have been completed. In addition, a modification to install new 6" vent line discharge paths (HV-22819, HV-22820, HV-22821 and HV-22822) has been completed.

Future modifications may be consider ~ed for justifying power levels beyond 82%. However, these modifications have not been defined at this time. Likewise, as we gain experience with the present system, we may find it advantageous to make more modifications to reduce operator manual actions, but such modifications are not presently in our plans.

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l Evaluation PSC's response'to NRC-Question No. 3 is acceptable, since these modifications were specifically made to reduce operator access to areas of high. temperature. Any planned future modifications should be

. evaluated by the NRC staff.' However, as stated above, these modifications have not b6en defined at this time.

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4. EVALUATION This evaluation was performed.in order to address the NRC staff's concern about operator roles in' mitigating an HELB. The documents reviewed duringLthe evaluation are. listed in the reference section at the end of this

. report.

Reference 4, a PSC letter dated March 31, 1987 (P-87131), states that equipment required for mitigation of an HELB is environmentally qualified and/or located outside a harsh environment or is operable from remote handwheels.

Only for assumed single failure of qualified valves would operator actions in harsh environments be required. PSC has verified that at least 1/2 hour would be available before operator actions were required.

Temperature profiles show that the area temperature would be 140 degrees F or less at that time.

The PSC approach appears to be acceptable. This determination is based on the relatively simple nature of the required operator actions (opening manual valves and reading gauges); on the results of the temperature profile, which indicates access with or without protective clothing (140 degrees F or less); and on the timeframe available. However, it should be verified that these required actions are addressed at some time

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5. CONCLUCION Based on the documents presented for review and evaluation, it is concluded that FSV has the capability to adequately deal with an HELB. It is also concluded that the operators' role in mitigating an HELB is acceptable, based on the staff's present position concerning operator involvement in accident scenarios.

It should be verified that the utility has addressed operator actions in some portion of their training program. The training should specifically address recognizing the need for manual operator action, as well as equipment location and equipment operation. Training should also address the location and proper use of cool suits, ice vests, and/or air packs in addition to possible hazards associated with entering the HELB atmosphere.

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6. REFERENCES

1. Letter, Heitner (NRC) to Walker (PSC), April 6, 1986.

- 2. Letter, Warembourg (PSC) to Berkow (NRC), June 26,.1986 (P-96438).

3. Letter, William:: (PSC) to Berkow (NRC), January 15, 1987 (P-87002).

4. Letter, Warembourg (PSC) to Berkow (NRC), March 31, 1987 (P-87131).

5. Letter, Walker (PSC) to Berkow (NRC), March 14, 1986 (P-86208).

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