ML20148Q757

Provides Info in Response to Various NRC Inquiries Re Adequacy of Facility Remote Shutdown Capability,Per NRC 871016 & s.Util to Conduct Integrated Testing to Demonstrate Primary Sys Inventory Control Prior to Restart

ML20148Q757
Person / Time
Site:	Rancho Seco
Issue date:	01/25/1988
From:	Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F Office of Nuclear Reactor Regulation
References
GCA-88-029, GCA-88-29, NUDOCS 8802010192
Download: ML20148Q757 (13)

Text

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$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT D 6201 S Street, P.O. Box 15830 Sacramento CA 958521830 (916) 452 3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA GCA 88-029 JAN 2 51933 U. S. Nuclear Regulatory Commission Attn: Frank J. Miraglia, Jr.

Associate Director for Projects Phillips Building 7920 Norfolk Avenue Bethesda, MD 20014 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 REMOTE SHUTDOWH CAPABILITY

References:

Correspondence, Crutchfield to Andognini, January 7,1988, Restart Inspection Findings.

Inspection Report 87-24, Zimmerman to Andognini, October 16, 1987, Resident Inspection Report.

Dear Mr. Miraglia:

In response to various NRC inquiries, including those contained in the referenced documents, the following information is provided with respect to the District's actions to assure the adequacy of Rancho Seco's remote shutdown capability.

Prior to restart, the District will conduct integrated testing to demonstrate primary system inventory control and secondary heat transfer control from the remote shutdown panel. The District will also conduct, prior to restart, an in-plant walkdown of OP-C.13a, "Plant Shutdown from Outside the Control Room",

with all operating crews.

The combination of these actions, along with post-modification component and functional testing, will provide assurance that the plant and personnel are capable of performing a remote shutdown evolution.

Attached is a description of accomplished and planned activities supporting remote shutdown capability.

Please contact me if you have any questions. Members of your staff with coments or questions requiring additional information or clarification may contact Mr. David Compton at (209) 333-2935, extension 4915.

Sincerely, An n

Chief Executive Officer, dO 47 Nuclear 1

8802010192 880125 i/l PDR ADOCK 05000312

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P PDR RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95633 9799;(209) 333 2935

ATTACHMENT'l REMOTE SHUTDOWN CAPABILITY ORIGINAL CAPABILITY AND TESTING The original plant design provided the capability to shut the reactor down and maintain it in a safe condition (i.e., hot shutdown) if access to the control room were lost. This remote shutdown capability, described in USAR Section 1.4.11, was provided to satisfy General Design Criterion ll, "Control Room",

as originally proposed in July 1967. The potential capability for placing the reactor in a cold shutdown condition is also described in USAR Section 1.4.11.

Subsequent to FSAR filing in 1971, the Commission conducted its technical review against 10 CFR 50, Appendix A, Criterion 19, "Control Room",

(see USAR Section 1.5.15), issued in 1971, and concluded that the plant design conformed to the intent of the revised criterion.

The District received its Operating License for Rancho Seco on August 16, 1974 Testing was performed during initial plant startup to demonstrate the remote shutdown capability.

This test was conducted from above 10 percent of rated gower with normal full power line up established (USAR Section 13.1.4, Initial Startup Programs").

10 CFR 50, APPENDIX R MODIFICATIONS AND TESTING Subsequent to the promulgation of 10 CFR 50.48, which requires plants licensed to operate before January 1,1979 to comply with Sections III.G, III.J, and III.0 of 10 CFR 50, Appendix R, the District re-evaluated the remote shutdown capability.

Paragraph 50.48(c)5 requires NRC review and approval of modifications planned to meet the requirements of Section III.G.3 of Appendix R, "Fire protection of safe shutdown capability".

Compliance with Section III.L of Appendix R, "Alternative and dedicated shutdown capability",

is considered included by Section III.G 3 and therefore also required.

Generic Letter 81-12 provides a summaiy of the information required by the Commission to evaluate the adequacy of the remote shutdown capability relative to Appendix R.

By letters dated November 30,1983, April 5,1985, July 12, 1985, September 27, 1985, and October 30, 1985, the District provided the information requested by Generic Letter 81-12, and responded to questions on l

the submittals (see also USAR Section 7.4.8.2, "Control Room Evaluation i

Emergency"). The information submitted by the District identifies and l

describes the modifications to the remote shutdown panel, the testing l

performed on the panel, and the procedures for safe shutdown from outside the control room.

The District's re-evaluation concluded that equipment required for alternative shutdown capability to be the same or equivalent to that relied upon in the original Safety Analysis. The scope of the Appendix R modifications to the remote shutdown capability was limited to addition of indication, control, and electrical isolation of existing systems.

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Page 2 of 4 These modifications were completed prior to Cycle 7 startup in June 1985.

Post-modification tests of the Control Room isolation switches and remote instrumentation and controls providing remote shutdown capability were performed following initial component installations.

The NRC Region V inspection of Appendix R compliance was conducted on August 12-16, 1985, and documented by Inspection Report 85-22, August 27, 1985. The inspection evaluated compliance with Appendix R Sections III.G.3 and III.L, and included a walkthrough of the hot shutdown procedures to verify effectiveness.

No items of noncompliance were identified.

By letter dated May 19, 1986, NRR provided their Safety Evaluation Report of Rancho Seco's post-fire alternate shutdown capability. The evaluation concludes that, allowing for the exemption for 205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br /> to reach cold shutdown, the safe shutdown capability meets the requirements of Sections III.G.3 and III.L of Appendix R.

RECENT MODIFICATIONS AND TESTING Since December 1985, modifications made to the remote shutdown panel / remote shutdown capability have been associated with installation of EFIC (Emergency Feedwater Initiation and Control). Remote EFIC control at the shutdown panel H2SD has been installed by Engineering Change Notice (ECN) A-5415P. The scope of ECN A-5415P includes:

0 addition of hand / auto stations HIC-20527B and HIC-20528B for AFW control valves FY-20527 and FY-20528 to provide control via the EFIC cabinets.

O addition of hand / auto stations HIC-205718 and HIC-20562B for the six (6) Atmospheric Dump Valves (ADVs) to provide control via the EFIC cabinets.

O design modifications to the Steam Generator (SG) wide range level and pressure indicators (LI-20509/LI-20510, and LI-20547/LI-20548, respectively) to provide signal inputs from isolated outputs of the EFIC cabinets in the Nuclear Services Electrical Building.

O deletion of hand switch HS-20578A which previously provided control for AFW Bypass Valve SFV-20578.

These H2SD changes have been functionally demonstrated during post-modification testing. All H2SD instrumentation and controls to be demonstrated in planned integrated testing have been verified functional on a component basis. provides a tabulation of H2SD instrumentation and controls with their associated pre-restart tests.

Page 3 of 4 INTEGRATED TESTING A recent review of the restart test program by the NRC (IR 87-24) concluded that planned testing was basically similar to that recomended by Regulatory Guide 1.68, "Initial Test Programs for Water-Cooled Nuclear Power Plants."

Regulatory Guide 1.68.2, "Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants" was issued in January 1977, after Rancho Seco's initial startup.

However, integrated testing of the Rancho Seco remote shutdown panel through the restart test program is planned using the guidance of Regulatory Guide 1.68.2.

Integrated testing of the remote shutdown panel will be accomplished as part of STP.1113, "EFIC Hot Functional Test", to be performed prior to plant restart criticality.

STP.1111 will provide a functional demonstration of primary system inventory control and secondary heat transfer control from the remote shutdown panel.

STP.1113 will functionally dernonstrate all remote shutdown panel indications and controls with the exception of the makeup isolation valve SFV-23508 and the DHR A BWST suction valve SFV-25003.

Normal RCS makeup lineups will be maintained with HPI Loop A SFV-23811 controlling makeup rather than the normal makeup control valve. Normal Reactor Coolant Pump (RCP) Seal Injection will be controlled from the Control Room, with RCS letdown control maintained in the Control Room and not terminated.

The Control Room will maintain sufficient control of plant equipment during testing to take effective action to assure plant safety in the unexpected case that the test does not proceed as planned. The Control Room will maintain control of the ADV block valves, AFW motor-operated isolation valves, the normal makeup valve, and the HPI pump trip.

The Control Room observers and test personnel at the H2SD panel will maintain an open channel for communications during testing.

STP.1113 will verify the following functions from panel H2SD, with the plant at hot shutdown and four RCPs running:

0 Demonstrate that the Turbine Bypass Valve (TBY) Emergency Close solenoids cause the TBVs to close.

O Demonstrate that the Attrospheric Dump Valves (ADYs) will automatically control SG pressure at setpoint and can be controlled manually.

O Demor, strate that the RCS temperature can be reduced in a controlled manner by manual operation of the ADYs.

0 Demonstrate that the RCS inventory can be maintained through the HPI LOOP A SFV-23811 control.

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Demonstrate that the EFIC signals to the AFW Flow Control Yalves will automatically control SG level at setpoint and can be controlled manually.

O Demonstrate that the planc can be maintained in a stable Hot Shutdown condition with ADY manual pressure control and EFIC automatic SG 1evel.

O Demonstrate the H2SD indicators are functional. provides the approved test outline for STP.1113, "EFIC Hot Functional Test". Attachment 4 provides a comparison of planned integrated testing with Regulatory Guide 1,68.2.

_ PROCEDURES AND TRAINING Final restart procedure revisions on both OP-C.13a, "Remote Shutdown from Outside the Control Room" and OP-C.13b, "Remote Cooldown from Outside the Control Room" are complete and the procedures issued with Plant Review Committee (PRC) approval.

Plant Operators have received over 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> of classroom, on-the-job and simulator training relative to the plant modifications made during the restart outage.

This has included training on both hardware and procedure revisions.

Operators have received extensive training on EFIC.

Operators have received training on all modifications made to the remote shutdown panel.

In addition to the hardware change training, Operators have received training on the two major procedures controiling the use of the shutdown panel, OP-C.13a and OP-C.13b.

Classroom training has been conducted on the procedures referenced above for all Operators. An in-plant walidown of OP-C.13a is scheduled to be conducted for all operating crews prior tu restart. This will consist of having each crew implement the procedure as if the event had actually occurred.

Adjustment of plant controls will be simulated.

Similar in-plant walkdowits of OP-C.13a were conducted in April and May of 1985 for all operating crews.

The effectiveness of these training exercises is demonstrated by the Operations personnel evaluations provided in Inspection Report 85-22 (August 27,1985).

Operator training on OP-C.13a is an annual requirement of the Licensed Operator Continuing (Requalification) Training Program and will be conducted in accordance with that periodicity.

This will ensure that the Operators remain proficient on the guidelines to be used to shut down the plant in a safe and controlled manner from outside the Control Room.

Page 1 of 1 ATTACHMENT 2 REMOTE SHUTDOWN PANEL INSTRUMENTATION AND CONTROL EQUIPMENT ID DESCRIPTION PRE-RESTART TESTING HIC-20527B AFW VALVE LOOP A STP-666 HIC-20528B AFW VALVE LOOP B STP-666 HIC-20562B ADY LOOP B STP-666 HIC-205718 ADY LOOP A STP-666 HS-20561 TBV CONTROL STP-1040 LI-20509 OTSG A LEVEL STP-666 LI-20510 OTSG B LEVEL STP-666 LI-21502C MU TANK LEVEL SP.200.14 LI-21503B i2R LEVEL STP-1115 (SP.200.14)

LI-21503D PZR LEVEL SP.200.14 PI-20547 OTSG A PRESSURE STP-666 PI-20548 OTS3 B PRESSURE STP-666 PI-21050 RC PRESSURE PM TASK #3922 PI-21051 RC PRESSURE PM TASK #3922 SWITCH SFV-23508 MU TANK ISOLATION STP-1091 SWITCH SFV-23811 HPI LOOP A INJECT STP-1091 SWITCH SFV-25003 DHR A BWST SUCTION STP-1087 TI-21024C COLD LEG TEMP PM TASK #3357 TI-21025E COLD t.EG TEMP PM TASK #3357 i

TI-21031C HOT LEG TEMP PM TASK #4515 TI-21032C HOT LEG TEMP PM TASK #4516 i

• All components have successfully completed functional testing, except j

Switch SFY-25003 which will be danonstrated prior to restart.

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ATTACllMENT 3 TEST O(IfLINE REV. 1 EFIC HCfr FUNCTIONAL TEST ECN(s)/REV NO:

A-5415 REV. 4. R-0861 Rev. 6 PREPARED BY:

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Dmr PROCEDURE NO:

__STP.1113 1.0 TEST OBJECTIVE 1.1 To functionally demonstrate the operability of the Emergency Feedwater Initiation and Control (EFIC) system controls and indication under closed loop conditions with the plant in Hot Shutdown mode.

1.2 To document the relationship between the EFIC system and the Non-Nuclear Instrumentation (NNI) system steam generator level and pressure indications.

1.3 To functionally demonstrate the Appendix 'R' Rat Shutdown Panel (H2SD) indications and controls (except the MU TNK ISOL SFV-23508 valve and the DHR PUMP A SUCT. FROM BWST SFV-25003 valve) as committed to in letter GCA 88-029.

1.4 To functionally demonstrate the EFIC system and the Auxiliary Feedwater (AFW) system response time from EFIC initiation of AFW to the time that 475 gpm AFW flow is i

achieved.

l 2.0 ACCEPTANCE CRITERIA l

2.1 Demonstrate, with the plant at Hot Shutdown (HSD), that the l

EFIC controller gain operating settings result in stable (not diverging) steam generator (SG) level control under closed loop conditions. (

Reference:

DBR A-5415 MAJOR, Calculation Z-EFI-0146) 2.2 Document the relationship between the EFIC and NNI steam generator level and pressure indications.

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Reference:

DBR A-5415 MAJOR) 1 I

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4 TEST OUTLINE, Rev.1 (Continued)

STP.1113 2.3 Verify, with the plant at HSD and four Reactor Coolant Pumps (RCPs) running, that EFIC AFW initiation and flow control will not result in excessive overcooling of the Reactor Coolant System (RCS).

RCS temperature should be maintained at > 5000 F and the cooldown rate within the technical specification limits.

Stable level control (not diverging) will be demonstrated. (

Reference:

DBR A-5415 MAJOR, Technical Specification 3.1.2 and Emergency Operating Procedure E.05) 2.4 Verify the following functions from Panel H2SD, with the plant at HSD and four RCPs running:

(

Reference:

DBR A-5415 MAJOR, DBR R-0861, Calculation Z-EFI-0146, USNRC Regulatory Guide 1.68.2) 2.4.1 Demonstrate that the Turbine Bypass Valve (TBV)

Emergency Close solenoids cause the TBVs to close.

2.4.2 Demonstrate that the Atmospheric Dump Valves (ADVs) will automatically control SG pressure at setpoint.

and can be controlled manually.

2.4.3 Demonstrate that the RCS temperature can be reduced in a controlled manner by manual operation of the ADVs.

2.4.4 Demonstrate that the RCS inventory can be maintained through the HPI LOOP A SFV-23811

Control, 2.4.5 Demonstrate that the EFIC signals to the AFW Flow

(

Control Valves (FCVs) will autcmatically control SG level at setpoint and can be controlled manually.

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2.4.6 Demonstrate that the plant can be maintained in a l

stable HSD condition with ADV manual pressure l

control and EFIC automatic SG level.

2.4.7 Demonstrate the remote indicators are functional.

2.5 Demonstrate that the time interval between initial EFIC AFW initiation and establishing an AEW flow of 475 gpm is not greater than 70 seconds, including the maximum EFIC Time j

Delay module setting.

(

Reference:

EAR SY87-119) 1 l

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TEST OUTLINE, Rev. 1 (Continued)

STP.1113 3.0 CONDITIONS PRIOR 'IU TEST 3.1

'Ihe EFIC and AW systems are operable in accordance with existing operating procedures.

3.2 The EFIC controls and associated components receive normal electrical power throughout the test.

3.3 STP.667, EFIC Cold Preoperational Test is complete and the test results are satisfactory.

3.4 The Remote Shutdown Panel, H2SD, is available for service.

3.5 The Interim Data Acquisition Display System (IDADS) is in service.

3.6 The plant is in Hot Shutdown per Operating Procedure B.2, Section 4.3.

4.0 TEST ME'IBOD 4.1 The NNI and'EFIC system SG levels and Main Steam line pressure outputs will be monitored by IDADS during the testing.

The pressure readings will be documented in the critical region near the TBV post trip setpoint and the ADV setpoint.

4.2 The EFIC low level initiate and control test will be started by reducing Main Feedwater to one SG until EFIC initiates AW on low SG level.

'Ihe other SG will be at a level sufficiently above the low level control setpoint such that the EFIC FCV demand signal will close the AW FCVs to that SG.

4.3 The functional testing of the Remote Shutdown panel, H2SD, will be accomplished by transferring the following controls from the Main Control Room to panel H2SD (All EFIC FCN, pressure and initiation controls and their respective indications will be isolated from the Control Room):

4.3.1 HPI IDOP A SW-23811 4.3.2 EFIC AW CONTROL VALVE LOOP A 4.3.3 EFIC ADV CONTROL IDOP A 4.3.4 EFIC AW CONTROL VALVE IDOP B 4.3.5 EFIC ADV CONTROL IDOP B 3

TEST OUTLINE, Rev.1 (Continued)

STP.1113 l

l 4.4 The ADV automatic control setpoint verification will be l

cocomplished from panel H2SD by isolating the TBVs through the use of the H2SD 'IURBINE BYPASS VALVE MANUAL CLOSE, HS-20561, handswitch.

The OTSG pressures will increase to the ADV setpoint and be verified individually.

I 4.5 The ADV manual control will be demonstrated from panel H20D after each ADV automatic setpoint is verified by using the H2SD HIC, in manual, for each valve.

0 F 4.6 The RCS temperature will be decreased a minimum of 10 through the manual control of the ADVs from the H2SD panel.

The RCS inventory will be controlled during the cool down through the use of the H2SD HPI IDOP A SFV-23811 control.

4.7 Demonstration of stable HSD control will be accomplished from the H2SD panel with EFIC maintaining SG level.

4.8 Demonstration of the time from EFIC initiation to minimum AW flow will be by manually initiating the EFIC Channel "B" to ~cause flow via the AW turbine. driven pump, P-318, i

through the most restrictive flow path, FV-20532, to the "B" SG.

The maximum EFIC time delay, 9.9 seconds, will be incorporated into the final test results.

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ATTACHMENT 4 COMPARISON TO REGULATORY GUIDE 1.68.2 REG. GUIDE 1.68.2 RANCHO SECO STP.1113 3.

Hot Standby Demonstration Procedure Testing will begin at stable hot shutdown (Tave greater then 5250F, The test should be initiated from a reactor subcritical). Data will be location outside the control room with the obtained at the hot shutdown panel reactor at a moderate power level (10-25%)

H2SD to verify that the plant is in a sufficiently high that plant systems are a hot shutdown condition.

All H2SD in the normal configuration with the indicators will be demonstrated-turbine generator in operation.

The functional, test should be performed with the mini-mum of personnel required to be at the Testing will be conducted prior to reactor unit at any one time (minimum plant restart criticality, with shift crew). Data should be obtained four RCPs operating and essentially at locations outside the control room to no decay heat in the core.

Remote verify:

shutdown testing initiated at power was performed during initial plant startup, a.

That the plant has achieved hot The remote shutdown Casualty Procedure standby status.

C.13a instructs operators to trip the reactor and verify rod insertion prior to Control Room evacuation.

In-plant walkdowns of C.13a in 1985 considered minimum shif t staffing.

Planned in-plant walkdowns of C.13a with all operating crews will be performed considering minimum shift staffing.

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Page 2 of 3 REG. GUIDE 1.68.2 RANCHO SEC0 STP.1113 b.

That the plant can be maintain-Stable hot shutdown conditions will be ed at stable hot standby conditions for maintained for at least 30 minutes with at least 30 minutes.

ADV manual pressure control and EFIC automatic SG control. All applicable in-During the demonstration only strumentation and controls at panel H2SD that equipment for which credit would will be demonstrated, with the exception be taken in performing an actual remote of the controls for makeup tank. isolation shutdown should be used.

valve SFY-23508.

During a CR evacuation, RCP seal injection would be controlled by local manual operation of the valve handwheel, observing a local flow meter.

During STP.1113, RCP seal injection control will be maintained in the Control Room, rather than transfer to local manual, to minimize unnecessary risk to the pump seals.

RCS letdown control will be maintained in the Control Room and not terminated.

Normal RCS makeup lineups will be main-tained with HPI Loop A valve SFY.23811 controlling makeup rather than through the normal makeup valve.

Timing and com-munications for manual valve lineups and electric power lineups required in C.13a will be demonstrated during in-plant walkdowns.

Certain train B equipment required to be isolated from the Control Room during performance of Casualty Procedure C.13a will not be isolated. The Control Room will maintain sufficient control of plant equipment during testing to take effective action to assure plant safety in the unexpected case that the test does not proceed as planned.

Page 3 of 3 REG. GUIDE 1.68.2 RANCHO SECO STP.1113 4

Cold Shutdown Demonstration Procedure Control of secondary heat transfer will be demonstrated by an RCS cooldown of at The licensee should demonstrate a least 100F.

The RCS temperature will be potential capability for cold shutdown reduced in a controlled manner by manual by partially cooling down the plant operation of the ADVs.

AFW flow control from the hot standby condition using will be demonstrated in both automatic controls and instrumentation located (EFIC) and manual.

RCS inventory will outside the control room, be controlled from panel H2SD using SFV-23811.

The test should demonstrate that:

This partial RCS cooldown is considered a.

The reactor coolant tempera-sufficient to demonstrate the potential ture and pressure can be lowered capability for remote cold shutdown.

sufficiently to permit the operation of the core decay heat removal system that is to be ultimately used to place the reactor in a refueling shutdown mode.

b.

Operation of this decay heat Transfer to the decay heat removal removal system can be init*ated and system and subsequent continued and controlled, cooldown are slowly-unfolding evolutions, well within the plant equipment and c.

A heat transfer path to the personnel capabilities.

Operators have ultimate heat sink can be established.

received training on the remote cooldown Casualty Procedure C.13b. Functionality d.

Reactor coolant temperature of panel H2SD control of DHR A BWST can be reduced approximately 500F suction valve SFV-25003 will be using this decay heat removal system demonstrated prior to restart, at a rate that would not exceed tech-The design adequacy of the decay heat nical specification limits.

This removal system is considered sufficiently cooldown should show that the potential demonstrated, therefore, a further for achieving cold shutdown from out-demonstration of this capability is side the control room is available, considered unnecessary. The benefits of performing this further demonstration are considered minimal in comparison to the time and manpower required for such a demonstration.