ML20094F303

Forwards Emergency Operating Procedure Generation Package,In Response to Requirements of Generic Ltr 82-33, Suppl 1 to NUREG-0737 - Requirements for Emergency Response Capability, Section 7

ML20094F303
Person / Time
Site:	Summer
Issue date:	07/30/1984
From:	Dixon O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20094F305	List: ML20094F303 ML20094F319
References
RTR-NUREG-0737, RTR-NUREG-737 GL-82-33, NUDOCS 8408100002
Download: ML20094F303 (11)

Text

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., _.O SOUTH CAROLINA ELECTRIC & GAS COMPANY POST OFFICE 764 CoLuustA. South CAROLINA 29218 O. W. O' mon. Ja.

suc$*.$.'"o','Ur7o . Ju1y 30, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Generic Letter 82-33 EOP Generation Package

Dear Mr. Denton:

In response to the requirements of Generic Letter 82-33,

" Supplement 1 to NUREG-0737-Requirements for Emergency Response Capab ili ty ," Section 7, the Emergency Operating Procedure (EOP)

Generation Package for the Virgil C. Summer Nuclear Station is herewith submitted as enclosures I through III to this letter.

South Carolina Electric and Gas Company's ( SCE&G) upgraded EOPs are based on Westinghouse Owners Group Emergency Response Guidelines ( ERG s ) , revision 1, and were developed with the assistance of Essex Corporation to ensure human factor concerns were adequately addressed.

Enclosure I compares the Virgil C. Summer Nuclear Station with the ERG generic plant described in the Executive Volume of the ERGS. Enclosure II outlines the specific areas where SCE&G's EOPs differ from the Westinghouse Owners Group ERGS, revision 1.

Enclosure III is the Virgil C. Summer Nuclear Station Administrative Procedure (SAP-207) which details the development of EOPs.

SAP-207 addresses requirements found in Generic Letter 82-33, Item 7.2.b. Section 6, " Technical Guidelines," details elements of the plant specific technical guidelines and how these guidelines are used for procedure development. The Emergency Operating Procettures Writers Guide, Attachment I to this procedure, is ba sed on guidelines published by the Institute of Nuclear Power Operations (INPO). Section 10.0, " EOP Validation,"

describes the procedure validation program which was also designed using INPO guidelines. The training program is addressed in Section 11.0, " EOP Training."

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Mr. Harold R. Denton '

Generic Letter 82-33 EOP Generation Package July 30, 1984 Page #2 SCE&G's plant specific procedure validation program is scheduled to begin in August 1984. Formal operator training of the upgraded procedures should begin in October 1984.

As stated in our March 22, 1983 letter to the Staff, justification for manual reactor coolant pump (RCP) trip was to be submitted along with this EOP generation package. This justification is found in the Westinghouse Report, " Justification of Manual RCP Trip for Small Break LOCA Events," transmitted to the NRC by the Westinghouse Owners Group in a letter dated March 12, 1984.

If you should have any questions, please advise.

Very truly yours,

/

O. W. Dir , Jr.

AMM/OWD/gj Enclosures cc: V. C. Summer C. A. Price T. C. Nichols, Jr./O. W. Dixon, Jr. C. L. Ligon (NSRC)

E. H. Crews, Jr. K. E. Nodland E. C.' Roberts R.-A. Stough W. A. Williams, Jr. G. Percival D. A. Nauman C. W. Hehl J. P. O' Reilly J. B. Knotts, Jr.

Group Managers NPCF O. S. Bradham File

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ENCLOSURE I CGSS-01-1004-N0 COMPARISON OF THE VIRGIL C. SUMMER NUCLEAR STATION TO THE HIGH PRESSURE REFERENCE PLANT The V. C. Sumer Nuclear Station is a Westinghouse Model 312 three loop nuclear power plant rated at 2775 MWT with a General Electric 950 Turbine / Generator rated at 950 ftWE.

Differences between the V. C. Sumer Nuclear Station and the High Pressure Reference Plant are as follows: (Refer to ERG Executive Volume).

2.2 Safety Injection Signal -

In addition to the reference plant SI signals there is a steamline Differenti&1 Pres-sure SI at 97 psi. This does not have a block / reset function.

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2.2 Safety Injection Block -

Low Pressure Ste  !

ignal block / reset is c }[

P-12, Low-Low R(

2.2 Turbine Driven EF - Loss of Power f ,

Pump Start buses versus 10 t RCP Buses.

2.2 Main Steam Isolation - High Steam Flow coincident with Low-Low RCS Tavg versus High Steam Pressure rate below P-11.

2.2 Main Steam Isolation - Dependant on Low-Low Tavg (P-12)

Block versus Low Pressure (P-11). Steam-lineAP SI is not blockable.

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$.7 Reactor Coolant System - 3 loop versus 4 loop. 3 Pzr PORV versus 2 PZR PORV. Cold Over-pressure Protection System is auto-matically placed in service. Proposal currently submitted to eliminate COPS.

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1 ENCLOSURE I CGSS-01-1004-N0 2.8 Hot Leg Recirculation -

Only recirculates from the Mode Rx. Bldg. sump to the RCS hotlegs.

2.8 Charging /SI Subsystem -

A seperate line is provided for cold leg recirculation in addition to thru the BIT.

2.8 High-Head SI - Not applicable to V. C.

Subsystem Summer.

2.8 SI-Accumulator Sub- - 3 Charging /SI pumps and no system positive displacement pump.

Seal return and charging pump mini-flow returns to the volume control tank, not charging pump suction.

2.11 Component Cooling -

Does not provide water to Water System the containment fan coolers.

2.12 Service Water System -

Provides cooling water to the containment fan coolers via Service Water Booster' Pumps.

Provides automatic emergency makeup water to the Emergency Feedwater System, i:

2.13 Containment Spray - Includes a Sodium Hydroxide System Systs.

2.14 Containment Atmosphere - Incl 0 des both charcoal and HEPA Control System filters.

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ENCLOSURE I CGSS-01-1004-N0 2.16 Main Feedwater & Condensate - The main feedwater system System is also isolated on a low flow, coincident with low feedwater temperature or high Intermediate Building sump levels. All main feedwater pumps are tur-bine driven. Between the condensate pumps and feed-waterpumpsarefour(4)

Feedwater Booster Pumps and a deaerator. Shutoff head of the Feedwater Booster Pumps is approximately 350 psig.

2.17 Emergency (Auxiliary) -

The Emergency Feedwater Feedwater System System is completely sep-arated from the Main Feed-water system and injects to steam generators via a separate nozzle. Any Emer-gency Feedwater Pump can supply all three Steam Genera-tors. The alternate water supply is the Service Water l System.

2.18 Steam Generator Blowdown - Includes automatic diversion to a holdup system in the l

event of high blowdown radia-l i

tion.

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ENCLOSURE I CGSS-01-1004-N0 2.24 Electrical Power System - The electrical power supply consists of two independent off-site power supplies feed-ing two independent on-site emergency power supplies.

During a blackout condition -

non-essential loads are locked-out on the emergency A. C. buses. l 2.25 Pneumatic Power System - Two of the three Pressurizer PORV's are supplied by a high pressure nitrogen system with accumulators instead of control air. ,

Instrument and Control Requirements (Table 3) are consistent with the reference plant.

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l ENCLOSURE II CGSS-01-1004-N0 Comparison of the Virgil C. Summer Nuclear Station Emergency Operating Procedures to the Westinghouse Owners Group Emer-gency Response Guidelines.

The Virgil C. Summer Nuclear ~ Station Emergency Operating Procedures (E0P's) are consistent with the Westinghouse Owners Group Emergency Response Guidelines (ERG's), Rev. 1, with the following exceptions:

1. ERG Procedures ES-0.3, " Natural Circulation Cooldown With Steam Void In Vessel (With RVLIS)", is not included in the E0P set. The difference between this procedure and ES-0.2, " Natural Circulation Cooldown", includes moni-toring reactor vessel level indication and provides di-rection to the Operators for dealing with reactor vessel void formation. This is included in E0P-1.3, " Natural Circulation". Since the Virgil C. Summer Nuclear Station is a T COLD plant as documented in the ERG Background document for ES-0.2, the lack of a plant specific E0P based on ES-0.3 is not considered a significant safety concern.

2. ERG Procedure FR-H.2, " Response To Steam Generator Over-Pressure", is not included in the E0P set. To reach S/G overpressure higher than the highest steamline safety valve setpoint requires the failure of five (5) code I

safety valves and a steamline power relief valve. The l ERG procedure response basically requires-the Operator l to dump steam to relieve the high pressu.re; however, if steam dump capabilities were available the conditions l

would not exist to begin with. Neither the initiating conditions nor the recommended Operator responses are considered credible. The ERG FR-H.2 Background document Page 1 of 2

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ENCLOSURE 11 CGSS-10-1004-N0 acknowledges that most plants have five (5) steamline safety valves and does not postulate 1:ow the entry conditions could conceivably be reached.

A complete listing of plant specific E0P's versus the appli-cable ERG procedure is included as Attachment I to this en-closure. Based on the results of the validation program, some consolidation of the procedures may become necessary.

Therefore the final E0P numbering as shown on this listing may change.

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• 4 6/8/84 A'ttachment I Enclosure II CGSS-01-1004-N0 E0P TITLES ~AND NUMBERING ERG NUMBER TITLE REV. 1 E0P-1.0 Reactor' Trip / Safety Injection E-0 Actuation E0P-1.1 Reactor Trip Recovery ES-0.1 E0P-1.2 Safety Injection Recovery ES-1.1 E0P-1.3 Natural Circulation ES-0.2 E0P-1.4 Rediagnosis ES-0.0 20P- 2.0 Loss Of Reactor Or Secondary Coolant E-1 E0P-2.1 Post-LOCA Cooldown And Depressurization ES-1.2 E0P-2.2 Transfer To Cold Leg Recirculation ES-1,3 E0P-2.3 Loss Of Emergency Coolant Recirculation ECA-1.1 E0P-2.~4 Total Loss Of The Residual Heat Removal System -

E0P-2.5 LOCA Outside Containment ECA-1.2 E0P-3.0 Faulted Steam Generator Isolation E-2 E0P-3.1 Uncontrolled Depressurization Of Generators ECA-2,1 E0P-4.0 Steam Generator Tube Rupture E-3 E0P-4.1 Post-SGTR Cooldown ES-3.1, 3.2, 3.3 E0P-4.2 SGTR With Loss Of Reactor Coolant: Subcooled Recovery ECA-3.1 E0P-4.3 SGTR With Loss Of Reactor Coolant: Saturated Recovery ECA-3.2 Page l of 3

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6/8/84 Attachment I Enclosure II

.E0P TITLES AND NUMBERING CGSS-01-1004-N0 ERG NAME TITLES REV. 1 E0P-4.4 SGTR. Without Pressurizer Pressure Control ECA-3.3 E0P-6.0 Loss Of All AC Power ECA-0.0 E0P-6.1 Loss Of All AC Power Recovery Without SI Required ECA-0.1 E0P-6.2 Loss Of All AC Power Recovery With SI Required ECA-0.2 E0P-7.0 Refueling Emergency -

E0P-8.0 Control Room Evacuation -

E0P-9.0 High Radiation Outside Containment -

E0P-10.0 Malfunction Of Control System -

E0P-11.0 Emergency Boration -

E0P-12.0 Monitoring Of Critical Safety Functions F-01 thru 6.0 E0P-13.0 Response To Abnormal Nuclear Power Generation FR-S.1 E0P-13.1 Response To Loss Of Core Shutdown FR-S.2 E0P-14.0 Response To Inadequate Core Cooling FR-C.1 E0P-14.1 Response To Degraded Core Cooling FR-C.2 E0P-14.2 Response To Saturated Core Cooling Conditions FR-C.3 E0P-15.0- Response To Loss Of Secondary-Heat Sink FR-H.1 Page 2 of 3

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6/8/84 Attachment I Enclosure II CGSS-01-1004-N0 E0P TITLES'AND NUMBERING

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ERG NUMBER TITLE REV. I E0P-15.1 Response To Steam Generator High Level FR-H.3 E0P-15.2 Response To Loss Of Normal Steam Release Capabilities FR-H.4 E0P-15.3 Response To Steam Generator Low Level FR-H.5 E0P-16.0 Response To Imminent Pressurized Thermal Shock FR-P.1 E0P-16.1 Response To Anticipated -

Thermal Shock FR-P.2 E0P-17.0 Response To High Reactor

• Building Pressure FR-Z.1 E0P-17.1 Response To Reactor Building Flooding FR-Z.2 E0P-17.2 Response To High Reactor Building Radiation Level FR-Z.3 E0P-18.0 Response To High Pressurizer Level FR-I.1 E0P-18.1 Response To Low Pressurizer Level FR-I.2 E0P-18.2 Response To Void In Reactor Vessel FR-I.3 Page 3 of 3