ML18026B121

Forwards Safety Analysis,Implementation Plan & Implementation Schedule for Safety Parameter Display Sys,Per DB Vassallo 840612 Confirmatory Order to Hg Parris

ML18026B121
Person / Time
Site:	Browns Ferry
Issue date:	07/30/1984
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8408030015
Download: ML18026B121 (12)

Text

REGULATORY FORMATION DISTRIBUTION SY EM (BIDS)

ACCESSION NBR:8008030015 DOC ~ DATE: 8Q/07/30 NOTARIZED: YES DOCKET FACIL:50-259 Browns Ferr y fiucl ear Power Stations Un) t 1< Tennessee 05000259 50-260 Browns Ferry Nuclear Power Stations Unit 2i Tennessee 05000260 50-296 Browns Ferry Nuclear Power Stations Unit 3, Tennessee 05000296 AUTH, NAME AUTHOR AFFILIATION MILLS L M0 E Tennessee Valley Authority RECIP ~ NAME RECIPIENT AFFILIATION DENTONEH ~ RE Office of Nuclear Reactor RegulationE Ditector

SUBJECT:

For wards safety analysisqimplementation plan implementation schedule for safety parameter displa sysrper DB Vassallo 800612 confirmatory order to HG Parris.

DISTRIBUTION CODE: A003S COPIES RECEIVED:LTR J ENCL J SI E' ~ m m w m m m w m ee TITLE: OR/Licensing Submittal: Suppl 1 to NUREG-0737(Generic Ltr 82-33)

NOTES:NI4lSS/FCAF 1cy, 1cy NMSS/FCAF/PM'Lo06/26/73 05000259 NMSS/FCAF 1cy. 1cy NMSS/FCAF/PM. 05000260 OL:06/28/7O NMSS/FCAF 1cy ~ 1cy NMSS/FCAF/PM ~ 0500029b OL:07/02/76 RECIPIENT RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB2 BC 7 7 INTERNAL: ADM<<LFMB 0 IE/DEPER/EPB 3 3 NRR PAULSONpN 1 1 NRR/DHFS/HFEB 5 5 NRR/DHFS/PSRB 1 1 NRR/DL/ORAB 1 1 NRR/DL/ORB5 '5 5 NRR/DSI/CP8 1 1 NOTES'OPIES NRR/DSI/ICSB NRR FILES B

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/EPRPB 1 1 EXTERNAI: LPDR 1 1 NRC PDR NSIC 1 1 NTIS 2 2 TOTAL NUMBER OF COPIES REUUIRED: LTTR 38 ENCL 37

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TENNESSEE VALLEY AVTHORITY CHATTANOOGA. TENNESSEE 3740l 400 Chestnut e

Street Tower II July 30, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

In the Matter of the Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 In accordance with D. B.;Vassallo's June 12, 1984 Confirmatory Order to H. G. Parris on Browns Ferry, we are enclosing the safety analysis, implementation plan, and implementation schedule for the Safety Parameter Display System (SPDS).

Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, anager Nuclear Licensing Subscr ibex'. an+ swor n t before me t a i~4~day of 1984.

Notary Public My Commission Expires Enclosure cc (Enclosure):

U.S. Nuclear Regulatory Commission Region II ATTN: James P. O'Reilly, Regional Administrator 101 Marietta Street;, NW, Suite 2900 Atlanta, Georgia 30323 Mr. R. J. Clark Browns Ferry Project Manager U.S. Nuclear Regulatory Commission 7920 Nor folk Avenue Bethesda, Maryland 20814 84080300f5 840730 PDR ADOCK 05000259 P PDR An Equal Opportunity Employer

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ENCLOSURE BROWNS FERRY NUCLEAR PLANT (BFN)

SAFETY PARAMETER DISPLAY SYSTEM (SPDS)

Introduction This SPDS safety analysis has been prepared to describe the basis on which the selected parameters are sufficient to assess the safety status of each identified function for a wide range of events as well as briefly describe the proposed methods of implementation.

This document responds to the SPDS requirements set forth in Supplement 1 to NUREG-0737, Item 4. 1.a, Page 8, which states (in addition to the requirement described above):

"The minimum information to be provided shall be sufficient to provide information to the SPDS user about:

( 1) Reactivity control (2) Reactor core cooling and heat removal from the primary system (3) Reactor coolant system integrity (4) Containment conditions (5) Radioactivity control."

The SPDS user for BFN dur ing accident situations will either be the shift engineer (SE), assistant shift engineer (ASE), or a senior reactor operator (SRO). The SE, ASE, or SRO monitoring the SPDS will oversee the control room activities within the main operating area and direct the operators whenever necessary. The SPDS will provide the user with a brief synopsis of the plant status, a progression of plant conditions and any approaching challenge to the five identified functions. The SE, ASE, or SRO will have the capability of obtaining a quick assessment, of the plant conditions without interrupting the control board operator and will be able to advise and aid.the operator whenever necessary. Specific operational information will, be provided by the instrumentation at the control boards and the vast majority of the decisionmaking will be done by the control board operators. The SPDS will provide the brief, concise summary of plant conditions required to give the control room SRO an overview of plant conditions.

TVA has initiated a project to replace the process computers for units 1, 2, and 3 with new, state-of-the-art process computer systems. This project is currently underway. To facilitate the implementation of SPDS, TVA intends to implement the SPDS displays in the control room by way of the new process computer display system. Once the process computer changeout is complete, the SPDS displays will be implemented in each of the unit control rooms and will undergo a series of operational tests to verify that the displayed data is current and is as specified. The SPDS implementation will be in accordance with the following schedule; unit 3 cycle 7 1987; unit cycle 8 - 1988, unit 2 cycle 8 1989.

1 III. Safet Anal sis Each of the five functions are discussed below as they apply to BFN.

Parameters sufficient to assess the safety status of the function are also identified.

1. Reactivit Control:

The purpose of this function is to ensure that the core can be maintained in a subcritical condition when required. This is accomplished by ensuring that all rods have been fully inserted into the reactor core. A reactor scram signal from the reactor protection system (RPS) and an all-rods-in indication from the rod position indication system (RPIS) will be available. Nhen all control rods are fully inserted, the reactor power will be reduced below the indicating range of the average power range monitors (APRM). Signals and position from the source range monitors (SRM) will be available as well as'APRM signals.

By verifying from SPDS that all rods are inserted and that the APRMs are downscale, the operators can assure the core is in the subcritical condition. As long as all rods are inserted the,,

reactor will remain subcritical. This is verified at the start of each fuel cycle by performance of the full core shutdown margin (SDM) demonstration.

2. Reactor Core Coolin and Heat Removal From the Primar S stem:

Reactor core cooling is accomplished solely by maintaining the reactor vessel water level above the top of active fuel. The only parameter that the operator can monitor to verify that this is being accomplished is the reactor water level. Cooling water flows or system operability status does not assure the operator that the cooling water is actually being supplied to the reactor. Only the reactor water level can provide core cooling assessment. Thus, the reactor water level is the only parameter needed on the SPDS to assess core cooling.

4 0 Heat is removed from the primary system by releasing steam through the relief valves to the torus or through the steamlines to the high-pressure cooling injection system (HPCI), reactor core isolation cooling system (RCIC), or the main condenser., If the steamlines are isolated, heat cannot be removed by way of steam to HPCI, RCIC, or the main condenser leaving the relief valves to provide the heat removal capability from the reactor vessel. Failure to relieve a sufficient amount of steam will result in an overpressurization of the primaxy system.

Therefore the reactor pressure and reactor water level parameters are sufficient to assess heat removal from the primary system.

3 . Reactor Coolant S stem Inte rit The reactor coolant system integrity .for BFN applies to a breach of the reactor coolant system (RCS) pressure boundary. A breach can be detected by a combination of several parameters. Along with a drop of reactor pressure, a decrease in reactor water level and/or an increase in dxywell pressure indicates a possible breach in the coolant system. Thus, by monitoring the reactor pressure, reactor water level, and drywell pressure, and observing their behavior, the operator can verify reactor coolant system integrity.

4. Containment Conditions:

Containment conditions are monitored to maintain containment design conditions and to maintain a heat sink for heat removal from the primary system as discussed in item 2 'above. The drywell is monitored for two purposes: (1) to detect a possible approach to a containment breach; and (2) to monitor the environmental conditions for equipment protection. The drywell pressure is the key parameter to alert the operator to an approach to the design pressure of the containment which could lead to a breach. Drywell temperature will inform the operator'hat the containment environment is approaching conditions that may degrade drywell equipment performance or exceed containment design tempexature limits. Thus, the parameters necessary for SPDS monitoring for the drywell conditions are drywell pressure and drywell temperature.

The torus is used as a heat sink for heat removal from the primary system during accident situations. Proper thermodynamic conditions within the suppression pool will ensure that an unacceptable energy accumulation within the torus is not taking place and that the heat sink remains available. The suppression pool water temperature will provide the operator all the necessaxy information needed to assess the thermodynamic conditions of the torus.

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5. Radioactivit Control:

In order to have an uncontrolled release of radioactivity, three things must occur: (1) the fuel cladding must degrade to allow radionuclides to escape to the coolant; (2) the radioactive water or steam must be transported to the dryell and/or torus, and (3) a breach in containment must occur in order to have an uncontrolled release to the environs. As long as the fuel cladding, the reactor pressure boundary, and the containment remain within analyzed conditions, any release of',radioactive material is maintained below analyzed levels. Hence, for the purpose of radioactivity contxol for SPDS monitoring, the four functions previously discussed provide full protection against an uncontrolled release to the environs. The parameters identified for monitoring the four functions are more than adequate for monitoring any possible uncontrolled release of radiation.

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Conclusions:==

The parameters identified above for monitoring the five functions specified by Supplement 1 to NUREG-0737 are listed in Table 1 . They will constitute the primary SPDS display. Additional parameters may

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be utilized by SPDS for secondary, on demand, displays to be defined at a latexdate. These secondary displays will be in support of the emergency procedures guidelines. Howevex, the parameters identified in Table 1 provide sufficient information to the SPDS user to assess..

the safety status of each of the five identified functions for a wide range of events. Keeping the primary display as simple as possible reduces chances of confusing the operator.- The goal is to automatically display the minimum number of parameters to the operator in order to monitor the specified functions.

Qd I ~W V t h'fl 4NAA'l4 TABLE 1 PARAMETERS SUFFICIENT TO ASSESS SAFETY STATUS OF THE RIVE FUNCTIONS AND IDENTIFIED BY SUPPLEMENT 1 TO NUREG-0737 Function Parameters Reactivity control'PRM, Scram SRM signal All-rods-in indication Reactor core cooling and heat Reactor water lever removal from the primary Reactor pressure system.

RCS Integrity Reactor pressure Reactor water level Drywell pressure Containment Drywell pressure Conditions Drywell temperature Suppression pool water temperature Radioactivity control All the above parameters