HL-1920, Forwards Design Summary & 10CFR50.59 Evaluation for Planned Addition of Pressure Sensor Actuation for Main Steam Srv,Per 910723 Meeting

Forwards Design Summary & 10CFR50.59 Evaluation for Planned Addition of Pressure Sensor Actuation for Main Steam Srv,Per 910723 Meeting

ML20091K819
Person / Time
Site:	Hatch
Issue date:	01/21/1992
From:	Beckham J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-1920, NUDOCS 9201270092
Download: ML20091K819 (15)

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J, i e,on m. a,. Gl'Of 'Qiu IDWCf vaens u p1,.<$y 4 W:m h ,.tt ill-1920 002495 January 21. 1992 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERAllNG LICENSES DPR-57, NPf-5 MAIN STEAM SAFETY RELIEF VALVE ER[HVRE SENSQR_ACTUAT10))

Gentlemen:

In response to your request following a meeting with representatives of the NRC staff on July 23, 1991, enclosed is the design summary and 10 CFR 50.59 evaluation for our planned addition of pressure sensor actuation for the Main Steam Safety Relief Valves (SRVs). Georgia Power Company (GPC) is in the process of adding a pressure sensor actuated logic system to enhance assurance of SRV actuation at the appropriate pressure setaoint.

Specifically, the system will improve reliability of the SRVs taereby minimizing the potential for forced outages as experienced by Plant Hatch, Units 1 and 2 during February, 1991. GPC plans to install the new systems during the next refueling outages for Units 1 and 2, currently scheduled for Spring 1993 and fall 1992, respectively.

As the new systein is not safety related, it will not be included in the Technical Specifications. However, the system will be designed to meet single failure criteria, and will use equipment and components equivalent to our safety related systems. Maintenance and calibration will be performed consistent with that of our safety related systems.

If you have any questions, please contact this office.

Sincerely,

.l / / \ --

J. T. Beckham, Jr.

JAW /CLT/cr

Enclosures:

(See next page.) .

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PDR ADOCK 00000321 P PDR 4

Geolgk!lbwer U.S. Nuclear Regulatory Commission January 21, 1992 l' age Two

Enclosures:

1. Narrative Design Summary - Sensor-activated SRV Initiation and Simplified Diagrams 111ustrating the New System

2. 10 CFR 50.59 Safety Evaluation cc: Egoraia Power _ComaaDX Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS U.S. Nuclett_RennlA10ry Commission. Washinaton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion 11 Mr. S. D. Ebnctor, Regional Administrator Hr. L. D. Wert, Senior Resident inspector - Hatch I

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e ENCLOSVRE 1 PLANT HATCH - UNITS 1,2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 MAIN STEAM SAFETY REllEF VALVE PRESSVRE SWITCH ACTUATIOR NARRATIVE D151GN

SUMMARY

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Page 1 of 3 NARRATIVLDIll(M_10M&RY SENSOR ACTIVATED SRVs INITIATION OBJECTIVE:

A new sensor initiated logic is being proposed to actuate the Safety Relief Valves (SRV) electrically at their respective mechanical setpoints (Ref:

Figure 1). This will provide a redundant method of preventing overpressurization of the Nuclear Steam Supply System in addition to the mechanical relief mode of the SRV(s).

DISCUS $10N:

The safety objective of the SRV system is to prevent overpressurization of the Nuclear Steam Supply System (NSSS).

Automatically controlled SRV(s) are installed on the main steam lines inside primary containment. The valves are dual purpose in that they will relieve pressure by normal mechanical action or by automatic action of an electro-pneumatic control system. A two position switch is provided in the Main Control Room (HCR) for the control of each SRV. The two positions are "0 PEN" and "AVT0". In the "0 PEN" position, the switch energizes the solenoid operated valve to open the SRV. In the "AUT0" position, the SRV is opened on Automatic Depressurization System (ADS) or Low Low Set (LLS) initiation signals, the normal mechanical action, or the new electrical actuation "backuy" mode.

The relief by normal mechanical action is intended to prevent potential overr essurization of the nuclear steam supply system following a main turbine trip or other pressurization transients. On a loss Of Coolant Accident (LOCA) wherein the High Pressure Coolant injection (liPCI) system can not supply adequate inventory makeup, the depressurization by automatic action is intended to reduce nucicar system pressure to allow Core Spray (CS) or Low Pressure Coolant injection (LPCI) to inject water into the reactor vessel.

The proposed modification will meet the following criteria:

1. 1... ;..a control logic will give the SRV(s) an electrical signal to open at the same setpoints that the valves are supposed to open mechanically.

Calculations will account for instrument loop uncertainties (including setpoint tolerances), thus allowing for a setpoint equivalent to the mechanical actuation setpoint.

2. The proposed addition of control logic will not change the existing function of the SRV(s), ADS, or LLS system.

3. One-out-of-two taken twice logic will minimize the potential for inadvertent actuation of the SRV while providing reliable operation of the SRV at the setpoint independent of mechanical action.

Page 2 of 3 NARRATIYL Dfi[6N

SUMMARY

SENSOR ACTIVATED SRVs INITIATION ,

4. The Emergency Safety System (ESS) Division I to Division 11 interface will be isolated by fuses.

5. A DC power socce will be used.

6. The logic will rematn operable in the event of Loss-of-Offsite Power (LOSP).

7. All new hardware will be procured to meet Class IE (environmental and seismic) requirements as appropriate. Similarly, any existing equipment used will also meet these requirements.

8. No single failure of the new logic shall cause the SRV(s) to lift inadvertently, or prevent the 50V(s) from lifting upon receiving an ADS /LLS signal.

9. This is a non-safety related design change. The proposed modification is intended to be used during abnormal plant operation to increase the reliability of the NSSS depressurization system.

10. One solenoid is provided for each SRV.

11. Calibration and maintenance requirements already established by Plant Hatch Technical Specifications for Nuclear Boiler System (B21) equipment will apply to this proposed modification.

CONCLUS!W:

ESS DIVISION I AND DIVISION II SENSORS (1-out-of-2 Taken Twice Logic):

In order for the Unit I and Unit 2 SRV(s) to actuate, two pressure signals will be required. The Unit 1 SRV(s), set to operate at 1000 psig, will require two signals from the trip units set at 1080 psig. The SRV(s) set to operate at 1090 psig will require one signal from the trip units set at 1080 psig and a second signal from the trip units set at 1090 psig. 1he remaining SRV(s) set to operate at 1100 psig will require one signal from the trip units set at 1090 psig and a second signal from the trip units set at 1100 psig (Ref. Table A, Hatch Unit 1). The Hatch Unit 2 SRV(s) are intended to function in a similar manner. The Unit 2 SRV setpoints are 1090, 1100, and 1110 psig respectively (Ref. Table A, Hatch Unit 2).

This modification will require installation of four new pressure transmitters (pts), two per panel,_on local panels H21-P404 and H21-P405. This modification will require the installation of four new cables (approx. 700' each), one per PT, from MCR panel Hil-P927 and Hil-P928 to the new pts, located on the local instrument panel H21-P404 and H21-P405 elevation 158' of

! the Reactor Building and four conduits (approx. 40' each), one per PT, from

! the new pts to the nearest tray. Eleven cables (approx.100' each) will run l between MCR panels Hil-P602 and Hil-P927. Eleven cables (approx. 50' each) will run between MCR panels Hil-P927 and H11-P928.

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l Page 3 of 3 '

l HARR&LIYLDUlGX_5U!Hi&RY SLNSOR ACTIVATED SRVs INITIATION Two Haster Trip Units (MTVs) and two Slave Trip Units (Slus) will be added per panel, to the Main Control Room (MCR) panels Hil-P927 and Hil-p928. Six existing spare relays (per panel) now located in panels lill-P927 and Hil-P928 will be used.

Figures 1, 2, and 3 show a simplified logic, P&l0, and elementary. Table A shows each SRV and its function, division, setpoint, and relays used to energize the solenoid in the proposed scheme.

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i ENCLOSURE 2 PLANT HA1CH - UNITS 1.2 HRC DPCLETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 MAIN STEAM SAFETY RELIEF VALVE PRESSURE SWITCH ACTUATION 10 CFR 50.51 SAFETY EVALUATIOS 002495

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.10 CFR 50,59 SAFETY EVALUAT1Dli E.1. IIATCil NUC1J.AR p1 ANT l page 1, of 4 Units 1 6 2  !

A. SAFETY SYSTEM APPLICABILITY; The document to which this evaluation applies represents:

1. [X ] Yes [ ] No A change to a safety related equipmt,nt or document?

Basis for answer: The l'uclear Boiler System (B21) Saf ety Relief Valves (SRV) are safety related equipment described in the Unit 1 and Unit 2  ;

FSARs. The proposed modification provides non safety related l sensor initiated, one.out of two, taken twice, logic to  !

actuate the SRV(s), for Units 1 and 2, at their respective  :

mechanical setpoints. The intent is to mintinize the potent tal for overpressurization of the Nuclear Steam Supply System  :

(NSSS), in each unit.

2. [X ] Yes ( ) No A change which could impact safety related equipment or l document?

Basis for answer: The proposed modification is a non safety related design that will meet Class 1E equipment and Seismic Class 1 i requirements. The Unit 1 and Unit 2 FSARs will be revind to reflect the added capability provided by this design c.;ange. '

3. ( ) Yes [X ] No A change-to a system, atructure, or component which handles /

controla radistion hazards?

Basis for answer: The SRV(s) functions and operation do not encompass handling or controlling radioactive effluents, j

-B.- 10CFR$0.59 APPLICABILITY The document to which this evaluation applies represents:

1. [X ] Yes [ ] No A change to the plant as described in the FSAR (i.e., will this change require a revision to some portion of the FSAR?)

Basis for answer: The proposed modifications provide an electrical signal to open the 3RV(s) at the setpoints used to open the valves by mechanical action. Section 4.4 of the Unit 1 FSAR and Section 5.2.2 of the Unit 2 FSAR will be impacted.  !

2. [ ] Yes [X ] No A change to procedures described in the FSAR? (i.e., is the e document a safety related procedure?)

Basis for answer: This evaluation addresses a design change, and not a procedure ,

change. Any plant procedures impacted by this design change,

-and requiring revisions, will be evaluated separately. '

3. [ ] Yes [X ) No A test or experiment not described in the FSAR which affects plant safety?

Basis for answer: This evalauntion addresses a design change, and not a test or .

experiment. l

4. [ ] Yes [X ) No A change to the Technical Specificationc and/or Environmental '

Technical Specifications incorporated in the operating license?

JBasis for answer: .This is a non-safety related inodification intended for use during abnormal plant operation to increase reliability by providins a reliable backup to the NSSS depressurization system. The non safety related equipment is properly isolated froin the safety relateu portion of the B21 system, so no i malfunction of this new equipment will adversely affect the safety related operation of the SRV(s). Therefore, since no licensing credit is sought for this modification, no limiting  ;

conditions for operation or

'SE. doc. 06/29/90 e

l 10 CFR 50.$9 SAFETY EVALUATIQH E.1. HATCH NUCLEAR piANT ppge 2 of 4 surveillance requirements will need to be added to the i

Technical Specifications. Also, since this modification will not impact the operation of any safety related system, no I change to any existing portion of the Technical Specifications 1 will be required, if the answer to all the questions in Sections A and B are "NO", skip section C. 1E the enswer to any question in section A or B is 'YES", complete section C.

C. MFREVIEVED SAFETY OUESTION CRITERIA:

-1. [ ] Yes [X ) No Does the proposed activity increase the probability ot occurrence of an accident previously evaluated in the FSAR7 Basis for answer: The proposed modification will not change the existing function of the SRV(s), ADS, or LLS systems. The sensor initiated logic being proposed to alectrically activate the ,

SRV(s) is in a one out of two taken twice logic scheme. The l worst case failure mode of this proposed logic modification is a short in the logic which causes an inadvertant opening of an SRV, which has previously been analyzed in chapter 15.1 17 of  ;

the Unit 2 FSAR.- The relays, transmitters and associated hardware for this modification will be procured and installed to the necessary 1E, seismic, environmental and/or design standards to ensure proper isolution from the safety related functions of the SRV(s). These required components will be designated as Class 1E to insure replacements are procured and installed to the same stant erds. Therefore, the likelihood of

• an inadvertant SRV opening transient is not significantly '

increased. Thus, this proposed modification does not increase the probability of occurrence of a previously reviewed accident.

2. ( ) Yes [X ] No Does the proposed activity increase the consequences of an .

accident previously evaluated in the FSAR7 Basis for swer; Because there is no change to SRV function, there is no increaso to the consequences of previously evaluated accidev' scenarios described in the Unit 1 and 2 FSARs.. s

3. [ -j Yes [X ] No Does the proposed activity increase the probability of  ;

occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR7

. Basis for answer: All new hardware proposed by this modification will be procurod to meet Class IE requirements (as appropriate). The new sensor initiated-logic is one-out of two, taken twice, so a singic failure cannot cause inadvertent SRV actuation.

Therefore, no increase in the ocetrrence probability of a previously evaluated equipment. malfunction results from the modification. The reliability of the SRV(s) is enhanced by the addition of the proposed modification.

4. ( ) Yes (X) No Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR?

Basis for answer: No single failure of the proposed SRV actuation logic will cause the SRV(s) to lift inadvertently, or prevent lifting upon ADS /LLS activation, t

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5. l ) Yes _

[X) No Does the proposed activity create the possibility of an  !

accident of a different type than any previously evaluated in j the FSAR7  :

Basis for answer: The pSAR evaluates pressurization transients and small pipe i t break 1DCA events, in which proper SRV operation is important, j one out of two twice t,iken logic will mintuire the potential ,

for inadvertent SRV actuation in case of a false signal, while  ;

providing for increased reliabilit.y of operation at the design ,

basis setpoints. Thus, no new types of accidents are j introduced, as a result of this modification, j i

6. [ } Yes [X) No Does the 1roposed activity create the possibility of a '

malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR7 Basis for answer: The ESS Division 1 to Division 11 interface will be isolated  ;

by the addition of fuses, hoss of power to the Units 1 and 2 control rooms, via Ups failure, could result in failure of the  ;

Master / Slave trip units, installed by this proposed i modification. The electrical actuation of the SkV(s) is i i

redundant to the mechanical actuation; therefore, their function is unimpaired. . Because of that redundancy, this -

change does not create the possibility of a malfunct. ion of the  ;

SRV(s) beyond that considered in the FSARs . l 7, [ ] Yes- [X) No Does the proposed act iity reduce the margin of safety as defined in the basis for any Technical Specification?

Basis for answer: As the function of the SRV(s), ADS, and LLS systems remain unchanged, and the addition of the electrical actuation logic L actually enhances SRV operability and reliability, the margin ,

of safety defined in the _ basis for Technical Specification is  ;

not reduced by the proposed modification, _ There is no change ,

to Unit 1 and Unit 2 Environmental Technical Specifications.  ;

There are no acceptance limits increased or failure points  !

decreased due to this proposed modification.

{

t II the answer to ADy of the questions in Section C is "YES", an unreviewed safety question is indicated. Approval from the NRC is required befort the document can be -

used; refer to project procedures for guidance on exceptions to this, r

D, ' TECHNICAL SPECIFICATIONS CONSIDERATIONS:

- l,

[ ).Yes (X ) No Does this document require a change to the Technical Specificat. ions and/or the Environmental Technical Specifications?

Basis-for answer: This is a non safety related modification intended for use during abn,rmal plant operation to increase reliability by r l' _ providing a reliable backup to the NSSS depressurization .

system. . As non safety related equipment, it i_s properly 4 isolated from the safety related portion of the Nuclear Boiler System, so that no malfunction of this new equipment. will l

adversely affect the safety related operation of the system.

Therefore, since no licensing credit is taken for the new  ;

design, no. limiting conditions for operation or surveillance requirements will need to be added to 'he t P--%h-- tr'--Estgv Tert' t' -fr* gowe ---gwiggrrwp,,y9,w+<,gey,-yw-g-gty-gog ge y,,,g,yy -tiy,erw my yei-'use='e>>rm- T'W=e mv' w'Tt'r-'weW7h-w r-*w--m.r**'w wW e-TNNT'-'t'M-'eW 'wf'*1f'uw'1r '

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10 CFR 50.$9 fAFETY I:\%111AT10N E. I . IIATCil NUCilAR p! ANT Pego 4 of 4 Technical Specifications for this modification. Also, since this modification will not itpact '.he operation of any safety related system, no chango to any existing portion of the Technical Specifications will be required.

2. If a change to the Tee belaal Specifications cnd/or the Environmental Technical Specifications is required!

a. Record associated i.* . :ent Change Request (DoCR) Number: To be determined,

b. ( ) Yes [ ] No Locs the amendment have to be inemented gr,1er to use of the document?

c. If the answet to D.2.b van "YES", specify controls on docurnent use:

d. If the answer to D.2.b was "N0", tell why it is permiss.fble to use it prior to implementation of the amendment:

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