ML18026A899

Proposed Changes to Tech Specs Lowering Group 1 Isolation Setpoint for Reactor Lower Water Level.W/Affidavit

ML18026A899
Person / Time
Site:	Browns Ferry
Issue date:	02/22/1984
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18026A898	List: ML18026A897 ML18026A899
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.16, TASK-TM NUDOCS 8403010313
Download: ML18026A899 (42)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS TVA BFNP TS 196 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 84030i03i3 840222 PDR *DOCK 05000259 p PDR

~ ~

0 UNIT 1 TECHNICAL SPECIFICATIONS

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

l. 1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY PoMer Translenv 0. Power Transient Tri Settin s To ensure that thc Safety Limits Scram and isola- 538 in established in Spccif ication tion (PCIS groups above

1. l.A are nn rxcceded, 2,3,6) reictor loM vessel each required scrnm shall bc wntcr level rero initiated by its expected scram 2. Scram--turbine 5 10 per si8nal. Thc Safety Limit shall stop valve cent valve bc assumed to be exceeded when c1osur e closure scram is accomplished by means other than the expected scram 3, Scram--turbine 550 psig signal. control valve fast clo ura or turbino trip

4. Scram--low con- I 23 inches vacua denser vacuum Hg

5. Scram main 5 1,0 per-steam line cent valve isolation ...=losuro 6, Hain steam isola- ~825 'psig tion valve closure nuclear system low pressure C. lh actor Vessel li'ntcr Level C. Water Level Tri Scttin s

<fh<<n<<ver ther<< is irrndi.ate<i . Core spray and 378 in.

iiwl in thc reactor vrsnrl, . LPCI actuation"- above th<< aeter level shall not be 'reactor low water vessel less than 17.7 in. above thc level rero top of the normal active fuel xone ~ and RCIC h 470 in.

'PCI actuation--reac- above, tor low water level

'easel rero

3. Hain steam isola- ' 378 in.

tion valve above closure reactor vessel low wa ter level rero

TABLE 3 '

PRIMARY Col AI:~"-;iT AND REACTOR BUILDING ISOLATIOS INSTRUNM+TATION Hinimum ho:

instrument Channels Oper'able Punction Tr Level Settin Action 1 Remarks Instrument Channel- 538~ above vessel zero A or 1 Belcw trip setting does the Reactor Lcw 'Mater Level (6) (B and E) following:

a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel >> 100 + 15 psig Above trip setting isolates the Reactor High Pressure shutdown cooling suc ion valves of the RHR sysrem.

Instrument Channel- 2378" above vessel zero A 1. Below trip setting initiates Main Reactor Low Mater Level Steam Line Isolation (LIS 3-56A-Dg SM ~ 1)

Instrument Channel- 2 ' pais A or l. Above trip settirg following:

do s the High Drywell Pressure (6) (B and E)

(PS 64 56A D) a. Initiates Reactor Building Isolation

b. Initiate Primary Containmen Isolation

c. Initiates SGTS Instrument Chanrel 3 times normal rated Above trip setting initiates >min High Radiation Main Steam full power background Steam Line Isolation Line Tunnel (6)

Instrument Channel 825 Psig (4) 1. Below trip setting initiates Mai'n Low Pressure Main St.ea.-. Steam Lire Isolat'on Line 2 (3) Instrument Chanrel 140% of rated steam flow B 1. Above trip setting initiates Main High Plow Main S earn 'ne Steam ~inc Isolation

3. 2 BASES In addition to reactor protection instrumentation which initiates reactor scram, protective instrumentation has been provided which in'itiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The ob)ectives of the Specifi-cations are (i) to assure the effectiveness of the protective instru-mentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate per-When necessary, one channel may be made inoperable for brief 'ormance.

intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2.A which senses the conditions for which iso-

' lation is required. Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolationis connected in a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (538" above vessel zero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water fevel is 109.7" (470" above vessel zero) above the top of the active fuel (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low water level instrumentation set to trip at 17.7" (378" above vessel above the active fuel (Table 3.2.B) closes the Main Steam Isolation Valves, 'ero) the'ain Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1).

Details of valve grouping and required closing times are given in Specf.fica-tion 3.7. These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

3.2 BASES I

The low reactor watet level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operati'on so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addit'on to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentacion; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the dr ywell, a trip setting of 1404 of rated steam flow in con)unction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of isolation valves. The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An al'arm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig.

112

~ '4 ~ 'yL Yltyl ~ ~ ~ ~

HOT ES FOR TABLE 3. 7.A Loy ' ~ Open C Closed SC ~ Stay'e Closed CC ~ Goes Closed Hote: Isolation groupings are ae follova:

Group 1: Th>> valv>>o in Croup 1 are actuated by any one of the f olloving conditions:

1. Reactor Vesee) Lov Mater I.evol (378~<)

Hain Stcaoline High Radiation

3. Hain Steamiine High Plov

4. Hain Staeoline Space High Teoperature

5. Hain Stcaaline Lov Pressure Croup 2: Thc va]ves in Group 2 are actuated by any of the follcwtng 1 condir iona:

Reactor Vee ~ ~ 1 Lov Mater Level (538")

.2. High Dryvell Pressure The valves in Croup 3 are actuated by any of the tolloving Group 3:

conditions:

Reactor Lov Mater Level (538")

2. Reactor Mater Cleanup Systeo High T>>erperatur ~

3. Reactor Mater Cleanup Systen High Drain Teuperature valves I'he Jn Group 4 are actuated by any of the folloving Group 4:

condit lone '.

1. HPCI Steenline Space High Teuperature

2. HPCI St>>amlins High F1ov

3. HPCI Steaoline Lov Pressure valves in Croup 5 are actuated by any of the Iolloving Croup 5: Th>>

condition:

I. RCIC St>>a~line Spac>> High Temperature

2. RCIC Steaol'n>> High Plov RCIC St>>saline Lov Pressure valv>>s in Croup 6 are actuated by any of the folloving" Group 6: Th>>

conditions:

Reactor Vr ~ eel Lov Mater Level (538")

2. High Dryvell Prcssure 3, Reactor Building Ventilation High Radiation f54

ra

~Grou 1 Process lines are isolated by reactor vessel low ~ater level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in group 1, except the reactor water sninple line valves, are also closed when process instru-mentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Grou 2 Isolation valves are closed by reactor vessel low water level (538") or high drywell pressure. The group 2 isolation signal also "iso>>

lates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.

~Grnu 3 Process lines ara normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non>>safety related causes. To protect the reactor from a possible pipe break in the sysrem, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.

Grou s 4 and 5 Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable.

~Grou 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water levy 1 (538"), high drywell pressure, or reactor building ventiPation high radiation which would indicate a possible accident and necessitate. primary containment isolation.

Cr o y - proc a l! ea are cloned only on reactor loo neer le el (i70'y.

These close on tile oayoe signal that initiatea HPCIS and RCICS to ensure tltat the valves are not open When HPCIS or RCICS action ia required.

Croup g - iine (traveling in-core probe) is isolated on high dryvell prea-

~ ure. This ia to assure that this line does not provide a leakage path vhcn <<ontainplent pressure indicatea a poaeible accident condition.

The maximum closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released

,fission products following pipe breaks inside the primary containment.

In satisfying this design intent, an additional margin has been included in

'specifying maximum closure times. This margin permits identification of degraded valve performance prior to exceeding the design closure times.

e 277

UNIT 2 TECHNICAL SPECIFICATIONS

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

l. l FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY PIIMcr Tr;lns I I'lu II. Foyer Transient Tri Settin s To ensure th;It rhc Safety Limit s Scram and isola 538 in established in Sprcification tion (PCIS groups above

l. L.A arc not rrcccded, 2 ',6)r 1rvel Matc reartor loM vessel zero e;IclI rcquircJ scram shall bc initiated by its expected scram 2. Scram--turbine 5 10 per-s iI',nal ~ Tile Safety Limit shel 1 stop valve cent valve be assumed to be exceeded Mhen c1osur e closure scram is accomplished by means other than thcexpected scram 3, Scram->>turbine 550 ps il, signal. control valve fast clo~ura or turbine trip

4. Scram--low con- 2 23 inches denser vacuum Hg vacuum

5. 'Scram--main 5 10 per-steam line cent valve isolation .. =losur~

6. Main steam isola>> ~ 825 psig tion valve closure nuclear system low pressure C. I!I actor Vcsscl I4atcr Level C. Ilatcr Level Tri Settin IIII<<llvvur thr r<< is irradi;Itcd . Core spray and 2 378 iIIvl in thc reactor vrsnr 1, LPCI actuation in.'bove th<<un'<<r level "hall not be reactor low water vessel less than 17.7 in. above thc level zero top of the normal active fuel zone. 2, HPCI and RCIC 470 in ~

actuation--reac- above tor low water vessel level zero 3, Hain steam isola- 2 378 in.

tion valve above closure--reactor vessel low water level zero

TABLE 3e 2~A PRIMARY COR AI:~:-:i AND REACTOR BUILDING ISOIATIOR IHSTRUNMTATIOH Minimum KO.

Instrument Channels 'Oper'able

'((( Function Tr Level Settin Action 1 Rema rks Instrument Channel- 538~ above vessel sero A or Belcw trip setting does the Reactor Lov Mater Level (6) (B and I) fel loving(

a. Initiates Reactor Building Isolation b Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel- 100 + 15 psig 1~ Above trip setting isolates the Reactor High Pressure shutdown cooling suc son valves of the RHR system.

Instrument Channel- > 378" above vessel sero A 1 Belch trip setting initiates Main Reactor Low Mater Level Stea~ Line Isolation (LIS 3-56A D~ SM ~ 1)

Instrument Channel- S 2.5 pa($ A or l. Above trip settirg "oes the folloving:

High Dryvell Pressure (6) (B and E)

(PS-6%-56A-D) a. Initiates Reactor Building Isolation b Initiates Primary Containment Iso)ation

c. Initiates R TS Instrument Chanr el 3 times normal rated 1. Above trip setting initiates Main High Radiation Main Steam full power background Steam Line Isolation Line Tunnel (6)

Instrument Channel- > S25 psig (u) 1. Helot trip setting initiates Main Low Pressure Main Stea.-. Steam Lire Zsolat on Line 2 (3) Instrument Chanrel- lu0$ of rated steam f1~ B Above trip setting initxates Main High Plov Main Steam 'ne Steam Line Isolation

3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifi-cations are (i) tc assure the effectiveness of the protective instru-mentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate per-formance. When necessary, one channel may be made inoperable fo'r brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2.A which senses the conditions for which iso-lation is required. Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (538" above vessel zero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains 'and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 109.7" (470" above vessel zero) above the top of the active fuel (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure oi the respective drain valves (Group 7).

The low water level instrumentation set to trip at 17.7" (378" above vessel zero) above the active fuel (Table 3.2.B) closes the Hain Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group I).

Details of valve grouping and required closing times are given in Specifica-tion 3.7. These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

3.2 BASES l

The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels wer e chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addit'on to initiating CSCS, causes isolation of Groups 2 and 8 isolation valves. 'For the breaks it discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

Ventur is are provided in the main steam lines as a means of measur ing steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function 'of the instrumentation is to detect a break in the main steam line. For the woist case accident, main steam line break outside the dr ywell, a trip setting of 140$ of rated steam flow in con)unction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000oF, and release oi radioactivity to the environs is

- well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of isolation valves. The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in 'the conti ol-rod drop accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig.

112

AI. '4 ~ U QI p '<<+j ~ I HOTES FOR TABLE 3. 7,A Xey: 0 <<Open C << Closed SC << Stays Closed CC << Goes Closed Note'. 'Isolation groupings are as tolloua:

Croup 1: The valveo in Croup 1 are actuated by any one of the f olloving conditions:

1. Reactor Yesse) Lov Mater Level (378 ')

2 ~ Main Stean',ine High Radiation

3. Main SteasLline High Flov

4. Main Steanline Space High Teaperature 5, Hain Steanline Lov Pressure Croup 2: The valves in Group 2 are actuated by any of the following conditions:

Reactor Vessel Lou Mater Level (538")

2, High Dryvell Pressure Croup 3: The valves in Croup 3 are actuated by any of the follcnring conditions:

1, Rehctor Lov Mater Level (538")

2. Reactor Mater Cleanup Systen High Teerpersture

3. Reactor Mater Cleanup Systen High Drain Teaperaturn Croup 4: The valves In Croup 4 are actuated by any of the folloving conditions:

1. HPCI Steenl.ine Space High Temperature

2. )PCI Steanii~e High Flov

3. HPCI Steamline LoM Pressure Croup 5: The valves in Croup 5 are actuated by any of the folloving cond' ion-:

1. RC I" 5 t ca~line Space High Temperature

2. RCIC Stesaline High Flov

3. RCIC Steaaline Lou Pressure Group 6: The valves in Group 6 are actuated by any of the follcwing conditions '.

1 ~ Reactor Vessel Lov Mater Level (538")

2, High Dryvell Pressure

3. Reactor Buildiag Ventilation High Radiation

~Grou 1 Process lines are isolated by reactor vessel low water level (3/8") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in group 1, except t reactor water saolple linc valves, are also closed when process instru-mentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature. The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Grou 2 isolation valves are closed by reactor vessel low water level (538") or high drywell pressure. The group 2 isolation signal also "iso-lates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 isolation signal by' transient or spurious signal.

Grn~u i 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywc] I prcssure res>>l ti>>g fr>>m no>>-"afety related ca>>ses. To protect the reactor from a possil>lo pipe break in the system, isolation is provided by high temperature in the clca>>-

Up system area or high flow through the inlet to thc cleanup'system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.

the consequences of an accident which results in the isolation of other

. process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable,

~Grou 6 Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation.

a~or - proc ~ ~ ll ca are closed only on reaecor lo Peer lc cl (ayd'i, There cloae on the oadlpe ~ igyl ~ I that initiates HPCIS and RCICS to ensure that the valvcr rre not open bdhen HFCIS or RCICS action is required.

Crau~8 - I<ne (traveling In-ccrc probe) i ~ isolated on high d".yvell prea-

~ ure. Thi ~ is to araure that this line doer not provide a leakage path vhcn contatnyaent pressure indicater a porrible accident conditions The maximum'closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary 'containment.

In satisfying this design intent, an additional margin has been included 'in specifying maximum closure times. This margin permits identification of degraded valve performance prior to exceeding the design closure times.

277

UNIT 3 TECHNICAL SPECIFICATIONS

0

~

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FllEL CLADDING INTEGRITY 2~1 FUEL CLADDING INTEGRITY PoMcr Transient B. Power Transient Tri Settings To ensure tinct the Safety Limits 1. Scram and isola- 538 in.

established in Specification tion (PCIS proups above

1. 1.A are not exrci dud 2,3,6) reactor low vessel each required scram shall bc watt r level zerv initiated by its expected scram 2. Scram--turbine 5 l0 per-signal. The Safety Limit shall stop valve cent val "e be assumed to be .exceeded when closure closure scram is accomplished hy means other than thc expected scram Scram--turbine

'50 psig signal. control valve fast clo"ura or turbine trip

4. Scram--low con- 2 23 inches denser vacu'um Hg vacu~l

5. Scram--main 5 1Q per-steam line cent va'.

isolation "losuro

6. Hain steam isola- ~825. psi",

tion valve closure

--nuclear system low prcssure C. Reactor Vessel l4atcr Level C. Water Level Tri Scttinps Nivncvcr there is irradiated Core spray .and 2 378 in.

fuvi in thc rca<<tor vessel, LPCI actuation-- above thc water level shall not l>e 'eactor low'water vessel'ero less than'7.7 in. above thc level top of the normal active fuel zone ~ 2. HPCI and RCIC in.

actuation"-reac- above tor low water vessel level zero 3~ Hain steam 2378 in.

valve isola-'ion above closure--reactor vessel low water level zero 13

t TABLE 3.2.A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOIATION INSTRUMENTATION Minimum Yio.-

Instrument Channels Operable er Tri S.s(l) gy Function Tri Level Settin Action 1 Remarks 2 Instrument Channel 2 538> above vessel cero A or 1. Below trip setting does the Reactor Low Mater Level (6) (B and E) following:

a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel- 100 + 15 psig D 1. Above trip setting i.solates the Reactor Hiqh Pressure shutdown cooling suction valves of the RHR system.

Instrument Channel- 2 378" above vessel zero A 1. Below trip setting initiates Main Reactor Low Mater Level Steam Line Isolation (LIS-3-56A-Dg SM 81)

Instrument Channel- 5 2.5 Pais A or l. Above trip setting does the High Drywell Pressure (6) (B and E) fol lowing:

(ps-6a-56A-D) a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel- S 3 times normal rated 1. Above trip setting initiates Hain Hiqh Radiation Main Steam full power background ,Steam Line Isolation Line Tunnel (6) 2 Instrument Channel- 825 psig (a) 1. Below trip setting initiates Main Low Pressure Main Steam Steam Line Isolation Line 2 (3) Instrument Channel- la0% of rated steam flew B 1. Above trip setting initiates Main High Flow Main Steam Line steam Line Isolation 2 (l2) Instrument Channel- 2004 F 1. Above trip setting initiates Main Steam Line Tunnel Main Steam Line Isolation.

High Temperature

3. 2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, pro tec tive ins trumenta tion has been provided which initiates action to mitigate the conse'quences of accidents which are beyond the operator's ability to control, or terminates operator errors

~ before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The obgectives of the Specifi-cations are (i) to assure the effectiveness of the protective instru-mentation when required by preserving its capability to tolerate a singLe failure of <<ny component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate per-formance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and 1ow values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mei>tation 'shown in Table 3.2.A which senses the conditions for which iso-lation is required. Such instrumentation must be available whenever primary containment integrity is required.

The instrumentatlon which initiates primary system isolation is connected

$ n a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (538" above .

vessel zero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 109.7" (470" above vessel zero) above the top of the active fuel (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low" water level instrumentation set to trip at 17.7" (378" above vessel zero) above the active fuel (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (G'oup 1).

Details of valve grouping and required closing times are given in Specifica-tion 3.7. These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

1'08

~2BASES The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray 'Pumps, contributes to ADS initiation, and starts the diesel generators.,

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be aocomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete break of a 28-inch recirculation line and with the"'ircumferential trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating 'CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line, For the worst case accident, main steam line break outside the dr ywell, a trip setting of 140$ of rated steam flow in con)unction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000oF, and release of'adioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of'solation valves. The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For .large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

High radiation monitors in the main steam line tunnel have been pi ovided to detect gross fuel failure as in the control, rod drop accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is pr ovided also.

109

HOTES FOR TABLE 3. 7. A Key: 0 ~ Open C ~ Closed SC ~ Stays Closed GC ~ Goes Closed Note: Xsolation groupings are as follows:

Group 1: The valves in Group 1 are actuated by any of the fol lowing conditions:

1.

2.

Reactor Vessel Main Steamline Low Wa ter Level High Radiation (37@ t)

3. Main Steamline High Flow

4. Main Steamline Space High Temperature

5. Main Steamline Low Pressure Group 2: The valves in Group 2 are actuated by any of the fol lowinq conditions:

1. Reactor Vessel Low Mater Level (538<<)

2. High Drywell Pressure Group 3: The valves in Group 3 are actuated by any of the following conditions:

1 'Reactor Low Mater Level (538>)

2. Reactor Water Cleanup System High Temperature

3. Reactor 'Water 'Cleanup System High Drain Tempera ture Group 4; The valves in Group 4 are actuated by any of the C

following conditions:

1. HPCX steamline Spac'e High Temperature

2. HPCX Steamline High Flow

3. HPCX steamline Low Pressure Group 5: The valves in Group 5 are actuated by any of the following conditions:

1. RCXC Steamline Space High Temperature

2. RCXC Steamline High Flow

3. RCXC Steamline Low Pressure Group 6: The valves in Group 6 are actuated by any of the following conditions:

1. Reactor Vessel Low Water Level (538>)

2. High Drywell Pressure

3. Reactor'uilding Ventilation High Radiation

'Group 7: The valves i.n Group 7 are automatically actuated by 266

3.7.D/4.7.D Primar Containment Isolation Valves Double isolation valves ar e provided on lines penetrating the primar y containment and open to the free space'f the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.

Group 1 - Process lines are isolated by reactor vessel low water level.

(378") in order to allow for removal of decay heat subsequent to a scram yet isolate in time for proper operation of the core standby cooling systems. The valves, in group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high The reactor water sample line valves isolate only on reactor 'emperature.

low water level at 378" or main steam line high radiation.

~Grou 2 - Xsolation valves are closed by reactor vessel low water. level (538") or high drywell pressure. The group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.

~Grou 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell from nonsafety-related causes. To protect the reactor from a pressure'esulting possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.

the consequences of an accident which results in the isolation of other

, process lines. The signals which initiate isolation of groups 4 and 5 process lines are therefor e indicative of a condition which would render them inoperable.

~Grou 6 - tines are connected 'to the primary containment but not directly; to the reactor vessel. These valves are isolated on reactor low water ,"

level (538"}, high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary c'ontainment isolation.

process lines ase closed or RCXCS action is line

~ at &e valveS are not oPen when HPCTS drywell pressure (tnavelin provide a 1 e &age path when 'X~ent s is to assure that this line does no possible accid en condi&on. con t pressure in@.ca es a 2g4

ENCLOSURE 2 I

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA BFNP TS 196 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 Descri tion of Amendment Re uest

-The amendment would lower the group 1 isolation setpoint for reactor low water level from 470 inches above vessel zero to 378 inches above vessel zero. The containment isolation valves involved are the main steamline isolation valves, the main steamline drain isolation valves, and the reactor water sample isolation valves.

Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48FR 14870). One example of actions involving no significant hazards consideration includes:

(vi) a change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to a system component specified in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method."

The abnormal operational transients events defined in Chapter 14 of the FSAR which could be affected by the proposed change in setpoint were reanalyzed. The results of the reexamination of the plant specific safety evaluation show that the lowering of the group 1 water level isolation setpoint will not have any adverse effects on MCPR, peak vessel pressure, radiation release, equipment integrity or shutdown capability.

In turn, the lower setpoint will delay or eliminate reactor isolations for certain events'allowing additional time for stabilization of the system. The overall effect will be a reduction in the number of safety/relief valve ohallenges. The Commission has endorsed reducing safety/relief valve (S/RV) challenges per NUREG-0737, Item II.K.3.16.

Therefore, since the application for amendment involves a proposed change that is similar to an example for which no significant hazards consideration exists, we have made a proposed determination that the application involves no significant hazards considerations.

~ ~

ENCLOSURE 3 DESCRIPTION AND JUSTIFICATION TVA BFNP TS 196 Units 1 and 2 - Pa es 11 55 111 112 254 and 277 Unit - Pa es 1 57 108 109 266 and 294 The group 1 isolation setpoint for reactor low water level is being lowered from 470 inches above vessel zero to 378 inches -above vessel zero. The valves involved are the main steamline isolation valves, the main steamline drain valves, and the reactor water sample valves. The technical specifications and bases are being revised accordingly.

By lowering the MSIV water level trip setpoint to 378 inches above vessel zero, reactor isolations for certain events will be delayed or eliminated. This will allow the operator additional time to stabilize the system and will also prolong the availability of the main steam system to the feedpumps and to the condenser as a heat sink. This will in turn reduce the number of safety/relief valve challenges as recommended by the NRC staff in NUREG-0737, item II.K.3. 16.

A safety analysis performed by the General Electric Company (GE) showed that the lowering of the group 1 water level trip setpoint will not have any adverse effects on MCPR, peak pressure, radiation release, equipment integrity, or shutdown capability. That safety analysis is provided in GE proprietary report NEDE-30012, December 1982 (enclosure 5).

ENCLOSURE 4 AFFIDAVIT OF PROPRIETARY INFORMATION FOR GE REPORT NEDE-30012

GENERAL ELECTRIC COMPANY AFFIDAVIT I, Joseph F. Quirk, being duly sworn, depose and state as follows:

l. I am Manager, BWR Systems Licensing, General Electric Company, and have been delegated the function of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withholding.

2. The information sought to be withheld is contained in the document NEDE-30012 entitled "Browns Ferry Nuclear Plant Units 1, 2 and 3 MSIV Water Level Isolation Setpoint Reduction".

3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement Of Torts, Section

'his definition provides:

757.

"A trade secret may consist of any formula, pattern, device or compilation of information which is used in one's business and which gives him an opportunity to obtain an advantage over who do not know or use it.... A substantial 'ompetitors element of secrecy must exist, so that, except by the use of improper means, there would be difficulty in acquiring informa-tion.... Some factors to be considered in determining whether given information is one's trade secret are: (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3) the extent of measures taken by him to guard the secrecy of the information; (4) the value of the information to him and to his competitors; (5) the amount of effort or money'.expended by him in developing the information; (6) the ease or difficulty with which the information could be properly .

acquired or duplicated by others."

4. Some examples of categories of information which fit into the definition of proprietary information are:

Information that discloses a process, method or apparatus where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information consisting of'upporting data and .analyses, includ-ing test data, relative to a process, method or apparatus, the application of which provide a competitive economic advantage, e.g., by optimization or improved marketability; C. Information which if used by a competitor, would reduce his expenditure of resources or improve his competitive position in h

0 the design, manufacture, shipment, installation, assurance of quality or licensing of a similar product;

d. Information which reveals cost or price information, production capacities, budget levels or commercial strategies of General Electric, its customers or suppliers;

e. Information which reveals aspects of past, present or future General Electric customer-funded development plans and programs of, potential commercial value to General Electric;

f. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection;

g. Information which General Electric must treat as proprietary according to agreements with'ther parties.

In addition to proprietary treatment given to material meeting the standards enumerated above, General Electric customarily maintains in confidence preliminary and draft material which has not been subject to complete proprietary, technical and editorial review.

This. practice is based on the fact that draft documents often do not appropriately reflect all aspects of a problem, may contain tentative conclusions and may contain errors that can be corrected during normal review and approval procedures. Also, until the final document is completed it may not be possible to make any definitive determination as to its proprietary nature. General Electric,'is not generally willing to release such a document to the general public in such a preliminary form. Such documents are, however, on occasion furnished to the NRC staff on a confidential basis because it is General Electric's belief that it is in the public interest for the staff to be promptly furnished with significant or potentially significant information. Furnishing the document on a confidential.

basis pending completion of General Electric's internal review permits early acquaintance of the staff with the information while.

protecting General Electric's potential proprietary position and permitting General Electric to insure the public documents are technically accurate and correct.

Initial approval of proprietary treatment of a document is made by the Subsection Manager of the originating component, the man most likely to be acquainted with the value and sensitivity o'f the information in relation to industry knowledge. Access to such documents within the Company is limited on' "need to know" basis and such documents at all times are clearly identified as proprietary.

The procedure for approval of external release of such a document is reviewed by the Section Manager, Project Manager, Principal Scientist or other equivalent authority, by the Section Manager of the cognizant Marketing function (or his delegate) and by the Legal Operation for technical'ontent, competitive effect and determination of the accuracy of the proprietary designation in accordance with the standards enumerated above. Disclosures outside General Electric

~

are generally limited to regulatory bodies, customers and potential customers and their agents, suppliers and licensees only in accordance with appropriate regulatory provisions or proprietary agreements.

8. The document mentioned in paragraph 2 above has been evaluated in accordance with the above criteria and procedures and has been found to contain information which is proprietaTy and which is customarily held in confidence by General Electric.

9. 'he information noted in the document given in Paragraph 2 that is considered proprietary to General Electric includes the results of a feasibility study of reducing the MSIV water level setpoint for the Browns Ferry Nuclear Plant Units 1, 2 and 3.

10. The information, to the best of my knowledge and belief, has consistently been held in confidence by the General Electric Company, no public disclosure has been made, and it is not availabe in public sources.

Also, disclosures to third parties have been made pursuant to regulatory provisions for proprietary agreements which provide for maintenance of the information in confidence.

'll. Public disclosures of the information sought to be withheld is likely to cause substantial harm to the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities because approximately 6 manmonths and

$ 100,000 in analysis costs were required to obtain the information.

STATE OF CALIFORNIA )

COUNTY OF SANTA CLARA )

Joseph F. Quirk, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and

'belief.

Executed at San Jose, California, this LLI day of 198$ .

Jo p F. Quirk Ge eral Electric Company Subscribed and sworn before ee this I o day of +~~~g198$

CSSCSDCS9GX5 Ce CSSCCS9CSDCSSCO CSSCSSCShII OFFICIAL SEAL KAREN S. VOGELHUBER I

NOTARY PUBLIC ~ CALIFORNIA ge SANTA CLARA COUNTY My Commission Expires Dec. 21, 1984 ARY PUBLIC, STAT F CALIFORNIA I Q4Q4Q4Q4Q4Q4Q4QCQ4Q4eQ4Q4Q485 CHscal/K02093

ENCLOSURE 5 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION IN SUPPORT OF TVA APPLICATION TVA BFNP TS 196 BROMNS FERRY NUCLEAR PLANT