ML18033B472

Proposed Tech Specs Revising Level 1 Low Reactor Pressure Vessel Water Level

ML18033B472
Person / Time
Site:	Browns Ferry
Issue date:	08/06/1990
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18033B471	List: ML18033B470 ML18033B472
References
TVA-BFN-TS-291, NUDOCS 9008090106
Download: ML18033B472 (54)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TS 291) 9008090i06 900 PpDR ADOCK 05'000260 PDC

4 UHIT 2 EFFECTIVE PAGE LIST REMOVE IHSERT 1.1/2.1-5 1.1/2.1-5 1.1/2.1-5a 1.1/2.1-10 1.1/2.1-10 1.1/2.1-11 1.1/2.1-11*

3.2/4.2-7 3.2/4.2-7 3.2/4.2-8 3.2/4.2-8*

3.2/4.2-14 3.2/4.2-14 3.2/4.2-15 3.2/4.2-15*

3.2/4.2-23 3.2/4.2-23*

3.2/4.2-24 3.2/4.2-24 3.2/4.2-65 3.2/4.2-65 3.2/4.2-66 3.2/4.2-66*

3.7/4.7-25 3.7/4.7-25 3.7/4.7-26 3.7/4.7-26*

3.7/4.7-33 3.7/4.7-33*

3.7/4.7-34 3.7/4.7-34 3.7/4.7-43 3.7/4.7-43*

3.7/4.7<<44 3.7/4.7-44

• Denotes overleaf or spillover page.

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.112 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.B. Power Transient Tri Settin s To ensure that the Safety 1. Scram and isola- g 538 in.

Limits established in tion (PCIS groups above Specification 1.1.A are 2,3,6) reactor vessel not exceeded, each required low water level zero scram shall be initiated by its expected scram signal. 2. Scram turbine g 10 per-The Safety Limit shall be stop valve cent valve assumed to be exceeded closure closure when scram is accomplished by means other than the 3. Scram turbine g 550 psig expected scram signal. control valve fast closure or turbine trip

4. (Deleted)

5. Scram main g 10 percent steam line valve isolation closure

6. Main steam g 825 psig isolation valve closure nuclear system low pressure C. Reactor Vessel Water Level C. Water Level r Settin s Whenever there is irradiated 1. Core spray and in.

fuel in the reactor vessel, LPCI actuation g 398 above the water level shall be reactor low vessel greater than or equal to water level zero 372.5 inches above vessel zero. 2. HPCI and RCIC Z 470 in.

actuation above reactor low vessel water level zero

3. Main steam 398 in.

isolation above valve closure vessel reactor low zero water level BFN 1.1/2.1-5 Unit 2

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THIS PAGE INTENTIONALLYLEFT BLANK BFN 1.1/2.1-5a Unit 2

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.1. 1'ASES (Cont The safety limit has been established at 372.5 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling.

REFERE CE

1. General Electric BMR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10938.

2. General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.

BFN 1.1/2.1-10 Unit 2

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2.1 BASES

IMI I G TY S ST M SE NGS RELATED T L CLADD NG EG~RI [

The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed throughout the spectrum of planned operating conditions up to the design thermal power condition of 3,440 MHt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR. In addition, 3,293 MHt is the licensed maximum power level of Browns Ferry Nuclear Plant, and this represents the maximum steady-state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and

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axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance.

Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model. The comparisons and results are summarized in Reference l.

The void reactivity coefficient and the scram worth are described in detail in Reference 1.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications as further described in Reference 1. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid inse'rtion of negative reactivity is assured by the time requirements for 5 percent and 20 percent insertion. By the time the rods are 60 percent inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect. The times for 50 percent and 90 percent insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

For analyses of the thermal consequences of the transients a MCPR > limits specified in Specification 3.5.k is conservatively assumed to exist prior to initiation of the transients. This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

BFN 1.1/2.1-11 Unit 2

TABLE 3.2.A PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.

Instrument Channels Operable Per T riR+ys( 1+~11 Functi on Tri L v 1 S tin Ac ion 1 R mark Instrument Channel- > 538" above vessel zero A or 1. Below trip setting does Reactor Low Water Level(6) (8 and E) the following:

(LIS-3-203 A-0) a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel 100 + 15 psig 1 . Above trip set ting i sol ates Reactor High Pressure the shutdown cooling suction (PS-68-93 and -94) valves of the RHR system.

Instrument Channel > 398" above vessel zero 1 Below trip setting Reactor Low Water initiates Hain Steam Level (LIS-3-56A-0) Line Isolation Instrument Channel < 2.5 psig A or l. Above trip setting does the .

High Drywell Pressure (6) (8 and E) following:

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

b. Initiates Primary Containment Isolation

c. Initiates SGTS

TABLE 3.2.A (Continued)

PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.

Instrument Channels Operable Per Tri~S~l~ll} Function Tri Lev 1 tin A i n 1 R m rk Instrument Channel < 3 times normal rated l. Above trip setting High Radiation Hain Steam full power background initiates Hain Steam Line Line Tunnel (6) Isolation Instrument Channel- > 825 psig (4) l. Below trip setting Low Pressure Hain Steam initiates Hain Steam Line Line Isolation (PIS-1-72, 76, 82, 86) 2(3) Instrument Channel- < 140Ã of rated steam flow l. Above trip setting High Flow Hain Steam Line initiates Hain Steam (PdIS-1-13A-D, 25A-D, Line Isolation 36A-D, 50A-D) 2(12) Instrument Channel < 2000F l. Above trip setting Hain Steam Line Tunnel initiates Hain Steam High Temperature Line Isolation.-

Instrument Channel < 100 mr/hr or downscale l. 1 upscale or 2 downscale will Reactor Building a. Ini ti ate SGTS.

Ventilation High b. Isolate reactor zone and Radiation - Reactor Zone refueling floor.

c. Close atmosphere control system.

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TABLE 3.2.B .

INSTRUHENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.

Operable Per Trip. SyS(.l} Function Trip Lev~~lS ~ei~n ~Aien Instrument Channel > 470" above vessel zero. 1. Below trip'setting initiated Reactor Low Hater Level HPCI.

(LIS-3-58A-D)

Instrument Channel > 470" above vessel zero. l. Hultiplier relays initiate Reactor Low Hater Level RCIC.

(LIS-3-58A-0)

Instrument Channel- > 398" above vessel zero. 1. Below trip setting initiates Reactor Low Water Level CSS.

(LS-3-58A-D).

Hultiplier relays initiate LPCI.

2. Hultiplier relay from CSS initiates accident signal (15).

2(16) Instrument Channel- > 398" above vessel zero. 1. Bel ow trip set tings, in Reactor Low Hater Level conjunction with drywell (LS-3-58A-D) high pressure, low water level permissive, ADS timer timed out and CSS or RHR pump running, initiates ADS.

2. Below trip settings, in conjunction with low reactor water level permissive, AOS timer timed out, ADS high drywell pressure bypass timer timed out, CSS or RHR pump running, initiates AOS.

1(16) Instrument Channel > 544" above vessel zero. 1. Below trip setting permissive Reactor Low Water Level for ini ti a ting s i goal s on ADS.

Permissive (LIS-3-184, 185)

Instrument Channel- > 312 5/16" above vessel zero. A l. Below trip setting prevents Reactor Low Water Level (2/3 core height) inadvertent operation of (LIS-3-52 and LIS-3-62A) containment spray during accident condition.

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TABLE 3.2.8 (Continued)

Hinimum No.

Operable Per

~Tri Ss1 Function Tri Level S ttin Action Remarks 2(18) Instrument Channel- 1< p<2.5 psig 1. Below trip setting prevents Drywell High Pressure inadvertent operation of (PIS-64-58 E-H) containment spray during accident conditions.

2(18) Instrument Channel- < 2.5 psig 1. Above trip setting in con-Drywell High Pressure junction with low reactor (PIS-64-58 A-D) pressure initiates CSS.

Hultiplier relays initiate HPCI ~

2. Hultiplier relay from CSS initiates accident signal. (15) 2(18) Instrument Channel- < 2.5 psig l. Above trip setting in Drywell High Pressure conjunction with low (PIS-64-58A-0) reactor pressure initiates LPCI.

2(16) (18) Instrument Channel- < 2.5 psig l. Above trip setting, in Drywell High Pressure conjunction with low reactor (PIS-64-57A-D) water level, low reactor water level permissive, ADS timer timed out, and CSS or RHR pump running, initiates ADS.

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NOTES OR TABLE 2 B

1. Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two OPERABLE trip systems except as noted. If a requirement of the first column is reduced by one, the indicated action shall be taken.

If the same function is inoperable in more than one trip system or the first column reduced by more than one, action B shall be taken.

Action:

A. Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take action B.

B. Declare the system or component inoperable.

C. Immediately take action B until power is verified on the trip system.

D. No action required; indicators are considered redundant.

E. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable channel(s) to OPERABLE status or place the inoperable channel(s) in the tripped condition.

2. In only one trip system.

3. Not considered in a trip system.

4. Deleted.

5. With diesel power, eachRHRS pump is scheduled to start immediately and each CSS pump is sequenced to start about 7 seconds later.

6. With normal power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about 7 sec. with similar pumps starting after about 14 sec. and 21 sec.,

at which time the full complement of CSS and RHRS pumps would be operating.

7. The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure. The'CICS setting of 450" of water corresponds to at least 150 percent above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.

Similarly, the HPCIS setting of 90 psi corresponds to at least 150 percent above maximum steady state flow while also ensuring the initiation of isolation following a postulated break.

8. Note 1 does not apply to this item.

9. The head tank is designed to assure that the discharge piping from the CS and RHR pumps are full. The pressure shall be maintained at or above the values listed in 3.5.H, which ensures water in the discharge piping and up to the head tank.

BFN 3.2/4.2-23 Unit 2

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NOTES FOR'TABLE 2 B ~ont'd)

10. Only one trip system for each cooler fan.

ll. In only two of the four 4160-V shutdown boards. See note 13.

12. In only one of the four 4160-V shutdown boards. See note 13.

13. An emergency 4160-V shutdown board is considered a trip system.

14. RHRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (g 39g" above vessel zero) originating in the core spray system trip system.

16. The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable initiated for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power shall be less than power reduction shall be and reactor 30 percent within four hours.

18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

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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 Specifications are (i) to assure the effectiveness of the protective instrumentation 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 performance.

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 low values are both critical and may have a substantial effect on safety. The setpoints 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 actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation 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 538 inches above vessel zero 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 470 inches above vessel zero (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 g 398 inches above vessel zero (Table 3.2.A) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1). Details of valve grouping and required .closing times are given in Specification 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.

The low reactor water level instrumentation that is set to trip when reactor water level is g 398 inches above vessel zero (Table 3.2.B)

BFN 3.2/4.2-65 Unit 2

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3.2 ~ BASES (Cont'd) 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 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 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.

ADS provides for automatic nuclear steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool.

ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level 1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, and the ADS timer timed out. In addition, at least one RHR pump or two core spray pumps must be running.

The ADS high drywell pressure bypass timer is added to meet the requirements of NUREG 0737, Item II.K.3.18. This timer will bypass the I

high drywell pressure permissive after a sustained low water level. The worst case condition is a main steam line break outside primary containment with HPCI inoperable. With the ADS high drywell pressure bypass timer analytical limit of 360 seconds, a Peak Cladding Temperature (PCT) of 1500 F will not be exceeded for the worst case event. This temperature is well below the limiting PCT of 2200 F.

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 drywell, a trip setting of 140 percent of rated steam flow in conjunction 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 200'F 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 BFN 3.2/4.2-66 Unit 2

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TABLE 3.7.A PRIMARY CONTAINHENT ISOLATION VALVES Number of Power Maximum Action on Operated Valves Operating Normal Ini ti ating Grogp Valve Idgn~if ication Inboard Outboard ~Time ~PP iti n ~nl Hain steamline isolation valves (FCV-1-14, -26, -37, & -51; 1-15,

-27, -38, & -52) 3 < T < 5 GC Hain steamline drain isolation valves (FCV-1-55 & 1-56) 15 GC Reactor water sample line isolation valves (FCV-43-13 & -14) SC RHRS shutdown cooling supply isolation valves (FCV-74-48 & -47) 40 SC RHRS LPCI to reactor (FCV-74-53 & -67) 40 SC Suppression chamber drain (FCV 75-57 & -58) 15 0Pth GC Drywell equipment drain discharge isolation valves (FCV-77-15A & -158) 15 GC Drywell floor drain discharge isolation valves (FCV-77-2A & -28) 15

'These valves isolate only on reactor vessel low low low water level (> 398") and main steam line high radiation of Group 1 isolations.

""These valves are normally open when the pressure suppression head tank is aligned to serve the RHR and CS discharge piping and closed when the condensate head tank is used to serve the RHR and CS discharge piping.

(See Specification 3.5.H)

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Number of Power Haximum Action on Operated Valves 'perating Normal Ini ti ating Valve Iden ti fi cati on Inboard Outboard TimeZmg.). Pepsi i n ~inal Reactor water cleanup system supply isolation valves (FCV-69-1, & 2) 30 GC HPCI warm-up (FCV-73-81) 10 SC HPCIS steamline isolation valves (FCV-73-2 & 3) 20 GC RCICS steamline isolation valves (FCV-71-2 & 3) 15 GC Orywoll nitrogen make-up inlet isolation valves (FCV-76-18) SC Suppression chamber nitrogen make-up inlet isolation valves (FCV-76-19) SC Drywell main exhaust isolation valves (FCV-64-29 and 30) 2.5 SC Suppression chamber main exhaust isolation valves (FCV-64-32 and 33) 2.5 SC Orywell/suppression chamber purge inlet (FCV-64-17) 2.5 SC Drywell purge inlet (FCV-64-18) 2.5 SC

TABLE 3.7.A (Continued)

Number of Power Maximum Action on Grnpp Valve Identification Operated Valves Inboard Outboard T~im ~~

Operating Normal

~Pition Ini ti ating

~in 1 N/A Demineralized water supply check N/A C N/A valve (2-1192)

N/A Demineralized water supply isolation N/A C N/A valve (2-1383)

N/A Service air supply isolation valves N/A N/A (33-1070)

N/A Service air supply check valve N/A N/A (33-785)

N/A Drywell control air inlet header N/A Process check valves (32-2163,32-336)

N/A Drywell control air inlet header N/A Process check valves (32-2516, 32-2521)

N/A Suppression chamber vacuum relief N/A N/A (64-20, 64-21)

N/A Suppression chamber vacuum relief N/A Process check valves (64-800,64-801)

N/A Recirculation pump A seal injection N/A Process check valves (68-508,68-550)

N/A Recirculation pump 8 seal injection N/A Process check valves (68-523.68-555)

N/A Reactor water cleanup system N/A discharge check valve (69-579) Process'C N/A 'eactor building closed cooling N/A water drywell returri isolation valve (70-47)

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OTES FOR ABLE 0

Key: 0 = Open C = Closed SC = Stays Closed GC = Goes Closed Note: Isolation groupings are as follows:

Group The valves in Group 1 are actuated by any one of the following conditions:

1. Reactor Vessel Low Low Low Water Level (g 398")

2. Main Steamline High Radiation

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 following conditions:

1. Reactor Vessel Low Water Level (538")

2. High Drywell Pressure Group 3 ~

The valves in Group 3 are actuated by any of the following conditions:

1. Reactor Low Water Level (538")

2. Reactor Water Cleanup (RWCU) System High Temperature in the main steam valve vault,

3. RWCU System High Temperature in RWCU pump room 2A,

4. RWCU System High Temperature in the RWCU pump room 2B,

5. RWCU System High Temperature in RWCU heat exchanger room,

6. RWCU System High Temperature in the space near the pipe trench containing RWCU piping.

Group 4 ~

The valves in Group 4 are actuated by any of the following conditions:

1. HPCI Steamline Space High Temperature

2. HPCI Steamline High Flow Steamline Low Pressure HPCI

4. HPCI Turbine Exhaust Diaphragm High Pressure Group 5 ~ The valves in Group 5 are actuated by any of the following conditions:

1. .RCIC Steamline Space High Temperature

2. RCIC Steamline High Flow

3. RCIC Steamline Low Pressure

4. RCIC Turbine Exhaust Diaphragm High Pressure BFN 3.7/4.7-34 Unit 2

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3.7/4.7 BASES (Cont'd) conditions. If the efficiencies of the HEPA filters and charcoal absorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal absorbers.

Only two of the three standby gas treatment systems are needed to clean up the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat'o the containment system performance and reactor operation or refueling operation may continue while repairs are being made. If more than one train is inoperable, all fuel handling operations, core alterations, and activities with the potential to drain any reactor vessel containing fuel must be suspended and all reactors placed in a cold shutdown condition, because the remaining train would provide only 50 percent of the capacity required to filter and exhaust the reactor building atmosphere to the stack. Suspension of these activities shall not preclude movement of a component to a safe, conservative position. Operations that have the potential for draining the reactor vessel must be suspended as soon as practical to minimize the probability of a vessel draindown and subsequent potential for fission product release. Draindown of a reactor vessel containing no fuel does not present the possibility for fuel damage or significant fission product release and therefore is not a nuclear safety concern.

4.7.B/4.7.C Standb Gas eatme t S ste a d Seep da Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the standby gas treatment system. Functionally testing the initiating sensors and associated trip logic demonstrates the capability for automatic actuation. Performing'these tests prior to refueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frequent enough to load the filters, thus reducing their reserve capacity too quickly. That the testing frequency is adequate to detect deterioration was demonstrated by the tests which showed shipboard environment on the US Savannah (~0 NL ~). Pressure drop across the combined HEPA filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability, pressure drop and air distribution should be determined at least ence per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform's evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with USAEC Report DP-1082. Iodine removal efficiency tests shall BFN 3.7/4.7-43 Unit 2

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,3.7/8.7 gASES (Cont'dg follow AS'3803. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tra , emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Guide 1.52. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1975. Any HEPA filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.

All elements of the heater should be demonstrated to b'e functional and operable during the test of heater capacity. Operation of each filter train for a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> each month will prevent moisture buildup in the filters and adsorber system.

With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets'epaired and test repeated.

If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significance shall be made by the operator on duty at the time of the incident.

Knowledgeable staff members- should be consulted prior to making this determination.

Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the operable systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Prima Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free. space of 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 LOCA.

~Gros l Process lines are isolated by reactor vessel low water level (l 398") 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 temperature. The reactor water sample line valves isolate only on reactor low water level at

2. 398" or main steam line high radiation.

BFN 3.7/4.7-44 Unit 2

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ENCLOSURE 2

SUMMARY

OF CHANGES (UNIT 2)

1. Revise Safety Limit 1.1.C (Reactor Vessel Water Level).

Existing Safety Limit 1.1.C reads in part:

. water level shall be greater than or equal to 378 inches above vessel zero."

Revised Safety Limit 1.1.C would read in part:

. water level shall be greater than or equal to 372.5 inches above vessel zero."

2. Revise Limiting Safety System Setting (LSSS) 2.1.C.

a. Revise LSSS 2.1.C.1.

Existing LSSS 2.1.C.1 reads in part:

> 378 in. above vessel zero" Revised LSSS 2.1.C.l would read in part:

> 398 in. above vessel zero"

b. Revise LSSS 2.1.C.3.

Existing LSSS 2.1.C.3 reads in part:

> 378 in. above vessel zero" Revised LSSS 2.1.C.3 would read in part:

> 398 in. above vessel zero"

3. Revise bases section 1.1 (Fuel Cladding Integrity Safety Limit).

a. Existing bases section 1.1 reads in part:

. The safety limit has been established at 378 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling. This point is the lower reactcr low water level trip."

Revised bases section 1.1 would read in part:

.The safety limit has been established at 372.5 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling."

b. Add the following to the references of bases section 1.1. "General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2."

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Enclosure 2 Page 2 of 3

4. Revise Table 3.2.A.

a. Delete the asterisk on page 3.2/4.2-7 beside the function "Instrument Channel Reactor Low Water Level (6) (LIS-3-203A-D)".

b. Revise the trip level setting for the function "Instrument Channel Reactor Low Water Level (LIS-3-56A-D)" on page 3.2/4.2-7.

Existing trip level setting reads:

"> 378" above vessel zero" Revised trip level setting would read:

"> 398" above vessel zero"

c. Delete the following note from the bottom of page 3.2/4.2-7.

"*The automatic initiation capability of this instrument channel is not required to be OPERABLE while the Reactor Vessel water level monitoring modification is being performed. Manual initiation capability of the associated systems will be available during that time the automatic initiation logic is out-of-service."

5. Revise Table 3.2.B.

a. Revise the trip level setting for both functions "Instrument Channel Reactor Low Water Level (LS-3-58A-D)" on page 3.2/4.2-14.

Existing trip level setting reads:

"> 378" above vessel zero" Revised trip level setting would read:

"> 398" above vessel zero"

b. Delete the asterisk on page 3.2/4.2-14 beside the function "Instrument Channel Reactor Low Water Level (LS-58A-D)".

c. Delete the following note from the bottom of page 3.2/4.2-14.

"*The automatic initiation capability of this instrument channel is not required to be OPERABLE while the Reactor Vessel water level monitoring modification is being performed. Manual initiation capability of the associated system will be available during that time the automatic initiation logic is out-of-service."

d. Revise note 15 of the notes for Table 3.2.B.

Existing note 15 reads in part:

(> 378" above vessel zero)

Revised note 15 would read in part:

.(> 398" above vessel zero)

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Enclosure 2 Page 3 of 3

6. Revise bases section 3.2.

a. Existing bases reads in part on page 3.2/4.2-65:

low water level instrumentation set to trip at 378 inches above vessel zero (Table 3.2.B)

Revised bases would read in part:

low water level instrumentation set to trip at > 398 inches above vessel zero (Table 3.2.A)

b. Existing bases reads in part on page 3.2/4.2-65:

set to trip when reactor water level is 378 inches above vessel zero .

Revised bases would read in part:

set to trip when reactor water level is > 398 inches above vessel zero

7. Revise a note at the bottom of page 3.7/4.7-25.

Existing note reads:

"The valves isolate only on reactor vessel low low water level (378") and main steam line high radiation of Group 1 isolations."

Revised note would read:

"The valves isolate only on reactor vessel low low low water level

(>398") and main steam line high radiation of Group 1 isolations."

8. Revise notes for Table 3.7.A.

Existing item 1 under Group 1 reads:

"1. Reactor Vessel Low Water Level (378")"

Revised item 1 under Group 1 would read:

"1. Reactor Vessel Low Low Low Water Level (> 398")"

9. Revise bases section 3.7.D/4.7.D (Primary Containment Isolation Valves).

Existing bases (Group 1) reads in part:

reactor vessel low water level (378") in order to allow isolate only on reactor low water level at 378" or main steam line high radiation."

Revised bases (Group 1) would read in part:

reactor vessel low water level (> 398") in order to allow isolate only on reactor low water level at >398" or main steam line high radiation."

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C ENCLOSURE 3 REASONS AND JUSTIFICATION FOR THE CHANGES Reasons for the Chan es During the process of generating setpoint and accuracy calculations for plant parameters for which no calculation basis could be found, it was determined that the trip setting given in the unit 2 technical specifications for the Level 1 low reactor pressure vessel (RPV) water level was not conservative based on the new calculation methodology. Revisions to the unit 2 technical specifications are necessary to incorporate the revised value for this parameter. A detailed description of the proposed changes is provided by Enclosure 2.

'Justification for the Chan es TVA committed in the Nuclear Performance Plan Volume III to ensure that calculations exist to support the safe shutdown basis of unit 2. During the process of generating accuracy and setpoint calculations, a parameter (Level 1 low RPV water level) was determined to have a technical specification trip setting which did not agree with the calculation results and based on new calculation methodology was not conservative.

Methodology for determination of instrument setpoints is addressed by NRC Regulatory Guide (RG) 1.105. RG 1.105 endorses Instrument Society of America (ISA)

Standard ISA-S67.04 1982 "Setpoints for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants" as an acceptable method for ensuring that setpoints stay within technical specification limits.

The trip level setting in the following table is the limiting value that the trip setpoint can have when tested periodically, beyond which the instrument channel is declared inoperable and corrective action must be taken. The analytical limit for the subject parameter was provided by General Electric (GE) and is provided in the following table:

Process Parameter Anal tical Limit Tri Level Settin Level 1 low RPV water level 372.5 inches above 398.0 inches AVZ vessel zero (AVZ)

The Level 1 low RPV water level instruments ensure that the core spray and low pressure coolant injection systems are initiated at low reactor water level to mitigate a loss of coolant accident. They also isolate the main steam lines to prevent inventory loss.

The analytical limit provided by GE was used as a design input to a scaling and setpoint calculation which determined the nominal trip setpoint and -trip level setting based on inaccuracies associated with the instrument loops. The scaling and setpoint calculation meets the guidance of RG 1.105. While the calculation meets its guidance, BFN is not committed to the regulatory guide. The allowance for instrument inaccuracies in determining the actual trip setpoint provides conservative assurance that the trip function will be performed at or before reaching the analytical limit.

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• Enclosure 3 Page 2 of 2 The nominal trip setpoint and trip level setting calculated by the scaling and setpoint calculation shows that the present value in the technical specifications for the subject parameter is not accurate. The current value for Level 1 low RPV water level is approximately 378 inches AVZ. This value is being revised to 398 inches AVZ to account for unmeasurable inaccuracies during calibration. This change is being made in Limiting Safety System Setting 2.1.C, Table 3.2.A, Table 3.2.B, notes for Table 3.2.B, bases section 3.2, Table 3.7.A, notes for Table 3.7.A, and bases section 3.7.D/4.7.D.

Utilizing the calculated trip level setting in the technical specification tables will provide a basis to determine operability and ensures adequate core cooling and minimal fission product release during a design basis event. The current technical specification value of 378 inches is being revised to the analytical limit of 372.5 inches in Safety Limit 1.1.C and bases section 1.1.

The analytical limit has been provided by General Electric and is the safety limit which must not be exceeded. The trip setpoint has been chosen for this parameter and will be established in plant instructions to ensure that the trip level setting in technical specifications is not exceeded.

The proposed change guarantees that core cooling is maintained and fission product loss minimized during a design basis event by ensuring that trips occur within the process parameter value (analytical limit) utilized to confirm the design bases of the plant. The proposed changes are justified because they are based on the value derived by approved calculation means," are sufficiently removed from the operating range such that the incidence of spurious trips should not be increased, and add accuracy to the existing text of the technical specifications.

TVA is also deleting temporary amendment TS 261T. In a letter to NRC dated October 17, 1988, TVA requested this temporary amendment to unit 2 technical specification Tables 3.2.A and 3.2.B to allow the automatic initiation capability of specific reactor low water level instruments to be inoperable during the time the reactor vessel water level monitoring system modification was being performed. NRC approved the temporary amendment in a letter to TVA dated December 15, 1988. This was part of amendment 158 for unit 2. The reactor vessel water level monitoring system modification is complete for unit

2. This change administratively removes the special asterisked notes on page 3.2/4.2-7 of Table 3.2.A and page 3.2/4.2-14 of Table 3.2.B.

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ENCLOSURE 4 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT (BFN)

Descri tion of Pro osed Technical S ecification Chan e The unit 2 technical specifications are being revised as follows.

1. Safety Limit 1.1.C is being revised to incorporate the analytical limit for the Level 1 low reactor pressure vessel (RPV) water level (372.5 inches above vessel zero).

2. Bases section 1.1. is being revised to incorporate the analytical limit for the Level 1 low RPV water level (372.5 inches above vessel zero), to delete a sentence which states incorrectly that this is the trip point, and to add a reference.

3. Limiting Saf'ety System Setting (LSSS) 2.1.C is being revised to reflect the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero).

4. Tables 3.2.A and 3.2.B are being revised to incorporate the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero).

5. Note 15 for Table 3.2.B is being revised to incorporate the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero).

6. A note on page 3.7/4.7-25 of Table 3.7.A is being revised to incorporate the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero) and to clarify that it is the "reactor vessel low low low water level."

7. The notes for Table 3.7.A are being revised to incorporate the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero).

8. Bases sections 3.2 and 3.7.D/4.7.D are being revised to incorporate the new trip level setting for the Level 1 low RPV water level (398 inches above vessel zero).

Basis for Pro osed No Si nificant Hazards Consideration Determination NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92 (c). A proposed amendment operating license involves no significant hazards consideration if to'n operation of the facility in accordance with the proposed amendment would not

1) involve a significant increase in the probability or consequences of an accident previously evaluated, or 2) create the possib'lity of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety.

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Enclosure 4 Page 2 of 3 The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

TVA committed in the Nuclear Performance Plan Volume III to ensure that calculations exist to support the safe shutdown basis of unit 2.

During the process of generating accuracy and setpoint calculations, a parameter (Level 1 low RPV water level) was determined to have a technical specification trip setting which did not agree with calculation results and based on the new calculation methodology was not conservative.

The Level 1 low RPV water level instruments ensure that the core spray and low pressure coolant injection systems are initiated at a low reactor water level () 398 inches) to mitigate a loss of coolant accident and isolate the main steam lines to prevent inventory loss.

No equipment changes are being made. The changes will not affect the probability or consequences of an accident previously evaluated. The actual trip setpoint will remain unchanged.

The design basis accident in chapter 14 of the Final Safety Analysis Report affected by this change is the loss of coolant accident. The change will ensure that the trip level setting for the Level 1 low RPV water level is properly established so that the analytical limit is not exceeded.

Temporary technical specification TS 261T allowed the automatic initiation capability of specific reactor low water level instruments to be inoperable during the time the reactor vessel water level monitoring system modification was being performed. Deletion of this temporary amendment is administrative in nature and will therefore not increase the probability or consequences of an accident previously evaluated.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to the Level 1 low RPV water level trip level setting does not involve any modificatiop to plant equipment. No new failure modes are introduced and no new system interactions have been introduced. The results of a low reactor vessel water level remain as before. The same protective functions will still occur at the Level 1 low RPV water level.

The change to delete TS 261T is administrative in nature. -No new or different kinds of accidents can be created by this change because it conservatively revises the technical specifications to delete an exception to the operability requirements for certain reactor low water level instruments.

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Enclosure 4 Page 3 of 3 The proposed change does not involve a significant reduction in a margin of safety.

The margin of safety will be increased by ensuring that the trip level setting for the Level 1 low RPV water level is sufficiently removed from the analytical limit so that the analytical limit is not exceeded. The trip level setting has been selected utilizing the methodology endorsed by Regulatory Guide 1.105 and provides a conservative assurance that the trip function will occur at or before the analytical limit. The new trip level setting is sufficiently removed from the normal operating range such that the incidence of spurious trips should not be increased. Neither the design nor function of the affected components has been changed.

The proposed change to delete TS 261T will remove an exception to the operability requirements for certain reactor low water level instruments that was needed during the reactor vessel water level monitoring system modification. The modification has been completed.

This change will not reduce the margin of safety but will actually increase requirements.

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