ML18039A798

Application for Amends to Licenses DPR-52 & DPR-68,reducing Allowable Value Used for Reactor Vessel Water Level - Low, Level 3 for Several Instrument Functions

ML18039A798
Person / Time
Site:	Browns Ferry
Issue date:	06/03/1999
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18039A799	List: ML18039A798 ML18039A800
References
TVA-BFN-TS-397, NUDOCS 9906150078
Download: ML18039A798 (29)

Text

CATEGORY 1 REGULATO S INFORMATION DISTRIBUTION STEM (RXDS)

ACCESSION NBR:9906150078 DOC;DATE: 99/06/03 FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee NOTARIZED: YES DOCKET 05000260 I

i50-'496 Browns Ferry Nuclear Power Station,,Unit 3, Tennessee 05000296 AUTH.. NAME AUTHOR Tennessee AFFILIATION'BNEY,T.E.

Valley Authority RECIP.NAPPE RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)

SUBJECT:

Application for amends to licenses DPR-52 & DPR-68,reducing allowable value used for reactor vessel water level - low, level 3 mfor several instrument functions.

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TITLE: TVA Facilities - Routine Correspondence NOTES:

RECIPIENT .

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0 XLE-CENTER RES/DE/SSEB/SES 1,1 1 1 EXTERNAL: NOAC 1 1 NRC D

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'E NOTE TO ALL "RIDS" RECIPIENTS:

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Tennessee Valley Authority. Post Office Box 2000, Decatur, Alabama 35609 June 3, 1999 TVA-BFN-TS-397 10 CFR 50. 90 U.S. Nuclear Regulatory Commission ATTN; Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket Nos. 50-260 Tennessee Valley Authority ) 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 AND 3 TECHNICAL SPECIFICATIONS CHANGE (TS) NO. 397 REQUEST FOR LICENSE AMENDMENT TO LOWER THE ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL LOW, LEVEL 3

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In accordance with the provisions of 10 CFR 50.4 and 50.90, TVA is submitting a request for an amendment (TS-397) to licenses DPR-52 and DPR-68 to change the TS for Units 2 and 3. The proposed change will reduce the Allowable Value (Av) used for Reactor Vessel Water Level Low, Level 3 for several instrument functions.

The primary purpose of this proposed TS change is to reduce the likelihood of unnecessary reactor scrams and the resultant engineered safety feature actuations by increasing the operating range between the normal reactor vessel water level and Level 3 trip functions. The increased range will provide additional time for operators or automatic features to respond to recoverable transients and, thus, may avert unnecessary reactor scrams.

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U.S. Nuclear Regulatory Commission Page 2 June 3, 1999 Industry studies have identified low water level scrams as being initiators of a significant number of plant trips. The Boiling Water Reactor Operating Group, Scram Frequency Reduction Committee identified some of these scrams as unnecessary, since the reactor water level would have stabilized above the top of active fuel and recovered to normal level even without the scram. To provide relief from unnecessary scrams, a possible solution is to lower the instrument Av at which the scram will occur. The safety analysis in Enclosure 1 shows that the Av may be lowered without: adversely affecting the plant response to postulated transients and accidents. to this letter also provides the description and evaluation of the proposed change. This includes TVA's determination that the proposed change does not involve a significant hazards consideration, and is exempt from environmental review. Enclosure 2 contains copies of the appropriate marked-up TS pages from Units 2 and 3 showing the proposed changes.

The BFN Plant Operations Review Committee and the BFN Nuclear Safety Review Board have reviewed the proposed change and determined that operation of BFN Units 2 and 3 in accordance with the proposed change will not endanger the health and safety of the public. TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this lett:er and enclosures to the Alabama State Department of Public Health.

U.S. Nuclear Regulatory Commission

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Page 3 June 3, 1999 Due to the risks associated with online implementation of this change, TVA requests that the revised TS be made effective during the Unit 3, cycle 9 refueling outage (April, 2000) and by the Unit 2, Cycle 11 refueling outage (March, 2001) for Units 2 and 3, respectively. To support this schedule, TVA requests NRC approval by February 1, 2000. If you have any questions about this change, please telephone me at (256) 729-2636.

T. E. ney '

Manager of ng and Indu try Af airs Subscribed and sw rn to before me on this 3rd g d of June 1999.

Notary Public My commission Expires 09/22/2002 i'c: See Page 4

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U.S. Nuclear Regulatory Commission Page 4 June 3, 1999 Enclosures cc (Enclosures):

Mr. William 0. Long, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North-11555 Rockville Pike Rockville, Maryland 20852 Chairman Limestone County Commission 310 West Washington Street Athens, Alabama 35611 Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S. W.

Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS-397 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE I. DESCRIPTION OF THE PROPOSED CHANGE The proposed change will lower the current Reactor Vessel Water Level Low, Level 3 Allowable Value (Av) in the Units 2 and 3 TS for several instrument functions. The following specific TS functions are affected by this proposed change:

~ Reactor Protection System Actuation (SCRAM)

~ Primary Containment Isolation (including Shutdown Cooling System Isolation)

~ Reactor Water Cleanup (RWCU) System Isolation

~ Secondary Containment Isolation and Standby Gas Treatment (SGT) System Initiation

~ Control Room Emergency Ventilation (CREV) System Initiation

~ Automatic Depressurization System (ADS) Reactor Vessel Water Level Confirmatory Signal The proposed change will provide an additional 10 inches of operating range between the normal reactor vessel water level and the level used for initiation of the above functions. The increased range will provide additional time for operators or plant systems to automatically respond to recoverable transients such as feedwater system malfunctions and, thus, may avert unnecessary reactor scrams. This change will similarly reduce the likelihood of the above engineered safety feature (ESF) system actuations without increasing the consequences of events that rely upon these functions.

The proposed changes to the TS are listed below.

Enclosure 2 contains copies of the appropriate marked-up TS pages for Units 2 and 3 showing the changes. The TS changes are the same for both Units 2 and 3.

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A. Table 3.3.1.1-1, Reactor Protection S stem Instrumentation Allowable Value Function Current Proposed

4. Reactor Vessel 2 538 inches 2 528 inches Water Level- above vessel above vessel Low, Level 3 zero zero B. Table 3. 3. 5. 1-1 Emer en Core Coolin S stem Instrumentation Allowable Value Function Current Proposed

4. ADS Trip System A

d. Reactor Vessel 2 544 inches 2 528 inches Water Level above vessel above vessel Low, Level 3 zero zero (Confirmatory)

5. ADS Trip System B

d. Reactor Vessel 2 544 inches 2 528 inches Water Level above vessel above vessel Low, Level 3 zero zero (Confirmatory)

C. Table 3.3.6.1-1 Prima Containment Isolation Instrumentation Allowable Value Function Current Proposed

2. Primary Containment Isolation

a. Reactor Vessel c 538 inches c 528 inches Water Level above vessel above vessel Low, Level 3 zero zero El-2

Allowable Value Function Current Proposed

5. Reactor Water Cleanup (RWCU)

System Isolation

h. Reactor Vessel > 528 inches Water Level- 2 538 inches above vessel Low, Level 3 above vessel zero zero

6. Shutdown Cooling System Isolation

b. Reactor Vessel 2 538 inches 2 528 inches Water Level- above vessel above vessel Low, Level 3 zero zero D. Table 3.3.6.2-1 Seconda Containment Isolation Instrumentation Allowable Value Function Current Proposed

1. Reactor Vessel 2 538 inches 2 528 inches Water Level- above vessel above vessel Low, Level 3 zero zero E. Units 2 and 3 Table 3.3.7.1-1, Control Room Emer en Ventilation S stem Instrumentation Allowable Value Function Current Proposed

1. Reactor Vessel 2 538 inches 2 528 inches Water Level Low, above vessel above vessel Level 3 zero zero REASON FOR THE PROPOSED CHANGE During reactor operation, there is approximately 23 inches between the normal reactor water level and the reactor scram initiation point. Plant systems are designed such that the reactor can usually automatically recover from many transients such as a trip of a feedwater system pump.

However, in some cases, with this tight water level range, reactor scrams may result that would have been avoidable if

plant control systems or operators had slightly more time to take control. In addition, since Boiling Water Reactors operate with a high steam void fraction, water level is sensitive to mild pressure perturbations. Often, the prompt water level drop due to rapid void collapse caused by a manual or automatic scram is large enough to cause a Level 3 trip. This initiates a primary and secondary containment isolation, and a SBGT and CREV system initiation. These system trips are an unneeded distraction factors for the operators in responding to scrams.

This proposed TS change increases the opex'ating range between the normal reactor vessel water level (561 inches above vessel zero) and the Reactor Vessel Level Low, Level 3 actuation Av by 10 inches (current value of 538 inches, proposed value of 528 inches). With the small increase in water level range, over the course of the reactor operating it life, is expected that several unnecessary scrams will be avoided. This also has a positive effect in that unnecessary challenges to other ESFs will likewise be'voided.

In addition to reducing the low reactor water level scram initiation point, several other instrument functions that occur at Level 3 are being lowered to maintain consistency with the low level scram trip setting as well as to provide a similar margin to unnecessary initiation of ESFs. This reduction in the Av can be achieved without increasing the consequences of events that rely on these instrument functions and without having an adverse effect on plant safety analyses.

The safety related systems and components that are initiated by a Reactor Vessel Water Level Low, Level 3 signal will still operate in the same manner as they currently do. There are no changes to component maintenance or testing associated with the proposed TS change.

ANALYTICALLIMIT/ALLOWABLEVALUE DETERMINATION The instrument function analytical limit (AL) is the value used in the safety analyses to demonstrate acceptable nuclear safety system performance is maintained. The Av and trip setpoints (SP) are then chosen/calculated such that the instrument will function before reaching the AL under the woxst case environmental/event conditions. Instrument SPs account for measurable instrument characteristics (e.g.,

drift, accuracy, repeatability).

I The Av/SP instrument calculations for this. proposed change were performed in accordance with the methodology in TVA procedure EEB-TI-28 (Reference 1). This methodology is consistent with NRC Regulatory Guide 1.105 and has been previously reviewed by the NRC (References 2 and 3). The same methodology was also used for TS Change TS-390 to exi end the instrument function surveillance frequencies for 24-month fuel cycle operation (Reference 4). The NRC approved TS-390 on November 30, 1998 (Reference 5).

The attached figure illustrates the relationship between the SP, the minimum and maximum acceptable Avs [Av(min) and Av(max)], and the AL for a process that decreases toward the setpoint. To provide operational reliability and to ensure that the instrument will perform its design basis function, the TS Av is established within'the "Av Band."

The current TS Av is based on an AL of 530 inches'bove vessel zero. In the safety evaluation for this proposed change, a conservatively low AL value of 512 inches above vessel zero was used. This 512 inches value is actually below the lower instrument tap located at 517 inches. Since the water level instruments cannot physically measure levels below the instrument tap, the proposed TS Avs and SP calculations are based on an assumed AL of 518 inches. This is a conservative approach and provides additional margin in the safety evaluation.

IV. SAFETY ANALYSIS A safety analysis was performed to support lowering the Reactor Vessel Water Level Low, Level 3 AL by 18 inches from the present 530 inches to 512 inches above vessel zero.

As discussed above, 512 inches is conservatively lower than the minimum measurable value for this instrumentation. For the RPS (SCRAM) actuation function, the following events were evaluated; abnormal operational occurrences, loss-of-coolant accident (LOCA), anticipated transient without scram (ATWS), Appendix R fire event, radiological release, and containment loading and heating. The effects of lowering the corresponding AL for the remaining Level 3 instrument functions were also evaluated. The results of the evaluations are summarized below.

A. Method of Anal sis The analysis for LOCA events were performed with the GE proprietary SAFER/GESTR-LOCA model which is the current licensing basis methodology used for BFN (Reference 6).

For ATWS events, abnormal operating occurrences, and

Appendix R fire events, radiological release, and containment loading and heating, and the other instrument functions, the engineering analysis reviewed previous analyses to determine any potential impact of a reduced Level 3 Av.

Pur ose of Anal sis The analysis was conducted to demonstrate that lowering of the Reactor Vessel Level Low, Level 3 AL by 18 inches from the present 530 inches to 512 inches did not affect the licensing safety limits and did not affect the ability of the plant to operate safely and mitigate the consequences of a design basis accident and transients.

Anal sis for Reduced Level 3 RPS Actuation A low water level in the reactor vessel indicates that reactor coolant is being lost through a breach in the nuclear system process barrier or th'at the supply of reactor feedwater is less than required to maintain normal level due to a system malfunction. Should the water level decrease too far, fuel damage could ultimately occur if the reactor core is uncovered. The purpose. of the reactor low scram is to reduce the rate of water inventory loss by shutting down the reactor.

Scramming the reactor drastically reduces the steaming rate and allows time for feedwater systems or emergency injection systems to operate to prevent core damage.

The setting of the water level scram signal is chosen far enough below normal operating level to avoid spurious scrams, but high enough above the top of active fuel to assure that adequate cooling will be available following the most severe abnormal operating transient including a level decrease.

The following evaluates the effects of the Reactor Vessel Water Level Low, Level 3 scram function for events in the safety analyses for the plant.

Abnormal 0 erational Occurrences The abnormal operational occurrences evaluated in the Updated Final Safety Analysis Report (UFSAR) for BFN were reviewed with respect to the proposed change. The scenario for each event was examined to determine if a RPS actuation was assumed to occur on low vessel water level. A reduced Level 3 Av has no effect on the El-6

events for which a reactor scram does not occur on low water level.

The only analyzed abnormal operational occurrence for which a Level 3 water level scram occurs is the total Loss-of-Feedwater (LOFW) event. In a LOFW event, the reactor water level decreases due to loss of feed flow resulting in a low water level scram at Level 3.

Reactor level continues to drop until it reaches Level 2 (470 inches above vessel zero), at which point the reactor core isolation cooling (RCIC) system and high pressure cooling injection system auto-initiate to restore the reactor water level.

The safety evaluation shows that the RCIC system alone continues to be able to maintain the reactor water level above Level 1 and refill the vessel (as is the case with the existing Av for the LOFW event). Level 1 is at 398 inches above vessel zero and is above the top of the core.

Therefore, no unacceptable safety consequences will result for abnormal operational occurrences for the reduced Level 3 Av and there is no significant impact on the plant response to abnormal operational occurrences'oss-of-Coolant Accident Current pipe break analyses (Reference 6) indicate that the limiting LOCA event is a design basis accident (DBA) recirculation suction line break with a battery failure. The DBA LOCA bounds the limiting small break LOCA which is a 0.08 ft'eactor recirculation system discharge line break with a battery failure.

For the DBA LOCA, the initial reactor water level is assumed to be the normal reactor water level and the reactor scrams on high drywell pressure at the same time the break occurs. Therefore, there is no impact on the DBA LOCA analysis associated with the reduced Level 3 RPS actuation Av.

For the limiting (0.08 ft ) small break LOCA, initial water level is assumed to be at the scram water level AL and the reactor has already scrammed due to high drywell pressure at the time the break occurs.

Therefore, reducing the Level 3 AL only lowers the assumed initial water level for the small break analysis (530 inches versus 512 inches). With this

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reduced AL, the calculated peak clad temperature (PCT) for the small break is 1346'F, which is less than the current value of 1367'F. This decrease in PCT is directly related to the earlier initiation of ADS on the Reactor Vessel Water Level Low Low Low, Level 1 signal due to the lower assumed initial water level.

Therefore, lowering the Level 3 RPS Av will not have an adverse affect on reactor performance for postulated LOCA events and no changes in the plant licensing limits are required.

Antici ated Transient Without Scram The four limiting ATWS events for BFN are:

1) Closure of all Main Steamline Isolation Valves,

2) Pressure Regulator Failure to Maximum Steam Demand Flow,

3) Loss of Normal Feedwater, and

4) Inadvertent Opening of a Relief Valve.

These events assume the failure of the reactor scram and instead utilize the alternate rod insertion, recirculation pump trip, and standby liquid control system equipment to reduce core thermal power.

Therefore, reducing the Level 3 RPS Av does not affect the ATWS evaluations.

A endix R Fire Event Anal sis The Appendix R fire event analysis for BFN assumes that the reactor is manually scrammed with reactor water level assumed to be at normal operating level.

Therefore, reducing the Level 3 RPS actuation Av does not affect the Appendix R analysis.

Radiolo ical Release The limiting pipe break for radiological releases inside the containment is the DBA LOCA. The DBA LOCA assumes that the reactor scram occurs at time zero due to high drywell pressure with a normal reactor water level. Therefore, reducing the Level 3 RPS Av has no impact on the radiological release analyses inside the containment for the DBA LOCA a'nalyses.

The limiting pipe break for radiological releases outside containment is the design basis main steam line

(MSL) break outside the containment. The MSL break outside the containment assumes a normal initial reactor vessel water level and that the reactor scrams when the main steam isolation valves close on high main steam line flow. Therefore, the reducing the Level 3 RPS Av has no effect on the calculated radiological releases for the MSL break outside containment event.

Containment Loads And Heatin Containment dynamic loads and main safety relief valve loads associated with the DBA LOCA were also reviewed.

These analysis assume the reactor scrams on high drywell pressure. Therefore, the DBA LOCA short-term and long-term containment loads, and drywell/wetwell temperature response for the DBA LOCA are not affected by a reduced Level 3 RPS Av.

Review of Other Level 3 Functions As listed previously, several other system functions are initiated by a Level 3 water level trip signal.

The Avs for these functions are also proposed to be changed to the Level 3 RPS actuation Av to maintain consistency with current TS. Impacts on these functions are addressed below.

Primar Containment Isolation S stems (PCIS) (Includin Shutdown Coolin S stem and Reactor Water Cleanu S stem Iso'lation)

A low RPV water level indicates that the capability to cool the fuel may be threatened if level continues to drop. Therefore, valves whose penetrations communicate with the primary containment or the reactor coolant system automatically isolate't Level 3 to limit the potential for loss of reactor coolant and to limit the potential release of fission products. The isolation of primary containment valves at Level 3 supports actions to ensure that onsite and offsite dose limits of 10 CFR 20 and 100 are not exceeded. The Reactor Vessel Water Level Low, Level 3 isolation function is assumed in the Chapter 14 FSAR pipe break analyses since these leakage paths are considered isolated post-LOCA.

The Level,3 low water level setting for primary containment isolation was selected to initiate isolation at the earliest indication of. a possible breach in the nuclear system process barrier, yet far

enough below normal operational levels to avoid spurious isolation. Historically, the containment isolation low level trip and the RPS actuation trip setpoints are set at the same value.

Isolation of the following is initiated on Reactor Vessel Low Water, Level 3.

~ Residual Heat Removal (RHR) reactor shutdown cooling supply

~ Reactor Water Cleanup (RWCU)

~ Drywell equipment drain discharge

~ Drywell floor drain discharge

~ Drywell purge inlet

~ Drywell main exhaust

~ Suppression chamber exhaust valve bypass

~ Suppression chamber purge inlet

~ Suppression chamber main exhaust,

~ Drywell exhaust valve bypass

~ Suppression chamber drain

~ RHR- Low Pressure Coolant injection (LPCI) to reactor (in shutdown mode)

~ Drywell make-up

~ Suppression chamber make-up

~ Exhaust to SGT

~ Drywell radiation monitor

~ Drywell control air compressor

~ Containment atmosphere monitor

~ Drywell differential pressure air compressor

~ Traversing incore probes During postulated accidents, significant radiation releases cannot occur until after the core is uncovered. Since the reduced Level 3 actuation is still approximately 12 feet above the top of the core, still the Level 3 PCIS actuation will occur well before core uncovery. Therefore, a small delay of this isolation signal due to the reduction in Av will not affect the ability of the containment isolation valves to perform their intended functions. For LOCA events inside containment, a high drywell pressure signal will also initiate a primary containment isolation for all the above systems (except Reactor Water Cleanup) very early in the event (prior to a Level 3 water level trip) .

The shutdown cooling mode of the RHR system is also isolated by the Level 3 water level trip for a malfunction of the RHR which results in a reactor coolant inventory loss. Shutdown cooling is in service only when the reactor is shutdown. Isolation of system will also cause any operating RHR pumps to trip on loss of suction path. These automatic actions prevent further coolant loss through the RHR shutdown cooling loop if the water level decrease is being caused by the an RHR system malfunction. The reduction of the Level 3 Av will not affect the intended function of these isolation valves since the system still isolates at a water level far above the top of core. Also, the emergency mode of the RHR system (low pressure coolant injection) is not required to function until vessel level has dropped to Level 1. Therefore, reducing the Level 3 Av has no impact on the ability of the shutdown cooling mode isolation to perform its intended functions.

The RWCU system also isolates on Level 3 water level trip in the event that reactor coolant is being lost though a RWCU system line break. The Level 3 RWCU isolation is not directly analyzed in the UFSAR because the RWCU system line break is bounded by breaks of systems (DBA LOCA and main steam line break 'arger outside the containment). Therefore, reducing the Level 3 actuation has no impact on the ability of the RWCU isolation valves to perform their intended functions. Additionally, from an operations perspective, in order to maintain reactor water quality it is beneficial not to isolate the RWCU system unnecessarily.

The remaining systems which are isolated by the primary containment isolation signal are not required immediately following a loss of water inventory event since they do not directly contribute to the.

replenishment of the vessel water inventory.

Therefore, lowering the water Level' Av for automatic isolation will not impact the ability to replenish inventory. As previously discussed, the release of fission products will not occur until after the core is uncovered. Since the Level 3 actuation will always occur well before core uncovery, the delay of this isolation signal will not affect the ability of the containment isolation valves to perform their intended functions. Also, as noted previously, these valves, except for RWCU, also automatically isolate on high drywell pressure for LOCA events prior to the water

level trip. In summary, the primary containment isolation function is not adversely affected by reducing the Level 3 actuation.

Secondar Containment Isolation and Standb Gas Treatment (SGT) S stem Initiation The isolation of the secondary containment and initiation of the SGT system support actions to ensure that any radiological releases to secondary containment do not result in exceeding offsite release limits. The LOCA provides the most severe radiological release and, thus, serves as the bounding design basis accident in determining the post-accident offsite dose. For LOCA events, secondary containment and SGT will actuate on high drywell pressure prior to reaching the Level 3 water level trip; therefore, a reduced Level 3 Av 'has no effects on the LOCA event analysis. For other loss of inventory events, as described previously, the Level 3 actuations will always occur well before any core uncovery which could result in potential radiological release. Therefore, the small delay introduced by a change in the Level 3 Av will not affect the ability of the secondary containment or SGT to perform their intended function.

Control Room Emer enc Ventilation (CREV) S stem Initiation The CREV system is designed to provide a radiologically controlled environment to ensure the habitability of the control room for all plant conditions. In the.

event of a Level 3 signal, the CREV system is automatically initiated to pressurize the control room to minimize the consequences of radiological releases to the control room environment. The LOCA provides the most severe radiological release to the primary and secondary containment and, thus, serves as the bounding DBA in determining the control room dose. For LOCA events, the CREV system will actuate on high drywell pressure prior to reaching the Level 3 water level trip. Therefore, a reduced Level 3 Av has no effects on the LOCA event analysis. For other loss of inventory events, as described previously, the Level 3 actuations will occur well before any core uncovery, which could result in potential radiological release. Therefore, the small delay introduced by a change in the Level 3 actuation will not affect the ability of the CREV system to perform its intended function.

I Automatic De ressurization S stem (ADS) Reactor Vessel Water Level Confirmator Si nal The proposed TS change lowers ADS confirmatory signal Level 3 Av from 544 inches to 528 inches to maintain consistency with the other Level 3 trip functions.

This Level 3 signal is a confirmatory low water level signal for ADS initiation, which serves to prevent unnecessary ADS initiation resulting from spurious Level 1 (398 inches) water level actuations or as a result of a break in the Level 1 instrument line. The intended function of this confirmatory signal will still beis successfully accomplished even if the Level 3 signal reduced since the Level 3 signal will occur well prior to Level 1. Therefore, reducing the Level 3 Av will not affect the ability of ADS to perform its intended function.

0 erational Concerns on Reduced Level The proposed Level 3 Av is slightly below the level of the steam dryer seal skirt. Long-term reactor operation with water level below the dryer seal could affect the steam separator-dryer performance since additional moisture might be carried over into turbine side equipment. However, plant operators. continuously monitor reactor water level and take actions promptly to ensure normal level is maintained. Also, there is a water level alarm at 555 inches (about 6 inches below normal level) which would prompt operators to restore normal level if automatic controllers were not operating properly. Therefore, the potential to operate with water level below the steam dryer seal skirt is not considered a practical concern. This condition would also not be a safety concern since the main and reactor feed pump turbines are not required for safe shutdown of the plant.

Effect Of Lowerin The Level 3 Av On The Probabilistic Safet Anal sis (PSA)

There are two minor effects on the BFN PSA, which will be addressed qualitatively. The first and more substantial effect is the reduction (i.e., improvement) in the initiating event frequencies due to the lowering of the Level 3 setpoint. This results from the reduction in number of inadvertent scrams from minor operational transients that are avoided by the lower level Av. The improvement in initiating event frequency will result in a slight improvement in the

core damage frequency and large early release frequency.

The other potential effect on the PSA is a small affect on the timing of mitigative operator actions after the scram and isolation functions of the Level 3 set point are completed. The reduction of the Level 3 Av by 10 inches will result in a small reduction in time between the scram and isolation function, and other follow on actions. This effect is considered insignificant and overshadowed by the risk reduction due to the initiating event frequency changes discussed above.

G. Conclusion Safety analysis to support lowering the Reactor Water Level 3 Av were performed for BFN Units 2 and 3. Based on the analysis, it is concluded that lowering the Level 3 Av to 528 inches above vessel zero is acceptable and has no significant impact on abnormal operational occurrences, LOCA, ATWS, Appendix R fire events, radiological releases, or containment loads and heating. Furthermore, lowering Level 3 will provide additional operating range to the Level 3 RPS actuation during plant operational transients which reduces the probability of undesired reactor scrams and other ESF actuations on low reactor water level. Therefore, it concluded that the proposed change has a beneficial effect on plant operations and safety.

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has concluded that operation of BFN Units 2 and 3 in accordance with the proposed change to the TS does not involve a significant hazards co~sideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR

50. 92 (c) .

A. The ro osed amendment does not involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.

The Reactor Vessel Water Level Low, Level 3 functions are in response to water level transients and are not involved in the initiation of accidents or transients.

Therefore, reducing the Level 3 Av does not increase the probability of an accident previously evaluated.

Additionally, the results of the safety evaluation

associated with the lowering of the Level 3 Av concludes that the previously evaluated transient and accident consequences are not significantly affected by the change. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

B. The ro osed amendment does not create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

The proposed amendment to lower the BFN Units 2 and 3 Reactor Vessel Water Level Low, Level 3 Av does not involve a hardware change and the purpose of the Level 3 function is not affected. The Level 3 functions will continue to fulfill their design objective. Therefore, reduction of the Av does not result in the possibility of a new or different kind of accident.

C. The ro osed amendment does not involve a si nificant reduction in a mar in of safet The results of the safety evaluation associated with the reducing the BFN Units 2 and 3 Reactor Vessel Water Level Low', Level 3 Av concluded that transient and accident consequences remain within the required acceptance criteria. Therefore, the margin of safety is not reduced for any event evaluated.

Therefore, the proposed Level 3 Av does not adversely affect the public health and safety, and does not involve any significant safety hazards.

ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

an environmental assessment of the proposed change is not required.

VII. REFERENCES EEB-TI-28, "Setpoint Calculations," Branch Technical Instruction, Revision 2, Tennessee Valley Authority, October 6, 1992.

2. NRC Regulatory Guide 1.105, "Instrument Setpoints for Safety-Related Systems," Revision 2, February 1986

3. NRC letter to TVA dated May 8, 1989, "Notice of Violation (NRC Inspection Report Nos. 50-259/89-06, 50-260/89-06, and 50-296/89-06)"

4. TVA letter to NRC dated August 14, 1998, "Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 TS Change TS-390 Supplement 1 Request for License Amendment to Support 24-Month Fuel Cycles"

5. NRC letter to TVA dated November 30, 1998, "Issuance of Amendments Browns Ferry Nuclear Plants Units 1, 2, and 3 (TAC Nos. MA2081, MA2082, and MA2083)"

TVA letter to NRC dated July 24, 1998, "Browns Ferry Nuclear Plant (BFN) Response to Request For Additional Information (RAI) relating to Units 2 and 3 Technical Specification (TS) Change No. TS-384 Power Uprate Operation" (Enclosure 5)

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Figure: Instrument Value Relationships AL (upper)

Region of untneasurable uncertainties Av (max)

Av Band Av (min)

Region of normal Setpoint (SP) measurable uncertainties Av (min)

Av Band Av (max)

Region of unmeasurable uncertainties AL (lower)

E1-17

0 ~ ~