ML040710859

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(BFN) Unit 1 - Technical Specification 434 - Lowering Allowance Value for Reactor Vessel Water Level - Low Level 3
ML040710859
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 03/09/2004
From: Abney T
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-BFN-TS-434
Download: ML040710859 (29)
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Text

IV Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 March 1, 2004 TVA-BFN-TS-434 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:

In the Matter of ) Docket No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 - TECHNICAL SPECIFICATION 434 - LOWERING THE ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL - LOW LEVEL 3 Pursuant to 10 CFR 50.90, Tennessee Valley Authority (TVA) is submitting a request for an amendment to license DPR-33 for BFN Unit 1. The proposed amendment will reduce the Allowable Value used for Reactor Vessel Water Level - Low, Level 3 for several instrument functions.

The primary purpose of this proposed Technical Specification 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.

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 Printed on recyced paw

U.S. Nuclear Regulatory Commission Page 2 March '9, 2004

instrument Allowable Value at which the scram will occur. The

-safety analysis in Enclosure 1 shows that the Allowable Value may be lowered without adversely affecting the plant response to postulated transients and accidents. As discussed in Section 3.3 of Enclosure 1, the proposed changes to the Unit 1 Technical Specifications are the same changes as that approved for Units 2 and 3 in Reference 1.

The proposed amendment is necessary to support the restart of Unit 1 and improves the fidelity with Units 2 and 3. In addition, TVA relies on the approval of this proposed change as part of the design assumptions used to justify an extended power uprate for Unit 1 prior to restart. Therefore, TVA requests that the amendment be approved by March 11, 2005.

TVA has determined that there are no significant hazards considerations associated with the proposed amendment and that the amendment qualifies for a categorical exclusion 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 letter and attachments to the Alabama State Department of Public Health.

Enclosure 1 provides TVA's evaluation of the proposed amendment.

Enclosure 2 provide mark-ups of the proposed change to the

-Technical Specifications. Enclosure 3 provide draft Technical Specification pages that have been updated to reflect the proposed change.

There are no regulatory commitments associated with this submittal. If you have any questions about this amendment, please contact me at (256)729-2636.

-I declare under penalty of perjury that the foregoing is true and correct. Executed on March 9, 2004.

U.S. Nuclear Regulatory Commission Page 3 March 9;, 2004

Enclosures:

1. TVA Evaluation of Proposed Amendment
2. Proposed changes to the Technical Specifications (mark-ups)

References:

Enclosure cc (Enclosures):

State Health Officer Alabama State Department of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017

  • 7 a ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATION CHANGE (TS-434) -

LOWERING THE ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL - LOW LEVEL 3 TVA EVALUATION OF PROPOSED AMENDMENT INDEX SECTION DESCRIPTION PAGE 1.0 Description E1-2 2.0 Proposed Amendment E1-2 3.0 Background E1-4 4.0 Technical Analysis E1-7 5-.0 Regulatory Safety Analysis E1-19 6.0 Environmental Considerations E1-21 7.0 References E1-21 El-1

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1.0 DESCRIPTION

This letter requests an amendment to license DPR-33 for BFN Unit 1. The proposed amendment will reduce the Allowable Value used for Reactor Vessel Water Level - Low, Level 3 for several instrument functions.

The primary purpose of this proposed Technical Specification 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.

The proposed amendment is necessary to support the restart of Unit 1 and improves the fidelity with Units 2 and 3. In addition, TVA relies on the approval of this proposed change as part of the design assumptions used to justify an extended power uprate for Unit 1 prior to restart. Therefore, TVA requests that the amendment be approved by March 11, 2005.

2.0 PROPOSED AMENDMENT The proposed change will lower the current Reactor Vessel Water Level - Low, Level 3 Allowable Value in the Unit 1 Technical Specification for several instrument functions. The following specific Technical Specification functions are affected by this proposed change:

  • Control Room Emergency Ventilation (CREV) System Initiation The proposed changes to the Technical Specifications are listed below. Enclosure 2 contains copies of the appropriate marked-up Technical Specification pages for Unit 1 showing the changes. No changes to the Technical Specification Bases are required.

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i. Table 3.3.1.1-1, Reactor Protection System 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
2. Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation Allowable Value Function Current Proposed
4. ADS Trip System A
d. Reactor Vessel Water Level - 2 544 inches 2 528 inches Low, Level 3 above vessel above vessel (Confirmatory) zero zero
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)

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.3. Table 3.3.6.1-1, Primary Containment Isolation Instrumentation Allowable Value Function Current Proposed

2. Primary Containment Isolation
a. Reactor Vessel 2 538 inches 2 528 inches Water Level - above vessel above vessel Low, Level 3 zero zero
5. Reactor Water Cleanup (RWCU)

System Isolation

h. Reactor Vessel 2 538 inches 2 528 inches Water Level - above vessel above vessel Low, Level 3 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
4. Table 3.3.6.2-1, Secondary 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
5. Table 3.3.7.1-1, Control Room Emergency Ventilation System 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 E1-4

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3.0 BACKGROUND

Provided in this section is the reason for this proposed change and a description of the modifications required to implement the proposed change. Also included at the end of this section is a comparison of the proposed change, reason for change and technical analysis submitted in support of this proposed amendment with the information provided by TVA and approved by NRC for the Units 2 and 3 license amendments (References 1 and 2).

3.1 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 primary and secondary containment isolation, and SGT and CREV system initiations. These system isolations and initiations are an unneeded distraction for the operators responding to scrams.

This proposed Technical Specification change increases the operating range between the normal reactor vessel water level (561 inches above vessel zero) and the Reactor Vessel Level - Low, Level 3 actuation Allowable Value by 10 inches (current value of 538 inches, proposed value of 528 inches).

The increased range will provide additional time for operators or plant systems to automatically respond to recoverable transients such as feedwater system malfunctions. With the small increase in water level range, over the course of tie reactor operating life, it is expected that several unnecessary scrams will be avoided.

This also has a positive effect in that unnecessary challenges to other Engineered Safety Features (ESFs) will likewise be avoided.

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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 Allowable Value 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.

3.2 Description of the Proposed Modifications The proposed change will lower the current Reactor Vessel Water Level - Low, Level 3 Allowable Value in the Unit 1 Technical Specification for several instrument functions.

The setpoints for the affected instruments will be adjusted and associated procedures revised. 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 Technical Specification change.

3.3 Comparison with previous Technical Specification changes for Unit 2 and 3 TVA has compared the proposed change, reason for change, background information, and technical analysis submitted in support of this proposed amendment with the information provided by TVA and approved by NRC in TS 397 (References 1 and 2) for lowering the Units 2 and 3 Allowable Value for the Reactor Vessel Water Level - Low Level 3 signal. The comparison for each of these areas is provided below:

  • The proposed changes to the Unit 1 Technical Specifications are the same changes as that proposed and approved for Units 2 and 3.
  • The reason for the Unit 1 Technical Specification change is the same as that which was previously submitted for the Units 2 and 3 Technical Specification change (i.e., 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).

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  • The background information provided in support of the Unit 1 Technical Specification change incorporates the same elements previously submitted in support of the Units 2 and 3 Technical Specification change.
  • The technical analysis submitted for this Unit 1 Technical Specification change incorporates the majority of the same elements which were previously submitted for the Units 2 and 3 Technical Specification change. The Units 2 and 3 submittal contained a qualitative evaluation of the effect of lowering the Level 3 Allowable Value on the Probabilistic Safety Analysis (PSA). A Unit 1 PSA is not currently available. Therefore, the PSA evaluation was based on design similarities between the units.

4.0 TECHNICAL ANALYSIS

4.1 Analytical Limit  ! Allowable Value Determination The instrument function Analytical Limit is the value used in the safety analyses to demonstrate acceptable nuclear safety system performance is maintained. The Allowable Value and trip setpoints are then chosen / calculated such that the instrument will function before reaching the Analytical Limit under the worst case environmental / event conditions. Instrument setpoints account for measurable instrument characteristics (e.g., drift, accuracy, repeatability).

The Allowable Value / Setpoint instrument calculations for this proposed change were performed in accordance with the methodology in TVA procedure EEB-TI-28 (Reference 3). This methodology is consistent with NRC Regulatory Guide 1.105 (Reference 4) and has been previously reviewed by the NRC (Reference 5). The same methodology was also used for Technical Specification Change TS-390 to extend the instrument function surveillance frequencies for 24-month fuel cycle operation (Reference 6). The NRC approved TS-390 on November 30, 1998 (Reference 7).

The attached figure illustrates the relationship between the setpoint, the minimum and maximum acceptable Allowable Values [Allowable Value (min) and Allowable Value (max)],

and the Analytical Limit 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 Technical Specification Allowable Value is established within the "Allowable Value Band."

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The current Technical Specification Allowable Value is based on an Analytical Limit of 530 inches above vessel zero. In the safety evaluation for this proposed change, a conservatively low Analytical Limit 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 Technical Specification Allowable Values and setpoint calculations are based on an assumed Analytical Limit of 518 inches. This is a conservative approach and provides additional margin in the safety evaluation.

4.2. Safety Analysis A safety analysis was performed to support lowering the Reactor Vessel Water Level - Low, Level 3 Analytical Limit 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. As also discussed in Section 2.0, several specific Technical Specification functions are affected by this proposed change. These functions and references to the associated Updated Final Safety Analysis Report (UFSAR) descriptions of these design functions is provided below:

For the RPS actuation function (SCRAM), 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 Analytical Limit for the remaining Level 3 instrument functions were also evaluated. The results of the evaluations are summarized below.

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1. Method of Analysis The analysis for LOCA events were performed with the SAFER/GESTR-LOCA model, which is the current licensing basis methodology used for BFN (Reference 8). 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 Allowable Value.
2. Purpose of Analysis The analysis was conducted to demonstrate that lowering of the Reactor Vessel Level Low, Level 3 Analytical Limit 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 abnormal operational occurrences.
3. Analysis for Reduced Level 3 RPS and ECCS Actuations 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 that 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.

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0 Abnormal Operational Occurrences The abnormal operational occurrences evaluated in the 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 Allowable Value has no effect on the 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 (HPCI) 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 Allowable Value 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 Allowable Value and there is no significant impact on the plant response to abnormal operational occurrences.

o Loss-of-Coolant Accident Current pipe break analyses (Reference 9) 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 ft2 reactor recirculation system discharge line break with a battery failure.

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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 Allowable Value.

For the limiting (0.08 ft2) small break LOCA, initial water level is assumed to be at the scram water level Analytical Limit and the reactor has already scrammed due to high drywell pressure at the time the break occurs. Therefore, reducing the Level 3 Analytical Limit only lowers the assumed initial water level for the small break analysis (530 inches versus 512 inches). With this reduced Analytical Limit, the calculated peak clad temperature (PCT) for the small break is also reduced. This reduction in the 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.

The proposed Technical Specification change also lowers ADS confirmatory signal Level 3 Allowable Value 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 be successfully accomplished even if the Level 3 signal is reduced since the Level 3 signal will occur well prior to Level 1. Therefore, reducing the Level 3 Allowable Value will not affect the ability of ADS to perform its intended function.

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

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o Anticipated Transient Without Scram The four limiting ATWS events for BFN are:

1) Closure of all Main Steam Line 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 utilizes the alternate rod insertion, recirculation pump trip, and the standby liquid control system equipment to reduce core thermal power. Therefore, reducing the Level 3 RPS Allowable Value does not affect the ATWS evaluations.

o Appendix R Fire Event Analysis 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 Allowable Value does not affect the Appendix R analysis.

o Radiological 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 Allowable Value has no impact on the radiological release analyses inside the containment for the DBA LOCA analyses.

The limiting pipe break for radiological releases outside containment is the design basis main steam line break outside the containment. The main steam line 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, reducing the Level 3 RPS Allowable Value has no effect on the calculated radiological El-12

releases for the main steam line break outside containment event.

o Containment Loads and Heating Containment dynamic loads and main safety relief valve loads associated with the DBA LOCA were also reviewed. These analyses 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 Allowable Value.

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

The Allowable Values for these functions are also proposed to be changed to the Level 3 RPS actuation Allowable Value to maintain consistency with current Technical Specification. Impacts on these functions are addressed below.

o Primary Containment Isolation Systems (PCIS)

(Including Shutdown Cooling System and RWCU System Isolation)

A low reactor vessel 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 at 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 UFSAR pipe break analyses since these leakage paths are considered isolated post-LOCA.

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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

- 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 Standby Gas Treatment

- Drywell radiation monitor

- Drywell control air compressor

- Containment atmosphere monitor

- Drywell differential pressure air compressor

- Traversing incore probes EI-14

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, the Level 3 PCIS actuation will still occur well before core uncovery. Therefore, a small delay of this isolation signal due to the reduction in Allowable Value 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 primary containment isolation for all the above systems (except RWCU) 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 Allowable Value 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 (LPCI) is not required to function until vessel level has dropped to Level 1.

Therefore, reducing the Level 3 Allowable Value 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 larger systems (DBA LOCA and main steam line break 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 El-15

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 3 Allowable Value 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.

o Secondary Containment Isolation 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 Allowable Value 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 Allowable Value will not affect the ability of the secondary containment or SGT to perform their intended function.

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o Control Room Emergency Ventilation System 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 Allowable Value 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.

5. Operational Concerns on Reduced Level The proposed Level 3 Allowable Value 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.

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6. Effect of Lowering the Level 3 Allowable Value on Probabilistic Risk There are two minor effects on probabilistic risk, 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 Allowable Value. 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 effect on the timing of operator actions after the scram and isolation functions of the Level 3 set point are completed. The reduction of the Level 3 Allowable Value 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.

7. Conclusion Safety analysis to support lowering the Reactor Water Level 3 Allowable Value were performed for BFN Unit 1.

Based on the analysis, it is concluded that lowering the Level 3 Allowable Value 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.

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a 5.0 REGULATORY SAFETY ANALYSIS The Tennessee Valley Authority (TVA) is submitting an amendment request to license DPR-33 for the Browns Ferry Nuclear Plant (BFN) Unit 1.

5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment", as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No 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 BFN Unit 1 Level 3 Allowable Value 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 Allowable Value 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.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment to lower the BFN Unit 1 Reactor Vessel Water Level - Low, Level 3 Allowable Value 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. The proposed changes do not create the possibility of any new failure mechanisms. No new external threats or release pathways are created. Therefore, reduction of the Allowable Value does not result in the possibility of a new or different kind of accident.

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3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The results of the safety evaluation associated with the reducing the BFN Unit 1 Reactor Vessel Water Level - Low, Level 3 Allowable Value concluded that transient and accident consequences remain within the required acceptance criteria. Therefore, the margin of safety is not reduced for any event evaluated.

Based on the above, TVA concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The safety analysis provided above evaluated the reduction in the BFN Unit 1 Reactor Vessel Water Level - Low, Level 3 Allowable Value in the mitigation of (a) abnormal operational occurrences, (b) loss of coolant accidents, (c) anticipated operational occurrences, (d) anticipated transients without scram, (e) Appendix R events (fires) and (f) other events involving a potential radiological release. Compliance with the following requirements is not changed:

  • 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram [ATWS] events for light-water-cooled nuclear power plants);

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In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 50.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. TVA letter, T.E. Abney to NRC, dated June 3, 1999, "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."
2. NRC letter, Long to Scalice, dated August 16, 1999, "Browns Ferry Nuclear Amendments Regarding Allowable Value for Reactor Vessel Water Level (TAC Nos. MA5697 AND MA5698)."
3. EEB-TI-28, "Setpoint Calculations," Branch Technical Instruction, Revision 2, Tennessee Valley Authority, October 6, 1992.
4. NRC Regulatory Guide 1.105, "Instrument Setpoints for Safety-Related Systems," Revision 2, February 1986.
5. 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)."

El-21

6. 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."
7. 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) ."
8. General Electric SAFER/GESTR-LOCA, Loss of Coolant Analysis, Browns Ferry Units 1, 2, and 3, NEDC-32484P, Rev. 5, January 2002.
9. General Electric, "Browns Ferry Nuclear Plant Units 1, 2 and 3 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis,"

NEDC-32484P. Revision 2, December 1997.

EI-22

Figure: Instrument Value Relationships AL (upper) -~ l - - -I Region of unmeasurab e uncertainties Av (max)

I~

Av Band Av (min)

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

Ii Av Band Av (max) I-Region of unmeasurab: e uncertainties AL I (lower)

EI-23

. a ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATION CHANGE (TS-434) -

LOWERING THE ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL - LOW LEVEL 3 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

AFFECTED PAGE LIST 3.3-7 3.3-46 3.3-58 3.3-60 3.3-64 3.3-69

. MARKED PAGES e attached.

RPS Instrunentaton 3.3.1.1 Tabb11l.141 1Pag2d.3)

Reactor Protction Syrermum aa APPLICABLE CONITKMNS MOES OR REQUIRED REFERENCED FUWCTION OTFER CmkNNELS FROM SURVEILLANCE ALLOWABLE SPEIFIED PER TRIP REQUIRED REQLJaEMTS VALUE CONDITIONS SYSTEM ACTION D.1 I AveragePworvReage Mmbm -mmC

d. Oaomscele 2 F BR 3 31.1.7 r RTP SR 11 .t.8 SR 11t.1.14
a. Ihop 1.2 2 o SR 3t1.1.7 ' NA.

SR 1311.1.8 SR 311.1.14 3 ReactorVessel Stearn Dome 1.2 2 O SR 311.1.1 i 1055 pug PboSVut. o aR 13.1.1.8 SR &11.1.10 SR W1.1.14

4. R aac~torVeWsWlW utinL". 1.2 2 G SR 3.&1.1. k 2 kw2r , ..... .. hi. d18 Law, Lale3 SR l i i , m ow wusu SR 131.1.1a zwe OR 11..1A14
a. Awn 8ImIsolston V" . 1 I F SR llt.1.8 %t0%dekd Cloure SR &.S.1.13 SR 3.311.14 S. Drywal Pressue- Hgh 11.2 2 O SR 33.1.1.8 is2.Spg SR 2.1.1.13 SR &.1.1.14 WSte L.'e - Hoh
a. Resistance Temperatre 1.2 2 SR 1.1.8 S 0o sg SR 111.1.13 SR 13.1.114 5s() 2 a SR a13.1.8 s 50 ~lons SR 33.11.13 SR 3.3.1.1.14
b. Flied Sv~te$h 1.2 2 H SR 311.1.3 s Sogsm SR 11.1.1.13 OR 111.1.14

!jWG 2 SR .1.1. S50gams SR 131.1.13 SR 131.1.14 4t-(e) i ono d d a frma e e otieg e rmeIi seue.

BFN-UNIT I 3.3-7 Amendment No. 234

I i ECCS Instumentaton 3.3.5.1 T" 3.1&&1.1 jp~q 5d 6)

EretMMUM C0m 0ockQ Srstsmrkmfnt"mbo APPLICADLE cCONDTlONS MODES REQUIRED REF04ENCED FUNJCTION OR OTHER CHANNELS FROIA SLA;VLUANCE ALLOWABLE SPECIFIED PER REQUIRED REQUIREMENTS VALUE CONDITONS FUNCTION ACTION AI d RaeetasmdIWgwLAWi 1* F SR 3151.1 -

kjm  : 544 Lamw Lowmi) (Corl5mutIm 2 (443(d) SR 3.3.5.1.2 SR 3.3.5.1.5 SR 3.3.S.1.6

1. o SR 3.&5.1.2 x 175 pctard 2(d). 3(4) SR 3.3.1.1.3 SR 3.3.5.1.6 0mmSpra Pmn I. jcharge 1,
  • SR 3.15.1.2 2!9Opsig w 2(d) 3(d) SR 3.3&5.1.3 Preamrse- High SR 3.315.1.6 C1.
  • SR 3..5.1.5 240, 3(d) SR 3.3.5.1.0 Pressure Bxpis Tmiv IADS Trip Systern B
a. ReactrWesdWaer Lwal- 1, F SR 3.3.5.1.1 hcow a 38 2(d). 3(4 SR 3a5.1.2 abme vessd LOW w LOW, LuN.IsI SR 3.5.1.5 2er0 SR 3.15.1.6 It. DryemIlPressure - IKgh 1, F SR 3.3.5.12  % .5Fapig 2(4,.3(d) SR 3.35.1.5 SR 3.3.5.1.6
5. AAuru~c Depressurtzdan 1* C SR 3.3S.1.5 c 1I5secinds Systerm erWato Thaw 2(4.3(4) SR 3.3.5.1.6 4 Reactoryeodwwe"L~amI* 1, F SR 33.5.1.1 011s 4 sbvcetow LVA Level 3 (Crifator)t 2K43M0 SR 33..S.1.2 vowe SR 3.35.1.5 SR 33.5.1.6 E-uru.]

(4 Withmasctorsteam don,.pfSatm vire315Psi BFN-UNIT I 3.3-46 Anendment No. 234

I It 9 'i

, 16, Primary Containment isolation Instrumentation 3.3.6.1 Tak" 3.36.1I I d 3)

Plinwy Ccntainwiut Islaio Insurnmmnihm APPLICABLE CO#JDfllONS MODESOR RECKARED REFER0NCED FUN4CTION OTHER CHANNELS FROM aSURVWAFICE ALLOWASLE SPECFIED PER TRIP REWIRED REQUIREMEN~TS VALUE CONCITIOII SYSTEM ACTION CA1 1.23 2 D SR 3.36.1.1 aage Mhuct LEMI LOW LOW LaeK SR 3.26.1.2 SR 3..6.1. wMM OR 3.6. 1.$

b. Main Sleam Line Preasure. 1 2 E SR &3.6.1.2 It625 pslg SR 33.3.1.5 SR 3.3..16S NLo 1.2,3 2pw D SR 3.3.6.1.1 s 140% 1ad MSL SR 3.3.6.12 slwflaw SR 3.3.6.115 SR 3.6.14.
d. Main Sham Tunnel 1.2.3 a O SR 3.3.1.2 Tw~erstue - Hg SR 3.3.6..

SR 3.3.6.1.6

2. Prhaey Cantorvnm koklton a R9&awVessaIWgete 1.2.3 2
  • SR 3.3A.1.1 wo J

.....- .e:S .

L" Low. 14,13 SR 3.3.12 gbaveessel SR 3.2.6..

SR 2.3.6.6

b. ~d Pmssv - High 1.23 2 O SR 3.3.6.1.2 SR 3.3.6.15.

SR 3.&SA.1.

3.High Pressure Coolant lajacdan Q4P1lj Sy-tun lackin a HPCIStwn LineFlow 1.2.3 F SR &.3..1.2 H~gh SR 3.3.6.1.5 SR &.3.6.1.6 3l90 psig b HPCStarnS upplyLine 1123 F SR 3.3.6.12 Ptsrmse . LGaw SR 3.X.&. 5 SR 3.3.6.16

c. MMPOTwbi~e 1,,23 F MR 3..6.12 %20 pat B"lUst Diatuamn SR 3..6.1.5 Pressure - H SR 3.L6.1.6 BFN-UNIT I 3.3-58 Amendment No. 234

Primary Containment Isolation Instrumentalon 3.3.6.1 Tdbllzt-l (Ma 3031)

Pftry Conbinwant lokon Intvwmon APPUCABLE CONDITIONS MOOES OR REQUIRED REFERENOED FUNCTION OTHER CHANNS FROM SURVLLANCE ALLWABLE SPECIFIED PER TRIP REQUIRED REQUREMENIS VALUE CONDITIONS SYSTEM ACTION CG S RactorWaer 0lanup (RWCU) Spn Irdadon

. Udn SbwnVi Vail 1.2,3 2 f SR 3.1 .t2 S 201 F bj Tow uv - High SR 3.16LIA SR 3.11 bL Pip.Tftnch Ara 1.Z3 2 F SR S..16.12 s 135PF Taipawtmu High SR 3.3.6.4 SR &3.6.1.

c. Pum Ron AAmn 1.23 2 F SR 3.&1.2 s 152F Tuiiputum - High SR "3.6.1A SR SS.1.1.
d. Pwnp Roon BAa 1.2.3 2 F SR 3..1.2 Twrperutume - Mh SR 3.I6IA SR 3.161.6

TwrwftmbHjr.gh SR 116.11 1.2 H SR 3.6.1.6

h. Redr NAd Wuvsr Lai - LOw,Lai 3 1.2,3 2 F SR 3.3.6.1.1 SR 34.3.1.2 k*57 --.1-I. stud: 53 SR 3.2.6.1.5 aso SR 3.36.1.6

. Shtwwffl Coding Spern Isokumb

a. Reatof Stoan Damm 1.X3 1 F SR 3.36.1.2 s It5 psu Prmum r High SR 3.161.5 SR 3.161.6
b. RmrtorVsd Wat.r 3.,JS 2(b. I SR 3.3.1.1 kfp.M.. _1.J-.Dwt:

- Su aid- LON, Id 3 SR 3.16.1.2 ~bw WSmd SR 3.16.1.3 a" OR .116 1.6 ac Drt Pissur -I Hih 1.2. 2 F SR 3.16.12 4125 Oslo SR 3.16.1.5 SR 3.16.1.6 (a) OrisSW Syatam iNbah ignaid palds ogi hpt iala.beth RWCIJ%uNs.

(b) Cidy Sm cina(W hipsayahn raqubed hiMODES 4 nd 5wlimRMi~StulomaCadhigSysamtuin eltsty i isd 8FN4UNIT I 3.3-60 Amendment No. 234

Iv 0 Z Secondary Containment Isolation Insrurmentafon 3.3.6.2 Table3.&&21 t Peze I d t) 8JcOda Cntornment kdai InsturnuI2io APPLIBLE MODES CR REQURED FUNCTION OTHER CHANNELS SURVELLU.CE ALLOWABLE 6PEC91ED PER REQUIREMENTS VALUE CONDITIONS TRIP SYSTEM

1. RowtarV IWA arLeI.- 1.23. 2 SR 31S221 . 5..- 538 Low. LagI 3 (a) SR 3.16I12 sdzmr .... fuefd:51 SR 3.16.13 SR 3.16.14 2 DwIPnosse - High 1,23 2 SR 316.22 s 2.psig SR 3.162.3 SR 3.162.4 4 ReuactrZoraeaust 1.2.3. 1 SR 3.16.21 sz nRMV Radna- H (a)(bl SR 3.162.2 SR 3.16.13 SR 31.1O.4 2, Rvegtukoloor~dfust 1.2A I SR 3.162+/-1 slO u00 Rh Rdain-H10 (8)Cb) SR 231622 SR 13.623 SR 3.162.4 (a) Duwba epaui Vb a pde fbrdInhg t rdofkeas (b) Dwkg CORE ALTERATONS and duIS mownummnntr g mfad Esadim bl hseonary canvUmint BNFN-UNIT I 3.344 Amendment No. 234

II a CREV System Instrumentabion 3.3.7.1 T"l 31.17.1-11OMW~I d 1)

CWMoRaoom Emgeraecy V*ollboe Syclai 11tuwntioo I

APPLICABLE CON CMTSON MOCESOR REQUIRED REPE .RENCED FL4CTION OTHER CHANNELS PRO SURVEINE ALLOWABLE SPECIFIED PER TRIP RE( CURED REOUIAREMETS VALUE CONDTIONS SYSTEM AC! r1ONAI

1. RaVmuIWatarLa.I.

1.2Aa) 2 B BR 3.17.1.1 af3us2 w -j ku.Irb M5 Loft Leal 2 SR S.27.12 ~abwmvas SR 3.1.1.6 ism OR 3.a7.1.6 I DyWal Piasum - h 1.2,3 2 B SR 117.12 s 2.5 rs4 SR 3..7.15 SR 3.17.1.

I RsalorZon Exhaict 1.2.3 C SR 3.17.1.1 s too nE~iI Radidan- Hg (9).(b) OR 2.17.12 SR 3.7.115 SR 3.17.1.6

4. PRduaL lngPw E)Mut~a 1.23. C SR S.17.1.1 %10 aiRAr Rada!mo- mich (8).(b) SR 12 7.12 SR 1.27.15 SR 6.17.11 I Couitd Room Air Supl Dudt Il.23. 0 SR .17.1.1 !S270 cpm Radabmc.- High (9).(b) SR 3.27.12 -ow SR L3.7.11. baciwmd SR 3.3.7.1A (a) Ourkgopurd&ons*Ma~pdfntaifordraainf.fmakwmsst Ib! 0wkfg ORE ALTERoATIONS d Aulng mwvmant of knucjdfd aasstiUcsIi U~ soon~ly ip~niwat BFN-USNIT I 3.3-69 Amendment No. 234

ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATION CHANGE (TS-434) -

LOWERING THE ALLOWABLE VALUE FOR REACTOR VESSEL WATER LEVEL - LOW LEVEL 3 PROPOSED TECHNICAL SPECIFICATION CHANGES (REVISED PAGES)

I. AFFECTED PAGE LIST 3.3-7 3.3-46 3.3-58 3.3-60 3.3-64 3.3-69 II. UPDATED PAGES See attached.

I V a RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPUCABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDlIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued
d. Downscale 1 F SR 3.3.1.1.7 2 3% RTP SR 3.3.1.1.8 SR 3.3.1.1.14
e. Inop 1.2 G SR 33.1.1.7 NA SR 3.3.1.1.8 SR 3.3.1.1.14
3. Reactor Vessel Steam Dome 1.2 G SR 3.3.1.1.1 S 1055 psig Pressure - High SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14
4. Reactor Vessel Water Level - 1.2 G SR 3.3.1.1.1 2 528 inches Low, Level 3 SR SR 3.3.1.1.8 3.3.1.1.13 above vessel I zer SR 3.3.1.1.14
5. Main Steam Isolation Valve - 1 F SR 3.3.1.1.8 S 10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14
6. Drywell Pressure - High 1,2 G SR 3.3.1.1.8 S 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14
7. Scram Discharge Volume Water Level - High
a. Resistance Temperature 1.2 G SR 3.3.1.1.8 S 50 galons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) H SR 3.3.1.1.8 S 50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14
b. Float Switch 1,2 G SR 3.3.1.1.8 S 50 galons SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) H SR 3.3.1.1.8 s 50 gabons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn !ron a core cel containing one or more fuel assembies.

BFN-UNIT I 3.3-7 Amendment No. 234

e' ;

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 5 of 6)

Emergency Core Cooling System Instrumentation APPUCABLE CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER REQUiRED REQUIREMENTS VALUE CONDITIONS FUNCTION ACTION A.1

4. ADS Trip System A (continued)
d. Reactor Vessel Water Level 1 1 F SR 3.3.5.1.1 2 528 inches l Low, Level 3 2 (d), 3 (d) SR 3.3.5.1.2 above vessel (Confirmatory) SR 3.3.5.1.5 zero SR 3.3.5.1.6
e. CoreSpray PumpDischarge 1 4 G SR 3.3S.1.2 2175psig Pressure - High 2 (d). 3 (d) SR 3.35.1.3 and SR 3.3.5.1.6 5 195 psig
f. LowPressureCooant 1 8 0 SR 3.3.5.1.2 290psigand Injection Pump Discharge 2 (d), 3(d) SR 3.3.5.1.3 S 110 psig Pressure - High SR 3.3.5.1.6
g. AutomaticDepressurization 1, 2 G SR 3.3.5.1.5 S322 System High Dywel 2(d) 3(d) SR 3.3.5.1.6 seconds Pressure Bypass Timer
5. ADS Tdp System B
a. Reactor Vessel Water Level 1, 2 F SR 3.3.5.1.1 2 398 Inches

- Low Low Low, Level I 2 (d). 3 (d) SR 3.3.5.1.2 above vessel SR 3.35.1.5 zero SR 3.3.5.1.6

b. Drywall Pressure - High 1 2 F SR 3.3.5.1.2 s 2.5 psig 2 (d). 3(d) SR 3.3.5.1.5 SR 3.3.5.1.6
c. AutomaticDepressurization 1. 1 G SR 3.3.5.1.5 s115 SystemInitiationTimer 2 (d) 3 (d) SR 3.3.5.1.6 seconds
d. ReactorVesselWater Level 1 1 F SR 3.3.5.1.1 2528 inches

- Low, Level 3 2 (d), 3 (d) SR 3.3S.1.2 above vessel (Confirmatory) SR 3.3S.1.5 zero SR 3.3.5.1.6 (continued)

(d) With reactor steam dome pressure 3150 psig.

BFN-UNIT I 3.3-46 Amendment No. 234

a Primary Containment Isolation Instrumentation 113.3.6.1 Table 3.3.6.1-1 (page 1 0 3)

Primary Containment Isolation InstrumentatIon APPUCABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.A

1. Main Steam Une Isolation
a. Reactor Vessel Water 1,2,3 2 D SR 3.3.6.1.1 2 398 inches Level - Low Low Low, SR 3.3.6.1.2 above vessel Level 1 SR 3.3.6.1.5 zero SR 3.3.6.1.6
b. Main Steam Line 2 E SR 3.3.6.12 2 825 psig Pressure - Low SR 3.3.6.1.5 SR 3.3.6.1.6
c. Main Steam Line Flow - 1,2,3 2 per D SR 3.3.6.1.1 S 140% rated High MSL SR 3.3.6.12 team flow SR 3.3.6.1.5 SR 3.3.6.1.6
d. Main Steam Tunnel 1.2.3 8 D SR 3.3.6.1.2 S 200°F Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.6
2. Primary Containment Isolation
a. Reactor Vessel Water 1.2.3 2 G SR 3.3.6.1.1 2 528 Inches Level - Low, Level 3 SR 3.3.6.12 above vessel I SR 3.3.6.1.5 zero SR 3.3.6.1.6
b. Drywell Pressure - High 2 G SR 3.3.6.12 S2.5 psig SR 3.3.6.1.5 SR 3.3.6.1.6
3. High Pressure Coolant Injection (HPCI) System Isolation
a. HPCI Steam Llne Flow - 1.2,3 F SR 3.3.6.12 S 90 psi High SR 3.3.6.1.5 SR 3.3.6.1.6
b. HPCI Steam Supply Line 1,2,3 3 F SR 3.3.6.12 2 100 psig Pressure - Low SR 3.3.6.1.5 SR 33.6.1.6
c. HPCI Turbine 1,2,3 3 F SR 3.3.6.1.2 S 20 psig Exhaust Diaphragm SR 3.3.6.1.5 Pressure - High SR 3.3.6.1.6 BFN-UNIT I 3.3-58 Amendment No. 234

I dW Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3)

Primary Containment Isolation Instrumentation APPUCABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITiONS SYSTEM ACTION C.1

5. ReactorWater Cleanup (RWCU) System Isolation
a. Main Steam Valve Vault 1,2.3 2 F SR 3.3.6.1.2 !52011F Area Temperature - High SR 3.3.6.1A SR 3.3.6.1.6
b. Pipe Trench Area 1.2,3 2 F SR 3.3.6.12 :51351F Temperature . High SR 3.3.6.1.4 SR 3.3.6.1.6
c. Pump Room A Area 1.2.3 2 F SR 3.3.6.1.2 :s 1521F Temperature - High SR 3.3.6.1.A SR 3.3.6.1.6
d. Pump Room B Area 1,2.3 F SR 33.6.12 :9152'F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6
e. Heat Exchanger Room 1,2.3 2 F SR 3.3.6.1.2  !~143'F Area (MestWal) SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6
t. Heat Exchanger Room 1.2,3 2 F SR 3.3.6.1.2 f. 1701F Area (East WalD SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6
g. SLC System Initiation 1,2 1(a) H SR 3.3.6.1.6 NA
h. Reactor Vessel Water 1,2.3 2 F SR 3.3.6.1.1 2528 inches Level - Low, Level 3 SR 3.3.6.12 above vessel I SR 3.3.6.1S5 zero SR 3.3.6.1.6
6. Shutdown Cooling System Isolation
a. Reactor Steam Dome 1.2,3 1 F SR 3.3.6.12 S 115 psig Pressure - High SR 3.3.6.1.5 SR 3.3.6.1.6
b. Reactor Vessel Water 3,4,5 I SR 3.3.6.1.1 2 528 Inches Level - Low, Level 3 SR 3.3.6.12 above vessel I

SR 3.3.6.15 zero SR 3.3.6.1.6

c. Drywell Pressure - High 1.2.3 2 F SR 3.3.6.12 S 2.5 psig SR 3.3.6.1.5 SR 3.3.6.1.6 (a) One SLC System Initiation signal provides logic hiput to close both RWCU valves.

(b) Only one channel per trip system required In MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

BFN-UNIT I 3.3-60 Amendment No. 234

C' -a i

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.B2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED FUNCTION OTHER CHANNELS SURVEILLANCE ALLOWABLE SPECIFIED PER RE( *UIREMENTS VALUE CONDiTIONS TRIP SYSTEM

1. Reactor Vessel Water Level - 1,2,3, 2 SR 3.3.62.1 2 528 Inches I Low. Level 3 (a) SR. 3.3.62.2 above vessel zero SR: 3.3.62.3 SR 3.3.6.2A
2. Drywell Pressure - High 1.2.3 2 SRs 3.3.622 S 2.5 psig SR 3.3.62.3 SR 3.3.6.2A
3. Reactor Zone Exhaust 1,2,3, 1 SR 3.3.6.2.1 S 100 mRlhr Radiation - High (a)(b) SR. 3.3.6.2.2 SRt 3.3.6.2.3 SR: 3.3.62.4
4. Refueling Floor Exhaust 1,2,3, 1 SR' 3.3.6.2.1 S 100 mR/hr Radiation - High (a)(b) SR 3.3.622 SR: 3.3.62.3 SR 3.3.6.2A (a) During operations vwith a potential for draining the reactor vessel.

O) Dudng CORE ALTERATIONS and dudng movement of irradiated fuel assemblies in secondary containment.

BFN-UNIT I 3.3644 Amendment No. 234

CREV System Instrumentation

%. 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Ventilation System Instiumentatbon APPUCABLE CON DITnONS MODES OR REQUIRED REFE RENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP RECWUIRED REQUIREMENTS VALUE CONDIlTiONS SYSTEM ACTiON A.1

1. ReactorVesselwater Level -
1. Reactor Vessel Hhter Level - 1,2,3,(a) 2 B SR 3.3.7.1.1 2 528 inches Low, Level 3 SR 3.3.7.1.2 above vessel I SR 3.3.7.1.S zero SR 3.3.7.1.6
2. Drywell Pressure - High 1.2,3 2 B SR 3.3.7.12 S 2.5 psig SR 3.3.7.1.5 SR 3.3.7.1.6
3. Reactor Zone Exhaust 1,2,3 I C SR 3.3.7.1.1 S 100 mR/hr Radiation - High (a).b) SR 3.3.7.12 SR 3.3.7.1.5 SR 3.3.7.1.6
4. Refueling Floor Exhaust 1,2.3, 1 C SR 3.3.7.1.1 S 100 mRlhr Radiation - High (a),.@) SR 3.3.7.12 SR 3.3.7.1.5 SR 3.3.7.1.6
5. Control Room Air Supply Duct 1,2.3, I 0 SR 3.3.7.1.1 S270 cpm Radiation - High (a)@b) SR 3.3.7.12 above SR 3.3.7.1.3 background SR 3.3.7.1.4 (a) During operations with a potential for draining the reactor vessel.

(b) Dudng CORE ALTERATONS and dudng movement of irradiated fuel assembiles hi the secondary containment.

BFN-UNIT I 3.3-69 Amendment No. 234

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