ML18037A457
| ML18037A457 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/15/1993 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18037A456 | List: |
| References | |
| IEB-93-003, IEB-93-3, NUDOCS 9309210275 | |
| Download: ML18037A457 (23) | |
Text
ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T) 9309210275 930915 PDR ADOCK 05000260 P . PDR
TABLE 3.2.8 INSTRUHENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEHS Hinimum No.
~Tr '~l Operable Per n ti Tri v 1 n rk Instrument Channel- > 470" above vessel zero. Below trip setting initiates Reactor Low Water Level HPCI.
(LIS-3-58A-D)
Instrument Channel > 470" above vessel zero. Hultiplier relays initiate Reactor Low Water Level RCIC.
(LIS-3-58A-0)
Instrument Channel-Reactor Low Water Level (LS-3-58A-D)
> 398" above vessel zero. Below CSS.
trip setting initiates ( ~
Hultiplier relays initiate LPCI.
- 2. Multiplier relay from CSS initiates accident signal ( 15).
2(16) Instrument Channel 2, 398" above vessel zero. Below trip settings, in Reactor Low Water Level conjunction with drywell (LS-3-58A-D) high pressure, low water level permissive, ADS timer timed out and CSS or RHR pump running, initiates ADS.
- 2. Bel ow tri p settings, in conjunction with low reactor water level permissive, ADS timer timed out, ADS high drywell pressure bypass timer timed out, CSS or RHR pump running, ini tiates ADS.
1(16) Instrument Channel > 544" above vessel zero. A Below trip setting permissive Reactor Low Mater Level for initiating signals on ADS.
Permissive (LIS-3-184, 185)
Instrument Channel-Reactor Low Mater Level
> 312 5/16" above vessel zero. A Bel ow tri p setting prevents (2/3 core height) inadvertent operation of (LIS-3-52 and LIS-3-62A) containment spray during accident condition.
"Only one trip system will be required to be OPERABLE during the period that the Reactor Vessel water level instrumentation modification is being performed. Hanual and automatic initiating capability of CSS and LPCI will be available, but with a reduced number of instrument channels.
ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIfICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIfICATION REASON FOR THE CHANGE NRC Bulletin 93-03, dated May 28, 1993, Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs, requested licensees implement hardware modifications necessary to ensure the level instrumentation system design is of high functional reliability for long-term operation. Each licensee was requested to implement these modifications prior to startup from the next cold shutdown after July 30, 1993.
TVA responded to Bulletin 93-03 by letter dated July 30, 1993.
This letter describes TVA's proposed modifications and also states that the implementation of these modifications would require part of the automatic initiation logic of the Emergency Core Cooling Systems (ECCS) be temporarily removed from service.
Specifically, Limiting Condition for Operation (LCO) 3.2.B, Core and Containment Cooling Systems Initiation & Control, states that the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B. This LCO requires this instrumentation be operable when the system(s) it initiates or controls are required to be operable as specified in Section 3.5. In Section 3.5:
LCO 3.5.A.4, for the Core Spray System, requires at least one core spray loop with one operable pump and associated diesel generator be operable when the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel, except with the reactor vessel head removed as specified in 3.5.A.5 or prior to startup as specified in 3 '.A.1,
Page 2 oS 10 ENCLOSURE 2 (CONTINUED)
BROMNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION LCO 3.5.A.5 states that core spray is not required to be operable when irradiated fuel is in the vessel and the reactor vessel head is removed, provided the cavity is flooded, the fuel pool gates are open, the fuel pool water level is maintained above the low level alarm setpoint, and one RHR service water pump and associated valves supplying the standby coolant supply are operable, and LCO 3.5.B.9, for the Residual Heat Removal System (Low Pressure Coolant Injection and Containment Cooling),
requires at least one Residual Heat Removal loop with two pumps or two loops with one pump per loop be operable when the reactor vessel pressure is atmospheric and irradiated fuel is'n the reactor vessel.
Table 3.2.B, Instrumentation that Initiates or Controls the Core and Containment Cooling Systems, requires a minimum of two operable trip systems for the reactor low water instrument channel (LS-3-58A-D). This instrument measures the reactor vessel water level emergency range (-155 to +60"). The safety functions initiated by this reactor vessel water level instrumentation and affected by the proposed modification are:
- 1) the Core Spray System,
- 2) Low Pressure Coolant Injection (LPCI), and
- 3) the multiplier relay from the Core Spray System that, in conjunction with drywell high pressure, initiates an accident signal.
pll Page 3 oS 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION This reactor vessel water level instrumentation is taken from two taps on opposite sides of the reactor vessel (Figure 1). Two level instruments are connected to the reference leg off of each tap (LS-3-58A-B and LS-3-58C-D). The initiating logic for this reactor vessel water level instrumentation is a one instrument out of two trip systems taken twice configuration (Figure 2).
Removal of one reference leg, including its two associated reactor vessel water level instruments, and placing these instruments in the tripped condition, would generate a trip signal on both channels and initiate the associated safety functions (Core Spray and LPCI).
Initiation of Core Spray and LPCI, when there is no drawdown in the reactor vessel water level, would result in the rapid filling and pressurization of the reactor vessel. Water would enter the main steam lines and steam lines which provide the motive force for the High Pressure Coolant Injection and Reactor Core Isolation Cooling turbines.
The alternatives for performing the proposed modification without initiating the safety functions when they are not required are to:
- 1) Perform the proposed modification on only one reference leg at a time while removing the associated (one) trip system from service. This option would require a temporary change to the Technical Specifications, or
- 2) Eliminate the need for the two reactor vessel low water level trip systems by removing the reactor head, flooding the vessel cavity, and removing the spent fuel pool gates.
This would allow performance of the modification in accordance with current Technical Specification requirements (LCOs 3.5.A.5 and 3.5.B.10).
LCO 3.5.B.10 states that if the conditions of Specification 3.5.A.5 are met, LPCI and containment cooling are not required.
Page 4 oS 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION Removing the reactor head, flooding the vessel cavity, and removing the spent fuel pool gates would significantly extend any forced or planned outage if these evolutions were not otherwise required. Therefore, a temporary change to the number of reactor vessel water level trip systems that are required to be operable is being proposed.
DESCRIPTION OF THE PROPOSED CHANGE The following changes are proposed for Unit 2:
- 1) In Table 3.2.B, Instrumentation that Initiates or Controls the Core and Containment Cooling Systems, the Minimum Number Operable Per Trip System column for the third reactor low water level instrument channel (LS-3-58A-D) will be annotated with an asterisk (*)
- 2) A footnote will be added to the bottom of Table 3.2.B as shown below:
Only one trip system will be required to be OPERABLE during the period that the Reactor Vessel water level instrumentation modification is being performed.
Manual and automatic initiating capability of CSS and LPCI will be available, but with a reduced number of instrument channels.
Page 5 oS 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION JUSTIFICATION FOR THE PROPOSED CHANGE Despite the fact that the minimum Technical Specification required number of operable reactor vessel water level trip systems will not be operable during the period that the reactor vessel water level instrumentation (LS-3-58A-D) modifications are being performed, only one reference leg will be removed from service and modified at a time. The other reference leg and its associated reactor vessel water level instrumentation will remain in service and be capable of initiating the Core Spray System, LPCI, and the multiplier relay from the Core Spray System that, in conjunction with drywell high pressure, initiates an accident signal. However, this resulting configuration is not single failure proof.
In addition to the remaining leg (either LS-3-58A-B or LS-3-58C-D) being available for automatic initiation, the following reactor vessel water level instrumentation will not be effected by the proposed modifications and will remain in service:
INSTRUMENT RANGE Flood-Up Range (2-LI-3-55) 0 to 400" While this instrumentation is not configured to substitute for the automatic initiation of the reactor vessel water level instrumentation that is being removed from service (either LS-3-58A-B or LS-3-58C-D), it will provide the operators with reactor vessel water level indication. In case of a single failure in the remaining reference leg (either LS-3-58C-D or LS-3-58A-B), the operators would manually initiate the Core Spray System and LPCI, as appropriate.
Page 6 of 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION During the performance of the reactor vessel water level instrumentation modifications, the plant will be in a cold shutdown condition with the primary system temperature at less than 212 degrees fahrenheit. While in a cold shutdown condition, the reactor vessel will be vented through the reactor vessel head vent. Therefore, the reactor vessel will be at atmospheric pressure.
In addition, the following compensatory measures will be required:
- 1) Any refueling operations, core alterations, or other activities with the potential to drain the reactor vessel will be suspended.
- 2) The reactor vessel water level instrumentation associated with the reference leg not being modified will be operable prior to removing the other reference leg from service.
- 3) The remaining reactor vessel water level instrumentation not affected by the removal of a reference leg from service will be operable.
- 4) Operators will receive special training to ensure that they are fully cognizant of the instruments which will be out of service and the degraded automatic response capability of the plant.
- 5) An operator aid will be posted on the control room panels to identify the instruments which will be out of service and the degraded automatic response capability of the plant.
'1 Page 7 oS 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION While in a cold shutdown condition, with the reactor vessel at atmospheric pressure, the worst case accident is considered to be an inadvertent, drain down of the reactor vessel due to a break in shutdown cooling piping, a leak in the reactor coolant'ressure boundary due to maintenance or valve mispositioning, or a recirculation pump seal failure. The BFN licensing basis does not assume a double-ended guillotine break of a recirculation discharge line when the vessel is at atmospheric pressure.
- Since,
- 1) piping failures are considered to have an extremely low probability of occurrence due to the low pressure and margins inherent in reactor system piping design, and
- 2) any refueling operations, core alterations, or other activities with the potential to drain the reactor vessel will be suspended, the probability of a loss of reactor vessel inventory and the need for automatic initiation of the Core Spray System and LPCI during the time period when one trip system is out of service is considered remote.
However, if a leak is assumed in the primary system piping during the performance of this modification, and there is a single failure in the remaining trip system that prevents automatic initiation of the safety functions, sufficient instrumentation will be available for the operators to identify the event. The equipment required to mitigate the event will be available and the operators will be able to manually initiate the safety functions.
Page 8 oS 10 ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION The operator would be initially aware of a loss of coolant inventory upon receipt of an abnormal water level alarm at approximately 555" above vessel zero (6" below normal water level). Other indications would subsequently occur if the loss of inventory was not stopped (e.g., an automatic depressurization system blowdown permissive alarm at approximately 546" above vessel zero). These two indications come from reactor vessel water level instrumentation which is unaffected by the proposed modifications. If the reactor water level reached 398" above vessel zero, the Emergency Operating Instructions direct the operator to manually initiate LPCI if it was not automatically initiated.
Therefore, requiring only one trip system for the reactor low water level instrument channel (LS-3-58A-D) be operable during the period that the Reactor Vessel water level instrumentation modification is being performed is justified due to:
- 1) the low primary system temperature and pressure,
- 2) the low probability of an event that would result is the drain down of the reactor vessel, I
- 3) the automatic initiating capability of the remaining reactor vessel water level trip system,
- 4) the compensatory measures, and
- 5) the other reactor vessel level instrumentation and equipment that is available for manual operator intervention in the event of a plant transient or accident.
ENCLOSURE 2 BROMNS FERRY NUCI EAR PLANT (CONTINUED) t(BFN)
Page tI oS 10 UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION FIGURE 1 Reactor Uessel Water Level Instrumentation LT 3.$ 6A Through 0 Close MSIVa LT 3.56A Through 0 Inltlate ATWS/AAI RPT, NPCI, RCIC, AHA, Core Spray, Start Diesel Generator, ADS Perrelaslre LT 3.203A Through 0 SCRAM, Primary Contalnmanl laol ~ lion LT 3.208A Through 0 HPIC sort ACIC Turtrtne Trip CONDENSINO CHAMBERS LT 3.55 .-. I7 9.3 LLJI 3 ~ Ss 599' L L L L L L L IT 2s JSA~LI 32Lul" 3'6 I
358 A
358 B,' 3 ~ 58 356 B
I.EVEI.
ABOVE 3se C
356 0
358 C
3 Sg 0
3.J6.-- 3~a 25 32 VESSEL F.W.
CONTROLS
~ 3.58A QLI 9 5 ZERO.
AUTO BLOWDOWN 1
LI 3 ~ 58B INTERLOCK 9.5 I
t LT LT LT LT LT LT LT LT LT T L 208 208 203 203 3.53 ~ 18 360 203 203 208 208 A B A 0 C 0 C 0 1
517'ee. F.w.i LI CONTROLS
"~ AHA INTEAL'K ICOHT. sPRAYI AHR INTKAL'K~
LI 3.208 9.5 LI 3 53 AUTO SLOWDOWN g.. 3-62 LT r LT 9.3 (CONY. SPRAY) IHTEALOCK JP IVI 3 ~ 62A e2 pj3,62 JP 8 143.5 QLI 3 ~ 52 LOWKA TAP LOWER TAP 9.3 BKLOW COAE PLATE gPT 3 ~ st
Page 10 of IO ENCLOSURE 2 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 343T)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION FIGURE 2 Reactor Vessel Water Level Instrumentation Initiating Logic DRYWELL PRESS. HIGH REACTOR PRESS. LOW LS-3-58A RX VESSEL LVL LOW LS"3-58B MOM PB (TYPICAL OF 4) 9"3 HS LS-3-58C LS-3-58D 74-I5I (I52)
RESET SEAL-IN RHR AUTO INIT SIGNAL LEGEND MD NOTE: The RHR automatic initiation logic is shown. The automatic initiation logic for Core Spray is similar.
ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION DESCRIPTION OF THE PROPOSED TECHNICAL SPECIFICATION CHANGE In Table 3.2.B, Instrumentation that Initiates or Controls the Core and Containment Cooling Systems, the Minimum Number Operable Per Trip System column for the third reactor low water level instrument channel (LS-3-58A-D) will be annotated with an asterisk (*). A footnote will be added to the bottom of Table 3.2.B as shown below:
Only one trip system will be required to be OPERABLE during the period that the Reactor Vessel water level instrumentation modification is being performed. Manual and automatic initiating capability of CSS and LPCI will be available, but with a reduced number of instrument channels.
BASES FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed TS change is judged to involve no significant hazards considerations based on the following:
Page 2 oS 4 ENCLOSURE 3 (CONTINUED)
BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION
- 1. The proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.
The temporary change to the number of reactor vessel water level instrumentation trip systems required to be operable does not increase the frequency of the precursors to design basis events or operational transients analyzed in the Browns Ferry Final Safety Analysis Report. Therefore, the probability of an accident previously evaluated is not significantly increased.
While in a cold shutdown condition and the reactor vessel is at atmospheric pressure, the worst case accident is considered to be an inadvertent drain down of the reactor vessel due to a break in shutdown cooling piping, a leak in the reactor coolant pressure boundary due to maintenance or valve mispositioning, or a recirculation pump seal failure.
The other reference leg and its associated reactor vessel water level instrumentation will remain in service and be capable of initiating the required safety functions. In addition, several other reactor vessel water level instruments will remain in service and will provide the operators with reactor vessel water level indication in case of a single failure in the remaining reference leg. Based on input from this instrumentation, the operators would manually initiate the required safety functions, as appropriate. Therefore, the proposed changes will not significantly increase the consequences of an accident previously evaluated.
~ 1 0
ENCLOSURE 3 (CONTINUED)
Page 3 of 4 BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION
- 2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The temporary change to the number of reactor vessel water level trip systems required to be operable does not create a new pathway for radioactive material to reach the environment. The out of service instruments are used to indicate plant parameters and to initiate systems required to mitigate accidents or plant transients. The remaining instrumentation will still be available to automatically initiate the required function. The operators will receive special briefings to ensure that they are fully cognizant of the instruments which will be out of service and the degraded automatic response capability of the plant. The operators will be have sufficient time to manually initiate the required functions if a single failure occurs in the remaining instrumentation. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed amendment does not involve a significant reduction in the margin of safety.
The proposed temporary change to the number of reactor vessel water level trip systems required to be operable does not change the licensing or design basis limits for the initiation of protective actions. Compensatory measures will adequately compensate for the lack of single failure proof automatic initiation capability of the Core Spray and Low Pressure Coolant Injection systems during the period in which the effected systems will be out of service.
Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.
ENCLOSURE 3 (CONTINUED) Page 4 of 4 BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION CONCLUSION TVA has evaluated the proposed amendment described above against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has
~
determined that the proposed amendment will not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility for a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Thus, TVA has concluded that the proposed amendment does not involve a significant hazards consideration.
W Ill% 4