ML20063B951
| ML20063B951 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/21/1994 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M85255, TAC-M85256, TAC-M85257, NUDOCS 9402030242 | |
| Download: ML20063B951 (8) | |
Text
V A
Tennessee Va' ley Autnonty. Post OEce Box 2000 Decatur, Alabama 35609 JAN 211994 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-259 Tennessee Valley Authority
)
50-260 50-296 EAOWNS FERRY NUCLEAR PLANT (BFN) - ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION (TS) 328, LOW PREESURE COOLANT INJECTION (LPCI) OPERABILITY WITH THE RESIDUAL HEAT REMOVAL (RHR) BYSTEM ALIGNED FOR SHUTDOWN COOLING (SDC) (TAC NOS. M85255, H85256, and M85257)
TVA submitted a proposed technical specification change (TS 328) on December 23, 1992, that wodld allow LPCI to be considered operable when the RHR system is aligned in the SDC mode with the reactor shut down.
TVA provided additional information on August 12, 1993, in response to an NRC request for additional information, dated June 9, 1993.
Per telephone conversations with BFN's Project Manager during the week of January 3, 1994, NRC requested additional information to support the proposed technical specification change. provides the requested information.
A summary of the commitment contained in this letter is provided in Enclosure 2.
If you have any questions, please contact me at (205) 729-2636.
Sincer ly g g2 Pedro Salas Manager of Site Licensing i
Enclosures cc:
See page 2
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9402030242 940121 j
PDR ADOCK 05000259 Q
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U.S. Nuclear Regulatory Commission Page 2 JAN 211994 Enclosures cc (Enclosures):
Mr. R. V. Crlenjak, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35611 i
Mr.
D.
C. Trimble, Project Manager U.S.
Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. J.
F. Williams, Project Manager U.S.
Nuclear Regulatory Commission i
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY:
BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2,
AND 3 P
ADDITIONAL INFORNATION - LOW PRESSURE COOLANT INJECTION OPERABILITY BACKGROUND TVA submitted Technical Specification (TS) 328 on December 23, 1992 (Reference 1).
This proposed TS change would allow the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) system to be considered operable when capable-of manual realignment from the shutdown cooling (SDC) mode.
The proposed change only applies when the plant is in hot shutdown condition with the primary system at less than 105 psig or when the plant is in the cold shutdown condition with the primary system at atmospheric pressure.
On June 9, 1993, NRC requested TVA describe the " limiting' accident" for each operating mode affected by the proposed' I
technical specifications and the analysis of the' postulated sequence of events, including time to core uncovery if mitigating actions were not taken (Reference 2).
TVA provided the requested additional information (RAI) on August 12, 1993 (Reference 3).
TVA demonstrated that the " limiting accident" for each. operating mode affected by the proposed technical specifications would be bounded by BFN loss of coolant accident (LOCA)-analyses.. TVA also postulated a " worst case accident" for both operating modes affected by the proposed TS change..
For the cold shutdown condition, TVA postulated that the most severe accident was an inadvertent draindown of the reactor vessel due to a break in SDC piping or a leak in the reactor coolant pressure boundary caused 1
l by maintenance or valve mispositioning.
TVA explained that any.
draindown of the reactor vessel would be terminated by isolation of the SDC isolation valves due to a primary containment isolation system (PCIS) signal on low water level (538" above-vessel zero).
TVA later clarified that BFN Technical Specifications do not require the PCIS to be operable in,the cold shutdown condition with the reactor at atmospheric pressure.
However, BFN operating practice is to maintain that portion of I
the PCIS (i.e., reactor vessel low water level instrumentation) that initiates an SDC isolation operable when RHR is in the SDC mode.
I 1
4 REQUEST FOR ADDITIONAL INFORMATION (RAI)
On January 7, 1994, BFN's Project Manager advised TVA that additional information was needed to support the proposed I
technical specification for the cold shutdown case.
The NRC requested TVA provide a time to core uncovery for a " postulated event" in cold shutdown and additional information concerning RHR System valve interlocks that would prevent reactor draindown to the supprecsion pool.
BFN's Project Manager also advised TVA that no further j
information was needed to support the safety evaluation for the proposed technical specification for the hot shutdown case.
TVA RESPONSE TVA cannot postulate any credible events that would uncover the core during cold shutdown.
In the initial RAI response, TVA postulated a pipe break or a reactor draindown due to maintenance or valve mispositioning errors as a " worst case accident" for the purpose of showing that the event would be terminated by a SDC isolation due to a PCIS signal on low water level.
However, TVA never considered them credible events while in cold shutdown with the reactor at atmospheric pressure.
Per BFN Final Safety Analysis Report (FSAR) Section 14.4.3, a pipe break is only assumed if the component to rupture is subjected to significant pressure.
Reactor draindown due to maintenance or valve mispositioning errors is not considered credible since RHR system valves that would allow a significant draindown path are equipped with interlocks that prevent inadvertent draindown.
The interlocks that provide protection against valve positioning errors that could cause an inadvertent reactor vessel draindown are as follows:
1.
When an RHR SDC pump suction valve (FCV-74-2, 13, 25, or 36) is open, the corresponding suppression pool pump suction valve (FCV-74-1, 12, 24, or 35) cannot be opened or vice versa (see attached figure).
2.
When an RHR SDC pump suction valve is open, the l
corresponding RHR suppression pool return line valve (2-FCV-74-57 or 71) cannot be opened or vice versa.
(Interlock will be installed on Units 1 & 3 prior to restart.)
3.
The RHR minimum flow valve (FCV-74-7 or 30) is designed to automatically open on low flow.
Prior to placing the RHR into the SDC mode, the operating procedure requires the operator to bypass the minimum flow valve logic, using a handswitch, to keep the minimum flow valve closed (precludes El-2
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draining of the reactor vessel to suppression pool).
The RHR minimum flow valve is also designed to automatically close if a SDC suction valve is not fully closed and neither pump on the corresponding loop is operating.
In addition, BFN's RHR system operating procedure contains precautions that address potential drain paths from the vessel to the suppression pool.
There are no other paths in which a valve misalignment error could cause the reactor vessel to drain rapidly enough to allow core uncovery.
There are one-inch lines off the RHR SDC piping that could drain the vessel, however, these are too small to be considered capable of draining the reactor vessel prior to operators manually realigning RHR to the LPCI mode.
TVA calculated approximate times to core uncovery for various sized pipe openings.
This calculation assumed the reactor is in cold shutdown at atmospheric pressure, the draindown starts at 538" above vessel zero (low water level), and core uncovery is at the top of the active fuel.
The calculated draindown times are listed in the table below:
j DRAINDOWN OPENING TIME i
(MINUTES)
(SQUARE INCHES)
(NOMINAL l
PIPE DIAMETER - IN.)
l 3.3 113.04 12 4.8 78.5 10 7.5 50.24 8
13.3 28.26 6
j l
29.9 12.56 4
As stated in TVA's initial RAI response, the operator would be prompted to manually realign RHR from the SDC mode to the LPCI l
mode upon receipt of reactor vessel low water level i
scram / isolation indication and alarms (at 538" above vessel zero).
The actions required for manual realignment can be completed in approximately 3.5 minutes from the receipt of the scram / isolation indication and alarms.
Therefore, sufficient time is available to manually realign RHR to the LPCI mode.
El-3
f REFERENCES 1.
Letter from TVA to NRC, dated December 23, 1992, Technical Specification (TS) No. 328 - Low Pressure Coolant Injection (LPCI) Operability When Residual Heat Removal System (RHRS) is Aligned to Shutdown Cooling (SDC) Mode - Units 1, 2,
and 3 2.
Letter from NRC to TVA, dated June 9, 1993, Request for Additional Information - Proposed Technical Specification Amendment Regarding Low Pressure Coolant Injection Operability With Residual Heat Removal Aligned for Shutdown Cooling (TAC Mos. M85255, M85256, and M85257) 3.
Letter from Itu to NRC, dated August 12, 1993, Response to Request for Additional Information - Proposed Technical Specification Amendment Regarding Low Pressure Coolant Injection Operability With Residual Heat Removal Aligned for Shutdown Cooling (TAC Nos. M85255, M85256, and M85257)
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ENCLOSURE 2
[
TENNESSEE VALLEY AUTHORITY.
BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2,
AND 3
SUMMARY
OF COMMITMENT I
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Electrical interlocks will be installed between the Residual Heat Removal shutdown cooling pump suction valves (FCV-74-2, 13, 25,
- 36) and the corresponding suppression pool return line valves (FCV-74-57 & 71) prior to the restart of Units 1 and 3.
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