ML20212C250
| ML20212C250 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/23/1997 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9710290125 | |
| Download: ML20212C250 (11) | |
Text
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Tennessee Valley Authority, Post office Box 2000, Oecatur, Alabama 3%09-2000 October 23, 1997 U.S. Nuclear Regulatory Commission 10 CFR 50.54 (f)
ATTN:' -Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-259 Tennessee Valley Authority
)
50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - REVISION 1, RESPONSE TO NRC GENERIC LETTER (GL) 96-06, ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENT CONDITIONS By letter dated January 28, 1997 (Reference), TVA responded to GL 96-06.
In that letter, TVA summarized actions taken, conclusions reached, and any follow-up actions required to ensure containment integrity during design basis accident conditions.
In order to keep the staff abreast of TVA's long term actions towards resolution of this '3L issue, TVA is providing this
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revised BFN response.
This response will discuss only those issues related to BFN. -Accordingly, TVA is revising the BFN portion of the GL response, and reissuing it in its entirety.
Those portio.s related to Sequoyah Nuclear Plant and Watts Bar Nuclear Plant are not impacted by this revision and; therefore, will not be discussed.
In the January 28, 1997 letter, TVA provided response for two speciric BFN system p'netrations based on the American Society of Mechanical Engineers (ASME)Section III Appendix F methodology, and one response was based on valve leakage.
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U.
S. Nuclear Regulatory Commission Page 2 October 23, 1997 The conclusions for these evaluations have been revised as shown below:
The Sampling System and Main Steam System drain lines, which were previously analyzed for overpressure utilizing the ASME Appendix F methodology, have been reevaluated under actual plant conditions for these systems.
The results of the analysis for the sampling system showed that the design pressure under the code of record for the system, USAS B31.1.0-1967 with addendums, would not be exceeded.
The analysis showed that the Main Steam drain lines are nut susceptible to thermal overpressurization.
The Drywell Floor and Equipment Drain Sump discharge lines were originally justified acceptable based on the fact that valve leakage will preclude overpressurization.
TVA has reevaluated this conclusion and confirmed that because of this leakage, overpressurization of the system will not occur and the containment pressure boundary will retain its design margin.
However, since pressure relief is not a design function of the valve, a design change will be implemented to provide overpressure protection in accordance with the plant design basis.
The remaining conclusions reached in TVA's original response are unchanged and are repeated below for completeness:
TVA conducted a technical evaluation to determine if containment air cooler cooling water system is susceptible to either water hammer or two-phase flow conditions during postulated accident conditions.
TVA has also evaluated piping systems that penetrate the containment to determine if they are susceptible to thermal expansion of fluid such that overpressurization of piping may occur.
BFN has Drywell Coolers which are served by the Reactor Building Closed Cooling Water (RBCCW) system.
Drywell Coolers provide cooling to the drywell during normal operation and are not credited for post-accident Primary Containment cooling.
The loss of cooling capability is not a concern.
The RBCCW system does not isolate during a postulated Loss-of-Coolant-Accident (LOCA); therefore, over pressure is not a concern.
However, the portion of the RBCCW system inside the drywell is credited for maintaining primary containment boundary during and following a LOCA and; therefore, the potential for water hammer was analyzed.
.. -.. -. - _ ~. - -.
U.S. Nuclear-Regulatory Commission
'~Page 3j October 23, 1997 TVA has determined that the RBCCW-system.is subject to' e
water. hammer upon resumption of cooling water. flow-following a LOCA concurrent with a Loss-of-Offsite-Power.
The water hammer induced pressure loads are within the design capacity of the containment air coolers.
Therefore, the primary containment boundary will be maintained, BFN_ penetrations that are isolated during an event.were e
also analyzed.-
It was-determined that post-accident
-the: mal expansion-of fluid will not cause a containment 3:,olation boundary failure.
However, a procedure change was implemented that ensures the Demineralized Water system piping is partially drained following each use and the
. piping inside primary containment is open to primary containment atmosphere during power operation.
This eliminated the potential for overpressurization.
BFN Unit 1 is shutdown, defueled, and under administrative hold.
The conditions described by this GL will be addressed prior to its return to service.
A detailed discussion of the information requested in the-GL is contained in Enclosure 1. contains a list of the commitments that remain
-open from the original--reply and those made in this letter.
If:you have any questions please contact me at (205)
~
-729-2636.
Spcerelyr t
~ ' O _.v T. E. Abney Manager o icensi g and Indu try Affa rs Subscribed and sworn.before me on'. this %3nd Day _of October.1997.
Q.hD GN\\ t%
i f Notary Public
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4 Comunisalon Empires 10/0644 My Commission Expires t
Enclosures cc: ' See page 5
l U. S. Nuclear Regulatory Commission Page 4 October 23, 1997
Reference:
TVA Letter to NRC dated January 28, 1997, " Browns Ferry Nuclear Plant (1 BEN), Sequoyah Nuclear Plant (SON), and Watts Bar Nuclear Plant (WBN) Unit 1 Response to NRC Generic Letter l
(' L) 96-06, Assurance of Equipment Operability and G
l Containment Integrity During Design-Basis Accident i
Conditions, Dated September 30, 1996."
l
U.S. Nuclear Regulatory Commission Page 5 October 23, 1997 l
- Enclosures cc (Enclosures)
Mark S. Lesser, Branch Chief I
U.S. Nuclear Regulatory Commission Region II-l 61 Forsyth Street, S. W.
Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 4
A
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS-1, 2, AND 3 NRC GENERIC LETTER (GL) 96-06 ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINEMNT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS, REVISION 1
Background
the following provides the response for BFN on the subject GL.
Pursuant to Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54 (f), all holders of operating licenses for nuclear power reactors, except for those licenses that have been amended to possession-only status, are requested to submit the following written information.
NRC Requested Information Within 120 days of the date of this GL, addressees are requested to submit a written summary report stating actions taken in response to the requested actions noted below, conclusions that were reached relative to susceptibility for.
water hammer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued operability of affectud systems and components as applicable, and corrective actions that were implemented or are planned te be implemented.
If systems were found to be susceptible to the conditions that are discussed in this GL, identify the systems affected and describe the specific circumstances involved.
Requested Action (s)
Addressees are requested to determine:
(1) if containment air cooler cooling water systems are susceptible to either water hammer or two-phase flow conditions during postulated accident conditions;
'(2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
In' addition to the individual addressee's postulated accident conditions, these items should be reviewed with respect to the scenarios referenced in the GL.
If systems are found to be susceptible to the conditions discussed in this GL, addressees are expected to assess the operability of affected systems and take corrective action as
i.
-appropriate in'accordance with the requirements stated in110 CFR Part 50 Appendix B and as required by the plant Technical Specifications.
TVA Response
-Evaluation Summary TVA performed an evaluation of each water filled system penetrating primary containment.
Each penetration was reviewed'for the potential to compromise the containment
-integrity from thermal expansion of water because of elevated-containment. temperatures in isolated sections of piping during and following a LOCA.
The evaluation determined the maximum pressure that an isolated section of piping would encounter due to elevated temperatures during and following a LOCA would not exceed-required structural limits for the affected systems; thus, containment pressure boundary integrity would be maintained.
For the Drywell Floor and Equipment Drains, valve leakage will preclude overpressurization.
Since pressure relief is not a design function of the valves, a design change will be implemented to provide a designed method of overpressure protection.
The Reactor Building Closed Cooling Water (RBCCW) system is the only closed cooling loop inside the containment.
Because the RBCCW system does not isolate during a postulated LOCA, it was analyzed for potential water hammer.
Evaluation Details Sampling System:
The Sampling system was previously analyzed for post-LOCA overpressure utilizing ASME Section III Appendix F methodology.
This system has piping inside primary containment and is normally isolated during power operation.
The concern is overpressure-of the containment penetration
-between the inboard and outboard isolation valves.
The containment penetration and isolation valves are pneumatic, fail-closed globe valves.
These are oriented by design such that the above seat is toward the Reactor Pressure Vessel (RPV).
Therefore, high pressure between the inboard and outboard valves will unseat the inboard valve and relieve back to the RPV.
The containment penetration, affected piping, and isolation. valves were evaluated and determined to satisfy their structural design basis under the code of record for the system, USAS B31.1.0-1967 with addendums, for the pressure required to ljft the valve off the seat and relieve back to the vessel.
Therefore, the sampling system has been found acceptable.
Main-Steam Drain Lines:
The main steam drain lines were El-2
- - -.-. - - - ~.
L..
i originally assumed to be water solid during power operation.
-These were previously analyzed utilizing ASME'Section III Appendix F methodology and-found to be acceptable.
Operation of the drains was re-reviewed and it was determined tny the i
lines would contain only a small amount of water from i
condensed steam when isolated and will have ample volume for i
post _ accident thermal expansion.
Therefore, the main steam
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drain line penetration is'not susceptible to thermal overpressurization.
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l Demineralized'Wa'.er:
This system has the potential to be l
affected by overpressurization during a postulated LOCA if the J
-piping inside containment is completely filled with water.
I TVA performed an evaluation which hLs determined that water is 2
normally drained between the outboard isolation valves and a blocking valve during Appendix J testing.
Air retained in the A
system following the Appendix J testing, prevents the primary containment penetration from being overpressurized during a postulated LOCA.
Therefore, because of the as-left configuration of the system, it-will not be effected by conditions described by the Gl.
However, plant procedures were revised, adding actions that assure the system is partially drained following use and that piping inside the primary containment is open to the primary containment during power operation.
Standby Liquid Control (SLC) system:
The SLC system does not automatically isolate during a LOCA.
The SLC system is open to the reactor vessel.
The thermal expansion of water in the primary containment penetration would be relieved back to the vessel-through an inboard check valve inside containment.
Therefore, the conditions described in the GL do not pertain to this system.
Drywell Floor and Equipment Drains:
The primary containment isolation valves are both located cutside of primary containment._ Therefore, the piping between the two valves (and water in the piping) is not subject to the high temperature environment.inside containment following a LOCA.
There'is a section of piping between the isolation valve outside containment and valve inside primary containment which could be exposed to the LOCA temperature environment.
-However, the valves inside containment.do not provide leak tightness and, as such, the maximum pressure _is bounded by the
-design margin since any thermally induced pressure increase will be bled off through the leaking valves to the floor. sump.
Since pressure relief is not a designed function of the
-valves,:a design change will be implemented to provide a designed method of overpressure protection.
Reactor Water Cleanup System:
The system will be operating normally.at the time of the postulated event.
Portions of the system piping and_ penetrations effected by this issue are El-3
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O approximately 500 F during normal operation.
There is no concern for heat gain and pressure increases resulting from temperature increases, as the line will actually cool.
Therefore, the system is not susceptible to conditions described in the GL.
Residual Heat Removal (RHR) Shutdown Cooling:
The RHR shutdown cooling supply line isolation valves will close on an accident signal.
The isolated pipe section is provided with a pressure relief line back into the upstream process piping inside primary containment which will prevent overpressuration.
Therefore, the system is not susceptible to conditions described.in the GL.
Reactor Building Closed Cooling Water (RBCCW) System:
This system serves the Primary Containment Drywell Coolers.
The coolers are used for normal operation to cool primary containment.
The RBCCW system continues to operate post-LOCA as the plant procedures allow the coolers to be used if available post-accident.
However, accident analyses do not take credit for the coolers.
The RBCCW is a closed loop system with a surge / head tank outside containment.
The RBCCW isolation is remote manual from the control room for the discharge line and an in-line check valve for the supply side.
RBCCW is not isolated automatically following a LOCA.
The RBCCW system piping is considered part of the containment boundary.
As such, there are no primary containment isolation valves that automatically respond to a LOCA or a concurrent Loss-of-Offsite-Power (LOOP).
The cooler coils are assumed to void (boil dry) at the onset of a LOCA concurrent with a LOOP.
The voiding that occurs in the tubes $s postulated to result in a water hammer in the system.
Evaluation of the postulated water hammer concludes the water hammer pressure wave and resulting dynamic loads are acceptable and are within the design capacity of the system.
Thermal expansion of the water following the LOCA is accommodated by the system vented surge tank and will have no affect on the RBCCW system.
Therefore, the primary containment boundary will be maintained.
Two-phase flow is not a consideration at BFN, as the coolers are not credited with accident mitigation.
Therefore, containment heat removal is unaffected.
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Conclusion The conditions described by GL 96-06 exist at BFN.
- However, these conditions were determined to pose no threat during the mitigation of an event.
Specifically the following has been established:
The containment air coolers served by the RBCCW system are susceptible to water hammer during postulated accident condition; however, the effects of the water hammer are contained within the cooler and the intensity of the water hammer is within the design capacity of the coolers.
Two-phase flow is not a consideration for the RBCCW system post-LOCA.
BFN accident analysis does not credit heat removal by RBCCW.
Affected piping systems that penetrate the primary containment have been evaluated / analyzed to determine susceptibility to the conditions described by Generic Letter 96-06:
- The Drywell Floor and Equipment Drains system is acceptable based on leakage through valves which will avoid thermally induced pressure increase above the design of the system.
A design change will be implemented to provide a designed method of overpressure protection.
- The Demineralized Water system has the potential to be effected by overpressurization during a postulated LOCA if the piping is completely filled with water.
In light of this, TVA implemented procedure changes that assure the system is sufficiently drained following use and is open to containment during power operation.
Other systems are either not susceptible, or can safely accommodate thermally induced overpressurization.
BFN Unit 1 is shutdown, defueled, and under administrative hold.
The conditions described by this GL will be addressed prior to Unit 1 return to service.
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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY.
BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, and 3 NRC GENERIC LETTER (GL) 96-06, ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS REVISION 1 Commitment (s) j 1.
A design change will be implemented to pro'71de overpressure protection on the Unit 2 Drywell Floor and Equipment Drain Sump discharge lines prior to the restart from cycle 10 refueling outage.
2.
A design change will be implemented to provide overpressure protection on the Unit 3 Drywell Floor and Equipment Drain Sump discharge lines prior to restart i'
from cycle 8 refueling outage.
3.
BFN Unit 1 is shutdown, defueled, and under i
administrative hold.
The conditions described in this GL will be addressed prior to Unit 1 return to service.
l
^l 1
.d 1.
Commitment 3 was identified as part of the January 28, 1997 letter. As such, no new
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commitment has been added.
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