ML20147D870
| ML20147D870 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/02/1988 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8803040189 | |
| Download: ML20147D870 (20) | |
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LU.S. Nuclear Regulatory Commission-jATTN: ' Document Control Desk
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'In the Matter of
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Docket Nos. 50-327
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' SRQUOYAH MUCLRAR PLANT (SQN! - An'*ENDIX R - REQUEST.FOR ADDITIONAL INFORMATION j.
References:
- 1. NRC letter t o S. A. White dated February !26,1988,-;"Appendix
,f R Concerns Related to Sequoyah Unit 2 Restart"
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- 2. TVA letter to NRC dated February 27, >1988,. "Sequoyah Nuclear Plant ~(SQN) - Appendix R Fire Protection' Safe Shutdown Logic Calculation" v
This letter provides a response to your letter requesting additional
'information concerning the Appendix R evaluation of SQN (reference 1).-
TVA-has alt recently provided NRC copies of the current cafe shutdown loSicand the 1.ppendix R Resolution Review Team Final Report (reference 2).
Rnclosed are responses to your 26 questions on the Appendix R program.
During our. review of these questions, I note that approximately one-fourth of:
the requested information was not directly tiedLto 10 CFR 50 Appendix R..
The questions raised go beyond the requirements'of Appendix R and the design-j.
basis for Sequoyah. Even so, we have indicated what actions would likely be taken if these circumstances existed.
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However, in the interest of resolving these issues in an' expeditious manner, I have enclosed information and references responding to each of the questions. Copies of the addi.tional documentation referenced in the responses are.available at the-TVA' office in Rockvill, Maryland. TVA can provide this information for the docket if desired.
There are several points which should be considered during review of these-resporses. First, review of SQN plant layout, equipment, and peccedures has P-
. demonstrated that. safe shutdown, which is the touchstone of NRC's. fire ~
protection requirements,'has been evaluated-by TVA and NRC staff by analysis and plant walidown.
In keeping with Appendix R, the evaluation assumes that euch of the important. equipment and diagnostic ~ capability normally available tothe' operator has been rendered unavailable.
F See md, since all possible consequences of. a fire cannot be. anticipated in adtanee, equipmen utilization and operator actions will depend ~on what is c
l ava ilable.. To ' choose certain isolated occurce:4es--to ask "What -if f"--as is -
the case with some~of thesp questions -is contrary to the bases upon.which" w
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. MAR 021988 U.S. Nuclear Regulatory Connission Appendix R was developed. See for example Generic 1.etter 86-10, enclosure 2, page 28, which recognizes that a reasonable approach to fire protection should be taken.
Specifically, it states that a "worst case fire nond not be postulated to be simultaneous with nonfire-related failures in safety systems, plant accidents, or the most severe natural phenomena." It should be noted that the SQN Appendix R program was au0lted several times by NRC in 1985-1987, and that the NRC has drafted an Appendix R Safety Evaluation Report indicating acceptability of the TVA program.
In rummary, NRC fire protection requirements specify that given a damaging fire at an unspecified but critical location, plant layout, equipment, and operator actions must ensure safe shutdown. This is without credit for installed fire suppression and the fire brigade which are in themselves likely to be effective in extinguishing fires in critical areas. At SQN, we have taken the steps necessary to ensure that the plant is adequately protected against the inception and growth of fires and that safe shutdown is ensured.
Finally, I note that you consider some of the issues restart items for SQN and therefore request your expeditious review of this information.
Very truly yours, TENNESSEE VALLEY AUTHORITY R. L. Gridley, irector Nuclear Licensing and Regulatory Affairs Enclosure ec: See page 3
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U.S.'Wuclear Regulatory Commission MAR 01198B
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36 Mr. K. P..Barr, Acting Assistant Director
>for Inspection Programs idb W TVA Projects Division U.S. Nuclear Regulatory Commission Reglon II
-101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
~4 Mr. C.'O. Zech, Assistant Director for. Proj ects TVA Projects Division "U.S. Nuclear Regulatory Commission f-(
one White Flint, North L11355 Rockville Pike Rockville, Maryland 20852 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 P
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~ 2-MAR 021988 U.S. Nuclear Regulatory Commission Appendix R was developed. See for example Generic Letter 86-10, enclosure 2, page 28, which recognizes that a reasonable approach to fire protection ~should be taken. Specifically, it states that a "worst case fire need'not be postulated to be simultaneous with nonfire-related failures in safety systems, plant accidents, or the most severe natural phenomena." It should be noted that the SQN Appendix R program was audited several times by NRC in11985-1987, and that the NRC has drafted an Appendix R Safety Evaluation Report indicating acceptability of the TVA program.
In w memey, NRC fire protection requirements specify that given a damaging fire at an unspecified but critical location, plant layout, equipment, and operator actions must ensure safe shutdown. This is without credit for installed fire suppression and the fire brigade which are in themselves likely to be effective in extinguishing fires in critical areas. At SQN, we have taken the steps necessary to ensure that-the plant is adequately protected against the inception and growth of fires and that safe shutdown is ensured.
Finally, I note that you consider some of the issues restart items for SQN &nd therefore request your expeditious review of this information.
Very truly yours, TENNESSEE VALLEY AUTHORITY R. L. Cridley, irector Nuclear Licensing and Regulatory Affairs Enclosure cc: See page 3 L
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' -U.S. Nuclear Regulatory Commission MAR 021988 cc (Enclosure):
Mr. K. P. Barr, Acting Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. G. G. Zech, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission one White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 I
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ENCLOSURE OUESTION 1.
Provide the calculations supporting the TVA Task Group's opinion'that the release of RCS water into the containment with only technical specification allowable levels of failed fuel does not result in an unacceptable off-site dose and thus the need for containment integrity as described in A.2 - Radiation Release to the Environment - of TVA's "Task Group Disposition of Issues."
TVA RESPONSE TVA has performed a conservative scoping calculation of reactor coolant system spills into containment during an un-isolated purge operation to determine the offsite dose expected.
The analysis performed (SQNAPS3-086) assumes an expected failed fuel fraction t
corresponding to ANSI /ANS Standard 18.1-1984, fission product activity in the primary system, an iodine peaking factor of 10, no holdup in containment, no dilution in containment, instantaneous spill of coolant, and an efficiency of the containment purge HEPA filters and charcoal beds consistent with that used for the containment fuel handling accident described in FSAR chapter 15.5.6.
A plot for various fractions of reactor coolant inventory spilled to containment is shown in figures 1 and 2.
Spill fractions up to 100 percent are shown even though fractions of that magnitude would be indicative of a f
LOCA size spill. As can be seen from the plot, even with the bounding i
assumptions, the offsite doses are considerably less than 10 CFR 100 limits.
In the rare event of initial RCS conditions approaching levels of technical specification failed fuel limits or other compounding failures, offsite doses would be expected to be higher. However, it should be noted that the failed fuel history for Sequoyah has been better than that tusumed in the ANSI standard. Also, the following defense in-depth nust be noted:
1.
From a practical standpoint, the only path for release of radioactivity from the containment to the environment is through the purge system.
2.
The exhaust line has three fail closed valves in series.
If any of these valves close, a release will ba terminated.
3.
Should the valves not close, there are HEPA filters and charcoal beds in the exhaust lines.
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On the supply side, the same three valve combination exists.,however, there are no HEpA filters or charcoal beds in the line. There are four other fail closed dampers in the supply line that the operator could close to terminate a release of radioactivity to the environment.
5.
All valves and dampers close automatically if a radiation signal is received.
6.
There are a variety of other actions the operator can take to preserve containment integrity or mitigate offsite releases.
Therefore, the release of radioactivity to the environment due to concurrent fire, purge at power, and all required spurious actions is not a probable event.
This is not an Appendix R requirement, nor a TVA Appendix R commitment.
Appendix R requirements are based on ensuring safe shutdown rather than-mitigating accident consequences. Appendix R states ".
. Because fire may affect safe shutdown systems and because the loss of function of systems-usud to mitigate the consequences of design basis accidents under postfire
-conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems to mitigate the l
consequences of design basis accidents." TVA has a guaranteed path, approved by the NRC, which ensures safe shutdown during Appendix R fires.
i OUESTION 2.
Provide the information supporting the TVA Task Croup's conclusion that boiling of the spent fuel pool is not a technical concern. With respect to this conclusion address the HVAC syscem's ability to rapidly and effectively mix and dissipate steam ingested into the ventilation system.
Provide justification that any resulting condensation would not adversely 5,
affect safety related equipment. Discuss effects of any resulting contamination.
TVA RESPONSE TVA has performed a qualitative study of the problem and documented that l
review in Quality Information Reicase (QIR) NTBSQN88016 (B45 880121 263).
This issue was previously considered as item A.4 by the TVA Appendix R review team. TVA also determined that this issue was outside the scope of Appendix R, and that it was not a TVA commitment :o consider pool boiling effects following an Appendix R fire.
In addition, equipment located in the spent fuel pool aree that is needed to mitigate a fire has been shown to remain functional for a fire in the room. A fire in the room is considered to be a more severe temperature extreme than open pool boiling.
The study of j
this low probability event shows:
1.
hater boiling ra'.es of approximately 55 gal / min.
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Local building volumes in the refueling area of 980 000 ft,
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Auxiliary building volume in excess of 2,000,000 ft.
4.
With HVAC in operation, steam would not be transmitted into other i
areas of the auxiliary building.
5.
Without HVAC a negligible temperature rise in the refueling area would be expected.
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Therefore, because of the large room volumes' involved in comparison.alth the small volumetric release rate, spent fuel. pool boiling is not consid6ted to be
.i a problem. Surface contamination is only an' economic issue since c
safety-related equipment is generally qualified to nuch higher radiatica
-l 1evels and since operator actions in the area.would proceed pool boiling.
h OUESTION t
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Provide results of the procedure review. conducted to review the
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I' coordination of procedures for fire effects to determine:if there are potentially confusing areas between procedures as discussed during the~
i NRC/TVA staff meeting in Knoxville on January 6,.1988.
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TVA RESPONSE
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As a postrestart item, the TVA Appendix R review team reconssended that the 2
procedures be reviewed for potential conflicts.- This request was captured by l
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~ a memorandum from J.~ B. Hosmer to R.'.J. Johnson and S. J. Smith dated
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February 10, 1988 (B25 880210 101). iThe review / training request was made'to 3.
- the operations procedures staff and the training. staff. Use'of both staffs l
should' maximize the expertise, minimize the impset on operations, and
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reinforce knowledge on special Appendix'R considerations. The review has not yet been conducted.
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j This issue was determined to be beyond the scope of Appendix R.
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P ants, (SQN) minimized the spurious actions that could occur as a result of a l
fire. This will in turn minimise the need to enter and exit multiple procedures, t
i TVA did agree infomally that this type o't procedure coordination review was a useful endeavor. The emergency operatir.g procedures ensure the operator knows i
l how to get to hot-standby independent of the cause.
In events where operator i
actions could be controlled from several procedures, the operators are j
required to assess the plant situation and based on their training and i
experience detemine what actions are of.the highest priority. This was j
confimed from discussion with the SQN operations staff. The procedures have.
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a structured hierarchy, backed by training, that ensures-the operators treat '
i the most important condition first.
l The general approach is to satisfy the status tree function (suberiticality.
l core cooling, heat sink, pressurized thermal shock, containment, and inventory), satisfy the emergency procedures, stabilize the plant, and then to t
proceed to cooldown.
It is important to note that in addition to the fire recovery procedure SOI 26.2, other procedures recognize that a fire may be
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f; the initiating event. An example is abnomal operating instruction (AOI) 27
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on control room inaccessibility. The first symptom mentioned in that.
procedure is a fire.
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4 OUESTION i
4.
Describe how SOI 26.3, Revision 1, provides adequate boron concentration for a cold shutdown condition after a worst case Appendix R fire. If "pressurizer level fluctuation" is used as a depressurizing mechanism provide the procedures controlling this evolution and provide the calculations showing its effectiveness, i.e. depressurization rate vs.
time. Provide the calculation justifying use of one train of RHR to cool 0
the plant to less than 200 F. Provide the Performance curvv for one train of RHR. Discuss availability of an operable pressurizer PORV, pressurizer heaters, auxiliary spray and normal spray for postulated fire scenarios.
TVA RESPONSE Additional boron is not required for at least 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and is aot required 0
until the RCS temperature is 200 F or below.
SOI 26.3 requires boron concentration sampling and feed and bleed to maintain adequate boron concentration for cold shutdown conditions.
Pressurizer level fluctuation is not used as a means for depressurization in this procedure.
Westinghouse has analyzed one train RHR cooldown as reported in Westinghouse letter FSA-II-TV-16446 dated July 21, 1975. TVA has requested an update to this analysis and will provide those final results as soon as available.
Pressurizer heaters, auxiliary spray, and normal spray are not required to
- support safe shutdown. See question No. 5 for a discussion of the availability of a pressurizer PORV.
QUESTION 5.
Provide the clarification of the minor revision to shutdown logic on Key 37 as discussed in A.9, Pneumatic Systees of the Task Group Disposition of Issues. Does TVA take credit for use of any control air system during an Appendix R event? Can the plant be shutdown and controlled including operating in a solid configuration if needed without i
the use of the air system? Please provide justification to support any statements regarding your conclusions.
l TVA RESPONSE 1
The safe shutdown analysis ir documented in revision 8 to calculation SQN-SQS4-127. The minor revision to Key 37 of that document added a note that manual control of the control room HVAC dampers can be taken to maintain control room temperature in the event control air is not available.
i The safe shutdown ar.alysis did not credit the use of air systems to reach safe shutdown. However, if air remains available, it would be used to normally operate valves. Manual actions are taken as a backup to manipulate devices in the event the air system is not available.
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A preliminary separation analysis on unit 2 indicates that eitt?r a head vent or pressurizer PORV letdown path is available for all fire scenarios that may require water solid operation,(i.e., spurious safety injection).
In the unlikely event the plant is placed in a solid water condition and air systems are *)ot available, pressure control can be maintained. The charging pumps are used to provide plant makeup and control pressure during plant cooldown whereas over pressure protection is provided by the primary system safety valves. Plant pressure can be lowered by the use of normal RCS shrinkage during cooldown until entry into RHR cooling is possible.
'OUESTION 6.
Explain why the primary plant will not lose the pressurizer bubble in any fire scenario such that 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> is a conservative value for requiring the availability of RHR. Does TVA take credit for plant operations in a water solid condition to cooldown following a fire and prior to entering cold shutdown? If so, describe the operating procedures and operater training-which is conducted for these water solid plant operationo.
TVA RESPONSE L
There is a remote chance that pressurizer bubble could be lost. However, loss of pressurizer bubble itself does not place the plant in an unsafe condition t
or prevent hot standby from being achieved and maintained.
i To ensure that there war sufficient time to perform the repair to the circuits for the RHR valves (FCV-74-1 and FCV-74-2), the earliest time the plant would i
be able to be put on RHR had to be determined. The 19-hour value represents the earliest timeframe that the RHR system could be placed in operation using reactor coolant system shrinkage as the method for providing boration from the
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In the event of a solid water condition, a much slower cooldown rate would be used to assure proper pressure control was maintained.
t There is no TVA or NRC requirement that the plant must be in cold shutdown or I
on RHR within 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />. The only case where the plant must be in cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is if an alternate shutdown method is used (i.e.,
abandonment of the main control room and transfer to the auxiliary control room). The only other time constraint imposed by Appendix R is that if equipment required for cold shutdown is damaged by a fire, it must be repaired 1
and made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
i Water solid operation has a low probability of occurrence and is not the preferred method of cooldown for SQN and is not emphasized in operator training; nevertheless, this method can be used. See the response to question No. 5 for the method of control in the event of a solid water condition.
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.l OURSTION 7.
Provide justification for repair times for FCV-74-1 and FCV-74-2. State why these valves are considered operable for fires inside containment.
In particular provide valve location information and describe how persons entering containment following a fire could be expected to repair and l
manually position these valves.
If reactor head vents, pressurizer block valves and the pressurizer PORV are spuriously opened, discuss the effect on containment environment and RHR valve acces' ibility.
J TVA RESPONSE rdnce these valves are not needed to achieve or maintain hot standby, the cables would not have to be repaired until the plant was ready to go onto RHR cooling or within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if using alternate shutdown. However, for a s' ire outside containment, a casualty procedure (SMI-0-317-18 section 4) was developed to ensure that damage to electrical cables associated with FCV-74-1 and FCV-74-2 can be repaired within 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
Inside containment thare is no in situ fire load in the area of FCV-74-1 and FCV-74-2 or their associated cables sufficient to damage the valves or their cables. This is documented in items 3, 4, and 5 of a fire hazards analysis QIR SQP-SQN-88-212-RO.
Therefore, the valves would remain operable and containment access is not required.
(See TVA drawings 47W432-3, -4 and 47W200-11. -12, -13 for valve locations.)
If the fire involved the valve itself (i.e., the valve motor burned up), this would not create a fire of sufficient magnitude to cause damage to any other components. As an example, normal letdown should be available.
In addition, a fire in the valve itself would not render the containment inaccessible.
Damage to the FCV 74-1,,two valve motor operators will not prevent manual operation. Therefore, manual opening of these valves is possible once access to containment is available. Operation of these valves is not required to maintain the plant in hot standby and repairs are therefore not time critical operations. With the plant being capable of being maintained in a hot standby condition for an extended period of time, operation of these valves can be postponed until containment access is available to perform valve alignmet t.
Additional information on time requirements is provided in our response to question No. 6.
Spurious opening of the reactor head vents, pressurizer block valves and pressurizer PORV are addressed in the safe shutdown analysis. The only credible fire that effects MOV 74-1,2 is a fire in the cables or MOV itself.
This fire will not cause spurious opening of PORV, head vents, etc.
QUESTION 8.
Discuss the possibility of lubrication oil from the Main Reactor Coolant System Pumps being thrown beyond the oil collection system due to a break in the shaft, shaft collar or lower oil pan assembly.
TVA RESPONSE A fire protection engineer examined the oil collection system on February 26, 1988, and determined no credible failure of the lubricating system would result in the oil leaking down onto the shaft where it could bo thrown beyond the oil collection system.
The oil collection system for the RCP has been reviewed and accepted by the NRC.
Consideration of a broken reactor coolant pump (RCP) shaft has a low probability of occurrence. Any further postulated condition that includes multiple faults is beyond the Appendix R requirements and beyond normal failure analysis assumptions.
QUESTION 9.
Describe the protection and provide a copy of the fire analysis for steam generator PORV controls. Are tho Boron Injection Tank, Boron Injection Pumps, Safety Injection Pumps and tho automatic actuation of charging pumps included in the Appendix R functional critoria?
TVA RESPONSE A modification (ECN 6689 and FCR 5717) to the steam generator power operated relief valves (SG PORVs) associated with steam generators 1 and 4 ensures that a fire local to the solenoids or controllerc will not cause the SC PORVs to spuriously open.
The other two SC PORVs associated with steam r;enerators 2 and 3 are not subject to a significant fire exposuro local to the solenoids or controllers to cause them to spuriously open. This is documented in item 8 of the QIR noted in response to question No. 7.
Safe shutdown analysis
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demonstrates that safe shutdown can be achieved and Appendix R requirements satisfied without the boron injection tank, boron injection pumps, and safety I
injection pumps.
The charging pump automatic actuation is also not necessary to achieve safe shutdown or to comply with Appendix R.
However, the manual operability of at least one charging pump has been assured.
QUESTION 10.
Describe the effects of a Main St<aam Line Break and a resulting steam generator PORY opening spuriously.
Describe the environmental qualification of the PORV including seismic. Is the PORY single failure proof? Discuss whether the Appendix R functional criteria specifically called for no blowdown of any steam generator.
S TVA RESPONSE A break in the main steam line and a simultaneous power operated relief valve (PORV) opening would result in a decreasing DNBR and a cooldown of the primary system. A main steam line break downstream of the flow restrictor plus a spurious steam geneettor (SG) PORV blowdown are bounded from a core response standpoint by a steam line break upstream of the flow restrictor.
This event is similar to the large steamline break described in FSAR section 15.4.2.
Westinghouse has analyzed a two-steam generator blowdown generically as part of the Emergency Response Guideline development. This best estimata analysis demonstrated that the DNBR remains above 1.30 and that the safety injection system capacity is adequate.
The problem addressed in this question is not an' Appendix R issue. Appendix R does not require postulation of coincident fire with other accidents nor does it specify environmental qualification requirements for equipment.
A CAQR has been written that addresses fire, seismic, and MSLB as initiating eventa for failure of more than one SG PORV. The-Appendix R initiating event has been addressed and corrective action is complete (see the response to quention No. 9).
With respect to seismic, MSLB,- and single failure concerns, the CAQR has been classified postrestart based on (1) low probability, (2) best estimate analysis by Westinghouse showing that safe shutdown can be obtained, (3) industry precedence, and (4) main steam line break evaluation described above The closing solenoid for the steam generator PORY is environmentally and seismically qualified.
Independent of Appendix R, the NRC does not-require postulation of concurrent multiple indepetident initiating events. Hawaver, the condition noted has been documented in a CAQR SQT 871198 and is being evaluated to determine if any corrective action is necessary.
The Appendix R functional criteria does not specifically exclude the possibility of a steam generator blowdown. Procedurally, the operators would '
determine from indications whether a PORV had spuriously opened and would take immediate action to close the valve. TVA is currently reevaluating a postulated fire scenario in control cabinets L11-A, -B and will update the NRC on any corrective action required.
QUESTION 11.
Provide assurance that the pressurizer block valve will close against full reactor coolant system pressure.
TVA RESPONSE P
The requirement for valve closure at full reactor coolant system pressure (2500 psi differential) is specified on TVA drawing 47A940 series. Tests conducted and documented in accordance with maintenance instruction (MI) 10.43, have been completed using the MOVATS testing system. Specifically the valve thrust was verified to meet or exceed that required for closure at 2500 paid.
QUESTION 12.
provide an explanation of how Appendix R related cables are provided protection from spurious actuation.
In particulae define the grounding mechanism of these cables. Do cables of a train for various required components share a common ground.
If so is spurious actuation from a wire to wire short between different cables prevented. Were credible faults considered between individual conductors within a given cable?
Cable to cable?
TVA RESPONSE SQN utilizes an ungrounded de power system which powers the majority of _ the control circuits. The ac power system is a grounded system.
The circuits needed to achieve safe-shutdown and those necessary to prevent or mitigate tha effects of spurious operation of equipment which could adversely affect the plant's safe shutdown capability (type II circuits) are defined and analyzed as Appendix R. "Required Circuits." These required circuits Wero for the most part protected by separation or fire barriers. As identified in various interactions where separation did not exist, TVA dispositioned the interaction by analyses of credible faults that would induce the spurious operation.
These interactions were reported as LERs and reviewed as part of the various NRC audits of Appendix R.
Credible faults were considered between conductors within a given cable, but not from cable to cable.
OUESTION 13.
Is there base line data to say whether Revision 6 to DNE Calculation SQN-SQSA-127, "Equipment Required for Safc Shutdown During Design Basis Fire," has been properly implemented? Wa,s a task force formed to establish the base line for Revision 6? Have audits been conducted to verify that breaker / fuse coordination efforts were conducted properly?
TVA RESPONSE Baseline data exists to show proper implementation of the safe shutdown analysis. This data is in the form of ARSK cable separation drawings, I
interactions, Safety Function position Statements, approved deviation requests, fico hazard evaluations, and engineering change notices (ECN).
Each ECN is followed through to completion by QA controlled procedures.
A task force was f ormed to review internally identified Appendix R concerns.
The findings of this task force have been previously forwarded for your review.
Audits performed on Appendix R and coordination calculations include:
Appendix R audit (NRC)
January and June 1985 Electrical calculations (NRC)
June 1987 Sargent & Lundy independent review of electrical calculations 1987 ID1 audit / review July-September 1987 EA technical audit January 1987
i TVA, therefore, believes that breaker / fuse coordination efforts have received i
sufficient attention by audits.
QUESTION l
l 14.
Provide the HVAC calculations or damper closure information which show that rooms outside the fire area can stay within equipment qualification limits during a fire. Provide the HVAC calculations showing when containment access can be accomplished following a fire.
l TVA RESPONSE Calculations which show that rooms outside the fire area stay within equipment qualification limits during a fire do not exist.
Fire dampers are designed to enintain fire barrier integrity and thereby prevent fire damage to adjacent rooms. The temperature rating of the fire damper fusible links are normally 160-165 F.
0 HVAC calculations are not an Appendix R requirement. Appendix R requires either physical (fire barriers) or spatial (20 feet horizontal) separation be l
provided between redundant safe shutdown components to ensure that safe l
shutdown ca.. be achieved.
l Equipment qualification is not required for compliance with Appendix R and has not been evaluated for a fire. However, equipment qualification for design basis events have been addressed under the EQ program.
Safe shutdown analysis demonstrates that we can maintain hot standby without entering containment. Calculations that show when containment access would be available following a fire do not exist and ar.e not required. However, for a fire inside containment, alternate or dedicated shutdown methods (e.g.,
shutdown from the alternate control room) are not used. Therefore, SQN is not required to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
For fires outside containment, containment access is not required. For additional information see the response to question No. 7.
QUESTION 15.
Provide the fire interac61on study for a fire in the immediate vicinity of the pressurizer.
TVA RESPONSE This information is provided by the fire hazards analysis documented in Calculation SQN-00-D052, EPM-EAC-11888 and the QIR referenced in the response to question No. 7.
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4 QUESTION i
16.
Provide a list of guaranteed Appendix R equipment which~is required for
. availability duringLperformance of 3DI [ sic] 26.3.-
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l TVA RESPONSE.
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l Appendix R. equipment for safe shutdown is described-in our shutdown logic.
The equipment required in 50I 26.3.is_found in Key 48-(page A149) of the shutdown logic analysis sqN-sQS4-0127 Rev. 8.
QUESTION I
17.
Provide information'if available regarding testing or analysis of passing
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liquid through a pressurizer code. safety valve and the resultant erosion and subsequent ability of the valve to reseat.
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TVA RESPONSE This is eutside the requirements of Appendix R.
However, testing has been performed by EPRI and is documented in "PWR Safety Valve Test Report "
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.NP-770-CD which was prepared by CE (project Number V102-2).
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OUESTION l
18.
Provide rationale for protection of CCP cavitation from a spurious actuation of the VCT isolation valve.
Provide calculations and i
I procedural references for protecting the VCT when' static pressure head-is l
lost from the RWST during cool-down and RWST inventory reduction, j
Discuss spurious actuation of the VCT Hydrogen blanket-makeup valve.
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TVA RESPONSE Rationale and procedural references are provided in the shutdown logic Keys 4' and 5 and Appendix B..The operator actions identified in Appendix B are also l
provided in 20I 26.2.
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Calculations have not been performed to demonstrate that the VCT is protected j
when static pressure head is lost from the RWST during cooldown and RWST l
j inventory reduction. At any time the charging pump supply from the VCT should be lost, 301 26.2 requires the appropriate operator action.
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should hydrogen makcup valves spuriously operate, the operator actions I
identified is SOIL 26.2 have ensured that the VCT valves will be closed. This i
j isolates the VCT from the charging pump suction.
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OUESTION 19.
Provide the basis for fire protection of Appendix R shutdown systems inside the containment.
TVA RESPONSE The basis for fire protection inside containment is Appendix R Section III.C.2.a-f.
All equipment inside contairAent that is necessary for safe shutdown was evaluated under the required basis and documented in the safety function position statements for inside containment.
QUESTION 20.
Discuss possibility of two low pressure signals causing an actuation-of the safety injection system. Also discuss the protection of circuitry required to assure proper alignment of the safety injection system.
TVA RESFCNSE The possibility of actuation of the safety injection system has been considered.
Safety injection is not required for safe shutdown at SQN.
In the event a spurious actuation of safety injection should occur, the Emergency Operating Procedures (EOP) provide instructions for operators to take corrective actions (Emergency Instruction E-0, step 4) to mitigate adverse effects. This is consistent with scope of 10 CFR 50 Appendix R, "Both trains of equipment necessary for mitigation of consequences following a design bases accident may be damaged by a single exposure fire."
Therefore, protection of the safety injection system circuitry is not required.
QUESTION i
21.
Has TVA evaluated the effect of fire on instrument sense linest Provide the result of the evaluation and its effect on the Functional Analysis Report. Discuss the effect on the safe shutdown analysis due to the fire effects on pressurizer level, steam generator level, and temperature instrumentation.
TVA RESPONSE TVA has evaluated the effect of fire on instrument sense lines. This is documented in calculation SQN-00-D052, EPM-EAC-Oll888 referenced in response to question No. 15.
The calculation demonstrates that adequate instrumentation, including pressurizer level and steam generator level, will i
be available to the operator.
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OUESTION 22.
Explain why the firo in containment would not affect the instrumentation (as discussed in the preliminary Task Group Disposition of Issues in B.2) used by the operator to distinguish between a fire and a LOCA.
TVA RESPONSE Refer to calculation SQN-00-D052, EPM-EAC-ll888 and the QIR referenced in our response to question No. 7.
In addition, although a separation analysis has not been performed for all-other instruments, it is reasonable to assume that some of the indicators will be available because of the redundancy and diversity of instrumentation type and location.
These indicators include:
containment isolation valve status, reactor coolant system (RCS) pressure; pressurizer level, power operated relief valve (PORV) acoustic monitor status. PORV/ safety valve tallpipe temperature, containment temperature, upper compartment radiation monitor Status, lower compartment radiation monitor status; nhield building vent stack monitor status, containmenc purge air exhaust radiation monitor status, ice condenser inlet door status, containment sump level, containment narrow range pressure, containment wide range pressure, plant area radiation monitor status, control room ventilation isolation status, fire water flow alarm, fire detection system alarm, etc. These indicators would be sufficient to diagnose the differance between a fire and a LOCA.
Information Notice 84-09 provided the guidelines for what instrumentation is required to comply with Appendix R.
SQN is in compliance with these requirements, or has approved deviations.
OUESTION 23.
Discuss how steam generator overfill from the main feedwater system is protected against following a fire in the control building.
In particular a4 dress response times for feedwater isolation following loss of the control building.
TVA RESPONSE Keys 22 and 23 identify the equipment necessary for isointion, and we ensuro
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one steam generator level indication is available for each steam generator.
AOI 27 provides that before the Main Control Room is abandoned, the reactor is j
tripped and the MSIVs are closed. This shuts off the steam supply to the main feedwater pump turbines, j
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QUESTION 24.
WCAP 10541 provides justification of RCP seal integrity under a station blackout condition where the RCP's would not be running.
If TVA has not assured the ability to promptly trip the RCP's during a fire, on what basis do you consider this analysis applicable to Sequoyah. WCAP 10541 provides for no less than one hour until failure for non-energized RCP's using high temperature elastomers.
It appears this is not true for all elastomers available. Provide information on elastomers installed in Sequoyah's RCP seals.
TVA RESPONSE If there is a loss of seal flow for any reason, the operator is instructed to shut off the pump.
These general instructions are contained in SOI 68.2, "Reactor Coolant Pumps." A re-review of WCAP 10541 and informal discussions with Westinghouse personnel indicate that the one-hour value applies to both qualified and non-qualified elastomers. Please refer to case 3 page 10-48.
QUESTION 25.
Given that TVA considers the spurious opening of a pressurizer PORV a credible event and relies on the manual closure of the block valvo to limit the consequences of this event, discuss for a control building fire how long will it take for the operator to take this action. What will the RCS conditions be at the time of PORV isolation? Does SI actuation i
occur and is it available? What restoration guidelines will be used?
What specific operator training and procedures have been provided?
TVA RESPONSE It is conservatively estimated that approximately two minutes would be required for one operator to reach the auxiliary control room and to begin taking action.
Cenerally, this action would occur as soon as the spurious opening was diagnosed. For a control building fire that prevented the operator from closing the block valve, the operator must take action from the auxiliary control room. Normally this is accomplished by overlapping operators between the amin and auxiliary control rooms for a short period of time before abandonment of the main control room. Once in the auxiliary control room, AOI 27 revision 9. "Control Room Inaccessibility," instructs the operator to immediately close the PORV, etc., if abnormalities in pressurizer pressure or level are noted (see step F).
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Final RCS conditions would be a function of RCS system parameters at the initiation of the uncontrolled PORV blowdown.
For nominal RCS conditions, the RCS response would be similar to the preliminary portion of the small break PORV LOCA described in WCAP-9600 Volume III.
The primary system pressure would decrease to approximately 1700 psia in two minutes.
This is below the setpoint to initiate safety irdection but above the pressure at which safety injection pump flow would occur.
During this portion of tho translant, only flow from the centrifu6al charging pumps would be possible.
If the safety injection signal was prevented by the fire, this charging flow would not take place unless manually initiated (this would correspond to a loss of approximately 700 gallons of injection out of an RCS volume of 90,000 gallons and would not significantly modify the RCS response).
Safety injection is not prevented or relied on for safe shutdown.
The restoration guidelines and procedures are covered by AOI 27.
This is part of the normal training for the operators for control room inaccessibility.
QUESTION 26.
Are the narrow range RCS pressure sensors included in the Appendix R analysis and has it been verified that they are sufficiently separated such that the 2 out of 4 logic required to actuate SI would not be jeopardized. Considering that the spurious actuation may lead the operator to think a LOCA is in progress, what other instrumentation may also be affected, i.e.,
(pressurizer level sensing, sump level sensors, reactor building temperature sensors, wide range RCS pressure sensors, f
and containment radiation monitors)? What procedures and/or operator training have been developed to aid the operator in distinguishing an actual RCS depressurization from fire induced spurious failures which falsely indicate a LOCA?
TVA RESPONSE This question falls under the Appendix R program and is similar to Task Croup issue B.2 and question No. 22 above.
The narrow range RCS pressure sensors are not required by the Appendix R safe shutdown analysis nor has verification of separation been performed. The root question is whether spurious safety injection can occur. Spurious safety injection is possible and has been considered in the safe shutdown analysis.
Spurious actions could also affect various instrumentation. However, TVA believes that because of the wide diversity of instrumentation available to the operator that the fire would not be misolasnosed as a LOCA and the operator would take appropriate corrective actions as previously discussed.
The response to question No. 22 provides a partial list of the diverso instrumentation available to the operator for distinguishing a LOCA event from a fire event.
It is interesting to note that fire scenarios conducted on the Bellefonte simulator with spurious instrumentation readings were properly diagnosed by the operators, without special training, as fire events. Although similar tests have not been conducted on the Sequoyah simulator, it is anticipated that equivalent results would be achieved.
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