ML20082P421
| ML20082P421 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/04/1991 |
| From: | Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9109100322 | |
| Download: ML20082P421 (14) | |
Text
n l'Illi ADEl.I'lllA ELECTRIC COMi%NY NUCLEAR GROUP llEADQUARTERS 955-65 CilESTERDROOK IILVD. 3 WAYNE, PA 19087 5691 l I
(215) 640 6000 NUCLEAR I;NOINI;f: RING A $1 RVICl3 D1I%RTMENT September 4, '991 Docket Nou. 50-352 ;
$0-353 l License Nos. NPF-39 l NPF-85 U.S. Nuclear Requlatory Commission ATTN Document Control Desk Washington, DC x0555
SUBJECT:
Limerick Generating Station, Unita 1 and 2 10CFR50.63, " Loss of All Aternating current" Response to NPC Safety Evaluation Gentlemen Our letter dated April 17, 1989, as supplemented by our letter dated April 9, 1990, provided the required response to 10CFR50.63, " Loss of All Aternating Current," for the Limerick Generating Station (LGS), Unita 1 and 2. A meeting between representatives of the NRC and Philadelphia Electric Company (PEco) was held on November 15, 1990, during which additional clarifying information was discussed. As requested by the NRC, this additional information was provided to the NRC on December 13, 1990.
NRC letter dated June 3, 1991, which was received by PECo on June 6, 1991, transmitted the results of its review of our response to 10CFR50.63 for LGS, Units 1 and 2, as documented in the Safety Evaluation (SE) and supporting Technical Evaluation Report (TER) enclosed with the June 3, 1993 letter. The NRC stated in its June 3, 1991 letter that, contingent upon the satisfactory resolution of the issues discussed in the NRC SE, l the design of LGS Unita 1 and 2 conforms with 10CFR50.63 and the guidance cf Regulatory Guide 1.155, " Station Blackout," dated August 1988, Nuclear Management and Resources Council (NUMARC) report 87-00, " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," and NUMARC 87-00 " Supplemental Questions / Answers and Major Assumptions," issued by NUMAkC letter dated January 4, 1990.
Accordingly, the NRC reque s t.ed that additional supporting analyses and confirmation of proposed actions as discussed in the NRC SE be provided within 90 days of receipt of the June 3, 1991 NRC letter.
9309100322 910904
, P PDR ADOCK 05000352 PDR h [O
U.S. Nuclear Regulatory Commission September 4, 1991
- Document Cf,? trol Desk Pago 2 l .
I attachment 1 provides the requested iiiformation by restating each specific request stated in the NRC SE followed by the responne for LGS, Unita 1 and 2. Procedure revisions identified in our responses provided in Attachment I will be completed within one year of receipt of final NRC approval of our response to 10CFR50.63, in accordance ith 10CFR50.63(c)(3). Also, upon review of the NRC SE and supporting TER, we noted a number of conclusions which we determined should be clarified.
Accordingly, Attachment 2 provides this additional clarifying information.
If you have any questions or require additional information, please contact us.
Very truly yours, f.)
,/j,f' , L,6 t < /'
G. J. Beck, Manager Licensing section Attachments cca T. T. Martin, Administrator, Region I, USNRC w/ attachmente T. J. Kenny, USNRC Senior Resident Inspector, lgs w/ attachmento
I ATTACHMENT 1 !
Lim 0 rick G;ntrcting Station, Units 1 cnd 2 )
R spon 3 to NRC Stction Blackout SCfoty Evoluction
1.0 INTRODUCTION
AND PURPOSE NRC letter dated June 3, 1991 transmitted the results of its review of Philadelphia Electric Company's (PEco's) response to 10CFh50.63 (i.e., the Station Blackout, SBO, rule) for the Li-erick Generating Station (LGS), Units 1 and 2 as documented in the NRC Safety Evaluation (SE) and supporting Technical Evaluation Report (TER). PEco's response tu the SDO rule was provided in our letter ,
dated April 17, 1989, as supplemented by our letter dated April 9, j 1990, and by additional information discussed with the NRC at a meeting held on November 15, 1990, and provided to the NRC on December 13, 1990. As stated in the June 3, 1991 NRC letter, contingent on the satisfactory resolution of the issues discursed in he NRC 1E, the design of LGS, Units 1 and 2, conforms with the SBO :
i rw r .d the guidance of Regulatory Guide (RO) 1.155, " Station Ol d sut," dated August 1988, Nuclear Management and Resources 5 .ncil (NUMARC) report 87-00, " Guidelines and Technical Basea for
!TJMAPC Ir.itiatives Addr/ssing Station Blackout at Light Water Reattoro," and NUMARC 87-00, " Supplemental Questions /Annwers and Hajor Assumptions," issued by NUMARC letter dated January 4,. 1990.
Accordingly, the following provides PEco's response to the NRC SE issues for LGS, Unita 1 and 2.
+
RESPONSE TO NRC SE ISSUES / QUESTIONS
- 1. CST Inventory SE: PEco should discuss what action they propose to ensure that 138,000 gc11ons of water por unit is maintained in each of the two- Condensate Storage Tanks (CSTs) during normal plant
^ operating conditions.
Response We have determined that actions to ensure that 138,000 gallons of water per unit be ;
maintained in each CST are not necessary. Based !
on our calculation, we have concluded that during an SBO event at LGS, the CST inventory will not be required at any timu during the assumed ,
four-hour duration of the event. This L
calculation demonstrates the acceptability of using only the suppression pool inventory for ;
reactor pressure vessel (RPV) makeup and reactor heat removal. !
I l-
- 2. Cont.o1 Room Heatup j NRC SE PECo is requested to justify the use of a nonconservative initial temperature in the '
control room heat-up calculations. In the response, PECo is also requested to describe what operator actions will be performed to maintain j.
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Attachment 1 Page 2 the bulk air temperature and temperatures in equipment cabinets below that which affects equipment operability.
Response We have reviewed the NRC question, and agree .i that a non-conservative initial Ms.in Control Room (MCR) temperature of 75'F was assumed in the.
calculations performed to support our previous submittals. However, based on the following discussion, we have determined that justification of the initial MCR temperature is no longer required.
As discussed in our April 9, 1990 submittal, wie stated that we were considering the performance of additional heatup analyses with the objective of minimizing operator actions. Additionally, during the November 15, 1990 PEco/NRC meeting, l
-the-.NRC.was informed that we were still working on room heatup calculations in order to reduce operator actions during the first hour of an 500 ,
event. The calculations used to support our previous submittals assumed operator actions to open the MCR doors and to remove ceiling tiles in ;
the MCR.
Additional HCR heatup calculations have been ,
performed which envelope the MCR conditions expected during an SB0 event. This calculation ;
determined the MCR heatup temperature profile - ,
during a lose ' of . MCR Heating, Ventilation, and Air. conditioning (HVAC),.under the conditions of a temporary partial loss-of-of f eite-power (LOOP)- '
on Unit 1 and a Loss of Coola'nt Accident (LOCA) on Unit 2. Although this calculation is a #
transient ~ (i.e. , non-NUMARC B7-00) analysis, the initial conditions are consistent with those discussed in NUMARC 87-00. The initial HCR temperature assumed in this calculation is 83'F -t with the adjacent rooms initially at 104*F. No i operator actions were- assumed in this calculation.
The-results of this calculation indicate that at .
time equals one hour (i.e.,-the time at which the Alternate .AC (AAC) powered ventilation is restored), the HCR temperature' reaches approximately 105'F. As such, consistent with .
our - two previous submittals, the MCR does not become a Dominant Area of Concern (DAC) during the assumed four hour SBO event at LGS since the temperature will . be below 120*F. Accordingly, equipment cabinet doors do not need to be opened f
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Attachment 1
- Page 3 in order to maintain equipment operability.
Additionally, based on the results and assumptions of this calculation and the guidance of NUMARC 87-00, we have concluded that no operator actions are required to ensure that the MCR does not become a DAC, and to ensure equipment operability during a four hour SDO event.
- 3. Containment Heatup NRC SE For the containment heatup, PECo is requested to verify that the conditions assumed for the design basic LOCA are representative of those expected during an 5B0 event per NUMARC 87-00 Supplemental Questions and Answers.
Response In our submittal dated April 9, 1990, we stated that the containment heatup during a 500 event is bounded by the design basis LOCA which leads to the most severe containment (i.e., drywell) conditions. This design basis LOCA was determined to be a small primary system rupture above the reactor water level that results in the release of reactor steam to the drywell. For this case, the drywell temperature profile was conservatively assumed to be 340*F for the first three hours and 320*F for the following three hours after the initiation of the event (i.e.,
LGS, Units 1 and 2 Updated Final Safety Analysis, UFSAR, Section 6.2.1.1.3.3.5.4). These temperatures were determined by the maximum steam temperatures possible under the expected reactor system conditions. Our submittal also stated that this event is assumed to occur concurrent with a loss of offsite power (i.e., see UFSAR Section 6.2.1.1) and no drywell cooling.
To respond to the hRC request, we reviewed the conditions assumed for the bounding design basis LOCA against those conditions expected during an SBO event in accordat:ce with NUMARC 87-00,
" Supplemental Questione/ Answers and Major Assumptions." This guidance cautions against making the assumption that LOCA/High Energy Line Break (HELB) containment temperature profiles bound the expected drywell temperature profile during an SBO event. This caution is provided because many LOCA/HELB analyses were performed assuming that drywell fans and coolers are operating; nowcver, during an SBO event, drywell fans and coolers may not be available.
4 Attachment 1
- Page 4 4 l As discussed in UFSAR Section 9.4.5.2, the l safety-related function of the drywell air cooling system is to maintain the drywell atmosphere in a thoroughly mixed condition after
' a LOCA to prevent stratification of oxygen that may be generated as a result of the accident. ,
Cooling water supply to the drywell coolers is not needed nor is cooling water assumed to ba ,
available for this function. As a result, no .
- Credit is taken for drywell coolers in any of the ,
LOCA and HELB analyses, as discussed in UTSAR 1' Sections 3.6.1.2.1 and 6.2. Use of drywell fans ;
to maintain air circulation in the drywell under accident conditions serves no cooling function.
Therefore, we have concluded that the containment heatup during a SB0 event is bounded by drywell ;
conditions during a LOCA, as discussed above. As ,
requested, we.have verified that the conditions i assumed for the design basis LOCA are . not only 1 representative of, but bound those expected >
during an SDO event in accordance with NUMARC guidance.
Additionally, we note that the LGS Environmental Qualification Program incorporates a bounding ;
drywell temperature profile of a constant 340*F i during the first six hours following a LOCA, as shown in NUREG-0588, " Interim Staff Position on Environmental . Qualification of Safety-Related Electrical Equipment," Revision 1. This ,
bounding drywell temperature profile envelopes ,
the bounding profile assumed for the design basis LOCA,- thus ensuring that the required safety-related equipment inside the drywell is -)
not rendered inoperable by the drywell conditions (
during.an SBO event.
- 4. Effects of Loss of Ventilation NRC SE: PEco is requested to describe what operator actions will be performed to maintain the bulk air temperature and temperatures in equipment cabinets below that which affects equipment. !
operability.
Response: The overall effects of loss of ventilation !
during an SBO event at LGS are a concern for the i MCR, Auxiliary Equipment Room (AER), Reactor Core Isolation Cooling (RCIC) pump room, containment, j and the inverter rooms. The effects on the MCR, containment, and the inverter rooms are discussed 1
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. Page L i
l in Responses to issues 2, 3, and 5, respectively.
The effects on the AER and RCIC pump room will be discussed here.
As discussed in the Response to icsue 2, in our efforts to minimize operator actions, an additional calculation which envelopes the AER conditions expected during an SBO event was performed. As with the MCR, although this calculation is a transient (i.e., non-NUMARC 87-00) analysis, the initial conditions are consistent with those discussed in NUMARC 87-00.
The results of this analysis indicate that at time equals one hour (i.e., the time at which AAC powered ventilation is restored), the AER temperature reaches approximately 108'F. The calculation shows that at no time during the assumed four hour SBO event will the AER temperature reach 120*F. We have, therefoce, concluded that the AER is not a DAC.
Accordingly, equipment cabinent doors do not need to be opened in order to maintain equipment operability. In our two previous submittals, we had stated that it was. Based on the results and assumptions of thin calculation and the guidance of NUMARC 87-00, we have concluded that no operator actions are required to ensure that the AER does not become a DAC, and to ensure equipment operability during a four hour SDO event.
For the RCIC pump room, we have performed a non-NUMARC B7-00 calculation which indicates that no operator actions are required to maintain the RCIC pump room bulk air temperature below that which effects equipment operability for the one hour priar to establishing the AAC powared ventilation. This calculation was pe r f orn.ed in order to minimize operator actions and to consider the effects of u failed RCIC pump barometric condenser. Based on the results of this calculation and the guidance of NUMARC 87-00, we have concluded that no operator actions, including the opening of equipment cabinet doors, are required to ensure RCIC system operability during the four hour SDO event.
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Attachment 1
- Page 6
- 5. ~ Inverter Operability NRC SE PECo.is requested to verify that the NUHARC i 87-00, Appendix F, values which they use, bound-the_ manufacturer's recommended operating j temperatures for the inverters in the Unit i 1 Auxiliary Equipment Room and the Computer Room. j Responses As discussed during the November 15, 1990, PEco/NRC meeting, the Unit 1 Average Power Range Monitor (APRH) inverters, which are located in the Unit 1 AER, were-located in a DAC. However, as discussed in-the Response to issue 4, we have determined that.this room is not a DAC. The Unit 1 Reactor Protection System (RPS) and computer inverter room has been determined te be a DAC.
For the inverters in each of thene two rooms, -l reasonable -assurance of operability was determined using the mottodology of NUMARC 87-00, l Appendix F. In reaching this conclusion, . the specified operating conditions for these inverters was incorporated in our calculation.
Each of these inverters has been designed for
-continuous operation at 104'F with a mean time between failures (MTBF) of at least 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. ,
This calculation used the Arrhenius equation,-as .
outlined- in Section F.3.4 of NUMARC '87-00, Appendix F, and incorporated the specified-inverter operating conditions as stated above.
The results of this calculation showed that for the Unit 1 RPS and computer inverter room, at the esiculated room temperature of 131'F, the inverters would remain operable . throughout the four hour SBO event. This calculation incorporated a bounding value for activation energy, in accordance- with NUMARC 87-00, Appendix F.
Although the Arrhenius equation was not required to be used in determining reasonable assurance of operability for the Unit 1 APRM inverters, the conclusions for the RPS and computer . inverters -
also apply to the Unit ' 1 APRM inverters based on the calculated heatup temperature for the AER being less than the calculated heatup temperature for the RPS and computer inverter room.
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' Page 7 j
Dased on the results of this calculation and using the guidance of NUMARC 87-00, Appendix F, we have verified that the inverters in the Unit 1 AER .and the RPS and computer inverter room will remain operable throughout the four hour SB0 event. The values used for activation energy in this calculation were bounding and in accordance with the NUMARC 87-00, Appendix F values.
Additionally, considering the discussion above, basing our conclusions on the SB0 Operability Temperatures and Durations given in Table F-1 of NUMARC 87-00, Appendix F, for these inverters is '
unnecessary. Ilowever, by using the Arrhenius equation and appropriate calculational inputs, and assuming a.. steady state room temperature of i 185'F during the SB0 e*ent, we determined that the HUMARC 87-00, Appendix F, Table F-1 I operability conditions do bound the specified operating conditions for the subject inverters.
- 6. Containment Isolation Valve (CIV) Closure Verification NRC SE: PEco is requested to verify that procedure E-10/20 identifies the action necessary to ;
confirm that containment isolation valves are fully closed, if needed.
Desponses LGS procedure E-10/20 will be revised to provide guidance as described below for the operators regarding CIV closure verification. This procedural revision will be completed within one year of receipt of. final NRC approval, in accordance with 10CFR50.63(c)(3). ,
If containment integrity is required during . an-SDO event, the following methods are available to ,
verify valve closure.
a) If the Motor Operated Valve (MOV) is powered by DC, then position indication- will be available in the MCR and valve closure can be verified.
b) For AC powered MOVs, once the AAC source is available many containment penetrations will.
have at least one isolation valve with power available. Therefore,- position indication ,
will be available in the MCR and valve closure can be verified. ,
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l Attachment 1 Pape 0 c) For AC- powered MOVs, if no power is avaliable to either valve in a penetration, or if there'is a single unpowered NOV in a penetration, then an operator can close the valve locally by using the handwheel. For each containment penetration covered by the valve list in our April 9, 1990 submittal, there'is at least one valve that would be 1 accessible outside of containment. ;
i In our April 9,-1990 submittal, we included a list of the LCS CIVs that fail as-is . and therefore do not moet the NUMARC 87-00 exclusion 4 criteria. Should conditions during an SBO event result in the need for an operator to close a CIV locally by use of the handwheel, although not required by the SB0 rule or associated guidance, l provisions should be included to ensure proper b lighting in - the valve area.- As such, we have reviewed the locations of the existing LGS ,
emergency lighting -areas.- For .many _ of .the !
valves listed in our spril 9, _ 1990 ' submittal,
_j emergency lighting will be . available if it is required .to manually verify ' valve closure.
However, because emergency lighting is not j available for all of the valves listed, the E-10/20 procedural revision will include provisions regarding the use of flashlights if it >
necessary to verify valve closure locally. >
- 7. Quality Assurance and Technical Specifications NRC SE: The licensee has not provided any information on how the plant conform; with the guidance of RG l 1.155, Appendices A and B.
Responset With the exception discussed below, the equipment i that. is assumed operational to achieve and- -!
maintain safe chutdown of both units during an- .; '
SBo event is safety-related and is covered by
-PEco'.s Quality Assurance (QA) program as required by Appendix B to 10CFR50. Equipment that is l relied upon during an SBO event, but.that is not l safety related, will be maintained in accordance .j with -- the - ' guidance of RG - 1.155, Regulatory Position C.3.5 and Appendix A. l The only non-safety related equipment relied.upon l during an SBO event at LGS is the 101 and 201 l safeguard buses, which will be used to backfeed l AC pover to the blacked out unit.
-l I
L
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Attachment 1
. Page 9
- 8. Emergency Diesel Generator (EDG) Reliability Program NRC SE: The licensee submittal did not specifically address a commitment to implement an EDG-reliability program in accordance with the 511 dance of RG 1.155, Regulatory Position c.1.2.
Responses A target EDG reliability of 0.95 was selected based on. Laving a nuclear unit average EDG-reliability for the last 100 demands greater than 0.95, consistent with NUMARC 87-00, Section 3.2.4.
An EDG reliability program will be implemented to monitor and maintain the EDG target reliability of 0.95 utilizing the guidance in RG 1.155, Regulatory Position C.I.2. If the EDO performance falls below the target reliability level of 0.95, action will be taken as required by the EDG reliability program to restore the affected EDG to the target reliability level.
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ATTACHMENT 2 Limerick Gansrating Station, Units 1 end 2
+ NRC Station Bicekout Sofoty Evoluction Additional Clarifying Information NRC letter dated June 3, 1991, transmitted the results of its review of Philadelphia Electric Company's (PEco's) response to 10CFRLO.63 (i.e.,
the Station Blackout, SBO, rule) for the Limerick Generating Station (LGS), Units 1 and 2 as documented in the NRC Safety Evaluation (SE) and supporting Technical Evaluation Report (TER). Upon review of the NRC SE and supporting TER, we noted a number of conclusions which we determined should be clarified. Accordingly, this attachment provides the identified SE and/or TER conclusions followed by clarifying information.
CLARIFYING INFORMATION A. Alternate AC Power Source The following three clarifications should be made regarding the Alternate AC (AAC) power source.
- 1. The NRC SE stated that instead of using one of the Emergency Diesel Generators (EDGs) on the blacked-out (BO) unit, the licensee decided to use one of the EDGs on the opposite non-blacked out (NBO) unit as the AAC source.
For the NBO unit, considering the single tailure criterion, three EDGs are assumed to be available. The SE statement j implies that one of these EDGs from the NBO und w l.l l be dedicated as an AAC source for the BO unit. Actually, we are crediting the excess capacity from the NBO unit as the AAC power source for the BO unit. We do not rely on any single EDG from the NBO unit as the AAC source for the BO unit.
The LGS SBO AAC configuration assumes no operable EDGs on the BO unit, and three operable EDGs on the NBO unit. The ausumption that no EDGs are operable on the BO unit is an assumption mhde for the purposes of analysis only.
- 3. The NRC SE stated that the NRC's assessment of PECo's proposed AAC power source indicates that it falls into the minimally capable AAC power source category.
Based on a review of our proposed AAC configuration, as discussed above and at the November 15, 1990 PECo/NRC meeting regarding the LGS SBO submittal, we have concluded that LGS should be categorized as a fully capable AAC power source design. As concluded in the NRC SE, the NRC's staff evaluation of the AAC power source indicates that there is adequate capacity and capability to power the essential loads in the BO unit without requiring any load shedding in the NBO unit. As defined in the NRC SE, although redundant capability is not available, a fully capable AAC source
9 Attachment 2 Page 2 would enable attainment of safe shutdown during an SBO event and recovery from the Main Control Room (HCR), as is the case with LGS. We have, therefore, concluded that LGS should be categorized as a fully capable AAC power source design.
B. Reactor Core Isolation Cooling (RCIC) Pump Suction Transfer The NRC SE and supporting TER states that the licensee will revise the SBO procedure (E-1) so that the operators will no longer be directed to shift the RCIC pump suction to the suppression pool during an SBO event. This statement is no longer true, and this procedural revision will not be incorporated.
As discussed in our response to SE issue 2, the Condensate Storage Tank (CST) inventory is not required at any time during the assumed four-hour duration of the SBO event. No credit for CST inventory has been taken in our SBO analyses, and as such, the CST inventory is not required during an SBO event.
C. Compressed Air The TER states that an instrument air compressor will operate when the AAC power source is established. Therefore, the TER states that sufficient compressed air will be available for bp needed air operated valves during an 500 event. Q To clarify, in our submittal dated April 9, 1990, we stated that the AAC power supply ir capable of energizing an instrument air compressor and an instrument gas compressor within one hour of an SBO event. The only air-operated valves relied upon during an SBO event are the Automatic Depressurization System (ADS) valves. Gas bottles with Seismic Category I supports are provided 'or long-term operation of the ADS valves. These gas bottles supply the ADS valves with an air supply adequate for seven days of operation (i.e., see LGS Updated Saf ety Analysis Report, UFSAR, Section 9.3.1.3.2).
Therefore, although the carability for energizing an instrument air and instrument gas cc.opressor within one hour of an SBO event exists, sufficient local air supply is available such that we do not need to rely upon the availability of an air or gas compressor during an SBO event.
D. Load Shedding The NRC SE and TER concluded that the AAC power source has adequate capacity and capability to power the essential loads in the BO unit without requiring any load shedding in the NBO unit.
' Attachment 2 3
+ Page 3- ;
I To clarify, we note that there is a number of non-essential loads in the NBO unit that could be shed at the operator's discretion; however, load shedding is not necessary, in order to rouintain EDO loading below the continuous rating (i.e., 2850 Kw) and power essential loads in the BO Unit.
E. Electrical Cross-Tie The NRC SE states, " . . . .In addition, there are croseties between the Unit 1 safeguard buses and the Unit 2 safeguard- l buses. . . ." This statement implies that there are electrical i crossties between units in addition to ths 101 and 201. safeguard ]
buses. This is not the case. All electrical connectibility I between units at LGS is asailable via the 101 or 201 safeguard i buses only.
i F. Class 1E Battery Capacity The NRC SE states that the class 1E batteries have suf ficient - ;
capacity to supply the Sao loads for four hours, and that after l
.one hour selected battery' chargers would be powered. The NRC SE also states that the AAC power source will be available to l support the required battery chargers after one hour. ;
4 Theae conclusions are not entirely accurate. As discussed in our submittal dated April 9, 1990, the batteries associated with ;
the DC system for each unit's electrical divisions are desAgMd I
-to have rufficient energy to supply power for four hcura. Each ;
Class 1E battery bank has sufficient capacity without it s l charger to independently supply the required loads for design basis accidents for four hours. As stated in our April 9, 1990 ,
submittal, since the SBo loads _are.a subset of the design basis }
accident loads, the station battery capacity is sufficient to l meet SBO shutdown: requirements for four hours. Four-hour battery capacity is necessary at LCS since some electrical l divisions are not . expected to have their battery chargers -j powered during an SBO event. .l
'I Baced on this discussion, we are providing the clarification j that battery chargers._are not required at any time during the f assumed four' hour SBO event; however, at one hour into the SBO f event, with the AAC power source available, the ability to power I selected battery chargers will be available, f
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