ML091550723

Emergency License Amendment Request Application for Technical Specification 3.8.1 Required Action B.4 Completion Time - (Draft)

ML091550723
Person / Time
Site:	FitzPatrick
Issue date:	06/03/2009
From:	Peter Dietrich Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
vaidya B, NRR/Dorl/lpl1-1, 415-3308
References
JAFP-09-00XX
Download: ML091550723 (27)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A Fdzpatnck NPP P O Box110 Lycoming, NY 13093 Tel 315-342-3840

-- Pete Dnelnch Sore Vice Preside-t June 3,2009 JAFP-09-00XX U.S. Nuclear Regulatory Commiss~on ATTN: Document Control Desk tllashington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, I James A. FitzPatr~c Docket No. 50-333 L~censeNo. DPR-59 Emerqency License Amendment Request Application for Technical Specification 3.8.1 Required Action 8.4 Com~letionTime

REFERENCE:

Techni perating Dear Sir or Mad (5), Entergy Nuclear Operations, Inc, on (NRC) review and approval of a James A. FitzPatrick Nuclear Power e to Technical Specification 3.8.1 est is to add a note allowing a

~ o r n ~ l e t i o n ~ of i m"21 e days", on a one-time basis. This one-time allowance will expire at 1015 on ~ u h e 16, 2009.;- ;.

he 2-Year EDG Preventive Maintenance (PM) a deficiency as identified. Through inspection and testing it has been ht poles on the rotor must be rewound. Review of test data determined that the deficiency does not extend to JAF's other three safety-related EDGs. The proposed change is required to complete rewind of the rotor pole and return the EDG to operable status without requiring a plant shutdown.

JAFP-09-00XX Page 2 of 3 provides a description and evaluation of the proposed TS changes. provides the proposed changes to the current TS on marked up . .pages.

- provides the proposed TS changes in final typed format. orovides a simolified diaarams of the Electrical Distribution Svstem brovides a list bf commithents made as part of this submittal:

June 8, 2009, with the amendment being implemented immediately.

In accordance with 10 CFR 50.91, a copy of this attachments, is being provided to the designated I declare under penalty of perjury t Executed on the -

Pete Dietrich cc: next page

JAFP-09-00XX Page 3 of 3 Regional Admln~strator,Reg~onI U S. Nuclear Regulatory Comm~ssion 475 Allendale Road K~ng of Pruss~a,PA 19406-1415 Res~dentInspector's Off~ce U S. Nuclear Regulatory Comm~ss~on James A. F~tzPatr~ck Nuclear Power Plant P.O. Box t 3 6 Lycom~ng,NY 13093 Mr Bhalchandra Va~dya,Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Llcens~n Office of Nuclear Reactor Regul U S Nuclear Regulatory Commi Mall Stop 0-8-C2A Washington, DC 20555-0001

JAFP-09-00XX Attachment 1 Description and Evaluation Emergency License Amendment Re Technical Specification 3.8.1 Required

JAFP-09-00XX Attachment 1 1.0 Description The proposed amendment would revise the Technical Specrfication (TS) 3.8.1 Required Action B.4 Completion Time, on a one-time basis by adding a footnote to the completion time. The proposed note would read "For the " A EDG subsystem only the Completion Time that the subsystem can be inoperable as specified by Required Action B.4 may be extended beyond the "14 days and 21 days from discovery of failure to meet LCO" up to a total of 21 days as part of the 93EDG-C rotor repair. Upon completion of the repair and restoration, this footnote is no lonqer applicable and will expire at 1015 on June 16, During the performance of the 2-Year EDG with the rotor on 93EDG-C was identified.

rotor has been transported to JAF plant personnel are on-si the repair activity.

has been completed to those EDGs.

Action B.4 unne rations personnel performing a plant During paver operation, the JAF Emergency power buses are normally supplied by Normal Station Service Transformer (NSST) 71T-4 through separate feeder breakers.

(See Figure 710-1in Attachment 4) Should the plant trip for any reason the feeder breakers from 71T-4 are o-tripped and feeder breakers from ~ e s e r v eStation Service Transformers (RSST) 73T-2 and 71T-3 are closed such that each emeraencv bus is supplied by a separGe RSST. The RSSTs are supplied by the JAF 1I<KV kffsite power system. The offsite power system is supplied by two (2) independent lines. One line (Line 4) receives power from the Oswego substation via the Nine Mile Point switchyard and the second line (L~ne3) is supplied directly from the Lighthouse Hill hydro-electric power station, These lines come into the JAF 115 KV switchyard and are connected through motor operated disconnect 10017. This normally closed disconnect allows either line to supply power to both RSSTs such that power would be available to both emergency buses in the event that one offsite power source is lost.

Page 1 of 15

JAFP-09-00XX Attachment 1 Emergency Power:

The JAF emergency power system (Figure 2 in Attachment 4) conslsts of four (4) EDGs each located in a separate room within the EDG building, connected to the " A and " B Emergency busses to supply emergency power during a loss of offsrte power. Each of the two (2) independent and redundant emergency power systems (i.e., divisions) conslsts of an EDG pair connected to emergency switchgear, which contalns the emeraencv bus, aenerator output and tie circuit breakers, and the ECCS load circuit breaKers. he E ~ G are S designed to provide an e source of reliable 41 60 VAC power for safe shutdown equ' to mitigate th~.consequences of a design basis accident in the event o he normal and?offsitepower sources. Each generator has a continuo theref&e,:the total loading capacity available per div EDGs in the divisional pair operating, to the emergency OOkw at 416OVAC and 60Hz. Each EDG also has short 000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, 2,950 KW for 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> and 3,050 KW for 3 The worst case loading (normal and emergency) for the " A emergency bus with a single EDG supplying power is 3179.1 KW, which excludes the second RHR pump that is blocked from starting if one EDG in a divisional pair fails to start. With regard to the blocked RHR pump, operators can manually start the RHR pump as needed within the EDG capacity, as directed by the emergency, abnormal and normal operating procedures. The RHR system would be capable of providing the 100% capacity divisional function that is required for the RHR system to perform the Low Pressure Injection function with a single EDG in the division. In the current configuration with one 93EDG-C out of service, ooerators have transferred normal loads to the " B emeraencv "

ithin the capac~tyof 93EDG-A (refer to This unique configuration (2 EDGs per emergency power train) allows JAF to maintaln emergency AC power to an emergency AC power bus with a single EDG out of servlce.

Each generator has sufficient capacity to supply the required loads necessary to achieve safe reactor shutdown during an operational transient [i.e., loss of offsite power (LOOP) or degraded 4160 VAC emergency bus voltage]. The JAF EDG tram availability is maintained by automatically limiting the initial loading of the slngle EDG while maintaining all emergency loads available.

In conclusion the JAF plant design provides multiple and diverse means of supplying both normal and emergency power to the 4160V buses.

JAFP-09-00XX Attachment 1 Coping Strategies:

Abnormal Operating Procedures (AOPs) address the loss of individual 41 60 VAC buses, the loss of station batteries and in the worst case Station Blackout. These procedures are periodically trained on in licensed operator requalification in the classroom, in the simulator, and through walkdowns. These procedures provide guidance for achieving a safe shutdown condition.

In addition to the AOPs the plant also has a strategy of extending the station blackout coping time, Technical Support Guideline (TSG) TSG-8, ""Extending Station Blackout Time", provides guidance on this strategy. TSG-8 provides direction to start the EDG manually without electrical power available, flashing the field if the EDG does not self-excite, and ensure cooling water supply. RClC o 'on time is extended by providing AC power to a Station Battery Charger usi generator. In addition, instructions are provided to manually opera ith no DC power is available. All necessary equipment is pre-staged. The op t per~odicallytrains on the implementation of the strategies in TSG-8.

The plant will also implement mea wing in accordance with Protected Equipment Program AP

/ Board

~ ~ T T Bus ~ K v 11 Reserve Station Transformer T-3 Page 3 of 15

JAFP-09-00XX Attachment 1 Noun Name Component ID South 115KV Bus / 71-EDSC-10025 I Reserve Station Transformer T-2

! Disconnect Switch

/

I Reserve Station Service 71T-2 1i Transformer T2 Reserve Station Service / 71T-3 1 Transformer T3 I 2.0 Assessment The James A. FitzPatrick Nuclear Power Pla "AC Sources - Operating," requires two qu transmission network and the onsite Class and two emergency diesel generator (EDG and 3.

At 1015 on May 26, 2009 the plant Gnte,ced~ondi CO 3.8.1 to support planned maintenance and inspection activitieson'63EDG- ond night of the maintenance activities during a preve&v.e ma&tenance t megger the EDG rotor a low reading was obtained. The readingindicated It on the rotor.

Subsequent inspection.andtesting has detemiin ight poles on the rotor was faulted. The rotorj%s been~!removed'tindtransported to an approved vendor facility where addit.ic$al inspectidsand testinghas determined that it is necessary to rewind the faultedp4epn therotor. . ...:.:+:.....,.

.Using

, intl$~stry

. standard repair methodologies the

~

rewound and-tested,,.: . ..,,

. . ... ~

G subsystem inoperable, Required Action B.4 states, rable status" and the associated Completion Time is very of failure to meet LCO. TS 3.8.1 Required at, ifjhe required actions and completion times of Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

rn of the EDG generator rotor from the rewind vendor is r will have to be received on-site, inspected and transported ation for re-installation. It is expected that re-installation activities will commence late on Sunday, June 7thonce the rotor is released for re-installation. The re-installation window for the rotor and reassembly of the generator is estimated to be approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Remaining maintenance activities which were part of the original maintenance window but that could not be performed with the generator disassembled are estimated to take an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Post-maintenance testing is estimated to be approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an estimated completion of June gth. Although this time line appears to be within the 14 day Page 4 of 15

JAFP-09-00XX Attachment 1 completion time allowed by L C 0 3.8.1 Condition B it is only an estimate that is based upon currently available information and does not include any allowances for unforeseen circumstances either at the rotor rewind vendor or the site.

Since the removal and re-installation of an EDG rotor is a first-time evolution at FitzPatrick, it is critical that the maintenance is performed in a deliberate manner without perceived time pressure. The pre-job briefings for the reassembly will clearly identify the expectation to stop work in the event that unanticipated circumstances arise or additional time is required to complete the specific task.

Based on the above, an extension of 7 days to the cur AOT is requested.

The 7 day extension will allow ample time to avoi first time evolution and will provide a reasonable unanticipated circumstances that may arise.

The Bases for TS 3.8.1 Condition B states, 'The 14 day Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and low probability of a DBA occurring during this period. While the JAF Probabilistic Risk Assessment (P del has not been through the Regulatory Guide 1.200 peer review process at this model was used to assess the proposed completion time and the delta Core ency (CDF) is 1.25E-061ry.

JAF's request to allow a one-time use of a 21 day completion time allows time to complete the required repairs without maneuvering the plant. During this period additional compensatory measures will be implemented to minimize risk to the plant these measures are described in section 3.1 below.

through the Regulatory Guide 1.200 detail, and qual~tyof the James A.

NPP) PRA are sufficient to support a technically tion of the risk change for this proposed completion time sses internal events at full power only.

The JAF PRA is based on the original JAF PRA that was performed to support the Individual Plant Examination (1991). Since 1991, several updates have been made to incorporate plant design and procedure changes, update plant-specific reliab~lityand unavailability data, improve the fidelity of the model, incorporate BWR Owners' Group (BWROG) peer review comments. and support other applications, such as on-line maintenance, ILRT extension, risk-informed in-service inspection, and Licensing Renewal.

The JAF PRA is maintained through a periodic review and update process, Peer certification of the JAF PRA using the BWROG peer review certification guidelines was Page 5 of 15

JAFP-09-00XX Attachment 1 performed in December 1997. Certification was performed by a team of independent PRA experts from U.S. nuclear utility PRA groups and PRA consulting organizations.

This intensive peer review involved approximately two person-months of engineering effort by the review team and provided a comprehensive assessment of the strengths and limitations of each element of the PRA. On the basis of its evaluation, the certification team determined that, with certain findings and observations addressed, the quality of all elements of the PRA would be of sufficient quality to support risk significant evaluations with defense-in-depth input.

Facts and Observation sheets documented the peer review teams' insights and potential level of significance. All issues and observations from the BWROG Peer Review (i.e., Level A, 0, C, and D observations) have been addressed and incorporated into the PSA model used for the JAF Licensing Renewal project SAMA (Severe Accident Mitigating Alternatives) analysis (JAFNPP PSA Model Revision 2, October 2004). The current PSA model (JAFNPP PSA Model Revision 3, May 2007) was updated to include the plant design and procedural component fallure data of the MSPl systems and used for June 2007 NRC ion.

To meet the requirement of the R e latest updated JAF PSA model (JAFNPP PSA Model Rev' or model changes that were incorporated lnto the JAF P summarlzed as follows:

o Updated the PSA and procedural changes.

o Updated the ini$kiing events'f&quencies,:!F.omponentfailure data by using generic data in NUREG&-6928, "lnc/;isfry . . ..~,.. Aver&& Performance for Components and Initiating Events at&.-,^. ..i;, ..ComrnCir~fal-Nuclear Power Plants", February 2007 and performed:the~ayes~an.update wiit%&plant data. Updated CCF data by using genexickfdta iiiken fro&-:U,S. Nuclear Regulatory Commission, "CCF Parameter

~ ~ f i & ~200F@pdate'i2il-t i ~ ~ ~ ,

h ~.p. : l / n r c o e . i n l . g o,,:.v / r e s ~.. ~..,. .t. ~ / ~ ~ ~ ~ ~ a r a m ~ s t 2 ~ ~ 7 / c c fSeptember p a r a m e s2008.

thtm,

.~..

o Updated the offsite power recovery model based on NUREGICR-6890, "Reevaluation of Station Blackout Risk at Nuclear Power Plants Analysis of Loss of Offsite Power Events: 1986-2004, December 2005, which contains data through 2004. in addition, JAF PSA Update supplemented with EPRl loss of offsite power events data from 2005 to December 2007.

c Revised the core damage definition from the peak clad temperatures greater than or equal to 2200°F to 1800°F defined in American Society of Mechanical Engineers, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications,"

ASME RA-Sb-2005, December 30,2005.

Page 6 of 15

JAFP-09-00XX Attachment 1 o Updated the Human Reliability Analysis methodology from THERP (NUREGICR-1278, "Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications," October 1983) to EPRl HRA Method (EPRI-TR-100259, An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment,").

o Updated the internal flooding frequencies from updated pipe failure data analyses provided in EPRl report TR-1013141, "Pipe Rupture Frequency for Internal Flooding PRAs, Revision I", March 2006.

0 The accident sequence quantification truncation limit lowered from 10. to lo-".

o Enhanced the PSA model to incorporate the in ont Yankee and Pilgrim BWROG Regulatory Guide 1.200 p This updated PSA model has undergone a up peer check and IS currently scheduled for BWROG Regulatory r review durlng September 2009.

Rlsk-tnformed suppo evaluation of PRA ge Frequency (CDF),

Incremental Con Incremental Cond~t~onal reased cornplet~ontime In order to support the change in the TS 3.8.1 requirements, a probab~listicrisk assessment was performed given a 93EDG-C Allowable Outage Time (AOT) of 17 days (min~mum),21 days and 24 days (maximum). The risk assessment involved use of the average maintenance unavailabil~tyPSA model from both the current PSA model (Revision 3) and the updated PSA model (Revision 4).

The risk assessment predicted no unacceptable increase in risk during the period of 93EDG-C inoperability using either the current or updated PSA model.

Page 7 of 15

JAFP-09-00XX Attachment 1 The risk of continued JAF operation with 93EDG-C out-of-service beyond the current 14 days AOT as measured by the delta core damage frequency (CDF), incremental conditional core damage probability (ICCDP), delta large early release frequency (LERF) and incremental conditional large early release probability (ICLERP) for internal events is shown below:

These values are less than the ICCDP and External Event Risk performed as a one-time ces of equipment being modification wou was re-quantified with 93EDG-C out of service.

P for fire events is 4.69E-08 for 21 days. As ommended relocating heat detectors in the cable spreading room to severely limit contribution from transient fires. In lieu of the hardware modification, a change was made to administrative procedures proscribinq -

unattended combustible material in the room. This change in procedure potentiallv reduces the CDF contribution from transient fires in the Cable spreading Room.

JAFP-09-00XX Attachment 1

2. In the IPEEE analysis, spurious actuation or failure due to hot shorts and open circuits within cable jackets was included with a conservatively high probability of occurrence of 1.O. However, in the latest fire PRA methodology for NFPA-805 compliance [NUREGICR-68501, this probability is addressed by assigning a probability of occurrence based on the configuration of the cabling and nature of the short circuit. Open circuits are no longer considered, therefore reducing the impact of the cable damage assessment.

JAF uses thermoset cables which have a high damage temperature.

A conservative estimate considering this new methodology for worst-case failure mode probabilities of hot short circuits for thermoset cables in trays with control power transformer (typical of MCC circuits) results in a probability of failure of 0.05. This change woul the CDF contribution from transient fires in Cable Spreading Ro Reactor Building.

3. In the IPEEE analysis, the dominant scenario in the Control alysis is a generic control room fire with a forced evacuation and failure P ~ ~ I Y shut down the plant by implement~ngabnormal operating procedures. The

~gnitionfrequency used for the IPEEE was 1.07E-02 per year. However, with almost 10 years of additional accumulated industry experience, this frequency has been reduced to 2.5E-03 per year [NUREGiCR-68501. This change would reduce the CDF contribution from fires in Control Room.

Additionally, a backup battery charger lant which could be utilized for scenarios which lead to battery de 10500 or 10600.

Seismic The JAF plant .h te a safe-shutdown earthquake (SSE) with 0.1,.$Q&eaKground seismic analysis performed in the IPEEE-:studyis i check on the design, estimating seismi~capacity

., beyond the The seismic analysis methodology implemented for JAF satisfied the NRC requirements for performing a seismic IPEEE as presented in Generic Letter 88-20, Supplement 4.

Seismic events were evaluated using the Seismic Margins Analysis (SMA) method. The SMA methodology uses a deterministic approach to identify the weakest components in terms of High Confidence Low Probability of Failure (HCLPF) peak ground acceleration.

A seismic margin can be expressed in terms of the earthquake motion level that compromises plant safety--the seismic margin assessment determines whether there is high confidence that the plant can survive a given earthquake. No core damage frequency sequences were quantified as part of the IPEEE seismic risk analysis.

The seismic analysis is dominated by seismic initiating events that lead to station blackout; specifically, seismic-induced station blackout sequences controlled by Page 9 of 15

JAFP-09-00XX Attachment 1 seismic-induced block wall failures, in the EDG Building For the proposed extended L C 0 3.8.1 Required Action B.4 Completion Time seismic-induced failure of the block walls remains the limiting failure. Since the block wall failure is the limiting failure the inoperable status of 93EDG-C during this period would not result in any significant change to the existing core damage contribution from seismic events.

Flood The analyses completed for the internal events PRA and IPEEE did not identify any risk significant contribution related to the 93EDG-C as a result of internal or external flooding.

Confiauration Risk Manaaement Changes to plant configuration due to corrective and preventive maintena ill be controlled in accordance with procedure EN-WM-104, On-Line Risk Assessment. This Entergy fleet procedure complies with the requirement of 10CFR50.65 (a)(4),

Regulatory Guide 1.182, and NUMARC 93-01 require that prlor to performing maintenance activities, risk assessment shall be performed to assess and manage the increase in risk that may result from proposed maintenance activittes.

As discussed previously the scope of the repair is limited to a single pole on the rotor.

This has been confirmed by inspection and testing at an approved vendor facility. The requested one-time allowance of a 21 day cotripletion time for L C 0 3.8.1 Required Action B.4 provides adequate time to complete the rewind activity, reassemble the rotor, test the rotor, transport the rotor to JAF, re-install the rotor, and perform the required post-maintenance testing to restore the EDG to Operable status. Dur~ngthe period of the extended out-of service time the " B train of emergency power wtll remain Operable, both qualified offsite circuits will be available, and the second EDG (93EDG-A) on the

" A train of emergency power wtll remain available. This configuration is discussed in the current destgn basis fsrthe plant and is allowed for limited perlods of time (LC0 3.8.1 Condition B). Operations trains on various scenarios relating to loss of power and off -normal plant conditions. The operations staff is familiar with this configuration and the limitations on an emergency bus with only one EDG available.

To ensure the health and safety of the public, the following risk management actions will be implemented to increase operator awareness of critical equipment to provide reasonable assurance that the assumptions in the risk model are maintained, and to minimize the likelihood of a transient for the duration of the proposed LC0 period.

Page 10 of 15

JAFP-09-00XX Attachment 1

1. The following equipment will be protected in accordance with the plant Protected Equipment Program, AP-12.12, during the time 93EDG-C is out of service:

c Emergency Diesel Generators 93EDG-A, 93EDG-B and 93EDG-D c Emergency Service Water Pumps 46P-2A and 46P-2B o 4160V Normal and Emergency Switchgear Buses 10300, 10400, 10500and 10600 o Station Batteries 71SB-1 and 71SB-2 o Main Transformers 71T-l A, and 71T1 o RHWRHRSW Loops 'A' &

o HPCl pump 23P-1 o RClC pump 13P-o Torus vent valve o Diesel Driven Fir o Diesel Driven Fir to the "B" emergency rt time capacity of 93EDG-sel generator on-site as a back-up g in these areas will be reviewed and any ure occurs that could affect the protected equipment noted ittee will convene a meeting to evaluate plant status and extended L C 0 period is affected.

7. Maintenance and surveillance activities which could lead to Main Turbine trip will be avoided.

Page 11 of 15

JAFP-09-00XX Attachment 1

8. The plant Operations crew and Maintenance staff will be briefed on these risk management measures.

9. As an enhancement to the existing communications protocols daily communications will take place between JAF Operations and the Grid Operator.

10.Just-in-time training will be provided to the operating shifts to heighten their awareness of challenges to the distribution system in this configuration.

11.Operations will monitor weather conditions t tential impacts on plant conditions due to adverse weather co 12.These compensatory measures an operations department stand1 Conclusion Based upon this review, there is n the incremental core damage probability or large early release extended L C 0 period while oper 3.0 Regulatory Anal onsideration determination" for the ment involve a significant increase in the Response: No.

ense amendment introduces a one-time 21 day completion for TS 3.8.1, Required Action 5.4. The urouosed comuletion time does not introduce any new' accident initiators. The probability o i an accident occurring is not affected by the proposed completion time and the consequences of the accidents evaluated in the UFSAR Accident Analysis are not affected by the proposed extension.

Page 12 of 15

JAFP-09-00XX Attachment 1

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

The proposed amendment makes a one-time allowance of a 21 day completion time for TS 3.8.1 Required Action 8.4. The proposed amendment does not introduce any new equipment, create any new failure modes for existing equipment, or create any new limiting single failures. The plant equipment considered when evaluating the existing completion time remains unchanged. The extended completion time will permit complet~onof repair activities without unnecessarily challenging the plant operators to maneuver the plant to perform a shutdown with one train of emergency power out of service.

3. Does the proposed amendment i t reduction margin of safety?

allowance of a 21 day e proposed complet~on I as discussed above. The ts in an ICCDP of 7.19E-08 of 5E-07, and an ICLERP of 3.68E-of 5E-08 Therefore the proposed nt reduction in any margin of safety.

3.2 While JAF was not built orlicensed to the 10 CFR 50 Appendix A, General Design Criteria (GDC), it was evaluated and determined to meet the intent of Appendix A.

GDC-17 requires two independent power sources; the proposed amendment does not alter JAF's compliance with the intent of that criterion. The one-time allowance of a 21 day completion time for TS 3.8.1 Required Action 8.4 does not change the requirement to restore the inoperable EDG to operable status.

In conclusion, based upon the considerations described above:

1. there is reasonable assurance that the health and safety of the public will not be adversely affected by operation in the proposed manner,

2. such activities will be conducted in compliance with the Commission's regulations, and, Page 13 of 15

JAFP-09-00XX Attachment 1

3. the issuance of the amendment will not be detrimental to the common defense and security or to the health and safety of the public.

4.0 Environmental Evaluation In accordance with 10 CFR 51.30 an environmental assessment for proposed actions, other than those for a standard design certification under 10 CFR 52 or a manufacturing license under Part 52, shall identify the proposed action and include:

1. A brief discussion of:

I. The need for the proposed action; ii. Alternatives as required by iii. The environmental impacts and alternatives as appropriate; and

2. A list of agencies and persons con nd identification Need for Proposed Action:

As previously stated in this submi ent to the Technical Specifications will provide sufficie " EDG. By granting a during a plant shutdown with only one emergency diesel gen as required by TS of NEPA for this action.

luents or increase in the amounts of effluents, to the environment.

y effluent, nor does it Radiation Exposure:

There is no increase in individual or cumulative, occupational or public radiation exposure or planned increase in radiation exposure as a result of the planned EDG repairs during the proposed TS amendment extended L C 0 period. The component and activities that relate to the proposed amendment do not affect plant radiation levels, and therefore, do not affect dose rates and occupational exposure.

Page 14 of 15

JAFP-09-00XX Attachment 1 Risk of Radioactive Release:

Although the JAF PRA Model has not been evaluated through the Regulatory Guide 1.200 peer review process at this time, it was used to evaluate the requested one-time allowance of a 21 day completion time for TS 3.8.1 Required Action B.4 from a probabilistic risk standpoint. This assessment considered the expected plant configuration during the period of the extended L C 0 and determined that it does not involve an unacceptable increase in risk. The risk of continued JAF operation with the "C"EDG out of service during the additional 7 day period beyond the technical Specification 14-day Completion Time, as measured by the Incremental Core Damage Probability (ICCDP), is 7.19E-08, and an ICLERP of 3.68E-09 which is below the ICLERP guidance of 5E-08 for internal events, This value is below the ICCDP guidance of 5E-07 and 5E-08 identified in NRC Regulatory Guide 1.177, "An Approach for Plant Specific, Risk Informed Decision making: Technical Specifications", 1998. The ICCDP for seismic, fire and flood external events is bounded by the ICCDP for internal events, and therefore, also meets the quidance threshold. Based upon this review, there is no significant increase in the incremental Core damage probab/lity or large early release probability during the proposed TS amendment extended L C 0 period while operating at power.

Therefore, Entergy ill not involve additional I, or historic resources, threatened or en itat. No environmental resources are on a finding of no impact.

Page 15 of 15

JAFP-09-00XX Attachment 2 Proposed Technical Specification Changes (Mark up)

AC Sources-Operating 3.8.1

@ ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 8.4 Restore EDG subsystem to OPERABLE status.

AND 21 days from discovery of failure to meet LC0 CLi C. Two offsite circuits C.l Declare required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from inoperable. feature(s) inoperable discovery of when the redundant Condition C required feature(s1 concurrent with are inoperable. inoperabil ity of redundant required feature(s)

AND C.2 Restore one offsite 7 days circuit to OPERABLE status.

D. One offsite circuit --.-.----.--NOTE-.-----.--.---

inoperable. Enter applicable Conditions and Required Actions of AND

- LC0 3.8.7, "Distribution Systems-Operating," when One EDG subsystem Condition D is entered with inoperable. no AC power source to any division.

..--.-..-*-...-.--.7.-.---.*-

Restore offsite circuit to OPERABLE I l2 hours I status.

I

JAFP-09-00XX Attachment 3 Proposed Technical Specification Changes (Final Typed)

JAFP-09-00XX Attachment 4 Simplified Electrical Distribution Diagrams

MINE-L$LE ON!? i

- -#47 SYSiE?d & SYSTEM 8 DIVISION 1 om LIG~T~OUSE 0-G -:Am FOR PROJECT 3~

TRAINING AC Distribution Mar. 1996 I 2 z-Entff~ PURPOSES FIG. x SDLP # DWG x

• SV as-e sb 06117102 Rev. 02

~yuclear~Vorthsast OIYLY, 71 0-'i SDLP 710 ~ 3 ~ - o ~ ~ J.

"";" . Paque::e

~ d ~

JAFP-09-00XX Attachment 5 List of Commitments

List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITMENT