NL-12-1192, Units 1 and 2, Technical Specification Revision Request for TS 3.7.14, Engineered Safety Features (ESF) Room Cooler and Safety-Related Chiller System

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Units 1 and 2, Technical Specification Revision Request for TS 3.7.14, Engineered Safety Features (ESF) Room Cooler and Safety-Related Chiller System
ML12271A229
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 09/26/2012
From: Ajluni M
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-12-1192
Download: ML12271A229 (49)
v • d • e


Text

Mark J. Ajluni. P.E. Southern Nuclaar Nuclear Licensing Oirector Operating Company, Inc.

40 Invemess Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 September 26,2012 SOUTHERN'\'

COMPANY Docket No.: 50-424 NL-12-1192 50-425 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 and 2 Technical Specification Revision Request for TS 3.7.14 Engineered Safety Features (ESF)

Room Cooler and Safety-Related Chiller System Ladies and Gentlemen:

By letter dated December 19, 2011 (Agency Document Access and Management System (ADAMS) Accession No. ML113550489), Southern Nuclear Operating Company (SNC) submitted a request to revise the Technical Specifications (TS). A request for additional information was received from the Nuclear Regulatory Commission (NRC) by letter dated May 21, 2012 (ML12129A297). Subsequent discussions with the NRC staff have prompted SNC to revise the request. Therefore, in accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," SNC proposes to revise the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 TS, Appendix A to Operating License Nos. NPF-68 and NPF-81 . The proposed TS change would revise TS 3.7,14, "Engineered Safety Features (ESF) Room Cooler and Safety-Related Chiller System" such that, with one ESF room cooler and safety-related chiller train inoperable, the allowed Completion Time for Condition A is extended from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. In addition, this proposed TS change would allow 14 days for overhaul maintenance of the safety-related chiller system to be performed. Also proposed is an editorial change to delete a note added as a one-time emergency change to TS 3.7.14 per SNC letter NL-10-1609, dated August 18, 2010 (ML102300574), which is no longer needed. Appropriate TS Bases changes would be made to reflect these TS changes.

The need for this TS change is that the current LCO of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may not provide sufficient time for unplanned maintenance, and does not provide sufficient time for planned maintenance. This has resulted in additional out-of-service time and delays to perform needed vendor recommended maintenance of both Unit's chillers, affecting overall reliability. In addition, on several occasions SNC has either had to seek or has been prepared to seek emergent regulatory relief due to the insufficient 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO time to complete needed maintenance. The 7 -day completion time was chosen to allow sufficient time to prepare and perform chiller maintenance; however, when tear down of the chiller is necessary, up to 14 days may be required. The two LCO times complement each other, and ensure that the 14-day time clock will only be used when necessary, such as to perform chiller overhaul maintenance.

U. S. Nuclear Regulatory Commission NL-12-1192 Page 2 A discussion of the proposed TS change, the basis for the change, and Significant Hazards Considerations are provided in Enclosure 1. Enclosure 2 provides the marked-up TS and TS Bases pages. Enclosure 3 provides the clean typed TS page. has a commitment table. Enclosure 5 supplements Enclosure 1 by providing a discussion of probabilistic risk assessment capability for VEGP. has a list of ESF room coolers served by the safety-related chiller system.

SNC has evaluated the proposed TS change and has determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92.

SNC requests that the proposed TS change be reviewed and approved on or before December 21, 2012. Following approval, overhaul maintenance on the Unit 2 Train B ESF chiller will be completed prior to the refueling outage scheduled to begin on March 10, 2013. The proposed change will be implemented within 60 days of issuance of the amendment This letter contains NRC commitments (reference Enclosure 4). If you have any questions, please contact Ken McElroy at (205) 992-7369.

Mr. M. J. Ajluni states he is Nuclear Licensing Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted,

~ 9 6j/--

M. J. Ajluni Nuclear Licensing Director Sworn to and su~scri effefore me this ~ day of t+/-' o

  • fl v4Lp lvnut./--e.,'l ,2012.

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Notary Public My commission expires: / ( - Z - Z( I ')

MJA I RMJllac

Enclosures:

1. Basis for Proposed Change
2. Marked-Up Technical Specifications and Bases Pages
3. Clean-Typed Technical Specifications Page
4. Commitment Table
5. Discussion of Probabilistic Risk Analysis Capability
6. ESF Room Coolers Served by Safety-Related Chiller System

U. S. Nuclear Regulatory NL-12-1192 Page 3 cc:

Mr. S. E. Kuczynski, Mr. D. G. Bast, Nuclear Officer Mr. T. Tynan, Vice - Vogtle Mr. B. L Ivey, Regulatory Affairs Mr. B. J. Adams, Vice - Fleet Operations RType: CVC7000 Mr. V. M. McCree, I-lor',,'\n<:>I Mr. R. E. Martin, NRR Manager - Vogtle Mr. L. M. Cain, Senior Inspector - Vogtle State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

Vogtle Electric Plant Units 1 and 2 Technical Specification Revision Request for TS 3.7.14 Engineered Safety (ESF)

Room Cooler and Safety-Related Chiller System 1

Basis for Proposed Change to NL-1 for Proposed (:nl~nt"IO Table of Contents

1. Description Description
3. Evaluation
4. Evaluation 4.1 Significant Consideration Applicable Regulatory Requirements/Criteria Conclusion 6.

-1

Enclosure 1 to Nl-12-1192 Basis for Proposed Change

1. Summary Description This licensing amendment request is to Vogtle Electric Generating Plant P) Units 1 and 2 Operating License Nos. NPF-68 and NPF-81.

The proposed change the Technical Specifications (TS) would revise 3.7.14, "Engineered Features Room Cooler and Safety-Related Chiller System" such that, with one room and safety-related chiller train inoperable, the allowed Completion Time (CT) for Condition A is extended 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. 7-day completion will allow sufficient time for routine maintenance and electrical controls component replacements. In addition, this proposed change would allow 14 days for overhaul maintenance of related chiller system to be performed. Also proposed is an editorial to delete a note as an emergency to 3.7.14 per Southern Nuclear Operating Company (SNC) letter NL-10-1609 dated 18,2010 (Agency Document and Management System (ADAMS) Accession No. ML102300574). That note was for a one-time use is no longer needed.

2. Detailed Description

Background

This proposed change is to extend the allowed 3.7.14 Condition A completion time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 7 days. 7-day completion time is for routine maintenance of safety-related Essential Chilled Water System (ECWS) and electrical controls component replacements.

Allowing a 7-day completion time for maintenance will allow for more thorough troubleshooting techniques and for resolution to and perform maintenance. When a tear-down the chiller is 7 days may not be enough to implement all required maintenance follow-up functional testing activities. such, this proposed change also to allow up to 14 days for overhaul maintenance of the Allowing up to 14 days for overhaul maintenance will permit activity to performed during power operation, and will allow more focused management attention for planning, oversight, maintenance implementation, and return to service. essential chillers are 300-ton units, type 13. Due to heavy structure, special rigging and lifting preparations are required for component removal and re-In addition, the room size constrains some to be performed in series.

The ESF room coolers included in Vogtle Probabilistic Risk Assessment (PRA) model due to their negligible impact on reliability of PRA-credited functions.

Incremental Conditional Core Damage Probability and Incremental Conditional Large Release Probability are to negligible for the model) when room and are in a degraded one-train mode of operation such as that proposed by Currently, to ensure unit reliability: of these chillers is run a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month to support control room emergency filtration (CREFS) operability quarterly ClI,A".JCI\I logic are performed to start chiller; maintenance is performed on controls and breakers for chillers on 18 and 54-month frequencies; inspections and motor megger testing are performed on a 54-month frequency; monthly engineering walk downs are performed; and oil sampling is done every 54 months to check for wear metals and oil degradation.

to NL-12-11 Basis for Proposed Change essential chiller overhaul maintenance, currently performed on a i80-month frequency, includes replacement and mechanical inspection of internal components. second round of this overhaul maintenance has been completed for essential chillers 1A, 2A, 1B. Completed overhaul maintenance activities enhance reliability of operating trains during the proposed on-line maintenance of an ECSW train.

Overhaul maintenance has performed at least one time for all of essential chillers. The first round of overhaul maintenance was completed in late 1990s. The second round of overhaul maintenance is currently in progress with three of the four essential chillers completed, including the overhaul maintenance performed on the 2A chiller during the repair of a tube leak in August, 2010.

The essential chillers have in service for approximately 25 years. During that time, a total of five tubes condenser tubes) have plugged in all of the chillers served by I\ISCW.

Specifically, two in the condenser have been plugged in chiller, two tubes in the condenser have plugged in 2A chiller, one tube in the condenser has been plugged in the 1B chiller. No tubes are plugged in the condenser in 1A chiller. Of five tubes, only one tube had failed four were plugged due to thinning of the tube wall. chiller condenser 479 tubes, the plugging limit is 10% of tubes (that is, a plugging limit of 47 tubes). current testing has demonstrated the Train 1A, and 1B evaporator and tubes are in good condition. Thinning the essential chiller condenser and evaporator tubes have historically been within acceptable parameters and have not impacted operation, with exception of the 2A chiller leak in 2010.

This proposed change to the is similar to the previous 2A chiller emergency TS revision request for the 2A chiller to inoperable 14 days to water into the replace chiller hermetic compressor motor (SNC NL-10-1 (ML102300574) NL-10-1 (ML10231 01 dated August 18, 2010), for which SI\IC received Nuclear Regulatory Commission (NRC) approval on August 1 2010. While the operational conditions associated with tl-lis request may different for performing overhaul maintenance (unplanned shutdown verses a refueling outage), request is similar to the August 0 amendment. The TS 3.7.14 completion are based on generic assumptions. SNC is addressing a non-risk significant activity, and providing justification an completion time for chiller overhaul maintenance for an additional completion time for routine maintenance. As stated above, Incremental Conditional Core Damage Probability Incremental Conditional Early Probability are to be negligible in PRA modeling when the room coolers and ECWS are in a one-train mode of operation.

the completion time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days for maintenance will allow a more thorough troubleshooting of any operational issue and effective Allowing up to 14 days for chiller overhaul maintenance (such as when tear-down of the chiller is necessary) will allow this activity to performed online, instead of during a planned or unplanned outage.

While it is possible to perform chiller overhaul maintenance during a planned refueling outage, this is not desirable. Performing chiller overhaul activity with the unit online instead of during a scheduled outage will allow for management to be more focused on work activities associated with the chillers. There instances of NRC approval of modification/maintenance being performed online in lieu of during outage periods. example, on November 19, 2003, the NRC approved a from Byron Station Units 1 & 2 and Braidwood Station Units 1 & 2 to Technical Specification to revise a completion time from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days for

Enclosure 1 to Nl-12-11 Basis for Proposed Change one inoperable instrument bus inverter (ML032830455). SER the change "provides operational flexibility in the scheduling and performance of online maintenance of an instrument inverter."

Also proposed is an editorial to delete a note added as an emergency change to TS 3.7.14 per SNC NL-10-1 dated August 18. 2010. That note was for a one-time use is no longer needed.

3. Technical Evaluation Proposed Change Revise the LCO 3.7.14 Condition A from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 In addition, add an OR statement to allow a change to the Condition A extend it 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 14 days for chiller overhaul maintenance. The chiller overhaul maintenance activity is to be performed at 10-15 intervals. Also is an editorial change delete a added as an emergency change to TS 3.7.14 per SNC NL-10-1609 dated August 18, 2010. note was for a one-time use and is no longer

System Description

The ECWS is comprised of two separate, redundant, 100% independent trains. The room coolers and ECWS provide chilled water to equipment room coolers during abnormal, accident, post conditions. The normal chilled water in certain room coolers provide cooling certain rooms during normal operations. The ECWS supplies chilled water to the cooling coils in ESF room coolers and the CREFS.

ESF room coolers are to maintain the ambient air temperature below environmental qualification rating of the equipment served by the system. equipment room is cooled by a room cooler and associated chiller that are powered from the same ESF train as that associated with the equipment in the room. Thus, a power failure or other single failure one cooling system train will not prevent the cooling redundant equipment in the other train. a list of room coolers and areas by the safety-related chiller see 6.

In addition to a manual capability, automatic cooling of ESF equipment rooms is initiated by three possible signals. room will start upon receipt of a high temperature signal from the associated room. Certain room coolers will start upon receipt of an equipment-running signal or a safety injection (SI) signal. The equipment-running signal is used provide supplemental cooling for the normal ventilation system in some ESF equipment rooms. The high room temperature signal supplements the normal cooling system function and not constitute a credited function. The SI signal or equipment-running signal is the credited function automatic start will start only those room coolers which are required to operate during an SI. ECWS an automatic start from Control Room Isolation (CRI) signal to provide chilled to the The containment spray pump room coolers start when the containment spray pumps Containment spray is actuated when containment pressure reaches the Hi-3 setpoint, which may occur following a loss-of-coolant-accident or a steam line break.

-4 to Nl-12-1192 for Proposed Change The fuel pool exchanger and pump rooms are cooled by train oriented room coolers with normal chilled water (NCW) and train-oriented essential chilled water (ECW) cooling for each spent fuel pool heat exchanger pump room cooling train. During normal plant operation, NCW cooling coil in the room coolers provides the necessary cooling for both the A and Train B spent fuel pool heat exchanger and pump rooms. In the of loss of NCW cooling source while one train is inoperable, the second train ECW coil can provide the necessary cooling function for the corresponding train spent fuel pool heat exchanger and pump room.

ESF room coolers and ECWS are category 1, are provided with 1E and are designed to remain operational during after a shutdown earthquake.

All of the loads served by the inoperable ECWS train will be considered inoperable as required by 3.0.6. However, the Conditions and Required Actions supported systems (that are not required to be as allowed by LCO 3.0.6. The unit's operable ECWS is a separate independent train that provides required cooling to the equivalent loads in operable and all of equipment by the ECWS will be the required cooling. Disregarding other unrelated failures, these train loads will be operable and capable of performing their intended function.

If 1 overhaul maintenance is necessary, the compensatory actions to put in enhance reduction of risks to work an room cooler and safety-related chiller train (that is, work which could potentially have an adverse impact on the normal chilled water system, the other train, or that could affect the availability of the offsite power are:

  • screening and limiting such work to only that deemed necessary to ensure regulatory compliance or to  :>UL'UUI safe continued plant
  • the opening of and the installation of fans, as necessary, in selected rooms to mitigate the potential rise in room temperature, in order to facilitate the availability of the associated housed equipment, in the event Unit's chilled systems are lost See Enclosure 4 for a complete list of commitments. compensatory measures are not credited with maintaining the equipment in affected room(s) operable but are to be prudent actions.

If the ECWS train is out-of-service for overhaul maintenance performance, the compensatory actions listed in Enclosure 4 will continue to demonstrate compliance with single failure criteria and the requirements of General Criterion -19

1. proposed action on the p<:l:<::pnl'I:::I chiller only affects the chilled water system. It does not affect airflow into and out of main control room. Hence, proposed does not have any impact on the unfiltered in-leakage to control room.
2. airflow to and from the control room will not be and hence contributing factors which affects the dose in the main control will not be impacted.

-5 to NL-12-1192 Basis for Proposed Change

3. The main control room HVAC system has two Unit 1 and two Unit 2 control room filtration/cooling units served by their respective train essential chilled water system. One of the four control room filtration/cooling unit is sufficient for maintaining control room habitable conditions. Additionally, the control room has two non-safety related normal air conditioning units which are served by the normal chilled water system. Thus, if one train essential chilled water system is inoperable due to the unavailability of the chiller, the remaining trains (along with the normal air conditioning units) will be available to provide the necessary COOling requirements for maintaining the main control room habitable.
4. The redundant filtration/cooling units served by their respective train essential chilled water system and powered from the diesel generator will be available to provide the necessary cooling to the main control room to maintain it habitable during loss of offsite power.

GDC- 5 states: "Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining unit." Vogtle FSAR Chapter 3 provides how GDC- 5 is met for the control room by stating:

The VEGP is a two-unit plant with the following common safety-related structures:

A. Control building.

B. Auxiliary building.

C. Fuel handling building.

Within these buildings are shared spaces, such as the control room, which contain physically separated safety-related equipment. A detailed description of plant structures is provided in section 3.8.

Safety-related systems are not shared, with the exception of the fuel handling building post accident exhaust system. (See subsection 9.4.2.) Common heating, ventilation, and air conditioning system (HVAC) ducting headers are used in some instances for redundant HVAC units. Systems or portions of systems and spaces that are shared by units 1 and 2 are listed in paragraphs 1.2.2.1 and 1.2.2.2. Where common structures, systems, and components are utilized, such sharing has been evaluated to ensure that there are no adverse impacts on safety functions."

Need for Technical Specification Change Completion Time Extension to Allow 7 Days for Routine Maintenance Past VEGP history has shown that a 72-hour completion time for maintenance may not always be sufficient. Due to the limited completion time, scheduling of preventive maintenance such as instrumentation and control system relay calibrations are a challenge and require multiple entries into the LCO to complete. In addition, if additional items requiring further corrective maintenance are discovered, the limited completion time forces the facility to address only immediate operability issues. As a result, maintenance activities during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> out-of service time are typically symptom-based and the time needed to address longer term reliability issues (such as air in-leakage) is not addressed. In addition, on several occasions SNC has E1-6 to NL-12-1192 Basis for Proposed Change either to seek or has been prepared to seek emergent regulatory relief to the insufficient hour LCO time to complete needed maintenance.

the is risk-significant as in PRA. completion for sufficient troubleshooting and to prepare and perform maintenance.

facilitate scheduling preventive and maintenance allow for longer term action, reduce number of LCO lower the overall out-of-service time for chiller. In addition, it will afford additional time for the availability of parts.

The chilled water system is robust, and the Normal Water System, a common system both units, should available if an taken offline. and rnolnr'\I Operating the loss of electrical provide mitigate temperature in the as propping open doors to enhance circulation. A procedurally controlled contingency will be in place for propping doors open for dissipation of heat from room if the remaining train's ESF room and safety related chiller and the normal chilled water system are out-of-service.

Completion Time Extension to Allow 14 Days for Overhaul Maintenance The of the essential chiller overhaul maintenance that are the focus of this proposed TS change are maintenance the chiller water (cleaning and eddy current testing of tubes) and of the refrigerant (disassembly. I inspection, and of the chillers). overhaul substantially more time than currently allowed by 3.7.14. Therefore, work is typically performed during refueling outages. With an allowance for contingencies, these activities may require up to 14 This duration would allow time such the maintenance could performed with time allowed for recovery in the event that there are complications or nT,-.,r"",,,,c,n problems.

Maintenance of an essential chiller train will be on an expedited but to the the 7-day TS 4 Condition A LCO will not allow sufficient time to complete all activities for overhaul maintenance. overhaul work is scheduled to be performed wh online, necessitating request for a TS amendment.

A number of activities are performed during the refrigerant overhaul maintenance. The rotlri,.,tlr!:llnt is sampled and for any moisture and particles. After the chiller, a visual inspection is performed on components that include volutes, shaft assemblies, impeller vane assemblies, motor included is the reassembly, chiller will be for any refrigerant If leaks are , the associated boundaries are and the leak check is performed again. a vacuum is drawn on the machine to remove moisture and also to demonstrate that there is no Then, the machine is charged with refrigerant. The chiller is functionally tested and returned to service.

sequential steps (not inclusive) of the chiller maintenance their approximate durations are:

  • "'Tn..,m clearance and 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> E1-7 to Nl-12-1192 Basis for Proposed Change

.. Chiller preparation and installation of ground breakers - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

.. Remove chiller end bells and tubes - 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />

.. Rigging I remove crossover piping 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

.. Disassembly and inspection

.. Reassembly - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

.. Leak check with N2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

.. Draw vacuum and hold (including use liquid N2) - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

.. Charge with - 12

.. Verify electrical connections - 2

.. Remove equipment I normal power to heater - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

.. Release clearance I function test and 10-hour run - 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> The scheduled duration for tag out work, which requires a need for entry into 14 Condition A, is approximately 9 days. A proposed completion time of 14 days is being to allow for contingencies such as:

  • the need to replace the hermetic compressor motor and dry the refrigerant side of chiller from a tube leak (requiring and rigging), which would require approximately an additional 3 days; and
  • filtering the refrigerant if it's with corrosive products, which would approximately an additional 2 system is not risk significant as lOOleIE!O in PRA, and it is expected that if this ""clrn:;;,

maintenance were to be work could be managed more 01'1',:>1"1',,\1011\1 with more management an essential chiller is not rU'\,"r<lhlo in Modes 5 6 per chiller does have to functional in Modes 5 and 6 to support Also proposed is an editorial change to a note added as an emergency change 14 per SNC letter NL-10-1609 August 18, 2010. That note was for a one-time use and is no longer needed.

Risk Assessment An assessment was performed to the acceptability of the proposed continued plant operation while performing the overhaul maintenance of a train of ECWS beyond the current allowed CT. The assessment used criteria from Regulatory Guide 1.1 Approach for Plant-Specific, Risk-Informed Making: Technical Specifications" for determination of the acceptability of Conditional Core Damage Probability (ICCDP) and the Incremental Release Probability merit under the proposed circumstances. ICCDP and ICLERP represent the Core Damage Frequency (delta Early Release Frequency multiplied by the proposed noted that the is not in reviewed VEGP (PRA) model due to the impact of the ECWS on the functions, as discussed below. It follows, to Nl-12-1192 Basis for Proposed Change therefore, that ICCDP and ICLERF are assessed to be negligible (zero, as modeled) when the ECWS system is in a degraded one train mode of operation such as that proposed by this request.

This assessment has been performed using the peer reviewed VEGP PRA model, and using the NRC's three-tier approach described in RG 1.177. The three tiers consist of:

Tier 1 - PRA Capability and Insights Tier 2 - Avoidance of Risk-Significant Plant Configurations, and Tier 3 - Risk-Informed Configuration Risk Management Tier 1: PRA Capability and Insights (Reference Enclosure 5 for a discussion of VEGP PRA capability.)

Risk Evaluation In the VEGP internal events PRA model, room cooling is only modeled for the Emergency Diesel Generator (EDG) Rooms. The EDG room cooling is provided by plant systems other than ECWS. This assessment addresses the justification for reduction in redundancy of ECWS supported room cooling.

Methodology The approach used in the assessment of risk increase included the following considerations:

  • potential for creating a new initiating event (IE),
  • potential for an increase in the frequency of an existing IE(s), and
  • the impact on the consequence of an IE NewlE As documented in the VEGP PRA Model (PRA-BC-V-11-01, Appendix 2A, Modeling Effects of Loss of Room Cooling on VEGP Components), a number of VEGP specific room heat-up calculations have shown that room heat-up occurs over time and the room temperature can readily be reduced below equipment operating temperature limits by opening appropriate doors. Therefore, at the worst case, a loss of room cooling will result in a controlled plant manual shutdown. Crediting the "door opening" operator action to prevent an initiating event is limited to those cases where:
  • at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are available prior to room heat up to a temperature at which damage might occur to supported equipment, and
  • the room temperature is found to be stabilized In the current PRA model documentation, room R-B61 , Unit 1 Train 1B "480V SWGR 1 BB06 and room R-B18, Unit 2 Train 2B "480V SWGR 2BB06", located in the Control Building, are E1-9 to Nl-12-1192 Basis for Proposed Change assessed to require operator action (that is, to open the door). According to the VEGP room heat-up calculations, the avaiiable time to implement the compensatory measure of opening that door is 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Thus, it was determined that no new IE will be created by the requested extension in completion time.

Impact on the Frequency of an Existing IE The ECWS maintains ambient air temperature in the ESF equipment rooms and switchgear rooms below the environmental qualification rating of the ESF equipment served by the system during all postulated accidents. The ECWS consists of two independent trains, each a closed loop system. Following a SI-inducing I E, both trains of the ECWS are automatically actuated. During normal operation, the ECWS is the backup to the Normal Chilled Water System (NCWS), which provides dlilled water throughout the plant to all air cooling units which are required during normal plant operation. Because VEGP specific room heat-up calculations have shown that room heat-up occurs over time and the room temperature can readily be reduced below equipment operating temperature limits by opening doors, the impact of the proposed CT extension on the frequency of an existing IE is negligible.

Note, neither ECWS nor NCWS are required to support PRA-credited accident mitigating functions. However, NCWS can be used as a backup for ECWS in those locations served by both systems for all events that do not result in loss of offsite power. NCWS capacity is much larger than ECWS capacity. NCWS is operating and ECWS is in standby during normal power operations. If a reactor trip occurs, NCWS continues to operate unless there is a LOSP. Therefore, during normal operation and after shutdown, NCWS is the primary source of chilled water. In case of a LOCA, the ECWS auto starts on a safety injection signal. However, NCWS could be used as a backup if it has power available or if it is manually restarted when normal power is available. (Note: At least one normal chiller is powered from each unit and turbine plant cooling water for cooling can be supplied from either unit. Therefore, some cooling from NCWS can be restored if either unit has off-site power available.)

Impact on Consequences of Other IEs As stated above, the ECWS is credited to provide cooling following a SI-inducing initiating event. Based on a detailed review of VEGP-specific room heat-up calculations and industry reference documents, it has been concluded that the ECWS-supported systems will be able to perform their safety function within the PRA credited mission time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). The basis for this conclusion was reached by using the industry reference documents to establish survivability and VEGP specific calculations to establish room heat up. Industry and VEGP specific reference documents used for establishing the basis for survivability include the following:

  • "Guidelines and Technical Bases for NLiMARC Initiatives Addressing Station Blackout at Light Water Reactors," I\lUMARC 87-00 Rev.1, Nuclear Management and Resource Council, Inc., 1991
  • "Equipment Operability During Station Blackout Events," NUREG/CR-4942, Sandia National Laboratories, 1987 E1-10 to NL-12-1192 Basis for Change
  • Qualification Report Long Component Aging " WCAP 8687 (VEGP document number: AX6AA10-001 Westinghouse, 1987
  • "Equipment Qualification Westinghouse Class 1 E Equipment," WCAP 8587, Westinghouse, 1987 The VEGP nor*,t"" room heat-up evaluations includes following:
  • Up Calculations," 95-VAA093, Company, 1996
  • of HVAC," REA VG-2007, Southern Nuclear Operating Company, 1992
  • 1 & 2 Room Temperature Heatup Calculation," GP-17289, Westinghouse, 2001

'GCH.-ULI evaluations were performed for every room that contains credited For one room (R-B18, "480V 2BB06,2B807 2BBC"), located in the Control Building, the requirement for room cooling following an accident was out by crediting operator action to open door. In this the available time to the action was 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

that the calculations are in this to similarities in room the Unit 1 Unit 2 rooms, the Unit 1 calculations are judged to be applicable to the Unit 2 rooms.

Operator actions to open doors were only credited the action would not be impacted by the post environmental condition (such as radiological concern). Also, the following should be noted:

  • guidance on room in an power.
  • non-LOSP initiating the NCWS is available to provide cooling to most rooms supported by the The most likely use of the door-opening compensatory measure is a during normal There are no radiological concerns in case.

Since the of ECWS does not in an initiating event or impact any a'_A... \..I"",.

mitigating <:"1<>'+0""" and, not impact core damage, external discounted in evaluation proposed extension of ECWS CT.

Therefore, the impact of loss of on the consequences of any initiating event (due to internal is considered to negligible.

Results and Conclusion

-11 to NL-12-1192 Basis for Proposed Change The results of the risk evaluation that the potential impact of the unavailability of the ECWS on the PRA figures of merit and is negligible because the PRA components can perform their nu:>nnc,n function within the PRA mission time.

the ICCDP and ICLERP for the change in the CT are well below the Guide 1.177 acceptance and ICLERP are negligible).

CT extensions, is to provide ra",,,,nr*.,,n.

configurations will not occur when equipment is out of configurations do occur, then enhancements the or procedures, such as limiting unavailability of backup systems, increased surveillance frequencies, or upgrading procedures or can be made that avoid, limit, or importance of these configurations.

Specifically, the following Tier 2 controls are implemented during the period of the proposed extended CT:

  • Increase reliability and availability the NCWS

- No work will be performed in U1 and U2 Low Voltage or High Voltage Switchyard that might result in a

- No work will be performed on NCWS components and their supporting components that would reliability.

  • Increase reliability and availability the remaining train of ECWS

- Availability of the remaining is verified.

No work will be no."" ...*........ ",,.... train components components that would

  • Increase the reliability of to the affected room.

- Contingency plans for propping doors and placing temporary cooling in place if the normal chillers remaining essential chiller are lost.

- Minimize work on components that support the control room.

The objective of the third tier is to ensure that the risk impact of out-of-service equipment is evaluated prior to performing any activity. As stated in RG 1.177, "a program would be one that is to uncover risk-significant plant equipment outage configurations as they evolve during normal plant operation." The third-tier requirement is an extension of requirement, but addresses the limitation of not being able to identify all plant configurations in the c:a/'nn1!_T, evaluation.

SNC has developed a as~)essment and management.

process and procedures ensures that of equipment unavailability is appropriately evaluated both prior to performing activity, and following an equipment

-12 to Nl-12-1192 Basis for Proposed Change or other internal or external event that impacts risk. Procedure NMP-OS-010, "Protected Train/Division Protected Program", and Procedure 00354-C, "Maintenance Scheduling" provides guidance for managing function, probabilistic, and plant trip risks as by 10 CFR 50.65(a)(4) of the Maintenance Rule. procedures risk practices in the maintenance planning phase and maintenance (real time) phase for Modes 1 through 4. Appropriate consideration is given to unavailability, operational activities as testing, weather conditions.

In general, risk from performing maintenance during power operation is minimized by:

.. Performing only those preventive and corrective maintenance items during power operation to maintain reliability of structures, or components (SSCs).

.. Minimizing cumulative unavailability of safety-related and risk-significant SSCs by limiting number of maintenance outage windows per cycle per train/component.

.. Minimizing the total number of out of service at same time.

  • Minimizing risk of initiating plant transients (trips) that could challenge safety systems by implementing compensatory measures.
  • Avoiding higher risk combinations of out of SSCs PRA insights.
  • Maintaining defense-in-depth by avoiding combinations of out of service that are related to similar safety functions or that affect multiple safety functions.
  • Scheduling in train/bus windows to avoid removing equipment from "'UT1'Orl:l"'T trains simultaneously.

In no"or"", risk is by:

.. Evaluating plant trip risk activities or conditions mitigating them by taking appropriate compensatory measures and/or ensuring defense-in-depth of safety systems that are challenged by a plant trip.

  • and controlling risk based on probabilistic and safety function defense-in depth evaluations.
  • Implementing compensatory measures and requirements for management authorization or notification for certain "high-risk" configurations.

Actions are taken and appropriate attention is to configurations and situations commensurate with the of risk. This occurs during both planning and the real time (execution) nl"l':I<>c,c>

The current for planned maintenance activities is for an assessment of the overall risk of the activity on plant safety (including benefits to reliability performance) to performed and documented prior to scheduled work. In this consideration is given to plant external the number activities being performed concurrently, potential for plant and the availability redundant Risk is evaluated, managed, and documented all activities or conditions on the plant

-13

1 to NL-12-1192 Proposed Change

.. planned or emergent maintenance is be

.. soon as possible when an emergent plant is

.. when an external or internal event or condition is measures are implemented as necessary and if the a course of action is determined to restore ru,,,,, ...<:>n thereby the probabilistic risk.

Enclosure 4) to implementing compensatory measures as follows for 14-day period required for the overhaul maintenance for the chillers, including designation of the remaining train ESF room coolers and ECWS as a Train." Procedure NMP-OS-010 defines the "Protected Train and Protected fundamental objective of the procedure is to enhance nuclear availability of equipment necessary to maintain plant emergency rO'c .....I'\.'cO rC\f,CnT inadvertent plant trips, transients, or safety system challenges. This lJVI\,,..,.., guidance for management of the protected train and for posting protected when redundant equipment is out of service. Additionally, operation or maintenance of rotE~ctE!a plant equipment is limited or prohibited.

room coolers and ECWS equipment will no elective or corrective maintenance, surveillance or availability of the remaining equipment would to ensure continued safe operation of plant and

~ .....r'I"I'\',/Orl by management. Additionally, major components/locations room coolers and ECWS will have signage placed to that equipment is 1"1'\"~l"rt:.n" Signage locations will include both entrances to room housing train essential chiller and CREFS, the entrance to the remaining train chiller supply and main control room handswitches for the remaining train chiller chilled water pump.

maintain plant personnel awareness of the protected train, at a minimum, the protected train is on the plant morning report, in the Main Control Room, Maintenance Shop areas, HP Control Point and in the Work Release office. The protected train is also discussed at the beginning of shift for each group.

Additional compensatory measures include maintaining the following equipment available (that no routine or maintenance activities will be performed):

E1-14 to Nl-12-1192 Basis for Proposed Change

  • High Voltage Switchyard Source to Reserve Auxiliary Transformers. Work that does challenge both feeders to power will permitted managed as a high Operational Risk Awareness job.
  • The Normal Chilled Water (NCWS).

A contingency plan will be in place for monitoring in the rooms below; propping doors, as deemed necessary, per procedure 19100-C; and putting temporary cooling measures (fans) in place if remaining train's Room the remaining train's ECWS, and the remaining train's NCW system are out of It is expected that only in the case of of all HVAC (in the Control Building) will opened installed.

On loss of the HVAC there would be no design airflow, hence only airflow patterns (apart from natural convection) would those established by the portable fans. In addition, would no driving force to introduce radioactivity into the control building, hence would be no impediment to operator actions in the control building due to radioactivity. Fans and the required cords are staged in vicinity of rooms. The will be positioned to draw air from area outside rooms and discharge into rooms.

rooms whose have identified to propped open in affected Control Building electrical equipment room are as follows:

UNIT 1 B47,B48,B52, B55,B61, B76,B63

, B36, B04, B30 Regulatory Evaluation 4.1 Significant Hazards Consideration The proposed changes will provide a to the Generating Plant (VEGP) Completion Time of 3.7.1 Condition A, to allow one inoperable Engineered Safety (ESF) Room Cooler and Safety-Related Chiller train for 7 days, or 14 for chiller overhaul maintenance. extended Completion Time to 14 days for chiller overhaul maintenance will permit performance of overhaul maintenance of a of chiller continuing plant operation. extended completion to 7 days for routine maintenance will allow for a more thorough problem identification and resolution.

Also proposed is an change to a note as an change to TS 4 Southern Nuclear Operating Company (SNC) letter NL-10-1609 dated August 18, 2010. That note was for a one-time use and is no needed.

1. Does proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

E1-15 to 192 Basis for Proposed Change The changes do not any plant equipment or operating in such a manner that the probability of an accident is increased. The proposed changes will not alter assumptions relative to the mitigation of an accident or transient Therefore, the proposed changes do not involve a significant in the probability or consequences of an accident previously evaluated.

2. oroiOm;ea license create of a new or kind of any accident evaluated?

The changes do not involve any physical alteration of the plant or significant change in methods governing normal plant operation. Therefore, the proposed changes do not create the possibility of a new or kind of from any accident previously evaluated.

Does the nl"t"\nt"\c amendment a significant '::;;U,n..lIUI in a of safety?

Based on operability of the remaining ESF Room Cooler and Safety-Related Chiller Train, the accident analysis assumptions continue to met with enactment of the proposed changes. The system and operation are not affected by proposed changes. safety analysis criteria are not altered by the nl"t"\,nt"\c changes. the proposed compensatory measures for the increase in Completion Time for overhaul maintenance work activities will provide further assurance that no reduction in a margin will occur.

The proposed changes provide assurance that the ESF Room Cooler and Safety-Related Chiller system will continue to perform intended safety function.

Therefore, proposed changes do not involve a significant reduction in a margin of safety.

Based on nUL,\fG. SNC concludes that the proposed present no significant hazards .nC'-"'Tlr,n under the set forth in 10 50.92(c), and, accordingly, a finding of significant hazards consideration" is justified.

Applicable Regulatory Requirements/Criteria

  • General A to 10 CFR Design Nuclear power plant systems, and components important to safety be to withstand effects of natural phenomena as earthquakes, tornadoes, hurricanes, floods. tsunami, and without loss of capability to perform their safety functions.

For VEGP, essential chiller, water pump, and are designed in accordance with Seismic Category 1 requirements.

  • 10 CFR 50.36(c)(2)(ii): A technical specification limiting condition for operation a nuclear reactor must be established for each item meeting one or more of the following criteria:

-16

Enclosure 1 to NL-1 192 Basis for Proposed Change Criterion A structure, system, or component which operating experience or probabilistic risk assessment shown to significant to public health and For TS 3.7.14 limiting condition for operation has established the ESF room cooler and safety-related chiller 4.3 Conclusion The proposed changes will provide a revision to the VEGP Completion of 3.7.1 Condition A to allow an inoperable Cooler and Safety-Related Chiller train for 7 or 14 days for chiller overhaul maintenance. The extended Completion Times will permit completion of the overhaul maintenance of a train of essential chiller while continuing plant operation, and will allow a more thorough troubleshooting resolution for routine maintenance. Also proposed is an editorial change to delete a note added as an change to 14 per letter NL-10-1609 dated August 18, 2010. That was for a one-time use and is no needed.

The Plant Review Board Q\/U:>\Alt:~rt the proposed change to Technical Specifications and concluded it does not a significant hazard consideration and will not endanger the health and safety of public.

5. Environmental Consideration proposed changes qualifies categorical exclusion from an environmental as set forth in 10 51.22{c)(9). Therefore, no environmental impact statement or environmental assessment is needed in connection with approval of the proposed changes.

amendment meets eligibility criteria for categorical exclusion set forth in 10 51.22(c)(9) as follows:

(i) The amendment involves no significant consideration.

described in ;:;elGIICln 4.1, the proposed changes involve no significant hazards consideration.

(ii) There is no significant in the types or significant in the amounts of any that may offsite.

proposed changes do not the installation of any new equipment or the modification of equipment may affect the or amounts effluents that may offsite. Therefore, there is no significant change in the types or significant r-"Q:':H~O in the amounts of any effluents may be offsite.

(iii) There is no significant increase in individual or cumulative occupation exposure.

-17 to Nl-12-1192 for Proposed Change The proposed changes do not plant physical changes or introduce any new mode of plant operation. Therefore, is no significant increase in individual or cu occupational radiation exposure.

Based on the above, SNC concludes the proposed changes meet the criteria in 10 CFR 51.22(b) for a categorical from requirements of 10 CFR to requiring a specific environmental by the Commission.

References

1. VEGP FSAR, Revision 18, 09/1 VEGP Units 1 and 2 Technical Amendments 167 and 149 respectively, Section 3.7.9 VEGP Units 1 and 2 Environmental Protection Plan, Amendments 97 and 75 respectively NUREG-1137, "Safety Evaluation related to the operation of Vogtle Electric Generating Plant, Units 1 and "dated 1985 "Room Heat Up Calculations," 95-VAA093, Southern Nuclear Operating Company, 1996
6. "Loss of HVAC," REA VG-2007, Nuclear Operating Company, 1
7. "VEGP 1 & 2 Room Temperature Calculation," GP-17289, Westinghouse, 2001
8. Vogtle procedure 191 OO-C, "ECA-O.O of All AC Power," Revision Vogtle letter NL-10-1609, "Vogtle ~or,or:!:Itir\n Plant Unit 2 Emergency Specification Revision nOQ,rOrl S!:IT'OT\I Features Cooler and Safety-Related I'"\U'-AU,",' 18, 2010.
10. Vogtle letter NL-10-1623, "Vogtle .ora-tin,... Plant Unit 2 Emergency Technical Specification Revision Request for ,no.e,ron Safety Features (ESF) Room Cooler and Safety-Related Chiller to Requests for Additional I "

dated August 18, 2010.

11. Vogtle letter NL-11-1297, "Vogtle Generating Plant -Units 1 and 2 Methods to used in the Implementation of Risk-Informed Technical Specifications Initiative 4b," dated September 27, 2011.

1 Vogtle letter NL-11-1628, "Vogtle Generating Plant - Unit 2 Technical Specification Revision Request for TS 3.7.14 Safety Features (ESF) Room Cooler and Safety-Related Chiller System," ,ol'c.rnn.ar 1 2011.

-18

Vogtle Electric Generating Plant Units 1 and 2 Technical Specification Request for TS 3.7.14 Engineered Safety (ESF)

Room Cooler and Safety-Related Chiller System 2

Marked-up Technical Dec::lfl,ca1:lor,s and Bases Pages

Room Cooler and Safety-Related Chiller .... ""'Tarn 3.7.14 3.7 PLANT SYSTEMS 4 Engineered .... '.>Tonl Features (ESF) Room Cooler and Safety Related Chiller System LCO 3.7.14 Chiller trains shall be Safety-Related Chiller train may removed from for

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> under administrative for surveillance of the Safety-Related Chiller train.

APPLICABILITY: MODES 1, and 14 days for chiller overhaul maintenance ACTIONS CONDITION RED ACTION A. room cooler A.1 Restore room safety-related chiller cooler safety-related train inoperable. chiller train to OPERABLE status.

Required Action and in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associated Completion Time not met.

Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Vogtle Units 1 2 3.7.14-1 Amendment No. (Unit 1)

Amendment No. +99 (Unit 2)

Room Cooler and BASES LCO The associated chilled water system, including chiller, (continued) pump, piping, valves, and instrumentation required to rform the safety-related function is The is modified by a Note that allows one chiller train to be removed from service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> administrative controls for surveillance testing of other chiller train. note is required to allow surveillance separately on each safety-related chiller testing may include individual automatic starts of Administrative controls must be in to ensure the train from service can be rapidly returned to if When this note is utilized, the train from is not required OPERABLE during the APPLICABILITY and 4, the ESF room cooler must be OPERABLE to provide a consistent with the OPERABILITY equipment it supports. In MODES 5 or 6, are no OPERABILITY requirements for the room cooler and chiller system. However, the functional of the ESF room cooler and safety-related system to provide supplemental cooling for normal HVAC are determined the systems it supports. In these cooling provided by the ESF room cooler and chiller system is not a required safety function of ACTIONS room cooler and safety-related chiller to perform the heat removal function for equipment.

overall reliability is reduced because a single room cooler and safety-related chiller train could result in loss of the ESF room cooler chiller system function. The ompletion on the redundant capabilities afforded by the train, and the low probability of a DBA occurring this time.

Vogtle Units 1 and 2 B 4-3

Insert 1:

In addition, 14 days is allowed for chiller overhaul maintenance. Chiller overhaul maintenance may include maintenance of the essential chiller water side (cleaning and eddy current testing of the evaporator tubes) and maintenance of the refrigerant side (disassembly, examination /

internal inspection, and reassembly of the chillers). These overhaul maintenance activities require substantially more time than typical corrective maintenance activities. Therefore, such work is typically performed at 10-15 year intervals. With an allowance for contingencies, these activities may require up to 14 days. This duration would allow time such that the maintenance could be performed with time allowed for recovery in the event that there are complications or unforeseen problems. If the 14-day LCO for chiller overhaul maintenance is to be entered, the following actions are necessary:

  • The remaining train ESF Room Cooler and Safety-Related Chiller System will be operated as a Protected Train.
  • The Unit 1 low voltage switchyards and the Unit 2 low voltage switchyards will be maintained available (that is, no routine testing or maintenance activities will be performed) .
  • High voltage switchyards will be maintained available (that is, no routine testing or maintenance activities will be performed) with the exception of work activities which do not challenge both feeders from offsite power sources will be permitted and managed as a high Operational Risk Awareness job.
  • The Unit 1 and Unit 2 Train A and Train B Emergency Diesel Generators will be maintained available (that is, no routine testing or maintenance activities will be performed).
  • The l\Iormal Chilled Water System will be maintained available (that is, no routine testing or maintenance activities will be performed).
  • The opposite Unit's Essential Chilled Water System and the opposite Unit's CREFS will be maintained available to support control room cooling (that is, no routine testing or maintenance activities will be performed).
  • A contingency plan will be in place tor propping open doors and putting temporary cooling measures (fans) in place it the remaining train's ESF room coolers, the remaining train's satety-related chiller system, and the normal chilled water system are out-ot-service.

Vogtle Electric Generating Plant Units 1 and 2 Technical Specification Revision Request for TS 3.7.14 Engineered Safety Features (ESF)

Room Cooler and Safety-Related Chiller System Enclosure 3 Clean-Typed Technical Specifications Page

ESF Room Cooler and Safety-Related Chiller System 3.7.14 PLANT 3.7.14 (ESF) Room Cooler and Safety 3.7.14 Two ESF Room Cooler and Safety-Related Chiller trains shall be OPERABLE.

One Safety-Related Chiller train may be removed from service for

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> under administrative controls surveillance of the other Safety-Related Chiller APPLICABILITY
1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION A. One ESF room A.1 Restore the room 7 days and safety-related chiller cooler and safety-related train inoperable. chiller train OPERABLE 14 days for chiller overhaul maintenance B. Required Action and 1 Be in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associated Completion Time not met. AND B.2 in MODE 36 Vogtle Units 1 and 2 3.7.14-1 Amendment No. (Unit 1)

Amendment No. (Unit

Vogtle Generating Units 1 and 2 Technical Specification Revision Request for TS 4 Engineered Features Enclosure 4 Commitment Table

Enclosure 4 to Nl-1 192 Commitment Table of Regulatory Commitments The following table the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions. statements are provided for information purposes and are not considered to be regulatory commitments.

Regulatory Commitments Duration remaining ESF Room Cooler and S~'rQT'J_

Duration of TS 3.7.14 Condition A Related Chiller System will be operated as a overhaul maintenance Protected per procedure NMP-OS-010.

The Unit 1 low voltage switchyards and Unit 2 low voltage switchyards will maintained available (that Duration of 4 Condition A for no routine or maintenance activities will be overhaul maintenance performed).

High voltage switchyards will be maintained available (that no routine or maintenance activities will performed) with exception work activities Duration of 14 Condition A for which do not challenge both feeders from offsite overhaul maintenance power sources will be permitted and managed as a high Operational Risk job.

Unit 1 and Unit 2 A and Train B Emergency Generators be maintained available (that Duration of 14 Condition A for no routine testing or maintenance activities will be overhaul maintenance performed).

The Normal Chilled Water System will maintained Duration of TS 14 Condition A available (that is, no routine testing or maintenance f"TlUI:TIOl:> will be performed). overhaul maintenance Duration of TS 4 Condition A for overhaul maintenance A contingency will be in place for propping open doors per procedure 191 OO-C for putting temporary cooling measures (fans) in place if the Duration of 4 Condition A remaining train's ESF room coolers, the remaining overhaul maintenance train's safety-related chiller system, and the normal chilled water system are out-of-service.

Vogtle Electric Plant Units 1 and 2 Technical Specification Revision Request for TS 3.7.14 nll:H:l''''' _aT""n, Features (ESF)

Enclosure 5 Discussion of Risk Analysis Capability to Nl-1 192 Discussion of Probabilistic Risk Analysis Capability PRA Capability SNC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of PRA models for all operating SNC generation This approach includes both a proceduralized PRA maintenance and update the use of self-assessments and independent The following information this approach as it to the VEGP Technical Adequacy of VEGP PRA Model SNC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated units. SNC risk management process delineates responsibilities and guidelines for updating full power events PRA models all operating SNC generation overall SNC program defines process for regularly and interim for identified as potentially affecting PRA models due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as operated plant, VEGP PRA model has updated according to the requirements in the SNC management

  • Modifications to physical the Base (BL PRA) calculated core frequencies or large early frequencies to a significant degree) shall to determine the scope and of a revision to the model within months following Unit 2 refueling or a specific major plant modification occurring outside a refueling outage. The should updated as necessary in accordance with a schedule app roved by the PRA Manager following the scoping Upon completion the lead Unit's the other will be by modification updated to account for Unit differences which significantly impact
  • Modifications plant procedures and Technical Specifications shall be reviewed annually for changes which are of statistical significance to the of the BL-PRA and those changes documented. Reliability data, failure data, initiating events frequency human reliability data. and other such PRA inputs shall reviewed approximately every three years for statistical significance the results of the Following annual BL-PRAs shaH updated to account for the statistically changes to PRA inputs in an approved schedule.
  • BL-PRAs shall updated to changes in methodology, phenomenology, and regulation as judged to be prudent by the PRA Model or as required by regulation.

In addition to SNC risk management procedures the guidance particular risk and PRA quality maintenance This

  • Documentation of the PRA model, products, and documents.

to NL-12-1192 Discussion of Probabilistic Risk Analysis Capability

  • The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.
  • Guidelines for updating full power, internal events models for SNC nuclear generation
  • Guidance for use of quantitative and qualitative risk models in support of the during operation Work Control Program for risk evaluations for maintenance tasks (corrective maintenance, minor surveillance and modifications) on and within the scope of the Maintenance (10 50.65 (a)(4>>.

In accordance with this regularly scheduled PRA model updates nominally occur on an approximate three-year however, longer intervals may justified if it can be shown that the PRA continues to represent the as-built, plant. Table 1 shows a brief history of the major PRA model updates.

E5-2 to Nl-12-1192 Discussion of Probabilistic Risk Analysis Capability Table 1: History of the Major VEGP PRA Model Updates Model Document No. Scope Updated Items CDF and LERF (/yr)

IPE WCAP-13553 At-power, internal and The original CDF: 4.9E-05 (Westinghouse report) by external, Level 1 and Level 2, LERF: 1.78E-06 Westinghouse and SNC, 11/1992 Rev. 0 SAIC prepared reports At-power, internal, CDF and Conversion from a large Event CDF: 3.62E-05 3/1998 LERF Tree/small Fault Tree approach to a LERF: 1.72E-06 small Event Tree/large Fault tree approach (linked fault tree model The CDF reduction was method). mainly due to changes, such as removal of PRA software changed from unrealistic SSO scenarios, WESQT/GRAFTER (Westinghouse addition of more realistic Event Tree and Fault tree software) assumptions regarding the to CAFTA. effect of loss of room cooling and removal of a

'guaranteed failure' assumption made during the IPE for event CON (operator action to depressurize one SG to cause feed flow from the condensate pumps if AFW failed).

Rev. 1 PSA-V-99-002 by SNC, At-power, CDF and LERF Enhanced the treatment of operator CDF: 3.702E-05 9/1999 action dependency, removal of LERF: 2.290E-06 circular logic, and minor corrections I improvements.

E5-3 to NL-12-1192 of Probabilistic CaDab Model Rev. 2 2 iH_ .....nlAIQr CDF and LERF

_ transient Incorporated was reduced from to after the data Also, items during revIsion and Oc, especially the crediting of the plant Wilson for a back up AC power source, contributed to the reduction in reduction in LERF was mainly due to reduced failure of some of the the tsayeslarl uj-Iucm::

failure data failure data.

ES-4

Discussion Model Rev.2a increase was due to the new seal failure modes.

LERF decrease was due to in success criteria for Rev.2b in and 2c by CDF and LERF I Peer reviewed model PRA peer review team.

Revised the LERF decrease was on the new WOG LERF to a decrease in modeling guidelines. of the initiating event using the more recent generic data.

data source (NUREG/CR-5750).

The decrease in Some SGTR scenarios were of some removed from the scenarios for the LERF model.

scenarios and minor were made to analysis. Removed circular in normal pump trees.

Table 1:

Model No.

Rev. 3 PRA-BC-V-06-001 by At-power, and LERF This is the most extensive 2/2006 upgrade of the PRA model the IPE.

1"''1''''' were due to All level 1 PRA from the effects of many changes selection and of revision 3.

events to the final re I The main cause of the LERF done. increase was the of all of the sequences back into the Resolved all containment and Owners the removal the credit for B Facts and for some (F&Os). There were no A F&Os. Interfacing LOCA scenarios resolution of peer review E5-6

PRA Model Updated Items Based on the Rev.3level1 PRA This model was used for the Severe Accident success Management Alternative LERF: 1.819E-07 Analysis for the VEGP license renewal which was in The increase in treati nn 2007. from revision 3 to model was due to the Upgraded the full level 2 PRA correction of a seal model, based on WCAP-16341 probability from P guidelines which aim for producing an ASME PRA The above LERF value is the sum of capability category II LERF four LERF release cateoc>rI model.

Incorporated success terms in level 1 and 2 logic. an error in the level 1 PRA failure data.

Rev. 4 PRA-BC-V-07 -003 Originally prepared in April 2009 for RG 1.200 R1 peer and ASME PRA

  • Performed internal tlru*...rI'nn May 2009. PRA.

Rev. 4 was reissued after Not Mef'

""' ..,...."" Capability VEGP PRA Model UD(jatE~s Model Rev. 4.1 .':;'VL.;.-ViJ, error factor

.",':;'L.;.-vu. error due to to NL-1 SClJSSIOn of Pro,batUI Risk Analysis (PRA) Capability ASME PRA Standard Requirements Previous Assessment for VEGP PRA Model In addition to independent internal and external review during each model development and assessments of the technical capability been made before the PWR Group (PWROG) peer review against the ASME PRA 1.200, Revision 1 in May of 2009. Listed below are the previous PRA:

!ill An independent review was conducted under the of Owners Group (WOG) in December 2001, following the peer review included an assessment of the update process. This assessment did not identify All "8" F&Os from the 2001 Industry PRA PRA model Revision 3.

PRA model results were evaluated in the WOG cross-performed in support of implementation of the mitigating e"efern indicator (MSPI) process. Results of this cross-comparison are Westinghouse Owner's Group Mitigating Systems PRA Cross Comparison Candidate MSPI base document. Noted in nor-iti,.. features, there are no

!ill In a was performed against the available versions of the Standard and 1.200, Revision 0 (2003 trial version).

!ill In 2008, P model (draft Revision 4) was benchmarked with PWRs (Comanche Callaway, Wolf Creek) as a part of MSPI margin benchmarking concluded that there were no significant issues in the which would impact MSPI calculations.

RG 1.200 PRA for VEGP PRA Model against ASME PRA Standard internal events (including intemal flooding) at power was Revision 4 in 2009 the gaps from the 2006 self-assessment, to PRA Standard supporting requirements, and to represent the as-built as-operated plant.

model Revision 4 was reviewed per 1.200 1 Requirements. A summary of this peer review is provided a total of 327 numbered supporting requirements nine technical configuration control element. Eleven of the deleted SC-A3, SY-A9, SY-89, HR-G8, IF-A2, and to be not applicable to the P applicable met Capability Category II or E5-9 to NL-12-1192 of Probabilistic Risk Analysis (PRA) Capability Capability Category Met  % of total a No. of SRs SRs 210 70.9%

o 0%

38 12.8%

7 2.4%

14 4.7'%

24 8.1%

3 1.0%

296 100%

were judged to be not are HR-G6, aU-D3, HR-G6 was not

....v'"'.........,'" the reasonableness check of Human Reliability (HRA) was done of the PRA revision. aU-D3 was met because the from similar VEGP PRA report of meeting this requirement, which is an outdated comparison. was not met the limitation of the that could impact risk-informed applications was not identified.

Resolution of Findings from RG 1.200 Peer Review Table 2 shows details of the three Not Met" findings and the peer review.

As in Table 2, the three not met been resolved.

o to Nl-1 1 Discussion of Probabilistic Risk Analysis (PRA) Capability Table 2: Resolution of the Review F&Os associated three "SR not Met" Review F&O# level Resolution The Status of Resolution by SNG Element HR-G6 HR-G6 (SR Finding Reasonableness check for all 01 not met GG- HRAs for Revision 4 model was 1/111Il1) re-performed. All HRAs have been determined to be reasonable or have been :::lnrlrnrm:::l1IAI\/

revised.

check, as is in Section 8.3 of the November 2005 HRA u date of Revision 3.

1

Discussion of Pr()b2Ibll Analysis (PRA) Capability Table 2: Resolution of PRA Peer Review F&Os associated three "SR not Met" SRs Review F&O# Resolution The Status of Resolution by SNC Element aU-D3- aU-D3 (SR Finding Reviewer asked the In order to resolve the F&O, a new 01 CC-II Not Met) Staff to provide evidence of comparison study was comparison of the by comparing VEGP PRA results results to those from similar with two PWR PRAs (Callaway plants. The benchmark and Wolf Creek) which are report for MSPI was considered relatively similar to presented as evidence of VEGP. In addition to the comparison. Reviewer of PRA reports, a concluded that the is visit was performed to identify not sufficient evidence for more details of Callaway <>"("0

demonstrating compliance and PRA modeling.

with this SR.

The comparison showed that all three plants have LOSP/Station Black Out as the most dominant contributors which indicate that the P PRA results are not an outlier as compared to similar PWRs. Differences in the dominant CDF contributors were InvestlQatea and it was found that these differences are due to differences in details of """",>>rn configuration, operation and

..,,,,,,,,,...,, barriers for internal flooding, and in the sources for generic initiating events frequency data (VEGP PRA used the latest generic initiating frequency and failure data along with VEGP specific experience data for its data update).

Therefore, this F&O is resolved.

ES-12 to NL-12-1192 Dls,cu!;sicm of Probabilistic Risk Analysis (PRA) Capability Peer Review Resolution of Resolution by Limitations in the LERF A comparison of VEGP LERF analysis that would impact scenarios with those in Table applications are not 4.5.9.3 of the ASME PRA standard identified. LERF analysis revealed that the VEGP PRA documentation is incomplete included more potential LERF because limitations in the scenarios than as required for a LERF analysis that would dry containment in the applications, as PAA standard.

required by SR are not identified. scenarios in PRA included containment

.HlIl' .... " ... core damage scenarios generator tube and Interfacing systems LOCA),

thermal or pressure induced steam generator tube rupture after core damage, containment isolation failure with core damage, and various early containment failure modes.

noroYnr<> this F&O is resolved.

3

Vogtle Generating Plant Units 1 and 2 Technical Specification Revision Request for 3.7.14 Engineered Safety Features (ESF)

Room Cooler and Safety-Related Chiller System 6

Room Coolers ~"'rv..'n by Safety-Related Chiller System

6 to Nl-12-1192 COlolelrs Served by Safety-Related ESF Room Cooler Room Number Area Served 1-1531-N7-001-000 Main Control Room Main Control Room (Common to Units 1 and No' Control Building Battery Rooms, Motor Control Center Switchgear

-1532-A7-001-000 B56, B55, Yes Yes Rooms, and Shutdown ~

A75 1 -1539-A7-001-000 A45 Control Building Auxiliary Relay Room Yes No 1 -1555-A 7-~Ol-O~~ 0105 Auxiliary Building Electrical Switchgear and Motor Control Center Room Yes Yes 1 -1555-A 7-003-000 Cl09 Auxiliary Building Electrical Switchgear and Motor Control Center Room Yes Yes 1 -1555-A7 -005-000 118 Yes Yes 1 -1555-A7-007-000 048 Yes Yes 1 -1555-A7-009-000 076 Yes No 1 -1555-A7 -011-000 A05 Yes No 1 -1555-A 7 -013-000 Cll5 Yes Yes 1 -1555-A7-015-000 B15 Yes No 1 -1555-A7-017-000 A53 Yes YE No 1 -1561-E7-001-000 Series of Rooms Yes 1-1539-A7-005-000 325 Control Buildina Normal AC Room Yes No

  • I\If'lrm:::l is provided by normal handling (A -1531-A7-001 A -1 -A7-002) served by NeW.

Served Served ESF Room Cooler Room Number Area Served ECW NCW 1 -1531-N7 -002-000 Main Control Room Main Control Room to Units 1 and 2) Yes B61. B53. B44.

1 -1532-A7-002-000 B48, B47, B49, A50.

Control Rooms. Motor Control ve,,,<::, nvv",. O""""'l:j'::"'U Yes A43 1 -1539-A7-002-000 226 Relay Room Yes 1 -1555-A7-002-000 207 Yes 1 -1555-A7-004-000 B16 Electrical Switchgear and Motor Control Center Room Yes 1 -1555-A 7-006-000 116 Electrical Switchgear and Motor Control Center Room Yes 1 -1555-A 7-008-000 D49 Residual Heat Removal Pump Room Yes 1 -1555-A7-010-000 D77 Containment Spray Pump Room Yes 1 -1555-A7-012-000 A03 Yes I "H*.oIAIl I"" I , ....,YIi\04I1I~ '\J11\;;ii111l¥\A1 \AI 1'1...01 .\,.11l1..li111'1...0 """"'1 HI VI "'-"J,,;.;nVIII '.'I'IIUI!:::f"I~ I .,.Uilp 1 -1555-A7-014-000 C118 Yes Room 1-1555-A7-016-000 B19 Injection System Room Yes 1 -1555-A7-018-000 A07 Fuel Handlina Buildina Soent Fuel Heat Exchanaer and Pumo Room Yes Yes No 1 -1561-E7-001-000 Series of Rooms Penetration Area Yes NS' 1-1539-A7-006-000 322 Control Buildina Electrical Eauioment Room Yes Yes

-1 Water NCW Normal Chilled Water NSCW - Cooling Water to NL-12-1192 ESF Room Coolers Served by Safety-Related Chiller System Engineered Safety Features (ESF) Room Coolers Served by Unit 2 Train 2A Safety-Related Chiller System Served By Served By ESF Room Cooler Room Number Area Served ECW NCW 2-1531-N7 -001-000 Main Control Room Main Control Room (Common to Units 1 and 2) Yes No

  • A 16, A76, B01, B04, Control Building Battery Rooms, Motor Control Center Room, Switchgear 2-1532-A7 -001-000 B16, B25, B26,B27, Yes Yes Rooms, and Shutdown Room
  • B29,B85 2-1539-A7-001-000 A22 Control Building Auxiliary Relay Room Yes No 2-1555-A7-001-000 0104 Auxiliary Building Electrical Switchgear and Motor Control Center Room Yes Yes 2-1555-A7 -003-000 C07 Auxiliary Building Electrical Switchgear and Motor Control Center Room Yes Yes 2-1555-A7 -005-000 149 Auxiliary Building Electrical Switchgear and Motor Control Center Room Yes Yes 2-1555-A7 -007-000 022 Auxiliary Building Residual Heat Removal Pump Room Yes Yes 2-1555-A7 -009-000 005 Auxiliary Building Containment Spray Pump Room Yes No 2-1555-A7-011-000 A98 Auxiliary Building Component Cooling Water Pumps Room Yes No Auxiliary Building Chemical and Volume Control System Charging Pump 2-1555-A7 -013-000 C16 Yes Yes Room 2-1555-A7-015-000 B119 Auxiliary Building Safety Injection System Pump Room Yes No 2-1555-A7-017-000 A91 Fuel Handling Building Spent Fuel Pool Heat Exchanger and Pump Room Yes Yes No 2-1561-E7 -001-000 Series of Rooms Piping Penetration Area Yes (served by NSCW) 2-1539-A7-005-000 325 Control Building Normal AC Room Yes Yes
  • Normal cooling is provided by normal air handling units (A-1531-A7-001 and Legend A-1531-A7-002) served by NCW. ECW - Essential Chilled Water NCW - Normal Chilled Water NSCW - Nuclear Service Cooling Water E6-3

\"Ivl 'lfv\".( LJ'j ESF Room Room Area NCW 2-1531 Main Control Room to Units 1 and Yes No

  • Control Rooms, Motor Room, 2-1532-A7-002-000 ~ Yes 2-1539-A7-002-000 Control Room No Auxiliary Switchgear and Motor Control Center 2-1 Yes Yes Room Auxiliary Center 2-1555-A7-004-000 B122 Yes Yes Room Auxiliary Center 2-1555-A7-006-000 147 Yes Yes Room 2-1 D121 Auxiliary 2-1555-A7 -010-000 D04 Auxiliary No 2-1 2-000 Auxiliary Yes No Auxiliary Building Chemical and Volume Control System Charging 2-1555-A7 -014-000 C17 Yes Pump Room 2-1555-A7-016-000 B117 Auxiliary Room No 2-1 8-000 A04 Fuel Handlina Yes Yes No 2-1 -000 of
  • Normal is provided by normal handling units (A-1 and A-1531-A7-002) served by NCW. - Essential Chilled Water NCW - Normal Chilled
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