ML15055A039

Response to Request for Additional Information Regarding License Amendment Request for Emergency Action Level Scheme Change

ML15055A039
Person / Time
Site:	South Texas
Issue date:	02/11/2015
From:	Capristo A South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-15003214, STI: 34039132, TAC MD4195, TAC MD4196
Download: ML15055A039 (22)

Text

Nuclear Operating Company South Texas Project Electric Generatin8 Station P.. BoHa 289 Wadsworth, Texas 77483

/

February 11, 2015 NOC-AE-15003214 10 CFR 50.90 File No. G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information -

South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergqency Action Level Scheme Change (TACs MD4195 and MF4196)

References:

1. Letter; G. T. Powell to USNRC Document Control Desk; "License Amendment Request for Revision to Unit 1 and Unit 2 Emergency Action Levels;" NOC-AE-14003087; dated May 15, 2014 (ML14164A341)

2.

E-mail; Balwant Singal to Lance Sterling; "Request for Additional Information (RAI) -

Revised Emergency Action Levels for South Texas Project, Units 1 and 2 (TACs MD4195 and MF4196);" dated December 18, 2014 (ML14352A180)

By Reference 1, STP Nuclear Operating Company (STPNOC) requested approval of a License Amendment Request for revision to Unit 1 and 2 Emergency Action Levels. By Reference 2, the NRC staff requested additional information (RAI) to complete its review. STPNOC's response to Reference 2 is provided in the Attachment to this letter. The No Significant Hazards Consideration determination provided in Reference 1 is not altered by the additional information provided in this correspondence.

While formulating the attached RAI response, STPNOC discovered several discrepancies in the calculation that was performed to determine EAL threshold values for Abnormal Rad Levels. As a result, the STPNOC response to RAI-04 is not complete. As discussed with the Staff, after the discrepancies are resolved, STPNOC will provide a response to RAI-04 by February 26, 2015.

Included with the response to RAI-04 will be a clean copy and a redline markup copy of the STPEGS Emergency Action Level Technical Bases Document.

There are no commitments in this letter.

STI: 34039132

NOC-AE-15003214 Page 2 of 3 If there are any questions, please contact Drew Richards at (361) 972-7666 or me at (361) 972-7697.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on Aldo Capristo Executive Vice President Chief Administrative Officer Date amr

Attachment:

Response to Request for Additional Information - South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change

NOC-AE-15003214 Page 3 of 3 cc:

(paper copy)

(electronic copy)

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 G9A) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P.O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 Morgan, Lewis & Bockius LLP Steve Frantz U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Energy Kevin Polio Cris Eugster L. D. Blaylock Crain Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

Attachment Response to Request for Additional Information - South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change NOC-AE-15003214 Page 1 of 18 REQUEST FOR ADDITIONAL INFORMATION SOUTH TEXAS PROJECT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST FOR EMERGENCY ACTION LEVEL SCHEME CHANGE DOCKET NUMBERS 50-498 AND 499 The NRC staff requires the following additional information to complete its review of the request:

RAI-01

Because the information in the basis document can affect emergency classification decision making, NEI 99-01, Revision 6, Section 4.6 contains an expectation that the basis document will be evaluated in accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.54(q). Please explain how this expectation will be clearly identified to ensure appropriate reviews are conducted for any potential changes to the basis document.

Note: The NRC staff does understand that appropriate administrative controls are in place to ensure that changes to Abnormal and Emergency Operating Procedures are screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required. This RAI is intended to ensure similar controls are in place for the STP EAL Basis Document.

STPNOC RESPONSE TO RAI-Ol Procedure OPGP05-ZV-0010, Emergency Plan Change, provides administrative controls for evaluating changes to the STP EAL Technical Bases in accordance with 1 OCFR50.54(q).

No further action is required.

Per OPGP05-ZV-0010, 4.1 Emergency Plan Change Process; changes shall be prepared as follows:

The proposed change shall be reviewed using a 50.54(q) screen to determine whether an effectiveness evaluation is required. The screen will be performed in accordance with Addendum 1 of this procedure and documented on Form 1, Screen Evaluation Form.

IF the change impacts an Emergency Action Level, conduct a review of the EAL in accordance with Addendum 3 and document the evaluation on Form 3, EAL Change Evaluation Form.

Per OPGP05-ZV-0010 Addendum 3, EAL Change Evaluation Instructions: "The Emergency Action Levels and the EAL technical bases manual are a part of the Emergency Plan. This section is. applicable to all changes to the EALs that include changes to the technical bases manual, initiating conditions, fission product barriers, mode applicability, bases, definitions, and other subjects necessary to establish the criteria for entering an emergency plan event."

1.0 Evaluate all changes to the EAL technical bases manual in accordance with these instructions.

2.0 Use the same CR number as on the Screen and Effectiveness Evaluations as the identifier for the EAL evaluation for traceability of all three evaluation forms to the same change.

NOC-AE-15003214 Page 2 of 18

RAI-02

The proposed Section 4.3, "Instrumentation Used for EALs," was changed to include "If instrumentation failures occur that have EALs associated with them (i.e., process radiation monitors) alternate means of implementation may be used as described in plant procedures."

For consistency, please use the term "compensatory," rather than "alternate," as it relates to means of implementation, or provide further rational for use of term "alternate."

Note: Licensees are required to maintain, in effect, their Emergency Plan as stated in 10 CFR 50.54(q)(2). Since all the probable methods of compensating for instrumentation are usually too numerous to effectively analyze, licensees should make every effort to compensate for instrumentation deficiencies when they occur, and should document this compensation in their corrective action program.

STPNOC RESPONSE TO RAI-02 STPNOC has revised the proposed Section 4.3, "Instrumentation Used for EALs" to replace the term "alternate" with the term "compensatory" as follows:

"If instrumentation failures occur that have EALs associated with them (i.e., process radiation monitors) compensatory means of implementation may be used as described in plant procedures."

NOC-AE-15003214 Page 3 of 18

RAI-03

Table R-1, "Recognition Category "R" Initiating Condition Matrix," was not modified with site-specific information for RU1, RS2, and RG2. Please provide justification or revise accordingly.

STPNOC RESPONSE TO RAI-03 STPNOC has revised Table R-1, "Recognition Category "R" Initiating Condition Matrix with site specific information for RU1, RS2 and RG2 as follows:

RU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

Op. Modes: All RS2 Spent fuel pool level at 40'-4" or lower.

Op. Modes: All RG2 Spent fuel pool level cannot be restored to at least 40'-4" for 60 minutes or longer.

Op. Modes: All NOC-AE-15003214 Page 4 of 18

RAI-04

The proposed EALs RU1 appears to be a different base-value than the escalation values (RA1, etc.). Please justify further or revise accordingly. If the values are correct, please note the discrepancy in the basis section.

STPNOC RESPONSE TO RAI-04 During the development of this RAI response, STPNOC identified several discrepancies in the EAL threshold calculation for RU1, RA1, RS1, and RG1 (reference Calculation STPNOC013-CALC-002). STPNOC is tracking this item in the STP Corrective Action Program (reference CRs 15-2022, 15-2576). Once these discrepancies are resolved, STPNOC will provide a response to this RAI by February 26, 2015.

Attachment I NOC-AE-15003214 Page 5 of 18

RAI-05

The proposed EAL RU2 uses potentially confusing logic associations to determine the applicable classification. Please explain how this EAL will be consistently evaluated pending approval, or revise to add clarity regarding the logic.

STPNOC RESPONSE TO RAI-0O STPNOC has revised EAL RU2 to reduce the potentially confusing logic associations by restructuring the multiple statements in a manner consistent with the STP Emergency Operating Procedure User's Guide, OPOP01-ZA-0018. The individual choices for the conditions identified in "a" and "b" are confined to bulleted statements separated by a logical "OR" conjunction. When multiple conditions are required to satisfy the criteria within the bulleted choices, these conditions are separated by a logical "AND" conjunction.

(1) a.

UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

Visual Observation OR Annunciator alarm on lampbox 22M02 Window F-5 "SFP WATER LVL HI/LO" OR Spent fuel in the ICSA AND Annunciator alarm on lampbox 22M02 Window F-6 "SFP Trouble" AND Plant Computer point FCLC1420 "REFLNG CAV LVL IN CNTMT" (ICSA Water Level HI/LO) is in alarm.

AND

b. UNPLANNED rise in area radiation levels on ANY of the following radiation monitors.

RE-8055 (68' RCB) - Mode 5 or 6 only OR RE-8099 (68' RCB) - Mode 5 or 6 only OR RE-8090 (68' FHB)

NOC-AE-15003214 Page 6 of 18

RAI-06

The proposed EAL RA3.2 includes a number of plant areas for all operating modes. Please verify the plant areas identified for EAL RA3.2 reflect only those areas required for normal plant operations, cooldown, or shutdown, and that access to these areas is required, i.e., cannot be operated remotely. Please provide evidence of this verification, or revise as necessary to support accurate and timely assessment. In addition, consider adding operating mode specificity to the listed areas if applicable.

STPNOC RESPONSE TO RAI-06 STPNOC has revised EAL RA3.2 to include a listing of plant areas that require access for normal plant operations, cooldown, or shutdown and components in these areas cannot be remotely operated. Additionally, modes of applicability have been included for each area as follows:

(2) An UNPLANNED event results in radiation levels that prohibit or impede access to ANY of the areas listed in Table H3/R2.

TABIE'H3/R2: Plant Areas Requirij64Ac),ess Other Aux Boilers, Lighting DG, LCs 12L & 12G, IVC 44 ft all areas LU-L TGB 83ft turbine area, 55ft all areas, 29ft all open areas, Deaerator Storage Tank Stand, Cold Chem Lab, Feed Booster Pumps and Startup Feedwater Pump MAB Cubicle 335, Hot Chem Lab, CVCS Chemical Mixing Tank Room, Boron Analyzer Room 235A

.. - P.'

• 1.....,

Other Aux Boilers, Lighting DG, LCs 12L & 12G, IVC 44 ft all areas 55 ft all areas, 29 ft all open areas, Deaerator Storage Tank Stand, Cold Chem Lab, and o

T TGB Startup Feedwater Pump MAB Hot Chem Lab, High Energy Valve Room 80 Other IVC 44 ft all areas, RCB all areas, ESF DGB all areas TGB Deaerator Storage Tank Stand, Cold Chem Lab, Startup Feedwater Pump Demineralizer Resin Loading Area 329, Demineralizer Valve Operating Row Area 238, Filter Valve Operating Row Area 237 SHot Chem Lab, Radwaste Control Room, Valve Mezzanine Room 64, GWPS Valve 0 m MAB Room 68J, Water Removal Skid Room 68K, High Energy Valve Room 80, CCW SRooms Common Header Room 106A, CCW HX Room 106E, SRST Entrance Area Room 111, SWPS Process Area Corridor Room 218K Other Cubicle 108C, 19 ft Penetration Space, Penetration Areas M-55 and M-76 EAB Penetration Space all levels, 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft, RX Trip Switchgear and Rod Control Power Cabinet Room, QDPS Rooms, Aux Shutdown Panel Other IVC 44 ft all areas, RCB all areas, ESF DGB all areas TGB Cold Chem Lab S

MB o

Hot Chem Lab, Radwaste Control Room, Valve Mezzanine Room 64, CCW Rooms o0 MAB Common Header Room 106A, CCW HX Room 106E 2Other Cubicle 335, Penetration Areas M-55 and M-76 Roof, Penetration Space all levels, 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft, QDPS Rooms, Aux Shutdown Panel TGB Cold Chem Lab 0 Ln MAB Hot Chem Lab EAB 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft NOC-AE-1 5003214 Page 7 of 18

RAI-07

The proposed EAL CU1.1 uses the term "reactor vessel flange" to identify the threshold.

However, Operating Mode 6 technical specifications may allow Reactor Coolant System (RCS) level below the flange under certain controlled conditions. Please justify further the deviation from the endorsed guidance or revise accordingly.

STPNOC RESPONSE TO RAI-07 STPNOC has revised EAL CU1.1 to accommodate the below-the-vessel-flange conditions allowed when operating in Mode 6 as follows:

(1) UNPLANNED loss of reactor coolant results in RCS level below the procedurally required limit for 15 minutes or longer.

STPNOC has revised the CUI.1 Selection Bases as follows:

CUI: EAL-1 Selection Basis:

RCS inventory is maintained above the reactor vessel flange (39'-3") during refueling outages per OPOP03-ZG-0007, Plant Cooldown. RCS level may be lowered below the vessel flange for specific purposes (e.g., head removal, mid-loop operations) as described in OPOP03-ZG-0009, Mid-Loop Operation. The 15 minute time frame allows for prompt operator actions to restore RCS level in the event of an UNPLANNED lowering of RCS level below the prescribed operating limit.

NOC-AE-1 5003214 Page 8 of 18

RAI-08

The proposed EAL CS1 uses a level that is less than 33% of plenum as the "site-specific level" for EAL declaration. This level is contrary to NRC endorsed guidance which describes the "site-specific level" as 6 inches below the bottom ID of the RCS loop and not the low point of the loop. Additionally, the proposed EAL CS1 does not include the conditions where containment closure is established. Please provide a justification for these deviations from NRC endorsed guidance, or revise accordingly.

STPNOC RESPONSE TO RAI-08 STPNOC maintains that the plenum level being less than 33% in CSl-EAL1 meets the intent of the NEI 99-01. The level described in the STP EAL is as close to the specified level of "6 inches below the bottom ID of the RCS loop and not the low point of the loop" without being below the specified level. STPNOC has revised the CS1-EAL1 bases with an improved description of the Reactor Vessel Water Level (RVWL) Monitoring System to support the EAL setpoint. STPNOC has revised the CS1 EALs to include the endorsed CSl-EAL 2 (the condition where containment closure is established) and reinstituted the endorsed CS1-EAL 3. The selection bases for CSl-EAL1, CSl-EAL 2 and CSI-EAL 3 have been revised to support the EAL changes.

(1) a. CONTAINMENT CLOSURE not established.

AND

b. RCS level less than 33% of plenum (2) a. CONTAINMENT CLOSURE established.

AND

b. RCS level less than 0% of plenum (3) a. RCS level cannot be monitored for 30 minutes or longer.

AND

b. Core uncovery is indicated by ANY of the following:

Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.

OR Erratic source range monitor indication OR UNPLANNED rise in ANY of the following sump or tank levels in Table C2 of sufficient magnitude to indicate core uncovery CSI: EAL-1 Selection Basis:

Per NEI 99-01 Rev. 6, the RCS level indication should be six inches (6") below the bottom inside diameter of the RCS loop penetration at the reactor vessel. Six inches (6") below the bottom inside diameter of the RCS hot leg nozzle (elevation 31'-0.5") is elevation 30'-6.5" per OPOP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram.

The nearest RVWL Monitoring System thermocouples are located 6 inches above (Sensor 6) and 4.9 inches below (Sensor7) the prescribed elevation of 30'-6.5". When water level is at the desired elevation of 30'-6.5", Sensor 6 will be dry and Sensor 7 will be wet. This condition corresponds to a reading of 33% of plenum per OPOP02-11-0002, RVWL Monitoring System, Addendum 1, RVWL Sensor Elevations.

NOC-AE-1 5003214 Page 9 of 18 CSI: EAL-2 Selection Basis:

Per NEI 99-01 Rev. 6, the RCS level indication should be approximately the top of active fuel (TAF). The RCS level which corresponds to the top of the active fuel is 28'-2" (OPOP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram). The nearest Reactor Vessel Water Level Monitoring System thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with the inherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of signaling that the threat to the public is reduced when CONTAINMENT CLOSURE is established.

CSI: EAL-3 Selection Basis:

As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. A rising trend on these monitors can be an indication that core uncovery is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can give further indication of core uncovery.

The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOC013-CALC-006 Rev.1. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9,000 mR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 mR/hr corresponds to a water level approximately 8 inches above the top of the active fuel with the reactor head on; which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in conjunction with the other threshold values allow for an accurate assessment of the EAL.

Core uncovery can be determined by the secondary indications listed in this EAL. The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since other system leaks could raise levels in various tanks and sumps, the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment.

NOC-AE-15003214 Page 10 of 18

RAI-09

The proposed EAL CG1 does not incorporate EAL CG1.1 from the endorsed guidance, and insufficient justification was provided for this deviation.

a. Please provide further explanation as to why this threshold was not developed, or revise accordingly to incorporate.

b. The exclusionary note, from the endorsed guidance, is important to maintain as intended to ensure consistent implementation. Please justify why it was not added, or revise accordingly.

STPNOC RESPONSE TO RAI-09 STPNOC has revised EAL CG1 to incorporate EAL CG1.1 from the endorsed guidance.

Additionally, STPNOC has renumbered the current EAL CG1.1 to CGI.2 and has restored the exclusionary note to its endorsed form.

(1) a. RCS level less than 0% of plenum for 30 minutes or longer.

AND

b. ANY indication from Table C1.

(2) a. RCS level cannot be monitored for 30 minutes or longer.

AND

b. Core uncovery is indicated by ANY of the following:

" Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.

OR

" Erratic source range monitor indication OR

" UNPLANNED rise in ANY of the following sump or tank levels in Table C2 of sufficient magnitude to indicate core uncovery.

AND

c. ANY indication from Table C1.

Table Cl: Containment Challenge

• CONTAINMENT CLOSURE not established*

• >4% hydrogen exists inside containment UNPLANNED rise in containment pressure

• If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

Table C2: RCS Leakage Containment Normal Sump Pressurizer Relief Tank (PRT)

Reactor Coolant Drain Tank (RCDT)

MAB Sumps 1 thru 4 Containment Penetration Area Sump SIS/CSS Pump Compartment Sump NOC-AE-15003214 Page 11 of 18

RAI-10

The proposed Loss of RCS Barrier due to Category 3, RCS Activity / Containment Radiation, Threshold A, contains a plant-specific basis discussion where temperature induced currents as the result of an RCS leak would preclude the use of containment radiation monitors (RT-8050 and RT-8051) for approximately 40 minutes, and a secondary system break would preclude the use of the containment radiation monitors for 90 minutes.

a. Please add this information to the table as the table is the decision-maker tool used for EAL determination, or justify how this information will consistently be used by EAL decision-makers.

b. Please explain why these limitations were not included for Containment Barrier Potential Loss threshold 3.A.1, or revise accordingly.

c. Please explain why these limitations were not included for Fuel Clad Barrier Loss threshold 3.A.1, or revise accordingly.

STPNOC RESPONSE TO RAI-10 In the EAL Fission Product Barrier Table (Table 9-F2), STPNOC has revised Loss of RCS Barrier due to Category 3, RCS Activity / Containment Radiation, Threshold A, to contain the plant-specific basis discussion of temperature induced currents (TIC) as a Note; see below. TIC is a noticeable impact to the radiation reading for a RCS Barrier Loss because the resultant radiation levels are near the radiation monitor background level.

NOTE: Temperature induced current (TIC) as the result of an RCS leak precludes the use of containment radiation monitors (RT-8050 and RT-8051) for approximately 40 minutes following an RCS leak and 90 minutes following a steam line break.

Additionally, STPNOC has revised the bases for the Containment Barrier Potential Loss threshold 3.A.1 and the Fuel Clad Barrier Loss threshold 3.A.1 to state that TIC is not a.

limitation for these events.

Temperature induced current (TIC) limitations are not applicable to the Containment Barrier Potential Loss threshold 3.A.1 (Fuel Clad Barrier Loss threshold 3.A. 1) because the expected radiation dose for this event overwhelms the TIC effect. This is discussed in the 10CFR50.59 evaluation 04-8245-60 associated with DCP 04-8245-33.

NOC-AE-15003214 Page 12 of 18

RAI-11

The proposed HU4 EAL does not include the independent spent fuel storage installation (ISFSI) as an area of concern for EAL HU4.3 and HU4.4. Please justify further or revise accordingly.

STPNOC RESPONSE TO RAI-11 STPNOC has revised HU4.3 and HU4.4 to include the ISFSI as an area of concern.

(3) A FIRE within the ISFSI OR plant PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.

(4) A FIRE within the ISFSI OR plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

NOC-AE-15003214 Page 13 of 18

RAI-12

The proposed EAL HA5 appears to cover a wide range of rooms or areas during all modes of operation.

a. Please verify the plant areas identified for EAL RA3.2 reflect only those areas required for normal plant operations, cooldown, or shutdown, and that access to these areas is required, i.e., cannot be operated remotely. Please provide evidence of this verification, or revise as necessary to support accurate and timely assessment. In addition, consider adding operating mode specificity to the listed areas if applicable.

b. For EAL HA5, please provide justification for the omission of the control room as a plant area where access is needed to support normal plant operations, cooldown, or shutdown.

STPNOC RESPONSE TO RAI-12 STPNOC has revised EAL RA3.2 to include a listing of plant areas that require access for normal plant operations, cooldown, or shutdown and components in these areas cannot be remotely operated. Additionally, modes of applicability have been included for each area. See the Response to RAI-6 for details.

STPNOC has revised EAL HA5.la to include the control room as a plant area where access is needed to support normal plant operations, cooldown, or shutdown.

1 a. Release of a toxic, corrosive, asphyxiant or flammable gas into the Control Room or ANY of the plant rooms or areas listed in Table H3/R2:

AND

b. Entry into the room or area is prohibited or impeded.

TABLE H3/R2: Plant Areas RequiringAccess.

Other Aux Boilers, Lighting DG, LCs 12L & 12G, IVC 44 ft all areas LU 0

83ft turbine area, 55ft all areas, 29ft all open areas, Deaerator Storage Tank Stand, Cold 0

TGB 2E Chem Lab, Feed Booster Pumps and Startup Feedwater Pump MAB Cubicle 335, Hot Chem Lab, CVCS Chemical Mixing Tank Room, Boron Analyzer Room 235A Other Aux Boilers, Lighting DG, LCs 12L & 12G, IVC 44 ft all areas o.I TGB 55 ft all areas, 29 ft all open areas, Deaerator Storage Tank Stand, Cold Chem Lab, and 2Startup Feedwater Pump MAB Hot Chem Lab, High Energy Valve Room 80 Other IVC 44 ft all areas, RCB all areas, ESF DGB all areas TGB Deaerator Storage Tank Stand, Cold Chem Lab, Startup Feedwater Pump Demineralizer Resin Loading Area 329, Demineralizer Valve Operating Row Area 238, Filter Valve Operating Row Area 237 Hot Chem Lab, Radwaste Control Room, Valve Mezzanine Room 64, GWPS Valve m

MAB Room 68J, Water Removal Skid Room 68K, High Energy Valve Room 80, CCW SRooms Common Header Room 106A, CCW HX Room 106E, SRST Entrance Area Room 111, SWPS Process Area Corridor Room 218K Other Cubicle 108C, 19 ft Penetration Space, Penetration Areas M-55 and M-76 Penetration Space all levels, 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft, RX Trip Switchgear and Rod Control Power Cabinet Room, QDPS Rooms, Aux Shutdown Panel NOC-AE-15003214 Page 14 of 18 T.ABL.E Fl3.R.2: Plant Areas Req umiing*Access - C6rttinued Other IVC 44 ft all areas, RCB all areas, ESF DGB all areas TGB Cold Chem Lab LU Hot Chem Lab, Radwaste Control Room, Valve Mezzanine Room 64, CCW 0Rooms o 0" MAB Common Header Room 106A, CCW HX Room 106E Other Cubicle 335, Penetration Areas M-55 and M-76 EAB Roof, Penetration Space all levels, 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft, QDPS Rooms, Aux Shutdown Panel TGB Cold Chem Lab 0o0 L MAB Hot Chem Lab EAB 4.16 KV Switchgear Rooms all levels, Relay Room 35 ft NOC-AE-15003214 Page 15 of 18

RAI-13

For EALs HS6 and HG1, please consider the addition of operating mode specificity to the listed safety functions to preclude event classification when these safety functions are no longer needed in accordance with site technical specifications, or provide rational as to why this would not be appropriate.

STPNOC RESPONSE TO RAI-1 3 STPNOC has revised HS6.1 and HG1.1 to include operating mode specificity to the listed safety functions to preclude event classification when these safety functions are no longer needed in accordance with site technical specifications.

HS6.1.b

b. Control of ANY of the following key safety functions in Table H2 is not reestablished within 15 minutes in MODES 1, 2 or 3 ONLY.

Table H2: Safety Functions

• Reactivity control

• Core cooling

• RCS heat removal HG1.1.b

b. ANY of the following safety functions in Table H2 cannot be controlled or maintained in MODES 1, 2 or 3 ONLY.

Table H2: Safety Functions

• Reactivity control

• Core cooling

• RCS heat removal STPNOC has revised the EAL Selection Bases for HS6: EAL-1 and HG1: EAL-1 to include the following:

The mode applicability conditioning statement for Table H2 is based on the Technical Specification Operability requirement for the following functions of the Remote Shutdown System:

Core reactivity control (initial and long term)

RCS pressure control

" Decay heat removal via the AFW System and the SG safety valves or SG PORVs

" RCS inventory control via charging flow, and Safety support systems for the above functions.

NOC-AE-15003214 Page 16 of 18

RAI-14

For EAL SU4, please explain why consideration should be given for SU4.3. The use of 150 gallons per day may be the technical specification limit for primary-secondary leakage, but this equates to -0.1 gallons per minute when the expectation for the Notification of Unusual Event classification is 25 gallons per minute for 15 minutes or longer. The NRC endorsed guidance allows for this value in situations where licensees do not have a specific steam generator tube leakage value applicable to the initiating condition. Please justify or revise accordingly.

STPNOC RESPONSE TO RAI-14 STPNOC has revised SU4.3 as follows:

(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.

STPNOC has revised EAL-3 Selection Basis as follows:

The STP Technical Specification limit for primary-to-secondary leakage is 150 gallons per day through any one steam generator, but the specification does not specify the type of leakage. Therefore, STPEGS will use the leakage outside containment; which may include SG Tube Leakage, at 25 gpm for 15 minutes or longer in accordance with NEI 99-01 Rev. 6 guidance.

NOC-AE-15003214 Page 17 of 18

RAI-15

For EAL SU5 and SA5, please provide justification for removing the basis discussion that was included in NRC approved guidance that actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels."

STPNOC RESPONSE TO RAI-15 STPNOC has revised the Bases discussion for EAL SU5 and SA5 to include the discussion included in the NRC approved guidance that actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels."

SU5 Bases A manual action at the reactor control panels is any operator action, or set of actions, which cadses the control rods to be rapidly inserted into the core (e.g.,

initiating a manual trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels".

SA5 Bases A manual action at the reactor control panels is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g.,

initiating a manual trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels".

NOC-AE-15003214 Page 18 of 18

RAI-16

The definition of "Containment Closure" is expected to be the STP site-specific definition, not the generic terminology from the endorsed guidance. Please justify or revise accordingly.

STPNOC RESPONSE TO RAI-16 STPNOC has revised the definition of CONTAINMENT CLOSURE to be site specific as below.

This definition is derived from OPGP03-ZO-0035, Reduced RCS Inventory Operations.

CONTAINMENT CLOSURE: Those actions necessary to place the RCB in the closed containment condition that provides at least one integral barrier to the release of radioactive material. Sufficient separation of the containment atmosphere from the outside environment is to be provided such that a barrier to the escape of radioactive material is reasonably expected to remain in place following a core melt accident.