Text

Response to Request for Additional Information: License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications
	ML11318A205
Person / Time
Site:	Kewaunee
Issue date:	11/08/2011
From:	Price J; Dominion, Dominion Energy Kewaunee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	11-025B, TAC ME7110
	Download: ML11318A205 (78)
	v • d • e

Dominion Energy Kewaunee, Inc.

N490 Hwy 42, Kewaunee, WI 54216 Dominion Web Address: www.dom.com November 8, 2011 U. S. Nuclear Regulatory Commission Serial No. 11-025B Attention: Document Control Desk LIC/CDS/R2 Washington, DC 20555 Docket No. 50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS (TAC NO. ME7110)

By application dated August 30, 2011 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK), requested an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). This proposed amendment (LAR 244) would revise the KPS Operating License by modifying the Technical Specifications (TS) and the current licensing basis (CLB) to incorporate changes to the current radiological accident analysis (RAA) of record. This amendment would also fulfill a commitment made to the NRC in response to Generic Letter 2003-01, "Control Room Habitability" (References 1 and 2). The commitment stated that DEK would submit proposed changes to the KPS TS based on the final approved version of TSTF-448, "Control Room Habitability."

Subsequently, on October 17, 2011 the Nuclear Regulatory Commission (NRC) staff transmitted a request for additional information (RAI) regarding the proposed amendment (Reference 3). The RAI questions and associated DEK responses are provided in Attachment 1 to this letter. Enclosures 1 through 5 to this letter provide the changes to the specific LAR 244 pages affected by the DEK response to the questions in Attachment 1.

If you have any questions or require additional information, please contact Mr. Craig Sly at 804-273-2784.

Sincerely, J. Al n rice V icý Pr iident -- Nuclear Engineering

Serial No. 11-025B LAR 244 RAI Responses Page 2 of 3 STATE OF WISCONSIN )

)

COUNTY OF KEWAUNEE )

The foregoing document was acknowledged before me, in and for the County and State aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering of Dominion Energy Kewaunee, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this ____day of - ' ,2011.

My Commission Expires: _JL1 N.*oi </

Notary Public Attachments:

1. NRC Request for Additional Information Questions and DEK Responses

Enclosures:

1. LAR 244, Attachment 1 Changed Pages

2. LAR 244, Attachment 2 Changed Pages

3. LAR 244, Attachment 3 Changed Pages

4. LAR 244, Attachment 4 Changed Pages

5. LAR 244, Attachment 5 Changed Pages Commitments made in this letter: None

References:

1. Letter from J. A. Price (DEK) to Document Control Desk (NRC), "License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications," dated August 30, 2011. [ADAMS Accession No. ML11252A521]

2. Letter from Craig W. Lambert (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01; Control Room Habitability - Supplemental Response," dated April 1, 2005. [ADAMS Accession No. ML050970303]

3. E-mail from Karl D. Feintuch (NRC) to Craig D. Sly, Jack Gadzala (DEK),

"ME7110 - Chi-over-Q amendment application - Non-Acceptance with opportunity to supplement," dated October 17, 2011. [ADAMS Accession No. ML11290A148]

Serial No. 11-025B LAR 244 RAI Responses Page 3 of 3 cc: Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. K. D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station Public Service Commission of Wisconsin Electric Division P.O. Box 7854 Madison, WI 53707

Serial No. 11-025B ATTACHMENT I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS NRC REQUEST FOR ADDITIONAL INFORMATION QUESTIONS AND DEK RESPONSES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 11-025B Attachment 1 Page 1 of 25 NRC REQUEST FOR ADDITIONAL INFORMATION QUESTIONS AND DEK RESPONSES By application dated August 30, 2011 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK), requested an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). This proposed amendment (LAR 244) would revise the KPS Operating License by modifying the Technical Specifications (TS) and the current licensing basis (CLB) to incorporate changes to the current radiological accident analysis (RAA) of record. This amendment would also fulfill a commitment made to the NRC in response to Generic Letter 2003-01, "Control Room Habitability" (References 1 and 2). The commitment stated that DEK would submit proposed changes to the KPS TS based on the final approved version of TSTF-448, "Control Room Habitability."

Subsequently, on October 17, 2011 the Nuclear Regulatory Commission (NRC) staff transmitted a request for additional information (RAI) regarding the proposed amendment (Reference 3). The RAI questions and associated DEK responses are provided below. Enclosures 1 through 5 of this letter provide the changes to the specific LAR 244 pages affected by the DEK responses below.

NRC Question 1 , Tables 3-2 and 3-3 of the proposed license amendment request (LAR)

(Adams Accession Number MLI 12520670) provides the currently approved design-basis accident calculated radiological consequences and the proposed new design-basis accident radiological consequences, respectively. The waste gas decay tank (WGDT) rupture and volume control tank (VCT) rupture doses are given in units of rem (Roentgen equivalent man) whole body dose, and rem Total Effective Dose Equivalent (TEDE).

The NRC acceptance for these accidents provided in the safety evaluation of Amendment No. 166 (ML030210062) is based upon the doses calculated by the licensee and by the NRC staff that are within relevant dose criteria specified in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, "Accident source term,"

and Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," (ADAMS Accession Number ML003734190). These dose criteria are provided only in rem TEDE. Furthermore, the application which provided the information reviewed by the staff for Amendment No. 166 (Adams Accession Number ML020870565) also provides the WGDT and VCT doses in rem TEDE.

Serial No. 11-025B Attachment 1 Page 2 of 25 , page 5 of the LAR states:

The evaluations documented herein have employed the detailed methodology contained in RG 1.183 [Regulatory Guide] for use in design basis accident analyses for the AST [alternative source term]. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 (Reference 2) or the supplemental guidance in RG 1.183.

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accident at Nuclear Power Reactors," (Adams Accession Number ML003716792), Regulatory Position C.1.1.3 states:

In either case, the facility design bases should clearly indicate that the source term assumptions and radiological criteria in these affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria[emphasis added].

This is consistent with the regulations that do not provide a way to go back to using the acceptance criteria in 10 CFR Part 100, "Reactor Site Criteria," acceptance criterion once a conversion to 10 CFR 50.67 has been completed.

The NRC staff is concerned that the proposed VCT and WGDT rupture accidents were calculated using both rem whole body (offsite calculations) and rem TEDE (control room). Please explain why this is acceptable or make the calculations consistent with , page 5 above.

DEK Response:

The presentation of the WGDT and VCT rupture analysis results should all be in units of Rem TEDE and the acceptance criteria for offsite doses should be 0.25 Rem TEDE, based upon Kewaunee License Amendments 166 and 172 (References 4 and 5). No change is required to the values reported because nearly all the dose is from noble gases and there is no appreciable difference between the Rem whole body and Rem TEDE results for the offsite dose locations.

Revised pages for Attachments 1 and 4 of LAR 244 reflecting these changes are provided in Enclosures 1 and 4, respectively, of this letter.

Serial No. 11-025B Attachment 1 Page 3 of 25 NRC Question 2 , page 5 of the LAR states:

The evaluations documented herein have employed the detailed methodology contained in RG 1.183 [emphasis added] for use in design basis accident analyses for the AST. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 (Reference 2) or the supplemental guidance in RG 1.183.

During the acceptance review the NRC staff found that exceptions to the methodology in RG 1.183 are taken (see example below). While methods different from RG 1.183 may be found acceptable, the staff needs a full analysis to justify why the proposed assumption is conservative. This information is needed to start staffs review.

Example , page 109 of the LAR, provides the following assumption:

Activity released from the break is assumed to participate with 50% of the turbine building volume.

Credit for the turbine building volume appears to conflict with the intent of RG 1.183, Regulatory Positions C 5.1.2 and 6.4 in Appendix A. Since this assumption is not consistent with the RG 1.183, additional information is needed to perform the review.

As a minimum, the staff needs a justification for methods different than those provided in RG 1.183 and those different than the current licensing bases. The methodologies and inputs used should also be provided so that staff can make a determination whether they are conservative.

DEK Response:

DEK agrees that assumptions and departures from AST guidance need to be justified.

Main Steam Line Break (MSLB) Turbine Building (TB) Model The RAI includes a specific example regarding the assumption of the volume used in modeling the turbine building in the MSLB analysis. As indicated in the RAI, credit may be assumed for Engineered Safeguard Features (Regulatory Position C 5.1.2, i.e.,

Serial No. 11-025B Attachment 1 Page 4 of 25 safety-related, required by TS, etc.) and holdup and dilution in the turbine building should not be assumed when modeling BWR LOCAs (Appendix A 6.4). The turbine building does not qualify as an ESF. The current licensing basis (CLB) MSLB assumes that the main steam line break occurs in the environment. The need to model the turbine building in the MSLB analysis is discussed in LAR 244, Attachment 4, Section 3.5 and includes the following points:

1. The main steam piping is located within the auxiliary and turbine buildings and the control room is in the auxiliary building with access from the turbine building.

(Section 3.5.1)

2. The use of a TB volume allows unfiltered inleakage directly into the control room without the benefit of atmospheric dispersion. (Sections 3.5.1 & 3.5.3)

3. After the initial blow down of the ASG, release flows from the TB volume are varied between 10 and 1 building volume/hr to maximize offsite and CR doses and minimize any benefit of having a TB volume. (Sections 3.5.1 & 3.5.5.1)

4. The model, as shown in Figure 3.5-1, includes 800 cfm of unfiltered CR inleakage from the TB volume without the benefit of atmospheric dispersion and an additional 800 cfm of unfiltered CR inleakage from the environment.

The Turbine Building Model is based on assumptions of volume and exhaust flow. The volume used in the analysis was assumed to be 1.6E6 ft3, or 50% of the actual building volume. A sensitivity study on the dependence of control room and offsite doses to the exhaust flow rate from the turbine building was performed with exhaust flow rates assumed to be between 26,700 cfm (1 volume/hr based on the 50% volume assumption) and 267,000 cfm (10 volumes/hr based on the 50% volume assumption).

These exhaust flow rates equate to 0.5 Air Changes per Hour (ACH) and 5 ACH based on the actual turbine building volume. The selection of the exhaust flow rate for the lower bound was based on a review of data presented in Table 12-10 of Reference 6 for intact residential structures. The use of a rate shown to be representative for an intact structure is conservative based on the expectation that the turbine building will not be intact after a MSLB pressurization event (e.g., blowout panels blown out, etc.). The selection of the exhaust flow rate for the upper bound was based on the relative insensitivity of the dose results to increases in the volumetric release rates. A summary of the results of the exhaust flow sensitivity studies are presented in the tables below.

Serial No. 11-025B Attachment 1 Page 5 of 25 MSLB in the Turbine Building - Pre-accident Spike - Exhaust Flow Rate Sensitivity (50% Turbine Building Volume)

Source CR EAB LPZ Rem TEDE Rem TEDE Rem TEDE SG Bulk Liquid 1.737 0.024 0.005 26,700 cfm or 1 vol./hr TS RCS activity 0.097 0.000 0.000 26,700 cfm or 1 vol./hr Spike 2.769 0.001 0.002 26,700 cfm or 1 vol./hr SG Bulk Liquid 1.636 0.024 0.005 267,000 cfm or 10 vol./hr TS RCS activity 0.024 0.000 0.000 267,000 cfm or 10 vol./hr Spike 0.713 0.001 0.002 267,000 cfm or 10 vol./hr Intact SG Bulk Liquid 0.004 0.000 0.000 Intact TS RCS activity 0.000 0.000 0.000 Intact Spike 0.001 0.000 0.000 Total - 1 volume/hr 4.6 0.03 0.01 Total - 10 volume/hr 2.4 0.03 0.01 MSLB in the Turbine Building - Concurrent Spike - Exhaust Flow Rate Sensitivity (50% Turbine Building Volume)

CR EAB LPZ

.Source RemCREBLZNotes TEDE RemTEDE Rem TEDE SG Bulk.Liquid 1.737 0.024 0.005 26,700 cfm or 1 vol./hr TS RCS activity 0.019 0.000 0.000 26,700 cfm or 1 vol./hr Spike 2.425 0.006 0.002 26,700 cfm or 1 vol./hr SG Bulk Liquid 1.636 0.024 0.005 267,000 cfm or 10 vol./hr TS RCS activity 0.008 0.000 0.000 267,000 cfm or 10 vol./hr Spike 0.935 0.006 0.003 267,000 cfm or 10 vol./hr Intact SG Bulk Liquid 0.004 0.000 0.000 Intact TS RCS activity 0.000 0.000 0.000 Intact Spike 0.006 0.001 0.000 Total - 1 volume/hr 4.2 0.03 0.01 Total - 10 volume/hr 2.6 0.03 0.01

Serial No. 11-025B Attachment 1 Page 6 of 25 The results of the exhaust flow sensitivity study were rounded, as shown in the table below, to obtain the results presented in LAR 244.

Summary of MSLB Dose Consequences & Acceptance Criteria Reported in LAR 244 CR EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE)

Pre-accident spike 4.7 0.1 0.1 RG 1.183 Acceptance Criteria 5.0 25 25 Coincident spike 4.2 0.1 0.1 RG 1.183 Acceptance Criteria 5.0 2.5 2.5 A sensitivity study to evaluate the dependence of the dose consequences on turbine building volume was not specifically performed in the preparation of LAR 244. Instead, the impact of turbine building volume on doses was inferred from the exhaust flow rate sensitivity study. The use of a turbine building volume to retain or holdup the affected SG releases is appropriate and conservative for a Kewaunee MSLB. As discussed in LAR 244, Section 3.1.1, the primary pathway and source of inleakage into the control room is through the control room doors, primarily from turbine building air. Holding the MSLB activity within the turbine building is conservative with respect to releasing the activity directly into the environment as assumed in the CLB. The benefit of atmospheric dispersion is reduced. The proposed MSLB model sums the Control Room dose contributions from both the environment (for activity released from the turbine building) and directly from the turbine building volume. While use of 50% of the turbine building volume was selected arbitrarily, based on previous similar modeling for other Dominion facility accident analyses, the penalty (increase in Control Room doses) that this modeling created was viewed as an appropriate measure to assure conservatism in the resulting control room doses with no reduction identified in offsite doses as a result of this assumption. Dominion believes any benefits that could be judged from the dilution and holdup of releases from the 50% TB volume assumption were negated when escape rate sensitivity studies were performed (discussed and shown in the previous paragraph and tables).

As a result of this RAI, the insensitivity of the dose consequences to changes in turbine building volume was confirmed by the performance of an informal sensitivity study that included cases using 25% volume, 50% volume, 75% volume, and no turbine building volume. The results of this sensitivity study are shown in the tables below.

Serial No. 11-025B Attachment 1 Page 7 of 25 MSLB in the Turbine Building - Pre-accident Spike - Volume Sensitivity (0.5 TB ACH = 26,700 cfm or 1 vol/hr based on 50% volume)

Source CR EAB LPZ Rem TEDE Rem TEDE Rem TEDE SG Bulk Liquid 1.970 0.024 0.005 No volume TS RCS activity 0.016 0.000 0.000 No volume Spike 0.474 0.001 0.002 No volume SG Bulk Liquid 1.706 0.024 0.005 25% volume TS RCS activity 0.098 0.000 0.000 25% volume Spike 2.798 0.001 0.002 25% volume SG Bulk Liquid 1.737 0.024 0.005 50% volume TS RCS activity 0.097 0.000 0.000 50% volume Spike 2.769 0.001 0.002 50% volume SG Bulk Liquid 1.898 0.023 0.005 75% volume TS RCS activity 0.096 0.000 0.000 75% volume Spike 2.741 0.001 0.002 75% volume Intact SG Bulk Liquid 0.004 0.000 0.000 Intact TS RCS activity 0.000 0.000 0.000 Intact Spike 0.001 0.000 0.000 Total - No volume 2.5 0.03 0.01 Total - 25% volume 4.6 0.03 0.01 Total - 50% volume 4.6 0.03 0.01 Total - 75% volume 4.7 0.02 0.01

Serial No. 11-025B Attachment 1 Page 8 of 25 MSLB in the Turbine Building - Concurrent Spike - Volume Sensitivity (0.5 TB ACH - 26,700 cfm or 1 vol/hr based on 50% volume)

Source CR EAB LPZ Rem TEDE Rem TEDE Rem TEDE SG Bulk Liquid 1.970 0.024 0.005 No volume TS RCS activity 0.006 0.000 0.000 No volume Spike 0.765 0.006 0.003 No volume SG Bulk Liquid 1.706 0.024 0.005 25% volume TS RCS activity 0.020 0.000 0.000 25% volume Spike 2.485 0.006 0.003 25% volume SG Bulk Liquid 1.737 0.024 0.005 50% volume TS RCS activity 0.019 0.000 0.000 50% volume Spike 2.425 0.006 0.002 50% volume SG Bulk Liquid 1.898 0.023 0.005 75% volume TS RCS activity 0.019 0.000 0.000 75% volume Spike 2.375 0.005 0.002 75% volume Intact SG Bulk Liquid 0.004 0.000 0.000 Intact TS RCS activity 0.000 0.000 0.000 Intact Spike 0.006 0.001 0.000 Total - No volume 2.8 0.03 0.01 Total - 25% volume 4.2 0.03 0.01 Total - 50% volume 4.2 0.03 0.01 Total - 75% volume 4.3 0.03 0.01 The results of this sensitivity study verify that for both the pre-accident and concurrent spike MSLB accidents, the selection of 50% turbine building volume is an appropriate assumption for the purpose of calculating a conservative control room dose without providing any dilution or holdup benefit in resulting offsite doses.

Further explanation for the insensitivity of control room dose to assumed turbine building volume is provided in the following Figure, which shows concentration of 1-131 versus time in the turbine building volume or at the control room intake for the concurrent iodine spike. The Figure demonstrates that concentration quickly achieves equilibrium within the assumed volume based on the conservative removal rate of 0.5 TB ACH that result in nearly identical control room dose consequences. A similar pattern can be observed in the pre-accident spike data, which was not plotted.

In summary, the use of the turbine building in the MSLB analysis model is an exception to standard methodology described in RG 1.183. DEK has presented in this response and LAR 244 why this exception is justified and conservative. Due to the nature of Kewaunee's control room design where tracer gas studies conclude that the majority of inleakage into the control room is through doorways that communicate primarily with the

Serial No. 11-025B Attachment 1 Page 9 of 25 TB, the use of a TB volume into which MSLB activity is assumed to be released and retained before release into the environment results in higher control room dose and no apparent decrease in offsite dose from the holdup and dilution. The MSLB analyses include assumptions of both TB volume and TB exhaust flows that are combined to make a conservative model. Two sensitivity studies performed by DEK (one on TB exhaust rate discussed in LAR 244 and another on TB volume performed for this response), confirm that use of a TB volume in the MSLB model results in conservative dose consequences when compared to the CLB and does not result in a benefit from holdup and/or dilution of activity released into the assumed 50% TB volume. The sensitivity studies discussed demonstrate the insensitivity of the model to TB volume and justify assumption bases and logic why this model is conservative.

Serial No. 11-025B Attachment 1 Page 10 of 25 Concurrent Spike 1-131 Concentration Vs Time 0.5 TB ACH (26700 cfm or 1 vol/hr based on 50% volume) 0 2 3 4 5 6 7 8 9 10 11 IOOE- 02 -_______

---Concunrrnt 50%I Concurrent NoT Enviromnent Concun-ent 25% B 03 i Concurrent 75%'

LO0E-1.00E-3 CiiM I.OOE-C05 1.00E-06 1,00E-07 Time (hr)

Serial No. 11-025B Attachment 1 Page 11 of 25 NRC Question 3 , Page 28 of the LAR states:

DEK [Dominion Energy Kewaunee, Inc.] is proposing to adopt TSTF -312 [Technical Specification Task Force], "AdministrativelyControl Containment Penetrations." , Page 30 of the LAR states:

Specifically, consistent with TSTF-312, the proposed change to allow the containment equipment hatch to be open to the outside atmosphere during movement of recently irradiatedfuel assemblies within containmentis based on: ....

2. A commitment to implement acceptable administrative procedures that ensure, in the event of a refueling accident .... that the equipment hatch can and will be promptly closed [emphasis added] following containment evacuation... , page B 3.9.6-3 and B 3.9.6-4 of the LAR states:

If it is determined that closure of the equipment hatch and/or containment penetrations would represent a significant radiological hazard to the personnel involve, the decision may be made to forgo [emphasis added] closure of the hatch and/or penetrations.

The above language from Attachment 3 is not contained in TSTF-312 nor does it appear to be consistent with the intent of the commitment cited above. No explanation is provided to justify this proposed deviation.

For each proposed deviation from the referenced TSTF models please provide a justification for why the deviation is conservative.

DEK Response:

DEK agrees that TSTF-312 does not contain the cited language from Attachment 3 of the submittal. The proposed language would be located in the KPS TS 3.9.6 Bases.

This language was introduced because similar language has been approved for other Dominion plants (References 7, 8, and 9). The intention of this language is to ensure that if radiological conditions represent a significant hazard to personnel who must close the equipment hatch, then the decision may be made to forgo closure of the hatch. The

Serial No. 11-025B Attachment 1 Page 12 of 25 proposed language is intended to ensure workers are protected in the event that radiological conditions would not allow closure of the hatch. The language is not intended to contradict the commitment to ensure administrative controls are in place that will ensure the hatch is closed in the event of a FHA.

DEK provides the following justification for this statement:

The revised RAA for the FHA assumes the containment hatch is open for the duration of the accident and the resulting calculated doses are well below 10 CFR 50.67 limits.

Therefore, the commitment to ensure closure of the containment hatch after a FHA provides additional defense-in-depth to the plant design. However, closure of the containment hatch is not a vital action to mitigate the consequences of, or recovery from a FHA. Since the hatch is assumed open in the revised RAA for the duration of the FHA, and results are well below 50.67 limits, this proposed change is considered to be acceptable.

Serial No. 11-025B Attachment 1 Page 13 of 25 NRC Question 4 The following is proposed to be added to the bases for to Technical Specification (TS) 3.7.10, "Control Room Post Accident Recirculation (CRPAR) System":

This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, dampers, hatches, floor plugs, and access panels.

This change appears to be consistent to TSTF-448, Revision 3, "Control Room Habitability" (ML062210095). Per the staffs safety evaluation for TSTF-448, Revision 3 (ML063460558), the allowance of this note was found acceptable because the administrative controls will ensure that the opening will be quickly sealed to maintain the validity of the licensing basis analyses of design basis accident (DBA) consequences.

For the Kewaunee the proposed DBA safety analyses does not appear to consider doors, dampers, floor plugs or access panels to be open at the time the accident occurs or the time to isolate them after an accident occurs. For each of the openings allowed to be open by the proposed note provide an analysis which justifies that the time to close these openings will not impact the design basis accident consequences and that 10 CFR 50.67 will continue to be met.

DEK Response:

On October 26, 2011, a telephone conference was held between the NRC staff and DEK concerning Question 4 above. This telephone conference is documented in a telephone call summary [ADAMS Accession No. ML11298A265]. During this telephone conference, the NRC staff indicated that a response to Question 4 is not necessary in order to complete the acceptance review of LAR 244.

Serial No. 11-025B Attachment 1 Page 14 of 25 NRC Question 5 The following note is proposed to be added to TS 3.7.10:

The CRE shall be isolated during movement of recently irradiatedfuel assemblies.

The note is added to an existing note which states that the control room can be opened intermittently under administrative controls.

The LAR states:

As a result of the analyses documented in this LAR, the alternate control room intake will be restricted from use. This restriction is required because of the X/Q that would result due to the close proximity of the alternate intake to various release points; one of which is < 10 m from the alternate intake. Administrative controls will be in place to assure the alternate control room intake is closed and prohibit its use during normal operation,following an accident, or while moving recently irradiatedfuel.

RG 1.183 states:

5.1.2 Credit for Engineered Safeguard Features Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operatingprocedures.

The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regardingthe occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulatedradiologicalconsequences.

a) RG 1.183, Regulatory Position 5.1.2 provides credit for mitigation features that are required to be operable by technical specifications. Justify why the proposed note is considered a technical specification which would require the system to be operable.

b) The two notes appear to conflict with one another since the control room cannot be both isolated and opened at the same time. Which note takes priority and how does the operator know which note takes priority?

Serial No. 11-025B Attachment 1 Page 15 of 25 c) How is the need for fresh air satisfied and how is this incorporated in the analysis used to demonstrate compliance with 10 CFR 50.67?

DEK Response:

a. LCO 3.7.10 requires two CRPAR trains to be operable. The Applicability of LCO 3.7.10 is Mode 1-4 and during movement of recently irradiated fuel assemblies.

Proposed TS SR 3.7.10.4 would require the CRE to be tested periodically in accordance with the proposed CRE habitability program. TS 3.7.10, Condition B addresses inoperable CRE boundary in Modes 1-4 and TS 3.7.10, Condition E addresses CRE boundary issues during movement of recently irradiated fuel.

Therefore, the proposed TS 3.7.10 would require the CRE boundary to be operable when the plant is in Mode 1-4 and during movement of recently irradiated fuel assemblies.

The new Note was intended to require that the CRE shall be physically isolated during movement of recently irradiated fuel. This requirement is necessary to accommodate no longer crediting the control room vent radiation monitor (R-23) for control room isolation after a FHA. The control room vent radiation monitor is a single train non-safety related radiation monitor and is currently credited for isolating the control room after a FHA. The proposed new RAA assumes the CRE is isolated at the initiation of the FHA rather than being isolated by the control room vent radiation monitor upon receipt of a high radiation signal after a FHA.

DEK has reviewed this proposed Note and its basis and has concluded that it is more appropriate to delete the proposed Note and include the requirement that the CRE be isolated during movement of recently irradiated fuel as part of the TS 3.7.10 LCO. This ensures that the proposed requirement is understood and uniformly considered to be part of the TS. In addition to adding this requirement to LCO 3.7.10 a new Condition has been added to Condition E. This new Condition would apply to situations where the CRE is not isolated during movement of recently irradiated fuel assemblies. If Condition E is not met the corresponding Required Action is to immediately suspend movement of irradiated fuel assemblies.

b. As stated in the response to item a. above, the Note requiring the CRE to be isolated during movement of recently irradiated fuel is being deleted and re-introduced as part of the TS 3.7.10 LCO. The current TS 3.7.10 LCO Note will

Serial No. 11-025B Attachment 1 Page 16 of 25 remain as modified. This Note allows the CRE to be opened intermittently under administrative controls. This Note does not conflict with the proposed requirement for the CRE be isolated during movement of irradiated fuel. The Note is intended to modify the LCO and allow the CRE to be opened intermittently under administrative controls. This Note would be applicable when the plant is in Mode 1-4 and when the control room is isolated during moving recently irradiated fuel. Both the requirement to isolate the control room during movement of recently irradiated fuel and the Note allowing the CRE boundary to be intermittently opened under administrative controls could be implemented at the same time, if necessary. That is, while the CRE is isolated during the movement of irradiated fuel, the CRE boundary could be opened intermittently under administrative controls using the Note.

c. Fresh air is supplied to the control room by either normal CRE inleakage or, if necessary, through the normal control room air intake. As stated in item b.

above, the current Note applicable to LCO 3.7.10 allows the CRE boundary to be opened intermittently under administrative controls. Therefore, if it is necessary to provide fresh air to the control room through means other than normal inleakage, the normal control room air intake can be opened under administrative controls to provide fresh air.

The measured CRE inleakeage during operation of each CRPAR train is provided in LAR 244, Attachment 1, Table 3-1 (see page 18 of 52). CRE inleakage is assumed to be 800 cfm in the revised radiological analyses (RAA).

The actual measured CRE inleakage is approximately half of the inleakage assumed in the RAA (i.e. about 400 cfm). The use of the LCO 3.7.10 Note to open the normal control room air intake and provide fresh air is not explicitly accounted for in the RAA. However, such a plant configuration would be infrequent and administrative controls are required to be in place. As detailed in Attachment 5 (see page 2-3) of the LAR, the manual actions necessary to isolate the CRE when in this configuration are simple and can be accomplished in a short period of time from the control room using control switches.

A copy of the revised proposed TS 3.7.10 is provided in Enclosure 2 of this letter.

Several pages of LAR 244, Attachment 1 were also revised to delete reference to the proposed change as an addition of a Note. These changed LAR 244, Attachment 1 pages are included in Enclosure 1 of this letter.

Serial No. 11-025B Attachment 1 Page 17 of 25 NRC Question 6 Page 65, Attachment 4 of the LAR states:

A critical parameter in the radiological-impactanalysis is the definition of a proper Partition Coefficient (PC) for the iodines in the RWST water. The PC applicable to the iodines in the RWST water was based on information in A. K. Postma, L. F.

Coleman and R. K. Hilliard (Reference 26), "Iodine Removal from Containment Atmospheres by Boric Acid Spray," Report No. BNP-100, Battelle Memorial Institute, Pacific Northwest Laboratories (PNL), Richland, WA 99352 (7/1970). Use of BNP-100 is discussed in SRP 6.5.2 (Reference 22). For this application, the RWST is assumed to behave like a closed system for the establishment of equilibrium conditions between the water and air. This is the same method Dominion has employed at Millstone Unit 3 and North Anna submittals (References 33 and 34) for RWST releasesdue to sump back-leakage.

The LAR uses the Reference 26 report to calculate decontamination factor for the Reactor Water Storage Tank (RWST).

Standard Review Plan (SRP) 6.5.2 states:

... Experiments with fresh sprays with no dissolved iodine were found to be quite effective at scrubbing elemental iodine at a pH as low as 5 (References 18 and 15)

[Reference 15 is the BNP-100 report]. Solutions with dissolved iodine such as the sump solutions that recirculate after an accident may revolatilize iodine if the solutions are acidic (References 17 and 10). [emphasis added] Chemical additives in the spray solution have no significant effect on aerosol particle removal because this removal process is largely mechanical.

The NRC safety evaluations for Millstone Unit 3 and North Anna submittals cited do not appear to explicitly mention or provide an NRC evaluation of the Reference 26 methodology. The NRC staff has the following concerns regarding the methodology in Reference 26:

The impact of radiation and pH do not appear to have been considered in the modeling of the RWST release pathway. The RWST solution is acidic (due to the presence of boric acid) and may revolatilize iodine as described above in SRP 6.5.2. More recent studies than Reference 26 (NUREG/CR-5950 and NUREG/CR-4697) have shown that the formation of volatile iodine as elemental iodine is dependent on radiation and the solution pH.

Serial No. 11-025B Attachment 1 Page 18 of 25 The cited Reference 26 was not developed for the RWST.

Please address these concerns in a justification why the Reference 26 methodology is applicable for use with the RWST back leakage or consider these concerns in the analysis. Please provide the pH vs. time for the RWST liquid.

DEK Response:

One of the contributions to the radiological consequences from a postulated LOCA is the release of radioactivity from leakage through valves that isolate the Engineered Safety Features (ESF) recirculation systems from tanks vented to atmosphere (e.g., the refueling water storage tank (RWST)). The Kewaunee current licensing basis for analyzing this release pathway is documented in the Safety Evaluation Report related to Amendment 166 for Kewaunee (Reference 4). Specifically, the leakage water which is at a temperature below 212 OF enters the RWST which is significantly below 212 OF and enters the tank from the bottom.

The current licensing basis for this release pathway, models the evolution of iodine from sump fluids that leak into the RWST by assuming 1 percent iodine partition. This method treats leakage into the RWST the same as any other ESF leakage. No credit was taken for dilution into existing 'clean' RWST boric acid fluid that remains in the tank nor release into the gaseous 'unfilled' region of the tank. As specified in the SER for Kewaunee License Amendment 166 (Reference 4), "The radiological consequence contribution from this pathway is less significant (less than 2 percent) at the EAB for the postulated LOCA."

In this dose re-analysis effort, DEK reviewed all inputs, assumptions, and methods used in the current analyses of record for appropriateness and consistency. As specified in LAR 244 and discussed with the NRC staff during a telephone conversation on October 20, 2011, DEK chose to model the release of radioactivity from RWST back-leakage consistent with similar modeling and use at other Dominion facilities (References 10 and 11; Millstone Unit 3 TAC NO. MC3333; and North Anna TAC No's MD3197 and MD3198). The applied DF values used to model RWST releases for these facilities were 100 and 40, respectively.

Dominion provided previous explanations as to the applicability and appropriateness for this method in a response to an NRC Request for Additional Information (Reference 12) when the method from Ref. 26 ("Iodine Removal from Containment Atmospheres by Boric Acid Spray," Report No. BNP-100, Battelle Memorial Institute, Pacific Northwest

Serial No. 11-025B Attachment 1 Page 19 of 25 Laboratories (PNL), Richland, WA 99352 7/1970, A. K. Postma, L. F. Coleman and R.

K. Hilliard)was first employed to estimate iodine partition coefficients in an RWST for Millstone Unit 3. The emphasis in that response focused on the fact that an RWST should be treated as a closed system for the establishment of achieving equilibrium conditions between the water and air in the tank. The environment in the RWST does not experience high temperature, radiation, or forced ventilation. It is a static environment that will experience partitioning between the higher iodine concentration in the water that will drive to achieve equilibrium conditions with the air. The studies documented in BNP-100 discuss that when the gas phase concentration achieves near equilibrium constant gas phase concentration, the behavior of the iodine mass to release from the aqueous phase will be mitigated. This mitigation was not credited in the Dominion release model.

The importance of pH in iodine evolution is understood and was considered in the application of iodine release from an RWST. The RWST environment does not experience high doses which relate to iodine radiolysis discussed in more recent studies performed on iodine evolution and the importance of pH control to hinder the formation of 12, (e.g., NUREG/CR-5950 and NUREG/CR-4697). It is the formation of hydrogen peroxide (H2 0 2 ) during the irradiation of water that primarily reacts with iodide (I ) to form 12. For pools with sufficient iodine concentration that contend with the interaction of iodine species and the products of water radiolysis, the effect of pH has been shown to be a sensitive parameter in iodine evolution, with lower pH promoting higher iodine evolution. Because the RWST does not experience high doses, water radiolysis is not a direct contributor to the evolution of iodine. The BNP-100 study was based on acidic solutions not including the effects of radiolysis, more applicable to Kewaunee's RWST.

Therefore, the partition coefficient curves from the BNP-100 study are directly applicable to this configuration.

The initial volume of "clean" RWST fluid remaining in the tank following Safety Injection is approximately 39,000 gallons of boric acid, with a pH of about 4.5. The pH in the RWST over the course of this event has not specifically been calculated, but using basic chemistry principles of adding a weak base (i.e., sump fluid above pH of 7) to an acidic solution, the resulting pH in the tank remains acidic over the entire event. The resulting contaminated fluid in the RWST can best be referred to as an acidic, dilute iodine aqueous solution.

Dominion's application of the iodine partition coefficient model to calculate releases from back-leakage to Kewaunee's RWST during the recirculation phase of the LOCA accident contains assumptions which provide conservatisms that yield higher than expected dose consequences from this release pathway. The total iodine concentration

Serial No. 11-025B Attachment 1 Page 20 of 25 in the RWST water is the critical factor that determines the partition coefficient. The concentration of iodine in the RWST increases over time with the maximum occurring at the end of the 30 days. The concentration of iodine in the RWST at the end of 30 days is about 3 milligrams/liter.

The partition coefficient (PC) will be at a minimum when the concentration of iodines in the RWST are the greatest at the end of 30 days. The smaller the PC, the larger the amount of iodine that will partition from the water phase and enter into the gas phase and be available for release. The PC corresponding to the maximum iodine concentration of about 3 milligrams/liter was taken from Figure 8 of the BNP-1 00 study.

The PC is approximately 600. Using the equation from SRP 6.5.2, Rev. 2, and the ratio of RWST liquid to air at the end of 30 days of approximately 0.6, the DF is calculated to be approximately 360.

DF = 1 + (Vliquid / Vair) x PC The absolute lowest calculated DF was calculated to occur around 300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months into the accident based on the effect of varying liquid to air volumes and a decreasing partition coefficient as concentration in the liquid increases with time.

Calculated DF in RWST based on Iodine Partition Coefficient 900 700 " ..-

8~00 -:

600 400 -

300 - "

200 100 0 I I I I I Ii 0.1 3 6 12 24 48 96 200 400 720 Hours Dominion selected to model the release of iodine from the RWST liquid throughout the entire accident using a DF value of 100 for conservatism (more than three times lower than the lowest calculated DF value). The volumetric flow of gas containing iodine

Serial No. 11-025B Attachment 1 Page 21 of 25 partitioned out of the liquid, is released out of the RWST at a rate equal to the liquid volumetric leakage rate into the tank. There is no mechanistic phenomenon acting upon the tank to promote accelerated or exchange of gas from within the tank. When compared to a dynamic partition coefficient which would start as high as 5000 and gradually decrease to the minimum value of 600, the integrated curies of iodine released under each modeling scenario would indicate the model selected using a DF=100, releases at least 5 times more iodine than actually anticipated. Even with the application Dominion has selected to model RWST releases with a conservative DF=100, the total contribution to EAB dose from the RWST is only 0.1%. If the DF were decreased to a value of 40, the contribution to EAB dose would increase to only 0.2%.

The Kewaunee RWST is not a significant contributor to offsite or control room dose during a LOCA. This is the same conclusion that was found in the SER to Amendment 166 (Reference 4).

In summary, DEK did address the impact of radiation and pH when modeling RWST releases. The RWST is best described as a nearly closed system that does not experience high temperature, radiation, or forced ventilation. Without the effects of high doses, water radiolysis is not a direct contributor to the evolution of iodine due to lower pH, as discussed in recent studies. The primary behavior of iodine in the RWST liquid will be to achieve equilibrium conditions with the air in the tank. Since the liquid entering the tank and the water in the tank are both less than 212 OF, flashing will not occur. Partitioning will occur to achieve equilibrium conditions in the RWST. The BNP-100 study addresses iodine partitioning in acidic solutions without the effects of radiolysis. The RWST static environment is applicable to the conditions discussed in the BNP-100 study. DEK calculated the lowest expected DF based on the lowest calculated partitioning coefficient over the entire duration of RWST releases. Using a DF value of 100 for the entire modeled release provides at least a factor of five conservatism compared to the expected integrated release that would result if dynamic partitioning were used over the 30-day release period.

Serial No. 11-025B Attachment 1 Page 22 of 25 NRC Question 7 This application is not risk-informed (even in Attachment 5) and DRA/APLA should not be involved in the review.

For reasons stated below the application is non-acceptable with an opportunity to supplement:

7.1 The licensee needs to review its utilization of Standard Review Plan (SRP) 19.2, Appendix D. The appendix is specifically for the staff to consider in determining if an application that is not risk-informed should be risk-informed because of "special circumstances:" It is not for the licensee to use to justify not being risk-informed.

This application is not acceptable with this discussion included.

7.2 The NRC staff notes that in Section 3.1 of Attachment 5, counter to their earlier statement in Step 1 of not being risk-informed, in Step 2 the licensee cites the definition from Regulatory Guide (RG) 1.200 (a PRA RG) on core damage to support their position. However, their interpretation of this definition is not complete and is flawed (that is, it is correctly quoted, but not correctly interpreted).

Since Probabilistic Risk Assessments (PRAs) do not model the specifically identified design bases accidents DBAs (Fuel Handling Accident (FHA) and Locked Rotor Accident (LRA)) related to control room habitability, they should not use the "estimated importance measure" since it needs an estimate of risk importance (and the needed PRA does not exist). Rather, the licensee should do the generic Human Action (HA) Method. The HAs need to be addressed/reviewed in a traditional, deterministic manner.

Based on the above comments, NRC staff recommends not accepting this application for review until the above issues are resolved.

DEK Response:

7.1 In LAR 244, Attachment 5, DEK followed the guidance contained in NUREG-1764, "Guidance for the Review of Changes to Human Actions." In LAR 244, Attachment 5, DEK provides a summary of the key aspects of the NUREG-1764 as part of the strategy for justifying the acceptability of the proposed new human actions necessary to meet the assumptions contained in the RAA. NUREG-1764 references SRP 19.2 Appendix D in several places. DEK did not use the guidance in SRP 19.2, Attachment D to justify the acceptability of the HAs. The

Serial No. 11-025B Attachment 1 Page 23 of 25 reference to SRP 19.2, Appendix D was made in Attachment 5 in order to provide a complete summary of the NUREG-1764 process. Section 2.4.1 of NUREG 1764 refers the reader to SRP 19.2 Appendix D. However, it also states this assessment is performed by a risk analyst. In order to eliminate potential confusion with respect to use of SRP 19.2, Appendix D, Attachment 5 has been revised to contain no mention of, or reference to, SRP 19.2, Appendix B.

7.2 NUREG-1764 contains two methods for assessing the safety significance of proposed human actions (HAs). The method used by DEK in Attachment 5 was the "Estimated Importance Method" (NUREG-1764, section 2.4.3.1). The second method is the "Generic HA Method" (NUREG-1764, section 2.4.3.1). DEK agrees that use of the "Generic HA Method" is appropriate for the specific HAs being proposed in LAR 244. LAR 244, Attachment 5 has been revised to show the assessment of safety significance using the Generic HA Method. The conclusions of the assessment were not affected by the change in method.

A revision to LAR 244, Attachment 5 is provided in Enclosure 5 of this letter.

Serial No. 11-025B Attachment 1 Page 24 of 25 Additional Information:

DEK notes that there is an error in one of the marked-up TS Bases pages submitted with the LAR. In LAR 244, Attachment 3, marked-up TS Bases page B 3.7.10-9 contains a Reference 5 as follows:

5. NEI 99-03, "Control Room Habitability Assessment," March 2003.

This reference is not correct and should read as follows:

5. NEI 99-03, "Control Room Habitability Assessment Guidance," June 2001.

A corrected copy of marked-up TS Bases page B 3.7.10-9 is provided in Enclosure 3 of this letter.

References

1. Letter from J. A. Price (DEK) to Document Control Desk (NRC), "License Amendment Request 244, Proposed Revision to Radiological Accident Analysis and Control Room Envelope Habitability Technical Specifications," dated August 30, 2011. [ADAMS Accession No. ML11252A521]

2. Letter from Craig W. Lambert (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01; Control Room Habitability - Supplemental Response," dated April 1, 2005 [ADAMS Accession No. ML050970303]

3. E-mail from Karl D. Feintuch (NRC) to Craig D. Sly, Jack Gadzala (DEK), "ME71 10

- Chi-over-Q amendment application - Non-Acceptance with opportunity to supplement," dated October 17, 2011. [ADAMS Accession No. ML1I1290A1 48]

4. Letter From John Lamb (NRC) to Tom Coutu (NMC), "

Subject:

Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC No. MB4596)," dated March 17, 2003. [ADAMS Accession No. ML030210062]

5. Letter from John Lamb (NRC) to Tom Coutu (NMC), "

Subject:

Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Stretch Power Uprate (TAC No.

MB9031)," dated February 27, 2004. [ADAMS Accession No. ML040430633]

6. R. H. Perry and C. E. Chilton, ed., "Chemical Engineers Handbook 5 th edition", pg 12-27, McGraw-Hill Book Company, 1973.

7. Letter from S. Monarque (NRC) to D. A. Christian (VEPCO), "North Anna Power Station, Units 1 and 2 - Issuance of Amendments on Implementation of Alternate Source Term," dated June 15, 2005 (pg 18).

Serial No. 11-025B Attachment 1 Page 25 of 25

8. Letter from G. E. Edison (NRC) to D. A. Christian (VEPCO), "Surry Units 1 and 2 -

Issuance of Amendments RE: Alternative Source Term," dated March 8, 2002 (pg 9).

9. Letter from V. Nerses (NRC) to D. A. Christian (Dominion Nuclear Connecticut),

Millstone Power Station, Unit No. 2 - Issuance of Amendment RE: Selective Implementation of Alternate Source Term," dated September 20, 2004 (pg 13)

10. Letter from S. P. Lingam (NRC to D. A. Christian (VEPCO), "North Anna Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Technical Specification Changes Per Generic Safety Issue (GSI) 191 (TAC NOS. MD3197 and MD3198)," dated March 13, 2007. [ADAMS Accession No. ML070720043]

11. Letter from V. Nerses (NRC) to D. A. Christian (DNC), "Millstone Power Station, Unit No. 3 - Issuance of Amendment RE: Alternate Source Term (TAC NO.

MC3333)," dated September 15, 2006. [ADAMS Accession No. ML061990135].

12. Letter from W. R. Matthews (DNC) to NRC Document Control Desk, "Response to Request for Additional Information Regarding Proposed Technical Specification Changes for Implementation of Alternate Source Term," dated March 23, 2005

[ADAMS Accession No. ML050950215]

Serial No. 11-025B ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS LAR 244, ATTACHMENT 1 CHANGED PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 11-025A Attachment 1 Page 6 of 52 TABLE 2-1 List of Technical Specification Sections Affected by Adoption of TSTF-51 Technical Specification Sections Affected by Adoption of TSTF-51 3.3.6 Containment Purge and Vent - Note applicable to Condition C Isolation Instrumentation - Table 3.3.6-1, Footnote (a) 3.3.7 Control Room Post-Accident - Condition D and Required Action 0.

Recirculation (CRPAR) System - Table 3.3.7-1, Footnote (a)

Actuation Instrumentation

- Proposed new LCO re uirement 3.7.10 Control Room Post-Accident - Applicability Statement Recirculation (CRPAR) System - Condition D and Required Action D.2

- Condition E and Required Action E.1 3.7.11 Control Room Air Conditioning - Applicability Statement (CRAC) Alternate Cooling - Condition C and Required Action C.2 System - Condition D and Required Action D.1

- Applicability Statement 3.8.2 AC Sources - Shutdown - Required Action A.2.1

- Required Action B.1 3.8.5 DC Sources - Shutdown - Applicability Statement

- Required Action A.1 3.8.8 Inverters - Shutdown - Applicability Statement

- Required Action A.1 Shutdown- Applicability Statement 3.8.10 Distribution Systems - Shutdown- Requic Action tionA

-Required A.2.1

- Applicability Statement 3.9.6 Containment Penetrations - Requic tionA I- Required Action A.1

Serial No. 11-025A Attachment 1 Page 8 of 52 2.2.2. Modify TS 3.7.10, Control Room Post-Accident Recirculation (CRPAR)

System DEK is proposing to modify TS 3.7.10 consistent with adoption of TSTF-448. The proposed changes to TS 3.7.10 are described below and shown in Attachment 2.

1. The existing NOTE in LCO 3.7.10 would be modified to change the current wording from; "The control room boundary may be opened intermittently under administrative controf' to; "The control room envelope (CRE) boundary may be opened intermittently under administrative control."

2. A new requirement would be added to LCO 3.7.10 which states; "and the CRE shall be isolated during movement of recently irradiated fuel assemblies."

3. The current APPLICABILITY for LCO 3.7.10 is Modes 1-6, and during movement of irradiated fuel assemblies. The APPLICABILITY would be changed to Modes 1-4, and during movement of recently irradiated fuel assemblies. Consistent with this change, TS 3.7.10, Condition D and Condition E are also modified by removing Mode 5 and 6 applicability.

4. TS 3.7.10, Condition A wording would be modified from "One CRPAR Train inoperable"to "One CRPAR Train inoperable for reasons other than Condition B."

5. TS 3.7.10, Condition B currently provides a Required Action when two CRPAR trains are inoperable due to an inoperable CRE boundary in Modes 1-4. The current Required Action B.1 is to restore the CRE boundary to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Condition B and its associated Required Action B.1 and Completion Time would be replaced with a new Condition B. The new Condition B would provide required actions and completion times when one or more CRPAR trains are inoperable due to an inoperable CRE boundary in Modes 1-4.

The new Condition B would include three new Required Actions when one or more CRPAR trains are inoperable due to an inoperable CRE boundary in Modes 1-4.

The three new Required Actions (B.1, B.2, and B.3) would require; (B.1) immediate initiation of action to implement mitigating actions; (B.2) verification that mitigating actions ensure CRE occupant exposures to radiological, chemical and smoke hazards will not exceed limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and; (B.3) restoration of the CRE boundary to operable status within 90 days.

6. TS 3.7.10, Condition E currently requires immediate suspension of movement of irradiated fuel assemblies when two CRPAR trains are inoperable in Modes 5 and 6 and during movement of irradiated fuel assemblies. The current Condition E would be modified by deleting Mode 5 and 6 applicability (see item 3 above) and adding the word "recently" so that the resulting Condition would read; "Two CRPAR trains inoperable during movement of recently irradiated fuel assemblies." In addition, Condition E would be expanded to include situations where the Required Actions and associated Completion Times of Condition B are not met during movement of recently irradiated fuel assemblies. The Required Action of Condition E would be

Serial No. 11-025A Attachment 1 Page 20 of 52 TABLE 3-2 Currently Approved Design-Basis Accident Calculated Radiological Consequences rem TEDE Design-Basis Accident EAB LPZ Control Room MSLB, Pre-existing iodine spike 0.030 0.01 0.70 Dose acceptance criteria 25 25 5 MSLB, Accident-initiated iodine spike 0.06 0.02 2.60 Dose acceptance criteria 2.5 2.5 5 Locked Rotor Accident 0.40 0.06 3.90 Dose acceptance criteria 2.5 2.5 5 Control Rod Ejection Accident 0.40 0.09 4.54 Dose acceptance criteria 6.3 6.3 5 SGTR, Pre-existing spiking 0.50 0.10 1.90 Dose acceptance criteria 25 25 5 SGTR, Accident-initiated spiking 0.80 0.20 2.80 Dose acceptance criteria 2.5 2.5 5 LBLOCA, total 0.52 0.09 4.95 Dose acceptance criteria 25 25 5 FHA 0.90 0.15 4.0 Dose acceptance criteria 6.3 6.3 5 WGDT Rupture 0.10 0.02 0.80 Dose acceptance criteria 0.25 0.25 5 VCT Rupture 0.10 0.01 0.40 Dose acceptance criteria 0.25 0.25 5

Serial No. 11-025A Attachment 1 Page 21 of 52 TABLE 3-3 Proposed New Design-Basis Accident Calculated Radiological Consequences rem TEDE Design-Basis Accident EAB LPZ Control Room MSLB, Pre-existing iodine spike 0.1 0.1 4.7 Dose acceptance criteria 25 25 5 MSLB, Accident-initiated iodine spike 0.1 0.1 4.2 Dose acceptance criteria 2.5 2.5 5 Locked Rotor Accident 0.3 0.2 4.7 Dose acceptance criteria 2.5 2.5 5 Control Rod Ejection Accident Containment Release Pathway 0.2 0.1 0.8 Dose acceptance criteria 6.3 6.3 5 Control Rod Ejection Accident Secondary Side Release Pathway 0.1 0.1 0.5 Dose acceptance criteria 6.3 6.3 5 SGTR, Pre-existing spiking 0.3 0.1 3.9 Dose acceptance criteria 25 25 5 SGTR, Accident-initiated spiking 0.2 0.1 1.1 Dose acceptance criteria 2.5 2.5 5 LBLOCA, total 0.5 0.5 4.1 Dose acceptance criteria 25 25 5 FHA 0.6 0.2 4.3 Dose acceptance criteria 6.3 6.3 5 WGDT Rupture 0.1 0.1 0.4 Dose acceptance criteria(') 0.25 0.25 5 VCT Rupture 0.1 0.1 0.6 Dose acceptance criteria(1 ) 0.25 0.25 5 (1) EorAth-_te.WG DT _Tn e t_ B and LPZ dose ace.ta.idatamr-smalU fspecified in 10 CFR 50.67 as JLcusse~d in the Safety Evaluatio.n frAm D.Q l eene l. n ro roo for these d st _ paed with ems f_ stn _A_PoG-ls 1___ -s e nce..1.)_0

Serial No. 11-025A Attachment 1 Page 22 of 52

4.0 TECHNICAL ANALYSIS

The proposed changes to the KPS TS are discussed and evaluated below. Section 4.1 addresses the proposed changes associated with the revised RAA. These are changes that result from the revised RAA, exclusive of those required for establishing a CREH program. Section 4.2 addresses the proposed TS changes associated with implementation of the CREH program.

4.1 Technical Specification Changes Proposed due to Revised Radiological Accident Analysis Based on revised inputs, assumptions and analysis, DEK is requesting changes to the TS to accommodate a proposed revision to the RAA. In conjunction with the changes necessary to accommodate the revised RAA, DEK is proposing changes that would adopt the following TSTFs:

TSTF-312, Revision 1, "Administratively Control Containment Penetrations."

TSTF-51, Revision 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations."

4.1.1 Revise Specific Activity Limits TS 3.4.16, "RCS Specific Activity" The proposed amendment would reduce the current Reactor Coolant System (RCS) specific activity limits in TS 3.4.16 to values that are consistent with the revised RAA.

The revised RAA assumes a DEI limit of < 0.1 pCi/gm, a pre-existing iodine spike limit of < 10 pCi/gm DEI, and a DEX limit of

16.4 pCi/gram.

TS 3.4.16 provides limits for the allowable concentration level of radionuclide's in the reactor coolant. The reactor coolant specific activity limits are established to minimize the dose consequences in the event of a main steam line break (MSLB) or steam generator tube rupture (SGTR) accident. TS 3.4.16 contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in Regulatory Guide (RG) 1.1831.

The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67, "Accident source term."

Doses to control room operators must be limited per 10 CFR 50, Appendix A, General Design Criteria (GDC) 19, "Control Room." The limits on reactor coolant specific activity eranlofth d ~c1~d segiedn OCER X6LKsAsqusd in.týaa._fet fratins(as~hn1 Evaluation for Amendment 166 (reference 7. Control room dose for these accidents JS compared with the limits ic GDC 19 (reference- L pjiable standards inRG t.1 (ref18er

Serial No. 11-025A Attachment 1 Page 37 of 52

2. A new requirement would be added to LCO 3.7.10 which states; "and the CRE shall be isolated during movement of recently irradiated fuel assemblies."

Currently, LCO 3.7.10 requires that two CRPAR trains shall be operable during Modes 1-6 and during movement of irradiated fuel assemblies within containment.

To ensure control room doses following a FHA remain below applicable acceptance criteria, the revised RAA in Attachment 4 assumes the control room is isolated at the initiation of a FHA. Pre-isolation of the control room minimizes infiltration of radioactive materials into the CRE prior to initiation of the CRPAR system in the emergency mode and ensures dose to CRE occupant's remains within applicable limits.

3. The current APPLICABILITY for LCO 3.7.10 is Modes 1-6, and during movement of irradiated fuel assemblies. The APPLICABILITY for LCO 3.7.10 would be changed to Modes 1-4, and during movement of recently irradiated fuel assemblies. Consistent with this change, TS 3.7.10 Condition D and Condition E are also modified by removing Mode 5 and 6 applicability.

Currently TS 3.7.10 requires the CRPAR system to be operable in MODE 5 and 6.

The current TS 3.7.10 Bases state that in Modes 5 and 6, the CRPAR system must be operable to; 1) control operator exposure during and following a DBA, and 2) to cope with the release from a rupture of an inside waste gas tank. DEK proposes to delete the Mode 5 and 6 applicability of TS 3.7.10 and revise the TS Bases consistent with this proposed change.

The KPS Waste Gas Decay Tanks (WGDT) (USAR Chapter 11.1) (reference 19) and Volume Control Tank (VCT) (USAR Chapter 9.2) are located inside the auxiliary building, where radioactive gases are collected and filtered prior to release.

The WGDT failure and VCT rupture (atmospheric release) radiological analyses are being revised to reflect revised X/Q values as discussed in Attachment 4 of this application. For the WGDT and VCT rupture accidents, the EAB and LPZ dose acceptance criteria are small fractions (less than 1 percent) of the dose criteria suecified in 10 CFR 50.67 as discussed in theSafety Evaluation for Amendment 166 (reference 17). Control room dose for these accidents is compared with the limits in GDC 19 (reference 15) and applicable standards in RG 1.183. The revised WGDT and VCT analyses demonstrate acceptable dose to control room operators without credit for the control room emergency ventilation filtration or CRE isolation. These analyses also demonstrate acceptable dose at the EAB and LPZ.

The only other design basis radiological accident postulated to occur when the plant is in Modes 5 and 6 is the Fuel Handling Accident. A FHA is postulated to occur only during movement of irradiated fuel and TS 3.7.10 will continue to be applicable during the movement of recently irradiated fuel, as discussed in item 2 above and in Section 4.1.2.

Serial No. 11-025A Attachment 1 Page 39 of 52 Condition B which states "OR Control Room Vent Radiation Monitor inoperable," would be deleted. In addition, SR 3.3.7.1, SR 3.3.7.2, and 3.3.7.4 would be deleted since these SRs are solely applicable to the Control Room Vent Radiation Monitor (Function 2 in Table 3.3.7-1). These changes are based on the revised RAA in Attachment 4.

The control room ventilation radiation monitor consists of a single radiation monitor (R-23) located on the common discharge of the outlet of the air conditioning fan units. A high radiation signal from the detector will initiate both trains of the CRPAR system.

The control room operator can also start the CRPAR fans by manual switches in the control room. The CRPAR system is also actuated by a safety injection signal. A detailed discussion concerning the CRPAR system and R-23 is provided in Section 3.1.1 of this Attachment.

DEK is proposing to delete R-23 as a required channel for CRPAR initiation. The revised RAA in Attachment 4 does not rely on or credit radiation monitor R-23 to isolate the control room during radiological events. In Section 4.2.2 above, a new requirement is being added to TS 3.7.10 which would require the control room to be isolated prior to I movement of recently irradiated fuel. This new requirement is consistent with the revised RAA, which assumes the control room is isolated prior to moving recently irradiated fuel.

Thus, reliance on R-23 to isolate the control room in the event of a FHA is no longer necessary. For other DBAs, the revised RAA assumes reasonable operator actions or a safety injection signal will perform the necessary control room isolation function and maintain doses within acceptable limits.

Specifically, in accordance with the revised RAA, DEK is proposing two manual actions to ensure post-accident control room dose is maintained within limits. The revised RAA indicates that manual actions are required to limit consequences of the FHA and LRA events. The proposed manual actions are as follows:

1. The revised RAA credits manual operator action to isolate the control room within one hour after initiation of a Locked Rotor Accident (LRA). This manual action is required to compensate for the proposed TS changes that would discontinue credit for control room auto-isolation using a high radiation signal from R-23.

2. The revised RAA assumes the CRE is isolated prior to movement of recently irradiated fuel assemblies (per new rNe uirement added to TS LCO 3.7.10). In addition, the revised RAA credits manual initiation of the Control Room Post Accident Recirculation (CRPAR) system within 20 minutes of occurrence of a FHA.

An evaluation of the acceptability of the proposed new manual actions is provided in Attachment 5.

The equipment necessary to initiate control room isolation and starting of the CRPAR trains is tested monthly as part of SR 3.7.10.1, which requires operation of each CRPAR train for greater than or equal to 15 minutes on a 31-day frequency. Because

Serial No. 11-025A Attachment 1 Page 41 of 52 4.3 Conclusions The proposed amendment would revise the KPS OL, TS, and USAR to incorporate changes resulting from a revised radiological accident analysis (RAA) and changes to implement a commitment relating to Control Room Envelope Habitability.

A revised RAA is included in Attachment 4. The RAA has been performed in accordance with RG 1.183, and concludes the plant meets the dose consequences acceptance criteria of 10 CFR 50.67.2 10 CFR 50.67(b)(2) states that the analysis must demonstrate with reasonable assurance that:

" An individuallocated at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 roentgen equivalent man (rem) total effective dose equivalent (TEDE).

" An individual located at any point on the outer boundary of the low-population zone (LPZ), who is exposed to the radioactivecloud resulting from the postulatedfission product release during the entire period of its passage, would not receive a radiationdose in excess of 25 rem TEDE.

" Adequate radiationprotection is provided to permit access to and occupancy of the control room (CR) under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

The revised RAA contains revised assumptions and requirements for meeting the acceptance criteria of 10 CFR 50.67 described above. The results of the revised RAA provide reasonable assurance of meeting these acceptance criteria. Therefore, the changes proposed as a result of the revised RAA are considered acceptable.

In addition, in accordance with commitments made in response to GL 2003-01, DEK is adopting TSTF-448 by incorporating applicable changes into the KPS TS. This requires adoption of a new OL condition. These changes are also based on and consistent with the revised RAA. Adoption of other TSTFs has been proposed that support the control room envelope habitability requirements of TSTF-448 and the revised RAA. These proposed changes have been evaluated above and are also considered acceptable.

2 Fq~e~GT VC-wptureBkientsjh -P-Q t -i-e-~~-m a~c~ti~o~n~s_(l~ssth~a.__.perqe~t)of.theosecteridaspcjfl ed n-10 CFR 500a discussed in the Safety E aliation for Amendment 166 (reference 17). Control room dose for these accidents is comDared with s GD rl ni 1 icable standards in RG 1,183

Serial No. 11-025A Attachment 1 Page 51 of 52

7.0 REFERENCES

1. NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000.

2. NRC Generic Letter 2003-01, "Control Room Habitability," dated June 12, 2003.

3. Letter from Craig W. Lambert (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01: Control Room Habitability - Supplemental Response," dated April 1, 2005. [ADAMS Accession No. ML050970303]

4. Letter from G. T. Bischoff (DEK) to Document Control Desk (NRC), "License Amendment Request 210

Subject:

Technical Specification Modifications Regarding Control Room Envelope Habitability," dated September 14, 2007.

[ADAMS Accession No. ML072620144]

5. TSTF-448, Revision 3, "Control Room Habitability," dated August 8, 2006.

(ADAMS Accession No. ML062210095). Including letter from TSTF to Document Control Desk (NRC), "Corrected Pages for TSTF-448, Revision 3, Control Room Habitability," December 29, 2006. [ADAMS Accession No. ML063630467]

6. Not Used.

7. TSTF-312, Revision 1, "Administratively Control Containment Penetrations,"

dated July 16, 1999. [ADAMS Accession No. ML040620147]

8. TSTF-490, Revision 1, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec," dated March 14, 2011. [ADAMS Accession No. ML110730473]

9. Not Used.

10. TSTF-51, Revision 2, "Revise Containment Requirements during Handling of Irradiated Fuel and Core Operations," dated July 31, 2003. [ADAMS Accession No. ML040400343]

11. Not Used

12. Not Used.

13. Letter from D. V. Pickett (NRC) to J. A. Spina (Constellation Energy), "Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2 - Amendment RE: Control Room Habitability (TAC Nos. MD 5928 and MD5929)," dated July 29, 2008. [ADAMS Accession No. ML082030173]

14. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

15. 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants,"

Criterion 19, "Control Room (GDC 19)".

16. Memorandum from C. Craig Harbuck (NRC) to Timothy J. Kobetz (NRC), "Model Application for TSTF-448, Control Room Habitability, Revision 3," dated February 2, 2007. [ADAMS Accession No. ML070330657]

Serial No. 11-025B ENCLOSURE 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS LAR 244, ATTACHMENT 2 CHANGED PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

AND CRPAR System 3.7.10 The CRE shall be isolated during movement 3.7 PLANT SYSTEM of recently irradiated fuel assemblies.

3.7.10 Control Ro m Post Accident Recirculation (CRPAR) System Ienvelope (CRE)

LCO 3.7.10 o CRPAR trai s shall be OPERABLE.

I Mi -I-r--

The control room boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1,2,3/I During move ent of irradiated fuel assemblies.

Sand -- r-e-centliy ACTIONS for reasons other CONDITION REQUIRED ACTION COMPLETION TIME than Condition B F\ L I I A. One CR R train A. 1 Restore CRPAR train to 7 days Insert 2

)

Sinoperable. OPERABLE status.

j B. Two CRPAR trains B. 1 Restore control room ....Uhec-b -*

inoperable due to boundary to OBEJRA -"-"

inoperable control room *_... .'

boundary in MO E4-**

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Kewaunee Power Station 3.7.10-1 Amendment No. 207 02/02/2011

CRPAR System 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Place OPERABLE CRPAR Immediately associated Completion train in emergency mode.

Time of Condition A not me ingr ORR [rentl-y during movement of irradiated fuel D.2 Suspend movement o Immediately assemblies, irradiated fuel assemblies.

[recently E. Two CRPAR trains E.1 Suspend movement of Immediately inoperable in, L* irradiated fuel assemb s.

5 ,during movement of,*adiated fuel assem-bfi receýntly]recently F. Two CRPAR trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1,

, 3, or 4 for reasons ther than Condition B.

Insert 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CRPAR train for > 15 minutes. 31 days S R 3.7.10.2 Perform required CRPAR filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP). with VFTP SR 3.7.10.3 Verify each CRPAR train actuates on an actual or 18 months simulated actuation signal.

Perform required CRE unfiltered air inleakage In accordance with testing in accordance with CRE Habitability CRE Habitability Program Program Kewaunee Power Station 3.7.10-2 Amendment No. 207 02/02/2011

Insert 2:

B. One or more CRPAR BA Initiate action to implement Immediately trains inoperable due to an mitigating actions.

inoperable CRE boundary in Modes 1, 2, 3, or 4. AND B.2 Verify mitigating actions ensure 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to OPERABLE status. 90 days Insert 3:

OR Required Actions and associated Completion Times of Condition B not met during movement of recently irradiated fuel assemblies.

OR CRE not isolated during movement of recently irradiated fuel assemblies.

Serial No. 11-025B ENCLOSURE 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS LAR 244, ATTACHMENT 3 CHANGED PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

CRPAR System B 3.7.10 B 3.7 PLANT SYSTEMS B 3.7.10 Control Room Post Accident Recirculation (CRPAR) System BASES BACKGROUND The CRPAR System provides a protected environment from which eperater-s-_c__uints can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or toxic gs.s.m.ok.

The CRPAR System consists of two independent, redundant trains that recirculate and filter the air in the control room envev QMRE and a GC.E boundar__thatalimitsth~einle~a.ageof unfiltered.,ai.eutsudea.i. Each

-P.A-R train consists of a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Common ductwork, valves or dampers, dooors, bAariers, and instrumentation also form part of the system.

The CRE is the area within the confines of the CRE boundar¥y that cnt.ains the s*.aces that control romom.c.c.pants.inhabit to control the unit durng, normal and accident conditions.Jbi reaaencornpasses the control room, and other non-critical areas to which fre-quent pen nel access or continuous occuancy is not necessary in the_ent of an cisdent. The CRE isrpmtetegd driianomlO 'pRurnaibmi yanL and accident conditions. The CRE bouda yis t e combination ofwals floor, roof. ducting, doors. penetrations and equipment that physically form_.heQ.CR-E. The _PERA BJLYJYTothe-_-REnEdar m-ust .bse rnmainlinnonsuetiuhathnle akae oL unfilteedW air into the CRE will not exceed the inleakage assumed in the licen.inqg basis analysis of de= sjg-n-basis accidenAt(DBA_ consequences to CRE occupants. The CRE and -d fnt.s-b.

i.ei*y

.n-th .n-treC LBoofn Envelope Hia-tabi[ity The CRPAR System is an emergency system, which is normally in the standby mode of operation. The CRPAR System is part of the Control Room Air Conditioning (CRAC) System. During normal unit operation, the CRAC System provides cooling of recirculated and fresh air to ventilate the control room. Upon receipt of the actuating signal(s), normal outside air intake supply to the co*trF!reemCRE is isolated, both CRPAR fans are started, the flow path through the Emergency Filtration System is opened, and a portion of the return air volume is filtered to remove airborne contaminants and airborne radioactivity, then mixed with the recirculated return air. The prefilters remove any large particles in the air to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The neutral pressure envelope design of the cAro!4 roemCRE minimizes infiltration of unfiltered air from the surrounding areas of the building. The CRPAR System fans are started upon receipt of a safety injection signal Kewaunee Power Station B 3.7.10-1 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 or manual initiation through switches in the control room hightadet~on sign"l deted by th rFadiatio n mnitor R R-n2cd 23 in thoemin h

control room emergency zone (CREZ) supply duct.

The CRPAR System operation in maintaining a habitable environment in.

the CRE centrol room habitable is discussed in the USAR, Section 9.6.4 (Ref. 1).

Redundant supply and recirculation trains provide the required filtration should an excessive pressure drop develop across the other filter train.

Normally open isolation dampers of the CRAC Alternate Cooling System provide double/redundant isolation capability so that the failure of one damper to shut will not result in a breach of control room ventilation isolation. The CRPAR System is designed in accordance with Seismic Category I requirements.

Isolation of the CRE during movement of recently irradiated fuel a*msJe~s (i.*.e._, fuel t~h~at has__ocqcuped -~art of a._rit~icalreact~ cr.e within he Dreyviousý_ 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months an-d mo1a.l~tuatiQn ofthe RAwi tht r 20 minutes after a fuel handling accident is the primary. means to ensure CREh_ abtaabiit¥in-the_ event of a fueI h lingtaccident whilepha n!dlring recen*ly irradiated fuel. Actuati.Qnofhe_.E.PA _.Stem a.dCRE iso__latkon-reperrmedjbyaSLuacttion siinal, either autoMa_[alIy__or manually initiated. Calculated doses to _RE occuDants from a volume tank r r__tu r wast - ca ytank pree surc1etsmal

........ t~B ......................

u ..........

M~tltdacdns Kewaunee Power Station B 3.7.10-2 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 BASES BACKGROUND (continued)

The CRPAR System is designed to maintain a ha**table environment in the CRE centrleroomenvironment for 30 days of continuous occupancy after a Design Basis Accident (DBA) without exceeding a 5 rem total effective dose equivalent (TEDE).

APPLICABLE The CRPAR System components are arranged in redundant, safety SAFETY related ventilation trains. The location of components and ducting within ANALYSES the control room ,-vel.pe CRE ensures an adequate supply of filtered air to all areas requiring access. The CRPAR System provides airborne radiological protection for... the ...centrel

. .. -1 room

-... operatersCRF_

eperat* o~ccupaints as E ccu antr~t, a demonstrated by the control rOom acidtQ.B L... p=aM dose analyses for the most limiting design basis loss of coolant accidentT fission product release presented in the USAR, Chapter 14 (Ref. 2).

The CRPAR System also provides protection frmrn smoke-and-hazardous chemicalst the RE occupants. The analysis of hazardous chemical releases demonstrates that.the toxicity limitsareot exceeded h the CRE following-a-hazard.ujs bhemca rgelearse Reff6.-The-evaluation of asmokecha_ len__ -alo demonstrates that it will not resut in the inabilit of the CRE occupants to control the reactor either from the control room or from the remote shutdowonpaneL(,~eLi .for tho control reoom peraters in rcotopossibihityoefa to fire inthc coentro-l room, as doccribed i RefeF*Re 4.

The worst case single active failure of a component of the CRPAR System, assuming a loss of offsite power, does not impair the ability of the system to perform its design function.

The CRPAR System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant CRPAR trains are required to be OPERABLE to ensure that at least one is available assuming-ifa single active failure disables the other train. Total system failuresuch .as frqa.

loss of both ventilation trains or from an inoperable CRE boundary, could result in exceeding a dose of 5 rem TEDE to the control room operator in the event of a large radioactive release.

The-Each CRPAR System-train is considered OPERABLE when the individual components necessary to limit eperatGe-CRE occupant exposure are OPERABLE-R both trFais. A CRPAR train is OPERABLE when the associated:

a. Fan is OPERABLE;

b. HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions; and Kewaunee Power Station B 3.7.10-3 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10

c. Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In addition, the CRAC fan in the same train must be OPERABLE when the CRPAR train is required. Furthormore, the control room bn, .dary must be maintaincd, including the integrity of the walls, floors,ciin, ductWOrk, and accocc dooers.

IneorLder.or the .RPAR trains to be cns-idered OPERALE-_he ORE...CRE bounnda must be maintained such that the ORE at osefrm

.1 rge Q~Qtdve-ele aae-daea noat mx.eed-1h~e Qal~Bto d dse in e licensing bisLFqons,.uence a fr DBan thatjdCRE Qc0.a0 n .

are protected from hazardous chemicals and smoke.

The CRE is also reoui-red.to be isolated duidng movement of recently irradiated fuel assemblj s. Thefuelhandlinqaccident analyssya.ssum a the control room is isolated at the initiation of the accident. Pre-isolation of the control room minimizes infiltration of radioactive materials into the dose troCREoccupiantre otoin i t ths dose to ORE o~ccupp nts' remains within applicable limit5.

Kewaunee Power Station B 3.7.10-4 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 BASES LCO (continued)

The LCO is modified by a Note allowing the ceneRtFe rmCeRE boundary to be opened intermittently under administrative controls. This Note only app*lies to openings in the cRE.boub r-narthalca-nbe-rajdlretored.to the d es iqn =on.iio* au doj a,.rjit~es_.[fo~or.p. u g s_*ac, access Dan.in..* For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls -hsgu.r* prn.,*-aedumzed an& consist of stationing a dedicated individual at the opening who is in continuous communication with the oDerators in the CREGGcontrl4roe0.

This individual will have a method to rapidly close the opening and restore the CRE boundary to a condition equivalent to the design co;nadatiwhen a need for centre roolmCRE isolation is indicated.

APPLICABILITY In MODES 1, 2, 3, and 4, 5, *,4 6, and during movement of recentl irradiated fuel assemblies, the CRPAR System must be OPERABLE to ensure that the ORE will, remain habitable Gcontro operator exposure during and following a DBA.

In MGDE 5 or 6, the CRPAR System is roquired toG cpo With tho roloaso from tho ruzpturie of an incido waste gas tank.

During movement of recently irradiated fuel assemblies, the CRPAR System must be OPERABLE to cope with the release from a fuel handling accident involving handling ofrecentl irradiated fuel. The CRPAR is only required to be OPERABLE duiLng fuel handlinginvolvintg handling of recently irradiated fuel (i.e., fuel that has occupied part of a criticraitvc.L reacto...r ewi inthejt 3Jhur .*..rad ioa*_tive deca9y ACTIONS A.1 When one CRPAR train is inoperable, for reasons other than an inoperable CRE boundary, action must be taken to restore OPERABLE status within 7 days. In this condition, the remaining OPERABLE CRPAR train is adequate to perform the centre! FemCR EGunt protection function. However, the overall reliability is reduced because a si4ge aGt-ive-failure in the OPERABLE CRPAR train could result in loss of CRPAR function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

B.-. B. and B-3 Kewaunee Power Station B 3.7.10-5 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 If the unfiltered inleakagoe of potentially conta minata9*dL past. tho-RE bondr iiýand into the CRE can result in CRE occupant radiological dose e.erI[tlhan the (calculated dose, of the licensing basis analyses of DBA

.¢_o_._n~~~~~~~~~~~sefl~~~~

Q ccuantsfro protctio d.*.*......up..

ofORE uen...c.*s*=*

-.-.[ . .I .-..-

hazrdos ~hmiqls r ...i.r...

e .....

or t-e the smnoke, ...

ORE b~oundoarv is ioperable. Ac ions mustibe taken to restore an OPERABLE ORE boundary within 90_day During the period that the CRE boundary is considered inoperable. acti Mushtbe initiated toim.plement miti ga Uac-t. Atheffet on

.C.R*` o.Qcqup[a ntsJrornt~he p!tent.i#````*al ar=ds***[ar=,ad,,o,,],g~i,,oLo,:,r,, *h==e-mi~c,,.,

.CvOt..-. .. a.. hall or .... n.. e.............

fr m sm ....ok e.

.......... tS,,

...... ,,sn.. ..... n..w i..thin 24 h,..,....our.s to verify_that in the event of a DBA, the miticating actions will ensure that ORE occu a ntr ..dio pqj calexUres will not ceed the calculated dose ftheicns.ng bassis ana yvses of DBA conse.uences. and that ORE occupants arprotected from hazardous chemicals _ansd smoke. These

atin__&ction sLeactions that are taken to offs_ the consequences of.-thes*s*ho-uld eorep-anbned for

_~~~~~~~~. .nprboR b.................

isintentional or unintentional. The 24-hour Completion Time is reasonable basaeeLon the Low probabiltoLDfA oqcurrinqdurin.this.

time_.eriodI,.and the use of miti..nq_.a0..ti.os .he9.:-day omptio.

Tmesrea~s~o.able ba~e~d~onthe determination that the migatingacftions wMil_ e_]nsure protection of CRE .Qcupants within l.i.r~tng~bArbaiithaLQ__R__O__gg~lants analyzed will have to imlimits while le~men..t

.ps .s th~at..maya dve.rseyafcbekiltAc-nrLh meia~

ractorLand maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to If the coAntro-l room boun~dary isioeal nMODE= 1, 2, 3, or 1-,-the CRPAR trains cannot pe-"'-m their nt*cnded functions. Action must be taken to restore an OPERABLE Got*r*l r... boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During the peried that the control room boundary is inoperable, appropriate comnpenSatorY measures (consistent with the intent ot GIDC 19) should be utilized to protcct control room operators fromn potential hazards su-ch as radioactive contamination, toxic chomicals, smoke, temperature and relative humidity,and physi*al Se.U.i.. .

Preplarmed .measuresshould be available to address these concerns for Kewaunee Power Station .... B 3.7.10-6 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 BASES ACTIONS (continued) intentional n~d uitntoa entry into the condition. The 24 hourf Complction Time is reasnable based OnRthe lo)W probability of a D6BA The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a typically reasonable time to diagnose, plan and possibly rcpair, and test moest problemns with the control room beundary.

C.1 and C.2 In MODE 1, 2, 3, or 4, if the inoperable CRPAR train or ceptrel r-em the C__ boundary cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

D.1 and D.2 In MOADE 0 or 6, or dD-uring movement of rLce ijrradiated fuel assemblies, if the inoperable CRPAR train cannot be restored to OPERABLE status within the required Completion Time, action must be taken to immediately place the OPERABLE CRPAR train in the emergency mode. This action ensures that the remaining train is OPERABLE and that any active failure would be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the eontFol roomMBE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

E._1 nQMODE 5 or 6, or dQuring movement of reanty irradiated fuel assemblies, with two CRPAR trains inoperable, or with one or more CRPAR trains inoperable due to an inoperable CRE boundary. action must be taken immediately to suspend activities that could result in a release of radioactivity that might eRteF-require isolaton of the CREIentrel F-em. This places the unit in a condition that minimizes the accident risk.

This does not preclude the movement of fuel to a safe position.

Kewaunee Power Station B 3.7.10-7 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 BASES ACTIONS (continued)

F.1 If both CRPAR trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable cton-trI FeemCRE boundary (i.e., Condition B),

the CRPAR System may not be capable of performing the intended function and the unit is in a condition outside the accident analyses.

Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month provides an adequate check of this system. Operating each CRPAR train for > 15 minutes demonstrates the function of the system. The 31 day Frequency is based on the reliability of the equipment and the two train red undancy-avaiwabiity.

SR 3.7.10.2 This SR verifies that the required CRPAR testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.10.3 This SR verifies that each CRPAR train starts and operates on an actual or simulated actuation (high radiation and safety injection) signal. The frequency of 18 months is based on industry operating experience and is consistent with the tyPiCal refueling cyclle. Operatin c.xpc. ie. has shown that theSo components usually pass the gurpolillancc whn performoed at the 18 month FrFequency. Theroforo, the FroequcncY was co-nclud-edIto be ac.optab-lo f-rom a reliability standpei*t.-

This SR verifies the OPERABILITY of the CRE boundar_ by testing fo )r unfiltered aiTr nleakage past the CRE.ioundary and into the._.RE. The e.iLs _thbe testinare s ecifiedin the__C.ointrolRomEnvelope Habitability Program.

Kewaunee Power-Station B 3.7.10-8 Amendment No. 207 02/02/2011

CRPAR System B 3.7.10 The .RE is considered habitable when the radiological dose.to C.E..

occupants calculated in the licensing basis analyses of DBA conseqg-ences is no more than 5 rem TEDE and the CRE occupantsr.e prote*_c.ted .cma.L from ha.zardousheIm sd so_*IRe...Ib~i yerifies that tfheunjtfite.red air inleakage into the CRE is no greater.than theflow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleaka e is greater than the assumed f*_w rate_ Condition B must be entered. Reguired Action B.3 allows time torest.ore theCE

Z oOEAL status provided mitigating actions can ensure that the CRE remains within the licenssing basis habitabiilty. imits for the

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exceptions. NEI 99-03, Section 8.4 and Appendix F (Ref. 55). These

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beadacorraensa measures su to restreQPERABILITY (_Ref.3.

Q.p.tions for restoring--the CREbounda._toPERABE status include Jangige.[ licensi cng basis BA c analsi r_.paring the testnm not.beQen ly s .ayr to establtýish thatt R o ryt ha. been restored to OPERABLE statu.,

REFERENCES 1. USAR, Section 9.6.4.

2. USAR, Chapter 14.

3. Letter from Eric J. Leeds RC) to James W. Davis NEI "NEI Draft White Paper Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability," dated 4~rA-[Q .0AlQA-MSAc~cessian No ML _4O3QN4]

4. Regulatory Guide 1.196, Rev. 2.

5.* NEI.99-03"iCo.ntrol Room Habitability Assessment Guidance." June

..... 200-O

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~rm.Q&tenhar-dt tJR.Q-. lSatmjnt.hl fevunme'-s Updated Control Room Habitability Evaluation Report to Address Concerns Over Control Room Ventilation." dated February_- 1989.

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Kewaunee Power Station B 3.7.10-9 Amendment No. 207 02/02/2011

Serial No. 11-025B ENCLOSURE 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS LAR 244, ATTACHMENT 4 CHANGED PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial Number 11-025A Attachment 4 Page 8 of 191

" Main Steam Line Break (MSLB) Accident,

Locked Rotor Accident (LRA) and

Rod Control Cluster Assembly (RCCA) Ejection Accident (REA).

The Waste Gas Decay Tank (WGDT) failure and Volume Control Tank (VCT) rupture (Atmospheric Release) radiological analyses are also being updated to reflect revised X/Q values determined in Section 3.1 of this application. Both analyses demonstrate acceptable dose to control room operators without credit of control room emergency ventilation or isolation as well as acceptable results to the EAB and LPZ..

The proposed licensing basis and plant operational changes are discussed in Section 2.0. These changes require appropriate changes to the KPS Technical Specifications, which are also described in Section 2.0 of this report. The key changes considered are listed below:

a. Revise definition of Dose Equivalent 1-131 in Section 1.1 of the Technical Specifications to reference Federal Guidance Report No. 11 (Reference 15) as the source of thyroid committed dose equivalent (CDE) dose conversion factors.

b. Revise Technical Specification 3.4.16, to decrease the RCS activity limits to 0.1 pCi/gm DE 1-131 and 16.4 pCi/gm DE Xe-1 33.

c. Revise Technical Specification 3.4.16, to decrease the pre-existing iodine spike limit from 20 pCi/gm DE 1-131 to 10 pCi/gm DE 1-131.

d. Revise Technical Specification 3.7.16, to decrease the SG bulk liquid concentration limit from 0.1 pCi/gm to 0.05 pCi/gm DE 1-131.

e. Revise Technical Specification 3.7.10, to require isolation of the control room prior to movement of recently irradiated fuel.

f. Revise Technical Specification 3.9.6, to allow ANY containment penetrations to be open under Administrative Control (including the equipment hatch) during Refueling Operations.

g. Revise 3.3.7 to remove Actions and Surveillance Requirements associated with R23 instrumentation.

Serial Number 11-025A Attachment 4 Page 26 of 191 Table 3.0-1 Accident Dose Acceptance Criteria Accident or Case Control EAB & LPZ Room(')

Design Basis LOCA 5 rem TEDE 25 rem TEDE Steam Generator Tube Rupture Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE(2)

Main Steam Line Break Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE(2)

Locked Rotor Accident 5 rem TEDE 2.5 rem TEDE(2)

RCCA Ejection Accident 5 rem TEDE 6.3 rem TEDE(2)

Fuel Handling Accident 5 rem TEDE 6.3 rem TEDE(2)

Waste Gas Decay Tank Failure 5 rem TEDE 0.25 rem TEDE(3)

Volume Control Tank Rupture 5 rem TEDE 0.25 rem TEDE (3)

(1) Based on 10CFR50.67 and 10 CFR 50, Appendix A, GDC 19 (2) Reduced from 10 CFR 50.67 criteria in accordance with RG 1.183 for higher probability events.

(3) Current licensing basis

Serial Number 11-025A Attachment 4 Page 168 of 191 3.8.6 WGDT Analysis Results The results of the design basis WGDT analysis are presented in Table 3.8-3. These results report the calculated dose for the worst 2-hour interval (EAB), and for the assumed 30-day duration of the event for the control room and the LPZ. The EAB and LPZ doses are calculated with RADTRAD and are compared with the applicable acceptance criteria used in the Safety EvaluationosotinA en nt166

rence=_l,0*). Control Room dose is compared with the limits defined in General Design Criteria 19 (Reference 31) and applicable standards in RG 1.183.

Table 3.8-3 Dose Results for the WGDT Accident Location (rem) Limits (rem)

EAB 0.1 (TEDE) 0.25 (TEDE)

LPZ 0.1 (TEDE) .250 -(TEDE)

Control Room 0.4 (TEDE) 5 (TEDE)

The results in Table 3.8-3 represent the highest control room and offsite doses that would result from a WGDT accident using worst case scenario conditions. As discussed previously, the control room consequences above assume control room isolation and unfiltered inleakage assumptions that maximize control room dose.

Control room dose in an unisolated control room will actually be less than the value listed in Table 3.8-3.

Serial Number 11-025A Attachment 4 Page 183 of 191 3.9.6 VCT Analysis Results The results of the design basis VCT analysis are presented in Table 3.9-5. These results report the calculated dose for the worst 2-hour interval (EAB), and for the assumed 30-day duration of the event for the control room and the LPZ. The EAB and LPZ doses are calculated with RADTRAD and are compared with the applicable acceptance criteria used in the Safety ,, vluatonsupporing Amendment 166 Ref~er~ee_10). Control Room dose is compared with the limit specified in General Design Criteria 19 (Reference 31) and applicable standards in RG 1.183.

Table 3.9-5 Dose Results for the VCT Accident Location (rem) Limits (rem)

EAB 0.1 (TEDE) 0.25 (TEDE)

LPZ 0.1 (TEDE) 0.25 (TEhDJ)

Control Room 0.6 (TEDE) 5 (TEDE)

The results in Table 3.9-5 represent the highest control room and offsite doses that would result from a VCT accident using worst case scenario conditions. As discussed previously, the control room consequences above were calculated using control room isolation and unfiltered inleakage assumption combinations that will maximize control room dose. Control room dose in an unisolated control room would be less than the value listed in Table 3.9-5.

Serial No. 11-025B ENCLOSURE 5 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

LICENSE AMENDMENT REQUEST 244, PROPOSED REVISION TO RADIOLOGICAL ACCIDENT ANALYSIS AND CONTROL ROOM ENVELOPE HABITABILITY TECHNICAL SPECIFICATIONS LAR 244, ATTACHMENT 5 CHANGED PAGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 11-025A Attachment 5 Page 1 of 19 EVALUATION OF PROPOSED NEW MANUAL ACTIONS Introduction In accordance with the revised RAA provided in Attachment 4, DEK is proposing two manual actions to ensure post-accident dose is maintained within limits. The revised RAA credits these manual actions to limit consequences of the Fuel Handling Accident (FHA) and Locked Rotor Accident (LRA). The proposed manual actions are as follows:

1. The revised RAA credits manual operator action to isolate the control room envelope (CRE) within one hour after initiation of an LRA. This manual action is required to compensate for the proposed TS changes that would discontinue credit for CRE auto-isolation using a high radiation signal from R-23.

2. The revised RAA assumes the CRE is isolated prior to movement of recently irradiated fuel assemblies (per new LCO Requirement added to TS 3.7.10). In addition, the revised RAA credits manual initiation of the Control Room Post Accident Recirculation (CRPAR) system within 20 minutes of occurrence of a FAA.

1.0 Manual Action for the Locked Rotor Accident This proposed manual action would require operator action to isolate the CRE within one hour after initiation of an LRA. This manual action is required to compensate for the proposed TS changes that would discontinue credit for CRE auto-isolation using a high radiation signal from R-23. R-23 is a single channel non-safety related instrument, and therefore DEK has proposed not crediting this radiation monitor in the revised radiological analyses. The LRA scenario is described in Attachment 4, section 3.6.

Verification of successful action is provided in KPS Emergency Operating Procedure (EOP) E-0, "Reactor Trip or Safety Injection." EOP E-0 provides direction regarding which status lights and annunciators will be illuminated if SI is actuated. In addition, verification of CRE isolation and CRPAR initiation is provided by observing status lights on the control board for the CRPAR fan and Control Room Air Conditioning (CRAC) fan.

If an automatic actuation of SI does not occur during this accident, the operators are directed to manually initiate both trains of SI once subcooling is lost. This action will isolate the CRE and start both CRPAR trains.

If the proposed manual action is accomplished within one hour of an LRA occurring, then control room doses will be maintained within the limits specified in 10 CFR 50.67.

2.0 Manual Actions for the Fuel Handling Accident This proposed manual action is based on the premise that upon initiation of a postulated FHA, the CRE will have been previously manually isolated in accordance with a

Serial No. 11-025A Attachment 5 Page 2 of 19 proposed new LCO Requirement in TS 3.7.10, "Control Room Post Accident Recirculation (CRPAR) System." The proposed new manual action consists of initiating one train of the CRPAR system within 20 minutes after the occurrence of a FHA. The revised FHA analysis is provided in Attachment 4, Section 3.3. The revised FHA analysis assumes that the CRE is isolated prior to moving recently irradiated fuel as required by TS 3.7.10. Upon occurrence of a FHA, the analysis assumes manual operator action to initiate one train of the CRPAR system. This manual operator action must be completed within 20 minutes following a FHA to ensure control room occupant dose remains within the limits specified in 10 CFR 50.67.

Control room operators would be promptly notified of a FHA by either of the following methods. These methods provide multiple and diverse means of alerting control room operators to the occurrence of a FHA.

1. KPS Procedure NF-KW-RRF-014, "Fuel Movement During a Refueling Outage,"

requires direct communication be maintained between the control room and the containment operating floor whenever changes in core geometry are taking place.

This ensures that control room operators would be promptly alerted if an FHA event occurs.

2. The KPS Technical Requirements Manual (TRM) 8.9.4, "Radiation Monitoring During Refueling Operations," requires continuous monitoring of radiation levels in the containment and spent fuel pool areas during refueling operations. TRM 8.9.4 is met by requiring radiation monitors R-2, R-5, R-12 and R-21 to be operating during refueling operations. Each of these radiation monitors alarms in the control room.

The proposed changes to TS 3.7.10 (see Attachment 1, Section 2.2.2) would require the CRE be isolated with no fresh air being supplied to the control room (outside air dampers closed and CRPAR fan off) during movement of recently irradiated fuel assemblies. In this configuration, the proposed manual action consists of initiating one train of the CRPAR system. One train of the CRPAR system is initiated by turning either the A (ES-46545) or B (ES-46546) control room hand switch for CRPAR Recirculation Fan to the ON position. Then, using control switch ES 40030 recirculation damper ACC3A is opened, or using control switch ES40031 recirculation damper ACC3B is opened, depending on which train is being started. Proper operation of the train would be verified by the operator using status lights on the control board for the CRPAR fan and CRAC fan. The revised RAA assumes one train of the filtration/recirculation system is placed in operation within 20 minutes of FHA initiation.

The proposed changes to TS 3.7.10 would require that the CRE be isolated with no fresh air being supplied to the control room (outside air dampers closed and CRPAR fan off) during movement of recently irradiated fuel assemblies. However, this TS can be modified with application of the existing TS 3.7.10 Note which permits the CRE boundary to be opened intermittently under administrative controls. In this less likely alignment, the manual action would consist of closing one outside air damper, in

Serial No. 11-025A Attachment 5 Page 3 of 19 addition to initiating one train of the CRPAR system. In this configuration, the revised RAA assumes one train of the filtration/recirculation system is placed in operation within 20 minutes of FHA occurrence and the CRE is fully isolated.

In this configuration, the operator must release the control room switches for outside air dampers ACC-2 (ES-46827) and ACC-1A/1B (ES-46833) to perform the required alignment. To align fresh air to the control room with the CRPAR system operating (either train) requires the operator to hold the selector switch for ACC-IA/1iB to the "Normal" (ACC-1A) or "Alt" (ACC-1B) position and hold the control switch for ACC-2 in the "Open" position. Since the operator is required to hold the switches in position, this configuration would be used sparingly and for short durations to provide fresh air to the control room. Therefore, this is considered an infrequent control room ventilation system configuration. Each of these control switches are spring return to "Auto" position and are interlocked to close the dampers when either CRPAR fan is running.

Therefore, when the operator releases the control switches, the CRE will return to the isolated configuration with at least one CRPAR fan running. No further actions would be required except to verify the correct alignment.

3.0 Acceptability of Proposed Manual Actions The NRC has provided guidance regarding the requirements for use of manual actions.

NRC RIS 2005-20, Revision 1, Section C.5 (reference 1), discusses the conditions under which temporary manual actions may be used in lieu of automatic actions in support of operability. NRC Information Notice (IN) 97-78 (reference 2) alerted licensees to the importance of considering the effects on human performance of such changes made to plant safety systems. Information Notice 97-78 states:

"The original design of nuclearpower plant safety systems and their ability to respond to design-basis accidents are described in licensees' FSARs [final safety analysis reports] and were reviewed and approved by the NRC. Most safety systems were designed to rely on automatic system actuation to ensure that the safety systems were capable of carrying out their intended functions. In a few cases, limited operator actions, when appropriately justified, were approved. Proposed changes that substitute manual action for automatic system actuation or that modify existing operator actions, including operator response times, previously reviewed and approved during the originallicensing review of the plant will, in all likelihood, raise the possibility of an unreviewed safety question (USQ). Such changes must be evaluated under the criteria of 10 CFR 50.59 to determine whether a USQ is involved and whether NRC review and approval is required before implementation.... In the NRC staffs experience, many of the changes [involving operator actions] proposed by licensees do involve a USQ."

It is recognized that the NRC updated 10 CFR 50.59, to remove the USQ wording.

Nonetheless, the intent of IN 97-78 is still pertinent. That is, licensees still need to

Serial No. 11-025A Attachment 5 Page 4 of 19 submit many of the changes in operator actions to the NRC for review and approval in accordance with 10 CFR 50.59.

The guidance presented in NUREG-1764 (reference 3) can be used to address safety-related operator actions (SROAs), as well as other required operator actions. The American National Standards Institute/American Nuclear Society defines "safety-related operator action" in ANSI/ANS-58.8-1994, as follows:

"A manual action required by plant emergency procedures that is necessary to cause a safety-related system to perform its safety-related function during the course of any DBE (design-basis event). The successful performance of a safety-related operator action might require that discrete manipulationsbe performed in a specific order."

Per NUREG-1764 changes in human actions (HAs) (synonymous with the term "operator actions") result from the following types of plant activities:

Plant modifications.

" Procedure changes.

Equipment failures.

Justifications for continued operations (JCOs)1 .

Identified discrepancies in equipment performance or safety analyses.

NUREG-1764 provides guidance for the review of human actions. This document provides guidance for use in determining the appropriate level of human factors engineering (HFE) review of HAs based upon their risk-importance. This guidance uses a graded, risk-informed (RI) approach consistent with RG 1.174, Rev. 1 (reference 4).

This guidance uses a two-phased approach to reviewing HAs. Phase 1 is a risk screening and analysis of the affected HAs identified to determine their risk-importance and the level of HFE review that is appropriate in Phase 2. Phase 2 is an HFE review of those HAs that are found to be risk-important.

3.1. Phase I - Risk Screening KPS has elected to provide this application using non-risk informed (non-RI) analysis techniques. The non-RI screening process consists of the following steps:

1. Verify that the non-RI change request is appropriate.

2. Assess safety-significance of the HAs.

3. Qualitatively assess the safety-significance of HAs involved in the change request.

1 NOTE: The term JCO is no longer recognized by the NRC as valid. Per RIS 2005-20 (reference 29),

"An SSC that is determined to be operable but degraded or nonconforming is considered to be in compliance with its TS LCO, and the operability determination is the basis for continued operation."

Serial No. 11-025A Attachment 5 Page 5 of 19

4. Make an integrated assessment of HA safety-significance to determine the appropriate level of HFE review (i.e., Level 1, 2, or 3).

The assessment of these four steps is provided below. Requirements are in normal text and responses are provided in italics text.

1. The non-RI chanqe request is appropriate In accordance with the guidance, the risk implications of a non-RI submittal would warrant further risk informed analysis if the submittal:

Significantly changes the allowed outage time (e.g., outside the range previously approved at similar plants), the probability of the initiating event, the probability of successful mitigative action, the functional recovery time, or the operator action requirement; Response: There is no significant change to allowed outage time. The proposed manual actions are limited to the response to the FHA and LRA. There is no change to the allowed outage time of any equipment designed to mitigate these accidents. DEK is proposing new TS requirements to isolate the control room prior to movement of recently irradiated fuel assemblies. This new requirement simplifies the necessary manual action in the event of a fuel handling accident.

Furthermore, the manual actions proposed do not change the probability of any initiating event or the probability of successful mitigation of events as discussed in Attachment 4. Finally, the proposed operatoractions do not change the functional recovery time of any other accident scenario or change other operator actions required to recover from anotheraccident.

Significantly changes functional requirements or redundancy; Response: The proposed manual actions do not significantly change functional requirements or redundancy. Operation of the CRPAR system is required for radiological accidents and a CRE isolation is assumed. The proposed manual actions do not change the system operation. CRPAR system starting, filtering and redundancy requirements are not changed. CRE isolation redundancy requirements have not changed. There is no change in redundancy as both CRPAR trains are still required to be operable as well as the CRE per the proposed TS changes. Therefore, there is no change in the functional requirements or redundancyfor the CRPAR system and CRE isolation.

Significantly changes operations that affect the likelihood of undiscovered failures; Response: The proposed operatoraction does not significantly change operations that affect the likelihood of undiscovered failures. Failures of the CRPAR system or dampers to isolate are indicated by lights in the control room. Therefore,

Serial No. 11-025A Attachment 5 Page 6 of 19 operators will be made aware of a failure of the CRPAR system or CRE through these lights. The manual actions proposed herein would not mask or hide any undiscovered failures of the CRPAR system or the CRE.

Significantly affects the basis for successful safety function; Response: The KPS current licensing bases (CLB) relied on a non-redundant, non-safety related radiation monitor to initiate the CRPAR system and perform a partial CRE isolation. The proposed manual action relies upon redundant, safety related control switches for CRPAR initiation and CRE isolation prior to movement of recently irradiatedfuel assemblies. The functionality of the CRPAR system is not changed and the safety function is enhanced by complete CRE pre-isolation.

" Could create "special circumstances" under which compliance with existing regulations may not produce the intended or expected level of safety and plant operation may pose an undue risk to public health and safety.

Response: No special circumstances are present in this application. There is no substantialincreasein the likelihood or consequences of accidents that are beyond the design and licensing basis for KPS. There is no change in the levels of defense or cornerstones of reactor safety with this application. The proposed change does not significantly reduce the availability, or reliability of structures, systems, components or other human actions that are risk significant but are not required by regulations. Finally, the proposed change does not involve a change for which synergistic or cumulative effects could pose an undue risk to the public health and safety.

The proposed HAs are simple and effective and are not subject to potential impacts of "specialcircumstances."

2. Assess the safety significance of the HAs NUREG-1764 discusses two methods for determining the safety-significance of HAs.

The first method, the "Estimated Importance Method," requires an estimate of the risk-importance of the HA. The second, the "Generic HA Method" is based upon general risk information and some plant-specific information.

The proposed HAs are related to recovery from a FHA or LRA event. The plant specific PRA does not model initiation of the CRPAR system after the FHA event or isolation of the control room after an LRA event. Therefore, DEK used the "Generic HA Method" in NUREG-1764, Section 2.4.3 to assess the safety significance of the proposed HAs.

DEK has performed a safety significance review based on the Generic HA Method.

This method is the most appropriate based on the fact that the KPS PRA does not

Serial No. 11-025A Attachment 5 Page 7 of 19 model initiation of the CRPAR system after the FHA event or isolation of the control room after an LRA event. The assessment using the Generic HA Method is as follows:

Generic HA Method - Preliminary Screening The proposed HAs were compared to the generic list in NUREG-1764, Table A.2, "Generic PWR Human Actions that are Risk-Important." A determination was made that the proposed human actions are not either HA Group 1 (HAs that are Risk Important) or HA Group 2 (HA's that are Potentially Risk Important).

The HAs were reviewed to determine if they involve a risk important system. The proposed manual actions are related to recovery from a FHA or LRA event (isolation of the control room or initiation of CRPAR). Since the plant specific PRA does not model isolation of the control room or initiation of CRPAR after these events it is concluded that the HAs do not involve a risk important system.

Based on the above, the proposed HAs are preliminarily considered to require a Level III review per the NUREG-1 764, Section 2.4.3, "Generic HA Method."

3. Qualitatively assess the safety-significance of HAs Three types of qualitative assessment are used:

a. Personnel Functions and Tasks

b. Design Support for Task Performance

c. Performance Shaping Factors Three types of assessments are discussed as follows:

a. Personnel Functions and Tasks This type of qualitative assessment examines the potential effects of the proposed HA for changes to operator tasks and the functions that they perform, under five major categories:

Operating Experience: Does the requested change adversely affect the performance of an action that was previously identified as problematic based on experience/events at that plant or plants of similar design?

Response: No, the requested change does not adversely affect the performance of an action that was previously identified as problematic.

Currently, there are no required actions associated with CRPAR performance or CRE isolation. Routine operator actions associated with CRPAR operation (verifying lights, verifying annunciators,damper position, etc.) are not impacted

Serial No. 11-025A Attachment 5 Page 8 of 19 by the proposed manual operatoractions. Manual initiation of SI during a LRA with fuel damage is already a required action to prevent loss of subcooling to the core. This manual action is not changed.

New Actions: Does the requested change introduce new HAs? Are the new HAs associated with new responsibilities for the success of safety functions (or additional actions associated with existing responsibilities)?

Response: Yes, the change does introduce new HAs. The proposed HAs are associatedwith success of the safety function for protection of the control room occupants. For the FHA, the safety function associated with the HA is initiation of at least one train of the CRPAR system. The proposed HA supports long-term operation of KPS during and after a FHA by ensuring that control room occupants are exposed to as low a radiologicaldose as possible. For the LRA, the safety function associated with the HA is manual initiation of SI, which causes isolation of the CRE and initiationof the CRPAR system. The proposed HA supports long-term operation of KPS during and after a LRA by ensuring that control room occupants are exposed to as low a radiological dose as possible.

Change in Automation: Has the requested change given personnel a new functional responsibility that they previously did not have and which differs from their normal responsibilities? For example, are operators now required to take an action in place of a previously automated one? Consider the example of simply being required to open a valve that previously was automatically operated, and where the action required to do so is similar to other valve-opening operations with which operators are familiar. This would not be a sufficient change (in and of itself) to warrant a "yes" to this question when considering task complexity. However, there may be increased workload if the aggregate of added actions is judged to be excessive, this may warrant a "yes."

Response: No, while new tasks are required as discussed above, the example in this question is directly applicable to the proposed HA. The HA for the FHA is simply to turn control room hand switches associated with CRPAR system.

This act is similar to the example of opening a valve that was previously automated. Operatorsare familiar with this type of action and routinely perform similar actions. This action is not complex and would be associatedonly with a FHA. For the LRA with fuel damage, the proposed HA is to manually initiate push buttons associated with Safety Injection. This is an action already required for loss of subcooling, and is therefore not a new action to the operators. Therefore, the proposed HAs are not excessive and are considered a minor increasein the workload for these events.

Change in Tasks: Has the requested change significantly modified the way in which personnel perform their tasks (e.g., making them more complex,

Serial No. 11-025A Attachment 5 Page 9 of 19 significantly reducing the time available to perform the action, increasing the operator workload, changing the operator role from primarily "verifier" to primarily "actor")? In this case, operators do not have a new functional responsibility; instead, the way that they perform their current functional responsibilities has significantly changed and is different from what they usually do.

Response: The proposed HAs do not significantly change the way in which operators perform their tasks. As described above, operators routinely monitor the control room indications and plant status during and after an event.

Initiation of one train of the CRPAR system is not complex and is consistent with the operator's role during an event. In addition, for the LRA with fuel damage, an expected response of the operatoris to manually initiate SI upon a loss of subcooling.

As discussed in Section 2.0, one train of the CRPAR system must be initiated within 20 minutes for the FHA event. The CRE is required by TS 3.7.10 to be isolated prior to moving recently irradiatedfuel assemblies. As discussed in Section 1.0, at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is available from the initiation of an LRA before the HA is required to be completed. The HA consists of manually depressing SI signal push buttons, which causes isolation the CRE and initiation of the CRPAR system. Once these HAs are complete, the control room occupants will be provided protection during the FHA and LRA. Therefore, there is no significant new functional responsibilityor significant change in responsibilities for operators during these events.

Change in Performance Context: Has the requested changed created, in some way, a new context for task performance? Or, does the change identify a previously unrecognized context? Or, does the request address a context previously not modeled or considered? If so, what are the important differences in context (e.g., different plant mode, plant behavior, timing of plant symptoms)?

Response: The proposed HAs will not create or modify the context for task performance. As described above, the proposed HAs would be performed after occurrence of a FHA or LRA event. The context of performing HAs during an accident scenario is a function that is requiredto be understood by operators in their training for accident response. Therefore, the context is expected and has not changed.

b. Design Support for Task Performance This type of qualitative assessment addresses how well the performance of the HAs is supported (e.g., with job aids):

Serial No. 11-025A Attachment 5 Page 10 of 19 Change in Human-System Interfaces (HSls): Has the requested change significantly changed the HSIs used by personnel to perform the task? For example, are personnel now performing their tasks at a computer terminal where previously they were performed at a control board with analog displays and controls?

Response: The proposed HAs would not change any HSIs. The proposed HAs require manipulation of controls that are known to the operators. No new controls or human-system interfaces are proposed. The proposed HAs are simple and routine for operators.

Change in Procedures: Has the requested change significantly changed the procedures that personnel use to perform the task, or is the task not supported by procedures?

Response: A significant change to the procedures that operators use to perform the proposed HAs is not necessary. For the LRA with fuel damage, plant emergency procedures currently require manual initiation of SI if a loss of subcooling occurs. For the FHA, the manipulation of CRPAR system switches will be directed by station operating procedures as part of implementation of this amendment.

Change in Training: Has the requested change significantly modified the training, or is the task not addressed in training?

Response: The proposed HAs have been provided to operators in training.

For the LRA with fuel damage, plant emergency procedures currently require manual initiation of SI if a loss of subcooling occurs. Operatortraining requires the operatorto memorize this step. For the FHA, the initiation of one train of CRPAR system is provided in trainingand the reasons/basisfor performing this HA is discussed in training.

c. Performance Shaping Factors This type of qualitative assessment addresses four performance shaping factors:

Changes in Teamwork: Has the requested change significantly changed the team aspects of performing an action. For example, (1) is one operator now performing the tasks accomplished by two or more operators in the past? (2) is it now more difficult to coordinate the actions of individual crew members? or, (3) is task performance more difficult to supervise after the modification?

Response: No changes in teamwork are required. No additional operatorsare requiredin the control room to perform the proposed HAs. There is no greater level of difficulty and no increase in the level of supervision necessary to

Serial No. 11-025A Attachment 5 Page 11 of 19 accomplish proposed HAs. Manipulation of control room switches is a routine evolution for operators,and no additionallevel of supervision is required.

Changes in Skill Level of Individuals Performing the Action: Has the requested change kept the same HA but made it necessary for an individual who is less trained and has lower qualifications to take the action than was the case before the modification? Here, context is defined as the overall performance environment, including plant conditions and behavior that, for example, affect the time available for the operator response and the effectiveness of job aids under these conditions that lead to the assessment of performance shaping factors.

Response: The skill level of the operatorperforming the proposed HAs and the performance environment for the operator will not change. For the LRA with fuel damage, the procedural requirement to initiate SI is required to be memorized by operators. This has not changed. For the FHA, initiation of one train of the CRPAR system via hand switches is a routine type task. Job aids consist of control switch identification placards on the control boards and understandingwhen the SI manual push buttons and CRPAR system switches need to be initiated. These are simple routine tasks for operators and do not requirenew skills or additionaltraining to accomplish.

Change in Communication Demands: Has the requested change significantly increased the level of communication needed to perform the task? For example, must an operator now communicate with other personnel to perform actions that previously could be taken at a local panel containing all necessary HSIs?

Response: The proposed HAs do not require significantly increased levels of communication to accomplish. Direction communicated by the unit supervisor during an event is considereda routine communication. Accomplishment of the proposed HAs is easily verified by lights and annunciatorsin the control room.

In addition, operatorsare accustomed to working in pairs for peer checking and independent verification of system alignments.

Change in Environmental Conditions: Has the requested change significantly increased the environmental challenges (such as radiation, or noise) that could negatively affect task performance?

Response: No, the operatorwill be performing the proposed HAs to ensure the CRE does not become a high radiation environment. There is no change in noise level associated with the proposed manual actions as these are the dampers and systems that normally provide air to the control room.

Serial No. 11-025A Attachment 5 Page 12 of 19

4. Make an integrated assessment of HA Safety-Significance The results of the qualitative assessment of HA Safety Significance have determined that the action is well defined and can easily be performed (it is clear when to perform the action), procedural direction exists, there is sufficient time and staff available to perform the action, and the action is similar to those routinely performed. Based on this, the level of HF review could be reduced to Level Ill. The Level Ill classification is warranted since most of the areas reviewed were answered "no" and the analysis indicates very little change is being made.

However, since the action involves support of a safety function and failure to accomplish the proposed HAs within the stated time limitations could potentially result in operators exceeding dose limits, the level classification will conservatively remain at Level II.

3.2. Phase II - HFE Review of ProposedHA using Level II Review Criteria Based on the results of the Phase I Risk Screening provided above, DEK has conservatively determined that the proposed HAs will be assessed using the Level II criteria identified in Section 4 of NUREG-1764. NUREG-1764 specifies that a Level II review include the following elements:

1. General Deterministic Review

2. Analysis

3. Design of Human System-Interfaces, Procedures and Training

4. Human Action Verification These four elements are assessed below:

1. General Deterministic Review Criteria Objective: The objective of this section is to verify that deterministic aspects of design, as discussed in RG 1.174, have been appropriately considered by the licensee. Deterministic aspects include verifying that the change meets current regulations and does not compromise defense-in-depth.

Scope:, The deterministic review criteria are applicable to all modifications associated with Level II HAs.

Criteria:

1) The licensee should provide adequate assurance that the change meets current regulations, except where specific exemptions are requested under 10 CFR 50.12 or 10 CFR 2.802. Examples of regulations that may be affected by a change, but that may be identified as risk-significant when using a standard

Serial No. 11-025A Attachment 5 Page 13 of 19 PRA to screen for risk include the following: 10 CFR Part 20, Criterion 19 of Appendix A to 10 CFR Part 50, and Appendices C through R to 10 CFR Part 50.

Response: See discussions located in Attachment 1 and Attachment 4 related to compliance with regulations and conformance to accident analyses criteria.

See section 5.2 for a discussion of compliance with General Design Criteria applicable to KPS.

2) The licensee should provide adequate assurance that the change does not compromise defense-in-depth:

Response: Defense-in-depth is one of the fundamental principles upon which KPS was designed and built. Defense-in-depth uses multiple means to accomplish safety functions and to prevent the release of radioactive materials.

It is important in accounting for uncertainties in equipment and human performance, and for ensuring some protection remains even in the face of significant breakdowns in particularareas.

Defense-in-depth is not compromised or altered as a result of the proposed HAs. Defense-in-depth is accomplished in this particular case by having multiple reliable methods to contain highly radioactive materials during a design bases accident. The containment structure, shield building ventilation system, and auxiliary building special ventilation system all minimize the release or act upon the release of highly radioactive materials should barriers fail. Each of these systems has the goal of protecting the health and safety of the public and the control room occupants during an event. The proposed HAs ensure that the CRE is isolated and a filtered source of air is available for the control room occupants. This function is not compromised by performance of the proposed HAs.

The proposedHAs do not lead to an over-reliance on programmaticactivities to compensate for weaknesses in plant design. System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,

no risk outliers).

The proposed HAs preserve defenses againstpotential common cause failures, and there is no potential for the introduction of a new common cause failure mechanism. The proposed HAs include initiation of an SI signal which causes actuation of the CRPAR system. Therefore, the independence of barriersis not compromised. Defenses against human errors are preserved because the proposed HAs are included in procedures and are included in operatortraining.

Human errors, should they occur, are easily detectable by control room annunciatorsand equipment status lights.

Serial No. 11-025A Attachment 5 Page 14 of 19

2. Analysis Objective: The objective of the review is to verify that the licensee has analyzed the changes to HAs and identified HFE inputs for any modifications to the HSI, procedures, and training that may be necessary.

Scope: The review criteria are applicable to all modifications associated with Level II HAs.

Criteria:

1) Functional and Task Analysis The licensee should identify how the personnel will know when the HA is necessary, that it is performed correctly, and when it can be terminated.

Response: The need for performing the proposed HA associated with the LRA will be identified by multiple alarm conditions. The following conditions indicate the need for operatoraction during a LRA event (details are provided in section

3.6.1 above)

" A sudden decrease in core coolant flow which results in fuel damage as indicated by the RCS subcooling monitor (loss of sufficient cooling to the fuel).

" Upon indication of loss of subcooling, operators enter emergency operationsprocedures and initiate SI based on loss of subcooling.

Initiation of an SI signal causes isolation of CRE dampers and initiation of the CRPAR system.

Indication of correct performance of this proposed HA is made by verifying that the train of SI selected to respond to the LRA is functioning by observing that the indicating lights associated with the selected control switch change color from green (Standby) to red (On). In addition, the train of CRPAR automatically selected by association with the selected SI train for the mitigation of a LRA is verified to be functioning by observing that indicating lights associated with the selected control switch change color from green (Standby) to red (On) and verifying that the associated CRAC fan automatically starts by observing the indicatorlight for this fan is red (On). Finally, the CRE boundary is verified to be intact by observing indication of damper positions in the closed position.

This can be done by observing indicationson damper controlswitches.

Serial No. 11-025A Attachment 5 Page 15 of 19 Termination of CRPAR operation is not necessary until radiation conditions return to normal background levels, indicating that a radioactive release is no longer occurring.

The following conditions indicate the need for operatoraction during a FHA:

Verbal communication from personnel on the containment operating floor or spent fuel pool area that indicates a FHA has occurred.

High radiationalarm indicatedon R-2, R-5, R-12 or R-21 indicates a FHA in containment or the spent fuel pool area has occurred.

Indication of correct performance of the proposed HA is made by verifying that the train of CRPAR selected for the mitigation of a FHA is functioning. This is done by observing that the indicating lights associatedwith the selected control switch change color from green (Standby) to red (On) and verifying that the associated CRAC fan automatically starts by observing the indicating light is red (On). The CRE boundary would have previously been verified to be intact by observing indication of damper positions in the closed position prior to moving recently irradiatedfuel.

Termination of CRPAR operation is not necessary until radiologicalconditions return to normal background levels, indicating that a radioactive release is no longer occurring.

Task analyses should provide a description of what the personnel must do.

The licensee should identify how human tasks or performance requirements are being changed. The task analysis should identify reasonable or credible, potential errors and their consequences.

Response: Refer to Sections 1.0 and 2.0 for a detailed description of the proposed HAs. The proposed HAs are only required for the LRA and the FHA.

LRA There are a limited number of credible failures or errors that the operator can make during the LRA event. As shown above, proposed HA is not complex (only requires manipulation of the manual SI push buttons) and requires little effort to complete. However, human errors do occur and the proposed HA accounts for the consequences of such errors. For example, if an operatorfails to initiate SI, then multiple alarms indicatinghigh radiation conditions may occur at various locations (e.g. R-9, R-11, R-12, or R-21). If the SI signal fails to isolate the control room, then the alarm from control room radiationmonitor R-1 will not clear without some action being taken. As a backup, R-23 (although not credited in the radiologicalaccident analyses) is still functional and will actuate

Serial No. 11-025A Attachment 5 Page 16 of 19 on a high radiation condition. Similarly, if the SI signal does not start the associated train of CRPAR, then a high radiation condition will continue to persist until the controlroom operatormanually starts a train.

FHA There are a limited number of credible failures or errors that the operator can make during the manipulations. As shown above, the proposed HA is not complex and requireslittle effort to complete. However, human errors do occur and the proposed HA accounts for the consequences of such errors. For example, if an operator fails initiate the CRPAR system, then the alarm from radiation monitor R-1 will not clear without some action being taken. As a backup, R-23 (although not credited in the analyses is still functional) will actuate on a high radiationcondition.

If the operator incorrectly manipulates the wrong control room switches, the most likely outcome will be that the CRPAR system will not be initiatedand high radiation alarms in the control room would continue. This would alert the operator that the incorrect switch was manipulated and could be corrected immediately.

2) Staffing:

The effects of the changes in HAs upon the number and qualifications of current staffing levels of operations personnel for normal and minimal staffing conditions.

Response: is the proposed HAs would have no effect on the number and qualifications of operationspersonnel required to support operations in a post-event condition. It is routine for operators to monitor control room conditions and verify properoperation of equipment in the control room.

3. Design of Human System-Interfaces, Procedures, and Training Objective: The objective of the review is to verify that the licensee has supported the HAs by appropriate modifications to the HSI, procedures, and training.

Scope: The review criteria are applicable to all modifications associated with Level II HAs.

Serial No. 11-025A Attachment 5 Page 17 of 19 Criteria:

1) HSIs:

Temporary and permanent modifications to the HSI should be identified and described. The modifications should be based on task requirements, HFE guidelines, and resolution of any known operating experience issues.

Response: No HSI modifications or new HSIs are required. The proposed HAs are simple and control room switches and push buttons are well marked.

2) Procedures:

Temporary and permanent modifications to plant procedures should be identified and described. The modifications should be based on task requirements and resolution of any known operating experience issues.

Justification should be provided when the plant procedures are not modified for changes in operator tasks.

Response: The appropriate modifications to plant procedures will be made as part of the implementation of this amendment request.

3) Training:

Temporary and permanent modifications to the operator training program should be identified and described. The modifications should be based on task requirements and resolution of operating experience issues. Justification should be provided when the training program is not modified for changes in operator tasks.

Response: Training lesson plans will be revised to incorporate the bases for performing the proposed HAs contingent upon approval of this amendment request. The requirements of the training will be developed using the process specified in DEK trainingdevelopment procedures.

4. Human Action Verification Objective: The objective of this review is to verify that the licensee has demonstrated that the HAs can be successfully accomplished with the modified HSI, procedures, and training.

Scope: The review criteria are applicable to all modifications associated with Level II HAs.

Serial No. 11-025A Attachment 5 Page 18 of 19 Criteria:

1) An evaluation should be conducted at the actual HSI to determine that all required HSI components, as identified by the task analysis, are available and accessible.

Response: DEK has performed a walkdown of the control room and has verified that components required to perform the proposed HAs are accessible and available to the operator.

2) A walkthrough of the HAs under realistic conditions should be performed to determine that;

The procedures are complete, technically accurate, and usable.

The training program appropriately addressed the changes in plant systems and HAs.

The HAs can be completed within the time criterion for each scenario that is applicable to the HAs. The scenario used should include any complicating factors that are expected to affect the crews' ability to perform the HAs.

Response: As part of the walkdown described above, DEK developed and verified the procedures to be used as guidance to the operatorsfor performing the proposed HAs. The procedures are used during the training of the operators to achieve a simulated performance of operator actions during training sessions.

3) The walkthroughs should include at least one crew of actual operators.

Response: Operationspersonnel were included in the walkdown of the control room.

4.0 Conclusions This evaluation has demonstrated that the proposed HAs are acceptable. The Phase I Risk Screening (Section 3.1) demonstrated that the safety significance of the proposed HAs is minimal and warranted only a Level III Human Factor review. However, since the proposed HAs involved action that support a safety function, and failure to perform the proposed HAs within the prescribed time limits could potentially result in operators in the control room exceeding dose limits, the classification level was conservatively left at Level IIfor the purposes of reviewing the HAs.

Serial No. 11-025A Attachment 5 Page 19 of 19 The results of the HFE review of the proposed HAs have determined that the four elements of the HFE review have been satisfied without identifying any obstacles to implementation. All the HFE elements of a Level II review were satisfied including:

The technical review provides adequate assurance that the proposed HAs meet current regulations.

The proposed HAs have been analyzed for their impact on current procedures, control room staffing, human system interfaces, and training.

The proposed HAs are captured in procedures and in training.

It has been demonstrated that the proposed HAs can be accomplished with the procedures and training provided.

5.0 References

1. RIS 2005-20, Revision 1, "Revision to NRC Inspection Manual Part 9900 Technical Guidance, Operability Determination & Functional Assessments for Resolution of Degraded and Nonconforming Conditions Adverse to Quality or Safety," dated April 16, 2008.

2. Information Notice 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times," dated October 23, 1997.

3. NUREG-1764, Revision 1, "Guidance for the Review of Changes to Human Actions," dated January 2005.

4. Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated November 2002. [ADAMS Accession No. ML023240437]

v t e Letter Sequence Response to RAI
TAC:ME7110, Control Room Habitability, Administratively Control Containment Penetrations, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Administration Questions 15 May 2012 15 May 2012 1 December 2011 9 January 2012 13 January 2012 30 January 2012 6 March 2012 13 April 2012 10 October 2012 Answers 8 November 2011 25 January 2012 27 February 2012 12 March 2012 30 April 2012 6 June 2012 3 August 2012 9 August 2012 Results Approval... Other: ML11290A148, ML12066A008, ML12166A006, ML12193A683, ML12193A684, ML12236A206
MONTHYEAR2012-02-27 [Table View]

ML11318A205

Text

Enclosures:

References:

Subject:

Subject:

4.0 TECHNICAL ANALYSIS

7.0 REFERENCES

Subject:

3.6.1 above)

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