ML18304A105

Request for Additional Information License Amendment Request for Transition to Westinghouse Core Design and Safety Analyses and Alternative Source Term (CAC No. MF9307; EPID L-2017-LLA-0211)

ML18304A105
Person / Time
Site:	Wolf Creek
Issue date:	11/05/2018
From:	Balwant Singal Plant Licensing Branch IV
To:	Heflin A Wolf Creek
Singal B, NRR/DORL/LPL4-1, 415-3016
References
CAC MF9307, EPID L-2017-LLA-0211
Download: ML18304A105 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 November 5, 2018 Mr. Adam C. Heflin President and Chief Executive Officer, Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION RE: LICENSE AMENDMENT REQUEST FOR TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSES AND ALTERNATIVE SOURCE TERM (CAC NO. MF9307; EPID L-2017-LLA-0211)

Dear Mr. Heflin:

By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), as supplemented by letters dated March 22, May 4, July 13, October 18, and November 14, 2017; and January 15, January 29, April 19, June 19, and August 9, 2018 (ADAMS Accession Nos. ML17088A635, ML17130A915, ML17200C939, ML17297A478, ML17325A982, ML18024A477, ML18033B024, ML18114A115, ML18177A198, and ML18232A058, respectively), Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a license amendment request to transition to Westinghouse core design and safety analyses and alternative source term for Wolf Creek Generating Station, Unit 1 (WCGS).

On September 27, 2018, a draft request for information (RAI) was transmitted to the licensee via e-mail. On October 4 and 11, 2018, RAI clarification calls were held between the staff from WCNOC, Westinghouse and the U.S. Nuclear Regulatory Commission (NRC). The final RAI is enclosed with this letter. It was agreed that WCNOC will respond to this request within 60 days from the date of the last clarification call (i.e., by December 10, 2018).

A. Heflin If you have any questions, please contact me at 301-415-3016 or via e-mail at Balwant.Singal@nrc.gov.

Sincerely, b~+-~""-~

Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure RAI cc: Listserv

REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSES AND ALTERNATIVE SOURCE TERM WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), as supplemented by letters dated March 22, May 4, July 13, October 18, and November 14, 2017; and January 15, January 29, April 19, June 19, and August 9, 2018 (ADAMS Accession Nos. ML17088A635, ML17130A915, ML17200C939, ML17297A478, ML17325A982, ML18024A477, ML180338024, ML18114A115, ML18177A198, and ML18232A058, respectively), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee), submitted a license amendment request (LAR) for the Wolf Creek Generating Station, Unit 1 (WCGS). The proposed amendment would, in part, revise the WCGS Technical Specifications {TSs) and the Updated Final Safety Analysis Report (UFSAR) Chapter 15, "Accident Analyses," radiological consequence analyses using an updated accident source term consistent with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.67, "Accident source term."

The U.S. Nuclear Regulatory Commission (NRC) staff issued a request for additional information {RAI} by letter dated December 4, 2017 (ADAMS Accession No. ML17331A178).

The licensee provided responses to the NRC staff request by letters dated January 15, January 29, April 19, and June 19, 2018. Based on the review of the information provided by the licensee, the NRC staff has identified the need for the following additional information in order to complete the review of the LAR. Please note that this request does not represent a new set of RAls and is either: ( 1) based on the review of the new information provided, or (2) the information provided conflicts with previously submitted information. The regulatory bases in general are the same as stated in the letter dated December 4, 2017. Any additional regulatory bases identified during the review have been stated with the individual RAls, as applicable.

Also, since this request is based on the original set of RAls issued by letter dated December 4, 2017, the original RAI designations have been retained. Draft RAls were transmitted to WCNOC on September 27, 2018, and clarification calls were held on October 4 and 11, 2018.

It was agreed that the licensee will provide a response to the NRC staff requests within 60 days from the date of the second call (i.e., by December 10, 2018).

RAI ARCB1-CONTROL ROOM-3 Paragraph 50.67(b )(2)(iii) of 10 CFR requires that the licensee's analysis demonstrates with reasonable assurance that "[a]dequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv [sievert] (5 rem [roentgen equivalent man]) total effective dose equivalent (TEDE) for the duration of the accident." (emphasis added)

Enclosure

Section 8 of Enclosure IV to letter dated January 17, 2017, "NRC Regulatory Issue Summary 2006-04 Comparison," states that the analysis conforms to Issue 1, which states that the amendment should identify and justify each change to the accident analysis. The licensee also states in the comments, that the submittal identifies the changes to the licensing basis analysis and includes sufficient analysis detail to allow for results verification through independent calculations.

The supplemental response for RAI ARCB1-CONTROL ROOM-3 by letter dated January 15, 2018 (page 54) stated (emphasis added):

For the ground shine dose, deposition velocities of 1.0E-02 m/sec [meters per second] for elemental iodine, 1.0E-04 m/sec for organic iodine, and 1.0E-03 m/sec for particulates are modeled, consistent with NUREG/CR-3332.

Noble gases are not assumed to deposit on the ground.

NUREG/CR-3332, "Radiological Assessment, A Textbook on Environmental Dose Analysis,"

dated September 1983 (ADAMS Accession No. ML091770419), page 2-51 states, in part (emphasis added):

Experimentally determined deposition velocities are also a function of wind velocity because the vertical profile of concentration changes with wind velocity.

Thus, the deposition velocity is not constant even for specific effluents. The variations in boundary conditions, such as sorption characteristics and roughness of the underlying surface, and variations in the wind velocity for a given chemical composition of effluent, can cause the deposition velocity to vary by more than one order of magnitude in different experiments (see Chapter 11 ). It should be realized, however, that these deposition velocities are derived from relatively short-term ([on] the order of one hour) experiments and thus vary more than their long-term averages which are expected to be the proper values to estimate long-term deposition.

For effluents from nuclear facilities, the following best-estimates of the deposition velocity, based on experimental data may be adequate: 10-2 m/sec for elemental iodine, 104 m/sec for organic iodine, and 10-3 m/sec for aerosols (approximately 1 micron in diameter). It is not clear that the use of the stated deposition factors from NUREG/CR-3332 is justified, in part, because of the following NRC staff observations:

• The loss-of-coolant accident (LOCA) duration considered is 30 days, but these deposition velocities appear to be derived for "relatively short term" (on the order of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).

• Although the deposition velocities were derived from experiments, it is not clear why the conditions of these experiments is applicable to the potential environmental conditions after a LOCA, including the possibility of rain moisture which would enhance deposition.

• The values assumed are constant over a 30-day period, but over time the deposition velocities would change.

Please provide a justification for the proposed change to the licensing basis analysis to include the deposition factors from NUREG/CR-3332.

RAI ARCB1-LOCA-3 Section 8 of Enclosure IV to letter dated January 17, 2017, "NRC Regulatory Issue Summary 2006-04 Comparison," states that the analysis conforms to Issue 1, which states that the amendment should identify and justify each change to the accident analysis. The licensee also states in the comments that the submittal identifies the changes to the licensing basis analysis and includes sufficient analysis detail to allow for result verification through independent calculations.

Section 50.36, "Technical specifications," of 10 CFR, requires the TSs to be derived from the analyses and evaluation included in the safety analysis report. Per WCGS TS Bases B 3. 7.13, "Emergency Exhaust System (EES)," the design basis is established by the consequences of the limiting design-basis accidents, which includes a LOCA.

In the supplemental response to RAI ARCB1-LOCA-3 by letter dated June 19, 2018, a new analysis (determining the offsite doses from a design basis LOCA and assuming that the EES is not credited) is discussed. Some of the details of the analysis described in the RAI ARCB1-LOCA-3 response needs to be confirmed or provided in order to enable the NRC staff to make a current finding of compliance with 10 CFR 50.67 and 10 CFR 50.36.

Accordingly, please confirm or provide the following information regarding the proposed new LOCA analysis:

1. Please confirm that the new analysis assumes a ground level release from the auxiliary building, and provide the corresponding atmospheric dispersion factor(s) used.

2. Please confirm that the release from the auxiliary building assumes no holdup or mixing of the radioactivity released into the auxiliary building (consistent with WCGS UFSAR Section 15.6.5.4.1.2).

3. Please confirm that the assumed releases into the auxiliary building and atmospheric dispersion factors bound any release from the auxiliary building without the EES credited.

4. Please provide the revised LOCA offsite dose results with the EES not credited.

RAI ARCB1-FHA-5 and ARCB1-FHA-6 Section 8 of Enclosure IV to letter dated January 17, 2017, "NRC Regulatory Issue Summary 2006-04 Comparison," states that the analysis conforms to Issue 1, which states that the amendment should identify and justify each change to the accident analysis. The licensee also states in the comments that the submittal identifies the changes to the licensing basis analysis and includes sufficient analysis detail to allow for results verification through independent calculations.

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792), Regulatory Position 5.1.3, "Assignment of Numeric Input Values," states, in part:

The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose.

Table A of Enclosure IV to the letter dated January 17, 2017, "Conformance with Regulatory Guide 1.183 Main Sections," states that the alternative source term (AST) analysis conforms to Regulatory Position 5.1.3, and that "[t]he numeric values that [were] chosen as inputs to the analyses required by 10 CFR 50.67 [were] selected with the objective of determining a conservative postulated dose."

In the letter dated June 19, 2018, in response to RAI ARCB1-FHA-5 and ARCB1-FHA-6, a revised analysis modeling the control room dose from a fuel handling accident in containment with an open personnel airlock is discussed. Additional information relating to the assumptions and inputs of this analysis is needed to enable the NRC staff to make a finding of compliance with 10 CFR 50.67 and 10 CFR 50.36. Please provide the following information regarding the new analysis:

1. AEC Research and Development Report NM-SR-10100, "Conventional Buildings for Reactor Containment," developed by Atomics International, is used to calculate the unfiltered inleakage through the various penetrations prior to the control room ventilation isolation signal. The document and equation used from this document may be a proposed change to your licensing basis. Please provide a technical justification for the use of this methodology for this intended application and why it is valid (i.e., an analysis showing how sensitive the control room dose is to varying amounts of unfiltered inleakage to show how important using this proposed methodology is, or justify why using NM-SR-10100 for determining the assumed unfiltered inleakage for control room habitability has been accepted by the NRC for your facility and is in your licensing basis),

or use clearly conservative or bounding and justified values of unfiltered inleakage to calculate a conservative postulated dose.

2. The revised analysis assumes that all inleakage (except that from ingress and egress) into and out of the control room following a control room ventilation isolation signal (emergency mode) is terminated. The justification provided for this assumption is that following a control room ventilation isolation signal the control room envelope is stated to be at a positive delta pressure of 0.25 inches of water pressure relative to the outside atmosphere. Also, the revised analysis proposes to assume that 300 cubic feet per minute (cfm) of air from the equipment room is filtered and transferred to the control room during the emergency mode of operation for the control room.

Utilization of the delta pressure as an indicator of the control room integrity has not been proven to be reliable. An inference is made from the delta pressure measurement that contamination will be unable to enter the control room if the control room is at a higher pressure than adjacent areas. This inference is based upon the assumption that the only source of pressurization flow to the control room is the pressurization flow through the emergency filtration unit. Experience with the nuclear industry control room integrity testing program since 1999 has shown that this may not be the case. Other unidentified

sources of air may be the origin of the pressurization flow. These unidentified sources of pressurization flow may originate from inleakage into the suction side of the fan or into ductwork located outside the control room that also traverses the control room. In the proposed analysis the equipment room is modelled and may contain vulnerabilities, such as fans that could provide sources of unfiltered inleakage into the control room.

Therefore, additional justification for the assumption of terminating the inleakage after the control room ventilation isolation is needed or the assumption needs to be revised and justified to show that this input determines a conservative postulated dose.

Also, please confirm the value of 300 cfm of filtered forced air flow from the equipment room to the control room during the emergency mode (or provide the value assumed) and justify the value assumed.

3. Chapter 3 of the WCGS UFSAR, Revision 30, Appendix 3A (ADAMS Accession Number ML17151A997), discusses the extent to which WCGS conforms to NRG-published regulatory guides. Exceptions to the guides are identified, and justification is presented or referenced. For RG 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," dated May 2003 (ADAMS Accession No. ML031490664), no exception to Regulatory Position 2.5, "lnleakage Test Acceptance Criteria," is noted.

This position states, in part:

Any analysis to demonstrate that a facility meets GDC 19 should include a value for inleakage that is due to ingress to and egress from the CRE [control room envelope]. This value is combined with the baseline test value for inleakage in the analyses. When integrity tests are performed to determine the CRE's integrity characteristics, the acceptance criterion for the test should be the licensing basis amount less the amount designated for ingress and egress. The staff considers 10 cfm as a reasonable estimate for ingress and egress for control rooms without vestibules.

The revised analysis proposes to assume 10 cfm unfiltered in leakage throughout the duration of the accident. The use of 1O cfm has generally been accepted to account for the sweeping action of the opening and closing of the door.

In the supplemental response to RAI ARCB1-FHA-5 by letter dated June 19, 2018, the unfiltered infiltration due to ingress and egress is stated to be 10 cfm throughout the duration of the fuel handling event. However, as noted above, the 10 cfm is to account for the sweeping action of the door and would not account for inflow into the equipment or control room due to a pressure gradient across the door caused by winds (as discussed in the section entitled "Maximum Wind Speed" in the June 19, 2018, supplemental response to RAI ARCB1-FHA-5). Not accounting for this inflow appears to be inconsistent with the stated conformance to select inputs for the analysis with the objective of determining a conservative postulated dose.

Therefore, please provide a revised value for the unfiltered inflow due to ingress and egress prior to the control room ventilation isolation signal, and justify the value used.

4. RG 1.183, Appendix B, Regulatory Position 5.3 states:

If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

Table C of Enclosure IV to the letter dated January 17, 2017, "Conformance with Regulatory Guide 1.183, Appendix B (Fuel Handling Accident)," states that the AST analysis conforms to Regulatory Position 5.3, and that "[t]he containment was assumed to be open and the release of radioactive material was modeled as a linear release over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period."

The supplementary response to RAI ARCB 1-CONTROL ROOM-6, by letter dated June 19, 2018, states, in part (emphasis added):

However, as documented within the supplemental response to ARCB1-FHA-5 and ARCB1-FHA-6 (contained within this letter), the FHA within containment with an open personnel air lock credited isolation of containment at two hours. Thus, in order to support the assumptions contained within the FHA within containment with an open personnel air lock analysis, markups of TS LCO [limiting condition for operation] 3.9.4, "Containment Penetrations," have been provided in Attachments Ill and IV. Attachments Ill and IV provide the Proposed Technical Specification Changes (Mark-up) and Revised Technical Specification Page, respectively.

The supplementary response dated June 19, 2018, to RAI ARCB1-FHA-6 states that TS LCO 3.9.4 changes are proposed to ensure that all the containment penetrations are isolated consistent with the safety analysis.

The NRC staff's understanding that in the fuel handling accident analysis (with the personnel airlock assumed to be open), the analysis assumes that the penetrations and the personnel airlock are assumed to be closed at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but the equipment hatch is assumed to be left open.

The NRC staff is unable to verify the calculated doses provided by WCNOC for this scenario (fuel handling accident in containment with an open personnel airlock) and it is not clear that the proposed markups of TS LCO 3.9.4 would be consistent with the personnel airlock and penetrations (other than the equipment hatch) being closed at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Please confirm that the fuel handling accident analysis conforms to Regulatory Position 5.1.3 and Appendix B to Regulatory Position 5.3, and that all the radioactive material in containment is released to the environment or auxiliary building over a 2-hour time period with the objective of calculating a conservative dose (such an analysis would consider the release through the pathway that maximizes the control room operator dose).

Note that with only the equipment hatch or a single penetration open winds can blow through the penetration while simultaneously exhausting the containment atmosphere

through the same penetration. Any open boundary can create a potential pathway for accident releases to the environment and should be accounted for in the accident analyses.

The proposed TS 3.9.4 does not explicitly require closure of the personnel door. It only requires that the personnel airlock be able to be closed, and the note proposed to be modified in TS 3.9.4 does not appear to have any impact on the equipment hatch, escape hatch, or personnel airlock.

Please justify how the proposed changes to the TS LCO 3.9.4 align with the proposed safety analysis (which assumes that the penetrations and personnel air lock are closed at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) or modify the TS LCO and the proposed safety analysis so that they are consistent.

Please state how the words "Penetration flow path(s) providing direct access from the containment to the outside atmosphere" are defined for TS 3.9.4.

RAI ARCB1-WT-5 According to the supplemental response to RAI ARCB1-WT-5 by letter dated June 19, 2018, the partition factor of 100 is removed from the calculations, and all iodine activity in the volume control tank is conservatively modeled to become airborne and is available for transfer to the waste gas decay tank. However, the values provided in Table 4.3-2a in Enclosure IV of the LAR to the letter dated January 17, 2017, does not appear to have been updated after the stated change in the assumed partition factor. Please provide the updates to Table 4.3-2a.

Note for all RAls: If changes are made (to the current proposed analyses as stated in the LAR) as a result of the following RAI questions, please provide the details of any revised analyses including the inputs, assumptions, methodology technical basis for the analysis and the results of the analysis. Also, please justify the assumptions and inputs used in the revised analysis.

Per Regulatory Position 1.5, "Submittal Requirements," "The staff recommends that licensees submit the affected Final Safety Analysis Report pages annotated with changes that reflect the revised analyses or submit the actual calculation documentation."

ML18304A105 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DRA/ARCB/BC NAME BSingal PBlechman KHsueh DATE 11/1/2018 11/1/2018 11/2/2018 OFFICE N RR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME RPascarelli BSingal DATE 11/5/2018 11/5/2018