ML12066A008
| ML12066A008 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 03/02/2012 |
| From: | Feintuch K Plant Licensing Branch II |
| To: | Sly C Dominion Energy Kewaunee |
| References | |
| TAC ME7110 | |
| Download: ML12066A008 (23) | |
Text
1 NRR-PMDA-ECapture Resource From:
Feintuch, Karl Sent:
Friday, March 02, 2012 6:07 PM To:
Craig D Sly Cc:
Jack Gadzala; Blumberg, Mark
Subject:
ME7110 Kewaunee Amendment Request Re: Chi-over-Q - AADB Request for Additional Information (RAI)
Attachments:
ME7110 AADB (Blumberg) RAII sent 2012-03-02.docx; ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx (DRAFT) REQUEST FOR ADDITIONAL INFORMATION KEWAUNEE POWER STATION LICENSE AMENDMENT REQUEST (TAC No. ME7110):
MODIFYING THE TECHNICAL SPECIFICATIONS (TS) AND THE CURRENT LICENSING BASIS (CLB)
TO INCORPORATE CHANGES TO THE CURRENT RADIOLOGICAL ACCIDENT ANALYSIS (RAA) OF RECORD (KNOWN AS CHI-OVER-Q)
DOCKET NO. 50-305 By letter dated August 30, 2011, Dominion Energy Kewaunee (DEK) submitted a license amendment request (LAR)-244 (ADAMS Accession No. ML11252A521) to revise the Kewaunee Power Station (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 proposed amendment would revise the current RAA for the design-basis accidents (DBAs) described in Chapter 14 of the KPS Updated Safety Analysis Report (USAR). This amendment would also fulfill a commitment made to the NRC in response to Generic Letter 2003-01, Control Room Habitability.
In the course of his technical review, the reviewer of the Accident Dose Branch (AADB) has requested further information items to enable completion of its respective Safety Evaluation efforts. These items are provided in draft form for you to review for clarification. We seek to confirm your understanding of the items and the determination of a firm date for response, typically within 30 days of the date of this Request for Additional Information (RAI), in this case sooner if practical. The items we seek are attached.
Please contact me by 03/05/2012 to confirm: (1) that the items are clear to you and to the responsive DEK staff without further discussion or (2) that a clarifying conference call is needed. Upon determination that the RAI items are clear and confirmation of when responses to these items are due, these draft RAI items will be considered to be in final form.
ME7110 is a complex project and we (Craig Sly of DEK and myself) have discussed methods for (1) improved movement of RAI information, (2) improved responsiveness to NRC staff requests, and (3) more flexibility for DEK to schedule RAI response activity, over that associated with more rigidly defined RAI milestone events.
This group of 17 AADB RAI items will be managed by the attached spreadsheet. This and subsequent RAI traffic will be tracked by an individual identifier to provide the associated response by the individualized request by date.
Docketing of this information by submittal under oath or affirmation will be managed by a reference to the associated ADAMS Accession No. (ML#) on the spreadsheet. Docketing will take place on groups of RAI item responses based on close completion schedules rather than close issuance schedules, as is now customary.
Thus, if DEK can respond with individual information in 5 days, the request by date will be shortened and will be received sooner than the rest of the items, although its docketing event might coincide with the original set of items or with those RAI items originating from another Technical Branch.
2 We will periodically assess when this new process is of mutual benefit while conforming to the regulation for processing amendment requests and their associated RAIs.
The attached AADB RAI items are assigned the following tracking numbers. The associated entries are defined in the Legend tab of the spreadsheet with the exception of the request by date information. The sent date is displayed until we can determine a target request by date. The RAI items of interest are designated:
- 1. ME7110-RAII-AADB-Blum-001-2012-03-02 (sent date)
- 2. ME7110-RAII-AADB-Blum-002-2012-03-02 (sent date)
- 3. ME7110-RAII-AADB-Blum-003-2012-03-02 (sent date)
- 4. ME7110-RAII-AADB-Blum-004-2012-03-02 (sent date)
- 5. ME7110-RAII-AADB-Blum-005-2012-03-02 (sent date)
- 6. ME7110-RAII-AADB-Blum-006-2012-03-02 (sent date)
- 7. ME7110-RAII-AADB-Blum-007-2012-03-02 (sent date)
- 8. ME7110-RAII-AADB-Blum-008-2012-03-02 (sent date)
- 9. ME7110-RAII-AADB-Blum-009-2012-03-02 (sent date)
- 10. ME7110-RAII-AADB-Blum-010-2012-03-02 (sent date)
- 11. ME7110-RAII-AADB-Blum-011-2012-03-02 (sent date)
- 12. ME7110-RAII-AADB-Blum-012-2012-03-02 (sent date)
- 13. ME7110-RAII-AADB-Blum-013-2012-03-02 (sent date)
- 14. ME7110-RAII-AADB-Blum-014-2012-03-02 (sent date)
- 15. ME7110-RAII-AADB-Blum-015-2012-03-02 (sent date)
- 16. ME7110-RAII-AADB-Blum-016-2012-03-02 (sent date)
- 17. ME7110-RAII-AADB-Blum-017-2012-03-02 (sent date)
Enclosure REQUEST FOR ADDITIONAL INFORMATION (RAI)
DOMINION ENERGY KEWAUNEE POWER STATION MODIFICATION TO TECHNICAL SPECIFICATION 3.3.7 (TAC# ME7110)
Accident Dose Branch (Dose Review Only)
Many of the following RAI questions might be answered by information already contained in the radiological accident analysis calculations. Where the information is already provided in the calculations it is acceptable to provide the calculations and state the location where the information located. When these calculations are provided to the staff it has been helpful because it has been found to increase the efficiency of the review.
Items are assigned the following tracking numbers:
ME7110-RAII-AADB-[Blum-001 to Blum-017 inclusive]-2012-0n-nn Blum-001 = RAI 1; Blum-002 = RAI 2; etc
RAI 1
ME7110-RAII-AADB-Blum-001-2012-03-02 (sent date)
, of the proposed license amendment request (LAR) (Adams Package No. ML11252A521), page 91 states:
The dose conversion factors used to calculate the TEDE doses and DE I-131 for the Steam Generator Tube Rupture accident were taken from Table 3.2-3 for the isotopes required by Regulatory Guide 1.183 for the SGTR analysis.
Table 3.2-3 provides the effective dose equivalent (EDE) and committed effective dose equivalent (CEDE) dose conversion factors for iodine. Therefore, according to the statement above, the total effective dose equivalent (TEDE) is calculated with either the EDE or CEDE dose conversion factors.
page 32 of the LAR states:
The dose conversion factors (DCFs) used to determine dose from iodine are from Federal Guidance Report No. 11 (FGR-11), Table 2.1 committed effective dose equivalent (CEDE) and the calculation of the Dose Equivalent I-131 from proposed technical specification surveillance are from FGR-11 Table 2.1 Thyroid Committed Dose Equivalent (CDE).
The two cited texts appear to conflict. Please clarify which dose conversion factors are used for each design basis (DBA) accident (CDE vs. CEDE). Since the TEDE is defined as the DDE plus the CEDE, please justify use of CDE dose conversion factors.
RAI 2
ME7110-RAII-AADB-Blum-002-2012-03-02 (sent date)
, page 45 states:
A reduction in airborne radioactivity in the containment by natural deposition within containment is credited. The model used is described in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, (Reference 25) and is incorporated into the RADTRAD-NAI computer code. This model is called the Powers model, set for the 10th percentile.
, page 43 also discusses credit for removal of aerosol by containment sprays.
Removal of aerosol by sprays and natural deposition are competing processes. Please justify crediting both spray removal and the proposed Powers natural deposition model. Describe how the Powers natural deposition model accounts for removal due to the spray model used. If any further credit for a reduction in aerosols is taken for any pathway, please provide a justification for that credit while considering the impact of any other removal mechanism credited. For example, with respect to spray removal and natural deposition ensure that the sprays do not remove aerosols that are also removed in the Powers model.
RAI 3
ME7110-RAII-AADB-Blum-003-2012-03-02 (sent date)
, page 46 states that the revised LOCA analysis contains some changes that include:
Replacement of the assumed 1% iodine evolution rate from RWST back-leakage to a conservative DF=100.
Justify this change and describe why it is conservative.
RAI 4
ME7110-RAII-AADB-Blum-004-2012-03-02 (sent date)
, page 49 states that whole body dose conversion factors are used in the LOCA dose calculation and that the reason for the change is Regulatory Guide (RG) 1.183. RG 1.183 does not discuss whole body dose conversion factors. Please clarify what was changed and justify the change.
RAI 5
ME7110-RAII-AADB-Blum-005-2012-03-02 (sent date)
, page 46 states that negative pressure in the shield building is established within 10 minutes. Please verify that negative pressure means 0.25 inch vacuum water gauge with one shield building ventilation system train operating (consistent with Technical Specification SR 3.6.8.2) or justify any proposed change.
RAI 6
ME7110-RAII-AADB-Blum-006-2012-03-02 (sent date)
, page 69 states:
The LOCA causes a Safety Injection (SI) signal, which also isolates the control room (per current Licensing Basis). The control room is isolated within 10 seconds after the SI signal.
Based on RG 1.183, the onset of the gap release does not start until 30 seconds post-LOCA.
Therefore, the control room will be isolated prior to the arrival of the radioactive release.
Technical Specification 3.4.16, RCS [reactor coolant system] Specific Activity allows radioactivity to be present in the RCS prior to an accident. The above justification, that states the control room will be isolated before the release of radioactivity, does not seem to consider that, by design, radioactivity may be present in the RCS prior to the gap release (at the start of the accident). Please provide a complete justification for not considering the impact of the RCS activity prior the release of gap activity for both the control room and offsite analyses.
RAI 7
ME7110-RAII-AADB-Blum-007-2012-03-02 (sent date)
, page 35 states:
Containment purge isolates within 37 seconds following the LOCA and is an insignificant contributor to control room and offsite dose.
, page 35 states:
KPS is a licensed leak before-break LBB plant (Reference 9). Per RG 1.183, the onset of gap release can be credited with a 10 minute delay for LBB. Containment purge isolation occurs within 37 seconds. Therefore, dose contribution from only TS RCS inventory is insignificant.
Reference 9 provides the citation for a letter to the NRC. Please provide a reference for the NRC safety evaluation which approved LBB methodology. Also, please provide a justification for applying the LBB methodology to the control room and offsite dose calculations.
ME7110-RAII-AADB-Blum-008-2012-03-02 (sent date)
RAI 8
Please explain why switchover to recirculation spray is not credited in the LOCA analysis. Also, please state if any operator action is credited in the assumption that the RWST switchover occurs at 0.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br />. If operator actions are credited please provide the NRC staff safety evaluation where these operator actions are approved for the design basis LOCA dose calculation.
RAI 9
ME7110-RAII-AADB-Blum-009-2012-03-02 (sent date)
, page 29 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 irradiated fuel. [emphasis added]
Regulatory Position (RP) 5.1.2 of 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 operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.
The licensees response to Question 5, of a letter dated November 8, 2011 (Adams Accession No. ML11318A205) provides proposed changes to limiting condition for operation (LCO) 3.7.10 to ensure the Control Room Envelope is isolated during movement of recently irradiated fuel.
The response to Question 5 and the license amendment submittal do not appear to provide a similar provision, consistent with RP 5.1.2 cited above, for ensuring that the alternative control room intake is closed during normal operation or following an accident.
RG 1.183, RP 5.1.2 provides credit for mitigation features that are required to be operable by technical specifications. Justify why credit for isolation of the alternative control room intake is assured (during normal operations or following an accident) by the proposed methods or propose a method consistent with RP 5.1.2.
RAI 10
ME7110-RAII-AADB-Blum-010-2012-03-02 (sent date)
, page 68 states:
Flows reduced from nominal values by a factor equal to the inverse of the partition coefficient derived from a DF of 100.
Why are the flows reduced by the partition factor rather than by using the decontamination factor (DF)? How are the DF and partition coefficient defined for this application (based upon volume or mass)?
RAI 11
ME7110-RAII-AADB-Blum-011-2012-03-02 (sent date)
, page B 3.9.6-3 and B3.9.6-4 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 proposed language seems to be contrary to the intent of the TSTF -312 [Technical Specification Task Force], Administratively Control Containment Penetrations [ADAMS Accession No. ML040620147]. TSTF-312 bases approval of the TSTF on whether, the hatch can and will be promptly closed. Per Title 10 of the Code of Federal Regulations Section 50.67 (10 CFR 50.67), the fission product release assumed for the design calculations are based upon a major accident that results in potential hazards not exceeded by any accident considered credible.
The need for a provision to forgo closure of the hatch appears to acknowledge that there is a credible scenario where the design source term is exceeded. Therefore, the source term used for the fuel handling accident does not appear to align with 10 CFR 50.67 which specifies the need for a source term that is not exceeded by any credible accident. Please justify the source term used for the fuel handling accident, remove the proposed provision to allow forgoing closure, or provide a source term for the fuel handling accident which is not exceeded by any accident considered credible.
RAI 12
ME7110-RAII-AADB-Blum-012-2012-03-02 (sent date)
For several design basis accidents the assumed time to isolate the control room is decreased (i.e., Attachment 4, Table 3.2-5, page 58). For example, the Loss of Coolant Accident assumes the control room isolation damper takes 10 seconds to close upon receipt of a safety injection signal, but the proposed value for control room isolation is zero seconds. Explain and justify why the proposed value for control room isolation for some accidents is less than the 10 seconds to close the isolation damper.
The revised time to isolate the control room for all accidents does not seem to include the time to start and load the diesel generators. Please justify that, given the worst case single failure, the isolation of the control room does not require diesel power and that the time to isolate the control room is not influenced by time to start and load the diesel.
RAI 13
ME7110-RAII-AADB-Blum-013-2012-03-02 (sent date)
The Technical Specification Ventilation Filter and Testing Program, 5.5.9 states that the High-Efficiency Particulate Air and charcoal absorbers are allowed to have a bypass of 1% by design. How is the allowed bypass of 1% accounted for in the design basis radiological calculations? If the bypass is not accounted for in the radiological design calculations please consider it in the design calculations or justify why it need not be considered.
RAI 14
ME7110-RAII-AADB-Blum-014-2012-03-02 (sent date)
, page 63, Figure 3.2-2 provides the effective filter efficiencies for filtered flow through the Auxiliary Building Special Ventilation zone filters. The 50 percent plate-out factor in the Auxiliary Building appears to have been used to derive the organic iodine filter efficiency.
When appropriate, plate out of iodine is typically only associated with elemental iodine and is conservatively not assumed for organic iodine. Please justify adjusting the organic iodine nominal iodine filter efficiencies by the 50 percent plate-out factor or remove the credit for organic iodine plate out.
RAI 15
ME7110-RAII-AADB-Blum-015-2012-03-02 (sent date)
, page 68, Figure 3.2-3 states that the unfiltered flow from the containment sump to the Reactor Water Storage Tank (RWST) is decreased by 50 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. While the containment sump is in direct contact with the containment atmosphere and the pressure of the containment atmosphere may decrease over time, the RWST backleakage may also be influenced by pumps that run during the loss of coolant accident. Please justify the 50 percent decrease in RWST backleakage after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
RAI 16
ME7110-RAII-AADB-Blum-016-2012-03-02 (sent date)
, page 73 states:
The core curies include a 6% increase to account for fuel management variations (493.6 +/- 10%
EFPD [effective full power days], average enrichment of 4.5 w/o +/- 10%, and core mass of 49.1 MTU [metric ton uranium] +/- 10%).
State whether the burnup of an assembly is limited to 493.6 + 10% EFPD. If not, justify why the assumed burnup is conservative for the fuel handling accident given the fuel is allowed to achieve higher burnups.
RAI 17
ME7110-RAII-AADB-Blum-017-2012-03-02 (sent date)
, page 84 states:
Based on the assumption that the fuel assembly will be horizontal once it comes to rest, it was determined that an assembly lying on the reactor vessel flange will have approximately 22.35 feet of water above the highest point of the assembly to the water surface. In the spent fuel pool, greater than 23 feet of water will exist.
The depth of 22.35 feet of water was evaluated to verify an effective decontamination factor of 200 using WCAP-7828 (Reference 27). Using the methods defined in the WCAP with conservative assumptions to minimize predicted decontamination factors for various depths of water, a DF of greater than 500 was determined for elemental iodine. The use of an overall effective DF of 200 was determined to be appropriate per RG 1.183.
Justify the assumption that the fuel assembly will be horizontal once it comes to rest for the accident in the containment. Are there any obstructions which could prevent it from becoming horizontal?
Provide all the input assumptions and the methodology used in the determination that a DF of 200 with less than 23 feet of water is appropriate. Justify why WCAP-7828 is appropriate for the fuel used at your facility.
ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx 1
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TAC Doc type Source TB Source TB Reviewer Request by date Status RAI Response ML#
MLnnnnnnnnn Description ME7110 RAII EICB Alva-001 1/27/2012 resp ML12031A138 dated 01/25/2012 1, In DEKs License Amendment Request (LAR)-210, DEK proposed incorporating the control room envelope operability and surveillance requirements, R-23 operability requirements, and the control room post-accident recirculation (CRPAR) system requirements into the KPS Technical Specification (TS) ensures the systems, structures, or components (SSCs) credited for mitigating the consequences of an accident for control room occupants were included in the TS. At the same time, DEK requesting removing crediting R-23 and the control room envelope boundary from the KPS Waste Gas Decay Tank (GDT) and Volume Control (VCT) rupture accident analysis, since it determined that occupant dose consequences are achieved without crediting the control room envelope boundary or the CRPAR system. Later DEK withdrew LAR-210. However, based on the information provided in LAR-210, it is not clear why DEK in LAR-244 is requesting deleting R-23 from the TS, even though in the accident analysis performed for both LARs, DEK stated that R-23 was not credited in the proposed accident analysis. Please explain the reason to remove R-23 and replace with analysis and manual operation of the isolation dampers.
ME7110 RAII EICB Alva-002 1/27/2012 resp ML12031A138 dated 01/25/2012
- 2. NUREG-0737, Clarification of TMI Action Plan Requirements, Item III.D.3.4, Control Room Habitability Requirements, required licensees to assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gas and that the plant can be safely operated or shutdown under design basis accident conditions. LAR proposed removing radiation monitor channel R-23 as a required channel for CRPAR initiation, modifying DEK previously approved by the NRC compliance with NUREG-0737. Please describe if R-23 is removed, how DEK will comply with NUREG-0737.
ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx 1
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TAC Doc type Source TB Source TB Reviewer Request by date Status RAI Response ML#
MLnnnnnnnnn Description 4
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ME7110 RAII EICB Alva-003 1/27/2012 resp ML12031A138 dated 01/25/2012
- 3. During the NRC staff review of LAR-210, EICB issued RAI January 30, 2008 letter (ADAMS Accession No. ML080280107). DEK provided a response on its April 3, 2008 letter (ADAMS Accession No. ML080950096); the response to question 1b included a logic diagram for operation of the control room ventilation radiation monitor. To assist NRC staff review, please address the following:
- a. Section 3.1.1 of Attachment 1 of LAR-244 (ADAMS Accession No. ML11252A521) states that radiation monitor R-23, as a single channel, initiates both trains of the CRPAR system and each SI train initiates the associated CRPAR fan and filtration unit train. If R-23 is removed from the logic, will it be necessary that both SI trains be actuated to initiate CRPAR fans, filtration unit trains, and close dampers ACC-1A, ACC-1B, ACC-2, and ACC-5?
- b. This logic shows that safety injection (SI) train A closes dampers ACC-1A and ACC-1B, and SI train B closes dampers ACC-2 and ACC-5. If R-23 is removed, how will dampers ACC-2 and ACC-5 close if the SI train B actuation signal fails?
- c. Provide a marked logic for the control room ventilation radiation monitor assuming that R-23 is removed from the logic.
ME7110 RAII EICB Alva-004 1/27/2012 resp ML12031A138 dated 01/25/2012
- 4. LAR-244 is requesting removal of R-23 from the CRPAR system. Please describe how DEK would reflect removal of R-23 from the CRPAR system in an update of the FSAR for the following items:
- a. Figure 9.6-6, Control Room Air Conditioning System-Flow Diagram, in the Final Safety Analysis Report (FSAR) shows R-23 location in the CRPAR system. Provide a marked diagram for an update of the FSAR after removal of R-23.
- b. Section 7.7.1, Control Room, in the FSAR describes how R-23 monitors and activates the control room ventilation.
ME7110 RAII EICB Alva-005 1/27/2012 resp ML12031A138 dated 01/25/2012
- 5. LAR-244, Attachment 1, Section 4.2.3 and Attachment 4, Section 2.7 state that revised radiological accident analysis (RAA) credits R-23 to limit consequences of the Locked Rotor Action (LRA) and Fuel Handling Accident (FHA). However, the RAA approved in license amendment 190 (current radiological analysis of record for KPS) credited R-23 high radiation signal for mitigating the radiological consequences to control room occupants for the LRA, GDT and VCT Rupture, and FHA. Please explain why the revised RAA (submitted in LAR-244) does not state whether credit for R-23 is considered for mitigating GDT and VCT rupture.
ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx 1
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MLnnnnnnnnn Description 7
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ME7110 RAII EICB Alva-006 1/27/2012 resp ML12031A138 dated 01/25/2012
- 6. LAR-244, Section 4.2.3 describes that removal of R-23 would require manual actions to ensure post-accident control room dose is maintained within limits and are required to limit consequences of the FHA and LRA events. Note that the current accident analysis does not credit operator action to isolate the control room during for FHA. Attachment 3, Section B.3.3.7 states that manual actuation of the CRPAR System is a backup for the SI signal actuation. To assist NRC staff review, please address the following:
- a. Manual actuation is not part of the logic diagram for operation of the control room ventilation radiation monitor (FSAR Figure 9.6-6). Please clarify if this would be included in the logic diagram.
- b. SI signal is not considered for all accident events (i.e., FHA, LRA, and GDT/VCT ruptures dont consider SI). In these cases manual action would be required. Please clarify if this would be included in the logic diagram.
ME7110 RAII EICB Alva-007 1/27/2012 resp ML12031A138 dated 01/25/2012
- 7. LAR-244, Attachment 4, Section 2.7, 3.3.1, and 3.6.1, state that full control room isolation require action by the operator to close monitor dampers that are not included in the isolation logic (of the control room ventilation radiation monitor). This was not discussed in previous LARs or in FSARs. Please explain the following:
- a. Where is this information described? Provide a logic diagram and a description for operation of all dampers required for the control room ventilation radiation system.
ME7110 RAII AHPB Lapin-001 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 1. Explain why it was found to be preferable to add manual actions rather than upgrade the quality classification and redundancy of Radiation Monitor R-23.
ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx 1
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MLnnnnnnnnn Description 10 11 12 13 ME7110 RAII AHPB Lapin-002 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 2. Among the proposed changes is a change to LCO 3.9.6.a to allow the containment equipment hatch to be open during handling of recently irradiated fuel when measures are in place which ensure the capability to close equipment hatch in the event of a fuel handling accident (FHA). As described, closing the equipment hatch requires special tools and equipment, such as, a trolley, a jactuator, chainfalls, etc.
- a. How will personnel ensure that all required tools and equipment needed to close the equipment hatch are pre-staged/available and operable?
- b. How will personnel know whether and what kind of radiation protection equipment and clothing is needed?
- c. Is a written procedure required and available?
- d. Is training provided to all personnel who may be called upon to close the equipment hatch?
ME7110 RAII AHPB Lapin-003 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 3. A proposed new Note, applicable to LCO 3.9.6.c, would allow penetration flow paths providing direct access from the containment to outside atmosphere to be opened under administrative controls. How will each of the administrative controls be implemented:
- a. How will containment penetration status be communicated? (to the CR and in-plant personnel)
- b. How will designated personnel know their assigned penetration(s)?
- c. How will designated personnel be cautioned about obstructions?
- d. How will Operations know that designated personnel are at their posts?
ME7110 RAII AHPB Lapin-004 3/2/2012, slipped from 2/13/2012 resp MLnnnnnnnnn dated 02/27/2012 4.
- a. Does Radiation Monitor R-23 perform any functions other than the isolation function that is being removed?
- b. If yes, what functions will remain?
- c. If no, will all controls, displays, and logic interfaces associated with R-23 be physically removed?
ME7110 RAII AHPB Lapin-005 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 5. In Attachment 5 of the licensees revised submittal, it is stated that: Operations personnel were included in the walkdown of the control room.
- a. Was at least one crew included in the walkdown?
- b. If not, what plans are being made to validate the procedures, training, and physical interfaces with a representative sample of operators, i.e., at least one crew.
ME7110 Chi-over-Q RAI status Tracking 2012-03-07.xlsx 1
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MLnnnnnnnnn Description 14 15 16 17 18 ME7110 RAII AHPB Lapin-006 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 6. The revised RAA credits manual initiation of the CRE isolation within 60 minutes of the occurrence of an LRA, and initiation of the Control Room Post Accident Recirculation (CRPAR) system within 20 minutes of occurrence of a FHA and within [60] minutes of an LRA. [In its response the licensee will clarify the differences between the current and requested licensing bases. Reviewer Lapinsky corrected the time to 60 minutes.]
- a. How were these completion times estimated?
- b. What are the actual times or the estimated required times for these actions?
- c. How much margin is built into the estimates of completion times?
ME7110 RAII AHPB Lapin-007 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012 7.[Deleted item] For FHA and LRA, how was ALARA factored into the mitigation strategies? Will operator training address the integration of ALARA into the mitigation strategies? [This item is deleted by Reviewer Lapinski]
ME7110 RAII AHPB Lapin-008 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 8. The licensee stated in Attachment 5 of the revised submittal, that The appropriate modifications to plant procedures will be made as part of the implementation of this amendment request. Identify all procedure changes that will be made in support of this LAR. Include the procedure numbers and titles of the affected [emergency operating]
procedures. [With the additional words emergency operating Reviewer Lapinski clarified the scope of the requested procedures]
ME7110 RAII AHPB Lapin-009 3/2/2012 (slipped from 2/13/2012) resp MLnnnnnnnnn dated 02/27/2012
- 9. Describe any changes to training that are necessary to support this LAR.
ME7110 RAII SRXB Sun-001 3/5/2012 (slipped from 2/28/2012) firm
- 1. In an email message from Craig Sly (DEK) to Karl Feintuch (USNRC) dated September 12, 2011 12:44 PM, DEK provided some information pertaining to the percentage (%) of Failed Fuel Following the Accident. SRXB seeks to apply information contained in the file response 9-12-11.pdf (one among six pdf attachments to the email of September 12, 2011 12:44, all of which are included with this message, for completeness). SRXB is providing assistance to another Technical Branch rather than using this information to prepare a safety evaluation of the requested licensing action. In your RAI Response to this item, please provide file response 9-12-11.pdf.
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MLnnnnnnnnn Description 19 20 ME7110 RAII SRXB Sun-002 3/5/2012 (slipped from 2/28/2012) firm
- 2. Application Attachment 4, page 154 indicates that the actuation time of the safety injection (SI) signal (in seconds) is changed from 52.5 to 240 during rod ejection accident (REA). The reason for the change, as stated by the licensee, is that the delay of the SI signal is conservative. The Current License Basis (CLB) assumption is based on a 2-inch diameter break. The REA is specified to have a smaller 1.6 inch diameter break. The SI signal generated from a 1-inch diameter break is 240 seconds.
It is not clear why a longer delay time of the actuation of the SI signal is conservative for the REA dose analysis.
Please provide justification of the longer SI actuation delay time used in the REA dose analysis.
ME7110 RAII AADB Blum-001 3/2/2012 draft, of the proposed license amendment request (LAR) (Adams Package No.
ML11252A521), page 91 states:
The dose conversion factors used to calculate the TEDE doses and DE I-131 for the Steam Generator Tube Rupture accident were taken from Table 3.2-3 for the isotopes required by Regulatory Guide 1.183 for the SGTR analysis.
Table 3.2-3 provides the effective dose equivalent (EDE) and committed effective dose equivalent (CEDE) dose conversion factors for iodine. Therefore, according to the statement above, the total effective dose equivalent (TEDE) is calculated with either the EDE or CEDE dose conversion factors. page 32 of the LAR states:
The dose conversion factors (DCFs) used to determine dose from iodine are from Federal Guidance Report No. 11 (FGR-11), Table 2.1 committed effective dose equivalent (CEDE) and the calculation of the Dose Equivalent I-131 from proposed technical specification surveillance are from FGR-11 Table 2.1 Thyroid Committed Dose Equivalent (CDE).
The two cited texts appear to conflict. Please clarify which dose conversion factors are used for each design basis (DBA) accident (CDE vs. CEDE). Since the TEDE is defined as the DDE plus the CEDE, please justify use of CDE dose conversion factors.
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MLnnnnnnnnn Description 21 22 23 ME7110 RAII AADB Blum-002 3/2/2012 draft, page 45 states:
A reduction in airborne radioactivity in the containment by natural deposition within containment is credited. The model used is described in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, (Reference 25) and is incorporated into the RADTRAD-NAI computer code. This model is called the Powers model, set for the 10th percentile., page 43 also discusses credit for removal of aerosol by containment sprays.
Removal of aerosol by sprays and natural deposition are competing processes. Please justify crediting both spray removal and the proposed Powers natural deposition model.
Describe how the Powers natural deposition model accounts for removal due to the spray model used. If any further credit for a reduction in aerosols is taken for any pathway, please provide a justification for that credit while considering the impact of any other removal mechanism credited. For example, with respect to spray removal and natural deposition ensure that the sprays do not remove aerosols that are also removed in the Powers model.
ME7110 RAII AADB Blum-003 3/2/2012 draft, page 46 states that the revised LOCA analysis contains some changes that include:
Replacement of the assumed 1% iodine evolution rate from RWST back-leakage to a conservative DF=100.
Justify this change and describe why it is conservative.
ME7110 RAII AADB Blum-004 3/2/2012 draft, page 49 states that whole body dose conversion factors are used in the LOCA dose calculation and that the reason for the change is Regulatory Guide (RG) 1.183.
RG 1.183 does not discuss whole body dose conversion factors. Please clarify what was changed and justify the change.
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MLnnnnnnnnn Description 24 25 ME7110 RAII AADB Blum-005 3/2/2012 draft, page 46 states that negative pressure in the shield building is established within 10 minutes. Please verify that negative pressure means 0.25 inch vacuum water gauge with one shield building ventilation system train operating (consistent with Technical Specification SR 3.6.8.2) or justify any proposed change.
ME7110 RAII AADB Blum-006 3/2/2012 draft, page 69 states:
The LOCA causes a Safety Injection (SI) signal, which also isolates the control room (per current Licensing Basis). The control room is isolated within 10 seconds after the SI signal.
Based on RG 1.183, the onset of the gap release does not start until 30 seconds post-LOCA. Therefore, the control room will be isolated prior to the arrival of the radioactive release.
Technical Specification 3.4.16, RCS [reactor coolant system] Specific Activity allows radioactivity to be present in the RCS prior to an accident. The above justification, that states the control room will be isolated before the release of radioactivity, does not seem to consider that, by design, radioactivity may be present in the RCS prior to the gap release (at the start of the accident). Please provide a complete justification for not considering the impact of the RCS activity prior the release of gap activity for both the control room and offsite analyses.
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MLnnnnnnnnn Description 26 27 ME7110 RAII AADB Blum-007 3/2/2012 draft, page 35 states:
Containment purge isolates within 37 seconds following the LOCA and is an insignificant contributor to control room and offsite dose., page 35 states:
KPS is a licensed leak before-break LBB plant (Reference 9). Per RG 1.183, the onset of gap release can be credited with a 10 minute delay for LBB. Containment purge isolation occurs within 37 seconds. Therefore, dose contribution from only TS RCS inventory is insignificant.
Reference 9 provides the citation for a letter to the NRC. Please provide a reference for the NRC safety evaluation which approved LBB methodology. Also, please provide a justification for applying the LBB methodology to the control room and offsite dose calculations., page 35 states:
Containment purge isolates within 37 seconds following the LOCA and is an insignificant contributor to control room and offsite dose., page 35 states:
KPS is a licensed leak before-break LBB plant (Reference 9). Per RG 1.183, the onset of gap release can be credited with a 10 minute delay for LBB. Containment purge isolation occurs within 37 seconds. Therefore, dose contribution from only TS RCS inventory is insignificant.
Reference 9 provides the citation for a letter to the NRC. Please provide a reference for ME7110 RAII AADB Blum-008 3/2/2012 draft Please explain why switchover to recirculation spray is not credited in the LOCA analysis.
Also, please state if any operator action is credited in the assumption that the RWST switchover occurs at 0.91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br />. If operator actions are credited please provide the NRC staff safety evaluation where these operator actions are approved for the design basis LOCA dose calculation.
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MLnnnnnnnnn Description 28 29 ME7110 RAII AADB Blum-009 3/2/2012 draft, page 29 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 irradiated fuel. [emphasis added]
Regulatory Position (RP) 5.1.2 of 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 operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.
The licensees response to Question 5, of a letter dated November 8, 2011 (Adams Accession No. ML11318A205) provides proposed changes to limiting condition for operation (LCO) 3.7.10 to ensure the Control Room Envelope is isolated during movement of recently irradiated fuel. The response to Question 5 and the license amendment submittal do not appear to provide a similar provision, consistent with RP 5.1.2 cited above, for ensuring that the alternative control room intake is closed during normal ME7110 RAII AADB Blum-010 3/2/2012 draft, page 68 states:
Flows reduced from nominal values by a factor equal to the inverse of the partition coefficient derived from a DF of 100.
Why are the flows reduced by the partition factor rather than by using the decontamination factor (DF)? How are the DF and partition coefficient defined for this application (based upon volume or mass)?
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MLnnnnnnnnn Description 30 31 ME7110 RAII AADB Blum-011 3/2/2012 draft, page B 3.9.6-3 and B3.9.6-4 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 proposed language seems to be contrary to the intent of the TSTF -312
[Technical Specification Task Force], Administratively Control Containment Penetrations
[ADAMS Accession No. ML040620147]. TSTF-312 bases approval of the TSTF on whether, the hatch can and will be promptly closed. Per Title 10 of the Code of Federal Regulations Section 50.67 (10 CFR 50.67), the fission product release assumed for the design calculations are based upon a major accident that results in potential hazards not exceeded by any accident considered credible.
The need for a provision to forgo closure of the hatch appears to acknowledge that there is a credible scenario where the design source term is exceeded. Therefore, the source term used for the fuel handling accident does not appear to align with 10 CFR 50.67 which specifies the need for a source term that is not exceeded by any credible accident. Please justify the source term used for the fuel handling accident, remove the proposed provision to allow forgoing closure, or provide a source term for the fuel handling accident which is not exceeded by any accident considered credible.
ME7110 RAII AADB Blum-012 3/2/2012 draft For several design basis accidents the assumed time to isolate the control room is decreased (i.e., Attachment 4, Table 3.2-5, page 58). For example, the Loss of Coolant Accident assumes the control room isolation damper takes 10 seconds to close upon receipt of a safety injection signal, but the proposed value for control room isolation is zero seconds. Explain and justify why the proposed value for control room isolation for some accidents is less than the 10 seconds to close the isolation damper.
The revised time to isolate the control room for all accidents does not seem to include the time to start and load the diesel generators. Please justify that, given the worst case single failure, the isolation of the control room does not require diesel power and that the time to isolate the control room is not influenced by time to start and load the diesel.
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MLnnnnnnnnn Description 32 33 34 35 ME7110 RAII AADB Blum-013 3/2/2012 draft The Technical Specification Ventilation Filter and Testing Program, 5.5.9 states that the High Efficiency Particulate Air and charcoal absorbers are allowed to have a bypass of 1%
by design. How is the allowed bypass of 1% accounted for in the design basis radiological calculations? If the bypass is not accounted for in the radiological design calculations please consider it in the design calculations or justify why it need not be considered.
ME7110 RAII AADB Blum-014 3/2/2012 draft, page 63, Figure 3.2-2 provides the effective filter efficiencies for filtered flow through the Auxiliary Building Special Ventilation zone filters. The 50 percent plate-out factor in the Auxiliary Building appears to have been used to derive the organic iodine filter efficiency. When appropriate, plate out of iodine is typically only associated with elemental iodine and is conservatively not assumed for organic iodine. Please justify adjusting the organic iodine nominal iodine filter efficiencies by the 50 percent plate-out factor or remove the credit for organic iodine plate out.
ME7110 RAII AADB Blum-015 3/2/2012 draft, page 68, Figure 3.2-3 states that the unfiltered flow from the containment sump to the Reactor Water Storage Tank (RWST) is decreased by 50 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. While the containment sump is in direct contact with the containment atmosphere and the pressure of the containment atmosphere may decrease over time, the RWST b
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ME7110 RAII AADB Blum-016 3/2/2012 draft, page 73 states:
The core curies include a 6% increase to account for fuel management variations (493.6 +/-
10% EFPD [effective full power days], average enrichment of 4.5 w/o +/- 10%, and core mass of 49 1 MTU [metric ton uranium] +/- 10%)
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MLnnnnnnnnn Description 36 37 38 ME7110 RAII AADB Blum-017 3/2/2012 draft, page 84 states:
Based on the assumption that the fuel assembly will be horizontal once it comes to rest, it was determined that an assembly lying on the reactor vessel flange will have approximately 22.35 feet of water above the highest point of the assembly to the water surface. In the spent fuel pool, greater than 23 feet of water will exist.
The depth of 22.35 feet of water was evaluated to verify an effective decontamination factor of 200 using WCAP-7828 (Reference 27). Using the methods defined in the WCAP with conservative assumptions to minimize predicted decontamination factors for various depths of water, a DF of greater than 500 was determined for elemental iodine. The use of an overall effective DF of 200 was determined to be appropriate per RG 1.183.
Justify the assumption that the fuel assembly will be horizontal once it comes to rest for the accident in the containment. Are there any obstructions which could prevent it from becoming horizontal?
Provide all the input assumptions and the methodology used in the determination that a DF of 200 with less than 23 feet of water is appropriate. Justify why WCAP-7828 is appropriate for the fuel used at your facility.
ME7110 RAII ME7110 RAII AADB Brow-00n mid-03/2012 (from Leta) tentative Check with Leta in mid-March for anticipated RAII.
TAC Doc type Source Tech Branch Source Reviewer and Ser#
Request by date Status ML#
Description ME7110 RAII EICB Alva-001 12/29/2011 draft MLnnnnnnnnn Assigned TAC No. This may be different than ME7110 if future sub-projects need other TAC No.
RAII RAII = Request for information item RAIR = Request for information response Suppl = docketed supplement EICB EICB = Instrumentation and Control Branch (Alvarado)
AHPB = Health Physics and Human Performance Branch (Lapinsky)
AADB = Accident Dose Branch (Blumberg, Brown)
ITSB = Technical Specifications Branch (no SE expected) (Hamm)
SCVB = Containment and Ventilation Branch (Torres)
SRXB = Reactor Systems Branch (no SE expected) (Sun, Guzzetta)
Alva-001 Alva-nnn = Items from Reviewer Alvarado (EICB)
Lapin-nnn = Items from Reviewer Lapinsky (AHPB)
Sun-nnn = Items from Reviewer Sun (SRXB)
Blum-nnn = Items from Reviewer Blumberg (AADB) 12/29/2011 Request by date (updated as mutually understood by PM, Reviewer and Licensee; maintained by PM and Licensee) draft draft = as issued prior to clarification pending = pending verification that no clarification is requested firm = as mutually understood and to be respond to by licensee resp = has dated response; is planned to or is already docketed ML#
If blank, then = not yet docketed in ADAMS If RAIR, then = docketed ML#
If RAII, then = issued RAI If Suppl, then = docketed supplement letter containing no RAIR information