ML18194A857

Korea Hydro & Nuclear Power Co., Ltd - Revised Response to RAI 432-8377 for Question 19-64 (Rev.7)

ML18194A857
Person / Time
Site:	05200046
Issue date:	07/13/2018
From:	Korea Electric Power Corp, Korea Hydro & Nuclear Power Co, Ltd
To:	Office of New Reactors
Shared Package
ML18194A854	List: ML18194A855 ML18194A857
References
KAW-18-0104, MKD/NW-18-0104L
Download: ML18194A857 (9)

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Non-Proprietary 19-64_Rev.7 - 1 / 5 KEPCO/KHNP REVISED RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION APR1400 Design Certification Korea Electric Power Corporation / Korea Hydro & Nuclear Power Co., LTD Docket No.52-046 RAI No.: 432-8377 SRP Section: 19 - Probabilistic Risk Assessment and Severe Accident Evaluation Section: 19 Application Section: 19 Date of RAI Issue: 03/08/2016 Question No. 19-64 10 CFR 52.47(a)(23) states that a design certification (DC) application for light-water reactor designs must contain an FSAR that includes a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass.

Provide the following regarding the discussion on ex-vessel steam explosion as provided in APR1400-E-P-NR-14003-P, Severe Accident Analysis Report, Rev. 0, Appendix D, Severe Accident Analysis Report for FCI and revise the design control document (DCD) as necessary

a. Figure 4-2 shows one dimensional nodalization of TEXAS-V for the ex-vessel steam explosion in the APR1400 RPV. Explain and justify using one-dimensional analysis for the cavity which has a large cross sectional area.

b. TEXAS-V code being one dimensional, assumed diameter for the mixing region would significantly affect the premixing results as shown in Figures 4-3 and 4-4. As stated in Section 4.5.3, mixing has an area of 7 m2, which is significantly larger than the cross-sectional area of the melt jet of 0.2 m2. Justify using one-dimensional analysis.

c. Provide the initial void fraction of the melt jet.

d. Explosion energy generated depends on melt fraction and void fraction before triggering an explosion, which are functions of time after the initiation of premixing.

Provide the timing and justify the time at which triggering was assumed.

e. Table 4-17 showing cavity structural analysis results lists number of cracks as 47,073 EA and a maximum crack width of 0.027 in. with a remark of considerable concrete damage. However, Table 5-1 remarks that ex-vessel steam explosion has no threat to APR1400 design. Explain what is meant by EA in listing number of cracks and why a

Non-Proprietary 19-64_Rev.7 - 2 / 5 KEPCO/KHNP possible concrete damage with 47,073 cracks would not cause a threat to the APR1400 cavity design.

Response - (Rev. 7)

a. TEXAS-V code is a one-dimensional code and the user is expected to input the area of the node as a user-defined parameter, ARIY, which corresponds to the cross-sectional area of the cavity. This user parameter is used to specify the amount of coolant at given node and its cooling capacity, in consequently. ARIY plays an important role in determining the vapor fraction during the mixing phase as well as the numerical convergence.

Instead of the actual cavity cross-sectional area (approximately 80 m2), ARIY is set to give a maximum energetic load based on the energy index concept, i.e. when the ratio of the given melts initial thermal energy and the coolant energy places in the optimal range the explosion pressure can be maximized. In other words, if user introduces the actual cavity floor area of APR1400 as ARIY, the excess of cooling capacity can produce the higher void fraction and eventually it can lead to the limited energetic load due to one-dimensional characteristics of TEXAS-V code. In contrary for the case with too small ARIY, the certain amount of the melt thermal energy may remain inside the melt and it can restrict the higher load.

Energy index, , is defined as below:

TS where, M is mass and e is specific energy. Subscript m and c represents melt and coolant, respectively. In the equation, Ac means the cross-sectional area of coolant (i.e.

ARIY). Under the given initial and boundary conditions the peak pressure and impulse predicted by TEXAS-V according to various ARIY is illustrated in the following figure.

The vapor fraction along the energy index is also given in the figure.

TS

Non-Proprietary 19-64_Rev.7 - 3 / 5 KEPCO/KHNP It is seen from this figure that the explosion energy increases along the energy index, and it begins to fluctuate as it reaches a transition region. After the transition region, the explosion energy decreases abruptly. As the index increases, the total vapor fraction in the cavity coolant also increases, leading to higher energetics. However, after the index exceeds a certain value (the optimal value), the vapor fraction increases much faster and the explosion energetics are reduced. This indicates that the vapor fraction and the energy index have a non-linear relationship, reflecting the jet break-up and several other explosion dynamical phenomena. If the vapor fraction increases rapidly, the explosion energy decreases quickly. As mentioned above, the calculated explosion energy fluctuates substantially in the transition region, due to the vapor fraction intermittently exceeding a certain threshold value. In this region, the area effect is minor, and the explosion energy is driven by the vapor fraction in accordance with the axial dynamic effects. Hence, the selection of the energy index value from the region that precedes the transition region appears to be a reasonable way to achieve a stable, converged solution. Based on this selected ARIY of 7.0 m2, energy index was calculated to be [ ]TS, as below.

m = [ ]TS (density of melt), Am=Dj2/4=0.071 m2 (melt jet area), um = 4 m/s (melt release speed)*tmix = [ ]TS (mixing time predicted from TEXAS-V),

TS cp,m = [ ] (specific heat of melt), Tm = 3000 K (initial melt temperature),

Tw= 351 K (water temperature), hfg,m= [ ]TS (heat of fusion of melt),

TS 2 c= [ ] , Ac = 7.0 m , Hc = 6.4 m (water pool height),

cp,c = [ ]TS, Tsat = 393 K (saturation temperature of water at cavity pressure, 2 bar), hfg,c = [ ]TS. Using given properties, = [ ]TS.

The influence of the large cross-sectional area of the cavity is eliminated in TEXAS-V study in this way from the conservatism standpoint. The selection of ARIY based on the energy index concept is appended in APR1400-E-P-NR-14003-P/NP, Severe Accident Analysis Report, Rev.1, Appendix D, Severe Accident Analysis Report for FCI, as shown in Attachment 6.

Predicted pressure profiles by TEXAS-V are then applied to the structural integrity assessment conservatively. The methodology, input pressure loading, and the results of analyses for APR1400 are also appended in APR1400-E-P-NR-14003-P/NP, Severe Accident Analysis Report, Rev.1, Appendix D, Severe Accident Analysis Report for FCI, as shown in Attachment 6.

In addition, with regarding In-Vessel Retention and External Reactor Vessel Cooling (IVR-ERVC) strategy, we did not give credit IVR-ERVC system at present. The adoption of this strategy is related with Accident Management (AM). Severe Accident Management Guideline (SAMG) contingent to activation of IVR-ERVC is also constructed in AM procedure. For example, applicant of APR1400 plants in United Arab Emirates (UAE) does not select IVR-ERVC strategy. In the case of Korean domestic APR1400 plants such as Shin Kori Units 3&4 the applicant selects the IVR-ERVC as SAMG. Then the studies for performance of IVR-ERVC and consequences from IVR-ERVC failure were done during Operating License Stage because they are highly related with Operator actions, ESFs possibility, and so on. Therefore the evaluation of the steam explosion load and consequential structural integrity

Non-Proprietary 19-64_Rev.7 - 4 / 5 KEPCO/KHNP assessment under the IVR-ERVC situation will be performed as COL item. To clarify this concern, DCD Section 19.2.7 and Table 1.8-2 is revised as shown in Attachment 5.

b. As discussed in Response a., ARIY represents the node area not the mixing area. The editorial error will be revised as Attachment (mixing replaced with node).

c. For melt jet, the initial void fraction is set to be zero.

d. The steam explosion energetics depends largely upon the corium mass participated in the interaction. Therefore, it is assumed that the artificial trigger is provided by the corium jet contact at the bottom of the reactor cavity. The less conservative results will be obtained if the corium jet is triggered before or after the bottom contact of corium leading edge to the cavity floor.

e. EA means each and it is already represented in number of cracks, so the table is revised like Table 4-20 in Attachment 6. The numbers of cracks described in Table 4-20 shown in Attachment 6 include all cracks having from a very small crack width to maximum 0.818 in crack width. In addition, there are no through cracks in concrete. It means that the possible concrete damage did not cause a threat to the cavity design even though cracks seem quantitatively much. In the scope of leakage, the damage of liner plate rather than concrete crack is more important. By ex-vessel steam explosion, the maximum stress in the liner plate is 55.5 ksi which is less than the ultimate tensile strength (75 ksi). In addition, the maximum effective plastic strain is around 1.5% which is less than the failure strain criteria of liner plate (5%). Therefore, it can be concluded that the APR1400 cavity structure remains intact from the ex-vessel steam explosion.

Section 19.2.3.3.5.2.2 of APR1400 DCD Rev.1 is revised to reflect the items in this RAI and given in Attachment 7.

Impact on DCD The changes that were proposed in the original response to this RAI have been incorporated into Revision 2 of the DCD; therefore, only the pages containing proposed changes as a result of Revision 7 of this response are included in the Attachment.

Section 19.2.5.1.2.1, 19.2.7 and Table 1.8-2 will be revised as shown in Attachment.

Impact on PRA There is no impact on the PRA.

Impact on Technical Specifications There is no impact on the Technical Specifications.

Non-Proprietary 19-64_Rev.7 - 5 / 5 KEPCO/KHNP Impact on Technical/Topical/Environmental Reports The changes that were proposed in the original response to this RAI have been incorporated into Revision 2 of the APR1400-E-P-NR-14003-P/NP; therefore, there is no impact on any Technical, Topical or Environmental Report..

RAI 432-8377 - Question 19-64_Rev.7 Non-Proprietary Attachment (1/4)

APR1400 DCD TIER 2 RAI 432-8377 - Qiuestion 19-64_Rev.7 The SIS is isolated during LPSD operations; however, at least two trains are kept in standby so that SI can be available to provide core cooling if necessary; when needed, the SIS is manually activated by the operators.

As described in the shutdown evaluation report, the minimum makeup flow of 481 L/min (127 gpm) is required to keep the core covered for the loss of DHR during the mid-loop operation. This makeup flow is calculated based on the boil-off condition with the decay heat of 4 days after shutdown. Even though each charging pump has a rated flow rate of 155 gpm during normal operation, one charging pump flowrate is limited to maximum 150 gpm during mid-loop operation to prevent boron dilution, which is greater than the minimum required makeup flow (127 gpm). Therefore, during POS 5, one charging pump is needed to keep the core covered. The minimum makeup flow depends on the decay heat level. Therefore, during the POS 3 and POS 4, the required charging pump to keep the core covered can be two in accordance with the decay heat level.

19.2.5.1.2 Retain the Core within the Reactor Vessel 19.2.5.1.2.1 During Operations at Power The onset of core damage is identified when core-exit temperature reaches 922.04 K (1,200 °F). The primary way to terminate the progress of core damage is inject water into the reactor vessel. This can be achieved by operation of the SI, SC, or CS pumps. Once core relocation to the lower plenum occurs, another option available to prevent accident progression is ex-vessel cooling.

If the SI pumps can be recovered prior to reactor vessel melt-through, injection may be capable of cooling the core and preventing failure of the reactor vessel. If the SI pumps cannot be recovered, the RCS can be depressurized using the POSRVs to allow injection using the SC or CS pumps. Successful cooling of the core depends on the configuration of core (i.e., whether it is intact, melted, relocated).

Once the core has relocated to the lower plenum, ex-vessel cooling can be established by the operators. This entails using the SC pumps to pump water from the IRWST into the reactor cavity to submerge the reactor vessel lower head in water. This action has the potential to remove decay heat through the lower head wall and prevent vessel failure.

If In-Vessel Retention and External Reactor Vessel Cooling is proposed for accident management by either the COL applicant and/or holder, then the COL applicant and/

or holder is to develop and submit to the NRC an evaluation of the effects of higher water level in the cavity on steam explosion loading (COL 19.2(3)).

19.2-51 Rev. 2

RAI 432-8377 - Question 19-64_Rev.7 Non-Proprietary Attachment (2/4)

APR1400 DCD TIER 2 RAI 432-8377 - Qiuestion 19-64_Rev.7 COL 19.2(3) The COL applicant and/or holder is to develop and submit an accident management plan including the evaluation of the effect of higher water level in the cavity on steam explosion loading when using In-Vessel Retention and External Reactor Vessel Cooling for accident management.

19.2.8 References Replace with "A" on next next page

1. SECY-93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, U.S. Nuclear Regulatory Commission, April 1993.

2. 10 CFR Part 100, Reactor Site Criteria, U.S. Nuclear Regulatory Commission.

3. 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Title 10, Code of Federal Regulations, U.S. Nuclear Regulatory Commission.

4. 10 CFR 50.34, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.

5. 10 CFR 50.44, Combustible Gas Control for Nuclear Power Reactors, U.S. Nuclear Regulatory Commission.

6. Regulatory Guide 1.216, Containment Structural Integrity Evaluation for Internal Pressure Loadings above Design Basis Pressure, U.S. Nuclear Regulatory Commission, August 2010.

7. SECY-90-016, Evolutionary Light-Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements, U.S. Nuclear Regulatory Commission, June 1990.

8. ASME Section III, Division 2, Rules of Construction of Nuclear Facility Components

- Code for Concrete Containments, American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.

9. NUREG/CR-5132, Severe Accident Insights Report, U.S. Nuclear Regulatory Commission, April 1988.

10. NUREG/CR-5597, In-Vessel Zircaloy Oxidation/Hydrogen Generation Behavior During Severe Accidents, U.S. Nuclear Regulatory Commission, September 1990.

19.2-59 Rev. 2

RAI 432-8377 - Question 19-64_Rev.7 Non-Proprietary Attachment (3/4)

RAI 432-8377 - Qiuestion 19-64_Rev.7 "A"

The COL applicant and/or holder is to develop and submit an accident management plan. If In-Vessel Retention and External Reactor Vessel Cooling is proposed for accident man-agement by either the COL applicant and/or holder, then the COL applicant and/or holder is to develop and submit to the NRC an evaluation of the effects of higher water level in the cavity on steam explosion loading.

RAI 432-8377 - Question 19-64_Rev.7 Non-Proprietary Attachment (4/4)

APR1400 DCD TIER 2 RAI 432-8377 - Qiuestion 19-64_Rev.7 Table 1.8-2 (37 of 39)

Item No. Description COL 19.1(19) The COL applicant is to describe the uses of PRA in support of licensee programs such as the reactor oversight process during the operational phase.

COL 19.1(20) The COL applicant is to perform the seismic-fire interactions walkdown to confirm a qualitative seismic-fire interaction assessment.

COL 19.1(21) The COL applicant is to develop outage procedures to ensure that in fire compartments containing post-seismic or post-fire safe shutdown equipment that: 1) the seismic ruggedness of temporary ignition sources is adequate, or that the duration that these temporary ignition sources are in these areas is minimized, 2) the seismic ruggedness of temporary equipment such as scaffolding in fire compartments containing potential seismic-fire ignition sources, or near fire protection equipment is adequate, and 3) either the duration of activities which could impact manual firefighting is minimized, or alternative firefighting equipment (e.g.,

pre-stage portable smoke removal equipment, prestage additional firefighting equipment, etc.) is supplied.

COL 19.1(22) The COL applicant is to demonstrate that failure of buildings that are not seismic Category I (e.g., turbine building and compound building) does not impact SSCs designed to be seismic Category I.

COL 19.1(23) The COL applicant is to ensure that asymmetric conditions due to modeling simplicity will be addressed or properly accounted for when the PRA is used for decision making.

COL 19.1(24) The COL applicant will demonstrate that maintenance-induced floods are negligible contributors to flood risk when the plant specific data are available.

COL 19.1(25) SAMGs are entered to initiate SI with the core exit thermocouple indicating 1200 °F.

COL 19.2(1) The COL applicant is to perform and submit site-specific equipment survivability assessment including flooding effect in accordance with 10 CFR 50.34(f) and 10 CFR 50.44 which reflects the equipment identified and the containment atmospheric assessments of temperature, pressure and radiation described in Subsection 19.2.3.3.7.

COL 19.2(2) The COL applicant will demonstrate that the covers for large penetrations such as equipment hatch and personnel airlocks meet the Service Level C requirements in Subsection NE-3220 of the ASME code and explain how the consideration of containment leakage is accounted for when modeling local regions of containment.

COL 19.2(3) The COL applicant and/or holder is to develop and submit an accident management plan including the evaluation of the effect of higher water level in the cavity on steam explosion loading when using In-Vessel Retention and External Reactor Vessel Cooling for accident management.

COL 19.3(1) The COL applicant is to perform site-specific seismic hazard evaluation and seismic risk evaluation as applicable in accordance with NTTF Recommendation 2.1 as outlined in the NRC RFI.

COL 19.3(2) The COL applicant is to address the flood requirements for wet sites COL 19.3(3) The COL applicant is to develop the details for offsite resources.

COL 19.3(4) The COL applicant is to address the details of selecting suitable storage locations for FLEX equipment that provide reasonable protection during specific external events as provided in NEI 12-06 guidance Sections 5 through 9, and the details of the guidance for storage of FLEX equipment provided in the Technical Report (Reference 5) Section 6.2.9.

COL 19.3(5) The COL applicant is to confirm, satisfy, or fulfill the specific design functional requirements of raw water tank including the associated instrument, capacity, location, flow path to on-site, the valve pit connected to FLEX equipment, and any other design features as described in Section 19.3 in support of BDBEE mitigation strategies.

The COL applicant and/or holder is to develop and submit an accident management plan. If In-Vessel Retention and External Reactor Vessel Cooling is proposed for accident management by either the COL applicant and/or holder, then the COL applicant and/or holder is to develop and submit to the NRC an evaluation of the effects of higher water level in the cavity on steam explosion loading.

1.8-41 Rev. 2