ML17324A346

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Meeting Slides for 11/15/2017, Public Meeting Between NRC and SNC Regarding Hatch License Amendment Request - Application of TSTF-423 to Drywell Spray Function Technical Specification End State
ML17324A346
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 11/15/2017
From: Burns P, Fox H, Julie Hughes, Scott O
Southern Nuclear Operating Co
To: Hall J
Plant Licensing Branch II
Hall J
References
Download: ML17324A346 (40)
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License Amendment Request - Application of TSTF-423 to Drywell Spray Function Technical Specification End State NRC Public Meeting 11/15/2017 Pamela Burns, SNC Owen Scott, SNC Heather Fox, JENSEN HUGHES 2

OUTLINE

  • Purpose
  • Problem Statement
  • Recommended Solution

- Qualitative Analysis

- Quantitative Analysis

- External Events and LERF

  • Plant Hatch Specific Technical Specification for RHR Drywell Spray Function
  • Applicability of NEDC-32988 and Justification for Changing the End State of the RHR Drywell Spray Function
  • Summary / Proposed Path Forward 3

PURPOSE

4

PROBLEM STATEMENT

  • In December 2015, Plant Hatch submitted a LAR to implement TSTF-423, which changed the required end states for various Technical Specification (TS) action statements. Among the TSs changed were those related to Residual Heat Removal (RHR) including ECCS, suppression pool cooling and torus sprays.
  • The LAR was approved 12/22/16 by the NRC.
  • This LAR did not include a change to the end states of TS 3.6.2.5 related to RHR Drywell Sprays.
  • Therefore, the end state of TS 3.6.2.5 still requires the plant to be taken to Mode 4 (Cold Shutdown) if the LCO Action Statements are not met.

5

PROBLEM CONTINUED

  • The change in RHR end states implemented by TSTF-423 are essentially negated by having a TS in place related to RHR that requires a more restrictive mode (i.e., cold shutdown vs. hot shutdown).
  • This impacts plant operational flexibility and regulatory efficiency.
  • This inconsistency in TS end states occurred because the drywell (DW) spray function of RHR is a unique TS to Plant Hatch and is not included in the Standard TSs.

6

SOLUTION

  • Apply the previously approved technical approach and methods used in BWROG Technical Report NEDC-32988, which supported TSTF-423, to change Plant Hatch TS 3.6.2.5 such that the required end state for the Required Action is Mode 3 (i.e., hot shutdown).
  • This will ensure that the Drywell Spray TS end state is consistent with the other RHR end states.

7

NRC Endorsement of NEDC-32988

  • The NRC provided a Safety Evaluation (SE) of Topical Report NEDC-32988, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required Action End States for BWR Plants, in letter dated September 27, 2002.
  • The NRC stated that NEDC-32988, Revision 2 is acceptable for referencing in licensing applications for GE-designated BWRs.
  • One conclusion of the SE was that Establishing Mode 3 (hot shutdown) instead of Mode 4 as the end state for several TS action statements could reduce operational costs without compromising safety and may actually enhance safety.
  • The NRC staff found that the risk assessment approach was comprehensive and followed staff guidance as documented in RG 1.174 and 1.177.

8

NEDC-32988 OVERVIEW

  • The objective of the report was to demonstrate that any risk increases associated with the proposed changes in TS end states are either negligible or negative (i.e., a net decrease in risk).
  • The report documents a risk-informed analysis of the proposed TS change which conforms to the guidance in Regulatory Guides (RGs) 1.174 and 1.177.
  • The three tiered approach outlined in RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision-Making: Technical Specifications, was followed.
  • The acceptance criteria outlined in RG1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, were used.

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NEDC-32988 OVERVIEW CONTINUED

  • The report outlined the following tasks:

- Performance of a qualitative generic risk assessment,

- Performance of a quantitative risk assessment for a pilot plant, which includes the following:

  • Comparison of baseline risks between Modes 3 and 4 (i.e., risk when no equipment outages are assumed),
  • Comparison of configuration-specific risks between Modes 3 and 4 (i.e.,

risk when certain equipment is assumed to be unavailable),

  • Performance of sensitivity studies to investigate the robustness of the results to uncertainties in data and modeling assumptions, and
  • Performance of sensitivity studies to ensure that the conclusions of the quantitative assessment for the pilot plant also apply to other BWR plants.

- Use of risk insights, derived from the qualitative and quantitative generic risk assessments, in the individual TS assessments supporting each of the proposed end state changes.

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QUALITATIVE RISK ASSESSMENT

  • Goal - Demonstrate that the proposed TS end state changes maintain defense-in-depth for expected initiating events.

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QUALITATIVE RISK ASSESSMENT

SUMMARY

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QUALITATIVE RISK ASSESSMENT

SUMMARY

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QUALITATIVE RISK ASSESSMENT

- All potentially risk significant initiating events that can occur while the plant is operating in Mode 3 can be represented (or subsumed into) by the following:

  • Loss of coolant accidents (LOCAs)
  • Loss of offsite power (LOOP)
  • Loss of power conversion system (PCS)

- All potentially risk significant initiating events that can occur while the plant is operating in Mode 4 can be represented (or subsumed into) by the following:

  • Loss of coolant inventory
  • Loss of offsite power (LOOP)

QUALITATIVE INSIGHTS

  • The risk impact of LOCAs, as pressure-driven initiating events, are not as significant in Modes 3 and 4 as they are in Mode 1. The major contributor to this initiator is loss of inventory caused by incorrect valve lineups. Since incorrect valve lineups are more likely during Mode 4 operation, the risk associated with LOCAs will be smaller if Mode 3 is adopted as the end state.
  • LOOP is an important initiating event in both Modes 3 and 4 with approximately the same frequency. Therefore, their risk impact is lower when there is more redundancy and diversity of the mitigating systems, as is the case when the plant is operating in Mode 3. Steam-driven high pressure injection systems are available in Mode 3.

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QUALITATIVE INSIGHTS

  • Loss of the power conversion system (PCS) in Mode 3 and loss of RHR in the shutdown cooling mode in Mode 4 are important initiating events of the same order of magnitude frequency. Since there is more redundancy and diversity of the mitigating systems when the plant is operating in Mode 3, the risk impact associated with the loss of the PCS initiating event (occurring in Mode 3) is lower than the risk impact associated with the loss of RHR SDC initiating event (occurring in Mode 4).

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QUALITATIVE INSIGHTS

  • Loss of service water is an important initiating event in both Modes 3 and 4 with approximately the same frequency.

This initiator disables all core and containment cooling systems using the service water system to transfer heat to the ultimate heat sink. In general , accidents initiated by loss of service water in either Mode 3 or Mode 4 are mitigated by using low pressure injections systems and containment venting. Since the SRVs are highly reliable, the risk impact associated with this initiating event is approximately the same for events occurring in Mode 3 and 4.

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QUALITATIVE INSIGHTS

  • The means used to achieve inventory control, reactivity control, reactor overpressure control, containment integrity control and power availability are approximately equally reliable in Modes 3 and 4.
  • More means are available to perform the decay heat removal function while the plant is operating in Mode 3 than when it is operating in Mode 4. For initiating events occurring in Mode 3, both the high and the low pressure systems can be used to provide core cooling. Although the high pressure systems can be used directly, the low pressure systems can be used following reactor depressurization which can be achieved reliably.

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QUALITATIVE INSIGHTS

  • More means are available to perform the containment heat removal function while the plant is operating in Mode 3 than when it is operating in Mode 4.

- In Mode 4, the RHR system or containment venting can be used to remove heat from the containment.

- These same means can be used for initiating events occurring in Mode 3 following reactor depressurization, which can be achieved reliably.

- The difference is that the power conversion system can provide heat removal in Mode 3, but not in Mode 4.

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SUMMARY

OF QUALITATIVE INSIGHTS

  • Plant operation in Mode 3 (hot shutdown) offers at least the same robustness to plant upsets as operation in Mode 4 (cold shutdown).
  • This is substantiated by the quantitative risk study.

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QUANTITATIVE ANALYSIS

  • Goal:

- Substantiate the conclusion of the qualitative risk assessment by providing numerical results for a representative plant.

- Investigate the robustness of the results to uncertainties in data and modeling assumptions through sensitivity analyses.

- Assess the applicability of the results to other BWR plants through sensitivity analyses accounting for design and operational differences.

  • Scope:

- Compare core damage risks associated with either staying in Mode 3 or going to Mode 4 to carry out equipment repairs. This comparison was made for a number of cases based on selected combinations of equipment outages.

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QUANTITATIVE ANALYSIS 22

QUANTITATIVE ANALYSIS 23

QUANTITATIVE INSIGHTS

  • The CDF estimates show that staying in Mode 3, rather than going to Mode 4 to carry out equipment repairs, does not have any adverse effect on plant risk and under some circumstances may actually reduce risk.

- When no equipment is taken out of service (base case), the CDF for Modes 3 and 4 are very similar. The Mode 3 CDF is slightly higher than Mode 4 CDF and the resulting difference is about 1E-08/year.

This is considered insignificant.

- For the majority of analyzed cases with equipment outages, the CDF for Modes 3 and 4 are approximately equal. In all such cases the resulting delta CDF, either an increase or a decrease, is insignificant.

- When one or more redundant trains or pieces of important equipment, such as the RHR, are taken out of service, the Mode 3 CDF is significantly lower than the Mode 4 CDF. This indicates that, for outages involving these important systems, the proposed end state change would lead to significant risk reductions.

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QUANTITATIVE INSIGHTS

  • Two accident initiating events are major contributors to risk in both Modes 3 and 4: (1) Loss of offsite power and (2) loss of service water. In addition, the loss of RHR in the shutdown cooling mode (SDC) becomes a significant risk contributor when certain equipment, such as one RHR loop is out during Mode 4.

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QUANTITATIVE INSIGHTS

  • Mode 3 and Mode 4 risks are essentially equal when no equipment is taken out even though more systems are available in Mode 3 than Mode 4 to mitigate accidents. This is due to the high redundancy and reliability of the low pressure systems which are used to mitigate accidents occurring in Mode 4 and also, following depressurizaton, in Mode 3.
  • When low pressure injection systems and their support systems are taken out of service, Mode 4 risks become significantly higher than Mode 3 risks. This is due to the increased importance of the high pressure systems, which are only credited in Mode 3.

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QUANTITATIVE INSIGHTS

  • A common element driving the risk of accidents occurring in both Mode 3 and Mode 4 is containment cooling.
  • All accident sequences initiated in Mode 3 and Mode 4 (except for Mode 3 transients with the PCS available) require containment heat removal using the RHR pumps or containment venting to avoid core damage.
  • If core cooling is lost in Mode 3 (i.e., the PCS is lost) while one RHR loop is out of service, successful operation of the second RHR loop OR successful containment venting is required to avoid core damage.
  • However, if core cooling is lost in Mode 4 (i.e., the operating RHR SDC loop is lost) while the other RHR loop is out of service, only containment venting remains to avoid core damage. Thus, when an RHR loop is out of service, the loss of core cooling initiating event is a larger risk contributor in Mode 4 than in Mode 3.

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LERF / EXTERNAL EVENTS

  • The quantitative risk assessment does not include risks from external events (seismic events, internal fire or internal floods), risks associated with transitions from one mode of operation to another or risk in terms of large early release (LERF). These risks were assessed qualitatively.
  • Risks associated with external events are smaller when Mode 3 instead of Mode 4 is selected as the end state for the following reasons:

- Seismic events, which are equally likely in either mode, have a larger impact on the plant accident mitigation capability during Mode 4 than during Mode 3. A seismic event is very likely to result in an unrecoverable loss of offsite power event. Also, a seismic event is more likely to disable the condensate and fire water systems than the emergency core cooling systems (ECCS). Since the RCIC and HPCI systems, which are designed for seismic loads, are available in Mode 3 and not Mode 4, the plant ability to prevent core damage is higher in Mode 3 than in Mode 4.

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EXTERNAL EVENTS

  • Internal fire and flood events are equally likely to occur during Mode 3 or Mode 4.
  • During either mode the same fire or flood event would impact the same equipment.
  • Because there are more systems available for accident mitigation in Mode 3 than in Mode 4, the plants ability to prevent core damage is at least as good in Mode 3 as is in Mode 4 29

LERF

  • During power operation, large early releases are the result of (1) energetic containment failure due to a high pressure core melt, (2) a containment bypass event, and (3) a core damage event occurring in combination with an unisolated containment.
  • Compared to power operation, Mode 3 or Mode 4 operation is associated with lower initial energy level, reduced fission product inventory level and reduced decay heat load.
  • Due to the combined effect of these factors, even though the initial RCS pressure during Mode 3 is higher than during Mode 4, the likelihood of large early releases in Modes 3 and 4 is very low.
  • These factors serve to provide time for the operator to respond to serious plant upsets and, consequently, the contribute to delaying the core melt progression and reducing radiation releases.
  • Therefore, any potential increase due to changing the end state is negligible.

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SENSITIVITY STUDY 31

NEDC-32988

SUMMARY

  • The risk assessment approach is comprehensive and follows the NRC staff guidance as documented in RG 1.174 and 1.177.
  • The analyses show that the criteria of the three-tiered approach for allowing TS changes (as documented in RG 1.177) are met.

- Tier 1: Risk Impact of Proposed Change. The risk changes in terms of CDF and LERF are risk neutral or risk beneficial.

- Tier 2: Avoidance of Risk-Significant Configurations. The performed risk analyses which are based on single LCOs, have shown that there are no high risk configurations associated with the proposed TS end state changes.

- Tier 3: Configuration Risk Management. Licensees have programs in place to comply with 10 CFR 50.65(a)(4) to assess and manage the risk from proposed maintenance activities.

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TS 3.6.2.5 RHR Drywell Spray

  • The end state of the RHR DW Spray TS was not specifically addressed as part of the LAR to implement TSTF-423 because this particular TS in not included in Standard Technical Specifications.
  • The qualitative, quantitative and other risk assessments from NEDC-32988 are applicable and relevant to TS 3.6.2.5.

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TS 3.6.2.5 BASIS

  • Drywell spray is a mode of the RHR system which may be initiated under post accident conditions to reduce the temperature and pressure of the primary containment atmosphere.
  • Drywell spray is also operated post-LOCA to wash, or scrub, inorganic iodines and particulates from the drywell atmosphere into the suppression pool.
  • At Plant Hatch, the drywell spray is credited post-LOCA for both the scrubbing function as well as the temperature and pressure reduction effects.
  • The drywell spray is not credited in determining the post-LOCA peak primary containment internal pressure; however, the Hatch radiological dose analysis does take credit for the drywell spray temperature and pressure reduction over time in reducing the post-LOCA primary containment leakage and main steam isolation valve leakage.

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APPLICABILITY OF NEDC-32988

  • This conclusion was based on the low probability of an event requiring the safety function, alternate methods to remove heat from primary containment, and the number of systems available in Mode 3.
  • The conclusion was that the risks of staying in Mode 3 were lower than or equal to going to the Mode 4 end state.

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NEDC-32988 Applicability to RHR DW Sprays

  • The topical report addressed suppression chamber sprays which has a similar function to DW sprays.
  • Following a design basis accident, the RHR suppression pool spray subsystem removes heat from the suppression chamber airspace. This is similar to the function of the DW spray function.
  • With respect to suppression chamber sprays, NEDC-32988 noted the following:

- Failure of suppression chamber sprays is not risk significant.

- Based on the low probability of an event requiring the safety function (i.e., LOCA), availability of alternate methods to remove heat from primary containment, and the number of systems available in Mode 3, it is concluded that the risks of staying Mode 3 are lower than or equal to going to the Mode 4 end state.

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NEDC-32988 APPLICABILITY TO DW SPRAYS

  • The fission product removal function of DW sprays is not quantified in the quantitative analysis.
  • However, the qualitative arguments related to LERF presented in NEDC-32988 are applicable in that LOCAs are not a significant initiating event in Mode 3 or Mode 4.
  • Compared to power operation, Mode 3 or Mode 4 operation is associated with lower initial energy level, reduced fission product inventory level and reduced decay heat load.
  • Due to the combined effect of these factors, even though the initial RCS pressure during Mode 3 is higher than during Mode 4, the likelihood of large early releases in Modes 3 and 4 is very low.
  • These factors serve to provide time for the operator to respond to serious plant upsets and, consequently, they contribute to delaying the core melt progression and reducing radiation releases.
  • Therefore, any potential increase in risk due to changing the end state is negligible.

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SUMMARY

  • NEDC-32988 provides a technical argument for changing the required end states of several standard Technical Specifications.
  • Plant Hatch adopted a TS that is not part of Standard Technical Specifications as part of the Alternate Source Term License Amendment.
  • The risk assessment presented in NEDC-32988 is applicable to TS 3.6.2.5, RHR Drywell Sprays even though it is not specifically referenced.
  • Changing the end state of TS 3.6.2.5 improves operational efficiency and reduces regulatory burden. This change is consistent with the intent of NEDC-32988 and TSTF-423. It is also consistent with Reg. Guides 1.174 and 1.177.

Defense-in-depth is maintained with the proposed change in end state.

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PROPOSED PATH FORWARD

  • SNC to submit LAR to change the end state of TS 3.6.2.5 to be consistent with the other end states of the TSs related to RHR; specifically, Mode 3.
  • NEDC-32988 will be used as the technical basis and justification. SNC asserts that no new risk assessments are required because NEDC-32988 appropriately addresses all risk aspects.
  • Plant Hatch has already committed to the implementation guidance for TSTF-423.

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FEEDBACK / QUESTIONS

  • Pamela Burns, Licensing Engineer, Southern Nuclear
  • pdburns@southernco.com
  • Owen Scott, Risk Informed Engineering Applications Manager, Southern Nuclear
  • Email: omscott@southernco.com
  • Heather Fox, Consultant, JENSEN HUGHES
  • Email: hfox@jensenhughes.com 40
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