ML20241A242
| ML20241A242 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/04/2020 |
| From: | Greg Werner Operations Branch IV |
| To: | Nebraska Public Power District (NPPD) |
| References | |
| Download: ML20241A242 (7) | |
Text
Attachment 6 OBDI 202 - IOLE Process Rev 3 PROPOSED OUTLINE COMMENTS Facility:
CNS First Exam Date:
Apr 27, 2020 Written Exam Outline Comment Resolution 1 NRC Generated None.
2 Replacement K/As approved by NRC None.
3 Replace RO Question 31:
K/A 259001 (SF2 FWS) Feedwater, Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including:
A1.06 Feedwater heater level with the following:
K/A 256000 (SF2 CDS) Condensate, Ability to (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.02 Valve closures.
Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including:
A1.10 Condenser vacuum Reason is that same system cannot be sampled on both RO and SRO exams until all systems in a tier/group have been sampled between the two exams.
First replaced K/A 259001 A1.06 with K/A 256000 A1.10 as noted in this comment, but could not write a discriminatory question at the RO level.
Later, replaced K/A 256000 A1.10 with K/A 256000 A1.07, System lineup, provided by NRC.
Administrative JPM Outline Comment Resolution 1 A1 says modified by type code is N Changed type code to M for A1 2 A3 is also type code D Added type code D for A3 3 Rename SA5-9 as A5-A9.
Renamed SA5-9 as A5-A9 4 A5 is also type code D Added type code D for A5
OBDI 202 - IOLE Process Rev 3 5
A8 on its face it appears to be a combination of the 2018 and 2017 A8 JPM tasks ensure pertinent conditions are changed so that this isnt just a combination of two previously-used JPMs, and that the spirit of non-repetition is met.
Discussed with NRC. The first part of this JPM is to determine the emergency exposure limit. The task is to manually close a RCIC steam isolation valve. The exposure limit for this task is 10Rem. In contrast, the task associated with 2018 JPM SA8 was to locally start RCIC and had an exposure limit of 25 Rem. The second portion of this task is to determine whether authorization of KI is required for the task. For the 2020 JPM, KI should NOT be authorized for the operator performing the task based on plant conditions. In contrast, 2017 JPM A8 required determining KI should be authorized for EMTs involved in the related task.
2020 JPM A8 is designated as new because it was created from the ground up without reference to any bank JPM and involves a new combination of task requirements related to completely different task conditions and expected results than used in 2017 JPM A8 or 2018 JPM SA8.
Control Room / In-Plant System JPM Outline Comment Resolution 1 Add S type code to all sim JPMs Added S type code to all sim JPMs 2
Combine ES-301-2s into one form (NRC will provide)
Combined the three ES-301-2 forms into one, specifying which are performed by RO, SRO-I, and SRO-U 3
For alt path JPMs, add a brief 1 sentence description of what the faulted condition is.
Added brief statement describing the alternate path for each alternate path JPM 4 Credit S2 as D as well, on 301-2 Added type code D for S2 Simulator Scenario Outline Comments Comment Resolution 1
SCEN1 - Is event 5 considered an ATWS? Should the reactor have tripped automatically but failed to?
If not an ATWS, Id like to credit this as a Component malfunction instead of a Major, esp since it is immediately followed by another Major.
Discussed with NRC. Event 5 is not an ATWS. The stability solution used at CNS does not include an automatic scram based on P/F Stability Exclusion Region Entry nor flux oscillation signature. As requested in comment #3, the automatic scram based on APRM flux high is defeated so that operator action is required to scram.
OBDI 202 - IOLE Process Rev 3 This was changed from a Major event to a malfunction per CE request.
2 SCEN1 - Event 5, does RRP B seize on shaft binding and then trip? Not clear.
Also:
This appears to be similar or identical to 2018 Scenario 4 event 6 Major event.
NUREG 1021 Rev 11 APP D states: if any major event is repeated from either of the previous two NRC initial licensing operating tests, the examination author should change the major event, the ICs, or subsequent malfunctions (or a combination) to alter the course of action (within the emergency procedures) for the given scenario(s).
The NRC expect that all major events would be broadly sampled over the course of several operating tests and that no major event will routinely be omitted without justification. If a facility licensee encounters difficulty meeting these requirements (e.g., because of large class sizes requiring the generation of more scenarios than normal), it should coordinate with the NRC chief examiner to meet the intent of this section to the extent possible.
Justify how this major event is different enough from 2018 THI event to satisfy above guidance, or replace.
The shaft binding malfunction causes a binding of RR Pump B rotor. Motor current, pump vibration, and motor vibration increase, resulting in annunciation on Panel 9-4. RR Pump B will trip ~20 seconds after Trigger 5 is activated.
This is considered to be a different event from that used in 2018 Scenario 4 event 6, because in that case RR Pump A experienced shaft binding, and in this scenario RR Pump B will experience shaft binding. Completely different annunciators and pump indications specific to RR Pump B are involved. Additionally, the course of action within the EOPs for this scenario is different from that in 2018 Scenario 4. In 2018 Scenario 4, normal scram response ensued, complicated by need to manually control RFPs due to a 480V Buss loss, until, after a 14 minute delay, a RCIC steam leak into secondary containment was inserted, and EOP-5A became the driver. No ECCS initiation signal was received in that scenario.
In this 2020 scenario, a drywell steam leak occurs 3 minutes after the scram, resulting in an ECCS initiation and multiple Group isolations and the need to verify initiations/isolations and override ECCS systems not required for injection IAW EOP-1A. The effect of the steam leak on DW pressure results in EOP-3A being the driver for the scenario.
In addition, this was changed to a component malfunction per CE request in comment #1. Discussed with NRC, and now there is no concern related to a repeat major event from 2018 Scenario 4 event 6.
3 SCEN 1 CT#1 - I like the 25% peak-to-peak bounding criteria, but I dont want the plant to be able to be put in a safe condition by an automatic scram on APRM hi power block the APRM high flux SCRAM such that the crew HAS to manually trip the reactor to put plant in safe condition. Dont count the blocked APRM scram as a separate event Discussed with NRC. The automatic scram based on APRM flux high is defeated so that operator action is required to scram. The overrides used to defeat the automatic scram are not counted as an additional malfunction.
OBDI 202 - IOLE Process Rev 3 because were not expecting a competent crew to get to the point where APRM auto scram is challenged, its more of an initial condition for the THI event.
4 SCEN 1 - Combine events 9 and 10, RHR pumps B and D trip, into one event, since they are both just used as drivers for Emx Depressurization.
Discussed with NRC. Combined events 9 and 10 and eliminated RHR Pump D failure to auto start.
5 SCEN1 Event 7 - what action is required by the crew in response to this malfunction, that they would not have to take if this malfunction did not occur?
Trying to determine if it should be creditable as a separate I/C bean, or if its more of a setup condition for the E.D.
Discussed with NRC. The Containment Spray cooling valve permissive switch will fail for the first RHR loop attempted to be placed in Torus spray. The summary was written assuming RHR A will be attempted first. This override will require the crew to transition to the opposite RHR loop to establish Containment Spray.
6 SCEN 1 - good job putting together and documenting OE for this scenario.
Will try to use this effort for OE as model for future exams.
7 SCEN 1 - Target Quantitative Attribute for Malfunctions After EOP Entry is 1-2.
Scenario 1 has 3. ES-301 states, The quantitative attribute target ranges that are specified on the form are not absolute limitations; some scenarios may be an excellent evaluation tool but may not fit within the ranges. A scenario that does not fit into these ranges shall be evaluated to ensure that the level of difficulty is appropriate. On evaluation, this scenario has an appropriately normal level of difficulty, the 1st two post-EOP malfunctions are straightforward, and the 3rd post-EOP malfunction is necessary to facilitate transition to an EOP contingency procedure, at least one of which is required per scenario set. In the opinion of the Chief Examiner, the deviation from target quantitative attributes is justified and the scenario is of an appropriate level of difficulty.
No action required.
8 SCEN 2 - Event 8 - what is the significance of Mode Switch failing in RUN position? Writeup sounds like scram in event 6 was successful, except for SDV vent/drain valves failure to close. Is this an ATWS?
Discussed with NRC. The Rx Mode Switch failing in RUN uses an override to simulate the switch being in RUN, with respect to electrical contacts, regardless of physical switch position. This scenario uses a low decay heat IC. When the reactor is scrammed, RPV pressure will decay below the Group 1 isolation setpoint on RPV OBDI 202 - IOLE Process Rev 3 Is this supposed to be the Mode Switch physically failing in RUN, or the RUN electrical contact remaining energized?
Pressure Low within < 1 minute. With the Rx Mode Switch failed in RUN, a Group 1 (MSIV) isolation will occur. This is necessary to prevent reactor pressure from depleting through paths to the condenser in order to maintain enough driving head for the HPCI steam leak to exceed Max Safe temperatures in more than one area in secondary containment. This will force the crew to ultimately transition to other injection systems than RFPs and prevents resetting the scram. The crew will have to control CRD flow to control reactor level below 9
SCEN 2 - Target Quantitative Attribute for Malfunctions After EOP Entry is 1-2.
Scenario 2 has 4. ES-301 states, The quantitative attribute target ranges that are specified on the form are not absolute limitations; some scenarios may be an excellent evaluation tool but may not fit within the ranges. A scenario that does not fit into these ranges shall be evaluated to ensure that the level of difficulty is appropriate. On evaluation, this scenario has an appropriately normal level of difficulty, the post-EOP malfunctions are straightforward and related, necessary to facilitate transition to an EOP contingency procedure. In the opinion of the Chief Examiner, the deviation from target quantitative attributes is justified and the scenario is of an appropriate level of difficulty.
No action required.
11 Scenario 3 event 5 hydraulic block ATWS: Explain how this is different from Hydraulic ATWS used on 2018 scenario 1 and 2017 scenario 1, to satisfy guidance: NUREG 1021 Rev 11 APP D: if any major event is repeated from either of the previous two NRC initial licensing operating tests, the examination author should change the major event, the ICs, or subsequent malfunctions (or a combination) to alter the course of action (within the emergency procedures) for the given scenario(s). The NRC expect that all major events would be broadly sampled over the course of several operating tests and that no major event will routinely be omitted without justification.
The 2018 ATWS was the low power scenario. Power was <3% following RR Pump trip, so reactor water level was not required to be lowered. SLC Pump A failed to start, but was of less consequence since power was low. For 2020, SLC B relief valve will fail open, which will be of significantly greater consequence, since power is much higher than in 2018.
The 2017 ATWS was ~40% hydraulic block malfunction severity, resulting in ~ 35%
power after RR Pumps were tripped. In that scenario, Bypass valves failed closed; however SLC was not impaired and some steam continued to go to the main condenser via BOP steam loads. RFPs remained available to control reactor water level. SLC OBDI 202 - IOLE Process Rev 3 If a facility licensee encounters difficulty meeting these requirements (e.g.,
because of large class sizes requiring the generation of more scenarios than normal), it should coordinate with the NRC chief examiner to meet the intent of this section to the extent possible.
How is the response to the hydraulic ATWS itself different from previous 2 exams?
reduced power to <3%within ~3 minutes.
RR Pumps automatically tripped on ATWS-RPT in 2017, but in 2020, RR Pump B will fail to trip automatically and must be tripped manually.
In this 2020 scenario, ATWS severity and initial resulting power level is comparable to that of the 2017 scenario. But in contrast, the 2020 scenario will involve a complete loss of condenser vacuum, resulting in loss of RFPs and MSIV closure. The condenser will be lost as a heat sink, and all steam will divert to the Suppression Pool. HPCI and RCIC will be required for level control due to loss of RFPs. SLC will be impaired, so power will remain higher significantly longer.
Manual control rod insertion and lowering reactor water level, which was not required to be executed in 2018, will necessitate higher priorities than in 2017 in order to limit the amount of energy the Suppression Pool must absorb. Execution of the legs of the EOPs will require a different prioritization than was required, necessarily, in 2017. The CRS will prioritize BOP actions differently than was necessary in 2017 due to loss of RFPs and MSIV closure.
Discussed with NRC, difference is acceptable.
12 SCEN 3 - Target Quantitative Attribute for Malfunctions After EOP Entry is 1-2.
Scenario 3 has 3. On evaluation, this scenario has an appropriately normal level of difficulty, the post-EOP malfunctions are straightforward and related, necessary to facilitate transition to an EOP contingency procedure. In the opinion of the Chief Examiner, the deviation from target quantitative attributes is justified and the scenario is of an appropriate level of difficulty.
No action required.
Scenario 2 - Event 4 (control rod double notch) is not listed on quantitative table of Instrument/Component Malfunctions, or Total Malfunctions Added control rod double notch to Instrument/Component Malfunctions and Total Malfunctions on qualitative attribute table.
OBDI 202 - IOLE Process Rev 3 General Comments Comment Resolution 1
ALL SCENARIOS: Make sure that none of the Major Events are exact repeats from last two exams. This is to satisfy guidance of NUREG 1021 Rev 11 APP D: if any major event is repeated from either of the previous two NRC initial licensing operating tests, the examination author should change the major event, the ICs, or subsequent malfunctions (or a combination) to alter the course of action (within the emergency procedures) for the given scenario(s). The NRC expect that all major events would be broadly sampled over the course of several operating tests and that no major event will routinely be omitted without justification.
If a facility licensee encounters difficulty meeting these requirements (e.g.,
because of large class sizes requiring the generation of more scenarios than normal), it should coordinate with the NRC chief examiner to meet the intent of this section to the extent possible.
Each justification for major events was discussed with NRC, and major events used in each scenario were determined to be different from those used in previous two exams, IAW NUREG 1021 Rev 11 App. D.
Form ES-201-2 3.a: provide revised form, signed, when audit exam is developed.
Audit exam will not be completely developed until after the proposed exam submittal date of 2/10/20.