ML21034A531
| ML21034A531 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/03/2021 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| Download: ML21034A531 (76) | |
Text
ML21034A531 Question 15 Original Comment:
Applicant 55-24559 comment submitted to the station:
The applicants understand that the Circulating Water Pump seal water system is designed to handle a loss of Intake Cooling Water with backup from the domestic water system, it is their opinion that the Circ Water seal system would still see a small temperature change. Therefore, accept answers C and D as both correct.
PSL Station Recommendation:
After receiving the feedback from the applicants and discussing system design and operation with operations personnel, it is possible that the system does see some perturbation and temperature change on the Circulating Water Pump seal coolers.
Therefore, the station recommends accepting C and D as correct answers.
Amplifying Information Provided Post NRC Feedback:
Given the different operating temperatures and pressures between Intake Cooling Water and Domestic Water, and the basic manually-controlled throttling mechanism of the Circulating Water Pump seal water flow; it is reasonable to assume that after SIAS, when Circulating Water Pump seal water swaps over to the Domestic Water backup supply, the Circulating Water Pump seal water flow and temperature will change, thereby causing a change to the packing gland temperature.
The lubricating water for the pump seals and bearings comes from the Intake Cooling Water (ICW) system through a series of self-cleaning and manually cleaned strainers. Flow is monitored via a flow indicating switch at each pump and is controlled via a manually throttled ball valve. An alternate supply of lube water comes from the domestic water system, which draws suction from the City Water Storage Tanks (CWST).
During a SIAS, ICW supply to the Circulating Water Pumps will be isolated by the automatic closure of MV-21-2/-3, A/B ICW TRAIN TO TCW HXS. A low-pressure condition created from isolation of ICW will cause PCV-21-26, B/U LUBE WATER TO CW PUMPS, to OPEN and supply backup cooling water to the Circulating Water Pump seals. This backup cooling water will be of a different pressure (75 psig vs 40 psig) and temperature (74°F vs 67-88°F) than the Intake Cooling Water system, and thus change the flow and temperature of water supplying the Circulating Water Pump seals.
Additionally, 2-NOP-21.02, Circulating Water System Operations, contains instructions for aligning alternate sources of lube water to the Circulating Water Pumps. Given a potential system pressure difference between the alternate vs normal seal water source, the procedure directs throttling of seal water flow to the 6-10 gpm band and validating adequate lube water leak off from the Circulating Water Pump packing gland. During alignment of the alternate water sources, changing seal water flow may require monitoring of packing gland temperature to validate no rising temperatures. Based on these instructions, when aligning alternate sources of lube water to the Circulating Water Pumps, some seal water temperatures changes are expected.
PSL Station Recommendation:
Therefore, the station still recommends accepting C and D as correct answers.
Intake Temperature has varied between 67-88°F over the last 512 days according to PI Points PSL2-TE-21-20, PSL1-TE-21-3A, PSL1-TE-21-3B.
Domestic Water is supplied by the City Water Storage Tanks (CWST). The CWST temperature is currently 74 °F as determined by measuring a water sample taken on 1/13/2021 at PX-21-13.
The CWSTs have a 353,000 gallon minimum operational limit (500,000 gallon capacity) and do not experience much temperature change over the course of the year.
Domestic Water Pumps operate between 100-120 psig (PS-15-16 Start Signal @ 100 psig, PS-15-15 Stop Signal @ 120 psig)
PCV-21-26 OPENS on LOW ICW Pressure (2998-G-082 Sh. 1)
Engineering Evaluation PSL-BFSM-16-002, UNIT 1 & 2 PCV-21-26 ACTUATOR SETPOINT EVALUATION, assumes the Domestic Water Upstream pressure before PCV-21-26 is 75 psig, and the ICW Downstream Pressure is 40 psig.
Circulating Water Pump seal water flow is manually adjusted by throttling a ball valve to achieve 6-10 gpm with normal ICW supply pressure of ~40 psig (2-NOP-21.02)
Post Exam Comments Question 12 Applicant comment submitted to the station:
Assumption was made that the timeline was too vague to determine when the EDG output breaker was closed. 4 applicants assumed that the EDG output breaker was closed in the 10 seconds between 00:05:00 and 00:05:10 and not at time 00:05:10 as the author intended. Therefore, the HPSI pump would have started by design on its 6 second block prior to 00:05:10. Therefore, accept answers A and C as both correct.
Station Recommendation:
The station does not have plant data that would support the EDG Output breakers closing any sooner than 00:05:08, therefore, at no time in the provided information would the HPSI Breakers be closed. The station recommends no changes to question 12.
Question 15 Applicant comment submitted to the station:
The applicants understand that the Circ Water seal water system is designed to handle a loss of ICW with backup from the domestic water system, it is their opinion that the CCW seal system would still see a small temperature change. Therefore, accept answers C and D as both correct.
PSL Station Recommendation:
After receiving the feedback from the applicants and discussing system design and operation with operations personnel, it is possible that the system does see some perturbation and temperature change on the Circulating Water Pump Seal coolers.
Therefore, the station recommends accepting C and D as correct answers.
Question 21 Applicant comment submitted to the station:
The question stem states that the Reactor has been determined to be critical. However, the applicants identified that the both of the Wide Range Nuclear Instruments (WRNI) indications were approximately TWO decades too low for where the Reactor would normally be while criticality is declared (normally near 1 x 10-5 %). Additionally, the information provided in the question only shows WR power rising ~ ONE decade (from 3 x 10-8 to 3 x 10 -7). This is also contrary to normal reactor behavior during an approach to criticality, where WRNI power typically rises by approximately two decades. Actual Unit 2 plant data from March of 2020 during a reactor startup shows that the Reactor went critical at a WRNI power of 1.89 x 10-5 (shown below) following a rise of over two decades. This expectation has also been validated during performance of 2-PTP-81, Reload Startup Physics Testing, which showed WRNI power (after raising power two decades above critical) was recorded as 1.0 e-3, so criticality likely occurred at ~1.0 e-5.
Therefore the response of the WR nuclear instruments would indicate that the reactor should only be half way to criticality. This would align with the response of S/U channel B (which has only shown 3 doublings). That means the response of 3 out of 4 Nuclear Instruments show the reactor only being half way to criticality, leading the applicants to determine that Startup Channel A is NOT reading correctly and Startup Channel B IS reading correctly. Therefore, accept answers A and C as both correct.
PSL Station Recommendation:
The question provided the candidates with 2 sets of nuclear instrument data for the candidate to determine which set of data was expected for the conditions. The correct answer is based on an original source range value that then doubles 6 times therefore by definition, the reactor is critical.
However in March of 2020, the most recent Unit 2 Startup, the Reactor went critical at 1.89 x 10-5.
PSL Station Recommendation:
The question provided the candidates with 2 sets of nuclear instrument data to determine which set was expected for the conditions. The correct answer was based on an original source range value which doubled 6 times, therefore by definition, the reactor is critical. However in March of 2020, the most recent Unit 2 Startup, the Reactor went critical at 1.89 x 10-5 with nuclear instrument response on par with the candidates description. Due to the new information related to the March 2020 reactor startup, the station is recommending accepting A and C as correct answers.
Question 32 Applicant comment submitted to the station:
The question stem doesnt contain enough information to determine if a HPSI pump restart would be allowed. 1 Attempt to restart the HPSI pump would be allowed to protect the health and safety of the public. Since the stem provides no mention of this and it is not the ROs function to determine if health and safety of the public is at risk, and there is no other information regarding why the HPSI pump tripped, it is rational that starting the HPSI pump 1 time would be allowed. Therefore, accept answers A and B as both correct.
PSL Station Recommendation:
The stem of the question does not provide enough information for the RO candidate to determine the impact on the HPSI system and associated pump. make the decision based on health and safety of the public. Therefore, the station recommends accepting A and B as correct answers.
The RCS heat removal safety function requires that either, an unisolated S/G level is between 60%-70% NR with Tcold stable or lowering, or feedwater is being controlled to restore the unisolated S/G level to between 60%-70% with Tcold stable or lowering.
Once Main Feedwater is lost, the RCS heat removal safety function is not being met, and immediate action is required to restore the RCS safety function.
ADM 11.16 specifically states:
If the safety function status check acceptance criteria are NOT met for a particular safety function, the operator should immediately report this to the Unit Supervisor, then
Additionally, ADM-11.16 states:
The caution states: IF Main Feedwater is lost and AFAS is left to automatically actuate, damage to the S/G feed ring will occur. Prompt restoration of Steam Generator levels using Auxiliary Feedwater is allowed in order to satisfy the Safety Function prior to formally addressing the entire RCS Heat removal Safety Function. This allows the crew to restore Steam generator water level with Auxiliary Feedwater.
IAW EPIP-02 a release is defined as follows:
The ruptured S/G is not being vented to atmosphere, and the C AFW pump steam admission valve from the ruptured S/G would have already been closed in accordance with Plant procedures. Therefore, a release is NOT in progress.
PSL Station Recommendation:
Based on the comment submitted by the Applicant, the Station agrees with the applicants statement that due to the information provided in the steam of question 99, the applicant, while following the guidance provided to the applicant during the NUREG-1021, Appendix E briefing, which states: If you have any questions concerning the intent or the initial conditions of a question, do not hesitate to ask them before answering the question. Note that questions asked during the examination are taken into consideration during the grading process and when reviewing requests for informal NRC staff reviews (appeals). Ask questions of the NRC examiner or the designated facility instructor only.
A dictionary is available if you need it.
When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question so states or the alarm is expected to activate as a result of the conditions that are stated in the question. Similarly, you should assume that no operator actions have been taken, unless the stem of the question or the answer choices specifically state otherwise. Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator operation or training references, you should answer the question based on the actual plant.
The assumptions made by the candidate are consistent with the guidance provided.
Therefore, the station recommends accepting A and C as correct answers.
Post Exam Feedback JPM A-2S and A-2R During administration of PSL L-20-1 NRC A-2S and A-2R; Perform RCS inventory Balance for the RO and SRO Candidates, an unintended typographical error was indentified which resulted in a math error in the JPM key.
The station has corrected the typographical error in the calculation in the JPM.
Station Recommendation: It is the stations request to update the two JPMs as noted below.
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 17 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 5.0 RESTORATION AND DOCUMENTATION 5.1 Restoration None 5.2 Documentation I have reviewed the requirements of this procedure and the results of the RCS Leak Rate Calculator Long Term Trend, including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s),
etc.) Any deviations, abnormal results, equipment problems, failures, or human performance issues must be documented via a Condition Report for each individual item.
C.R.# ________________________
Reviewed By:
SM/US (Print/Sign) Date Remarks:
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 31 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 1 of 4)
- 1. CALCULATE the RCS Leak Rate using PI-VIEW as follows. ______
- 2. IF PI-View is available, THEN PERFORM the following:
A. SELECT START at RO LAN workstation. ______
B. SELECT ALL PROGRAMS ______
C. SELECT PI-View. ______
D. CLICK PI-View ______
(1) ENSURE PI Server PSL-ERDADS is selected. ______
(2) SET START/END as desired ______
(3) SELECT desired sample increment using slide bar. ______
(4) SELECT Get Data. ______
E. RECORD end and start time data for desired leak rate duration. ______
F. CALCULATE and RECORD differences. ______
G. ENTER the data in the Hand Calculated Section. ______
NOTE Only one data sample should be saved in each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. If more than one data sample is taken in a calendar day, then a daily average value shall be determined using Attachment 5, Daily Average RCS Leak Rate Calculation. If the RCS Leak Rate Calculator shows a sample has already been saved for the current calendar day, a new value may be entered by pressing YES, when asked to overwrite the last saved data point. This should only be required when Attachment 5, Daily Average RCS Leak Rate Calculation, is being performed along with this section, in which case Attachment 4, Step 2.H and Attachment 4, Step 2.I are NOT Applicable (N/A).
H. CLICK on the Hand Calculated Section Results, Save Data button. ______
I. VERIFY that no action levels are displayed. ______
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 32 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 2 of 4)
- 2. (continued)
J. IF an action level exists, THEN ENSURE a narrative log entry is created to document this condition. ______
- 3. IF PI-View is unavailable OR at US/SM discretion, THEN RECORD data using RTGB indications. ______
RECORD results in Hand Calc Section. ______
Initial Conditions Final Conditions Notes Difference 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> minimum desired Date:____/____/____ Date:____/____/____ unless not at steady state.
_________Min Time:__________ Time:__________ (Chg. Pump data excluded)
(Attachment 2 data excluded)
TCOLD: _________°F TCOLD: _________°F Held constant (<.1°F) ** _________°F B.A. Integrator:_______ B.A. Integrator:______ Always Negative -_________gal.
PMW Integrator:______ PMW Integrator:_____ Always Negative -_________gal.
Notify SNPO to mark Notify SNPO to mark Zinc Injection Tank Zinc Injection Tank 60 minute minimum required.
gauge glass (LG-02-2) gauge glass (LG-02-2) -_________gal.
Always negative (gal) (gal)
+ if final > initial VCT Level:_______% VCT Level:_______% - if final < initial ( )________gal.
(% change x 33.8 gal. per %)
+ if final > initial Pzr Level:________% Pzr Level:________% - if final < initial ( )________gal.
(% change x 67.04 gal. per %)
Always Negative Q.T. Level:_______% Q.T. Level:_______% N/A if final < initial -_________gal.
(% change x per Table 2) 30 minute minimum required.
Notify SNPO to mark Notify SNPO to mark Always Negative.
Seal Tank level for ALL Seal Tank level for ALL -_________gal.
N/A if final < initial available Chg. Pps.
- available Chg. Pps. *
(1 inch = 0.59 gal.)
- Data should be recorded for any pump hydraulically connected to the system.
- If TCOLD is NOT held constant, consideration should be given to repeating calculation when conditions permit.
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 33 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 ATTACHMENT 4 Manual Leak Rate Calculation (Page 3 of 4)
- 3. (continued)
Table 2 Q.T. Level Range Q.T. Volume (gals/%)
60-61% 19.0 gals/%
61-66% 18.0 gals/%
66-67% 17.0 gals/%
67-68% 18.0 gals/%
68-69% 17.0 gals/%
69-70% 18.0 gals/%
- 4. DETERMINE Total Leakage of RCS by COMBINING differences of the following:
B.A. Integrator - _______gal.
PMW Integrator - _______gal.
Zinc Injection - _______gal.
VCT Level ( )_______gal.
Pzr Level ( )_______gal.
Total Leakage - _______gal.
- 5. DETERMINE Identified Leakage by COMBINING differences of the following components: ______
Q.T. Level - _______gal.
Chg. Pp. Seal Tanks - _______gal.
Attachment 2 (if used) - _______gal.
Identified Leakage = _______gal.
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 34 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 4 of 4)
- 6. DETERMINE Unidentified Leakage by SUBTRACTING Identified Leakage from Total Leakage. ______
Total Leakage ( )_______gal.
Identified Leakage - ( )_______gal.
Unidentified Leakage = ( )_______gal.
- 7. DETERMINE Unidentified Leakage rate by DIVIDING Unidentified Leakage value in Attachment 4, Step 6 by elapsed time in minutes. ______
Unidentified Leakage - _______gal.
Elapsed time _______min.
Unidentified Leakage rate =_______gpm
- 8. DETERMINE Identified Leakage rate by DIVIDING Identified Leakage value in Attachment 4, Step 5 by elapsed time in minutes. ______
Identified Leakage - _______gal.
Elapsed time _______min.
Identified Leakage rate =_______gpm
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 17 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 5.0 RESTORATION AND DOCUMENTATION 5.1 Restoration None 5.2 Documentation I have reviewed the requirements of this procedure and the results of the RCS Leak Rate Calculator Long Term Trend, including other surveillances performed during this procedure, if any (i.e., datasheet(s), PMT sheet(s),
etc.) Any deviations, abnormal results, equipment problems, failures, or human performance issues must be documented via a Condition Report for each individual item.
C.R.# ________________________
Reviewed By:
SM/US (Print/Sign) Date Remarks:
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 31 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 1 of 4)
- 1. CALCULATE the RCS Leak Rate using PI-VIEW as follows. ______
- 2. IF PI-View is available, THEN PERFORM the following:
A. SELECT START at RO LAN workstation. ______
B. SELECT ALL PROGRAMS ______
C. SELECT PI-View. ______
D. CLICK PI-View ______
(1) ENSURE PI Server PSL-ERDADS is selected. ______
(2) SET START/END as desired ______
(3) SELECT desired sample increment using slide bar. ______
(4) SELECT Get Data. ______
E. RECORD end and start time data for desired leak rate duration. ______
F. CALCULATE and RECORD differences. ______
G. ENTER the data in the Hand Calculated Section. ______
NOTE Only one data sample should be saved in each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. If more than one data sample is taken in a calendar day, then a daily average value shall be determined using Attachment 5, Daily Average RCS Leak Rate Calculation. If the RCS Leak Rate Calculator shows a sample has already been saved for the current calendar day, a new value may be entered by pressing YES, when asked to overwrite the last saved data point. This should only be required when Attachment 5, Daily Average RCS Leak Rate Calculation, is being performed along with this section, in which case Attachment 4, Step 2.H and Attachment 4, Step 2.I are NOT Applicable (N/A).
H. CLICK on the Hand Calculated Section Results, Save Data button. ______
I. VERIFY that no action levels are displayed. ______
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 32 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 2 of 4)
- 2. (continued)
J. IF an action level exists, THEN ENSURE a narrative log entry is created to document this condition. ______
- 3. IF PI-View is unavailable OR at US/SM discretion, THEN RECORD data using RTGB indications. ______
RECORD results in Hand Calc Section. ______
Initial Conditions Final Conditions Notes Difference 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> minimum desired Date:____/____/____ Date:____/____/____ unless not at steady state.
_________Min Time:__________ Time:__________ (Chg. Pump data excluded)
(Attachment 2 data excluded)
TCOLD: _________°F TCOLD: _________°F Held constant (<.1°F) ** _________°F B.A. Integrator:_______ B.A. Integrator:______ Always Negative -_________gal.
PMW Integrator:______ PMW Integrator:_____ Always Negative -_________gal.
Notify SNPO to mark Notify SNPO to mark Zinc Injection Tank Zinc Injection Tank 60 minute minimum required.
gauge glass (LG-02-2) gauge glass (LG-02-2) -_________gal.
Always negative (gal) (gal)
+ if final > initial VCT Level:_______% VCT Level:_______% - if final < initial ( )________gal.
(% change x 33.8 gal. per %)
+ if final > initial Pzr Level:________% Pzr Level:________% - if final < initial ( )________gal.
(% change x 67.04 gal. per %)
Always Negative Q.T. Level:_______% Q.T. Level:_______% N/A if final < initial -_________gal.
(% change x per Table 2) 30 minute minimum required.
Notify SNPO to mark Notify SNPO to mark Always Negative.
Seal Tank level for ALL Seal Tank level for ALL -_________gal.
N/A if final < initial available Chg. Pps.
- available Chg. Pps. *
(1 inch = 0.59 gal.)
- Data should be recorded for any pump hydraulically connected to the system.
- If TCOLD is NOT held constant, consideration should be given to repeating calculation when conditions permit.
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 33 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 ATTACHMENT 4 Manual Leak Rate Calculation (Page 3 of 4)
- 3. (continued)
Table 2 Q.T. Level Range Q.T. Volume (gals/%)
60-61% 19.0 gals/%
61-66% 18.0 gals/%
66-67% 17.0 gals/%
67-68% 18.0 gals/%
68-69% 17.0 gals/%
69-70% 18.0 gals/%
- 4. DETERMINE Total Leakage of RCS by COMBINING differences of the following:
B.A. Integrator - _______gal.
PMW Integrator - _______gal.
Zinc Injection - _______gal.
VCT Level ( )_______gal.
Pzr Level ( )_______gal.
Total Leakage - _______gal.
- 5. DETERMINE Identified Leakage by COMBINING differences of the following components: ______
Q.T. Level - _______gal.
Chg. Pp. Seal Tanks - _______gal.
Attachment 2 (if used) - _______gal.
Identified Leakage = _______gal.
REVISION NO.: PROCEDURE TITLE: PAGE:
36 REACTOR COOLANT SYSTEM INVENTORY BALANCE 34 of 35 PROCEDURE NO.:
2-OSP-01.03 ST. LUCIE UNIT 2 INITIAL ATTACHMENT 4 Manual Leak Rate Calculation (Page 4 of 4)
- 6. DETERMINE Unidentified Leakage by SUBTRACTING Identified Leakage from Total Leakage. ______
Total Leakage ( )_______gal.
Identified Leakage - ( )_______gal.
Unidentified Leakage = ( )_______gal.
- 7. DETERMINE Unidentified Leakage rate by DIVIDING Unidentified Leakage value in Attachment 4, Step 6 by elapsed time in minutes. ______
Unidentified Leakage - _______gal.
Elapsed time _______min.
Unidentified Leakage rate =_______gpm
- 8. DETERMINE Identified Leakage rate by DIVIDING Identified Leakage value in Attachment 4, Step 5 by elapsed time in minutes. ______
Identified Leakage - _______gal.
Elapsed time _______min.
Identified Leakage rate =_______gpm
Post Exam Feedback - JPM S-7 The JPM directs the candidate to perform a Containment Purge for Refueling Operations, in accordance with 2-NOP-06.20, beginning with Section 4.2.1, Step 5. Performance Step 3 of the JPM is identifies a critical and states, ensure the Purge Mode selector switch is in the Refuel position prior to fuel movement.
This is contrary to 2-GOP-365, Refueling Sequencing Guidelines. The Purge Mode selector switch position, in support of refueling, is maintained/ aligned per 2-GOP-365, Refueling Sequencing Guidelines, Section 4.0, step 31. Specifically, step 31 states, ensure the following is complete prior to performing 0-NOP-67.05, Refueling Operation. . .Ensure the Purge Mode Selector switch is selected to the Refuel position. During refueling operations at PSL, this Purge Mode selector switch is positioned prior to refueling in accordance with the procedure for Refueling Sequencing Guidelines.
Station Recommendation: It is the stations position that reference to the purge mode selector switch in the task standard be removed from the JPM and performance step 3 be changed to not critical.
Recommended changes to JPM S-7 (Amplifying Information is outlined below)
Task Standards: The applicant will align and initiate a Containment Purge, and then secure the running Containment Purge Exhaust Fan when a valid Plant Vent alarm go into high alarm on RMCS by performing the following critical tasks:
PLACE PURGE MODE SELECTOR SWITCH in the REFUEL POSITION START either HVE-8A or HVE-8B CNTMT PURGE EXHAUST FAN, Start 1 FAN ONLY STOP the Running Exhaust Fan, either HVE-8A or HVE-8B CNTMT PURGE EXHAUST FAN, whichever was started earlier in the JPM
Performance Step: 3 1. IF purge is for refueling operation, THEN ENSURE PURGE MODE Critical YES NO SELECTOR is in REFUEL position prior to fuel movement.(Section 7.1.3, Management Directive 1 Standard: Applicant will place PURGE MODE SELECTOR SWITCH in the REFUEL POSITION NOTE:
This step is not critical to performing the purge, its position is also controlled by 2-GOP-305, Refueling Sequencing Guidelines