ML20017A108
| ML20017A108 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/06/2017 |
| From: | Tekia Govan NRC/NRR/DRO/IRSB |
| To: | |
| Govan T, 415-6197, NRR/DRO | |
| References | |
| Download: ML20017A108 (6) | |
Text
Plant: Nine Mile Point Unit 1 Date of Event: 9/6/2017 Submittal Date:
Licensee
Contact:
Rose Demko Tel/Email: rose.demko@exeloncorp.com NRC
Contact:
Eric Miller Tel/Email: eric.miller@nrc.gov Performance Indicator:
Unplanned Scrams with Complications Site-Specific FAQ (see Appendix D)? ()Yes or (X) No FAQ to become effective when approved.
Question Section:
NEI 99-02 Guidance needing interpretation:
page 27, lines 1-9 Question 6, NEI-99-02 states, "Following initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs?
This step is used to determine if the scram was uncomplicated and did not require using other procedures beyond the normal scram response. Following the initial transient, maintaining reactor and drywell pressures below the Emergency Procedure entry values while ensuring reactor water level is above the Emergency Procedure entry values allows answering "No".
The requirement to remain in the EOPs because of reactor pressure/water level and drywell pressure following the initial transient indicates complications beyond the typical reactor scram."
Event or circumstances requiring guidance interpretation:
NRC POSITION:
The inspectors reviewed Nuclear Energy Institute 99-02; N1-EOP-2, RPV [reactor pressure vessel] Control, Revision 01600; the post transient review for the scram on September 6, 2017; and IR 04049445 and its associated root cause report, and determined that it appeared that a Scram with Complications should have been classified.
Question 6, NEI-99-02 states "Following the initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry condition for EOPs?"
The pressure control leg of N1-EOP-2 states "stabilize RPV pressure below 1080 psig using the Main Turbine Bypass Valves (TBVs)."
However, operators didnt have TBVs available because Main Steam Isolation Valves (MSIVs) were closed. Instead operators utilized an "alternate pressure control system" listed in N1-EOP-2, the Emergency Condenser (EC). Pressure was controlled using the EC for approximately 8 minutes before the MSIVs were opened and reactor pressure control was re-established using Page 1 of 5
the TBVs. Therefore, Question 6 should have been Yes in the Unplanned Scram with Complications (USwC) Performance Indicator.
The Frequently Asked Questions Log states (FAQ 18-01):
"Initial Transient is intended to envelope the immediate and expected changes to BWR parameters as a result of a scram (e.g., pressure, level, etc.) because of the collapsing of voids in the core and the routine response of the main feedwater and turbine control systems. For example, at some BWRs the reflected pressure wave resulting from the rapid closure of turbine valves during a turbine trip may result in a pressure spike in the reactor vessel that causes one or more safety-relief valves (SRVs) to briefly lift. The intent is to allow a licensee to exclude the momentary operation of SRVs when answering "Was pressure control unable to be established?" The sustained or repeated operation of SRVs in response to turbine control bypass valve failures or Main Steam Isolation Valve (Group I) isolations are not a part of routine BWR scram responses and are therefore not considered to occur within the initial transient."
Based on the inspectors review it appeared that Question 6 should have been answered "Yes,"
because the ultimate heat sink was lost with main steam isolation valves closed following the initial transient requiring additional time for the use of the alternate pressure control system, emergency condensers, as defined in N1-EOP-2.
SITE POSITION:
To answer NEI 99-02 BWR Flowchart Question 6, we will discuss the conditions of the scram and the design basis of the emergency condensers and then discuss each sentence of the question separately as follows:
The initiating event for the scram was a loss of all feedwater flow which caused a scram on Lo RPV Water Level (<53"). As part of this transient, RPV Water Level reached the Lo-Lo RPV Water Level setpoint of (<5") due to initial loss of feedwater and shrink following the scram and prior to RPV Water Level being restored as expected by High Pressure Coolant Injection (HPCI). The Lo RPV Water Level is an entry condition into N1-EOP-2, "RPV Control" and is entered as part of the normal scram response in addition to N1-SOP-1, "Reactor Scram" and N1-OP-43C, "Plant Shutdown." Once N1-EOP-2 is entered, the operators follow the EOP to control both RPV water level and reactor pressure. Due to initiating transient which caused loss of feedwater flow reaching the Lo-Lo RPV Water Level, a Vessel and Containment Isolation Signal occurred as expected which caused the Main Steam Isolation Valves to close. Operators established pressure control in accordance with N1-EOP-2 by manually initiating Emergency Condenser (EC) 11 during the initial transient (as shown in Attachment 1) and maintained reactor pressure below any further EOP entry conditions. This did not require using other procedures beyond the normal scram response and therefore did not require additional time for the use of EC 11. No ERVs/SRVs lifted during the scram response. Once pressure was stabilized (~8 minutes), the Main Steam Isolation Valves were re-opened and pressure control was transferred from the Emergency Condensers to the Turbine Bypass Valves. N1-EOP-2 was later exited as there had been no re-entry conditions and no other EOPs were entered as part of the scram response. As shown on Attachment 1, RPV Water Level and RPV Pressure responded as expected to the scram and no further equipment issues occurred during the scram response that cause complications required additional operator action to address. The entire transient spanned from the initiating event to the time the ECs were placed into operation.
In accordance with Nine Mile Point Unit 1 Technical Specifications the design basis of the emergency cooling system is to provide a redundant backup for core decay heat removal following reactor isolation and scram.
Page 2 of 5
In accordance with Nine Mile Point Unit 1 UFSAR, the design basis for the emergency condensers is to provide decay heat removal from the reactor fuel in the event that reactor feedwater capability is lost and the main condenser is not available. The emergency condensers serve as an alternate heat sink when the reactor is isolated from its normal heat sink (the main condenser).
As discussed above, the following is Question 6 and the associated response to each portion of the question.
"This step is used to determine if the scram was uncomplicated and did not require using other procedures beyond the normal scram response."
NMP Response:
For normal SCRAM recovery, the following procedures are used:
- 1. N1-EOP-2, "RPV Control"
- 2. N1-SOP-1, "Reactor SCRAM"
- 3. N1-OP-43C, "Plant Shutdown" No procedures were utilized during scram that were not part of the normal scram response.
N1-EOP-2 was entered due to Lo RPV Water Level. Use of the ECs to control pressure is allowed per N1-EOP-2 and N1-SOP-1, with allowance to cooldown using the ECs if the main condenser is not available. Response to a Lo-Lo RPV Water Level condition is part of N1-OP-43C which has guidance to reset and restore from this condition in conjunction with the normal scram recovery and cooldown procedure sections. Therefore, the answer is to this statement is "No."
"Following the initial transient, maintaining reactor and drywell pressures below the Emergency Procedure entry values while ensuring reactor water level is above the Emergency Procedure entry values allows answering "No."
NMP Response:
The EOP entry condition associated with Lo RPV Water Level is expected and occurs as part of the normal plant response to a scram. Lo Lo RPV Water Level is not an entry condition in any EOP procedure. RPV Water Level was restored as expected using HPCI. Operators established pressure control in accordance with N1-EOP-2 by manually initiating EC 11 during the initial transient (as shown in Attachment 1) and maintained reactor pressure below any further EOP entry conditions. Pressure control was maintained using EC 11 and then TBVs following restoration of the MSIVs. Following the initial transient, reactor pressure and drywell pressures remained below EOP entry conditions. The highest RPV pressure following the transient was 1005 psig which is well below the entry condition of 1080 psig. No other EOP entry conditions were met during the scram response. Therefore, the answer is to this statement is "No."
"The requirement to remain in the EOPs because of reactor pressure/water level and drywell pressure following the initial transient indicates complications beyond the typical reactor scram."
NMP Response:
The initial Lo RPV Water Level, post scram, was the only EOP entry condition setpoint met during or after the transient. The Lo RPV water level recovered to within normal operating band without Operator Actions as expected. No reactor or drywell pressure EOP entry conditions occurred during or after the transient. The highest pressure in the reactor during the transient Page 3 of 5
was 1005 psig, well below the EOP entry condition of 1080 psig. No conditions or equipment issues existed during the duration of the scram response requiring re-entry or extended operation in the EOP. Therefore, the answer is to this statement is "No."
Additionally, reactor water level scram signal(s) during the scram response indicate level could not be stabilized and require this question be answered "Yes".
NMP Response:
Once reactor water level recovered from the initial transient, reactor water level remained stable throughout the scram response. Therefore, the answer is "No.
Additional Clarifying Information Regarding use of Alternate Pressure Control System:
The NRC indicated that the NMP1 Operators used an "alternate pressure control system" as defined in N1-EOP-2 by using the ECs. The term "alternate pressure control system" is terminology used in N1-EOP-2. The plant responded as expected to an RPV Lo-Lo Level containment isolation. Operator's use of ECs with reactor vessel isolated is a procedural step in N1-EOP-2 and N1-SOP-1.
Note, in FAQ 18-01, SRVs are considered acceptable to momentarily lift during the initial transient, without being considered a scram with complications. Unlike SRVs, ECs are part of acceptable manual pressure control when MSIVs close. ECs are clearly referenced as part of the pressure control system in NEI 99-02, Question #2 of the BWR flowchart, while SRVs are not.
The relevant portion of NEI 99-02, Question 2 of the BWR flowchart, is provided below for clarity of the pressure control system components.
"The failure of the pressure control system (i.e., turbine valves / turbine bypass valves / HPCI / RCIC/isolation condenser) to maintain the reactor pressure or a failed open SRV(s) counts in this indicator as a complication beyond the normal reactor trip response and would result in a Yes response."
CONCLUSION:
In conclusion, the use of ECs as an "alternate Pressure Control System," as identified in N1-EOP-2, is a normal reactor trip response. As a result of a low RPV water level scram, at no time during the initial transient and during the scram response did the pressure control system, as described in N1-EOP-2, fail. No additional EOP entries were met after the expected initial entry on Lo RPV Water Level. There was no delay in exiting the EOP and SOP procedures due to the use of EC 11. Therefore, NMP maintains that question #6 is a "No" response.
If licensee and NRC resident/region do not agree on the facts and circumstances, explain:
The licensee and NRC concur on the facts and circumstances surrounding the event.
Potentially relevant FAQs:
FAQ 18 "Definition of Initial Transient" Page 4 of 5
Response Section:
Proposed Resolution of FAQ:
N/A If appropriate, provide proposed rewording of guidance for inclusion in next revision:
N/A PRA update required to implement this FAQ?
No.
MSPI Basis Document update required to implement this FAQ?
No.
Page 5 of 5
NMP1 Scram 9/6/17 Lo Reactor Water Level
~1157 - 13 Feedwater Flow Control Valve rapidly closed 11:57:15.401 - Reactor automatically scrams on low RPV level, HPCI initiation signal received 11:57:21 - 11 and 12 Feedwater Pumps start 11:57:34.003 - Lo-Lo RPV Water Level reached, vessel and containment isolation signals received 11:57:35.503 - All MSIVs shut 11:58:45 - 11 Emergency Condenser placed into service for RPV pressure control 12:03:31 - MSIVs reopened, Main Condenser re-established for RPV pressure control 12:04:14 - 11 Emergency Condenser removed from service
- Line denotes N1-EOP-2 entry condition for RPV Pressure (1080 psig)