L-MT-09-006, Response to Request for Additional Information for the License Amendment Request Revising the Required Actions to Specification 3.5.1, Emergency Core Cooling System
| ML090280576 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/27/2009 |
| From: | O'Connor T Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-09-006, TAC MD9170 | |
| Download: ML090280576 (11) | |
Text
January 27,2009 L-MT-09-006 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Response to Request for Additional Information for the License Amendment Request Revising the Required Actions to Specification 3.5.1, Emergency Core Cooling Svstem
/TAC No. MD9170)
On June 26,2008, the Nuclear Management Company, LLC a predecessor license holder to the Northern States Power Company - Minnesota (NSPM),") submitted a request (Reference Iof Enclosure I) to revise the Required Actions for Specification 3.5.1, "Emergency Core Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) System, ECCS - Operating," within the Monticello Nuclear Generating Plant (MNGP) Technical Specifications to more accurately reflect the assumptions of the MNGP Loss of Coolant Accident analysis.
Enclosure Ito this letter contains the NSPM response to a U.S. Nuclear Regulatory Commission request for additional information (Reference 2 of Enclosure 1). On November 6,2008, a teleconference was held between the NRC and NSPM personnel, in which the NRC clarified their proposed question as discussed herein.
Summary of Commitments No new commitments or changes to any existing commitments are proposed by this letter.
In accordance with 10 CFR 50.91, a copy is being provided to the designated Minnesota Official.
- 1. NSPM is incorporated as a wholly owned subsidiary of Xcel Energy, Inc. Transfer of operating authority from the Nuclear Management Company, LLC to NSPM occurred on September 15, 2008.
2807 West County Road 75 a Monticello, Minnesota 55362-9637 Telephone: 763.295.5151 Fax: 763.295.1454
Document Control Desk Page 2 of 2 rjury that the foregoing is true and correct. Executed lo Nuclear Generating Plant any Minnesota Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce
ENCLOSURE I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION On June 6,2008, the Nuclear Management Company, LLC a predecessor license holder to the Northern States Power Company - Minnesota (NSPM),(') submitted a license amendment request (Reference I ) to revise the Required Actions for Specification 3.5.1, "Emergency Core Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) System, ECCS - Operating," within the Monticello Nuclear Generating Plant (MNGP) Technical Specifications (TS) to more accurately reflect the assumptions of the MNGP Loss of Coolant Accident (LOCA) analysis.
The U.S. Nuclear Regulatory Commission (NRC) requested additional information (RAI) on the basis for this proposed change by e-mail (Reference 2). On November 6, 2008, a teleconference was held between the NRC and NSPM personnel, in which the NRC clarified their proposed question.
Your application identifies the worst-case event as a recirculation line break event. For this event, protection systems appear to respond with acceptable consequences. Please discuss a feedwater line break event and protection systems response, showing that the consequences are not greater than those of the recirculation line break event. In a feedwater line break event, is the RPS qualified for the environment? Does the operator have to take actions? if so, show that the operator can act on a timely basis to depressurize the reactor vessel so that other mitigating equipment can function.
November 6, 2008, NRC Conference Call Clarification For a feedwater line break event injection capability via the High Pressure Coolant Injection System will be lost. For this situation low pressure ECCS injection will be dependent on Automatic Depressurization System actuation or operator action to depressurize the reactor. Discuss this situation in detail.
A. Design Basis LOCA Determination 10 CFR 50, Appendix K requires that all potential break locations be considered when evaluating the plant response to a LOCA. As described in Section 14.7.2 of the MNGP Updated Safety Analysis Report (USAR), all possibilities for pipe break sizes and locations were investigated including severance of small pipe lines, the main steam lines upstream and downstream of the flow restrictors, and the recirculation loop lines. These evaluations included a feedwater line break.
- 1. NSPM is incorporated as a wholly owned subsidiary of Xcel Energy, Inc. Transfer of operating authority from the Nuclear Management Company, LLC to NSPM occurred on September 15, 2008.
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ENCLOSURE 'I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION The BWR Emergency Core Cooling Systems (ECCS) performance analyses demonstrate that the most limiting breaks are liquid line breaks (those below the elevation of the top of the core). The limiting break determined in the generic evaluations which results in the most severe nuclear system effects and greatest release of radioactive material to the primary containment was determined to be a complete circumferential break of one of the recirculation loop suction lines. The recirculation line is the largest line connected to the reactor vessel at a low elevation relative to the core. This accident was established as the MNGP design basis LOCA (DBA-LOCA).
The MNGP ECCS performance evaluation (Reference 3) considered various breaks ranging in size from a 0.05 square-foot to a maximum recirculation suction line break. The analysis also evaluated ECCS performance for four non-recirculation line breaks for their maximum break areas. The maximum recirculation line break size in addition to the four non-recirculation line break sizes are listed below. These non-recirculation line breaks are not limiting, in terms of Peak Cladding Temperature (PCT),
because they are located at a relatively high elevation (in comparison to the top of the core).
Break Location (Maximum Area) Break Area (in square-feet)
Recirculation Suction Line 4.095 Main Steam Line (Inside Containment) I.81(')
Main Steam Line (Outside Containment) I.67(')
Feedwater Line 0.51 Core Spray Line 0.21 (I)Steam line break areas are prior to Main Steam Isolation Valve closure.
B. Feedwater Line Break and Protection Svstems Response As discussed above, the ECCS performance evaluation considered a feedwater line break but determined that it was bounded by other events with respect to break size area and PCT response, and hence was not a limiting event with regards to ECCS performance capability.
For a feedwater line break inside or outside containment the reactor protection system (RPS) instrumentation will detect a decrease in reactor water level. Low reactor water level (plus 9 inches) is indicative that the reactor core is in danger of being inadequately cooled. Additionally, for a feedwater line break inside containment the RPS instrumentation will detect an increase in containment pressure. High primary containment pressure Page 2 of 6
ENCLOSURE l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (1.84 psig) is indicative of a break in the primary system process barrier.
These are the same instruments utilized to detect a DBA-LOCA, i.e., a recirculation suction line break. A reactor scram is initiated if either setpoint is exceeded.
Depending on the size and location of the feedwater line break, the High Pressure Coolant Injection (HPCI) may be able to provide makeup to the reactor pressure vessel (RPV),(*)however, HPCl System operation does not have a significant effect on the overall ECCS performance for large breaks.
The feedwater line break maximum flow area of 0.51 square-feet is larger than the HPCl System was designed to mitigate. Due to this condition, the reactor pressure vessel (RPV) will depressurize for this event without operator action. This effect can be seen from a review of Figure C-3b and Figure C-3e, provided in Enclosure 2. These figures are taken from Reference 3, which supported the 1996 MNGP power uprate, and provides the reactor vessel pressure and ECCS flow for the nominal case in response to a feedwater line break (with worst case equipment out-of-service assumptions).
The required operator actions for a feedwater LOCA are unchanged from those required for a recirculation suction line break. For the DBA-LOCA accident, no operator action is credited for the first ten minutes of the event.
At ten minutes into the event for a DBA-LOCA the reactor has depressurized and the low pressure ECCS pumps have injected to restore adequate core cooling.
C. Environmental Qualification The RPS instruments for detecting a decease in reactor water level or high primary containment pressure are environmentally qualified (EQ) in accordance with 10 CFR 50.49. The qualification of these instruments is to the bounding temperature, pressure and radiation conditions the instruments may experience when required to perform their safety-related functions. The feedwater line break event was considered when determining the bounding environmental conditions within and external to containment.
- 2. For example, if the break does not occur on the feedwater line that HPCl injects into, HPCl can supply make-up to the RPV since the inboard feedwater isolation check valve will prevent reactor coolant being lost through the break. Another example, is if the break size is sufficiently small HPCl can maintain reactor inventory.
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ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION D. Discussion of HPCl and the Combined Operation of ADS and the Low Pressure ECCS The HPCl System is designed to provide core cooling at high reactor pressures. The Automatic Depressurization System (ADS) provides a backup to HPCl by automatically depressurizing the reactor vessel to permit operation of the low pressure ECCS.
Hiqh Pressure Coolant lniection System The HPCl System is designed to provide core cooling over a wide range of reactor vessel pressures (150 psig to 1120 psig). The HPCl steam driven turbine pump transfers water to the core utilizing some of the feedwater system piping, where the coolant is then distributed within the RPV by the feedwater sparger. The steam supply to the HPCl turbine is from a main steam line (MSL) upstream of the inboard main steam isolation valve (MSIV).
Reactor Pressure Relief Svstem The Reactor Pressure Relief System (RPRS) consists of eight safetylrelief valves (SRV) located on the MSLs within the drywell between the RPV and the inboard MSIVs. The SRVs are designed to be self-actuating on overpressure and are capable of remote operation by an air actuator.
Three of the SRVs are assigned to the Automatic Depressurization System (ADS) function. Another three of the SRVs are assigned to the Low-Low Set System (LLS) function.
Automatic Depressurization Svstem The function of the ADS is to provide a backup to the HPCl System by automatically depressurizing the reactor vessel to permit the low pressure coolant injection systems (LPCI) andlor the core spray operation.
The ADS, which is a subset of the RPRS, is an automatic actuation logic which remotely operates three of the eight SIRVs via an air actuator. Since the ADS is made up of multiple independent SRVs, it is single-failure proof and is considered highly reliable. There are two ADS trip systems. Either of which will cause all the ADS relief valves to open. Automatic initiation occurs when signals indicating Reactor Vessel Water Level - Low Low and Core Spray or LPCI Pump Discharge Pressure - High are all present and the ADS initiation timer has timed out. The LPCI and Core Spray pumps discharge pressure is Page 4 of 6
ENCLOSURE 7 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION used as a permissive for ADS actuation, indicating a source of coolant is available once ADS depressurizes the vessel.
No operator action is required for ADS actuation. However, if both independent trip systems of the initiation logic were to fail, the operator could manually initiate ADS (by manipulation of each ADS valve control switch). Additionally, the operator can open any of the SRVs to blow-down the reactor vessel to the suppression pool if so required.
Therefore, diverse means of reactor depressurization are available to the operator.
The ADS is designed to automatically depressurize the RPV during a small break LOCA if the HPCl System fails or is unable to maintain required water level in the RPV. The ADS operation depressurizes the RPV to within the operating pressure range of the low pressure ECCS subsystems (Core Spray and LPCI), so these subsystems can provide coolant inventory makeup. For large breaks, including a feedwater line break, the RPV depressurizes through the break without assistance, i.e., ADS operation is unnecessary.
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ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REFERENCES
- 1. NMC letter to NRC, "License Amendment Request: Revision to Required Actions for Specification 3.5.1, Emergency Core Cooling System,"
(L-MT-08-034), dated June 26, 2008.
- 2. Email from P. Tam (NRC) to R. Loeffler (NMC) dated September 23, 2008, "Draft RAI re: proposed amendment to TS 3.5.1, ECCS (TAC MD9170).17
- 3. General Electric Licensing Topical Report, NEDC-32514P, Revision 1, "Monticello SAFERIGESTR-LOCA Loss-of-Coolant Accident Analysis," dated October I997.
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ENCLOSURE 2 MONTICELLO NUCLEAR GENERATING PLANT FIGURES C-3b AND C-3e FROM NEDC-32514P, REVISION 1, MONTICELLO SAFERIGESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS
System Response Curvesfor Nominal Non-Recirculation Line Break C-I6
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