NRC 2011-0086, Clarification/Comments on NRC Safety Evaluation Report, Amendment Nos. 238 (Unit 1) and 242 (Unit 2), Auxiliary Feedwater System Modification
| ML112770439 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/16/2011 |
| From: | Meyer L Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2011-0086, TAC ME0181, TAC ME0182 | |
| Download: ML112770439 (6) | |
Text
Y BEACH September 16, 201 1 NRC 201 1-0086 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Clarification/Comments on NRC Safetv Evaluation Report Amendment Nos. 238 (Unit I) and 242 (Unit 2)
Auxiliary Feedwater Svstem Modification TAC Nos. ME-0181 (Unit 1) and ME-0182 (Unit 2)
On March 25,201 1, the Commission issued License Amendment Nos. 238 and 242 to Renewed Facility Operating License Nos. DPR-24 and DPR-27, respectively, for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Enclosure Reference 1). The amendment provides changes to the auxiliary feedwater (AFW) system design and the Technical Specifications in response to the NextEra Energy Point Beach, LLC (NextEra) application dated April 7, 2009, as supplemented by letters dated June 17 (two letters), September II , September 25, October 9, November 20 (two letters), November 21 (two letters), November 30, December 8, December 16 of 2009; January 7, January 8, January 22, February 11, February 25, March 3, April 15, April 22, April 28, July 8, July 28, August 2, August 9, August 24, October 15, November 1, November 12 (two letters),
November 30, and December 21 of 2010. The proposed changes were originally included as part of the April 7, 2009, extended power uprate (EPU) license amendment request, but subsequently divided into a separate licensing action for independent technical review.
The amendment changed the AFW system design and Technical Specifications (TS) 3.7.5, "Auxiliary Feedwater (AFW)," and TS 3.7.6, "Condensate Storage Tank (CST)," resulting from:
- 1) modifications to the AFW system to support requirements for transients and other accidents at EPU conditions; 2) automatic AFW switchover from a CST suction source to a safety-related Service Water source; and 3) instrumentation setpoint changes supporting the aforementioned physical modifications. The upgrades and modifications to the AFW system have been installed to provide additional capacity and reliability for the system. Although the changes are also designed to support the requirements for transients and other accidents at EPU conditions, the changes for this amendment were evaluated by the NRC staff using the licensing basis prior to approval of the extended power uprate license application.
NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
Document Control Desk Page 2 During August 2011, representatives of NRC Region Ill Division of Reactor Safety conducted a component design basis inspection (CDBI) at PBNP. During this inspection, questions arose concerning the capability of the upgraded motor-driven auxiliary feedwater pumps (MDAFWPs).
The questions stemmed from ambiguity in the characterization of the system capabilities as described in the Safety Evaluation Report (Enclosure Reference 1). The enclosure of this letter documents the discussion held during a teleconference with the NRC staff on September 14, 201 1, to clarify the license basis analyses and design capabilities of AFW system.
The information contained in this letter does not alter the no significant hazards consideration contained in License Amendment 261, Extended Power Uprate, or subsequent submittals thereto, and continues to satisfy the criteria of 10 CFR 51.22 for categorical exclusion from the requirements of an environmental assessment.
This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on September 16,201 1.
Very truly yours, NextEra Energy Point Beach, LLC Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW Mike Verhagen, State of Wisconsin
ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 CLARIFICATIONICOMMENTS ON NRC SAFETY EVALUATION REPORT AMENDMENT NOS. 238 (UNIT 1) AND 242 (UNIT 2)
AUXILIARY FEEDWATER SYSTEM MODIFICATION TAC NOS. ME-0181 (UNIT I ) AND ME-0182 (UNIT 2)
Background
During August 201 1, representatives of NRC Region Ill Division of Reactor Safety conducted a component design basis inspection (CDBI) at PBNP. During this inspection, questions arose concerning the capability of the upgraded motor-driven auxiliary feedwater pumps (MDAFWPs).
The questions stemmed from ambiguity in the characterization of the system capabilities as described in the NRC Safety Evaluation Report for the auxiliary feedwater system (AFW) modifications dated March 25, 201 1 (Reference 1). This letter formally documents the telephone conference held with representatives of Region Ill and the NRR staff on September 14, 201 1, to clarify the license basis analyses and design capabilities of the AFW system.
In Section 3.4 of the safety evaluation report (SER) (Reference I ) , Design Basis Accidents, AFW is credited with addressing six design basis accidents:
Loss of Non-Emergency AC Power to the Station Auxiliaries (LOAC)
Loss of Normal Feedwater Flow (LONF)
Steam Generator Tube Rupture (SGTR)
Secondary System Pipe Ruptures 0 Anticipated Transients Without Scram (ATWS)
Small break Loss of coolant accident (SBLOCA)
With the exception of ATWS (because AFW has no effect upon the PBNP ATWS safety analysis),
each of the above design basis accidents were discussed during the telephone conference on September 14, 201 1.
LONF I LOAC Section 3.4.1 of the SER addresses both LONF and LOAC. As analyzed, these events do not result in a loss of a steam generator (SG), and the AFW system is credited with delivering a minimum of 275 gpm to the two steam generators with the flow split approximately equally between the two. If limited to only the MDAFWP (as is assumed in the analysis), the maximum flow that can be delivered to a single SG is approximately 230 gpm. This is due to the presence of a cavitating flow venturi installed in the MDAFWP discharge lines feeding each SG. The design function of the venturi is to limit the AFW flow to a faulted SG in the event of a main steam line break (MSLB). Please refer to the section on MSLB for further details of the venturi installation.
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The third paragraph of Section 3.4.1 of the SER states in part that "...The licensee states that with the new MDAFW pumps. .. 100 percent flow, minimum of 275 gpm, will be obtained within 150 seconds to one or split [emphasis added] equally between two SGs." While 275 gprn of flow to a single SG would be sufficient to successfully mitigate the subject transients, it is not within the capability of the AFW system as analyzed for the LONF / LOAC events.
Section 3.4.2 of the SER addresses a postulated SGTR event. The analysis of a SGTR event is primarily concerned with potential radiological consequences, and not with the maintenance of a secondary heat sink. AFW performs a supporting role in maintaining the heat sink, but this is typically not a limiting transient for AFW because the SG secondary side inventory is not depleted prior to a reactor trip. The substantial initial inventory in the SGs, combined with break flow from the RCS to the ruptured SG, preclude a significant challenge to being able to maintain an adequate secondary heat sink.
The fourth paragraph of Section 3.4.2 in the SER states:
"The licensee states that with the new MDAFW pumps, AFW flow will be initiated 60 seconds after the low-low SG water level setpoint is reached. Full AFW flow (i.e,,
100 percent flow) at a minimum of 275 gprn will be obtained within 150 seconds to one SG, or split equally between two SGs. For a SGTR event, operator action is required to isolate AFW to the ruptured SG once SG level has recovered and provide AFW flow to the intact SG to effect a rapid RCS cooldown."
Similar to the LONFILOAC event discussed previously, assuming that only a single MDAFWP is available results in the minimum 275 gprn flow being split to the two steam generators. After realignment by the operator, up to approximately 230 gprn could be delivered to the intact SG with the flow limited by the cavitating venturi.
In support of the upgrade of the AFW system and extended power uprate (EPU) license submittals, Point Beach Nuclear Plant (PBNP) introduced a new analysis that demonstrates the capability of the installed systems and operator actions to mitigate a postulated SGTR prior to the ruptured SG filling water-solid. This margin to overfill (MTO) analysis included a scenario in which the AFW system delivered only 275 gprn to the unit, split evenly to both SGs initially, and only 137.5 gprn to the intact SG following manual operator action to isolate AFW to the ruptured SG.
The analysis demonstrated that under such a scenario, the limited AFW flow to the intact SG, combined with the pre-existing SG secondary inventory, was sufficient to ensure an adequate heat sink in support of decay heat removal and the forced cool down needed to restore RCS subcooling prior to terminating break flow.
Section 3.4.3 of the AFW SER addresses a postulated MSLB event. The analyzed event is an overcooling transient and AFW flow is maximized in order to conservatively bound adverse consequences. Adverse consequences from the stand point of AFW flow are limited to the potential for containment over-pressurization for a break occurring inside containment. The discussion of the MSLB event describes a conservatively high assumed AFW flow rate of 1200 gprn used in the analysis of containment pressure.
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This 1200 gpm analytical limit was reduced to 1139 gpm by analysis in support of the EPU. To ensure that a faulted SG in combination with an additional single failure would not result in exceeding this flow rate, cavitating venturis had been incorporated into the design of the MDAFWP discharge piping. The venturis prevent flow from the MDAFWP to a faulted SG from exceeding approximately 230-240 gpm. One venturi is installed in each of the feed lines from the MDAFP to the SGs (two per unit; one per SG). They are located downstream of the flow control valves, and are sized to cavitate at approximately 240 gpm of flow. The cavitation chokes flow and prevents higher flows regardless of downstream pressure.
When operating in the cavitating regime, the venturis are insensitive to downstream pressure, and prevent excessive flow from the MDAFWP when a SG is partially or fully depressurized. This limits the AFW flow to a faulted SG, and ensures that some AFW flow remains available to an intact SG prior to manual isolation of flow to the faulted SG.
Test data has demonstrated that when operated at flow rates below the cavitation regime, the venturis function as constant hydraulic resistance devices; their flow coefficient (C,) remains essentially constant over the full range of sub-cavitation flow rates.
Small Break Loss-of-Coolant Accident Section 3.4.4 of the SER addresses a postulated small break LOCA. AFW flow is not credited in mitigating a LOCA, and no further clarification is needed.
Secondary System Pipe Ruptures Section 3.4.5 of the AFW SER addresses feedwater system pipe breaks (FLB) inside and outside of containment. This section concludes that postulated secondary system pipe failures are bounded by the analyzed MSLB.
In the subsequent NRC SER on EPU dated May 3,201 1 (Reference 2), the subject of a postulated feedwater line break (FLB) was discussed on Pages 180-182. In that discussion, the Commission affirmed that PBNP is not required to be analyzed for such an event, and the discussion also summarized a review that NRC had conducted of a feedwater line break at a similar facility (Ginna) (Reference 3). The conclusion was that the analysis performed for Ginna provided assurance that a postulated FLB at PBNP would not cause a safety concern.
Based on the contents of the Ginna SER and the supporting submittals (ML060540349 and ML060810218), the Ginna Nuclear Power Plant credits 235 gpm of AFW flow starting at 870 seconds following the postulated FLB. The delay is caused by the need to manually start a pump when prompted by procedure because the automatic safety related pumps are rendered inoperable by the combination of the initiating event and a single failure.
A similar combination of an initiating event and a single failure could similarly cause a loss of all automatic AFW flow at PBNP. However, as directed by procedure, approximately twice as much AFW flow could be delivered by manually aligning and restarting either the turbine-driven AFW pump (TDAFWP) or the MDAFWP (whichever has not failed), and by supplementing the flow with a standby steam generator feed pump (SSG).
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These details are not provided in the PBNP Safety Evaluation Report, but are relevant to understanding the acceptability of the PBNP AFW design without further detailed analysis of a FLB.
References
- 1. NRC Safety Evaluation Report for Amendment Nos. 238 and 242, Point Beach Nuclear Plant, Auxiliary Feedwater System Modifications (TAC Nos. ME-0181 and ME-0182)
(MLI 10230016)
- 2. NRC Safety Evaluation Report for Amendment Nos. 241 and 245, Point Beach Nuclear Plant, Extended Power Uprate (TAC Nos. ME-1044 and ME-1045 (MLI 10450159)
- 3. NRC Safety Evaluation Report for Amendment No. 97, R. E. Ginna Nuclear Power Plant, Inc.,
Extended Power Uprate (ML061380249)
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