NLS2009022, Response to NRC Request for Additional Information Alternative Source Term
| ML091040565 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/08/2009 |
| From: | O'Grady B Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2009022, TAC MD9921 | |
| Download: ML091040565 (8) | |
Text
Nebraska Public Power District "Always there when you need us" 50.90 NLS2009022 April 8, 2009 U.S. Nuclear Regulatory Commission.
Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Nuclear Regulatory Commission Request for Additional Information Re: Alternative Source Term (TAC No. MD9921)
Cooper Nuclear Station, Docket No. 50-298, DPR-46
References:
- 1. Letter from Carl F. Lyon, U. S. Nuclear Regulatory Commission, to Stewart B. Minahan, Nebraska Public Power District, dated March 3, 2009, "Cooper Nuclear Station - Request for Additional Information Re:
Alternative Source Term (TAC No. MD9921)"
- 2. Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S. Nuclear Regulatory Commission, dated October 13, 2008, "License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences"
Dear Sir or Madam:
The purpose of this letter is for Nebraska Public Power District to submit a response to requests for additional information (RAI) from the Nuclear Regulatory Commission (NRC) (Reference 1). The RAI requested information in support of NRC's review of a license amendment request for the Cooper Nuclear Station facility operating license and technical specifications to adopt the Alternative Source Term for use in calculating the Loss-of-Coolant Accident dose consequences (Reference 2).
Responses to the specific RAI questions are provided in Attachment 1. No regulatory commitments are made in this submittal.
The information submitted by this response to the RAI does not change the conclusions or the basis of the no significant hazards consideration evaluation provided with Reference 2.
If you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.
COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 A U wwvw.nppd.com
NLS2009022 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed onf /'el ?
(date)
Sincerely, Brian J. O'Grady Site Vice President
/em Attachment cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS Nebraska Health and Human Services w/ attachment Department of Regulation and Licensure NPG Distribution w/o attachment CNS Records w/ attachment
NLS2009022 Attachment 1 Page 1 of 5 Attachment 1 Response to Nuclear Regulatory Commission Request for Additional Information Re: Alternative Source Term (TAC No. MD9921)
Cooper Nuclear Station, Docket No. 50-298, DPR-46 NRC Question #1 (from Reactor Systems Branch)
If Cooper Nuclear Station (CNS) intends to take credit of the standby liquid control (SLC) system for the purpose of controllingpH level in the suppressionpool (SP)water by injecting boron solution into the lower plenum of the reactorvessel during the long-term coolingphase of a loss-of-coolant accident (LOCA), concerns regardingthe possibility of boron precipitationin the CNS core should be addressed. The U.S. Nuclear Regulatory Commission staff understands that when the SLC system is initiated,boron solution injectedfrom the SLC system will mix with the coolantfrom the residualheat removal (RHR) system in the vessel lower plenum, and a portion of the mixture will then flow into the core where it will undergo sustainedboil-off in orderto remove the decay heat. The remainderof the boron solution can overflow through the jet pump and reach the SP. Duringthis process, the core should remain covered to two-thirds of its height, while core boil-off removes the decay heat. The staff requests the licensee to provide following additionalinformation:
a) Describe the flow path of boron solution mixture after initiationof the SLC system following a LOCA, including the path through the core.
b) Justify how sustained and continuous boil-off of water mixed with the boron solution inside the core will not cause boron concentrationto increasewith time, and ultimately, precipitatein the core during the RHR cooling mode of a LOCA.
c) Describewhat measures are taken to monitor boron concentration in the core after the SLC system is initiatedfollowing a LOCA, and how to prevent potential boron precipitationin the core due to sustainedboil-off in the core.
Response
a) The sodium pentaborate solution from the Standby Liquid Control (SLC) tank is piped into the reactor vessel and discharged near the bottom of the core shroud so it mixes with the cooling water rising through the core. The boron solution will mix with the Emergency Core Cooling Systems (ECCS) flow and spill over to the drywell through the break and then to the Suppression Pool. See Figure 1.
NLS2009022 Page 2 of 5 Post LOCA - RHR System takes suction from Torus and injects VE CORE JET PUMP into the RPV via RRP discharge kTETAP line, through Jet Pumps, into VESSEL WALL LCINJECTION lower plenum, up through core, SPARGERIN SLCINJECTION LINE up & out Steam Seperators and WER PLENUM Dryers, down the downcomer JET PUMP DELTA-P CORE DELTA-P section of RPV, out the break in OTHER INDICATORS the RR Suction Line, into Drywell, STANDBY LIQUID CONTROL- CORE nxp LINE (Simplified Diagram) and drains back to Torus.
I I RECIRCULATION SYSTEM PRIMARY CONTAINMENT SYSTEM (Simplified Diagram) (Simplified Diagram)
Figure 1.
NLS2009022 Page 3 of 5 In the event of a Design Basis Accident Loss of Offsite Power concurrent with a Loss of Coolant Accident (DBA-LOOP-LOCA), two Residual Heat Removal (RHR) pumps and one Core Spray (CS) pump will be initially available to circulate at least 18,050 gpm of water from the suppression pool to the reactor vessel. Based on the maximum pool inventory of 103,979 ft3, this ECCS flow represents one complete exchange of the Suppression Pool volume every 43 minutes. After 600 seconds, it is assumed that one RHR pump is reconfigured to Suppression Pool cooling. On this basis, the Suppression Pool is assumed to be well-mixed such that a single pool pH value can be applied. After injection, a fixed quantity of boron and water continue to be circulated in a closed loop flowpath.
The injection of the flow from the RHR pump and the CS pump will also keep the sodium pentaborate solution in the reactor vessel well mixed, both during the initial injection and post-injection periods. The total ECCS flow is substantially more than the SLC injection rate of 38.2 gpm per pump (76.4 gpm total), and will be well mixed before it overflows to the drywell and then to the Suppression Pool.
b) Since Alternative Source Term (AST) in this application only applies to the DBA-LOCA, as stated above, the sodium pentaborate solution will be well mixed in the reactor vessel with the circulating ECCS flow from RHR and CS pumps. Once initiated, the contents of the SLC tank (up to 4416 gallons) will be injected in less than two hours, assuming only one pump in operation. During the time the sodium pentaborate is being injected one to two complete exchanges of the pool volume will occur. This will result in a steady increase in the boron concentration in the pool during the injection process.
SLC is designed to provide a minimum of 660 ppm boron in the reactor vessel after injection. To ensure this minimum concentration is achieved, the SLC volume is determined based on a target value of 125%, or 825 ppm. The total weight of water in the reactor (including the recirculation loops) is 706,000 pounds, and the RHR system is an additional 256,000 pounds, for a total of 962,000 pounds. This is equivalent to approximately 15,506 ft3 of water. The volume of the Suppression Pool is estimated to be 103,979 ft3 . When these two volumes are combined after a LOCA, the total volume is approximately 119,485 ft3 . This will provide a dilution factor of about 7.7, and will result in a final boron concentration of about 10 ppm in the water being circulated by the ECCS system. This is significantly below the saturation concentration for the sodium pentaborate solution. At a temperature of zero degrees centigrade, the saturation concentration of sodium pentaborate in solution is eight percent.
Based on the low concentration of boron in the Suppression Pool and Reactor Vessel, the short period during which boron injection takes place, the fact that the solution is well mixed from the start of injection, and will remain well-mixed due to the high volume of ECCS flow between the Suppression Pool and the Reactor Vessel, and the internal mixing due to natural convection in the core, boron precipitation is not considered to be a concern for Boiling Water Reactors.
NLS2009022 Page 4 of 5 c) As described in Item b) above, due to the low boron concentrations, high flow rates and good mixing of sodium pentaborate in the water being circulated, precipitation in the core due to sustained boil-off is not a concern for the BWR reactors. Post-accident sampling during a LOCA will be performed using the Post Accident Sampling System, which permits sampling the Reactor Coolant and the RHR System water. Samples are analyzed for pH, and may also be analyzed for boron. Since the water will be well-mixed, no additional sampling in the core region is necessary.
NRC Question #2 (from Mechanical and Civil Engineering Branch)
Page 1.183-8 of Regulatory Guide (RG) 1.183 (Section 1.3.2, "Re-Analysis Guidance) denotes the responsibility of licensees to evaluate the nonradiologicalimpacts of an alternative source term implementation and the impacts of any associatedplant modifications. In Section 4.1.7 of your October 13, 2008, submittal, you indicated that variousportions of the heating, ventilation, and air conditioning (HVA C) systems for the Control Room (the Control Room air conditioning system and the Control Room emergency filtration system) would be credited in calculatingthe LOCA dose consequences. Please discuss the methodology used for the seismic qualification of the Control Room HVAC systems.
Response
The portions of the Control Room Heating Ventilation and Air Conditioning (HVAC) system and Control Room Emergency Filter System (CREFS) credited for dose reduction in the AST LOCA analysis are classified as essential and seismic Class IS.
The analysis summarized in Attachment 1, Section 4.1.7 of the October 13, 2008 submittal conservatively assumes that the Control Room HVAC system (normal ventilation system) remains in operation for the first minute of the event. So, no credit is taken for dose reduction during operation of this system. The air conditioning function of the Control Room HVAC is non-essential, and the Air Conditioning Unit (HV-AC-(AC-C-1A)) is non-essential. The ductwork, Supply Fan (HV-FAN-(SF-C-1A & 1B) are classified as essential and Class IS. By design, the normal Control Room ventilation will isolate upon receipt of a Group 6 signal, and CREFS is started within 11 seconds.
The CREFS system is credited for the filtration of air into the Main Control Room. The components of the CREFS system credited in the evaluation, including the Control Room Emergency Bypass Pre-Filter (HV-F-(PF-C-1A)), the High Efficiency Filter (HV-F-(HEF-C-1A)), the Carbon Filter for Control Room Emergency Bypass (HV-F-(CF-C-1A)), the Control Room Emergency Supply Fan (HV-FAN-(BF-C-1A)), and the Emergency Supply Fan Motor (HV-MOT-(BF-C-1A)) are classified as essential and Class IS.
NLS2009022 Page 5 of 5 As stated in USAR XII. 2.3.5.1.1, Class I Structures and Equipment, the seismic design for Class I structures and equipment is based on dynamic analyses using acceleration response spectrum curves which are based on a ground motion of 0.1 g for the Operating Basis Earthquake. In addition, the seismic design for Class I structures and equipment required for safe shutdown is also based on dynamic analyses using acceleration response spectrum curves which are based on a ground motion of 0.2 g for the Safe Shutdown Earthquake. Seismic design may also be based on the Generic Implementation Procedure Revision 3 (GIP-3) earthquake experience based method.
I I
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© 4 4
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© Correspondence Number: NLS2009022 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None 4 .4 4 4 4 .4 4 -4 4 .4 1 .4 I 4 4 4 4 4 I 4 4 .4 A