ML051920032
| ML051920032 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/30/2005 |
| From: | Spina J Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP1L 1961, TAC MC6989 | |
| Download: ML051920032 (8) | |
Text
Constellation Energy-P.O. Box 63 Lycoming, NY 13093 Nine Mile Point Nuclear Station June 30, 2005 NMP1L 1961 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Nine Mile Point Unit 1 Docket No. 50-220 Facility Operating License No. DPR-63 Response to NRC Request for Additional Information Regarding Cold Water Reactivity Event of January 9, 2004 (TAC No. MC6989)
Gentlemen:
In a letter dated May 26, 2005, the NRC requested additional information regarding the cold water reactivity event of January 9, 2004, at Nine Mile Point Unit 1. The Nine Mile Point Nuclear Station, LLC (NMPNS) response to the request for additional information (RAI) contained in the NRC's May 26, 2005, letter is provided in Attachment 1. This letter contains no new regulatory commitments.
If you have any questions about this submittal, please contact M. Steven Leonard, NMPNS General Supervisor Licensing, at (315) 349-4039.
Very truly yours, C4.,
James A. Spina Vice President Nine Mile Point Attachments:
Response to NRC Request for Additional Information (RAI) Regarding Cold Water Reactivity Event of January 9, 2004 JAS/TFS/sc cc:
Mr. S. J. Collins, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Aco Mr. T. G. Colburn, Senior Project Manager, NRR Mr. J. P. Spath, NYSERDA
ATTACHMENT I Nine Mile Point Nuclear Station, Unit 1 Response to NRC Request for Additional Information (RAI)
Regarding Cold Water Reactivity Event of January 9, 2004 This attachment provides the Nine Mile Point Nuclear Station, LLC (NMPNS) response to the request for additional information contained in the NRC letter dated May 26, 2005, regarding the Cold Water Reactivity Event of January 9, 2004. Each question contained in the NRC RAI is repeated, followed by the NMPNS response.
RAI (1)
Have you previously analyzed the initiation of the emergency condenser (EC) atfillpower with an idle loop (or with allfive recirculation pumps in service) with regard to expected reactivity transient effect on the core? Is this event enveloped by such an analysis and do operating procedures address recovery from such an event? DER 1-1999-3510/3551/3583, page 4 of 5, indicates that no such analysis has ever been conducted. Without such an analysis, is the fill effect of the reactivity transient and resultant thermal margin~s) decreases well understood?
Response
The analysis of the reactivity transient effect on the core associated with the inadvertent initiation of a cold emergency condenser from 100% power was previously analyzed and determined to be' bounded by at least two of the design basis transients (100 degree Fahrenheit loss of feedwater heating, start of a cold recirculation loop), as described in the Unit 1 Final Safety Analysis Report Updated (UFSAR), and is considered an analyzed event.
The Emergency Cooling (EC) system test is enveloped by these design bases analyses since the EC test is initiated from reduced power in addition to the fact that the loss of feedwater heating associated with the inadvertent initiation is eliminated in the test procedure by placing the reactor pressure control on the turbine bypass valves prior to the EC system initiation. The following three operations procedures address recovery from an inadvertent EC initiation:
- Special Operating Procedure N1-SOP-1.5, Unplanned Reactor Power Change
- Alarm Response Procedure N1-ARP-K1, Control Room Panel KI
- Operating Procedure N1-OP-13, Emergency Cooling System The Deviation Event Reports (DERs) referenced (1-1999-3510, 1-1999-3551 and 1-1999-3583) were initiated to address the apparent omission of specific design basis thermal transient loading on the recirculation nozzles (e.g., cold recirculation loop start, idle recirculation loop start, cold EC initiation). The DER (1-1999-3551) disposition established that the emergency condenser initiation thermal transient was an analyzed event with regard to the maximum potential Page 1 of 7
C reactivity transient. The DER concluded no additional corrective actions were warranted with regard to the reactivity transient. The issue described in these DERs which required corrective action was confirmation that the design basis stress analysis adequately addressed the thermal loading on the recirculation nozzles. While the design basis stress analysis clearly states that the feedwater nozzle thermal transient is bounding for the recirculation nozzles, the basis for this conclusion was not discussed. The corrective actions reconstituted the design basis stress analysis specifically accounting for the design basis events such as the emergency condenser initiation and isolated recirculation loop starts. This reconstituted analysis explicitly addressed all the thermal transient event combinations and concluded that the original design basis statement was correct and the feedwater nozzle thermal stress analysis was bounding.
Note: DERs 1-1999-35 10 and 1-1999-3583 were duplicate DERs and were closed referencing DER 1-1999-3551.
RAI (2)
Wiat is your basis for believing that the average power range monitor (APRM) flux spike had reached its maximum peak? The nuclearffuel vendor's reports evaluating this event relied heavily on this assertion. Your evaluation relied heavily on the belief that APRMflux had peaked at approximately 17%. What was the peak flux response for local power range monitors (LPRMs) in the core quadrant most affected by this event? Thie General Electric analysis notes that at high powver, the reactors response to a cold water reactivity transient is magnified. This necessitated additional steps prior to conducting the second EC test at higher power level to ensure adequate margin. Does this indicate that the effects of the test on reactivity and thermal margin were not well understood before the conduct of the first EC surveillance test at high powver?
Response
The design basis for the NMPNS Unit 1 APRM system relies on a quadrant based local power range monitor (LPRM) assignment. Each APRM channel has eight (8) A&C or B&D level LPRMs assigned from a quadrant. This system is designed to monitor the core on a quadrant basis and capture potential quadrant based core inlet transients that could occur because of the design of the recirculation system. The Boiling Water Reactor (BWR)-2 recirculation system provides for mixing of the recirculation pump discharge in the lower plenum and effectively mixes the flow on a quadrant basis. This system is fully capable of monitoring the maximum flux response associated with the core inlet sub-cooling transient associated with an emergency condenser system initiation. The EC system initiation transient duration is less than one minute.
The APRM channels were recorded on a high speed data acquisition system with a sample rate of greater than 10 samples per second and on analogue chart recorders capable of capturing the maximum peak associated with the EC system initiation.
Prior to the first EC test, the evaluation of the peak APRM response was performed applying a conservative assumption that the core inlet temperature was the same as the recirculation discharge temperature assuming no mixing in the lower plenum. The reactivity assessment conservatively applied a maximum change in core inlet temperature to calculate the flux Page 2 of 7
w increase. This assessment confirmed the measured APRM response was not biased low (see the response to RAI (5)).
The flux response associated with a sub-cooling event is proportional to the initial power level and this flux response was anticipated for both the first and second EC tests. The reactivity transient associated with the first EC test was understood and sufficient margin to reactor scram and thermal limits was established for both EC tests. Additional margin was added for the second EC test to ensure that the flux increase would conservatively remain below the flow biased rod block setpoint. The rod block was used as an abort criterion as a conservative threshold which was easily implemented and provided a good test control method. Preventing a rod block was not required to ensure adequate margin to thermal limits.
The second EC test was also performed with additional turbine control bypass valve margin. The first test maintained stable reactor pressure control without any pressure transient. The second test increased the regulation margin as a precaution. This additional regulation margin was not required to ensure adequate pressure control.
RAI (3)
What is your basis for now concluding, pursuant to 10 CFR 50.59(c)(2), that you did not need to obtain a license amendment prior to implementing this surveillance test procedure change?
Have you reached that same conclusion for both the initial test at high power (with one idle recirculation loop) and the second test at high power (with allf ive recirculation loops in operation)? Has the effect of EC initiation at high power been adequately analyzed to ensure adequate core thennal margin?
Response
The NMPNS Unit I technical specification requires an EC system performance test on a five year interval. Performance of this test requires operation of the EC system with the reactor in the power operating condition. No restrictions are placed on EC operations in the technical specification or in the UFSAR. The NMPNS Unit 1 design basis and transient analysis addresses operation of the EC system during the power operating condition relative to the thermal transient on both the reactor vessel nozzles and the reactivity transient associated with the EC system initiation from a cold condition (i.e. inadvertent start of an EC is an analyzed event from 100%
power). The EC test was reviewed through a 10CFR50.59 screen which concluded that the EC system surveillance test was within the design basis. The review of the EC surveillance history concluded that the EC test had previously been performed above 25% power with pressure control using the turbine bypass valves. The overall conclusion reached was that no unreviewed safety question exists associated with the EC test procedure for either the four loop or five loop configuration.
The EC system initiation subcooling transient was conservatively evaluated considering a maximum temperature change assuming the entire core experienced the maximum subcooling event. This ensured the effect of the EC initiation was conservatively analyzed to ensure core thermal margin. The EC test procedure ensured that the test was performed within the required thermal limit margins and ensured margin to reactor scram for both EC system tests. Post EC test Page 3 of 7
review confirmed that adequate thermal limit margin and margin to scram were maintained for both EC tests performed in January 2004.
RAI (4)
Have you considered the potential impact on the reactivity transient Ma turbine bypass valve(s) fails to modulate during the EC surveillance test? Ifnot, considering the significant influence of the turbine bypass valve operation on the magnitude of any reactivity transient, what is your basis for not considering such afailure?
Response
Yes, the potential impact on the reactivity transient if a turbine bypass valve(s) fails to modulate during the EC surveillance test was considered. When the EC system is initiated, steam flow is diverted from the main turbine, and the pressure regulating valves controlling (bypass valve or control valve) modulate closed to maintain reactor pressure stable. The failure of the valves to modulate close would result in a pressure reduction transient. In the limiting case, a low pressure main steam isolation and scram will occur. This transient, if assumed concurrent with the subcooling initiation transient, would reduce the severity of the reactivity transient associated with the subcooling initiation transient. Reactor pressure is maintained during the EC surveillance test by normal operation of the reactor pressure regulator. Failure of pressure regulation coincident with the EC system surveillance test, while not assumed, remains bounded by the design basis pressure regulator failure transient analysis.
The EC test prerequisites establish stable reactor pressure control with the bypass valves controlling prior to the initiation of the EC system. This test method eliminates the loss of feedwater transient associated with performance of the EC test when the test is performed with turbine control valves controlling pressure. This test method reduces the subcooling reactivity transient severity associated with the EC system initiation and also demonstrates stable bypass valve modulation and control prior to the reduction in turbine steam flow associated with the EC system operation.
RAI (5)
What was the "rate of reactivity addition" and "amount of reactivity added" during the January 9, 2004 event, in contrast to the January 21, 2004, EC surveillance tests? What was the amount of timefor the EC cold water slug to travel through the core and how was this time determined?
What was the duration and magnitude of the reactivity transient?
Response
The rate of reactivity addition is defined based on the rate of change in inlet subcooling. The rate of change in inlet subcooling during the January 9, 2004 EC surveillance test was approximately the same as the January 21, 2004 EC surveillance test. The January 9, 2004 test was performed for EC loop 12, which returns condensate to recirculation loop 11 suction side, with recirculation loop 15 out of service (i.e., four recirculation loop operation). The resulting recirculation loop discharge temperature was therefore defined based on the mixing of the EC return flow with the Page 4 of 7
recirculation loop 11 flow. The rate of reactivity change is controlled by the rate of change of the recirculation flow exit temperature which is defined based on the recirculation loop flow, EC flow and the EC return valve opening stroke rate. The January 9, 2004 EC test was performed during four-loop operation at 67% power / 60.7% core flow (approximately 15% rated core flow per loop). The January 21, 2004 test was run during five-loop operation at 60.6% power / 77.7%
core flow (approximately 15% rated core flow per loop). Since the EC return flow is fixed based on static head and the opening rate of the EC return valve, and since the recirculation loop flows were essentially the same for both the January 9, 2004 test and the January 21, 2004 test, the rate of change in recirculation discharge temperature is essentially the same for both EC surveillance tests.
The rate of reactivity addition and the amount of reactivity added is estimated based on the rate of change in recirculation loop outlet temperature and mixing of the recirculation loop discharge flow into the lower plenum. The maximum reactivity is bounded by a conservative assumption that the recirculation loop flow will behave as a slug when it enters the lower plenum. The lower plenum of the NMPNS Unit 1 reactor has a flow diffuser that is designed to promote mixing of the recirculation loop discharge flow into the lower plenum. In addition, the recirculation loop exit velocity was approximately 20 ftl/sec during the EC test flow conditions and the average lower plenum velocity in the lower plenum was approximately 2 to 4 f/sec. As a consequence, mixing of the recirculation loop flow would be anticipated.
A supplemental assessment of the transient has been performed using a 3D thermal hydraulic model of the EC system, the reactor recirculation system and reactor vessel lower plenum. This model accounts for mixing of the EC return flow with the recirculation loop flow. The model considers the flow in the lower plenum assuming the lower plenum is divided into 5 equal channels (e.g. 1/5 pie shaped segments of the lower plenum volume). The model assumes no radial mixing for the channel where the recirculation loop discharges into the lower plenum. This model is used to estimate the maximum potential subcooling transient.
This evaluation shows the maximum recirculation loop discharge temperature rapidly decreases to a minimum within 8-10 seconds of EC system initiation (approximately 23 degree reduced recirculation temperature), and then increases rapidly over the next 10 seconds by approximately 12 degrees. Figure 1 shows the maximum core inlet subcooling transient. The model predicts the maximum change in core inlet temperature is 17 degrees Fahrenheit within 10 seconds based on these assumptions.
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Figure 1 Maximum potential Lower PLenum Temeprature at EC discharge location 530-T.-TL9 Reclrc Discharge EC Injection loop 525
-'-ht-TL4s6 lower plenum Recic loop discharge
-TLsl 1 Lower plenum Intermediate 5X0 TLs16'lowerplenumcoreinletelevation c
Row
-'TL~s1' Lower Plenum Below RR eeain-515 E
510 500 -
0 5
10 15 20 25 30 Time (seconds)
For the EC tests performed on January 9, January 21, and January 23, 2004, the measured responses for the APRMs in the quadrants where the EC was initiated resulted in APRM spikes from initial steady state to peak within 5 to 8 seconds. The durations of the "slug" transients were approximately 10 seconds, based on measured response, which were consistent with the calculated slug transient.
The General Electric Nuclear Engineering (GENE) approximation of the neutron flux transient, based on core inlet subcooling transient, is a 1% neutron flux effect for each 1 degree Fahrenheit temperature change. Based on this approximation, the maximum delta T at core inlet is between 17 degrees and 23 degrees and therefore the maximum flux increase would be approximately 17% to 23%. For example, the January 23, 2004 EC test was for EC loop 11 which injects into recirculation loop 15. Recirculation loop 15 injects into the lower plenum at the 288 degree azimuth. This quadrant is monitored by APRMs 12 and 16 and has LPRMs assigned for the quadrant between 270 degrees and 360 degrees. These APRMs showed a maximum flux increase of between 17% and 18% as measured on the high speed Data Acquisition System (DAS) and as recorded by operations monitoring of the APRM channels during the test initiation transient.
In conclusion, the anticipated flux transient based on the maximum potential subcooling is consistent with the measured APRM response, and the duration of the measured flux transient is Page 6 of 7
consistent with the duration of the calculated subcooling transient. The duration of the cold water slug transient also confirms that the transient duration was short such that the localized thermal power remained significantly below the measured APRM flux increase. The overall conclusion is that the thermal margins assessment remains conservative and bounds the maximum potential flux increase of 23% if the maximum recirculation discharge temperature reduction is assumed at the core inlet in a single quadrant.
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