ML112220098
| ML112220098 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/09/2011 |
| From: | Schimmel M A Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-11-079, TAC ME2609, TAC ME2610 | |
| Download: ML112220098 (24) | |
Text
Xcel Energye AUG 0 9 2011 L-PI-11-079 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units I and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 Response to Requests for Additional lnformation (RAI) Associated with Adoption of the Alternative Source Term (AST) Methodolo~v (TAC Nos. ME2609 and ME2610) In a letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 27, 2009 (Agencywide Documents and Management System (ADAMS) Accession No. ML093160583), the Northern States Power Company, a Minnesota corporation doing business as Xcel Energy (hereafter "NSPM"), requested an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP). The proposed amendment requested adoption of the Alternative Source Term (AST) methodology, in addition to TS changes supported by AST design basis accident radiological consequence analyses.
The NRC Staff sent Requests for Additional lnformation (RAI) in a letter dated May 12, 201 1 (ADAMS Accession No. ML103540433) regarding the steam generator tube rupture (SGTR) event radiological consequence analysis. In a letter dated June 22, 201 1 (ADAMS Accession No. MLI 11740145), NSPM provided a response to these RAls. On July 20, 201 1 (ADAMS Accession No.
MLI 12081967), the NRC Staff sent a consolidated request for clarification of the June 22, 201 1 RAI response. The request modifies or supercedes earlier draft requests for clarification sent by electronic mail on June 24, 201 1 (ADAMS Accession No. MLI 11822631) and June 29,201 1 (ADAMS Accession No.
ML111822669). The enclosure to this letter provides the response to the July 20, 201 1 SGTR radiological consequence analysis consolidated request for clarification.
1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone:
651.388.1 121 Document Desk Page 2 NSPM submits this supplement in accordance with the provisions of 10 CFR 50.90. The supplemental information provided in this letter does not impact the conclusions of the Determination of No Significant Hazards Consideration and Environmental Assessment presented in the October 27, 2009 submittal, supplemented by letters dated April 29, 201 0 (ADAMS Accession No. ML101200083), May 25,201 0 (ADAMS Accession No.
ML101460064), June 23, 201 0 (ADAMS Accession No. ML101760017), August 12,201 0 (ADAMS Accession No. ML102300295), December 17,201 0 (ADAMS Accession No. ML103510322), June 22,201 1 (ADAMS Accession No.ML111740145) and July 11,201 1 (ADAMS Accession No. MLI 11 9301 57).
In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of this LAR supplement by transmitting a copy of this letter to the designated State Official.
If there are any questions or if additional information is needed, please contact Mr. Gregory Myers, P.E., at 651-267-7263.
Summary of Commitments This letter contains no new commitments or revisions to existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
AUG 0 9 P@R! Mark A. Schimmel Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc: Administrator, Region Ill, USNRC NRR Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC State of Minnesota ENCLOSURE Response to Requests for Additional lnformation Associated with Adoption of the Alternative Source Term Methodologv - Response to Clarification Questions In a letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 27, 2009 (Agencywide Documents and Management System (ADAMS) Accession No.
ML093160583), the Northern States Power Company, a Minnesota corporation doing business as Xcel Energy (hereafter "NSPM"), requested an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP). The proposed amendment requested adoption of the Alternative Source Term (AST) methodology, in addition to TS changes supported by AST design basis accident radiological consequence analyses.
The NRC Staff sent Requests for Additional lnformation (RAI) in a letter dated May 12, 201 1 (ADAMS Accession No. MLI 03540433) regarding the steam generator tube rupture (SGTR) event radiological consequence analysis.
In a letter dated June 22,201 1 (ADAMS Accession No.
MLI 11740145), NSPM provided a response to these RAls. On July 20,201 1 (ADAMS Accession No.
MLI 12081 967), the NRC Staff sent a consolidated request for clarification of the June 22, 201 1 RAI response.
The request modifies or supersedes earlier draft requests for clarification sent by electronic mail on June 24,201 1 (ADAMS Accession No.
MLI 11 822631) and June 29,201 1 (ADAMS Accession No. MLI I1 822669). The enclosure to this letter provides the response to the July 20, 201 1 SGTR radiological consequence analysis consolidated request for clarification.
Question: (I) Table 2 lists the safety-related (SR), non-SR (NSR) and augmented quality (AQ) systems, components and instruments (SCls) available for steam generator tube rupture (SG TR) mitigation. Please provide a definition of the AQ SCls, identify the NSR and AQ SCls in Table 2 that are credited in the margin-to-overfill (MTO) or mass release analyses, discuss the functions of each identified SCls, address the acceptability of each of the NSR and AQ SCls for consequences mitigation assumed in the analyses for supporting the licensing applications, and demonstrate that the use of NSR or AQ SCls meets the intent of use of the SR SCls for design-basis analysis. NSPM Response Consistent with the NRC Safety Evaluation Report (SER) for WCAP-10698, the list of systems, components, and instruments (SCI) identified in Table 2 of Reference 1 was developed based on review of the Emergency Operating Page 1 of 22 Enclosure Response to Clarification Questions NSPM Procedures (EOPs) for mitigating a SGTR. The list of SCI in Table 2 of Reference 1 contains more SCI than are credited in the MTO analysis or mass release analysis. In response to this request, the list of SCI in Table 2 of Reference 1 was reviewed to determine the SCI that are credited in the analysis and classified as non-safety related (NSR) or augmented quality (AQ).
The SCI credited in the analysis that are classified NSR or AQ are identified in the following two tables; the first table is for systems and components and the second table is specifically for instrumentation.
For each component or instrument the tables provide the following information:
e Function of each component or instrumentation as used in mitigation of a Steam Generator Tube Rupture. e Although consideration of a limiting single failure is not in the PlNGP SGTR licensing basis, a listing of backup components or instrumentation that are available to meet the above function is provided from a defense- in-depth perspective. Description of testing and maintenance performed to ensure the reliability of the component or instrument. Additional details regarding power supply and other pertinent considerations.
As used in the following two tables, augmented quality (AQ) describes a subset of non-safety related items for which one or more of the governing NSPM Quality Assurance Topical Report (QATR) requirements apply. This includes items for which specific licensing commitments or additional controls over quality are deemed necessary by NSPM management, but are not explicitly defined in the QATR. As shown in Table 2, the pressurizer water level indications, RCS pressure transmitters and indicators are AQ.
These devices are qualified in accordance with Regulatory Guide (RG) 1.89, Rev. 0, "Qualification of Class 1 E Equipment". The following tables identify and discuss the functions of the NSR and AQ SCIs credited in the SGTR analyses, address the acceptability of these SCls for consequence mitigation assumed in the analyses and demonstrate that the use of these SCls meets the intent of use of the SR SCls for design-basis analysis. Page 2 of 22 Enclosure Response to Clarification Questions NSPM Table 1 Non-Safety Related or Augmented Quality Systems and Components Utilized for Steam Generator Tube Rupture Mitigation Page 3 of 22 Equipment1 Name SG PORV Control Board Controllers (PORV) Board Controllers Safety Classification NSR Backup Available No Function The control board controllers are used to control PORV position.
The functions of the SG PORVs (steam generator power operated
relief valves) are described in the response to Reference 1, RAI Question A.2.d. Reliability Features The controllers are reliable devices. Operability and testing requirements for the SG PORVs are specified per Technical Specification (TS) 3.7.4. As part of this testing the valves are operated monthly from the control room using the hand controller and verifying valve operation using the position indication lights on the control board. As identified in Technical Specification Surveillance Requirement (SR) 3.7.4.1, the SG PORV is tested per the ASME Section XI lnservice Testing Program.
Remarks The SG PORV control board controllers are powered from 120 Volts Alternating Current WAC) instrumentation buses, which are safety related (battery backed).
Thus, power will be available to the controllers following a loss of offsite power (LOOP).
Enclosure Response to Clarification Questions Equipment1 Component Name SG PORV Control Board lndicating Lights Safety Classification NSR Backup Available The emergency response computer system (ERCS) provides backup indications (PORV position, downstream flow rate and downstream temperature) that can be used to verify valve position.
The ERCS is supplied by an Uninterruptible Power Supply (UPS) and redundant power sources to network switches and remote multiplexing units. Function The control board indicating lights provide valve position indication. The functions of the SG PORVs are described in the response to Reference 1, RAI Question A.2.d. Reliability Features SG PORV Control Board lndicating Lights are tested with the Control Board Controllers as described above. NSPM Remarks The SG PORV control board indicating lights are powered from 125 volts direct current (VDC) buses, which are powered from safety-related batteries.
Thus, power will be available to the control board indicating lights following a LOOP. Page 4 of 22 Enclosure Response to Clarification Questions NSPM Page 5 of 22 Equipment1
'Omponent Name Pressurizer PORV Control Switches Chemical and Volume Control System (CVCS)
Letdown System Safety Classification NSR NSR Backup Available There are two PORVs. Either PORV can be used to perform this function. The ERCS provides backup indications (PORV downstream temperature) that can be used to verify valve position and indicate possible switch malfunction.
The ERCS is supplied by an Uninterruptible Power Supply (UPS) and redundant power sources to network switches and remote multiplexing units. If normal letdown cannot be established, excess letdown is established.
All equipment used to place excess letdown in service is supplied by safety related power. Function The control switches operate the PORV. A pressurizer PORV is used to depressurize the reactor coolant system to stop break flow through the ruptured tube. Balance letdown with charging flow to maintain equilibrium conditions across the break. Reliability Features Operability and testing requirements for the pressurizer PORVs are specified per Technical Specification 3.4.1 1. As part of testing the valves are operated from the control room every 24 months using the control switch and valve operation is verified using the position indication lights on the control board. The letdown system is in operation when the unit is operating at power.
Thus, reliability of the letdown system and components is demonstrated through normal operation of the system. Remarks The pressurizer PORV control board control switches are powered from 125 VDC buses, which are powered from safety related batteries.
Thus, power will be available to the control switches following a LOOP. All equipment used to place letdown in service is supplied by safety related power. The air supply to system control valves is from the Instrument Air System.
Enclosure Response to Clarification Questions NSPM Compressors Equipment!
Component Name Instrument Air air compressors are provided as described in the response to Reference 1, RAI A.2.d. In addition to the air supply from the lnstrument Air compressors, a Seismic Category I passive air accumulator is provided inside of containment as a back- up air supply for the pressurizer PORVs. Safety Classification NSR 1 Function 1 Reliability Features I Remarks I Backup Available Redundant instrument Provide air supply for operation of SG PORVs, pressurizer PORVs, and letdown system valves. The instrument air compressors are included in the Maintenance Rule Program. The instrument air compressors and other components in the system are part of a periodic maintenance and testing program to ensure that the system will function when required. This includes preventative maintenance, inspections, setpoint calibrations, and testing. The reliability of the lnstrument Air System is discussed in the response to Reference 1, RAI A.2.d. Page 6 of 22 Enclosure Response to Clarification Questions Table 2 Non-Safety Related or Augmented Quality Instruments Utilized for Steam Generator Tube Rupture Mitigation Safety SG Water Level Indication - Narrow Range E-dLL.- NSR Equipment.
Component Name NSPM Function Three separate indication channels for each steam generator are displayed in the control room. Reliability Features classification (Process LOOP) The ERCS provides backup indications for SG narrow range water level. The ERCS is supplied by an UPS and redundant power sources to network switches and remote multiplexing units. Identify Ruptured SG. aaieLy Classification (Indicator)
Operability and testing requirements are specified per Technical Specification 3.3.1. Maintenance and calibration of the control room indication is accomplished every 24 months as part of the testing performed to meet the TS surveillance requirement. Backup Available Remarks The instrument loops from the sensing line through the transmitter are safety related.
The indication is on the non- safety related side of the loop.
The indicator power supply is safety related and will be available during a LOOP. Page 7 of 22
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Enclosure Response to Clarification Questions NSPM Pressure Indication Equipment!
Component Name Two separate indication channels per unit are available in the control room.
The ERCS provides backup indications for reactor coolant system (RCS) pressure. The ERCS is supplied by an UPS and redundant power sources to network switches and remote multiplexing units.
Safety Classification (Process LOOP) Function Provides indication to the operators that supports securing depressurization and termination of SI. Safety Classification Reliability Features Backup Available Operability and testing requirements are specified per Technical Specification 3.3.1. Maintenance and calibration of the Control Room indication is accomplished every 18 months as part of the testing performed to meet the TS surveillance requirement.
Remarks The RCS pressure indicators and transmitters are augmented quality. The indicator power supply is safety related and will be available during a LOOP. SG Pressure lndication NSR Three separate indication channels per steam generator are available in the control room. The ERCS provides backup indications for SG pressure. The ERCS is supplied by an UPS and Provides indication to the operators that supports determination of target cooldown temperature.
Operability and testing requirements are specified per Technical Specification 3.3.2. Maintenance and calibration of the control room indication is accomplished every 24 months as part of the testing performed to meet the TS The instrument loops from the sensing line through transmitter are safety related. The indicators are non- safety related.
The indicator power supply is safety related and will be available during a LOOP. The SG pressure indicators are Page 9 of 22 Enclosure Response to Clarification Questions NSPM I Air Ejector I Radiation Equipment]
Component Name Monitor bt Generator Blowdown Liquid Radiation Monitor Main Steam Line Radiation Monitor Safety Classification (Process LOOP) NSR NSR Safety Classification (Indicator)
NSR redundant power sources to network switches and remote multiplexing units. Backup Available In the event that the operators do not transition to the SGTR EOP based on the radiation monitor signal, the operators would transition to the SGTR EOP based on SG narrow range level indication increasing in an uncontrolled manner. Although these monitors are not credited in the SGTR analyses, they provide early indication that a tube Function The radiation monitors provide an early indication of a SGTR. In addition, the main steam line radiation monitors provide indication of which steam generator is ruptured.
surveillance requirement. Reliability Features The radiation monitors are maintained and tested in accordance with the Offsite Dose Calculation Manual. This requires periodic calibration and testing (performed every 18 months) to confirm that the radiation monitors are functioning properly.
Remarks credited indications for meeting RG 1.97, Revision 2. Thus, these indications also have quality requirements imposed by RG 1.97. The air ejector radiation monitors and the main steam line radiation monitors are credited indications for meeting RG 1.97, Revision 2. These indicators are qualified in accordance with RG 1.89, "Qualification of Class 1 E Equipment".
The power supply to the condenser air ejector and steam generator blowdown radiation monitors is safety related and will be available during a Page 10 of 22 Enclosure Response to Clarification Questions NSPM Page 11 of 22 Remarks LOOP. The power supply to the main steam line radiation monitor is non-safety related, but is backed by a non-safety related diesel generator and expected to be available during a LOOP. Equipmentl Component Name Reliability Features Safety Classification (Process LOOP) Backup Available rupture has occurred.
The ERCS provides backup indications for these radiation monitors. The ERCS is supplied an UPS and redundant power sources to network switches and remote multiplexing units.
Safety Classification (Indicator)
Function Enclosure Response to Clarification Questions NSPM Question:
(2) The operator action delay times for reactor coolant system (RCS) depressurization are 4 minutes used in the MTO analysis and 7 minutes in the mass release analysis listed in Tables 5, and 6, respectively.
Provide bases for the different operator action times used for RCS depressurization, while the operator action times for RCS cooldown initiation (1 9 minutes following reactor trip), safety injection (SI) termination (2 minutes) and charging flow termination (15 minutes) remain the same in the MTO and mass releases analysis.
Discuss the acceptance criteria for the results of the simulator exercises in support of the operator action times (4, 2, 19, 15 minutes above) credited in the analysis. NSPM Response The supplemental steam release analysis used a conservative time of 7 minutes to model the operator action time delay between the end of cooldown and the start of depressurization, whereas, the margin-to-overfill (MTO) analysis used a more representative time of 4 minutes for this operator action time delay. The times were based on SGTR simulator validation exercises. The limiting MTO time of 4 minutes will be identified as an operator action time requirement and will be incorporated into the NSPM process described below. A NSPM fleet administrative procedure establishes the process to capture analysis-credited operator actions, such as SGTR analysis operator actions. As part of this process, the limiting time critical operator actions used in the margin-to-overfill analysis listed by Table 5 of Reference 1 will be validated and documented in plant procedures. Time critical operator action requirements must be met by all individuals and crews during validation. Whenever a time critical operator action requirement is not achieved during a simulator training scenario, the discrepancy is placed into the PlNGP corrective action program. The cause of the discrepancy will be determined and corrected. This process ensures that the SGTR time critical operator action times listed in Table 5 of Reference 1 will be met. Page 12 of 22 Enclosure Response to Clarification Questions NSPM Question:
(3) Figures 11 and 16 show the RCS and steam generator (SG) pressures, and steam releases for the mass release analysis.
I. Figure 11 indicates that at about 2850 seconds, and 3100 seconds, there are sudden decreases in the intact SG pressure and RCS pressure, respectively.
Explain the thermal hydraulic phenomena, mitigating systems, or operator actions that contribute to the decreases. NSPM Response The decrease in pressure in the intact steam generator at approximately 2850 seconds corresponds to the time when the intact steam generator PORV is re-opened to maintain the RCS target temperature. Following completion of the RCS cooldown the intact SG PORV is opened, as needed, to maintain the RCS cooldown target temperature achieved during the rapid cooldown in order to maintain the required subcooling margin. This is consistent with the Prairie Island EOPs. The change in RCS pressure reduction rate at approximately 31 00 seconds results from voiding in the upper head. It is noted that the Prairie Island-specific EOPs caution the operators that upper head voiding could occur during the SGTR response, if the reactor coolant pumps are not running, as is in this case. Question 3 (cont.): 2. Figure 16 shows that for the ruptured SG, the steam releases suddenly increase and decrease between 100 to 200 seconds into the transient.
At 800 seconds, the releases suddenly increase and then decrease gradually, following a deep gradual decrease at 1300 second until 1500 seconds. After 1500 seconds, the releases become a small, constant rate.
Explain the changes identified above for the steam releases through the ruptured SG. For the intact SG, the first significant steam releases occur between 1200 - 1900 seconds. This is due to the operator action that opens the SG power operated relief valve (PORV) for RCS cooldown. Explain the causes for the second significant release to occur after 2800 seconds until 4000 seconds when the presented results end.
Page 13 of 22 Enclosure Response to Clarification Questions NSPM NSPM Response The initial spike in SG steam releases corresponds to the steam release rate through the intact and ruptured SG PORVs following reactor trip and LOOP, as the pressure in the SGs increases rapidly. This release is reduced as the reactor power rapidly decreases to decay heat levels, safety injection (SI) flow cools the RCS and the AFW flow cools the SG secondary. The increase in ruptured steam generator steam releases at approximately 800 seconds corresponds to the time when the AFW flow to the ruptured SG is terminated at 738 seconds as shown in Table 7 of Reference
- 1. The energy that was being absorbed by the inflow of AFW into the ruptured SG is then released as steam once AFW flow is terminated. The decrease of the steam flow rate from the ruptured SG between approximately 1300 and 1500 seconds corresponds to the cooldown using the intact SG. As the RCS is cooled via the intact steam generator, energy removal that was taking place via steaming from the ruptured SG is terminated.
After this time, the steam release from the ruptured SG corresponds to the displacement of steam from the ruptured SG by the incoming break flow. The intact SG steam release after approximately 2800 seconds is due to the continued RCS cooling being performed with the intact SG. Question:
(4) Observing the time-temperature plots (Figures 5 and 6) provided in your RAI response dated June 22, 201 1, the transient associated with the SGTR MTO analysis shows what could be a pressurized thermal shock (PTS) event with respect to the reactor pressure vessel. Please explain, based on your understanding of the technical basis for the NRC's PTS rule, 10 CFR 50.61, and the condition (RTPTS values) of the limiting materials in the Prairie island, Unit I and 2 reactor pressure vessels, why you understand these reactor pressure vessels to be adequately protected from failure due to the transients shown in your RAI response for the operating lifetime of the facility.
NSPM Response The pressurized thermal shock (PTS) evaluation provides a means for assessing the susceptibility of the reactor vessel beltline materials to PTS events to assure that adequate fracture toughness is provided for supporting reactor operation.
Page 14 of 22 Enclosure Response to Clarification Questions NSPM The PTS rule, 10 CFR 50.61, was published in the Federal Register, December 19, 1995, with an effective date of January 18, 1996. This amendment made the procedure for calculating RTpTs values consistent with the methods given in Regulatory Guide I .99, Revision 2. The PTS rule establishes the following requirements for all domestic, operating PWRs: For each PWR that has had an operating license issued, the licensee will have projected values of RTpTs accepted by the NRC, for each reactor vessel beltline material for the end of life fluence of the material. The assessment of RTpTs must use the calculation procedures given in the PTS Rule and must specify the bases for the projected value of RTpTs for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material. This assessment must be updated whenever there is a significant change in projected values of RTPTS, or upon the request for a change in the expiration date for operation of the facility. Changes to RTpTs values are significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any license renewal term, if applicable for the plant. The RTpTs screening criteria values for the beltline region are: o 270°F for plates, forgings, and axial weld materials o 300°F for circumferential weld materials An evaluation of the impact of the EPU on PTS was performed for PlNGP Units 1 and 2. RTpTs calculations were performed for all the reactor vessel materials of the PlNGP reactor vessels under EPU conditions at 54 EFPY (end of life) using the rules from 10 CFR Part 50.61. The limiting circumferential weld material for the PlNGP Unit 1 reactor vessel is the Nozzle Shell to lntermediate Shell Circumferential Weld - Seam W2, with an RTpTs value of 160°F. The limiting forging material for the PlNGP Unit 1 reactor vessel is the lntermediate Shell Forging C, with an RTpTs value of 127°F. The limiting circumferential weld material for the PlNGP Unit 2 reactor vessel is the Upper Shell to lntermediate Shell Circumferential Weld - Seam W2, with an RTpTs value of 139°F. The limiting forging material for the PlNGP Unit 2 reactor vessel is the Lower Shell Forging D, with an RTpTs value of 116°F. Page 15 of 22 Enclosure Response to Clarification Questions NSPM Note that the RTpTs values detailed above were calculated considering EPU conditions at PINGP. However, these RTpTs values are conservative when compared to the current operating conditions at PINGP as documented in the PINGP License Renewal Application, Section 4.2.3, Tables 4.2-4 and 4.2-5 (Reference 2). Nevertheless, taking into consideration the uprated (conservative) neutron fluence values, the PINGP Units 1 and 2 reactor vessel beltline material RTpTs values remain well below the screening criteria values in 10 CFR 50.61 through 54 EFPY (end of life). Therefore, PTS is not a concern for the PINGP Units 1 and 2 reactor vessels. The technical basis for the PTS Rule is contained in NRC Letter SECY 465 dated November 23, 1982. This letter forms the technical basis for the PTS Rule, as stated above, and Steam Generator Tube Rupture (SGTR) was an analyzed transient. SGTR is discussed in Section 6.2.3 of SECY-82-465 and the parameters of the transient analyzed are defined in Table 6.5. Table 6.5 shows that the final reactor coolant temperature for this transient is 200°F at the reactor vessel wall.
Per Figure 6 of Enclosure 2 to Reference 1, the minimum temperature reached for the PI SGTR transient is approximately 250°F. Therefore, PINGP remains bounded by the technical basis for the PTS Rule, 10 CFR 50.61, as defined in SECY-82-465. The conclusions of Section 10 in SECY-82-465 confirm the screening criteria limits in 10 CFR 50.61. Therefore, since the SGTR transient was analyzed as part of the PTS technical basis development, and PINGP is under the limits of the rule, the Units 1 and 2 reactor pressure vessels are adequately protected from failure due to this transient.
Question:
(5) Table 1 indicates that the reactor trip occurs at 49 seconds following the event initiation.
Specify the signal that actuates the reactor trip, discuss the trip point assumed in the analysis and show that it is within the technical specifics tion (TS) value with inclusion of instrumentation uncertainties. NSPM Response The reactor trip denoted in Table 1 of Reference 1 is generated in response to the overtemperature delta T (OTDT) trip signal. The OTDT setpoint constants from the Prairie Island core operating limits report were used as shown in Table 3 below. Consistent with the Westinghouse implementation of the overall conservative methodology approved in WCAP-I 0698-P-A, the nominal values used do not consider instrument Page 16 of 22 Enclosure Response to Clarification Questions NSPM uncertainties.
As discussed below, the Prairie Island SGTR MTO analysis is not highly sensitive to the reactor trip time and margin to overfill exists regardless of reactor trip time. For the Prairie Island SGTR analysis, inclusion of instrument uncertainties on the OTDT trip signal setpoint constants would not have a significant impact on the calculated MTO.
Prior to the reactor trip, the level in the ruptured SG is maintained as a result of automatic feedwater control. For an SGTR accident, an earlier reactor trip would correlate to a higher RCS pressure at the time of trip and a slightly higher break flow rate into the ruptured SG immediately following the trip. From this, delayed reactor trip is less limiting since it would have the opposite effect on the post-trip break flow rate.
As shown by Table 5 of Reference 1, AFW flow into the ruptured SG is isolated based on the level in the ruptured SG. As such, the water volume in the ruptured SG at the time that AFW flow is terminated will be the same regardless of the exact reactor trip time and break flow rate immediately following trip. Shortly after reactor trip, SI flow is initiated (1 19 seconds from Table 1 of Reference 1). After initiation of SI, the RCS pressure is being controlled by the SI flow (as can be seen in Figure 1 of Reference
- 1) and the RCS pressure at the time of trip is no longer important.
From this, the reactor trip time would have a minimal impact on the calculated MTO. It is also noted from Table 5 of Reference 1, that the Prairie Island SGTR analysis models the initiation of the RCS cooldown 19 minutes from the time of reactor trip. As a result, operator response to the event would not be impacted by an earlier reactor trip time. An earlier reactor trip would result in earlier cooldown initiation leading to earlier break flow termination, which would result in a benefit when compared to a calculation with a later reactor trip time. However, the time from reactor trip until break flow termination, during which the secondary side water volume in the ruptured SG is increasing, would be approximately the same between the two calculations. Based on these discussions, the slightly higher RCS pressure that would result from an earlier reactor trip would not significantly impact the ruptured SG water volume at the time of break flow termination. Furthermore, the rate at which the RCS depressurizes would be similar between calculations and would differ only in the time of reactor trip. Thus, for a calculation with an earlier reactor trip time, the break flow rate would be higher immediately following the reactor trip compared to that calculated for the reported case at the time of reactor trip, but would be expected to be similar at the time the reported case generates the reactor trip (49 seconds). Page 17 of 22 Enclosure Response to Clarification Questions NSPM In addition to the above, the bounding impact of an earlier reactor trip can be estimated from the integrated break flow into the ruptured SG during the period until the time of reactor trip in the reported case. This method conservatively overestimates the impact on MTO of an earlier reactor trip by assuming reactor trip concurrent with the event initiation and adding the pre-trip integrated break flow from the reported case to the final calculated ruptured SG water volume, effectively assuming that all primary to secondary mass transfer takes place after reactor trip. The volume that would be taken up by the amount of integrated break flow into the ruptured SG during the first 49 seconds is lower than the reported margin to overfill. Thus, margin to overfill would be maintained for a calculation with an earlier reactor trip time even with this conservative adjustment method.
Table 3: OTDT Trip Setpoint Constants Page 18 of 22 Setpoint K1 K2 K3 T' P' Input Value 1.17 0.014 /"F 0.001 /psi 560°F 2235 psig Enclosure Response to Clarification Questions NSPM Question:
(6) The first paragraph on page 47 discusses the basis to support the conclusion that the original hand-calculated releases are the bounding result and remain valid. Please provide data of the dose releases through the intact and rupture SGs for the pre-trip and post- trip, and other applicable conditions to support the stated conclusion. NSPM Response: An explicit dose analysis was not performed using the supplemental thermal hydraulic analysis. However, it is demonstrated through the use of comparative analysis of key input that the original hand-calculation methodology results in the bounding consequences. The inputs for the SGTR dose consequences impacted by the supplemental thermal hydraulic analysis are the mass released and the duration of the release. Dose is approximately proportional to both of these inputs.
Table I1 in the August 12,201 0 AST Supplemental Response (ADAMS Accession No.
MLI 02300295) (Reference
- 3) provided the control room (CR) doses due to Primary to Secondary (P-T-S) iodine release. These results are presented below with the percentage of the total effective dose equivalent (TEDE) dose.
Table 4: Steam Generator Tube Rupture Control Room Doses This table demonstrates that the dose to the CR is controlled by the P-T-S flow. This flow is sum of flashed break flow, unflashed break flow, and Technical Specification leakage. Considering the magnitude of the break flow, Technical Specification leakage contributes an insignificant amount. P-T-S Iodine Release Total CR dose Of this flow, the flashed break flow contributes nearly the entire dose due to iodine. This is due to the fact that the iodine in the flashed break flow is not subject to retention by either scrubbing or partitioning, whereas the unflashed break flow is reduced by a factor of 100. Based on the comparison of the hand calculation break flows it can be shown that over Page 19 of 22 CR TEDE Dose (Rem)
Pre-Accident Iodine Spike 4.14 4.67 % of Total CR Dose 88% - CR TEDE Dose (Rem)
Concurrent Iodine Spike 3.33 3.45 % of Total CR Dose 96% -
Enclosure Response to Clarification Questions NSPM 90% of the dose due to iodine releases is from the post-trip flashed break flow. This is estimated by dividing the post-trip flashed break flow by the sum of the reduced post-trip unflashed break flow and the flashed post-trip break flow (i.e., iodine to the ruptured SG from flashed break flow divided by total iodine released to the ruptured SG). From Table 8 of Reference 1 a comparison of the total post-trip break flows, post-trip flashed break flows, and unflashed break flows are as follows: Table 5: Steam Generator Tube Rupture Mass Releases "Post-trip unflashed break flow is the Total post-trip break flow minus the post-trip flashed break flow. , **Reduced unflashed break flow is the unflashed break flow reduced by a factor of 100 to account for SG partitioning. ***For conservatism, the total break flow modeled included the pre-trip flashed break flow. Total Post-Trip Break Flow (I bm)*** Post-Trip Flashed Break Flow (Ibm) Post-Trip Unflashed Break Flow (Ibm)* Reduced Post-Tri p Unflashed Break Flow (Ibm)** This comparison demonstratesthat total flashed break flow determined by the supplemental thermal hydraulic analysis represents a reduction in release to the environment by more than a factor of 4 when compared to the flashed break flow modeled in the licensing basis analysis. This flow is the same for both the pre-accident iodine spike and concurrent iodine spike. Therefore, it is reasonable to estimate that the primary-to-secondary iodine release dose contribution using the flashed break flow determined by the supplemental thermal hydraulic analysis would result in a dose reduction by a factor of 4 for both the pre-accident iodine spike and concurrent iodine spike cases.
Hand Calculation 140,000 15,050 124,950 1249.50 Page 20 of 22 Supplemental Thermal Hydraulic 159,300 3,100 156,200 1562 Ratio .88 4.85 .80 .80 Enclosure Response to Clarification Questions NSPM The unflashed break flow for the supplemental thermal hydraulic analysis is slightly higher, which will result in a higher dose contribution from this source. However, since this flow is subject to significant activity retention in the steam generator, the increase in the release from the unflashed break flow is still bound by the flashed break flow dose. The duration of the break flow for the hand calculation method is 30 min as documented in Reference
- 1. The duration of the break flow for the supplemental thermal hydraulic analysis is at about 64 min post accident as documented in Table 7 of Reference
- 1. Table 7 of Reference 1 also documents that the flashing break flow stops at approximately 24 min post accident. The significance of the termination of flashing break flow is that all of the iodine in the break flow is assumed to be mixed with the bulk of the liquid in the steam generator and subject to scrubbing and partitioning, resulting in a reduction in the amount of iodine released. However, ignoring this consideration, it is conservatively assumed that the extended time of the break flow results in a dose increase of a factor of 2 for both the pre-accident and concurrent iodine spike scenarios.
The net result is that the supplemental thermal hydraulic analysis would result in dose consequence about a factor of 2 lower than the current licensing basis hand calculation method:
i.e., factor of 4 decrease due to lower flashed break flow with an assumed factor of 2 increase due to break flow time duration.
Offsite doses'would be reduced in a similar manner since dose consequences are calculated with the same basic dose equation but factoring in the applicable physical modeling factors, such at atmospheric dispersion factors. Therefore, the doses previously reported in the AST LAR remain bounding. Page 21 of 22 Enclosure Response to Clarification Questions NSPM References
- 1. NPSM Letter to US NRC, "Response to Additional Information (RAI) Associated with Adoption of the Alternative Source Term (AST) Methodology (TAC NOS. ME2609 and ME2610)," dated June 22,201 1 (ADAMS Accession No. MLI 1 1740145).
- 2. NMC Letter to US NRC, "Application for Renewed Operating Licenses," dated April 1 1,2008 (ADAMS Accession No.
- 3. NSPM Letter to US NRC, "Response to Requests for Additional Information Re: License Amendment Request to Adopt the Alternative Source Term Methodology (TAC NOS. ME2609 and ME2610)," dated August 12,2010, (ADAMS Accession No.
ML102300295). Page 22 of 22