ML13331A571
| ML13331A571 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/25/2011 |
| From: | Ameren Missouri, Union Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13331A633 | List: |
| References | |
| ULNRC-06038 | |
| Download: ML13331A571 (14) | |
Text
Enclosure to ULNRC-06038 UNION ELECTRIC COMPANY (dba AMEREN MISSOURI)
CALLAWAY PLANT DOCKET NO. 50-483 10 CFR 50.59
SUMMARY
REPORT Report Period: November 25, 2011 to May 28, 2013.
Enclosure to ULNRC-06038 EXECUTIVE
SUMMARY
In accordance with 10 CFR 50.59(d)(2), a summary report has been prepared which provides summaries of the 10 CFR 50.59 evaluations of changes, tests, and experiments approved and implemented for activities at Callaway Plant.
This report covers all 10 CFR 50.59 evaluations for changes that were implemented from November 25, 2011 to May 28, 2013. During this period there were seven changes implemented that required a 10 CFR 50.59 evaluation. For each of these changes it was determined that prior NRC approval was not required, in accordance with 10 CFR 50.59(c)(1),
and therefore, summaries of those seven 10 CFR 50.59 evaluations are hereby provided.
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Enclosure to ULNRC-06038 10 CFR 50.59 EVALUATIONS Evaluation Number: Activity:
11-01 Use of GOTHIC code for High Energy Line Break outside of Containment 11-02 Modification to change the LOCA Sequencers (for sequencing loads onto the safety buses following a load shed) to delay opening valve EFHV0037 and valve EFHV0038 11-04 Licensing Document Change Notification to reduce the minimum Auxiliary Feedwater (AFW) system flow assumed for the Feedwater Line Break (FLB) event described in the FSAR 12-01 Compensatory actions to support a Prompt Operability Determination (POD) performed to address a non-conforming condition identified for the Ultimate Heat Sink (UHS) retention pond 12-03 Installation of New Westinghouse Reactor Coolant Pump (RCP) Shutdown Seals 13-01 Main Feedwater Pump Turbine Control System Replacement 13-02 Solid State Protection System (SSPS) Printed Circuit Board Replacement LOCA = Loss of Coolant Accident IHA = Integrated Head Assembly RRVCH = Replacement Reactor Vessel Closure Head Page 2 of 13
Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 11-01: Use of GOTHIC Code for High Energy Line Break Outside of Containment Activity
Description:
This evaluation addresses a change in analysis methodology used to determine the effects of High Energy Line Breaks (HELBs) and Moderate Energy Line Breaks (MELBs) outside of containment. It is desired to use the industry standard for thermal hydraulic analysis, which is the GOTHIC computer code. GOTHIC is already used for containment pressure/temperature and Main Steam Line Break (MSLB) analysis at Callaway Plant.
The original pressure and temperature profiles for HELBs outside of containment were established under Calculation M-YY-49 and Addenda using Bechtel Topical Report BN-TOP-
- 04. Bechtel used the COPATTA code and FLUD6 for the original analysis. FLUD 6 was later replaced by PCFLUD Version 3.7 for Ameren Missouri calculation M-YY-49. Callaway Plant now wishes to use the GOTHIC code for future updates of HELB and MELB analyses outside of containment. The methods are described in FSAR Sections 3.6.1.2, 3.6.1.3, 3.6.2.5, 6.2.1.2, and Appendix 3B.
The River Bend GOTHIC HELB evaluation models were reviewed and approved by the NRC (May 20, 2004, Docket No. 50-458), but the Safety Evaluation had limitations that the model use homogenous equilibrium flow through vent paths and 100 percent entrainment except where it is more conservative not to employ these assumptions. The NRC also reviewed and approved (March 21, 2008, Docket No. 50-482) Wolf Creeks use of GOTHIC for the MSLB outside of containment.
NRC Information Notice 2011-17 references a GOTHIC model that was not sufficiently qualified through benchmarking against test or plant data to demonstrate applicability to the type of analysis. The following addresses each of the issues raised:
- 1. Benchmarking: Callaway commissioned Numerical Applications, Inc. (NAI) to benchmark the GOTHIC code against Revision 1 of Calculation M-YY-49 (based on PCFLUD models) as part of the validation/verification of the GOTHIC model. The peak pressure calculated by GOTHIC was 15.25 psia compared to 15.2 psia. The results were essentially the same or slightly conservative compared to the original analysis.
- 2. Initial Conditions and Vent Paths: The initial conditions and assumptions are the same for both the original and the GOTHIC model; however, the GOTHIC model includes areas of the plant that were not evaluated in the original model.
- 3. Heat Transfer Models: The GOTHIC model uses the Drop Liquid Conversion model (DLM) to model condensation. This is appropriate and acceptable as all of the cases presented in the benchmark performed by NAI are for steam line breaks (single-phase flow) and do not include two-phase flow from the break. The limitations on using the DLM method discussed in the River bend Safety Evaluation apply to breaks where a portion of the break is sub-cooled liquid (two-phase flow).
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Enclosure to ULNRC-06038
- 4. Mass and Energy Release: Mass and energy data for the 8-inch Auxiliary Steam line breaks are taken from the original PCFLUD model. Mass and energy data for the smaller breaks were derived from the 8-inch break flow.
Summary of Evaluation:
The GOTHIC models used for Callaway meet the requirements and restrictions set forth in the above-described approvals. Further, the results have been shown to be essentially the same as those obtained from the previous methodology (PCFLUD). Therefore, the proposed use of GOTHIC for the intended applications does not constitute a departure from a method of evaluation, and consequently, this change may be implemented without (additional) prior approval by the NRC.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 11-02: Calculation ZZ-525, Rev. 1 Add. 4 in Support of Plant Modification MP 10-0004, Revision of Sequencer Operation for Valves EFHV0037 and EFHV0038 Activity
Description:
The activity/change evaluated is Modification Package MP 10-0004 which will change the LOCA Sequencers (for sequencing loads onto the safety buses following a load shed) to delay opening valve EFHV0037 by 20 seconds and valve EFHV0038 by 25 seconds. This change will minimize water hammer effects on the Containment Coolers. The additional delay in providing cooling water to the coolers will have a slight effect on heat removal from Containment during a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB).
Summary of Evaluation:
Ameren Missouri calculation ZZ-525, Revision 1 Addendum 4 has been performed to re-analyze post-accident containment temperatures and pressures (for LOCA/MSLB) assuming that EFHV0037 (A Train Essential Service Water supply to the Ultimate Heat Sink) and EFHV0038 (B Train Essential Service Water supply to the Ultimate Heat Sink) are opened 25 seconds later than in the current analysis. Although the valves have different delay times, 25 seconds was used in the re-analysis for conservatism.
Results show that the limiting peak temperature and pressure are unaffected by the delay. The MSLB Double-Ended Rupture provides the limiting peak Containment temperature and the Double-Ended Hot Leg Shear provides the limiting peak Containment pressure. These peaks occur prior to cooling water flow initiation to the Containment Coolers in the current analysis.
Therefore, the current and updated analyses are identical up to that point. The main effect of the delay is that it takes slightly longer to remove heat from the containment atmosphere.
Temperature and pressure in the updated analysis trend slightly higher during the cool-down phase of the accidents with a difference of approximately 0.5 psig and 1.5 °F between the two sets of curves for the current and updated analysis. Although containment temperature and pressure are slightly higher during the cool-down phase, the potentially affected equipment remains bounded by the current EQ envelope.
Since there are no changes to the calculated peak pressures and temperatures, and the post-peak pressures and temperatures are bounded by the current EQ envelope, the change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety, does not result in more than a minimal increase in the consequences of an accident previously evaluated, and does not create the possibility of a malfunction with a different result. This change may thus be implemented without NRC approval.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 11-04: Licensing Document Change Notice 09-0010, Incorporation of Westinghouse AABD Module 6.0, Rev. 5 in FSAR for Feedwater Line Break (FBL)
Activity
Description:
The proposed activity is a reduction in the minimum Auxiliary Feedwater (AFW) system flow assumed for the Feedwater Line Break (FLB) event described in the FSAR Section 15.2.8.
Specifically, the change is a reduction in AFW flow from 664.4 gpm to three intact steam generators to 543.2 gpm to three intact steam generators.
Summary of Evaluation:
The reduction in AFW flow from 664.4 gpm to three intact steam generators to 543.2 gpm to three intact steam generators for the FLB event does not require prior NRC approval. The proposed change lowers an analysis assumption (in the non-conservative direction) for an input parameter for the FLB analysis but is not a change to the analysis methodology itself. Since the method of FLB analysis and the FLB event acceptance criteria remain unchanged, the proposed change does not result in a departure from a method of evaluation used in establishing the design bases or in the safety analysis.
The propose change does not impact any key input assumptions that determine the radiological consequences for postulated accidents, nor does it impact the calculation methodology in any radiological analysis. The method used to establish the design basis for mitigation of the most limiting postulated accident, i.e. the FLB event, is not changed. Furthermore, this analytical change involved no physical changes to the facility that could introduce a new malfunction or new accident not previously evaluated, or result in an increase in the likelihood of occurrence of a malfunction or accident previously evaluated. This evaluation also concludes that all acceptance criteria for the applicable accidents continue to be met with no increase in the consequence of the accidents or malfunctions previously evaluated.
Based on these evaluations, this change may be implemented without NRC approval.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 12-01: Compensatory Actions in Support of a Prompt Operability Determination (POD), Revision 006 Activity
Description:
The changes evaluated are compensatory actions put in place to support a Prompt Operability Determination (POD) performed to address a non-conforming condition identified in the design analysis of the Essential Service Water (ESW) system flow return path to the Ultimate Heat Sink (UHS) retention pond. The non-conforming condition involves identification of a limiting single active failure not previously considered in calculation EF-54 (Ultimate Heat Sink Thermal Performance Analysis) for determining the maximum UHS pond temperature during a Large Break Loss-of-Coolant Accident (LBLOCA), which is the limiting Design Basis Accident (DBA) for the UHS.
Revision 006 of the POD credits the analysis performed by calculation EF-123, which supersedes calculation EF-54, for ensuring acceptable peak ESW and UHS pond temperatures, as well as adequate UHS inventory, over the required postulated accident timeline (30 days).
Long-term operation to the requirements of EF-123 requires a license amendment which will permanently resolve the non-conforming condition. (Reference Ameren Missouri Letter ULNRC-05867 dated December 13, 2012, as supplemented by Ameren Missouri letter ULNRC-05995 dated June 11, 2013). Once NRC approval of the requested license amendment is received, the compensatory actions described below will become permanent, operational requirements.
To support operation of the plant in the interim, the requirements of EF-123 will be implemented as compensatory actions associated with the POD. The POD requires establishment of the following compensatory actions:
- 1. Operator Action - Prior to 7 days after LBLOCA initiation, manual operator action is taken to secure a train of ESW, given no equipment failures.
- 2. Operator Action - Prior to 70 minutes after LBLOCA initiation, manual operator action is taken to identify an ESW equipment failure and secure the applicable ESW train.
- 3. Operator Action - Prior to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after LBLOCA initiation, manual operator action is taken to swap UHS Cooling Tower (CT) Bypass Valve and Fan Speed control from the UHS CT inlet temperature to the ESW Pump discharge temperature.
- 4. Operator Action - Manual operator action is taken to secure a train of ESW given a failure of EFHS0067/68 to swap UHS CT temperature control loops (Compensatory action 3).
- 5. Acceptance Criterion - Initial UHS pond temperature for a design basis LBLOCA has been lowered to 89F.
Compensatory actions 1-5 were deemed adverse, requiring a 50.59 Evaluation. Action 6 was determined to be non-adverse and thus screened out.
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Enclosure to ULNRC-06038 Summary of Evaluation:
The POD requires 6 compensatory actions to ensure the UHS and ESW equipment can perform their specified safety function in the limiting LBLOCA DBA, as shown in calculation EF-123.
The 10 CFR 50.59 Evaluation concluded that all of the compensatory actions may be implemented without prior NRC approval.
Compensatory Actions 1-4 Each of the credited operator actions (1-4) has been tested and/or analyzed, determining prior NRC approval is not required for implementation. These actions will enable Control Room personnel to diagnose and mitigate the effect of a valid single failure that has previously not been analyzed (i.e., the failure of either ESW UHS Cooling Tower Bypass Valve (EFHV0065 or EFHV0066) such that, for one train, ESW flow would be returned directly to the UHS pond without being directed to and over the Cooling Tower). The bypass valve single failure, if left uncorrected following a LBLOCA, has the potential to result in unacceptably high UHS temperatures and an unacceptably low UHS inventory (which would have an adverse impact on the likelihood of malfunction of SSCs including the Emergency Diesel Generators and Emergency Core Cooling System pumps). Also, failure to secure an ESW train has the potential to result in unacceptably low UHS inventory (which would have an adverse impact on the likelihood of occurrence of malfunctions of these SSCs).
The new/required operator actions (and the required completion times) are to be reflected in plant procedures as necessary and in operator training programs. Simulator exercises demonstrate that the actions can be completed in the time required, and have proven that the new actions can be accomplished when taking the aggregate operator workload into consideration.
The operator actions will have no adverse impact on day-to-day operation and normal operating procedures or work. In the event of an accident, the actions will allow Control Room personnel to diagnose and respond to a single failure of the type considered, prevent UHS bulk temperatures from going beyond the analyzed maximum value, and ensure the UHS has adequate inventory over its expected 30-day mission time. Maintaining the UHS bulk temperature and inventory within its analyzed limits will ensure the SSCs supported by the UHS will not become subject to an increase in the likelihood of malfunction above the levels currently analyzed within the FSAR.
In summary, the new operator compensatory actions (1-4) do not introduce a more than minimal increase in the likelihood of a malfunction of SSCs important to safety. Given the nature of these procedural changes, the measures to ensure successful completion of these actions, and the lack of barriers to successful completion of these actions, there is assurance that these activities can be successfully performed on demand.
Additionally, the proposed procedural changes (to reflect the new operator actions) do not create a possibility for a malfunction with a different result. These actions will ensure the UHS/ESW equipment can perform its specified safety functions for the limiting LBLOCA DBA, as shown in calculation EF-123.
Compensatory Action 5 Page 8 of 13
Enclosure to ULNRC-06038 The change to the day-to-day maximum allowed UHS pond temperature (5) has been analyzed and determined to not require prior NRC approval for implementation. The adverse effect of lowering the maximum allowed UHS pond temperature is a slight increase in how often the ESW pump and UHS Cooling Tower have to be operated to maintain the lower temperature. The increase is minimal with respect to wear and tear such that there is no more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. This equipment has been qualified for the life of the plant, and has a robust Preventative Maintenance Program which will identify and correct any potential failures due to additional cycling caused by the temperature reduction. Additionally, the lower initial UHS DBA temperature by itself does not increase the likelihood of a malfunction because it is an administrative limit change that is not an initiator of any new malfunctions, and no new failure modes are introduced.
Compensatory Action 6 The change to the day-to-day minimum required UHS pond level (6) was deemed to be non-adverse, and a 10 CFR 50.59 Evaluation was not required for this compensatory action.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 12-03: Plant Modification MP 10-0009, Installation of New Westinghouse Reactor Coolant Pump Shutdown Seals Activity
Description:
The proposed activity is a modification (MP 10-0009) to replace the Reactor Coolant Pump (RCP) Number 1 seal insert with the SHIELD Shutdown Seal (SDS) for each of the four RCPs at Callaway Plant.
The Number 1 seal is a film-riding face seal located above the lower RCP radial bearing. The film is produced by the system pressure drop across the seal and does not require seal rotation to establish the sealing function. To maintain the film, a controlled leakoff flow passes between the radially tapered seal faces. The SDS integrates new features into the number 1 seal insert and is located downstream of the film-riding face seal. The modified number 1 insert design includes a shoulder machined into the inner diameter at the top flange and a bore machined into the groove diameter above the shoulder. SDS sealing rings and a thermal actuator are placed into this shoulder and bore, respectively.
The SDS is designed to function only when exposed to an elevated fluid temperature downstream of the RCP number 1 seal, such as would occur as a result of the coincident loss of all thermal barrier heat exchanger cooling and number 1 seal injection cooling. SDS activation occurs over the temperature range of 250°F to 300°F as described in WCAP-17541-R, Rev 0, Implementation Guide for the Westinghouse Reactor Coolant Pump SHIELD Passive Thermal Shutdown Seal, March 2012. In its installed and non-activated state, the SDS resides completely out of the normal seal injection and shaft seal leakage flow paths.
When activated as designed, the SDS limits RCP shaft leakoff to 1 gallon per minute (GPM) per pump.
The Callaway FSAR was reviewed, and Sections 5.4.1.2.1, 5.4.1.3.1, 5.4.1.3.2, 5.4.1.3.4, and 5.4.1.3.10 were identified for revision. The Callaway Technical Specifications were reviewed, including their associated Bases. Implementation of the SDS does not require changes to the Technical Specifications or their associated Bases.
Summary of Evaluation:
The evaluation concluded that the proposed activity may be implemented without NRC approval.
The SDS will have no effect on the normal operation of the RCP Seals. The SDS is installed so as not to interfere with the seals or cooling water flow paths. The worst-case failure or inadvertent actuation of the SDS would not result in leakage or malfunctions that are any more severe than those currently evaluated.
The inadvertent SDS actuation frequency is very low compared to moderate frequency faults, and therefore, the change would not yield a more than a minimal increase in the frequency of occurrence of accidents and malfunctions evaluated in the FSAR. Further, any SDS malfunctions that could possibly occur (including inadvertent actuation) have been evaluated to be less severe Page 10 of 13
Enclosure to ULNRC-06038 than existing postulated events (such as an RCP locked rotor) and would not yield a different result than any previously evaluated in the FSAR. The SDS has no impact on the radiological consequences of postulated accidents such as the Partial Loss of Forced Reactor Coolant Flow or Locked Rotor event. For the Complete Loss of Reactor Coolant Flow event, actuation of the SDS favorably reduces RCS leakage to the containment atmosphere. However, credit is not being taken for the reduced leakage, and thus, SDS failure would not result in leakage greater than what is currently analyzed. In total, therefore, implementation of the SDS modification will have no effect on the radiological consequences of accidents previously evaluated in the FSAR.
Based on the above, this change may be implemented without NRC approval.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 13-01: Plant Modification MP 03-1002, Main Feedwater Pump Turbine Control System Replacement Activity
Description:
Due to equipment obsolescence, the existing Westinghouse Steam Generator level control and Main Feed Pump Speed/DP control system (which is primarily non-safety related) is being replaced with a digital feedwater control system per plant modification MP 03-1002. The digital feedwater control system allows for removal of multiple single point vulnerability concerns and provides additional functionality for improving plant control. The Main Feedwater Pump (MFP) controls on the main control board will be replaced by touch screen controls. The MFP Servo Positioner control equipment, Main Feedwater Regulating Valve (MFRV) positioners, MFRV Bypass Valve (MFRVBV) positioners, steam/feed flow recorders, and MFP local speed gauges will additionally be replaced. MFP speed, steam header pressure (SHP), feedwater header pressure (FHP), steam flow, feedwater flow, and narrow range SG level transmitter logic will be modified. Redundant equipment will be installed for the MFRV positioners and FHP/SHP/MFP speed transmitters.
Summary of Evaluation:
For the digital feedwater control system (DFWCS) to be installed at Callaway during Refuel 19, a failure modes and effects analysis was performed, including evaluation for common mode software failure (CMSF) effects on accident and hazard analyses. Changes are being made to FSAR Section 15.1.2 for the feedwater malfunction (FWM) accident analysis described therein to account for flow increases due to a CMSF occurring at both full power and hot zero power.
DNBR values require revision, but the resultant DNBR values remain below the analysis limits (as all acceptance criteria for the accident analysis continue to be met). CMSF effects that would decrease feedwater flow remain bounded by the existing FSAR Section 15.2.7 analysis for a complete loss of feedwater. In addition, the flooding hazard event for Area 5 (main steam tunnel between the reactor and turbine buildings) was reanalyzed. Although the analysis determined that the flood level would be increased, the capability to achieve safe shutdown would still be maintained.
Thus, although the CMSF addressed in this evaluation is determined to be of very low likelihood such that the change does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR, re-performance of the accident/hazard analyses with the incredible CMSF assumed yields results that still meet the acceptance criteria of those analyses. Consequently, this modification can be made to the plant without prior NRC approval.
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Enclosure to ULNRC-06038 10 CFR 50.59 Evaluation 13-02: Plant Modification MP 10-0053, SSPS Printed Circuit Board Replacement Activity
Description:
The proposed activity is a modification (MP 10-0053) to replace designated boards within the Solid State Protection System (SSPS) at Callaway Plant, including computer demultiplexer cards. The replacement circuit boards have been developed for the SSPS by Westinghouse.
The existing SSPS cards are aging and becoming obsolete due to obsolescence of Motorola High Threshold Logic (MHTL) devices used on the cards supplied by the Original Equipment Manufacturer (OEM), Westinghouse.
Summary of Evaluation:
In light of the extensive testing performed by Westinghouse for the new designed cards inputs and outputs, as well as extensive evaluation of the failure modes and effects for the new cards (which yielded a very low potential for common mode failures), the replacement of circuit cards for the Solid State Protection System represents a minimal to negligible change in frequency of occurrence, or likelihood, of a malfunction and less than a minimal increase in consequences of an accident or malfunction. Since the SSPS will continue to function in accordance with the assumptions of the accident analysis, the proposed changes do not affect a design basis limit for a fission product barrier. The changes involve no changes to the FSAR-described accident analysis itself, including all elements and assumptions, and as such, they do not involve any change to a method of evaluation described in the FSAR. The failure modes that may be potentially attributed to the SSPS are unchanged, as no new credible failure modes are introduced.
Therefore, the proposed changes do not create a new type of accident or malfunction of a structure, system, or component with a different result. As such, no prior NRC approval is needed in accordance with 10CFR50.59(c)(1).
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