RS-23-048, Exigent Amendment Request to Revise Design Basis Related to Seismic Requirements
| ML23066A292 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/07/2023 |
| From: | Lueshen K Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RS-23-048 | |
| Download: ML23066A292 (1) | |
Text
Constellation RS-23-048 March 7, 2023 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, D.C. 20555-0001 LaSalle County Station, Unit 1 Renewed Facility Operating License No. NPF-11 NRC Docket No. 50-373 4300 Winfield Road Warrenville, IL 60555 630 65 7 2000 Office 10 CFR 50.90 10 CFR 50.91 (a)(6)
Subject:
LaSalle County Station, Unit 1 - Exigent Amendment Request to Revise Unit 1 Design Basis Related to Seismic Requirements In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," and 10 CFR 50.91(a)(6), Constellation Energy Generation, LLC (CEG) is submitting an exigent license amendment request for LaSalle County Station (LSCS), Unit 1 to revise the Updated Final Safety Analysis Report (UFSAR) to allow the use of a temporary cable installation that does not meet seismic qualifications to enable restoration of the Reactor Protection System Turbine Control Valve #2 Channel B1 Scram Channel to OPERABLE until industrial safety and radiological concerns are minimized during the next refueling outage on Unit 1.
The request is subdivided as follows:
- provides a description and evaluation of the proposed change.
- provides a mark up of the affected U FSAR pages.
The proposed amendment to the LSCS Unit 1 UFSAR is being requested on an exigent basis consistent with 10 CFR 50.91(a)(6)(vi). The reasons for the exigency are discussed in to this letter.
The proposed changes have been reviewed by the LSCS Plant Operations Review Committee in accordance with the CEG Quality Assurance Program. provides a description and assessment of the proposed changes. Attachment 2 provides the proposed UFSAR markup.
Approval of the proposed amendment is requested by March 22, 2023. Site implementation will occur immediately upon Nuclear Regulatory Commission approval.
March 7, 2023 U.S. Nuclear Regulatory Commission Page 2 In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Officials.
There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Jason Taken at (630) 657-3660.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 7th day of March, 2023.
Respectfully, Kevin Lueshen Sr. Manager - Licensing Constellation Energy Generation, LLC : Description and Assessment : UFSAR Markup cc:
NRC Regional Administrator, Region III NRC Senior Resident Inspector - LaSalle County Station NRC Project Managers - LaSalle Station Illinois Emergency Management Agency - Division of Nuclear Safety
- Lueshen, Kevin Digitally signed by Lueshen, Kevin Date: 2023.03.07 15:08:27 -06'00'
ATTACHMENT 1 LaSalle County Station, Unit 1 Renewed Facility Operating License No. NPF-11 NRC Docket No. 50-373 Description and Assessment
ATTACHMENT 1 Description and Assessment
Subject:
LaSalle County Station, Unit 1 - Exigent Amendment Request to Revise Unit 1 Design Basis Related to Seismic Requirements 1.0 Summary Description 2.0 Detailed Description
2.1 Background
2.2 System Description 2.3 Description of the Proposed Change 2.4 Circumstances Establishing need for the Proposed Exigent Amendment 3.0 Technical Evaluation 3.1 Defense-In-Depth 3.2 Safety Impact for the Proposed Design Basis Change 3.2.1 Impact on Failure Modes 3.2.2 Risk Insights 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements 4.2 No Significant Hazards Consideration Analysis 4.3 Conclusion 5.0 Environmental Evaluation 6.0 References
ATTACHMENT 1 Description and Assessment 1 of 14 1.0
SUMMARY
DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," and 10 CFR 50.91(a)(6), Constellation Energy Generation, LLC (CEG) is submitting an exigent license amendment request for LaSalle County Station (LSCS), Unit 1 to revise the Updated Final Safety Analysis Report (UFSAR) to allow the use of a temporary cable installation that does not meet seismic qualifications to enable restoration of the Reactor Protection System Turbine Control Valve Channel B1 Scram Channel to OPERABLE until industrial safety and radiological concerns are minimized during the next refueling outage on Unit 1.
The proposed change to the LSCS UFSAR would modify the following requirements:
- 1) Section 7.2.2.1 of the LSCS UFSAR would be modified to allow the Unit 1 instrumentation cable for Turbine Control Valve Fast Closure SCRAM Channel B1 to be routed in non-seismically qualified conduit until startup following the next refueling outage on Unit 1.
- 2) Table 3.2-1 of the LSCS UFSAR will be modified to allow the Unit 1 instrumentation cable for Turbine Control Valve Fast Closure SCRAM Channel B1 to be routed in non-seismically qualified conduit until startup following the next refueling outage on Unit 1.
2.0 DETAILED DESCRIPTION
2.1 Background
On February 21, 2023, LSCS Unit 1 received a B1 Reactor Protection System (RPS) signal resulting in a Half Scram condition on Unit 1. Troubleshooting identified the cable that provides the signal associated with the Turbine Control Valve #2 Fast Closure RPS B1 Scram signal was degraded such that its reliability, and operability, were in question.
The proposed change is to install a temporary cable installation associated with the Turbine Control Valve #2 Fast Closure Reactor RPS B1 scram signal and restore the reliability and operability of the scram signal channel. The proposed change will remain in effect until the temporary cable is removed during the next refueling outage on Unit 1.
The temporary cable will be safety related and meet all the design and licensing basis requirements for the RPS instrumentation, but will be routed such that the Seismic Category I requirements required by the current licensing basis will not be met.
Installing the temporary cabling as non-Seismic Category I was determined to have a potential more than minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. This potential adverse effect was reviewed in a 10 CFR 50.59 evaluation. The review determined that the proposed activity results in a more than minimal increase in the likelihood of an occurrence of a malfunction of an SSC important to safety and NRC approval is required.
ATTACHMENT 1 Description and Assessment 2 of 14 Currently LaSalle is in a 30-day Risk Informed Technical Specification timeclock for the Turbine Control Valve Fast #2 Closure B1 Function that expires on March 23, 2023 at 1843 CDT, at which point the RPS B1 logic channel must be placed in the tripped condition resulting in a half-scram condition.
2.2 System Description
RPS is designed to initiate a reactor scram with precision and reliability to prevent or limit fuel damage following abnormal operational transients, to prevent damage to the reactor coolant pressure boundary as a result of excessive internal pressure, and to limit the uncontrolled release of radioactive materials from the fuel assembly or reactor coolant pressure boundary by detecting conditions that threaten the fuel assembly or reactor coolant pressure boundary with inputs derived from variables that are true, direct measures of operational conditions.
RPS includes the motor-generator power supplies, sensors, relays, bypass circuitry, and switches that cause rapid insertion of control rods (scram) to shut down the reactor. It also includes outputs to the process computer system and annunciators, although these latter two systems are not part of the reactor protection system. The reactor protection system is classified as Safety Related, Seismic Category I, and Quality Group B (Electric Safety Class 1E), with the exception of the motor-generator power supplies which are non-Class 1E.
RPS is comprised of two independent trip systems (A and B), with two logic channels in each trip system (channels A1 and A2, B1 and B2). The outputs of the logic channels in a trip system are combined in a one-out-of-two logic so either logic channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram.
Turbine Control Valve Fast Closure (RPS)
With the reactor and turbine generator at power, fast closure of the turbine control valves can result in a significant addition of positive reactivity to the core as nuclear system pressure rises.
The turbine control valve fast closure scram initiates a scram earlier than either the neutron monitoring system or nuclear system high pressure. It is required to provide a satisfactory margin to core thermal-hydraulic limits for this category of abnormal operational transients. The scram counteracts the addition of positive reactivity resulting from increasing pressure by inserting negative reactivity with control rods. Although the nuclear system high-pressure scram, in conjunction with the pressure relief system, is adequate to preclude over pressurizing the nuclear system, the turbine control valve fast closure scram provides additional margin to the nuclear system pressure limit. The turbine control valve fast closure scram setting is selected to provide timely indication of control valve fast closure.
Turbine control valve fast closure inputs to the reactor protection system come from oil line pressure switches on each of four fast-acting control valve hydraulic mechanisms. These hydraulic mechanisms are part of the turbine control, and they are used to effect fast closure of the turbine control valves. These pressure switches provide signals to the reactor protection system. If hydraulic oil line pressure is lost, a turbine control valve fast closure scram is initiated.
ATTACHMENT 1 Description and Assessment 3 of 14 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation The EOC-RPT instrumentation initiates a recirculation pump trip (RPT), if operating in fast speed, to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to the MCPR Safety Limit (SL).
The EOC-RPT instrumentation is comprised of sensors that detect initiation of closure of the Turbine Stop Valves, or fast closure of the Turbine Control Valves, combined with relays and logic circuits, to actuate reactor recirculation pump downshift logic to trip each pump from fast speed (60 Hz). The channels include instrument switches that actuate pre-established setpoints.
Each EOC-RPT trip system is a two-out-of-two logic for each Function; including the Turbine Control ValveFast Closure, Trip Oil PressureLow signals are required for a trip system to actuate. If either trip system actuates, both recirculation pumps, if operating in fast speed, will trip.
The Turbine Control Valve-Fast Closure function is designed to trip the recirculation pumps, if operating in fast speed, in the event of a turbine trip or generator load rejection to mitigate the neutron flux, heat flux, and pressurization transients and to increase margin to the MCPR SL.
The OPERABILITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions.
2.3 Description of the Proposed Change The proposed change adds safety-related, but non-seismically qualified, temporary cabling to RPS to restore OPERABILITY to the RPS Instrumentation TS 3.3.1.1. A simplified one-line diagram is provided below for illustrative The temporary cable installation is associated with the Turbine Control Valve #2 Fast Closure B1 Scram signal. Turbine control valve fast closure inputs to the reactor protection system originate from oil line pressure switches on each of four fast-acting control valve hydraulic mechanisms. These hydraulic mechanisms are part of the main turbine control, and they are used to effect fast closure of the turbine control valves. These pressure switches provide signals to RPS. If hydraulic oil line pressure is lost, a turbine control valve fast closure scram is initiated.
Cable 1RP038 connects Turbine Control Valve #2 EHC Oil Line pressure switch 1C71-N005B to relay 1C71A-K008B. The temporary cable will be spliced to cable 1RP038 at junction box 1JB304A and routed to the pressure switch.
Cable 1RP038 is the RPS B1 logic channel cable. The LSCS UFSAR requires that an RPS cable be Seismic Category I and Safety Class 1E. The temporarily installed cable will be Safety Class 1E standard but will not be mounted to the Seismic Category I standard. The temporarily installed cable will be installed in flexible stainless-steel conduit to meet the cable separation requirements for an RPS cable.
ATTACHMENT1 Description and Assessment The new conduit path was developed and deviates from the LaSalle Electric Installation Standard for seismic conduit installations. The proposed change will restore the RPS B1 channel for Turbine Control Valve #2 Fast Closure to OPERABLE.
768' Turbine Floor Pull Sleeve 743' 12 R
--+------< 1JB304A AUX BLD.
A TDRFP Room n*==::..._- ~============-11 t Open;ng ;nwall
~ater Bay Floor 13 L
1JB164A f--------+-~
Original Route Proposed Route Figure 2. 1 - Current and Proposed Design for Affected Turbine Control Valve RPS Channel 2.4 Circumstances Establishing need for the Proposed Exigent Amendment As provided for in 10 CFR 50.91 (a)(6)(vi), CEG is required to explain the exigency and why this cannot be avoided.
The NRC refers to "exigent situations" as those in which a licensee and the Commission must act quickly and that the time does not permit the Commission to publish a Federal Register notice allowing 30 days for prior public comment. Under such an exigent situation, the Commission may either provide notice of an opportunity for public hearing allowing at least two weeks from the date of the notice for prior public comment or use local media to provide reasonable notice to the public if it also determines that the amendment involves no significant hazards considerations.
At 1743 CST on February 21, 2023, Unit 1 experienced a half-scram condition on channel B1 of its Reactor Protection System. This condition places the system in the Limiting Condition for Operation, Reactor Protection System Instrumentation, 3.3.1.1 and requires the affected channel to be placed in a trip condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR in accordance with its Risk Informed Completion Time (RICT). The RICT completion time for this action is 30 days.
This half-scram was bypassed in accordance with station abnormal operating procedures and the channel, as of the date of this application, remains not tripped. The 30-day RICT expires at 1843 CDT on March 23, 2023, at which point the channel must be placed in trip to comply with TS 3.3.1.1 Required Action A.1.
4 of 14
ATTACHMENT 1 Description and Assessment 5 of 14 On February 25, 2023, troubleshooting revealed the cause of the half-scram condition to be a degraded cable, 1RP038, for the Turbine Control Valve #2 RPS B1 logic channel of the Reactor Protection Instrumentation System.
On March 3, 2023, after performing a 10 CFR 50.59 evaluation, station personnel concluded that a license amendment request (LAR) and Nuclear Regulatory Commission (NRC) approval would be required to revise the LSCS Unit 1 licensing basis to allow the use of non-seismically qualified cabling as a temporary modification to RPS instrumentation to restore RPS Instrumentation to an OPERABLE status within the TS required completion time.
CEG determined the safest and most reliable option is to revise the current licensing basis and to set in place compensatory actions to ensure that the risks introduced from the non-seismic qualifications of the temporary modification are mitigated.
This exigency cannot be avoided because without approval of this license amendment application, LSCS will be required to place the affected Turbine Control Valve #2 B1 RPS channel in trip. The consequence of tripping the affected channel is that any spurious, or valid trip condition during normal operation or performance of required surveillance testing on the A RPS channels, would result in an unnecessary full reactor scram and shutdown transient without a corresponding health and safety benefit.
Should LSCS be required to repair the originally installed cabling instead of installing this temporary modification, LSCS Unit 1 would have to shutdown. During at-power operations, the higher elevation of the heater bay, where the current RPS cabling is routed, has elevated temperatures of approximately 150F. Notwithstanding the industrial safety risk to personnel associated with working at elevation coincident with high temperatures, the area in question also experiences dose rates in the range of 500-2100 mR/hour. There is no safe way to complete this repair during at-power operation, and thus the safest option is temporarily revising the current licensing basis to avoid the unnecessary plant shutdown.
3.0 TECHNICAL EVALUATION
3.1 Defense-In-Depth During the timeframe which the proposed change will be in effect, RPS and its Instrumentation will remain compliant with the respective TS Limiting Conditions for Operation (LCO). Should an event or failure mode occur (discussed in Section 3.2) related to the temporarily installed RPS B1 logic channel, the remaining channels of RPS Instrumentation would be available to ensure that a reactor scram would still occur if required.
Four channels of Turbine Control Valve Fast Closure, Trip Oil PressureLow Function, with two channels in each trip system arranged in a one-out-of-two taken twice logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
Thus, a failure of the temporarily installed cabling would not preclude the safety function associated with the affected channel from occurring. The remaining three logic channels of
ATTACHMENT 1 Description and Assessment 6 of 14 Turbine Control Valve Fast Closure Scram are available to successfully complete the safety function.
Functional diversity is provided by monitoring independent reactor vessel variables. Reactor pressure, water level, and neutron flux are all independent and are separate inputs to the system. Turbine control valve fast closure is anticipatory of a reactor vessel high pressure and is a separate input to RPS. Operability of the anticipatory signals from the turbine control valve fast closure following a safe shutdown earthquake is not a safety design criteria. If the gross failure of the trip occurred where the safety function failed to occur, the reactor would scram on high neutron flux or high reactor pressure.
Therefore, based on the redundancy, functional diversity, and design of the currently installed equipment along with the temporarily installed proposed change, defense-in-depth is sufficient for the proposed change.
3.2 Safety Impact for the Proposed Design Basis Change 3.2.1 Impact on Failure Modes The failure modes of the proposed Turbine Control Valve #2 RPS B1 cabling remain unchanged from the existing installed cable and are:
- 1) Open Conductor
- 2) Short Conductor to Ground
- 3) Conductor to Conductor Short Open Conductor An open conductor will cause the Turbine Control Valve #2 fast closing scram relay to de-energize. De-energizing this relay will initiate a half scram on the RPS B1 channel. This failure mode meets the fail-safe requirement.
Short Conductor to Ground A short conductor to ground will cause the in-line fuse to clear resulting in a loss of power to the Turbine Control Valve #2 fast closing scram relay. Clearing of the fuse will de-energize the relay resulting in initiation of a half scram on the RPS B1 channel. RPS is designed to be fail-safe. This failure mode meets the fail-safe requirement.
Conductor-to-Conductor Short The remaining failure mode would be a conductor-to-conductor short. This failure mode would disable the RPS B1 channels ability to initiate a half scram if the Turbine Control Valve #2 EHC pressure is lost. The remaining channels (A1, A2, B2) are unaffected and are not susceptible to this failure mode. Therefore, if RPS channel B1 incurred this failure mode, the remaining three
ATTACHMENT 1 Description and Assessment 7 of 14 channels are capable of initiating a reactor scram from Turbine Control Valve Fast Closure even if RPS channel B1 were to fail.
LSCS is designed to the IEEE 379-1972 standard IEEE Standard Application of the Single Failure Criterion to Nuclear Power Generating Station Class 1E Systems. Detectable failures will be identified through periodic testing or will be revealed by alarm or anomalous indication as defined in IEEE 379-1972.
The conductor-to-conductor failure mode meets the single failure criterion because it is detectable by testing and therefore is classified as detectable.
Turbine control valve surveillance testing tests the turbine control valves by energizing the fast-acting solenoid, bleeding off EHC hydraulic oil pressure until the RPS B1 half scram signal is initiated.
Based on the current cable having the same failure mode as the replacement cable, the conductor-to-conductor short has already been evaluated for the single failure criteria.
Therefore, the temporary installation does not alter the single failure criteria.
The most credible events resulting in a conductor to conductor short are either 1) Fire or 2)
Seismic Event.
Fire Hazards LSCS has fire protection equipment that provides the Main Control Room Operator indication of a potential fire condition along the proposed temporary routing path. The temporary routing path will go through multiple fire zones within Fire Zone 5.
Fire Zone 5A1 and 5B1 Both fire zone 5A1 and 5B1 have the automatic sprinkler protection designed to produce a density of 0.3 gpm/ft2 over 3,000 ft2 and 0.2 gpm/ft2 over 10,000 ft2. Upon actuation, each sprinkler system alarms both locally and in the control room.
Fire Zone 5A3 The turbine bearing pre-action sprinkler system deluge valve and the alternator-exciter CO2 systems can be actuated either automatically (utilizing heat detectors) or manually.
Both systems activate alarms locally and in the control room up on actuation. This zone contains at least 15 dry chemical and 6 carbon dioxide portable fire extinguishers and 2-150 lb. dry chemical units.
ATTACHMENT 1 Description and Assessment 8 of 14 Cable/Conduit Design The temporary cable will be a Type 2146 safety related cable that meets the safety related cable requirement within IEEE 383-1974, IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations. The temporary cable will be installed in a stainless-steel flexible conduit. The temporary cable will be inside the steel flexible conduit and therefore there is negligible impact to the fire load in the area.
In conclusion, there are two main factors that limit the probability of a conductor-to-conductor short resulting from a fire. First, the fire protection equipment throughout zones 5A1, 5A3, and 5B1 is designed to quickly extinguish any fires along the temporary routing path. Second, the temporary stainless-steel conduit and safety related cable protect the Turbine Control Valve #2 B1 RPS function if a fire was to occur within the designated fire zones.
Seismic Considerations The cable routing per the current licensing basis consists of a rigid conduit being run along the ceiling of the Turbine Building Heater Bay, down the columns and to the Turbine Control Valve Fast Closure pressure switches. The proposed temporary cable routing path will be from the Auxiliary Building to the Turbine Building Heater Bay (including a partial route across the floor) and finally to Turbine Control Valve #2 Fast Closure B1 pressure switch.
The temporary cable will be routed in a stainless-steel flexible conduit and supported where it cannot become a hazard to any additional equipment. The temporary cable routing path does not have any additional safety related equipment in the vicinity except the RPS limit switches which already comply with separation criteria. The greatest risk for a conductor-to-conductor short will be the temporary route across the heater bay floor. This path increases the possibility of an object falling and crushing the conduit, pinching the conductors together. The possibility of this occurring is unlikely because the stainless-steel flexible conduit meets the minimum yield strength of 25 ksi for a rigid steel conduit and the rigidity of components in the routing path.
Lastly, heater bay access is strictly controlled during at-power operations due to the industrial and radiological risks.
Notwithstanding the above, to further mitigate the impact of a crushing risk on the temporary cabling, horizonal runs will be inspected monthly. Vertical runs are less likely to experience this crushing risk because they are more susceptible to a strike, not a crushing force.
In the unlikely occurrence of a Safe Shutdown Earthquake seismic event, both the horizontal and vertical runs will need a full temporary cable route inspection to ensure there is no visible damage.
ATTACHMENT 1 Description and Assessment 9 of 14 Upon implementation of this amendment, station alarm response procedures will direct inspection of the temporary cabling modification following receipt of a valid main control room seismic alarm.
Structural Support of the Temporary Cable Installation The proposed cable routing will be from the Auxiliary Building to the Turbine Building Heater Bay to Turbine Control Valve Fast Closure B1 pressure switch. The temporary cable is procured safety related. It will be routed in a stainless-steel flexible conduit and supported where it cannot become a hazard to any other equipment. The cable being temporarily secured to the Turbine Building floor exposes the cable to less seismic forces than being secured to the ceiling. The sensors associated with the Turbine Control Valve Fast Closure themselves are procured Safety Related and Seismically qualified. In response to LSCS FSAR question 031.135, LSCS stated that these sensors in the Turbine Building are not relied upon for the protection of public health and safety during faulted plant conditions. These inputs serve to provide adequate fuel design safety limit margin during such operational transient conditions as Generator Load rejection.
The temporary, stainless-steel flexible conduit will be supported from existing equipment/pipe supports, structural steel, steel handrail, steel bar grating or other structures that are rigidly anchored to the building structural steel or reinforced concrete approximately every six feet. The total weight of the cable/conduit combination is approximately 7 lbs. per attachment point.
Where the flexible conduit must run across a floor slab, vertically up a wall or horizontally across a ceiling, the flexible conduit will be anchored to the concrete at intervals approximately every six feet.
Where the flexible conduit must run across a steel beam flange, the flexible conduit will be anchored to the beam flange at intervals approximately every six feet.
Mitigating Actions Due to the increased risk associated with the temporary cabling not meeting Seismic Category 1 qualifications, upon implementation of this amendment station procedures will direct performance of monthly visual inspections of the horizontal runs of the
ATTACHMENT 1 Description and Assessment 10 of 14 temporary cable routing during the time that the temporary licensing basis change is in effect.
Upon implementation of this amendment, station alarm response procedures will direct inspection of the temporary cabling modification following receipt of a valid main control room seismic alarm.
3.2.2 Risk Insights This temporary non-Seismic Category 1 modification of the cable routing of the Turbine Control Valve #2 Fast Closure RPS B1 signal has no significant impact on the plant seismic risk profile.
The likelihood of seismic-induced failure to scram is a very low hazard and that likelihood is non-significantly impacted by this temporary modification. RPS is designed with multiple diverse channel inputs; if seismic-induced failure of the Turbine Control Valve #2 Fast Closure RPS B1 signal were postulated to occur, the reactor would still scram on other diverse signals (e.g., High Neutron Flux and High Reactor Pressure).
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements 10 CFR 50.91(a)(6) provides the requirements to be met to allow the NRC to perform expedited approval of a license amendment under exigent circumstances. As discussed in Section 2.4 of this letter, CEG is requesting exigent processing of this license amendment request, as a delay in approval of the proposed change would place LSCS Unit 1 in a condition that should routine surveillance testing take place on the other channel of RPS, or a spurious, invalid condition occur on that other channel, LSCS Unit 1 would undergo an unnecessary full reactor scram and unnecessary plant transient without a corresponding health and safety benefit.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plans, Criterion 2, Design bases for protection against natural phenomena, states that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed. The design considerations and effects on failure modes have been evaluated such that this change complies with the requirements of GDC 2.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plans, Criterion 20, Protection system functions, states that the protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of
ATTACHMENT 1 Description and Assessment 11 of 14 systems and components important to safety. RPS continues to satisfy this design criteria because RPS will continue to sense accident conditions and to initiate the operation of systems and components important to safety.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plans, Criterion 23, Protection system failure modes, states that the protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. The proposed amendment will not affect the compliance of LSCS Unit 1 to GDC 23. Specifically, as described in this letter, the failure modes have been evaluated and no new failure modes not previously evaluated will be introduced as a result of this amendment with respect to RPS.
10 CFR 50.59 allows licensees to make changes to the plant as described in the UFSAR only if the changes do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety. As discussed in Section 3.0 above, the proposed change results in a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety and therefore requires NRC approval.
4.2 No Significant Hazards Consideration Analysis In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Constellation Energy Generation, LLC (CEG) is submitting an exigent license amendment request for LaSalle County Station (LSCS), Unit 1 to revise the Updated Final Safety Analysis Report (UFSAR) to allow the use of a temporary cable installation that does not meet seismic qualifications to restore the Reactor Protection System Turbine Control Valve #2 Channel B1 Scram Channel to OPERABLE until industrial safety and radiological concerns are minimized during the next refueling outage on Unit 1.
CEG has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises the LSCS UFSAR to allow a temporary cable installation that does not meet seismic qualifications to restore the Reactor Protection System Turbine Control Valve Channel #2 B1 Scram Channel to OPERABLE until industrial safety and radiological concerns are minimized during the next refueling outage on Unit 1.
The proposed change does not introduce the possibility of a change in the frequency of an accident because Turbine Control Valve Fast Closure is not an initiator of any accident,
ATTACHMENT 1 Description and Assessment 12 of 14 there is no change to the pressure switch model, setpoint or mounting. The proposed change only affects the Seismic Category I requirements of the signal cable.
No technical specification requirements will change because of the proposed change and all surveillances and maintenance that are performed consistent with the current licensing basis will be performed with the proposed change on the temporary cable installation.
As part of the single failure design criteria, loss of the affected channel of the Reactor Protection System Instrumentation system does not prevent the safety function from being performed.
The proposed change does not adversely affect accident initiators or precursors, and does not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated or maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed change, and the assumptions used in determining radiological consequences of previously evaluated accidents are not affected.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation.
The proposed change to the design basis does not alter or involve any design basis accident initiators. The accident, Decrease In Heat Removal By The Secondary System, is initiated from a Generator Load Rejection which the Electro-Hydraulic Control system sends a signal to the Turbine Control Valve Fast acting solenoid. The fast-acting solenoid bleeds EHC hydraulic oil lowering the pressure monitored by the Turbine Control Valve Fast Closure pressure switch until the setpoint is reached which initiates the B1 RPS signal.
Therefore, the proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR.
No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced by the proposed change. The proposed change does not invalidate assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
ATTACHMENT 1 Description and Assessment 13 of 14
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change does not involve a significant reduction in a margin of safety because:
- 1. The inherent defense-in-depth of the Reactor Protection System Instrumentation system ensures that a single failure of any channel will not result in a failure of the system to perform its design function.
- 2. The one-time change to the design basis affects only one channel of one function of the Reactor Protection System. The remaining channels and functions are unaffected by the proposed temporary change to the licensing basis.
- 3. Only one failure mode for the proposed temporarily installed cable is not fail-safe.
To ensure this one failure mode does not result in adverse consequences, LSCS will implement mitigating actions (e.g., periodic inspections) to ensure the remaining failure mode is monitored during the timeframe the one-time change is in effect.
- 4. There are no changes to the current licensing basis with respect to setpoints and the system response will remain the same. The temporarily installed cable will be required to meet all Technical Specification requirements imposed consistent with the other three channels of this function.
- 5. Functional diversity is provided by monitoring independent reactor vessel variables.
Pressure, water level, and neutron flux are all independent and are separate inputs to the system. Turbine control valve fast closure is anticipatory of a reactor vessel high pressure and is a separate input to RPS. Operability of the anticipatory signals from the turbine control valve fast closure following a safe shutdown earthquake is not a safety design criteria. If the gross failure of the trip occurred where the safety function failed to occur, the reactor would scram on high neutron flux or high reactor pressure.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, CEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.3 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations,
ATTACHMENT 1 Description and Assessment 14 of 14 and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVIRONMENTAL EVALUATION The proposed change would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
6.0 REFERENCES
- 1. LaSalle Updated Final Safety Analysis Report, Rev. 025
- 2. IEEE 379-1972, IEEE Standard Application of the Single Failure Criterion to Nuclear Power Generating Station Class 1E Systems.
- 3. LaSalle Units 1 and 2 Technical Specifications
- 4. IEEE 383-1974, IEEE Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations.
ATTACHMENT 2 LaSalle County Station, Unit 1 Renewed Facility Operating License No. NPF-11 NRC Docket No. 50-373 UFSAR Markup
LSCS-UFSAR In addition to the above safety design requirements, the reactor protection system instrumentation and controls comply with the specific regulatory requirements shown in Tables 7.1-2 and 7.1-3.
7.2.1.2 Power Generation Design Bases The reactor protection system has one power generation objective. The setpoints, power sources, and controls and instrumentation are arranged in such a manner as to preclude spurious scrams.
7.2.2 7.2.2.1
System Description
General The reactor protection system includes the motor-generator power supplies, sensors, relays, bypass circuitry, and switches that cause rapid insertion of control rods (scram) to shut down the reactor. It also includes outputs to the process computer system and annunciators, although these latter two systems are not part of the reactor protection system. Trip signals are received from the neutron monitoring system; however, other portions of this system are treated in Sections 7.5, 7.6, and 7.7.
The reactor protection system is classified as Safety Related, Seismic Category I, and Quality Group B (Electric Safety Class lE), with the exception of the motor-generator power supplies which are non-Class 1 Table 7.2-1 lists the limits for instruments tli rovide signals for the system. Figure 7.2-2 summarizes the reactor protection signals t 7.2.2.2 Power Sources and the Unit 1 instrumentation cable for Turbine Control Valve
- 2 Fast Closure SCRAM The reactor protection system receives power from two hi Channel B 1 which will be routed sets (Figure 7.2-1). A flywheel provides high inertia suffic non-seismic until MODE 2 frequency within 5% of rated values for at least 1 second£ following startup from refueling to the drive motor.
outage L 1 R20.
Alternate power is available to either reactor protection system bus. The alternate power switch is interlocked to prevent simultaneous feeding of both buses from the same source. The switch also prevents paralleling of a motor-generator set with the alternate supply. The station batteries supply d-c power to the backup scram valve solenoids.
An electrical protection assembly (EPA) consisting of Class lE protective circuitry is installed between the reactor protection system and each of the power sources (two 7.2-3 REV. 19, APRIL 2012
LSCS-UFSAR TABLE 3.2-1 (SHEET 4 OF 32)
QUALITY (4a)
QUALITY (4b)
(4c)
(2)
(3)
SEISMIC (5)
GROUP ASSURANCE ELECTRICAL PURCHASE PRINCIPAL COMPONENT(])
LOCATION CATEGORY CLASSIFICATION REQUIREMENT CLASSIFICATION DATE COMMENTS
- 8. Piping, beyond isolation valves RB I
B I
NA 9-74
- 9. Electrical and instrument modules RB I
NA I
NONIE
- 10. Cable RB,A II NA I
NONIE 10-75 (15) 11. Serv. System valves and piping RB II D
II NONIE 6-73 VI. Neutron Monitoring System
B I
NA 1-74
B I
NA 1-74
NA I
lE 1-74 (15)
IE 5-75
- 5. LPRM, incore detector assemblies PC I
A I
IE VII. Reactor Protection System
- 1. Electrical and instrument modules T,PC,RB,A I
NA I
IE 2-74
- 2. Cables T,PC,RB,A I
I IE 10-75 VIII. Process Radiation Monitors
- 1. Electrical and instrument modules RB,A I
NA I
IE 5-74 (15) main steam line and reactor building ventilation monitors
- 2. Cable, main steamline and reactor RB,A I
NA I
IE 10-75 building ventilation monitors
- 3. Electrical and instrument modules A,T,RB II NA II NONIE 7-74 (15) for process liquid, process ventilation, air ejector and off-gas radiation monitoring systems IX. Residual Heat Removal (RHR) System
- 1. Heat exchangers, primary side RB I
B I
NA 5-71 (35)
- 2. Heat exchangers, secondary side RB I
C I
NA 5-71 (36)
A I
NA 9-74 (10) outermost isolation valves
- With the exception of Unit 1 instrumentation cable for the Turbine Control Valve #2 Fast Closure SCRAM Channel B1 which will be routed non-seismic until MODE 2 following startup from refueling outage L 1 R20 TABLE 3.2-1 REV. 18, APRIL 2010