ML15343A128
| ML15343A128 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 12/11/2015 |
| From: | Lisa Regner Plant Licensing Branch IV |
| To: | Koehl D South Texas |
| Regner L | |
| References | |
| TAC MF7142 | |
| Download: ML15343A128 (39) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Dennis L. Koehl President and CEO/CNO STP Nuclear Operating Company South Texas Project P.O. Box 289 VVadsworth, TX 77483 December 11, 2015
SUBJECT:
SOUTH TEXAS PROJECT UNIT 1 - ISSUANCE OF AMENDMENT RE:
REVISION TO TECHNICAL SPECIFICATIONS FOR ONE OPERATING CYCLE OPERATION VVITH 56 CONTROL RODS (EMERGENCY CIRCUMSTANCES)
Dear Mr. Koehl:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 208 to Facility Operating License No. NPF-76 for the South Texas Project (STP) Unit 1.
The amendment consists of changes to the Technical Specifications (TS) in response to your application dated December 3, 2015, as supplemented by letter dated December 9, 2015.
The amendment revises STP, Unit 1, TS 5.3.2 "Control Rod Assemblies," to add a footnote reflecting that for one operating cycle, Unit 1 Cycle 20, the core will contain 56 full-length control rod assemblies instead of 57 full-length assemblies. This change will allow Unit 1 to operate for Cycle 20, approximately 18 months, while repair plans are completed for the control rod drive mechanism. This amendment was necessitated by emergent issues identified during control rod testing conducted at the conclusion of the recent refueling outage.
The license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations due to the time critical nature of the amendment. In this instance, an emergency situation exists in that the proposed amendment is needed to allow the licensee to resume operations while planning to repair the damaged control rod drive mechanism, and could not have been anticipated by the licensee.
A copy of the related Safety Evaluation is enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the final no significant hazards determination. A Notice of Issuance addressing the final no significant hazards determination and opportunity for a hearing associated with the emergency circumstances, will be included in the Commission's next biweekly Federal Regi r notice.
Docket No. 50-498
Enclosures:
- 1. Amendment No. 208 to NPF-76
- 2. Safety Evaluation cc w/encls: Distribution via Listserv Lisa M. Regner, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-498 SOUTH TEXAS PROJECT UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 208 License No. NPF-76
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by STP Nuclear Operating Company (STPNOC)*
acting on behalf of itself and for NRG South Texas LP, the City Public Service Board of San Antonio (CPS), and the City of Austin, Texas (COA) (the licensees), dated December 3, 2015, as supplemented by letter dated December 9, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- STPNOC is authorized to act for NRG South Texas LP, the City Public Service Board of San Antonio, and the City of Austin, Texas, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-76 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. STPNOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of issuance.
Attachment:
Changes to the Facility Operating License No. NPF-76 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: December 11, 2015
ATTACHMENT TO LICENSE AMENDMENT NO. 208 FACILITY OPERATING LICENSE NO. DPR-76 DOCKET NO. 50-498 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 5-6 5-6
SOUTH TEXAS LICENSE (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. STPNOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Not Used (4)
Initial Startup Test Program (Section 14. SER)*
Any changes to the Initial Test Program described in Section 14 of the Final Safety Analysis Report made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
(5)
Safety Parameter Display System (Section 18. SSER No. 4)*
Before startup after the first refueling outage, HL&P[**] shall perform the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to issues as described in Section 18.2 of SER Supplement 4.
(6)
Supplementary Containment Purge Isolation (Section 11.5. SSER No. 4)
HL&P shall provide, prior to startup from the first refueling outage, control room indication of the normal and supplemental containment purge sample line isolation valve position.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
- The original licensee authorized to possess, use and operate the facility was HL&P.
Consequently, historical references to certain obligations of HL&P remain in the license conditions.
Amendment No. 208
DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies. Each fuel assembly shall consist of a matrix of zircaloy, ZIRLO' or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy, ZIRLO' or stainless steel filler rods for fuel rods, in accordance with NRG-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 57* full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.9 inches of absorber material. The absorber material within each assembly shall be silver-indium-cadmium or hafnium. Mixtures of hafnium and silver-indium-cadmium are not permitted within a bank. All control rods shall be clad with stainless steel tubing.
5.4 (NOT USED) 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE 5.6.1 CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- The Unit 1 Cycle 20 core shall contain 56 full-length control rod assemblies with no full-length control rod assembly installed in core location D-6.
SOUTH TEXAS - UNITS 1 & 2 5-6 Unit 1 -Amendment No. 2,10,16,43, 61,65, 89,92,98, 104, 198, 208 Unit 2 - Amendment No. 2,6,32,50 54,76,79,85, 91, 186
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 208 TO FACILITY OPERATING LICENSE NO. NPF-76 STP NUCLEAR OPERATING COMPANY, ET AL.
1.0 INTRODUCTION
SOUTH TEXAS PROJECT, UNIT 1 DOCKET NO. 50-498 By application dated December 3, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15343A347), as supplemented by letter dated December 9, 2015 (ADAMS Accession No. ML15344A304), STP Nuclear Operating Company (STPNOC, the licensee) requested changes to the Technical Specifications (TSs), Appendix A to Facility Operating License No. NPF-76, for the South Texas Project (STP), Unit 1.
The amendment would modify TS 5.3.2, "Control Rod Assemblies" to add a footnote allowing the use of 56 full-length control rods for operating cycle Unit 1 Cycle 20, instead of 57 full-length control rods, and that there is no full-length control rod in the assembly at core location D-6.
This amendment was necessitated by emergent issues identified during control rod testing following completion of the recent refueling outage. The proposed amendment would allow the plant to operate for one cycle, approximately 18 months, while repair plans are completed for the control rod drive mechanism (CROM).
The licensee requested that the Nuclear Regulatory Commission (NRC) staff process this submittal under emergency circumstances as described in 10 CFR 50.91 (a)(5).
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act (Act) requires applicants for nuclear power plant operating licenses to include TSs as part of the license. These TSs are derived from the plant safety analyses.
In Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR), the NRC established its regulatory requirements related to the content of TSs.
Pursuant to 1 O CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements;
l (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs.
The regulations in 10 CFR 50.36(c)(2)(iii)(4) state that Design Features are required to be in Technical Specifications to show features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety.
The number of Control Rod Assemblies is included in this section.
Licensees may propose revisions to the TSs. The NRC staff reviews proposed changes and will generally issue changes provided that the plant-specific review supports a finding of continued adequate protection of public health and safety because: (1) the change is editorial, administrative, or provides clarification (i.e. no requirements are materially altered), (2) the change is more restrictive than the licensee's current requirement, or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. The detailed application of this general framework, and additional specialized guidance, is discussed in Section 3.0 of this safety evaluation in the context of the proposed TS changes contained in the licensee's license amendment request (LAR).
The 10 CFR 50, Appendix A, General Design Criteria (GDC) applicable to this LAR are discussed below:
GDC 2, "Design bases for protection against natural phenomena," states, in part, that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes 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.
GDC 4, "Environmental and dynamic effects design bases," as it relates to the structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions during normal plant operation as well as during postulated accidents.
GDC 10, "Reactor design," states that that the reactor core and associated coolant, control, and protection systems shall be designed to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
GDC 11, "Reactor inherent protection," states that the reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
r-
- GDC 12, "Suppression of Reactor Power Oscillations," states that the reactor core and associated, coolant, control, and protection systems shall be designed to assure that power oscillation which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
GDC 14, "Reactor Coolant Pressure Boundary," states the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
GDC 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.
GDC 25, "Protection system requirements for reactivity control malfunctions,"
states that protection system shall be designed to assure that specified acceptable fuel design limits are not exceed for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.
GDC 26, "Reactivity control system redundancy and capability," states that two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
GDC 27, "Combined reactivity control systems capability," states that the reactivity systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained; GDC 28, "Reactivity limits," states that reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can be neither ( 1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structure or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in the reactor coolant temperature and pressure, and cold water addition.
GDC 29, "Protection against anticipated operational occurrences," states that the protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operation occurrences.
NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP) Section 4.6, "Functional Design of Control Rod Drive System," informed the regulatory requirements and areas of review for the proposed change.
3.0 TECHNICAL EVALUATION
3.1 Description Unit 1 Cycle 20 Reactor Core STP, Unit 1, is a Westinghouse four-loop pressurized water reactor (PWR) design. The licensee proposed to change the number of full-length control rod assemblies specified in the TS Section 5.3.2 from 57 control rods to 56 control rods for Unit 1 Cycle 20 with no full-length control rod assembly installed in core location D-6. STP units operate for approximately 18 months between refueling cycles.
This amendment supports the STP, Unit 1 Cycle 20 core design. The STP, Unit 1, Cycle 20 core loading consists of all robust fuel assemblies in a checkerboard loading pattern. The design of the STP, Unit 1, reactor including fuel design and control rods is covered in detail in Chapter 4 of the Updated Final Safety Analysis Report (UFSAR).
The removal of the control rod assembly moved the highest worth control rod from core location F-8 to core location C-5 as shown in the licensee's submittal in Figure 1, "Control Rod Locations." The licensee analyzed the effects of the removing the control rod assembly utilizing its reload methodology WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology" (referred to as the Reload Safety Analysis Checklist (RSAC); this document is proprietary) referenced in the STP, Unit 1, TSs. The reload methodology assists in either confirming or changing the Core Operating Limits Report (COLR), which sets a collection of limiting values that bound the safe operation of the core (e.g., Moderator Temperature Coefficients, Shutdown Margin, and Control Rod Insertion Limits). The licensee chose to use the RSAC as a portion of the basis for demonstrating the acceptability of the changes outlined of this LAR. The RSAC was not specifically approved for the purposes of making plant design changes. The LAR addresses a single shutdown rod at the peripheral of the core for one cycle.
The staff reviewed the LAR and the specific circumstances addressed in the LAR while considering the use of the RSAC.
[ _____ __ _
The licensee applied the reload design change process to determine the effects to the nuclear design of the reactor core and whether the values in the COLR needed to change to account for the removed control rod. The rest of this safety evaluation details the effects observed and their impact on the COLR and UFSAR.
Control Rod Drive Mechanism for Rod D-6 In the licensee's submittal, STPNOC stated that STP, Unit 1, normally contains 57 full-length rod cluster control assemblies (RCCAs) divided into four control banks (A, B, C, and D) and five shutdown banks (A, B, C, D, and E) as shown in Figure 1 of the licensee's submittal. When the reactor is at normal power, all the RCCAs are fully withdrawn, except for Control Bank D, which performs short-term control to respond to normal temperature changes. The rest of the control banks assist in reactor startup and normal reactor shutdown. The five shutdown banks provide even more negative reactivity to meet the NRC's shutdown margin (i.e., keep the reactor shutdown with additional safety margin) requirements to ensure the reactor will shut down under abnormal and accident conditions. The shutdown banks are fully withdrawn in accordance with TS 3.1.3.5 and as specified in the COLR.
The removed RCCA, located in core location D-6, is within shutdown bank A as indicated on Figure 1 of the licensee's submittal. Each RCCA has a CROM that uses three electromagnetic coils called the stationary gripper, movable gripper, and lift pole to move the RCCA into and out of the core. The rods are moved in steps by sequencing power to the different electromagnets.
At all points during a rod movement, an electromagnet is holding the control rod up. If the electromagnets de-energize, the rod drops into the core under the force of gravity.
STP, Unit 1, has a unique fourth electromagnetic coil installed in the CROM as part of a Rod Holdout Control System to allow for rapid refueling of the reactor. STP, Units 1 and 2, are the only reactors in the United States that have the capability to lift the reactor vessel head, upper internals, and RCCAs with one polar crane lift during refueling outages. Using the standard three electromagnets was not deemed reliable for the rapid refueling operation, so a separate rod holdout ring acts to mechanically lock the RCCAs in place.
Damage to Rod Holdout Ring The licensee stated in the submittal that deformation of the RCCA holdout ring occurred in 2012 during refueling outage 17. During control rod unlocking operations, RCCA D-6 failed to fully insert into the core when unlatched. The licensee stated that hydraulic forces caused the rod holdout ring to move behind the stationary gripper latches at the same time as the latches engaged the control rod drive shaft. The combination of the gripper latches engaging on the falling control rod shaft deformed the holdout ring. Following this event, RCCA D-6 was tested and performed as designed during operating cycle 18; the damage to the holdout ring previously did not impact the RCCA normal operation. The damage to the rod holdout ring did, however, prevent the RCCA D-6 from being mechanically locked for rapid refueling operations. The licensee instead performed normal refueling operations since the damage to the holdout ring occurred.
In the submittal, the licensee stated that during the Unit 1 refueling outage 19 control rod startup testing, RCCA D-6 was not able to be moved using normal methods. It stated that the licensee was later able to move the control rod to the bottom of the core, which is the fully inserted
- --** - - --*---- - -------- position. If the RCCA 0-6 is not able to be moved using normal methods, then it is not reliable enough to perform its intended function of inserting into the core for normal and emergency operations.
3.2
NRC Staff Evaluation
3.2.1 Extent of Condition The licensee performed inspections on all 57 CROMs during the Unit 1 Cycle 19 refueling outage to determine the extent of the condition and insure that only the 0-6 CROM was affected by a deformed holdout ring. The licensee found that no similar deformation on the other 56 Unit 1 CROMs.
The licensee's inspection of the remaining 56 Unit 1 CROMs provides reasonable assurance that no other CROMs will be affected by a deformed holdout ring. Additionally, normal control rod testing will be re-performed prior to start up. This satisfies the intent of GOC 25 as reactor trip function remains fully capable of inserting the remaining 56 RCCAs. The intent of GOC 26 is also satisfied as appropriate margin remains allowing for single failure malfunctions such as an additional stuck rod.
3.2.2 Structural Evaluation
- 3. 2. 2. 1 Dynamic Analysis Removal of the control rod 0-6 drive shaft and RCCA reduces the weight of the CROM, which can impact the dynamic analyses that predict the stresses in the CROM, reactor vessel, vessel supports, and reactor internals when subjected to seismic or loss-of-coolant-accident (LOCA) excitations.
In the supplement, the licensee stated that the current dynamic analysis of the CROM was performed using its reactor equipment system model, which does not include the weight of the control rod drive shaft in the CROM assembly. Since the model is consistent with the proposed configuration, the NRC staff agrees that the current CROM dynamic stress evaluations, due to seismic and LOCA excitations, remains valid.
Other current reactor vessel, vessel supports, and reactor internals analyses are also unaffected by the mass impacts of removing the Control Rod 0-6 drive shaft and RCCA due to the relatively small change in mass when compared to the mass of the reactor vessel head.
Review of Section 3.9.1 of the UFSAR reveals that the models used in these analyses are lumped-mass models where the mass of the CROMs, including the drive shaft and the RCCA are lumped at the center of gravity of the reactor vessel head. In the submittal, the licensee estimated that the mass of the drive shaft and RCCA is approximately 300 pounds, while the mass of the reactor vessel head is approximately 350,000 pounds. Given the small percentage change (less than 0.09 percent) of the lumped mass to be modeled, the NRC staff concludes that the impact of the weight reduction on the model used to analyze the reactor vessel, vessel supports, and reactor internals is negligible.
- 3. 2. 2. 2 Reactor Coolant Pressure Boundary The NRC staff has reviewed the proposed changes and determined that the reactor coolant pressure boundary is unaffected. The reactor coolant system is unaffected since the changes are all within the reactor coolant pressure boundary and will not result in a change to the pressure rating or operating pressure of the reactor coolant system, nor an increase in the design temperature of the reactor coolant system. The installation of the thimble plug and the flow restrictor are to maintain thermal hydraulic conditions similar to the condition with 57 full-length control rods. The thimble plug and flow restrictor are discussed in Sections 3.2.2.3 and 3.2.2.4 of this safety evaluation.
- 3. 2. 2. 3 Thimble Plug A thimble plug assembly will be installed on the fuel assembly in core location 0-6. Thimble plugs are described in Section 4.2.1.6 of the UFSAR and help to accommodate thermal expansion and positive contact between the fuel assembly and core internals. The UFSAR also states that the thimble plug modifies the flow to acceptable design values. The NRC staff has determined that the use of the thimble plug is acceptable because it is described and analyzed in the UFSAR in Section 4.2.1.6. Thus, the use of the thimble plug is consistent with the design basis as described in the UFSAR.
- 3. 2. 2. 4 Flow Restrictor The licensee's submittal stated that a flow restrictor will be temporarily installed at the top of the 0-6 guide tube. The flow restrictor is intended to preserve the thermal-hydraulic characteristics of the upper internals as hydraulically equivalent to the previous reactor coolant system flow configuration. In the submittal, the licensee stated that the flow restrictor materials and design meet the intent of the American Society of Mechanical Engineers (ASME) Code, Subsection NG, and were analyzed using material properties that were taken from Section II, Part A of the ASME Code to meet the allowable stress limits. The NRC staff has reviewed the flow restrictor design and determined that it is structurally adequate because it meets the allowable stress limits of the ASME Code.
Thermal expansion and material compatibility must also be assessed to provide assurance of the structural adequacy of the added flow restrictor. In the supplement, the licensee stated that the flow restrictor is made of the same grade of stainless steel as used to fabricate the guide tubes to which the flow restrictor is mounted. Because the same material is used, there is no differential thermal expansion between these two components, and material compatibility of the flow restrictor with the fluid conditions in the reactor vessel upper head are assured.
The licensee provided a description of a number of design features and installation procedures in the supplement that provide assurance against the generation of loose parts. The hex bolt that is used for attachment is preloaded to a specified torque, which, in conjunction with the locking fingers, securely locks the flow restrictor to the top of the guide tube. A locking cup, which is tack-welded to the flow restrictor, is crimped onto the hex bolt to prevent loosening due to vibration. The NRC staff has reviewed the design features of the flow restrictor provided in the submittal and concludes that the design features provide assurance that the flow restrictor is securely installed and will not result in the generation of loose parts.
***---- - -- 3. 2. 2. 5 Conclusion The NRC staff concludes that the licensee's proposed removal of the control rod D-6 drive shaft and RCCA, and installation of a flow restrictor and thimble plug in the D-6 location, is acceptable from a structural standpoint because 1) the change does not invalidate the current models used in the dynamic stress evaluations, 2) the change does not affect the reactor coolant pressure boundary, and 3) the added components have been adequately analyzed for structural adequacy and therefore satisfied the intent of GDCs 2, 4, and 14.
3.2.3 Impact on Normal Operations Since RCCA D-6 would normally be fully withdrawn during Modes 1 and 2 according to TS 3.1.3.5 and the COLR, normal operations of the reactor are not significantly impacted. For the fuel assembly located in D-6, thimble plug assemblies (as described by Section 4.2.1.6.4 of the UFSAR) are installed in the guide tubes to restrict bypass flow. The licensee will install a flow restrictor on top of the control rod guide tubes to maintain reactor coolant flow in the upper internals. The licensee is adding a flow restrictor in the upper internals to maintain thermal hydraulic characteristics. The flow restrictor takes the place of the control rod drive shaft for the purpose of fluid dynamics and provides the same flow characteristics in the area of the guide tube.
3.2.3.1 Thermal Hydraulic Analysis The licensee stated that in order to preserve the thermal-hydraulic effects through the fuel assembly located at D-6, thimble plugs will be installed to maintain proper reactor coolant flow through the fuel assembly. The thimble plugs are described in Section 4.2.1.6 of the UFSAR and help to accommodate thermal expansion and positive contact between the fuel assembly and core internals, and limit flow through each occupied thimble to acceptable design values.
The use of the thimble plug is acceptable since the device is already used in other core locations, and is described in the UFSAR in Section 4.2.1.6. Thus, the use of the thimble plug is consistent with the licensee's design basis as described in the UFSAR.
The removal of the control rod drive shaft from the D-6 CROM could also impact rod drop times at other core locations, as well as alter the thermal-hydraulic characteristics locally around the D-6 CROM guide tube. The submittal details the installation of a flow restrictor in the upper guide tube to preserve the thermal-hydraulic characteristics of the upper internals, which is hydraulically equivalent to the previous reactor coolant system flow configuration. The NRC staff concludes that this is acceptable because limiting the flow through the empty guide tube to flow rates similar to what would be achieved through the previous annulus between the drive shaft and the guide tube provides assurance that the rod drop times and thermal-hydraulic characteristics assumed in current safety analyses remain valid.
In the course of the evaluation of the license amendment request, the NRC staff reviewed a Westinghouse Advisory Letter, NSAL-14-5, "Lower than Expected Critical Heat Flux Results Obtained during Departure from Nuclear Boiling Testing." This advisory letter discusses non-conservative critical heat flux calculations discovered during testing. The effect was presented to the NRC staff in the 2015 Westinghouse Fuel Performance Update meeting to make the NRC staff aware of the error. The Westinghouse NSAL was considered as part of the review, however, the staff believes that the NSAL has no impact on the removal of RCCA D-6 and the corresponding re-analysis of departure from nucleate boiling ratio (DNBR) because the non-conservatisms identified will not impact the UFSAR safety analysis.
- 3. 2. 3. 2 Nuclear Design Parameters According to the licensee's COLR and as discussed above, shutdown bank control rods are fully withdrawn from the reactor core during normal operation. As RCCA D-6 is a shutdown bank rod, the absence of RCCA D-6 does not change the nuclear design parameters such as burnup or axial power distributions during normal operations. A core loading pattern change is not necessary. The normal operation of the core and control of core parameters will not be affected since the removed control rod D-6 is normally not used to control core parameters but rather is utilized for shutdown actions.
Removal of RCCA D-6 is acceptable with respect to normal operations as the rod is fully withdrawn during normal full power operations as described above and due to the requirements of TS 3.1.3.5 and the COLR. General Design Criterion 10 is satisfied as the removal of the rod does not affect nuclear design parameters for at-power operation.
3.2.4 Parameters Assumed in Safety Analysis The licensee stated in the submittal that the purpose of the control rods in the shutdown bank is to provide additional negative reactivity to meet shutdown margin requirements and provide reactivity control during accident conditions. A shutdown rod is most important in accident and transient scenarios. With the highest worth rod in the core moving from core location F-8 to C-5 (C-5 is directly adjacent to the removed rod in D-6), the removal of RCCA D-6 has a direct impact on transients and key parameters assumed in safety analyses. The key parameters assumed in transient and accident analyses in Chapter 15 of the UFSAR that are impacted by the removal of RCCA D-6 include:
shutdown margin, boron worth, rod worth of adjacent rods, trip reactivity, and moderator density coefficient.
The impact to these parameters is described below.
- 3. 2. 4. 1 Shutdown Margin Removal of a shutdown bank control rod has a direct impact on the shutdown margin available.
Section 2.3 of the STP COLR provides the shutdown margin limits in order to keep the safety analyses valid. For Modes 1 and 2, the shutdown margin has to be greater than the 1.3 percent delta rho(% /l.p) limit stated in the licensee's COLR. The licensee recalculated the shutdown margin after removal of RCCA D-6, and the value changed from 2.42% ll.p to 2.17% f1p. Given the necessary margin is a minimum, the new shutdown margin value of 2.17% ll.p is greater than the COLR limit for Modes 1 and 2 of 1.3% ll.p, and is, therefore, acceptable.
The licensee described in the submittal that for Modes 3, 4, and 5, Sections 2.3.2 and 2.3.3 of the COLR specify that the shutdown margin shall be greater than the limits specified in Figures 2 and 3. Figures 2 and 3 provide limits for shutdown margin in percent llp for reactor coolant system critical boron concentration in parts per million (ppm) for all rods inserted into the reactor core, or 'all rods in' (ARI), but without the most reactive rod. Critical boron concentration is the amount of boron that would make the reactor critical with ARI minus the most reactive rod and without xenon and samarium effects (these reactivity 'poisons' reduce the amount of reactivity following shutdown). The critical boron concentration value is then used in Figures 2 and 3 of the COLR to determine a minimum required shutdown margin. More boron (a higher boron concentration) is needed in the reactor coolant system to meet the minimum required shutdown margin of the Cycle 20 reactor core.
This relationship of critical boron to minimum required shutdown boron concentration is covered in the license amendment request. The licensee described the lower limit of 1.14 as the minimum required shutdown boron concentration to critical boron concentration with ARI minus the most reactive rod. Figures 2 and 3 of the COLR provide the required shutdown margin values for immediately after the reactor is shutdown (i.e., at hot zero power) through progressively lower modes of operation including refueling mode. To keep the reactor subcritical throughout the different conditions, a minimum shutdown margin of 1.3% llp is required. This requirement is valid even at lower boron concentrations (i.e., when less than 600 ppm of boron is in the primary reactor coolant). To maintain shutdown margin requirements across all boron concentrations, a minimum ratio of 1.14 is required by the UFSAR (the ratio is shutdown margin boron concentration divided by critical boron concentration) to keep the reactor shutdown. Higher ratios of boron concentrations means more shutdown margin and are therefore desirable.
If the shutdown margin limits in the COLR are maintained, the shutdown margin assumptions used in the safety analyses listed in Chapter 15 of the UFSAR remain valid and applicable. This meets the requirements of GDC 10, since acceptable fuel design limits will not be exceeded during normal operation, including the effects of anticipated operational occurrences.
- 3. 2.4. 2 Boron Worth Since the removal of RCCA D-6 has a direct effect on the ARI condition, the licensee evaluated the impact on differential boron worth as a function of boron concentration for ARI. The licensee stated in the submittal that the removed control rod increases the boron worth as a function of boron concentration for the ARI condition. However, the licensee further stated that for the chemical and volume control system (CVCS) malfunction that decreases the boron concentration in the primary coolant, described in Chapter 15 of the UFSAR, the bounding boron worth is calculated assuming all RCCAs are withdrawn from the core. The larger the boron worth, the more reactivity is inserted as a result of the eves malfunction. The maximum boron worth assumed in the various accident evaluations is not impacted by the increased boron worth calculated for the ARI condition. The boron concentration change based on meeting shutdown margin requirements is still within the assumed values for the UFSAR events. The boron worth is not impacted for the analyses in Chapter 15 of the UFSAR. Thus, the impact on boron worth as a result of the removal of RCCA D-6 is acceptable because it stays within the bounds of the assumed values in Chapter 15 of the UFSAR. This satisfies the intent of GDC 11 by maintaining the negative reactivity feedback, and GDC 26 for maintaining a redundant and capable reactivity control system.
3.2.4.3 Rod Worth of Adjacent Rods The licensee determined that the rod worths of the adjacent RCCAs to the removed RCCA 0-6 increased when all RCCAs are inserted. With RCCA 0-6 in the core, the most reactive stuck rod was a control bank rod (Bank A) at core location F-8 with a rod worth of 0.973% llp. When RCCA 0-6 was removed and the re-analysis was completed, the most reactive stuck rod became a shutdown bank rod (Bank 0) at core location C-5 with a rod worth of 1.071 % llp.
This change of rod worth impacted different safety analyses including an RCCA ejection, uncontrolled RCCA bank withdrawal from subcritical or low power startup conditions, and main steam line break from zero power. The acceptability of the impact will be discussed in the respective events in Section 3.2.5, "Impact on Accident Analyses" of this safety evaluation.
- 3. 2. 4. 4 Trip Reactivity The trip reactivity as a function of rod insertion position decreased due to the removal of RCCA 0-6. This impacts trip reactivity as a function of time once RCCAs begin to insert on a trip signal. The UFSAR states the assumed minimum trip reactivity based on rod position in Figure 15.0-4 and the negative reactivity insertion based on time in Figure 15.0-5. Table 3 of the submittal details a change in trip reactivity values comparing the original values, modified values, and assumed UFSAR values. The values decreased slightly with the removal of RCCA 0-6, with no change greater than 0.1 % llp over the course of the rod position. No values approached the limit outlined in Figure 15. 0-4 and Figure 15. 0-5 of the UFSAR. The new trip reactivity values are acceptable because they are still bounded by the assumed values in Chapter 15 of the UFSAR. This satisfies the intent of GOC 10 by keeping an appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. This satisfies the intent of GOC 26 by keeping the assurance that the control rods will control reactivity under conditions of normal operation and anticipated operational occurrences. This satisfies GDC 28 by maintaining the rate of reactivity insertion in postulated reactivity accidents.
Each of the following reactivity insertion accidents has key safety parameters as described in the RSAC that are compared to the assumed values in the UFSAR safety analysis to determine if the conclusions stated in the UFSAR are still bounding. The only key safety parameter impacted for these events was the assumed trip reactivity. As explained in this section above, the trip reactivity decreased, however the values used in the safety analysis of the UFSAR were still bounding compared to the new values. The NRC staff concludes that the licensee's analysis of the events below are bounded by the UFSAR limits and are, therefore, acceptable.
Reactivity insertion accidents bounded by UFSAR limits (the event number refers to Table 3.2 in this safety evaluation):
Turbine Trip (Event 8) which is the bounding event for:
o Loss of External Electrical Load (Event 7) o Loss of Condenser Vacuum and Other Events Causing Turbine Trip (Event 10) o Inadvertent Closure of Main Steam Isolation Valves (Event 9)
L -
Complete Loss of Forced Reactor Coolant Flow (Event 15), which is the bounding event for:
o Partial Loss of Forced Reactor Coolant Flow (Event 14) o Loss of Non-Emergency AC Power to the Plant Auxiliaries (Loss-of-Offsite-Power) (Event 11)
Reactor Coolant Pump Shaft Seizure (Locked Rotor) (Event 16), which is the bounding event for o
Reactor Coolant Pump Shaft Break (Event 17)
Rod Cluster Control Assembly Mis-operation (Dropped Rod) (Event 20)
Inadvertent Opening of a Pressurizer Safety or Relief Valve (Event 27)
Steam Generator Tube Rupture (Event 29)
Anticipated Transient Without SCRAM (Event 32)
- 3. 2. 4. 5 Moderator Density Coefficient The maximum positive moderator density coefficient (MDC) is calculated assuming that all RCCAs are inserted into the core. Figure 15.0-6 in the UFSAR has a limit of 0.54 Llk/(g/cm3) used for the safety analyses in Chapter 15 of the UFSAR. With RCCA D-6 in the core, the MDC is 0.3911 Llk/(g/cm3). Without RCCA D-6 in the core, the MDC is 0.3892 Llk/(g/cm3). As detailed in Section 4.3.2.3.2 in the UFSAR, the effect of control rods is to make the moderator temperature (and density) coefficient more negative by reducing the required soluble boron concentration and by increasing the "leakage" from the core. This effect is why the maximum positive MDC is calculated with all rods in. To determine the least positive MDC, it follows that all rods out condition would be more conservative as removing the more negative effect of the control rods would drive the MDC to the least positive. Since the least positive MDC is assumed with all rods out, the least positive MDC in Figure 15.0-6 is not affected by the removal of RCCA D-6.
The new value for most positive MDC is still conservative as it is below the assumed most positive MDC used in the UFSAR Chapter 15 safety analysis and is acceptable. This satisfies the intent of GDC 10 by maintaining the conservative assumptions in accident analyses. This also satisfies GDC 11 as the core maintains its inherent negative reactivity feedback.
The Feedwater System Pipe Break described in Section 15.2.8 of the UFSAR is impacted by the change in most positive MDC. This event, listed as Event 13 in Table 3.2 below, includes key safety parameters as described in the RSAC that are compared to the assumed values in the UFSAR safety analysis to determine if the conclusions stated in the UFSAR are still bounding. The only key safety parameter impacted in this event was the assumed most positive MDC. As explained above in this section, the most positive MDC decreased; however, the values used in the safety analysis of the UFSAR still bounds the new most positive MDC value.
Based on the above, the NRC staff concludes that the licensee's analysis of the Feedwater System Pipe Break is consistent with the UFSAR and is, therefore, acceptable.
3.2.5 Impact on Accident Analyses Table 1, in the licensee's submittal, details UFSAR Chapter 15 events that the licensee reviewed for impacts due to removal of RCCA 0-6. The licensee's table provided the event number, UFSAR section, description of the event, and the licensee's comment on the accident impact. The NRC staff created a similar Table 3.2 below adding the applicable RSAC section and a staff determination on the licensee's statements concerning the impacts due to removal of RCCA 0-6.
The NRC staff reviewed the licensee's statement in the LAR, the referenced NRC-approved reload methodology in RSAC, and the relevant Chapter 15 UFSAR sections. The NRC staff compared the impacted parameters stated in the license amendment request with the corresponding key safety parameters in the reload methodology to verify that the parameters for the event were addressed and that they were bounded by the UFSAR. The details are presented below.
Table 3.2: Impact on UFSAR Chapter 15 Accident Analysis 1
15.1.1 5.3.9 2
15.1.2 5.3.9 3
15.1.3 5.3.9 I
Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature Bounded by Feedwater Malfunction See Section 3.2.5. 1 of this increasing Feedwater Flow (15.1.2). safety evaluation.
Feedwater System Malfunctions Causing an Increase in Feedwater Flow Most positive moderator density coefficient and trip reactivity remains bounding. All other analysis parameters not impacted.
Excessive Increase in Secondary No impact. Core DNBR limits are Steam Flow not challenged by this event.
See Section 3.2.5.1 of this safety evaluation.
The "Excessive Increase in Secondary Steam Flow" is consistent with the evaluation in the UFSAR. The event is recalculated in the reload design change process as noted in 15.1.3.2 of the UFSAR.
The impact of the most positive MDC is still captured by the original accident. See Section 3.2. 5.10 of this safety evaluation.
The UFSAR event points to 15.1.5 for analysis of the entire Inadvertent Opening of a Steam event. No impact is made to 4
15.1.4 5.3.14 Generator Relief or Safety Valve Bounded by spectrum of steam UFSAR evaluation and the
' Causing Depressurization of the system piping failures (15.1.5).
licensee's statement is Main Steam System acceptable. See Section 3.2.5.2 of this safety evaluation.
Spectrum of Steam System Shutdown margin remains 5
15.1.5 5.3.14 Piping Failures Inside and bounding. MDC remains See Section 3.2.5.2 of this Outside Containment acceptable. All other analysis safety evaluation.
parameters remain acceptable.
Based on the UFSAR section, Steam Pressure Regulatory this event is not applicable to 6
15.2.1 N/A Malfunction or Failure that Not applicable to STP STP and the licensee's Results in Decreasing Steam comment is acceptable. See Flow Section 3.2.5.10 of this safety evaluation.
7 15.2.2 5.3.7 Loss of External Electrical Load Bounded by turbine trip (15.2.3)
See Section 3.2.4.4 of this safety evaluation.
Event UFSAR WCAP-9272 Description Submittal Comment NRG Staff Determination Number Section Section Based on the UFSAR section and RSAC, the key safety parameters assumed in the UFSAR are still bounded despite the impact of the Trip Reactivity remains bounding, removed rod. The trip reactivity 8
15.2.3 5.3.7 Turbine Trip all other analysis parameters not and moderator density impacted.
coefficient used in Section 15.2.3 are still bounded.
The licensee's statement is acceptable. See See Section 3.2.4.4 of this safety evaluation.
Inadvertent Closure of Main Bounded by Loss of Condenser See Section 3.2.4.4 of this 9
15.2.4 5.3.7 Steam Isolation Valves Vacuum, which is bounded by safety evaluation.
Turbine Trip (15.2.5, 15.2.3)
Loss of Condenser Vacuum and See Section 3.2.4.4 of this 10 15.2.5 5.3.7 Other Events Causing Turbine Bounded by turbine trip (15.2.3) safety evaluation.
Trip Bounded by turbine trip (UFSAR Loss of Non-Emergency AC Section 15.2.3) for DNBR, and Loss See Section 3.2.4.4 of this 11 15.2.6 5.3.8 Power to the Plant Auxiliaries of normal feedwater flow (UFSAR, safety evaluation.
Section 15.2. 7) for pressurizer overfill
12 15.2.7 5.3.8 13 15.2.8 5.3.15 14 15.3.1 5.3.5 15 15.3.2 5.3.5 Loss of Normal Feedwater Flow Feedwater System Pipe Break Analiysis parameters not impacted.
Most positive moderator density coefficient remains bounding. All other analysis parameters not impacted.
The trip reactivity assumed in the UFSAR Section 15.2. 7 analysis is not challenged by the change in the trip reactivity.
The licensee's statement is acceptable. See Section 3.2.5.10 of this safety
.evaluation.
The most positive moderator density coefficient is still bounding. The licensee's statement is acceptable. See Section 3.2.4.5 of this safety evaluation.
Partial Loss of Forced Reactor Coolant Flow Bounded by complete loss of forced S S ct' 3 2 4 4 f th' ct I t fl (UFSAR ee e ion... o 1s rea or coo an ow
~ t 1 t' Section 15.3.2) sa.e y eva ua ion.
Com lete Loss of Forced Reactor Trip Reactivity ~emains bounding.
c 1 p t Fl All other analysis parameters not oo an ow impacted.
The assumptions in the UFSAR are not challenged. The licensee's statement that trip reactivity is bounding and other analysis parameters not impacted is acceptable. See Section 3.2.4.4 of this safety evaluation.
11r*
EV8M IJFSAR INCAP-9212 I
J
-~
~~
~Stall~ ' I
~
Seotlon Section 11*1 r.rlfll I
.;I The assumptions in the UFSAR are not challenged. The Trip Reactivity remains bounding, licensee's statement that trip 16 15.3.3 5.3.16 Reactor Coolant Pump Shaft all other analysis parameters not reactivity is bounding and other Seizure (Locked Rotor) impacted.
analysis parameters not impacted is acceptable. See Section 3.2.4.4 of this safety I
evaluation.
17 15.3.4 5.3.16 Reactor Coolant Pump Shaft Bounded by RCP Shaft Seizure See Section 3.2.4.4 of this Break (UFSAR Section 15.3.3) safety evaluation.
Trip reactivity remains bounding.
I Uncontrolled RCCA Bank AH other analysis parameters not 18 15.4.1 5.3.1 withdrawa:1 from sub-critical or low impacted. RCCA D-6 is in a See Section 3.2.5.3 of this power startup condition shutdown bank which is assumed safety evaluation.
withdrawn from the core for this analysis.
Trip reactivity remains bounding.
Most positive MDC remains bounding. All other analysis Uncontrolled RCCA Bank parameters not impacted. Control See Section 3.2.5.5 of this 19 15.4.2 5.3.2 withdrawal at power Rod D-6 removal does not impact safety evaluation.
the assumed reactivity rate because it is in a shutdown bank which is withdrawn from the core at power conditions The assumptions in the UFSAR Trip reactivity and other analysis are not challenged. The 20 15.4.3 RCCA Misalignment (Dropped parameters remain bounding.
licensee's statement that trip 5.3.3 Change in reactivity insertion due to reactivity is bounding and other Rod) removal of RCCA D-6 bounded by analysis parameters not current analysis impacted is acceptable. See Section 3.2.4.4 of this safety evaluation.
Startup of an Inactive Reactor Trip Reactivity remains bounding.
21 15.4.4 5.3.6 Coolant Loop at an Incorrect Most positive MDC remains See Section 3.2.5.5 of this Temperature bounding. All other analysis safety evaluation.
parameters not impacted.
CVCS Malfunction resulting in Required shutdown margin limits 22 15.4.6 5.3.4 decreased boron concentration in remain bounding. Boron worth See Section of 3.2.5.6 of Safety reactor coolant (Boron Dilution) remains bounding. All other Evaluation analysis parameters not impacted.
The assumptions in the UFSAR No impact. Inadvertent loading is Not in Inadvertent Loading of a Fuel are not challenged. The 23 15.4.7 WCAP-9272 Assembly into an Improper detected using in-core licensee's statement that there Position instrumentation when the shutdown is no impact is acceptable. See banks are withdrawn Section 3.2.5.10 of this safety evaluation.
rr 111*
Event UFSAR WCAP-9272 o.c:ripllon
&lblnltlal cc.nmant I
NRo Stair DetetmJnatJon Number 8*;ticm Section
~
Analysis parameters not impacted.
The core with RCCA D-6 removed would remain subcritical with an ejected rod and an additional RCCA Spectrum of RCCA Ejection with the highest worth not inserted See Section 3.2.5. 7 in Safety 24 15.4.8 5.3.17 into the core. Shutdown Bank Accidents RCCAs are withdrawn from the Evaluation core to maximize the power I increase due to these events, therefore, removal of RCCA D-6 has no impact The assumptions in the UFSAR No impact. The UFSAR concludes are not challenged. The 25 15.5.1 5.3.11 Inadvertent Operation of the "Spurious SI [safety injection]
I licensee's statement that there ECCS during Power Operation without immediate reactor trip has is no impact is acceptable. See no effect on RCS."
Section 3.2.5.10 of this safety evaluation.
- Trip reactivity remains bounding.
I 26 15.5.2 N/A eves Malfunction that Increases Most positive MDC remains See Section 3.2.5.8 in Safety Reactor Coolant Inventory bounding. All other parameters not Evaluation impacted The assumptions in the UFSAR are not challenged. The Trip reactivity remains bounding. All licensee's statement that trip 27 15.6.1 5.3.10 Inadvertent Opening of a other analysis parameters not reactivity is bounding and other Pressurizer Safety or Relief Valve impacted.
analysis parameters not impacted is acceptable. See Section 3.2.4.4 of this safety evaluation.
28 15.6.2 29 15.6.3 30 15.6.5 31 15.7 Not in F~ilure of Small Lines Carrying WCAP-9272 Primary Coolant Outside Containment Not in WCAP-9272 Steam Generator Tube Rupture 5.3.13 LOCA {Large and Small) from Spectrum of Pipe Breaks Not in Radioactive Release From a WCAP-9272 Subsystem or Component
.Analysis parameters not impacted Trip reactivity remains bounding.
AH other analysis parameters not impacted.
None of the parameters impacted are referenced in the UFSAR. The licensee's statement is acceptable. See Section 3.2.5.10 of this safety evaluation.
The assumptions in the UFSAR are not challenged. The licensee's statement that trip reactivity is bounding and other
' analysis parameters not impacted is acceptable. See Section 3.2.4.4 of this safety evaluation.
Analysis parameters not impacted.
Impacted of change in RCS liquid See Section 3.2.5.9 of Safety volume is negligible.
Evaluation Analysis parameters not impacted.
None of the parameters impacted are referenced in the UFSAR. The licensee's statement is acceptable. See Section 3.2.5.10 of this safety evaluation.
32 15.8 WCAP-8440 and WCAP-8330 ATWS WCAP-8440 and WCAP-8330 Trip reactivity remains bounding.
t
- t d d t
- All other analysis parameters not arectn~.timpac e dan n~
ct d All d
t MTC t
rea 1v1 y assume remains
~mpa e
- ro s ou no bounding. The licensee's impacted by removal of Control Rod t t t
- t bl s 0 _6 s a emen 1s accep a e.
ee Section 3.2.4.4 of this safety evaluation.
3.2.5.1 Feedwater System Malfunctions: Increase in Feedwater Flow and Reduction in Feedwater Temperature The licensee stated in its submittal that the most positive moderator density coefficient (MDC) has an impact on the assumed values for feedwater system malfunctions. The assumed most positive MDC is still bounding for the analyzed event as detailed in Section 3.2.4.5 of this safety evaluation. The limiting trip reactivity has not been impacted and the analyses in the UFSAR remain valid and bounding. Therefore, the licensee has properly accounted for the impact of the changed key safety parameter specified in RSAC and the UFSAR Sections 15.1.1 and 15.1.2. The licensee's statement that most positive MDC and trip reactivity remains bounding and all other analysis parameters are not impacted is acceptable. Based on the above, the NRC staff concludes that the licensee's analysis of the impacts to the feedwater system malfunction events is consistent with the design basis as described in the UFSAR and is, therefore, acceptable.
3.2.5.2 Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System and Spectrum of Steam System Piping Failures Inside and Outside Containment The "Spectrum of Steam System Piping Failures Inside and Outside Containment" event bounds inadvertent valve openings (see Table 3.2 above, Event Number 4) and steam line breaks described in UFSAR, Sections 15.1.4 and 15.1. 5. This event assumes the additional single failure of the Worst Stuck Rod (WSR or N-1) stuck in the top of the CROM with the analysis of the accident. With the removal of RCCA D-6, the assumption changes to Worst Stuck Pair (WSP or N-2); in other words, two rods are assumed to be fully out of the core. The WSR identification looks at three scenarios:
- 1.
The highest reactivity worth RCCA is fully withdrawn,
- 2.
The stuck RCCA is in the assembly location with highest fuel pin peaking (fuel section with the highest heat load) with its RCCA withdrawn, and
- 3.
The stuck RCCA in the assembly location with the lowest inlet temperature.
The first two scenarios are affected by the change to the WSP assumption; therefore, the licensee fully reanalyzed the accident and concluded that the limiting parameters for the event are still bounded with respect to the DNBR or the peak cladding temperature parameters.
The licensee addressed the impact on the steam line break accident in the "Rod Worth" section of the submittal. The removed rod impacts local power distributions around the removed rod since the highest worth rod is the one adjacent to the removed RCCA D-6. Since peaking factors change, the power distributions change, and the licensee performed a new DNBR analysis. The DNBR decreases from 3.011 to 1.811 with an analysis limit of 1.495 stated in the submittal.
The licensee explained in its December 9, 2015, supplement that the decrease in DNBR was due to the increase in power in the local region of the core due to the removed rod. DNBR is defined as predicted critical heat flux divided by actual heat flux. The predicted critical heat flux calculated was not impacted by the removal of the rod; however, the local calculated power increased. There are other factors, such as the radial and axial power distributions, that partially offset the impact of core power increases. The increase in power caused a decrease in DNBR; however, the value remains bounded by the safety analysis limit of 1.495, and the UFSAR Section 4.4.1.1 limit of 1.43, and is, therefore, acceptable.
The licensee stated that the assumptions used in the UFSAR analysis for steam line break analysis remain acceptable for the new core configuration. Specifically the shutdown margin value remains bounding. The MDC and trip reactivity are assumed in the UFSAR. The licensee performed analyses using the methodology in the RSAC for the MDC calculated for steam line break accidents. The NRC staff confirmed that the licensee's analysis in the submittal and confirmed that they remain bounded by the safety limits in the UFSAR. The calculation method ensures that the MDC assumed in the transient analysis remain applicable. The licensee stated that for steam line break accidents, the MDC remains acceptable with RCCA D-6 removed.
Based on the above, the NRC staff has determined that the licensee's methodology to evaluate the impacts to the spectrum of steam demand increase events is consistent with the design basis as described in the UFSAR, and is, therefore, acceptable.
- 3. 2. 5. 3 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition The licensee covers in the "Rod Worth" section of its submittal that the removal of RCCA D-6 does not impact the "Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition" analyzed in Section 15.4.1 of the UFSAR. The UFSAR assumes that the event happens at hot zero power where the shutdown banks are withdrawn. Thus the removal of RCCA D-6 does not impact the results of the RCCA accident since the analysis already assumes this bank of control rods are out of the core. The trip reactivity is impacted, but as discussed in Section 3.2.4.4 of this safety evaluation, the trip reactivity is bounded. The licensee's statement that the trip reactivity remains bounding and all other analysis parameters are not impacted, is acceptable because the assumptions in Section 15.4.1 of the UFSAR are still valid. Based on the above, the NRC staff has determined that the licensee's analysis of the Uncontrolled RCCA Bank Withdrawal form a Subcritical or Low Power Startup Condition event is consistent with the design basis as described in the UFSAR, and is, therefore, acceptable.
- 3. 2. 5. 4 Uncontrolled RCCA Bank Withdrawal at Power In the "Most Positive Moderator Density Coefficient" section, the licensee stated that the "Uncontrolled RCCA Bank Withdrawal at Power" event has an assumption that could be impacted. The assumptions stated in the UFSAR state that the Uncontrolled RCCA Bank Withdrawal at Power event is analyzed with a maximum and minimum reactivity feedback. The maximum reactivity feedback is impacted since the most positive MDC changed. However, the values assumed in the UFSAR have not been changed. The licensee's statement that the trip reactivity and most positive MDC are bounding is acceptable as the safety analysis has not been challenged by the changing values with RCCA D-6 removed. Based on the above, the NRC staff has determined that the licensee's analysis of the impacts to an Uncontrolled RCCA Bank Withdrawal at Power event is consistent with the design basis as described in the UFSAR, and is, therefore, acceptable.
- 3.2.5.5 Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature In the "Most Positive Moderator Density Coefficient" section of the submittal, the licensee stated that the "Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature" event is impacted by the assumed most positive MDC. However, the MDC assumption stated in the UFSAR and RSAC is still bounding for the event. The NRC staff finds the licensee's statement that trip reactivity and MDC assumed in the analysis of the UFSAR is still valid and the conclusions reached in the UFSAR remain true. Based on this discussion, the NRC staff has determined that the licensee's analysis of the impacts to the Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature are consistent with the design basis as described in the UFSAR, and are, therefore, acceptable.
- 3. 2. 5. 6 CVCS Ma/function Resulting in Decreased Boron Concentration in Reactor Coolant (Boron Dilution)
The licensee detailed the impact of shutdown margin, boron worth, and most positive moderator density coefficient in the respective sections of the submittal. The licensee stated that boron worth and shutdown margin limits are bounding and the other analysis parameters are not impacted. As detailed above, the changed values do not change the assumed values in the UFSAR analyses. Therefore, the licensee's statement is acceptable as the assumptions and results of the UFSAR analysis has not changed. Based on this discussion, the NRC staff has determined that the licensee's analysis of the impacts to a Boron Dilution Event is consistent with the design basis as described in the UFSAR, and is, therefore, acceptable.
- 3. 2. 5. 7 Spectrum of RCCA Ejection Accidents The licensee detailed in the "Rod Worth" section of the submittal that the RCCA Ejection accident is impacted, notably how hot zero power conditions are impacted. As detailed in Section 15.4.8 of the UFSAR, the shutdown banks are withdrawn at hot zero power and this fact is enforced by the rod insertion limits in the COLR. The licensee's comment that the analysis parameters are not impacted is acceptable as the changes in rod worth do not impact the analysis parameters covered in the UFSAR methodologies WCAP-7588 Revision 1-A "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods" and WCAP-9272. Based on this discussion, the NRC staff has determined that the licensee's analysis of the impacts to the Spectrum of RCCA Ejection Accidents is consistent with the design basis as described in the UFSAR, and is, therefore, acceptable.
3.2.5.8 CVCS Ma/function that Increases Reactor Coolant Inventory The licensee details that the event "CVCS Malfunction that Increases RCS Inventory" has an assumption related to the most positive MDC in the submittal. The trip reactivity is an input to Section 15.5.2 of the UFSAR and the licensee stated in the submittal that trip reactivity is impacted due to removal of RCCA D-6, as well. Despite the changing values, the assumptions used in the safety analysis remain bounding. The licensee's statement that trip reactivity and most positive MDC remain bounding and all other analysis parameters remain bounding is acceptable because it does not change the results of the UFSAR analysis. Based on the discussion above, the NRC staff has determined that the licensee's analyses for impacts to the CVCS Malfunction that Increases Reactor Coolant Inventory are consistent with the design basis as described in the UFSAR, and are, therefore, acceptable.
- 3. 2. 5. 9 Loss of Coolant Accident (Large and Small) Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary The Large Break (LB) LOCA accident does not credit any control rod inventory in the STP evaluation model per UFSAR 15.0.5. Therefore, the removal of RCCA D-6 does not have an impact on any portions of the current Analysis of Record (AOR) for LB LOCA.
The Small Break (SB) LOCA does credit control inventory in the STP evaluation model per UFSAR 15.0.5. The reactivity insertion rate and total scram reactivity as function of total RCCA insertion position and time is a key input to the SB LOCA evaluation model. Scram reactivity is one the major terminating actions in the event that occurs on low pressurizer level. The reactivity insertion rate has a direct impact on two-phase level swell and, as such, the maximum peak cladding temperature for the event. The reactivity insertion rate supplied by the licensee in Table 3 of LAR submittal shows the insertion rate with and without D-6 RCCA inserted as part of the SCRAM. Both rates are well bounding of the SB LOCA reactivity insertion rate for the STP AOR by factor nearing 2. Therefore, the current STP SB LOCA AOR remains bounding, and all other portions of the evaluation model are unaffected by removal of RCCA D-6. Based on the discussion above, the NRC staff has determined that the licensee's evaluation for these events acceptable since the licensee is consistent with the design basis as described in the UFSAR.
3.2.5.10 UFSAR Chapter 15 Events Not Impacted Specific events in the UFSAR were not impacted by the changes described in Section 3.2.4 of this SE. These events include:
Excessive Increase in Secondary Steam Flow (Event 3)
Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (Event 6)
Loss of Normal Feedwater Flow (Event 12)
Inadvertent Loading of a Fuel Assembly into an Improper Position (Event 23)
Inadvertent Operation of the Emergency Core Cooling System During Power Operation (Event 25)
Failure of Small Lines Carrying Primary Coolant Outside Containment (Event 28)
Radioactive Release From a Subsystem or Component (Event 31)
Each of these events has a set of key safety parameters as described in the RSAC that the NRC staff compared to the assumed values in the UFSAR safety analysis to determine if the conclusions stated in the UFSAR are still bounding. None of the key safety parameters nor assumptions in the RSAC were impacted by the changes stated within Section 3.2.4 of this SE and thus remain consistent with the design basis. Based on the discussion above, the NRC staff has determined that these events are not impacted by the proposed change and are, therefore, acceptable.
3.3 Technical Evaluation Conclusion
The WCAP-9272 for the Reload Safety Analysis Checklist (RSAC) process and the respective safety evaluation indicate that this process was designed to evaluate cycle-to-cycle loading patterns and minor fuel design changes. This process was not intended for use in dispositioning events and margin assessments for LARs related to plant configuration changes.
The NRC staff has reviewed the applicability of the RSAC for use in justifying the specific circumstances addressed in this LAR (e.g., Cycle 20 of STP Unit 1, the location of the rod in the core, the rod worth, the moderator density coefficient and the impact on the accident analyses).
The licensee does not intend for the removal of the RCCA in location D-6 to be a permanent plant configuration. As such, a full scope update of the licensee UFSAR Chapter 15 events is not warranted or needed to make a safety determination for one operating cycle. Additional plant design changes under different circumstances for different cycles may require additional staff review.
The NRC staff concludes that the licensee's proposed use of 56 control rod assemblies in STP, Unit 1, is acceptable because the design change is consistent with the current design basis and does not challenge the safety analyses detailed in Chapter 15 of the UFSAR. The design change will not result in modification to the Unit 1 Cycle 20 COLR beyond a revision number and a note detailing that RCCA D-6 has been removed. The staff concludes that the licensee used methods consistent with regulatory requirements and guidance identified in Section 2.0 above. The staff also concludes, based on the considerations discussed above, that (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 Commission's regulations, and (3) No operational changes are required with this TS change. Therefore, the proposed change is acceptable.
4.0 EMERGENCY CIRCUMSTANCES The NRC's regulations in 10 CFR 50.91 (a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 1 O CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.
In its submittal, the licensee requested that the amendment be treated as an emergency amendment since the situation could not have been predicted or avoided. The licensee stated that during the previous operating cycle, the D-6 CROM was monitored during monthly testing which indicated normal operation. Additionally, during a previous attempt to start up Unit 1 reactor following refueling activities were completed, the D-6 control rod operated acceptably during rod testing (subsequently, Unit 1 had to cool down to repair reactor coolant pump seal leakage). It was after these pump repairs were completed and the plant was heated and pressurized, that the licensee stated that CROM testing revealed D-6 as unreliable. Unit 1 was again cooled down and depressurized to remove the reactor head and allow inspection of the D-6 CROM. Once it was determined that the CROM could not be repaired or replaced without extensive planning and manufacture of special tooling as well as fabrication of mockups to test the tooling, the licensee requested an emergency amendment to allow it to operate Unit 1 without RCCA D-6. This will allow normal operation for one cycle - approximately 18 months -
while repairs plans are completed.
The NRC staff reviewed the licensee's basis for processing the proposed amendment as an emergency and agrees that an emergency situation exists. The licensee could not have foreseen the RCCA D-6 malfunction since it was operating normally during testing and was able to perform its intended function of dropping into the core.
The NRC staff agrees that due to the complex nature of a CROM replacement, an operation that has not been done at a nuclear power plant in the United States, it will take several months for STPNOC to make plans for repair or replacement of the CROM. It is not unreasonable to allow Unit 1 to operate during this planning phase since Unit 1 is able to operate safely, as concluded above, with the removal of the RCCA D-6.
The NRC staff concludes that the licensee's actions were reasonable, and that the identification of the problem, the actions of the licensee to inform the NRC staff, and the emergency amendment was provided in a reasonable timeframe. Further, the NRC staff agrees that the situation could not have been avoided. Finally, the NRC staff concludes that failure to issue this license amendment would result in prevention of the resumption of operation for STP, Unit 1,.
The NRC staff agrees that an emergency situation exists consistent with the provisions in 10 CFR 50.91 (a)(5). The NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91 (a)(5). Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the emergency provisions of 10 CFR 50.91 (a)(5).
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
Pursuant to 10 CFR 50.91 (a)(5), if the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in prevention of resumption of operation or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. As noted in Section 4.0 of this safety evaluation, the NRC staff has concluded that an emergency situation does exist, in that failure to issue the amendment would result prevention of STP, Unit 1, to resume operation. Therefore, a final finding of no significant hazards consideration follows.
The Commission has made a final determination that the amendment request involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
As required by 10 CFR 50.91 (a), the licensee provided its analysis of the issue of no significant hazards consideration, in its letter dated December 3, 2015, as supplemented by letter dated December 9, 2015. The licensee's analysis is presented below:
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No. Operation of STP, Unit 1, Cycle 20 with Control Rod D-6 removed will not involve a significant increase in the probability or consequences of an accident previously evaluated. STPNOC has evaluated the reactivity consequences associated with removal of Control Rod D-6 and determined that the amount of shutdown margin would be reduced but remains bounded by the limit provided in the Core Operating Limits Report. The impact on adjacent control rod worth was also evaluated and there is an increase in the rod worth of the most reactive stuck rod assumed in some accident analyses; however, the Updated Final Safety Analysis Report (UFSAR) accident analysis limits continue to be met. The probability of occurrence of a previously evaluated accident is not impacted by removal of Control Rod D-6. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No. Operation of STP, Unit 1, Cycle 20 with Control Rod D-6 removed will not create the possibility of a new or different kind of accident from any accident previously evaluated. To preserve the reactor coolant system flow characteristics in the reactor core, a flow restrictor will be installed at the top of the D-6 guide tube housing and a thimble plug will be installed on the fuel assembly located in core location D-6.
Installation of these components will not prevent the remaining 56 control rods from performing the required design function of providing adequate shutdown margin. No new operator actions are created as a result of the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response: No. Operation of STP, Unit 1, Cycle 20 with Control Rod D-6 removed will not involve a significant reduction in a margin of safety. The margin of safety is established by setting safety limits and operating within those limits. The proposed change does not alter a UFSAR design basis or safety limit and does not change any setpoint at which automatic actuations are initiated. STPNOC has evaluated the impact of the proposed change on available shutdown margin, boron worth, rod worth, trip reactivity as a function of time, and the most positive moderator density coefficient; the results of these evaluations show that the proposed change does not exceed or alter a design basis or safety limit.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and has determined that this amendment involves no significant hazards determination.
6.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendment on December 4 and December 10, 2015. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation and use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
8.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) the amendment does not (a) involve significant increase in the probability or consequences of an accident previously evaluated or, (b) create the possibility of a new or different kind of accident from any previously evaluated or, (c) involve a significant reduction in a margin of safety and, therefore, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (3) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (4) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: L. Regner, NRR/DORL J. Dean, NRR/DSS/SNPB W. MacFee, NRR/DSS/SRXB G. Thomas, NRR/DSS/SRXB I. Tseng, NRR/DE/EMCB M. Hamm, NRR/DSS/STSB Date: December 11, 2015
ML15343A128 Sincerely,
/RA/
Lisa M. Regner, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDssStsb Resource RidsNrrLAJBurkhardt Resource RidsNrrPMSouth Texas Resource RidsRgn4MailCenter Resource MHamm, NRR/DSS/STSB
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NAME LRegner JBurkhardt RElliott JDean DATE 12/11/15 12/11/15 12/11/15 12/11/15
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