ML14325A069

CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report

ML14325A069
Person / Time
Site:	Sequoyah
Issue date:	11/17/2014
From:	John Carlin Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
Download: ML14325A069 (11)

Text

Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 November 17, 2014 10 CFR 50.54 10 CFR 72.48 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034

Subject:

10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report In accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), enclosed is the Sequoyah Nuclear Plant, Units 1 and 2, Summary Report regarding the implemented changes, tests, and experiments for which evaluations were performed in accordance with 10 CFR 50.59(c) and 10 CFR 72.48(c). The evaluations occurred since the previous submittal dated July 5, 2013.

There are no commitments contained in this letter. If you have any questions concerning this issue, please contact Erin K. Henderson at (423) 843-7170.

Enclosure 10 CFR 50.59, and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report cc (Enclosure):

NRC Regional Administrator- Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant

ýi4h7 I,1 L (L

o. " 55

ENCLOSURE SEQUOYAH NUCLEAR PLANT 10 CFR 50.59 AND 10 CFR 72.48 CHANGES, TESTS, AND EXPERIMENTS

SUMMARY

REPORT

Design Change Notice (DCN) DESCRIPTION SAFETY ANALYSIS Design Change Notice DESCRIPTION SAFETY ANALYSIS (DCN) 22497 DCN 22497 will re-gear the actuators for containment The re-gearing of the containment sump isolation valves sump isolation valves 1-FCV-63-72 and 1-FCV-63-73. 1-FCV-63-72 and 1-FCV-63-73 changes the maximum This will allow each valve to develop additional allowable stroke time of the valves from 45 seconds to 120 torque/thrust that is needed to overcome the expected seconds. Analysis shows this is acceptable and will have no differential pressure across the valve. This re-gearing adverse affect on emergency core cooling system (ECCS) results in a longer opening time for these valves. The operations. Sufficient net positive suction head (NPSH) will spring packs will be modified or replaced to preclude still be available to the residual heat removal (RHR) pumps any potential issues associated with hydraulic lock as during the switchover from ECCS injection to ECCS a result of grease infiltrating the spring pack. DCN recirculation. For the limiting loss of coolant accident (LOCA),

22497 will ensure compliance to the NRC Generic the ECCS sump water inventory present during the switchover Letter 89-10/96-05 program including the new Joint from injection to recirculation mode, at the time suction is first Owners Group Motor Operator Valve Program taken from the sump, will continue to conservatively meet all requirements. ECCS flow requirements and will preclude unacceptable vortexing. The time to accomplish operator actions to complete the remainder of the switchover to the recirculation phase of ECCS core cooling is not affected by this change.

22688 Add main control room (MCR) handswitches to allow This activity adds the capability to manually open the DGB the operator to remotely open the Diesel Generator intake dampers from the MCR. This change does not impact Building (DGB) HVAC dampers. These dampers the frequency of any accident. Based upon the fact that need to be open prior to a tornado event to prevent opening the dampers is a simple action performed from the damage to the HVAC ductwork and/or dampers to MCR on a tornado watch, this change does not change the ensure availability of the diesel generator. Normal likelihood of any safety systems, structures and components operation of the dampers is not impacted. The (SSC) malfunction. This change does not impact the updated final safety analysis report (UFSAR) does not consequences of any accident or malfunction, does not impact currently describe the operation of the system except any fission product barriers, and does not alter any analysis in Section 3.3.2.2, "Inaddition, the DGB and the method. There will be no increase in the likelihood of essential raw cooling water (ERCWV) pumping station accidents, malfunctions, or consequences due to this change; depressurize due to the vent areas provided by the therefore, this activity does not require NRC approval.

ventilation openings in those buildings." The operation of the dampers is shown on UFSAR Figure 8.3.1-14 (schematic diagram). Based on analysis performed for problem evaluation report (PER) 217787, the dampers are required to be open to prevent damage due to a tornado depressurization.

E-1

Design Change Notice DESCRIPTION SAFETY ANALYSIS (DCN) Cont.

D23085 This design change removes the existing kirk key Implementing D23085A does not create any need for licensing interlocking scheme from the feeder breakers and tie amendment or additional NRC requirements. The breakers for ERCW motor control center (MCC) 1A-A, malfunctions remain bounded by the UFSAR and there exists 1 B-B, 2A-A and 2B-B and replaces the existing a less than minimal increase in the occurrence of an accident.

exterior rotary style breaker handle and breaker mechanism on the main and alternate feeds with a model that will properly function with the replacement breakers. The physical control of the ERCW MCC feeder breakers is replaced with administrative controls. The existing ERCW feeder breakers are obsolete. The replacement breakers have been evaluated through the equivalency process (PEG 1071185AO), but the replacement breaker's physical footprint has slight variations that prevents the existing kirk key interlocking scheme to be mounted onto the breaker.

D23209 The purpose of DCN 23209 is to replace the failed The installation of the replacement 125V Vital Battery Charger 150A-rated 125V Vital Battery Charger I. The I can be implemented without prior NRC approval because the replacement charger has an output rating of 200A. modification has no impact on the performance of the 125V (Note: Although the new charger is equipped with a Vital Battery System during any of the UFSAR Chapter 15 larger output capacity, the output current will be design basis accidents. The new charger introduces no new limited to that of the remaining 15A-rated 125V Vital credible failure modes that result in an increase in the Battery Chargers II, Ill, and IV). The new charger is likelihood of an accident or malfunction or result in increased procured as a Class 1 E, safety related device and is consequences of an accident or malfunction.

seismically qualified for installation in the Category I structure. All existing design functions for the replacement charger described in Sequoyah Nuclear Plant design criteria SQN-DC-V-1 1.2 "125V Vital Battery System" and the UFSAR will be met.

E-2

Design Change Notice DESCRIPTION SAFETY ANALYSIS (DCN) Cont.

D23216 The proposed activity will modify the Auxiliary "C" The change being made to the CCPs, AFW pumps, and PZR handswitches for the centrifugal charging pumps heaters removes an automatic function that is causing (CCP), auxiliary feedwater pumps (AFW), and the additional operator burden during MCR abandonment for an pressurizer heater (PZR) to support 10 CFR 50 Appendix R fire. The automatic function is already being Appendix R fire safe shutdown analysis. No design defeated. by the operator pulling the control power fuses functions or accident condition functions to these because the automatic actuation of these components causes components were affected with this modification while FSSD mitigation strategy problems. This modification allows the transfer switch is in the normal position. This the operator to maintain control of the FSSD mitigation modification impacts only the Fire Safe Shutdown strategy by using the handswitches located on the 6.9kV (FSSD) mitigation strategy during a 10 CFR 50 Shutdown board. The design functions of the pumps and Appendix R fire that results in the abandonment of the heaters will remain the same. This modification does not affect MCR. The circuit changes affect only the portions of the normal plant operation from the MCR, only during MCR the circuit that are active when the transfer switch is in abandonment. There will be no increase in the likelihood of AUX position. The design function of the transfer accidents, malfunctions, or consequences due to this change; switch is to isolate the MCR portion of the circuit therefore, this activity does not require NRC approval.

during an Appendix R fire. I__

E-3

PROCEDURE DESCRIPTION SAFETY ANALYSIS E-3 R(21) Procedure E-3, "Steam Generator Tube Rupture", Based upon the design of the reactor head vent system, the implements the steam generator tube rupture proposed change does not result in a greater likelihood of any mitigating actions that are summarized in UFSAR accident or malfunction of an SSC. This change does not Section 15.4.3.2. This revision modifies Step 24 for result in more severe radiological consequences from any isolating the centrifugal charging pump injection tank design basis accident or malfunction. This change is bound (CCPIT) flow path to incorporate additional by existing UFSAR accident analyses and the UFSAR contingency actions in the event that the CCPIT flow descriptions of head vent system malfunctions. This path cannot be isolated from the MCR. evaluation concludes this change does not require prior NRC approval.

FHI-19 R(4) FHI-19, "Fuel Handling Abnormal Incident Procedure," The proposed procedure change is not part of the UFSAR is being revised to add section 5.8 for manually described procedure. The procedure affects the hoist on the lowering a fuel assembly on the Manipulator Crane manipulator crane and is only used for moving fuel that is upon the loss/failure of the hoist. typically performed during refueling or forced outages. The proposed change to FHI-1 9 allows for the power to be removed from the manipulator crane hoist and added steps to raise/lower a fuel assembly to a safe location via a motor hand wheel. There is not more than a minimal frequency of occurrence, more than a minimum likelihood of occurrence of a malfunction, more than a minimum increase in consequences of a malfunction, or creation of a different type of accident, or malfunction of a SSC important to safety.

E-4

Temporary Alteration DESCRIPTION SAFETY ANALYSIS Control Form (TACF)

SQN-0-2013-011 The 125-volt Battery Charger I is ineffective and This modification is safe from a nuclear standpoint and does needs to be replaced as documented in PER 708758. not require prior NRC approval since this change does not With Battery Charger I out of service, Units 1 and 2 increase the frequency of an accident discussed in this safety will shutdown if either spare 125-volt direct current evaluation or create any new failures not previously evaluated.

(DC) Charger 1-S (currently used as the main charger Also, the design function of the system is maintained with no for Channel I) or 125-volt DC Battery Charger II additional radiological release to plant personnel or the becomes inoperable or not energized (Units 1 and 2 general public.

technical Specification (TS) Section 3.8.2.3).

Therefore, this temporary modification will provide a solution to eliminate this vulnerability and add a new backup charger for Channels I and II. Normally, spare Charger 1-S is the backup charger for Channels I and II.

COLR DESCRIPTION SAFETY ANALYSIS Ul COLR The transition to the Advanced W17 HTP (HTP) fuel It is shown that the change does not constitute a departure involves a change in the departure from nucleate from a method of evaluation. Use of the Biasi correlation does boiling (DNB) correlation employed in one of the not constitute a departure because (1) the Biasi correlation is safety analyses. For HTP assemblies in the main approved by the NRC specifically for this type of accident steam line break (MSLB) analysis, the Biasi analysis and (2) the Biasi correlation was used under the correlation is applied over the full length of the HTP terms, conditions, and limitations of the NRC approval.

assembly, replacing the BHTP correlation above the lower most spacer grid and the BWU-N correlation below the lower most spacer grid. The Biasi correlation is approved by the NRC for application in the MSLB for the thermal-hydraulic conditions that occur throughout the core during DNB limiting periods of the MSLB accident, including the lowest coolant pressure reached during the transient and including conditions that occur in the region below the lower most spacer grid.

E-5

SAR DESCRIPTION SAFETY ANALYSIS FSAR Table 5.2.1-1 PERs 422244 and 666088 document that the number The evaluation for this UFSAR Table 5.2.1-1 change of loading and unloading power changes for the determined that it does not increase the frequency of the replacement steam generators (RSGs) was not occurrence of an accident; nor does it increase the likelihood identified as being affected by the replacement of the of the occurrence of a malfunction; nor does it increase the Unit 1 and Unit 2 steam generators in DCN D20672A consequences of either an accident or a malfunction; nor does and DCN D22478A. The number of loading and it create the possibility of an accident of a different type; nor unloading power changes the Unit 1 and Unit 2 RSGs does it create the possibility of a malfunction with a different were designed should have been reduced as a result result; nor does it exceed or alter any design basis limits for of these DCNs, and the screening reviews should any fission product barriers. As such, no License Amendment have evaluated the change. or NRC approval is required.

LTR-CRA-02-219 R1 This is a change to the fuel handling accident (FHA) The FHA analysis of record for a FHA inside containment has analysis of record for a FHA inside containment. The changed. The mixing volume was changed from 5 percent to volume of primary containment being credited for 50 percent of the containment free volume consistent with the mixing is being changed from 5 percent to 50 percent. requirements in Regulatory Guide 1.183. In addition, the The containment purge isolation time is being duration of containment purge was increased from 30 seconds changed from 30 seconds to 300 seconds. to 300 seconds to coincide with the control room isolation and maximize dose to the control room for the scenario. As a result of these two changes the containment FHA analysis of record control room dose changed from 4.2 rem to 3.9 rem, an approximately 7 percent decrease, this is not more than a minimal increase in the consequences of a FHA.

PER 748496 The evaluation captures ramifications of steady state Removing the 10 minute criteria from UFSAR, Section 8.2.2 estimator including eliminating the 10 minute does not create any adverse impacts. SQN remains bounded notification requirement from Transmission Operations by GDC-1 7 and NERC requirements. In addition, to the site. TVA-SPP-10.010 "NERC Standard Compliance Processes Shared By TVA's Nuclear Power And Energy Delivery Organizations" complies with NUC-001 requirements. TVA is still compliant with NERC I SERC and GDC-17.

TVA-SPP-10.010 states that the Transmission Operator has 15 minutes to regain adequate voltage and power; if adequate voltage and power cannot be achieved, then the Transmission Operator must notify Site Operations within 15 minutes. As such, no license amendment or NRC approval is required.

E-6

SAR (Cont.) DESCRIPTION SAFETY ANALYSIS Penetrations As a result of PER 800432 and LERs 05000327/2013- The evaluation concluded the plant processes and procedures 004-00 and 05000327/2013-004-01, a prudent action will ensure the safety functions of the penetrations are is being taken to review the site's ability to comply with maintained during refueling. The evaluation concludes that the requirements specified in LCO 3.9.4. This review the proposed activity would not result in an unacceptable examines the aspects of having a containment release of radioactive material from the containment penetration open with fuel movement in progress, atmosphere to the outside atmosphere. It is concluded there relative to the requirements of LCO 3.9.4. A is a no more than a minimal increase in the consequence of description of the maintenance instruction is provided, any fuel handling accident.

along with the penetration identifiers that this instruction addresses. The Operations procedures verifies the containment penetrations are in their proper alignment before fuel movement is also described. The Operations procedure ensures that the manual action is in place to close any penetration isolation valves that require closure. The evaluation examines the design functions of the containment penetrations during the time period of applicability of LCO 3.9.4, refueling. These design functions are minimal, and it is stated containment and its penetrations are not required to withstand the normal containment design pressure during any event while refueling. The only analyzed accident during refueling is a fuel handling accident. The failure modes of this activity are that any of the involved penetrations might not be isolated if the need arises.

E-7

SAR (Cont.) DESCRIPTION SAFETY ANALYSIS T/S Bases 3/4.5.3 This safety evaluation is being performed for a change The potential design basis accident involved is a LOCA that to the Bases of Technical Specification Limiting might occur with the Unit in Mode 4. The RHR system may or Condition for Operation (LCO) 3.5.3 for Sequoyah may not be in service for shutdown cooling at the onset of the Units 1 and 2, with a corresponding change to UFSAR LOCA. A LOCA in Mode 4 is somewhat unique in the Section 6.3.2.2. These changes are being made as a technical specifications as it is recognized that the probability result of difficulty experienced in previous outages in of the LOCA is much lower than when the systems are fully completing leak tests of the reactor coolant system pressurized, and also the core decay heat is much less.

(RCS) pressure isolation valves, particularly the RHR Therefore, the technical specifications allow suspension of primary and secondary check valves, within the one single failure criteria. Only one train of ECCS is required to be hour time allowed by the LCO. Note 2 under the operable to mitigate the event. The analysis of the Mode 4 action section of LCO 3.5.3 allows the required RHR LOCA essentially allows the RHR injection valves to be in any subsystem to be inoperable for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in order to position for the duration of the event. Therefore, there really perform leak testing of these valves. The basic are no credible failure modes relating to this event. As such, change that is proposed is to allow one train of the no license amendment or NRC approval is required.

cold leg injection lines to be out of service for testing during Mode 4. This means that cold leg injection flow would be into 2 of the 4 RCS cold legs. A new paragraph is added to the applicable safety analysis section. This paragraph states that one train of ECCS injecting into two cold legs provides sufficient flow for core cooling. Under the LCO section it is repeated that either of the two cold leg injection flow paths may be isolated for testing in Mode 4. A new reference has been added. This reference is from an NRC SER pre-licensing in which it is concluded that a large break LOCA is not credible in Mode 4.

E-8

DOCUMENT DESCRIPTION SAFETY ANALYSIS NUMBER/72.48 EVALUATION TRACKING NUMBER PER 452027 This activity will revise Independent Spent Fuel The cask heat load for MPC-103 was originally calculated, storage Installation Final Safety analysis Report using the wrong methodology, to < 28.74kW and thus (FSAR) Table 1.2.2 to allow MPC-103 to remain in a backfilled in accordance with the requirements of the condition where the helium backfill pressure is in certificate of compliance (CoC), technical specifications and accordance with the backfill requirements of a cask FSAR for the calculated heat load. When the heat load was with a heat load 28.74kW, as defined in FSAR Table recalculated using the correct methodology given by the 1.2.2 (a 29.3 psig and < 48.5 psig) FSAR, it was determined that the heat load was > 28.74kW, and therefore the MPC did not meet the applicable backfill requirements. The CoC contains required actions to bring the MPC back into compliance in the event the backfill requirements are not met. These actions include an analysis, using the models and methods from the FSAR, to demonstrate that the limits for cask components and contents will be met during normal long-term operation, short-term operation, off-normal operation, and any postulated accident condition. After completion of those actions, MPC-103 was in an analyzed condition that proved it meets the applicable acceptance criteria set forth in the FSAR, and that it will prevent the occurrence of an unsafe condition that would threaten the health and safety of the public during all normal long-term operation, short-term operation, off-normal operation, and postulated accidents.

E-9