ML083250324
| ML083250324 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/18/2008 |
| From: | James Smith Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| S64 08119 800 | |
| Download: ML083250324 (9) | |
Text
S64 081119 800 November 18, 2008 10 CFR 50.59 10 CFR 72.48 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT - UNITS 1 AND 2 - 10 CFR 50.59, AND 10 CFR 72.48 CHANGES, TESTS, AND EXPERIMENTS
SUMMARY
REPORT The purpose of this letter is to provide the summary report of the implemented safety evaluations, performed in accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48. The evaluations occurred since the previous submittal dated June 12, 2008. There were no 10 CFR 72.48 evaluation performed in this timeframe.
If you should have any questions, please contact me at (423) 843-7170.
Sincerely, Original signed by James D. Smith Manager, Site Licensing and Industry Affairs Enclosure
U.S. Nuclear Regulatory Commission Page 2 November 18, 2008 JDS:JWP:SKD Enclosure cc (Enclosure):
Mr. Brendan Moroney, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G-9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 G. Arent, EQB 1B-WBC W. R. Campbell, LP 3R-C T. P. Cleary, OPS 4A-SQN C. R. Church, POB 2B-SQN T. Coutu, LP 3R-C D. E. Jernigan, LP 3R-C K. R. Jones, OPS 4A-SQN M. J. Lorek, LP 3R-C L. E. Nicholson, LP 3R-C M. A. Purcell, LP 4K-C L. E. Thibault, LP 3R-C S. A. Vance, ET 10 A-K G. E. Vickery, OPS 4A-SQN B. A. Wetzel, BR 4X-C E. J. Vigluicci, ET 10A-K T.J. Bradshaw, NSRB Support, BR 4X-C WBN Site Licensing Files, ADM 1L-WBN EDMS, WT CA-K (Re: B38 081117 806)
I: License TS/ FSAR 21/ FSAR 21-5059 Summary report
ENCLOSURE SEQUOYAH NUCLEAR PLANT 10 CFR 50.59 AND 10 CFR 72.48
SUMMARY
REPORT
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 DCN
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS D-21831-A This modification involves the removal of an original The original canister type penetration was replaced to address an adverse canister type containment electrical penetration trend of increased leakage through the penetration. The adverse trend device and replacement with a modular type resulted from equipment aging issues associated with the original penetration assembly for Containment Penetration penetration assembly. Replacement of the assembly eliminated the No. X-128E on Sequoyah Unit 2. observed leakage and restored the penetration integrity.
To support the interface with the new modular Evaluation of the replacement penetration assembly confirmed compliance penetration, changes to the electrical cable with the original component design standards (i.e., ASME Section III - 1986 connections and the fusing for the electrical loads and IEEE 317-1983). Also, the electrical interfacing changes were passing through the penetration were also evaluated against the existing design basis requirements and were found to implemented. be acceptable in terms of component and system qualification.
Performance of the penetration replacement was limited to the core empty period of a refueling outage when containment integrity requirements support the in-process configuration.
E-22316-A This modification established a revised setpoint This change was implemented to address the tendency of the nominal relief tolerance for relief valves which provide piping valve setpoint to drift downward during the established component test system overpressure protection for a portion of interval. The nominal setpoint for each of the affected relief valves is 220
- 1) the centrifugal charging pump (high head) suction psig. Each valve functions to relieve system pressure in the event the piping and 2) the safety injection pump (intermediate pump suction isolation valves experience small amounts of in-leakage from head) suction piping. The original tolerance for the the connected higher pressure piping systems when closed. The design spring loaded relief valves was +/- 3 percent of the capacity of the relief valves is approximately 20 to 25 gallons per minute nominal relief setpoint for both as-found and as-left (gpm).
test conditions. This change established a revised setpoint tolerance of +3 percent,-5 percent for as- Evaluation of this change focused on the effects of the lower valve found conditions. The original +/-3 percent setpoint actuation pressure afforded by the -5percent setpoint tolerance. The tolerance was not changed for as-left conditions. evaluation concluded that 1) sufficient margins exist between the range of normal operating pressures in the affected piping and the revised relief valve lower setpoint to prevent spurious operation of the relief valves during normal operation and 2) the valve reset (blowdown) pressure following actuation (the valves reseat at approximately 10 percent of the nominal lift pressure) has no effect on system operation of functional capabilities. The change did not affect the overpressure protection function of the relief valves.
Since the change has no effect on the system or component functional capabilities and is consistent with the component Code of Record, the change does not adversely affect the safe operation of the plant.
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SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 PROCEDURE
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS 0-SI-OPS-068-137. This change involves a revision to the primary This procedure requires the collection of measured RCS data and 0, R22 system water inventory surveillance procedure. The contains the data reduction requirements to quantify RCS leakage in revision involves a change to two of the three accordance with plant Technical Specification Surveillance equations used to determine total leakage from the Requirement No. 4.4.6.2.1. It is used as a surveillance tool to quantify reactor coolant system (RCS). Calculation of the RCS leakage and to backup the leakage detection systems located total RCS leakage value under this procedure inside of the containment building polar crane wall. This procedure is involves the primary system makeup rate, volume not relied upon to detect Reactor Coolant Pressure Boundary (RCPB) control tank (VCT) leakage rate, pressurizer leakage of 1 gpm in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less, as specified in Regulatory leakage rate, and RCS temperature correction. The Guide 1.45. That function is provided by the leakage detection systems changes involve normalization of two of the inputs monitoring RCPB external leakage located inside of the polar crane for total leakage to an established reference wall which include containment humidity, charging pump operation and temperature. The first change alters the RCS excessive makeup volume detection, sump level detectors, and temperature correction factor such that the radiation monitors.
volumetric leak rate is calculated using the density of water at 70 degrees F and atmospheric pressure The surveillance procedure changes are designed to reduce the (instead of using the average density of water standard deviation of the primary inventory leakage calculations. The based on RCS normal operating temperature and changes are based on the best industry practices and are consistent pressure). The second change alters the with the recommendations of the Pressurized Water Owners Group pressurizer leakage rate equation to use the actual (PWROG) as documented in Topical Report No. WCAP-16423-NP, pressurizer pressure instead of the primary system Revision 00.
design pressure (2235 psig). These changes allow for a more direct comparison of the leak rates that The procedure revisions improve the accuracy of the primary system make up the total system leakage. leakage surveillance calculations and do not negatively impact nuclear safety.
The changes affect the unidentified leakage rate values as total RCS leakage is one of the two components used to quantify unidentified leakage.
The changes do not affect the identified RCS leakage calculations.
0-SO-202-1, R12 This activity involves a one time only change Evaluation of the one time only start bus alignment involved 1) review (OTOC) (OTOC) to the system operating instruction for the of the modified loading on CSST B, 2) review of power source plant 6.9kV start buses. The procedure contains independence and 3) review of the postulated failure of one of the instructions for 1) energizing the start buses from active CSSTs.
available sources, 2) taking a start bus out of service for maintenance, modification or testing, 3) The CSST B load review confirmed that several large unit board motor transferring power supplies to the normal start loads aligned to Start Bus 2A would be removed from service and buses and 4) transferring power supplies to the DC administratively controlled (locked out) during the modified alignment.
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SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 PROCEDURE
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS control bus on the start buses. Review of the remaining loads applied to the CSST B in this configuration for both normal and accident operation confirmed that Under normal conditions, Start Bus 1A and Start they were within the capacity of the CSST B winding such that an Bus 2A are fed from common service station overload condition would not occur.
transformer (CSST) A and Start Bus 1B and Start Bus 2B are fed from CSST C. The CSST design is As a result of the one time procedure change, Start Bus 1B and 2A such that a loss of a CSST (either A or C) will were aligned to the Y winding of CSST B while the 2B start bus was generate a fast transfer to CSST B. Breakers are aligned to the X winding of CSST C and Start Bus 1A was aligned to electrically interlocked such that 1) only one start the X winding of CSST A. In this configuration, the 1A and 1B start bus is aligned to each CSST B winding and 2) only buses remain powered from separate CSST transformers as well as one start bus power train (either the A or B) is the 2A and 2B start buses. This configuration maintained independent aligned to CSST B at a time. These interlock power sources to the trained start buses under expected operating features prevent overloading of the windings on conditions.
CSST B for normal operating loads.
If failure of the either CSST A or CSST B were postulated to occur such To minimize the risk to Unit 1 operation during the that off-site power is lost to one of the Unit 1 power trains, off-site CSST C maintenance outage which was conducted power can be restored to the affected power train by manual operator during the Unit 2, Cycle 15 refueling outage (Spring alignment of the 6.9kV shutdown board normal/alternate feeder 2008), Start Bus 1B was aligned to CSST B. In breakers.
support of planned maintenance activities, a transfer of Start Bus 2A from its normal feeder Based on these results, the loading on CSST B remains within the breaker to its alternate feeder breaker was also functional capabilities of the component. One immediate source of off-required. This transfer placed Start Bus 2A on the site power remains available for the 1A and the 1B start buses as well Y winding of CSST B at the same time the 1B as one delayed source (based on manual operator action if failure of Start Bus was aligned to the same winding. This the credited CSSTs is postulated) in accordance with the requirements one time only procedure change permitted the of IEEE-308 and General Design Criterion (GDC)-17. As such, this one interlock in the control circuit of the alternate feeder time only CSST alignment change had no effect on nuclear safety.
breaker for Start Bus 2A to be bypassed to allow CSST B to carry both Start Bus 2A and Start Bus 1B on the Y winding during the maintenance activity.
This configuration also prevented the automatic or manual transfer of Start Bus 1A and Start Bus 2B to the X winding of CSST B.
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SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 UFSAR
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS Section 2.4.12 This activity involves a revision to the UFSAR which Section 2.4.12 of the UFSAR evaluates the ability of the surface waters evaluates the dilution and disbursement effects of near the plant to dilute and disburse a number of postulated accidental radiological effluents released to the surface waters radioactive liquid releases. The evaluation confirms that the level of in proximity to the plant. The revision is required to activity at the first downstream surface water intake will be below address changes in river flow control operations 10CFR20 limits. The evaluation is performed in accordance with the associated with a downstream dam which result in a requirements of 10CFR50, Appendix I. In this evaluation, the river flow reduction in the river flow rate assumed in the is specified as the flow which is equaled or exceeded approximately 50 original evaluation. percent of the time. Based on recent changes to the operational practice for flow through a downstream dam, the average river flow consistent with this criterion has been reduced from 29,000 cubic feet per second (cfs) to 27,474 cfs.
The reduction in the credited river flow was incorporated into the effluent disbursement and dilution analysis. Revised dilution values were calculated using the existing evaluation methodology and model.
The calculation established the revised concentration for a continuous plane source release as 3.4E-11 mCi/gm. There was no change in the instantaneous plane source release concentration.
The revised continuous plane source release concentration value remains well below the acceptable limit of 1.0E-9 mCi/gm for a liquid effluent release of Iodine-131.
Section 9.4.2.2.6 This activity involves a revision of the environmental Section 9.4.2.2.6 of the UFSAR contains a general description of the conditions described for the shutdown transformer shutdown transformer room ventilation system. The environmental rooms. The current UFSAR indicates that the design criteria for these rooms conservatively establish the minimum ventilation system for these rooms is designed to room temperature as 15 degrees F. A review plant records since initial maintain a minimum room temperature of 60ºF. operation indicate that the lowest recorded temperature within these This change revises the description to indicate that rooms is 24 degrees F.
the ventilation system is designed to maintain the room temperature within a range of 15°F to 97°F The subject rooms contain the 480V shutdown transformers. Review of consistent with ventilation equipment functional the transformer design indicates that the limiting low temperature capabilities and services equipment requirements. consideration is the transformer oil. The pour point of the transformer oil is established as -50 degrees F and represents the temperature at which the oil will not flow. There is no concern with moisture entering the transformer since there is a positive pressure nitrogen blanket over the transformer. An oil temperature of 15 degrees F has been established to be acceptable based on a pour temperature limit of E-4
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 UFSAR
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS
-50 degrees F.
Since the minimum 15 degrees F design basis room temperature is conservative with respect to actual conditions and the functional capabilities for the transformers are maintained with the 15 degrees F minimum temperature limit, the proposed change to the UFSAR is consistent with the established design basis. The change to the UFSAR represents a correction to be consistent with the design and has no effect on nuclear safety Section 15.3.1 The emergency core cooling system (ECCS) pump The ECCS pumps are designed to provide emergency core cooling and performance capabilities assumed in the plant reactivity control in the event of high energy piping breaks or other accident and transient analyses have been revised transients which involve primary system inventory losses or reactivity to 1) reduce the minimum developed head values insertions. Evaluation of the minimum pump performance assumption for the charging (high head) and safety injection changes resulted in a complete re-analysis of the small break LOCA (intermediate head) pumps by 5 percent and 2) transient. For the revised ECCS pump performance characteristics, the increase the maximum charging pump developed calculated peak fuel cladding increased from the previous value of head values for pump flows which exceed 250 gpm. 1162 degrees F to 1403 degrees F. This result remains well below the These changes have been previously made to the established analysis acceptance criteria of 2200 degrees F. The realistic large break loss-of-coolant accident (LOCA) re-analysis also confirmed that the revised ECCS performance analyses documented in Sequoyah Technical assumptions are sufficient to demonstrate compliance with the balance Specification Change Requests TS-07-04 and TS- of the small break LOCA analysis acceptance criteria.
08-01. The large break LOCA analysis was the limiting design basis transient for minimum ECCS The evaluation of the balance of the plant transients and accidents pump performance. Application of the realistic concluded the current analyses of record are either 1) not affected by analysis methodology to the large break LOCA the changes, 2) bounded by existing conservatisms built into the transient demonstrated acceptable results with the analysis or 3) bounded by crediting the full capability of the ECCS revised ECCS pump performance characteristics. which was conservatively minimized in the analysis of record through This activity extends the applicability of the revised the use of simplifying assumptions.
performance assumptions to the balance of the plant transient and accident analyses. Based on this result, the existing analyses were established to be conservative and bounding for the revised ECCS performance assumptions.
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SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 Work Order
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS 99-007763-000 Under this work order, one of the essential raw Evaluation of this activity included review of the actual lifting operation cooling water (ERCW) pumping station traveling as well as the maneuvering of surface craft near the ERCW pumping water screens was dissembled and replaced. The station. The evaluation concluded:
disassembly operation consisted of raising the entire screen assembly and removing one section at 1. The screen lifting operation placed loads in excess of 2100 lbs a time as required for replacement. Re-assembly above safety-related equipment. As such, lifting of the traveling was accomplished by reversing the process. The screens was performed in accordance with all the requirements of lifts were performed by a crane supported on a NUREG-0612 for a heavy load lift.
barge parked adjacent to the ERCW pumping station. The traveling screen sections were placed 2. Design and operation of the barge and tug boat were consistent on the barge and were moved to and from a staging with applicable industry standards. Controls and restrictions placed area. The following administrative controls were on the maneuvering of the craft are conservative and effective in applicable to the work order. preventing impact with the ERCW pumping station.
- 1. All tornado missile protection roof panels 3. Blocking flow to the ERCW intakes due to the postulated sinking of remained in place during the load lifts except for the barge/tug has been evaluated and is not credible. A trench those permitted to be removed by prior analysis. exists immediately in front of the intakes. The width of the barge/tug is greater than the trench and the slope of the trench
- 2. Load lifts were not permitted during a tornado walls is steep such that it is not possible for the submerged barge watch, warning, or any other periods of high to block the intakes.
winds.
As such, nuclear safety was not affected by this work activity.
- 3. The barge was not permitted to be positioned near the ERCW pumping station when work was not in-progress or during any flood conditions.
- 4. The barge and tug boat were not anchored during the lifts. The stationary position was better maintained by using the vessel motors and provided the ability to move the barge/tug quickly.
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