ML083250324
| ML083250324 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/18/2008 |
| From: | Smith J D 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 t he 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
C. R. Church, POB 2B-SQN
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 E-1 D-21831-A
This modification involves the removal of an original canister type containment electrical penetration
device and replacement with a modular type
penetration assembly for Containment Penetration
No. X-128E on Sequoyah Unit 2.
To support the interface with the new modular
penetration, changes to the electrical cable
connections and the fusing for the electrical loads
passing through the penetration were also
implemented.
The original canister type penetration was replaced to address an adverse trend of increased leakage through the penetration. The adverse trend
resulted from equipment aging issues associated with the original
penetration assembly. Replacement of the assembly eliminated the observed leakage and restored the penetration integrity.
Evaluation of the replacement penetra tion assembly confirmed compliance with the original component design standards (i.e., ASME Section III - 1986
and IEEE 317-1983). Also, the electr ical interfacing changes were evaluated against the existing design bas is requirements and were found to be acceptable in terms of com ponent and system qualification.
Performance of the penetration replacem ent was limited to the core empty period of a refueling outage when c ontainment integrity requirements support the in-process configuration. E-22316-A This modification established a revised setpoint tolerance for relief valves which provide piping
system overpressure protection for a portion of
- 1) the centrifugal charging pump (high head) suction
piping and 2) the safety injection pump (intermediate
head) suction piping. The original tolerance for the
spring loaded relief valves was +/- 3 percent of the
nominal relief setpoint for both as-found and as-left
test conditions. This change established a revised setpoint tolerance of +3 percent,-5 percent for as-
found conditions. The original +/-3 percent setpoint
tolerance was not changed for as-left conditions.
This change was implemented to address the tendency of the nominal relief valve setpoint to drift downward during the established component test interval. The nominal setpoint for each of the affected relief valves is 220 psig. Each valve functions to relieve system pressure in the event the pump suction isolation valves experi ence small amounts of in-leakage from the connected higher pressure piping systems when closed. The design capacity of the relief valves is appr oximately 20 to 25 gallons per minute (gpm).
Evaluation of this change focused on the effects of the lower valve actuation pressure afforded by the -
5percent setpoint tolerance. The 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 spuri ous operation of the relief valves during normal operation and 2) the valve reset (blowdown) pressure following
actuation (the valves reseat at approx imately 10 percent of the nominal lift pressure) has no effect on system operat ion of functional capabilities. The change did not affect the overpressure protection function of the relief valves.
Since the change has no effect on t he system or component functional capabilities and is consistent with the component Code of Record, the change does not adversely affect t he safe operation of the plant.
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 PROCEDURE
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS E-2 0-SI-OPS-068-137.
0, R22 This change involves a revision to the primary system water inventory surveillance procedure. The
revision involves a change to two of the three
equations used to determine total leakage from the
). Calculation of the total RCS leakage value under this procedure
involves the primary system makeup rate, volume
control tank (VCT) leakage rate, pressurizer
leakage rate, and RCS temperature correction. The
changes involve normalization of two of the inputs
for total leakage to an established reference
temperature. The first change alters the RCS
temperature correction factor such that the
volumetric leak rate is calculated using the density of water at 70 degrees F and atmospheric pressure (instead of using the average density of water
based on RCS normal operating temperature and
pressure). The second change alters the
pressurizer leakage rate equation to use the actual
pressurizer pressure instead of the primary system
design pressure (2235 psig). These changes allow
for a more direct comparison of the leak rates that
make up the total system leakage.
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.
This procedure requires the collection of measured RCS data and contains the data reduction require ments to quantify RCS leakage in accordance with plant Technical Specification Surveillance
Requirement No. 4.4.6.2.1. It is us ed as a surveillance tool to quantify RCS leakage and to backup the l eakage detection systems located inside of the containment building pol ar crane wall. This procedure is not relied upon to detect Reactor C oolant Pressure Boundary (RCPB) 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 le ss, as specified in Regulatory Guide 1.45. That function is prov ided by the leakage detection systems monitoring RCPB external leakage located inside of the polar crane
wall which include containment hum idity, charging pump operation and excessive makeup volume detection, sump level detectors, and
radiation monitors.
The surveillance procedure changes are designed to reduce the standard deviation of the primary inv entory leakage calculations. The changes are based on the best industr y practices and are consistent with the recommendations of the Pressurized Water Owners Group (PWROG) as documented in T opical Report No. WCAP-16423-NP, Revision 00.
The procedure revisions improve the accuracy of the primary system leakage surveillance calculations and do not negatively impact nuclear safety.
0-SO-202-1, R12 (OTOC) This activity involves a one time only change (OTOC) to the system oper ating instruction for the plant 6.9kV start buses.
The procedure contains instructions for 1) energizing the start buses from
available sources, 2) taking a start bus out of
service for maintenance, modification or testing, 3)
transferring power supplies to the normal start
buses and 4) transferring power supplies to the DC Evaluation of the one time only start bus alignment involved 1) review of the modified loading on CSST B, 2) review of power source independence and 3) review of the postulated failure of one of the active CSSTs.
The CSST B load review confirmed that several large unit board motor loads aligned to Start Bus 2A would be removed from service and
administratively controlled (locked out) during the modified alignment.
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 PROCEDURE
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS E-3 control bus on the start buses.
Under normal conditions, Start Bus 1A and Start
Bus 2A are fed from common service station
transformer (CSST) A and Start Bus 1B and Start
Bus 2B are fed from CSST C. The CSST design is
such that a loss of a CSST (either A or C) will
generate a fast transfer to CSST B. Breakers are
electrically interlocked such that 1) only one start
bus is aligned to each CSST B winding and 2) only
one start bus power train (e ither the A or B) is aligned to CSST B at a time. These interlock
features prevent overl oading of the windings on CSST B for normal operating loads.
To minimize the risk to Unit 1 operation during the CSST C maintenance outage which was conducted during the Unit 2, Cycle 15 refueling outage (Spring 2008), Start Bus 1B was aligned to CSST B. In support of planned maintenance activities, a transfer of Start Bus 2A from its normal feeder breaker to its alternate feeder breaker was also required. This transfer placed Start Bus 2A on the "Y" winding of CSST B at the same time the 1B Start Bus was aligned to the same winding. This one time only procedure change permitted the interlock in the control circuit of the alternate feeder 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.
Review of the remaining loads applied to the CSST B in this configuration for both normal and accident operation confirmed that they were within the capacity of the CSST B winding such that an overload condition would not occur.
As a result of the one time procedure change, Start Bus 1B and 2A were aligned to the "Y" winding of CSST B while the 2B start bus was aligned to the "X" winding of CSST C and Start Bus 1A was aligned to the "X" winding of CSST A. In this configuration, the 1A and 1B start buses remain powered from separate CSST transformers as well as the 2A and 2B start buses. This configuration maintained independent power sources to the trained start buses under expected operating conditions.
If failure of the either CSST A or CSST B were postulated to occur such that off-site power is lost to one of the Unit 1 power trains, off-site power can be restored to the affected power train by manual operator alignment of the 6.9kV shutdow n board normal/alternate feeder breakers.
Based on these results, the loading on CSST B remains within the functional capabilities of the component. One immediate source of off-site power remains available for the 1A and the 1B start buses as well as one delayed source (based on manual operator action if failure of the credited CSSTs is postulated) in accordance with the requirements of IEEE-308 and General Design Criterion (GDC)-17. As such, this one
time only CSST alignment change had no effect on nuclear safety.
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 UFSAR
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS E-4 Section 2.4.12 This activity involv es a revision to the UFSAR which evaluates the dilution and disbursement effects of
radiological effluents released to the surface waters
in proximity to the plant. The revision is required to
address changes in river flow control operations
associated with a downstream dam which result in a
reduction in the river flow rate assumed in the
original evaluation.
Section 2.4.12 of the UFSAR evaluates the ability of the surface waters near the plant to dilute and disburse a number of postulated accidental
radioactive liquid releases. The eval uation confirms that the level of activity at the first downstream surface water intake will be below 10CFR20 limits. The evaluation is performed in accordance with the
requirements of 10CFR50, Appendix I. In this evaluation, the river flow is specified as the flow which is equaled or exceeded approximately 50 percent of the time. Based on recent changes to the operational practice for flow through a downs tream 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 anal ysis. Revised dilution values were calculated using the existi ng evaluation methodology and model.
The calculation established the revi sed 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 invo lves a revision of the environmental conditions described for the shutdown transformer
rooms. The current UFSAR indicates that the
ventilation system for these rooms is designed to
maintain a minimum room temperature of 60ºF.
This change revises the description to indicate that
the ventilation system is designed to maintain the
room temperature within a range of 15°F to 97°F
consistent with ventilation equipment functional
capabilities and services equipment requirements.
Section 9.4.2.2.6 of the UFSAR c ontains a general description of the shutdown transformer room ventila tion system. The environmental design criteria for these rooms cons ervatively establish the minimum room temperature as 15 degrees F.
A review plant records since initial operation indicate that the lowest recorded temperature within these rooms is 24 degrees F.
The subject rooms contain the 480V s hutdown transformers. Review of the transformer design indicates t hat the limiting low temperature 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 pos itive pressure nitrogen blanket over the transformer. An oil tem perature of 15 degrees F has been established to be acceptable based on a pour temperature limit of SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 UFSAR
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS E-5
-50 degrees F.
Since the minimum 15 degrees F desi gn basis room temperature is conservative with respect to ac tual 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 performance capabilities assumed in the plant
accident and transient analyses have been revised
to 1) reduce the minimum developed head values
for the charging (high head) and safety injection (intermediate head) pumps by 5 percent and 2)
increase the maximum charging pump developed
head values for pump flows which exceed 250 gpm.
These changes have been previously made to the
realistic large break loss-of-coolant accident (LOCA)
analyses documented in Sequoyah Technical
Specification Change Requests TS-07-04 and TS-
08-01. The large break LOCA analysis was the
limiting design basis transient for minimum ECCS
pump performance. Application of the realistic
analysis methodology to the large break LOCA
transient demonstrated acceptable results with the
revised ECCS pump performance characteristics.
This activity extends the applicability of the revised performance assumptions to the balance of the
plant transient and accident analyses.
The ECCS pumps are designed to provide emergency core cooling and reactivity control in the event of high energy piping breaks or other transients which involve primary syst em inventory losses or reactivity insertions. Evaluation of the minimum pump performance assumption
changes resulted in a complete re-analysis of the small break LOCA transient. For the revised ECCS pum p performance characteristics, the calculated peak fuel cladding increas ed from the previous value of 1162 degrees F to 1403 degrees F. This result remains well below the
established analysis acceptance cr iteria of 2200 degrees F. The re-analysis also confirmed that the revised ECCS performance assumptions are sufficient to demonstrate compliance with the balance
of the small break LOCA anal ysis acceptance criteria.
The evaluation of the balance of the plant transients and accidents concluded the current analyses of reco rd are either 1) not affected by the changes, 2) bounded by existing c onservatisms built into the analysis or 3) bounded by crediting the full capability of the ECCS
which was conservatively minimized in the analysis of record through the use of simplifying assumptions.
Based on this result, the existi ng analyses were established to be conservative and bounding for the revised ECCS performance
assumptions.
SEQUOYAH NUCLEAR PLANT CHANGES IN THE FACILITY FOR AMENDMENT - 21 Work Order
SUMMARY
OF DESCRIPTION SAFETY ANALYSIS E-6 99-007763-000 Under this work order, one of the essential raw cooling water (ERCW) pumping station traveling
water screens was dissembled and replaced. The
disassembly operation consisted of raising the
entire screen assembly and removing one section at
a time as required for replacement. Re-assembly
was accomplished by reversing the process. The
lifts were performed by a crane supported on a
barge parked adjacent to the ERCW pumping
station. The traveling screen sections were placed
on the barge and were moved to and from a staging
area. The following administrative controls were
applicable to the work order.
- 1. All tornado missile protection roof panels remained in place during the load lifts except for
those permitted to be removed by prior analysis.
- 2. Load lifts were not permitted during a tornado watch, warning, or any other periods of high
winds.
- 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. Evaluation of this activity included re view of the actual lifting operation as well as the maneuvering of su rface craft near the ERCW pumping station. The evaluation concluded:
- 1. The screen lifting operation placed loads in excess of 2100 lbs above safety-related equipment. As such, lifting of the traveling
screens was performed in accordance with all the requirements of
NUREG-0612 for a heavy load lift.
- 2. Design and operation of the bar ge and tug boat were consistent with applicable industry standards.
Controls and restrictions placed on the maneuvering of the craft are conservative and effective in
preventing impact with t he ERCW pumping station.
- 3. Blocking flow to the ERCW intakes due to the postulated sinking of the barge/tug has been evaluated and is not credible. A trench
exists immediately in front of the intakes. The width of the
barge/tug is greater than the trench and the slope of the trench
walls is steep such that it is not possible for the submerged barge to block the intakes.
As such, nuclear safety was not affected by this work activity.