ML090210728

10 CFR 50.59, Changes, Tests, and Experiments and Commitment Summary Reports, Supplemental Report

ML090210728
Person / Time
Site:	Watts Bar
Issue date:	01/21/2009
From:	Brandon M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090210728 (13)

Text

10 CFR 50.59(d)(2)

U. S. Nuclear Regulatory Commission

ATTN: Document Control Desk

Washington, D.C. 20555-0001

Gentlemen: In the Matter of ) Docket No. 50-390 Tennessee Valley Authority )

WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - 10 CFR 50.59, CHANGES, TESTS, AND EXPERIMENTS AND COMMITMENT

SUMMARY

REPORTS, SUPPLEMENTAL

REPORT This letter supersedes TVA's letter, "Watts Bar Nuclear Plant (WBN) Unit 1 - 10 CFR 50.59, Changes, Tests, and Experiments and Commitment Summary Reports," dated September 24, 2008. This letter includes two additional 50.59 safety evaluations that should have been

included in the September report. These omissions were identified in a recent self-

assessment and have been entered for resolution into TVA's corrective action program as

Problem Evaluation Report 160317. Additional clarifications were also made regarding the

conclusions of two evaluations.

Pursuant to 10 CFR 50.59(d)(2), this letter provides the complete Summary Report of the

implemented changes, tests, and experiments in which evaluations were performed in accordance with 10 CFR 50.59(c). The enclosure provides a summary of the evaluations

associated with Updated Final Safety Analysis Report Amendment 7 (provided separately)

and includes other evaluations implemented during the period from March 17, 2007 to June

30, 2008.

During this reporting period, there were no previous commitments that TVA has evaluated

and revised using administrative controls that incorporate the Nuclear Energy Institute's (NEI) 99-04 "Guideline For Managing NRC Commitments."

U.S. Nuclear Regulatory Commission Page 2

There are no regulatory commitments associated with this submittal. If you have any

questions concerning this matter, please call me at (423) 365-1824.

Sincerely,

M. K. Brandon

Manager, Site Licensing

and Industry Affairs

Enclosure

cc: See Page 3 U.S. Nuclear Regulatory Commission Page 3

Enclosure

cc (Enclosure): NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

U.S. Nuclear Regulatory Commission Page 3

MKB:

Enclosure

cc (Enclosure): NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

G. Arent, EQB 1B-WBN

G. A. Boerschig, MOB 2R-WBN

W. R. Campbell, LP 3R-C

W. T. Coutu, LP 3R-C

D. E. Jernigan, LP 3R-C

M. T. McFadden, ADM 1Q-WBN

G. W. Mauldin, EQB 2A-WBN

L. E. Nicholson, LP 3R-C

M. A. Purcell, LP 4K-C

NSRB Support, LP 4K-C (including Advisors)

A. J. Scales, MOB 2M-WBN

J. E. Semelsberger, EQB 2W-WBN

M. D. Skaggs, ADM 1V-WBN

J. F. Tortora, Jr., ADM 1B-WBN

E. J. Vigluicci, WT 6A-K

K. W. Whittenburg, SP 2B-C

Sequoyah Licensing Files, OPS 4C-SQN

EDMS, WT 3B-K

M:\SUBMIT\Cycle 8 50.59 Report Supplement.pdf

E-1 of 9 ENCLOSURE WATTS BAR NUCLEAR PLANT UNIT 1 10 CFR 50.59

SUMMARY

REPORT

1. SA-SE Number: WBPLMN-07-008-1

2. SA-SE Number: WBPLMN-05-026-0

3. SA-SE Number: WBPLEE-07-022-1

4. SA-SE Number: WBPLMN-07-014-0

5. SA-SE Number: WBPLMN-08-006-0

6. SA-SE Number: WO 07-814239-000

7. SA-SE Number: WBPLMN-08-004-0

8. SA-SE Number: WBPLMN-08-005-0 SA-SE Number: WBPLMN-07-008-1 Implementation Date:

03/29/2007 Document Type: Affected Documents: Title: Temporary Alteration Change Form (TACF) Temporary Alteration 1 0002-065, R1 FSAR Change Package 1915 TS Bases change package

2007-02 Post-Accident Safety Function of Emergency Gas Treatment

System (EGTS) Description and Safety Assessments:

The following issue was documented in TVA's correct ive action program as a problem evaluation report (PER):

"Constant Operator monitoring will be required if, following an accident, the containment isolation phase A (CIA) signal is reset prior to annulus differential pressure decreasing to the

setpoint needed for the EGTS pressure control loop's isolation valves to open. EGTS will not

perform its safety function without these valves opened as required. This was identified by

Operators during simulator exercises and has been identified as a Priority 1 Operator Work

Around."

The EGTS consists of two separate control loops. The hand switches of the control loops can be

placed in the A-AUTO or A-AUTO STANDBY positions. The annulus pressure setpoint for the A-

AUTO STANDBY position is -0.812 inches water gauge (wg). If a large break Loss of Coolant

Accident (LOCA) occurs, the annulus pressure will increase from -5 inches wg to approximately -1

inch wg due to effects from the LOCA. The annulus pressure will continue to increase to the -0.812

inches wg setpoint since neither the annulus vacuum control system nor EGTS is operating to reduce

the annulus pressure; i.e. the annulus vacuum contro l system is turned off by a safety injection system signal initiated subsequent to a LOCA. When the -0.812 inches wg EGTS setpoint is reached, both EGTS control loops will operate to filter the airborne vapors and particulates in the annulus. The

off site and main control room doses will remain less than the 10 CFR 50, Appendix A, GDC 19

control room limits and the 10 CFR 100 offsite dose limits.

If a small break LOCA occurs, the annulus pressurization effects will not be as pronounced as with a

large break LOCA. In this event, the annulus pressure may remain less than the -0.812 inches wg

setpoint until the CIA signal is reset by operations. When the CIA signal is reset, the EGTS control

loops in the A-AUTO STANDBY position are disabled. In this case, no EGTS train will operate.

Consequently, the EGTS will not function as descri bed in the Updated Final Safety Analysis Report (UFSAR) and the EGTS safety functions will not be performed.

The primary safety function of the EGTS is to ma intain the offsite and Main Control Room (MCR) doses less than the allowable 10 CFR 100 and 10 CFR 50, Appendix A General Design Criteria 19

limits during a LOCA. This safety function is a ccomplished during a LOCA by (1) keeping the annulus pressure below the outside atmospheric pressure and (2) removing airborne particulates and vapors

that may contain radioactive nuclides from air drawn from the annulus and exhausted to the

atmosphere. These safety functions will be accomp lished with implementation of the TACF corrective actions assuming a credible failure of one train of EGTS exhausting to the outside environment for the

duration of a LOCA accident.

These changes will increase the reliability because active components in the auxiliary relay circuit that

are currently required to open the valves will be eliminated. No failure assumptions are changed, and

no accident frequencies are increased. Specific changes addressed in this TACF do not create the

possibility of a malfunction with different than currently evaluated results. Additionally, no fission

product barriers are affected.

SA-SE Number: WBPLMN-05-026-0 Implementation Date:

06/25/2007 Document Type: Affected Documents: Title: Engineering Document Change (EDC) EDC 51619-A Main Steam Valve Vault Airflow Rates Description and Safety Assessments:

The airflow rates of the exhaust fans serving the Main Steam Valve Vaults (MSVVs) are outside of the

+/-10% requirement stipulated by General Specification G-37. The fans are designed to maintain an

adequate temperature environment for the main steam safety related valves, but are non-safety related. Design Change Notice (DCN) M18106-A added ductwork to the suction side of these fans, which caused the fan static pressures to be considerably increased. Because of this, air flow rates

were significantly reduced. During the periodic tests, the fan speeds were increased to meet the

design airflow rates.

Problem Evaluation Report (PER) 03-006901-000 revealed that the fan speeds were higher than

those permitted in Technical Instruction-5.002, "Flow Testing of Ventilation Systems." Further

investigation revealed that the fan speed limits provided in Table 1 of the Technical Instruction may

exceed the structural classification limits of the MSVV exhaust fan housings. This was confirmed through a review of the contract documentation and discussions with the manufacturer. It was

advised to limit the fan speed to 535 rpm.

Temperature data was collected for a limited time period within the WBN MSVVs to justify the

acceptability of the reduced total flow rates. In the north MSVV, temperature values remained

unchanged, so the reduced exhaust fans' test flow rates are considered acceptable. In the south

MSVV, hot spot locations were generally greater than current values. These changes were updated

via Drawing Change Authorization-51619-002. The hot spot changes do not affect the Minimum, Normal Average, Normal Maximum and/or Abnormal Maximum temperatures. Quality of life of components was re-evaluated based on the new hot spot temperatures. The changes required were

as follows: 1) Four cables in binder WBNEQ-CABL-049 reduced their qualified life span from 34.5

years to 25.2 years. 2) The qualified life for the o-ring seals within the motor operated valves was

changed from 35.39 years to 30.47 years. 3) Sol enoid valve life span was reduced from 30.49 years to 27.32 years. 4) EQ Binder WBNEQ-IZS-003 was re vised from 9.17 years to 8.17 years, EQ Binder WBNEQ-IZS-007 was revised from 7.94 years to 7.06 years, and the qualified life for certain limit

switches listed in EQ Binder WBNEQ-IZS-005 was revised from 7.76 years to 7.05 years and from 9.89 years to 7.67 years.

The proposed changes do not introduce the possibility of a change in the frequency of an accident

and no new failure modes are introduced. Operation of the MSVV exhaust fans within the acceptable

limits based on the manufacturer's specifications inherently reduces the possibility of their

malfunction, and therefore this change will decrease the likelihood of a malfunction. Based on the

analyses, the Heating Ventilation Air Conditioning sy stem will still perform its intended design basis functions. No changes are made that would increase the consequences of any existing postulated

malfunction, and the reduction in air flow rates does not have the potential to create a new accident.

No fission product barriers are unduly challenged due to this change.

SA-SE Number: WBPLEE-07-022-1 Implementation Date:

10/30/2007 Document Type: Affected Documents: Title: Design Change Notice (DCN) DCN 52233-A Tornado Watch/Warning Final Safety Analysis Report (FSAR) Change Pkg. 1919 Description and Safety Assessments:

During the Component Design Basis Inspection (CDBI), the NRC identified a potential problem with

the low temperature switches associated with the Diesel Generator (DG) room exhaust fans and the requirement that the fans be energized and running during a tornado watch/warning condition. Since

the low temperature switches have a re-set setpoint value of 73°F and a setpoint of 68°F, the potential

exists that attempts to start the fans and open the associated dampers may be unsuccessful, or that

the fans may start initially, but later stop and close the dampers when they are required to be open. If

the temperature switches have cooled down below thei r setpoint value of 68° F, the switches will prevent manual start of the fans until the re-set set point value of 73°F is reached. Problem Evaluation Report (PER) 120005 was initiated to document the issue and establish the required corrective

actions. As part of the immediate actions for the PER, Abnormal Operating Instruction (AOI) 8, "Tornado Watch or Warning," was revised to require installation of a jumper wire in each of the DG

room exhaust fan Motor Control Center panels during a tornado watch. This was done in accordance

with a Temporary Alteration (TA) created by a technical procedure per Site Programs and

Procedures-9.5, "Temporary Alterations." DCN 52233-A supersedes this TA by installing tornado

bypass hand switches for each of the exhaust fans as explained below.

The design change is being issued to install a new tornado bypass hand switch in the control circuit

for each of the diesel generator room exhaust fans in order to bypass the low temperature switch function during a tornado watch or warning in accordance with AOI-8. During normal operating

conditions, the new hand switch will be placed in the "NORMAL" position allowing the low temperature

switch to control their respective fan, as necessary, in order to maintain the diesel generator room

exhaust temperature between 50°F and 120°F. In response to the requirements being added to

Section 5.6 of System Description N3-30DB-4002 and captured in AOI-8, the new tornado bypass hand switch will be placed in the "BYPASS" position to allow the DG room exhaust fans to start and

cause the associated inlet and exhaust dampers to open at temperatures less than the 68°F setpoint

during tornado conditions.

If the outside temperature is expected to drop below 50°F during a tornado warning, then a start of

each DG is required to be performed to ensure DG functionality. Monitoring the outside air

temperature is required during tornado warnings to assure the DGs are started prior to a drop in

temperature below 50°F. When tornado conditions are cancelled, the exhaust fans are required to be

turned off and the hand switches will be returned to "NORMAL" promptly, and independently verified in accordance with AOI-8. Based on the evaluation of historical data, it is very unlikely that a tornado

will occur with a temperature concurrent or lower than 50°F. Therefore, this change does not result in

more than a minimal increase in the frequency of occurrence of an accident previously evaluated in

the Updated FSAR.

The change supports successful completion of AOI-8 actions to open each DG room exhaust fan

supply and dampers and start each exhaust fan. The change does not result in new accidents or

malfunctions, and it does not increase the frequency or consequences of accidents or malfunctions

evaluated in the UFSAR. Additionally, no fission product barriers are challenged by the change.

SA-SE Number: WBPLMN-07-014-0 Implementation Date:

03/13/2008 Document Type: Affected Documents: Title: Design Change Notice (DCN) DCN 52220-A ABI/CVI Intertie Description and Safety Assessments:

DCN 52220-A modifies the Train A and Train B electrical circuits for high radiation in the refueling

area logic bus and the Train A and Train B Solid State Protection System (SSPS) input for high

radiation in the containment purge air exhaust, which initiates Containment Vent Isolation (CVI). The

modification to these circuits allows WBN to ma intain the Auxiliary Building Secondary Containment Enclosure (ABSCE) in the event of a high radiation signal in either the refueling area or the

containment purge exhaust during refueling operations while the containment and/or annulus is open

to the auxiliary building ABSCE spaces. This change will permit, but not require, operation of

containment purge when moving irradiated fuel in the Auxiliary Building with containment and/or

annulus hatches, personnel hatches, or penetrations open to the ABSCE spaces.

This change does not allow the equipment hatch to be open during fuel movement inside containment

since Technical Specification 3.9.4, "Refueling Operations, Containment Penetrations," requires the

penetration to be closed during this time.

The electrical circuit modifications are implement ed in two Auxiliary Relay Racks. One hand switch for each train will be added to the appropriate rack to allow the circuits to be swapped from a normal

mode to a refueling mode. During normal operation, the circuits will perform the same logic as before

the modification. Prior to starting refueling operations, the hand switch will be placed into the

refueling mode, and it will be returned to the nor mal mode prior to entering Mode 4 from Mode 5.

During the refueling mode, the ABSCE isolation valves will be closed by receiving either the current high radiation signal from the spent fuel pool accident radiation monitors or upon a CVI signal. The

Containment Ventilation Isolation Valves that are currently closed by a CVI signal will also be closed upon receiving a high radiation signal from the spent fuel pool accident radiation monitors. Isolation

of these valves will establish the ABSCE boundary, prevent the potential for back flow from the Shield

Building exhaust vent, and ensure that radioactive releases due to fuel handling accidents in

containment will be processed by the Auxiliary Bu ilding Gas Treatment System (ABGTS) if they migrate into the Auxiliary Building.

The response time requirement for the spent fuel pool radiation monitors is not affected by this

modification. A new response time requirement for the containment purge monitors during refueling

operations is required. The new response time requirements do not affect the calculated off site or

main control room dose analyses for the fuel handling accident.

This change complies with the safety and functional requirements specified in the applicable design

bases documents and does not adversely affect the performance of any safety related equipment.

The proposed modifications do not increase the frequency or likelihood of accidents or malfunctions, increase the consequences of an accident or malfunction beyond the ten percent allowed by

10CFR50.59, or create a new type of accident. The design bases for fission product barriers will not

be altered or exceeded and no new methods of evaluation were used in evaluating the proposed

modifications.

SA-SE Number: WBPLMN-08-006-0 Implementation Date:

04/24/2008 Document Type: Affected Documents: Title: Design Change Notice (DCN) DCN 52307-A Lower Compartment Coolers Blank Off Plate Installation Description and Safety Assessments:

The design function of the Lower Compartment Coolers (LCCs), along with the Control Rod Drive

Mechanism (CRDM) coolers, is to maintain acceptable temperature within the reactor building lower

compartment areas for the protection of equipment and control during normal reactor operation and

shutdown. Any combination of three LCCs, or any combination of two LCCs and two CRDM coolers

are required for safe shutdown per 10CFR50 Appendix R to prevent containment temperatures from exceeding established Equipment Qualificati on profiles used in the 10CFR50.49 program.

The LCC 1D-B upper outboard coil leaks at or near the header/tube interface and cannot be repaired

in place. This design change will blank off the supply and return of the Essential Raw Cooling Water (ERCW) to this cooling coil, resulting in seven of the eight coils remaining functional. Blank off plates

of stainless steel that are capable of withstanding the ERCW system design pressure will be installed

at the supply and return flanges associated with the leaking coil. The change will isolate ERCW flow

to the offending coil only and will not adversely impact continued flow to the remaining seven coils.

The piping system will remain seismically qualified with the blanking plates installed.

If the water leak in the offending coil is allowed to continue, an increase in the relative humidity levels

inside lower containment is expected. This condition could mask water leaks inside lower

containment. Failure of the blank off plates is less likely to occur than the failure of the existing coil

components, and the installation of the plates eliminates the potential for the coil to rupture. This

change will prevent one of the coils from performi ng its cooling function. Calculations have been revised to note that the capacity reduction of 1/8 on the cooler will not affect the current requirements.

Any combination of three LCCs or two LCCs and two CRDM coolers remains acceptable and does

not affect Technical Specification requirements.

Proper operation of the LCC cooling coils is not a safety related function and is not required for the

mitigation of any Updated Final Safety Analysis Report (UFSAR) Chapter 6 or 15 accidents. The

blanking plate will be designed to meet the design conditions of the ERCW system. The design basis

functions of the LCC 1D-B is maintained and is not adversely impacted by this proposed design

change. Therefore, this change does not result in any new accidents or malfunctions, and does not

result in increased frequency or consequences of accidents or malfunctions evaluated in the UFSAR.

In addition, no fission product barriers are challenged by this change. Safety related functions of the

LCCs are not affected by this proposed modification.

SA-SE Number: WO 07-814239-000 Implementation Date:

05/29/2008 Document Type: Affected Documents: Title: Maintenance Work Order (WO) Work Order 07-814239-000 Areva Procedure FS-235 Fuel Cleaning Description and Safety Assessments:

During Cycle 7 operation, the WBN Unit 1 core exhibited symptoms of Crud Induced Power Shift (CIPS). The root cause of this is boiling near the upper surfaces of high-powered fuel rods, leading to

the depositing of crud in these locations. The Ultrasonic Fuel Cleaning System (UFC) was developed

by Dominion Engineering, Inc. with Electric Power Research Institute (EPRI) to remove the corrosion

products from Pressurized Water Reactor (PWR) fuel using ultrasonic energy. Laboratory and field

experience has shown an average crud removal rate of greater than 80%. All UFC operations will be

performed underwater in the Unit 2 side of the transfer canal at WBN. Lifting and handling of the UFC

system and its support stand will be performed in accordance with NUREG 0612 to mitigate the

potential for damage to stored fuel.

The existing potential for pellet-clad interaction resulting from agitation of fuel pellet fragments that

were cracked prior to cleaning has been evaluated and determined to be very unlikely. Normal

administrative and procedural controls will be employ ed to ensure fuel is not damaged. Because of this evaluation and administrative control, the ri sk of damaged fuel or an increased risk of a fuel handling accident is negligible. Additionally, the cleaning apparatus is not located in an area where it

can fall on any safety-related equipment besides the transfer canal liner.

The increased risk of crud particle release into the Unit 2 transfer canal is mitigated by the current

operating procedures that require the fuel to be secured safely and cleaning suspended until pool

visibility is restored. The potential to release dissolved radioactive gasses is minimized by design.

The average flow velocity in the cleaning apparatus is sufficiently high to entrain all gas bubbles

smaller than 3 mm in diameter. These bubbles are mixed vigorously in the connection hose between

the cleaner and the filter skid. This mixing action causes most of the gasses to re-dissolve into the

bulk fluid. Any that are not re-dissolved have been shown to not cause a significant radiological

hazard.

The possibility of a fuel assembly becoming stuck in the cleaning fixture is very remote. It is protected

against by the small amount of force required to straighten a 1 inch bow in a fuel assembly, which can

be provided by the sidewalls of the cleaning fixture. Also, the fuel can be indefinitely cooled in the

cleaning apparatus by natural convection, as previously discussed.

The Spent Fuel Cooling and Cleaning System is a safe ty related system required to mitigate the consequences of Design Basis Events. During operation of the UFC system, fuel assemblies are

placed in the fuel cleaning chambers and bombarded with high frequency pressure oscillations. This

adds negligible heat loads to the system and will not prevent the spent fuel cooling system from performing its intended design function. Additionally, the UFC system is designed such that natural

circulation will cool a fuel assembly in the UFC chamber in the event of a station blackout.

All fuel handling will be performed within the bounds of Updated Final Safety Analysis Report (UFSAR) procedures and interlocks. This change does not affect the consequences of an accident or

malfunction because it does not affect the dose results of the accident analyses reported in the

UFSAR. Because of the normal administrative and procedural controls in place, the increase in

probability of a fuel handling accident is negligible. Potential accidents that have not been previously

evaluated are not introduced by this change. No fission product barriers are affected by this activity.

SA-SE Number: WBPLMN 08-004-0 Implementation Date:

03/27/2008 Document Type: Affected Documents: Title: Design Change Notice (DCN) DCN 52211-A Safety Analysis Report (SAR) Change Package 1927 Relocation of Radiation Monitor for Transfer Canal Fuel Handling

Accident Detection Description and Safety Assessments:

A fuel handling accident (FHA) in the fuel transfer canal may not be detected by the radiation monitors

located on the wall of the spent fuel pool below t he 757' floor elevation. Consequently, the Auxiliary Building Gas Treatment System (ABGTS) would not be aut omatically established. As a result of this condition, calculation WBNTSR-009 was revised. The revised calculation determined that such an

accident can be mitigated if spent fuel pool accident radiation monitor 0-RE-90-103-B is relocated

above the refueling floor elevation and the exhaust vents located around the fuel transfer canal

perimeter are isolated during movement of irradiated fuel in that area. Air intake must be isolated to

ensure the activity released by a FHA in the tr ansfer canal is not removed and exhausted by the ventilation system prior to radiation monitor detection.

This DCN is only relocating the radiation monitor, not changing its design functions. The mounting

method for the new location has been evaluated and has been determined to meet seismic

requirements.

Additionally, this DCN includes the isolation of air flow from the transfer canal by closing either a fire

damper or a balancing damper during irradiated fuel movement in the transfer canal. Alternate

ventilation requirements are required to ensure t he fuel handling area ventilation system remains balanced to support both normal and post-accident conditions. The DCN requires opening one or two

selected fuel handling area exhaust ductwork access panels and adding a sheet metal plate to restrict

air flow through one access opening as needed. The access panel(s) will be secured in place using

heavy gauge wire, and stainless steel screens will be installed to prevent debris from entering the

ductwork. Static pressure will be measured upstream of the access panel both before and after

isolation of flow from the transfer canal. This will verify required normal or post accident air flows are

maintained.

Relocation of the spent fuel pool radiation monitor will not affect the safety related operation of the

monitor. The proposed changes to the fuel handli ng area ventilation system as described above will not adversely affect establishment of the Aux iliary Building Secondary Containment Enclosure or operation of the ABGTS during FHA or Loss of Coolant Accident (LOCA) conditions. Thus, the

proposed changes do not result in an increase in frequency of occurrence of accidents or likelihood of

occurrence of malfunctions evaluated in the Updated Final Safety Analysis Report (UFSAR), and do

not result in an increase in the consequences of an accident or malfunction. The proposed changes

do not result in an accident of a different type or a malfunction with a different result. Further, the

changes do not affect fission product barriers and the method of evaluation described in the UFSAR.

SA-SE Number: WBPLMN 08-005-0 Implementation Date:

04/14/2008 Document Type: Affected Documents: Title: Engineering Document Change (EDC) Design Change Notice (DCN) 52270-A Number 2 Heater Drain Bypass to Condenser Level Control Valves Description and Safety Assessments:

When the Number 2 Heater Drain Bypass to Condenser level control valves (LCVs) are opened

during high turbine load conditions, violent water hammer occurs. This design change of

documentation only allows for the bypass LCVs to be opened only during low power operations to

protect against severe water hammer incidences.

The three bypass to condenser lines from the No. 2 heater were not part of original plant design. A

condition was identified where the No. 2 heaters did not drain to the No. 3 Heater Drain Tank (HDT)

during rapid load reductions at low power operating conditions. To account for this, the bypass lines

were installed. The LCVs are normally in the closed position and were designed to automatically

open if the heater water level reaches a prescribed high setpoint. This design change will still allow

the LCVs to be opened at low power levels. The change will require the bypass lines to be isolated

during normal high turbine load operation to prevent water hammer.

If maintenance is needed on the main drain valves, the main drain path bypass valves can be used to

maintain No. 2 heater water level. In the event the normal drain path malfunctions during high power

operation, the bypass to condenser LCVs will no longer open automatically, and a heater string

isolation could occur. It was determined that the reduced feedwater temperature caused by a heater

string isolation will not compromise plant safety.

This design change has the potential to increase the frequency of loss of normal feedwater if all three

strings were to isolate. Isolation of main feedwater from the steam generators (SGs) would require

Operations to start the Auxiliary Feedwater (AFW) system manually, or AFW will automatically start on a low low SG level. The "Loss of Normal Feedwater" is categorized as a fault of moderate

frequency, implying once a year occurrence. Historical operating experience at WBN shows that

heater drain failures do not typically occur at a frequency greater than once per year. Further, no new

failure modes are introduced that could increase the frequency of occurrence of an isolation of all

three intermediate pressure heaters and no new components are added with this design change.

The design basis accident analyses evaluated in the Updated Final Safety Analysis Report (UFSAR)

are not affected, and the system is not credited for accident mitigation. No equipment required for

safe operation or shutdown is changed by this design. No fission product barriers are challenged by

this change.