ML15040A593

Initial Exam 2015-301 Post Exam Comments

ML15040A593
Person / Time
Site:	Browns Ferry
Issue date:	02/09/2015
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
50-259/15-301, 50-260/15-301, 50-296/15-301 50-259/OL-15, 50-260/OL-15, 50-296/OL-15
Download: ML15040A593 (18)

Text

Post Operating Exam Comment #1 Scenario 2 Critical task 3 states:

After all RPV level instruments flash (level unknown), inject into the RPV with all available sources until one of the conditions in C-4, Note 7 is met.

1. Safety Significance Prevent fuel damage by establishing adequate core cooling

2. Cues Procedural compliance Loss of all RPV level indications

3. Measured by Observation-Indications that the Main Steam Lines are flooded are listed in C-4 Note 7

4. Feedback MSRV tail pipe temperature MSRV acoustic monitor RPV Pressure trend This Critical Task is not met if the Crew fails to continue to raise level and maintain pressure on the vessel.

Crew Performance:

One of the Crews did not inject into the RPV with all available sources until one of the conditions in C-4, Note 7 is met.

Bases for Post Exam Comment:

EOl step C4-25 states: Flood the RPV to the elevation of the main steam lines with the following. It does not say to use all available sources.

EOIPM Section O-V-J Discussion of step C4-25 states:

aligned to flood the RPV and establish core cooling by submergence. The list of flooding methods includes all sources capable of injecting into the RPV, including the alternate injection subsystems defined in EOIPM Section 0-I-C. Any or all of the listed methods may be used as necessary to flood the RPV to the elevation of the main steam lines.

The Facility recommends rewording the critical task as follows:

After all RPV level instruments flash (level unknown), flood the RPV with any or all of the listed methods to the elevation of the main steam lines.

1. Safety Significance Prevent fuel damage by establishing adequate core cooling

2. Cues Procedural compliance Loss of all RPV level indications

3. Measured by Observation-Indications that the Main Steam Lines are flooded are listed in C-4 Note 7

4. Feedback MSRV tail pipe temperature MSRV acoustic monitor RPV Pressure trend This Critical Task is not met if the Crew fails to continue to raise level and maintain pressure on the vessel.

BFN CONTINGENCY #4, RPV FLOODING EOIPM SECTION 0-V-J Unit 0 BASES Rev. 0001 Page 64 of 79 I0 CONTINGENCY #4, RPV FLOODING BASES (continued)

ISOLATE ANY of the following NOT needed for RPV injection:

• Main steam lines

• Main steam line drains

• HPCI

• RCIC C4-24 CAUTION

1. 2 Pump NPSH and Vortex Limits I

1. 3 Elevated suppr chmbr press may trip RCIC

1. 6 HPCI or RCIC suction temp above 1 40F L

I, FLOOD the RPV to the elevation of the main steam lines with the following:

INJ SOURCE APPX ss CNDSand AN 5A 1210 psig CRD 5B 1640 psig RCIC with CST suction if available 5C 1200 psig HPCI with CST suction if available SD 1200 psig CNDS 6D, 68 480 psig CS 6A 330 psig LPCI 6B, 6C 320 psig Cs 6D 6E 330 psig Stby coolant 7D 160 psig RHR crosstie to other units 7C 320 psig Fire Protection system 7K 150 psig CNDStransferpumpstoRHR 7A ilOpsig RHR drain pumps 7E, 7F 50 psig PSC head tank pumps 7G 30 psig RCIC (aux boiler steam) with CST 7H 1200 psig suction if available EDMG-24 RCIC manual start 1200 psig APPX B HPCI (aux boiler steam) with CST 7J 780 psig suction if available SLC (boron tank) 7B 1450 psig SLC (test tank) 78 1450 paig EDMG-24 Portable pump 150 psig APPX E C4-25 L

BFN CONTINGENCY #4, RPV FLOODING EOIPM SECTION 0-V-J Unit 0 BASES Rev. 0001 Page 65 of 79 1.0 CONTINGENCY #4, RPV FLOODING BASES (continued)

I DISCUSSION : C4-25 I

Once the MSRVs have been opened to depressurize the RPV, injection sources are aligned to flood the RPV and establish core cooling by submergence. The list of flooding methods includes all sources capable of injecting into the RPV, including the alternate injection subsystems defined in EOIPM Section 0-I-C. Any or all of the listed methods may be used as necessary to flood the RPV to the elevation of the main steam lines.

Steam-driven systems may be used, if needed and available, even though RPV water level may eventually exceed the elevation of the steam supply nozzles. While flooding the steam lines of operating systems could result in extensive damage, it is unlikely that RPV water level would actually reach the steam lines while the systems are in use. With the reactor shutdown, the RPV will likely depressurize to below the turbine stall pressures as a result of steam flow through the MSRVs and operation of the steam-driven systems themselves.

If suppression pool water level cannot be maintained above the top of the HPCI exhaust opening, HPCI must be secured, irrespective of whether the core is adequately cooled, in accordance with SP/L of the Primary Containment Control flowchart. Operation with the exhaust exposed would directly pressurize the suppression chamber. No such restriction exists for RCIC operation since the RCIC exhaust flowrate is much smaller and elevated suppression chamber pressure will trip the RCIC turbine before primary containment integrity is challenged.

High RPV water level logic may be defeated as necessary to permit use of available flooding sources. If RPV water level cannot be determined, the inputs to the logic must be considered invalid. Other system operating details are consistent with those listed in Steps RC/L EOI-1.

Post Operating Exam Comment #2 Scenario 3 Critical task 3 states:

During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MSCP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation - No ECCS injection prior to being less than the MSCP.

AND Observation - Feedwater terminated and prevented until less than the MSCP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms.

Injection system flow rates into RPV.

This Critical Task is not met if the Crew injects too fast and causes power oscillations or APRM downscale clear (>5% power).

Crew Performance:

Crew B did terminate and prevent RPV injection from ECCS and Feedwater until reactor pressure was below the MSCP however; they did momentarily exceed the APRM downscale setpoint of 5% power once RPV injection was reestablished.

Bases for Post Exam Comment:

The following statement was added to the critical task at the request of an NRC examiner during the exam preparation week:

This Critical Task is not met if the Crew injects too fast and causes power oscillations or APRM downscale clear (>5% power).

The Facility disputes this statement as bases for failing to meet this critical task.

The actions required by the critical task and measured against the above listed Measured by criteria in the critical task were satisfied by the crew.

In accordance with EOI step C5-25 the Crew started injection in a controlled manner using 2!JJY the Condensate system and increased flow using the startup bypass (throttle) valve to restore Reactor water level. Once the determination was made that Reactor water level could be restored and maintained above (-) 180 inches the crew was directed by step C5-28 back to C5-14 which was already completed.

The next step, C5-1 5 states:

IF Reactor power is above 5% or unknown And Reactor water level is above (-) 50 inches Then go to step C5-5. (Note that the APRMs could be above 5% and no action is required as long as Reactor water level is below (-) 50 inches, see EOIPM section 0-V-K). The Crew recognized that that Reactor power and Reactor water level were rising and injection was stopped prior to exceeding (-) 50 inches. Reactor water level lowered rapidly to approximately (-) 87 inches and the operator resumed injection with condensate. Reactor power momentarily exceeded 5% and Reactor water level reached (-) 50 inches. The conditions for step C5-5 were not meet because with Reactor pressure less than 50 psig above suppression chamber pressure the MSRVs would be closed. The crew immediately secured injection lAW C5-9 and Reactor water level immediately lowered to below (-) 50 inches allowing injection to the Reactor to be resumed lAW C5-10.

The actions of the crew were in accordance with CS and EOIPM 0-V-K, Contingency #5 Level/Power Control Bases.

Based on C5 and EOIPM 0-V-K there was no Safety Significance to momentarily exceeding 5% power on the APRMs.

The facility recommends removing the statement below from the Critical task.

This Critical Task is not met if the Crew injects too fast and causes power oscillations or APRM downscale clear (5% power).

The Critical task could be reworded to include failure criteria as follows.

During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MSCP, then restore and maintain Reactor water level above

(-) 180 inches.

This Critical task is not met if the crew fails to recognize and take action to secure injection to the Reactor if the conditions of C5-5 or C5-1 S are met.

See attached Plots of Reactor Power, Pressure, and Water level.

BFN CONTINGENCY #5 EOIPM SECTION 0-V-K Unit 0 LEVEL/POWER CONTROL BASES Rev. 0001 Page 35 of 99 1.0 CONTINGENCY #5, LEVEL/POWER CONTROL BASES (continued)

I DISCUSSION: C5-6 I

To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling.

For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

If reactor power is at or below the APRM downscale trip setpoint (**A.23**), it is highly unlikely that core bulk boiling boundary would be below that which provides suitable stability margin for operation at high powers and low flows. (A minimum boiling boundary of 4 ft above the bottom of active fuel has been shown to be effective as a stability control because a relatively long two-phase column is required to develop a coupled neutronic/ thermal-hydraulic instability.) Furthermore, flow/density variations would be limited with reactor power this low since the core has a relatively low average void content. Therefore, there is significant stability margin with power at or below the APRM downscale trip setpoint.

BFN CONTINGENCY #5 EOIPM SECTION 0-V-K Unit 0 LEVELIPOWER CONTROL BASES Rev. 0001 Page 39 of 99 1.0 CONTINGENCY #5, LEVEL/POWER CONTROL BASES (continued)

I DISCUSSION: C5-10 and C5-11 I

Twenty-four inches below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band (**A82**) This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that control of RPV water level with feedwater pumps can preclude the MSIV low water level isolation.

RPV water level must drop to this level before RPV injection can be restarted. Since this level will become the upper limit of the RPV water level control band in subsequent steps, space is provided next to step C5-1 I to record the level.

BFN CONTINGENCY #5 EOIPM SECTION 0-V-K Unit 0 LEVEL/POWER CONTROL BASES Rev. 0001 Page 87 of 99 1.0 CONTINGENCY #5, LEVELIPOWER CONTROL BASES (continued)

I DISCUSSION: C5-25 I

Injection into the RPV is re-established to maintain adequate core cooling.

Irrespective of whether the reactor is shutdown, injection is controlled to make up the mass of steam being rejected through open MSRVs and, if possible, to keep the core submerged. Injection is increased slowly to preclude the possibility of large reactor power excursions due to the rapid injection of relatively cold, unborated water under conditions where the reactor may not be shutdown. The Minimum Steam Cooling RPV Water Level is specified as the lower limit for control of RPV water level to provide the widest possible control band.

The systems listed in this step are the same preferred systems listed earlier in Steps C5-13 and C5-16those that are relatively easy to align, provide high quality water, and either inject outside the core shroud or inject borated water. When RPV injection is restored or increased under failure-to-scram conditions, injection systems should be aligned and operated in the manner that minimizes the potential for core instabilities and power excursions while still accomplishing the objectives of the EOI ATWS strategies.

Injection locations should be selected to optimize mixing and preheating of injection flow and minimize the potential for boron removal from the core.

Injecting into the downcomer steam space provides effective mixing and preheating, thereby reducing core inlet subcooling and minimizing the potential for core instabilities and power excursions. Injecting into the downcomer region below water level and through the jet pumps provides good mixing of injection and recirculation flows but results in greater core inlet subcooling than injection into the downcomer steam space. Injection paths discharging directly into or over the core region should be used only as a last resort due to the higher potential for boron displacement and cold water induced power excursions.

• Systems providing good flow control capability should be used to avoid large, rapid increases in core flow.

• The injection rate should be increased slowly and only when reactor power is not increasing. (Refer to the definition of slowly in EOIPM 0-I-C.)

BFN GLOSSARY OF KEY WORDS AND EOIPM SECTION 0-I-C Unit 0 TERMS Rev. 0001 Page 33 of 35 3.0 ABBREVIATIONS, ACRONYMS, AND DEFINITIONS (continued)

TERM/PHRASE MEANING/DEFINITION Denotes NRC/C [RIMS L44 900706 802]

Slowly Perform a specified action in a careful, controlled manner, making incremental adjustments to avoid undesirable consequences that may result from rapid changes in parameter values. Used specifically to identify the need for caution when restoring or increasing RPV injection under failure-to-scram conditions. If a potential for power excursions exists, RPV injection should be increased gradually while monitoring reactor power. If a sustained increase in reactor power is observed, the injection rate should be stabilized until reactor power is no longer increasing.

A direction to slowly increase RPV injection moderates the rate of change in injection rate but does not itself limit the final steady state flow. No quantitative limit on the rate of change can be defined since the appropriate rate is dependent on the nature of the event and the operating characteristics of available systems. Injection should be increased as slowly as reasonably achievable while still accomplishing the objectives of the PSTG ATWS strategies. Systems providing effective control capability over a wide range of flows permit more gradual changes than those with no throttling capacity. If flow cannot be throttled, options for increasing injection slowly may be limited to initiating systems individually at full flow, starting with those of lowest capacity.

SM Shift Manager SP/L Suppression Pool Level Control SPIT Suppression Pool Temperature Control SRM Source Range Monitor SRO Senior Reactor Operator SRV Main Safety Relief Valve SRVTPLL MSRV Tail Pipe Level Limit STBY Standby

c1I(o3 C/e)j 1501 NRC Exam Unit 2 Simulator X-23 YUAAPRM1 % APRM 1 POWER [2.77209 : 3e-06 6.60201 0.086672}

120 1 1 1 -

2-23 100 80 - .......

60 -

40 -

20 -

0 L_________ A 0 20 40 60 80 100 120 140 160 180 MINUTES 1-X /usr/tmp/plout.25441 2-X NRC-3B.d Tue Jan 20 13:48:09 2015 Tue Jan20 15:27:25 2015

CYe-k) 13 1501 NRC Exam Unit 2 Simulator X-4 ZAOLIO358A INCHES REACTOR LEVEL EMERG SYS RANGE [32.9539: -90.2832: 88.2989 : -48.8395]

I I 1 1 2-4 50 -

0 -

-100

-150 20 40 60 80 100 120 140 160 180 0

MINUTES 1 -X /usr/tmp/plout.22458 2-X NRC-3B.d Tue Jan 20 13:48:09 2015 Tue Jan20 15:24:26 2015

(fJ 1501 NRC Exam Unit 2 Simulator X ZAOPRO353 PSIG REACTOR PRESSURE [0.012213 : 0.012213 958.035 : 38.1275j 1 I I I 21 1400 1200 1000 800 600 400 200 I I 0

160 180 0 20 40 60 80 100 120 140 MINUTES 1-X !usr/tmp/plout.2 1987 2-X NRC-3B.d Tue Jan 20 13:48:09 2015 Tue Jan20 15:23:58 2015

Post Operating Exam Comment #3 JPM Number 631 Restore Offsite Power to 4KV shutdown board at PNL 9-23 U-2 Step 17.1 as currently written isa Critical Step and states:

USE the associated Diesel Generators governor control switch and voltage regulator control switch to reduce generator load to approximately 300 kW and 250 kVAR:

Applicant Performance:

At least one Applicant did not lower the Diesel Generator to 250 kVAR prior to performing Step 17.2 PLACE the associated Diesel Generator breaker control switch in TRIP.

Bases for Post Exam Comment:

There is a caution at 0-01-82 step 8.3[17] that states:

When unloading the Diesel Generator, failure to slowly approach the 300 kW/250 kVAR limit may result in a reverse power trip of the Diesel Generator output breaker.

This caution is the bases for the lower limit while unloading the Diesel Generator.

According to the Siemens 5KV breaker vendor manual (BFN \/TD-S1 06-0040) the Diesel Generator breaker has a normal interrupting capability of 10,000 amps and a short circuit interrupting capability of 29,000 amps.

While Unloading the Diesel Generator to approximately 300KW and 250 kVAR limits the electrical arch produced when the Diesel Generator breaker is opened, the breaker is designed to open and extinguish the electrical arch under more than rated load conditions. No bases was found for an upper limit on kW or kVAR prior to opening the Diesel Generator breaker.

See the attached E-mail from Ron Stowe Mgr, Engineering Operations Support.

Based on a review of the Facility procedures and the vendor manual for the Diesel Generator breaker, the facility recommends that Step 17.1 of the JPM Not be considered a Critical step.

Nichols, Keith A From: Nichols, Keith A Sent: Friday, January 30, 2015 12:51 PM To: Barton, Michael Alan

Subject:

FW: DIG breaker From: Stowe, Ronald H Sent: Friday, January 30, 2015 10:23 AM To: Nichols, Keith A

Subject:

RE: DIG breaker The Diesel Generator Output Breaker is a Siemens 5KV Horizontal Vacuum Circuit Breaker. Type 5-3AF-GEH-250-1200-58 and Type 5-3AF-GEH-250-2000-58 (of which the UNID of the Diesel Generator Output Breaker references) are discussed in Vendor Manual BFN-VTD-S106-0040 The vendor manual discusses 2 parameters that are important concerning the breaker. The first is that the breaker has a normal interrupting capability of 10,000 amps. It also has a Short Circuit interrupting capability of 29,000 amps. Both of these limits are far above the full load amps going thru the circuit breaker. In other words the breaker is well capable of being opened at full load at any time.

Any limit to when the breaker should be opened (separating from the grid) would be just to minimize system perturbations. There is no other limits or restrictions that we could find on this subject.

Post Operating Exam Comment #4 JPM Number 631 Restore Offsite Power to 4KV shutdown board at PNL 9-23 U-2 Step 2 states: VERIFY the associated 4kV shutdown board auto transfer lockout relay is tripped to MANUAL.

Applicant Performance:

At least one Applicant placed the 4KV shutdown bus 1 auto transfer lockout relay in manual.

Bases for Post Exam Comment:

Placing 4KV shutdown bus I auto transfer lockout relay in manual was not in accordance with the procedure in use.

There was no immediate adverse consequence to this action however, if the normal power supply to the 4KV shutdown bus was lost in this condition the shutdown bus would be de-energized and the A and B Diesel Generators would receive an auto start and would supply the A and B 4KV shutdown boards.

O-Ol-57A and 1-45E721 addresses the 4KV shutdown bus I auto transfer logic being inoperable. The only restriction in this condition is to block auto transfer of 4KV unit board IA (see attached).

An EOOS (risk analysis) what if case has been performed for the condition in which the 4KV shutdown bus I is de-energized since this is the worst case without assuming additional failures. The analysis showed that both core damage frequency (CDF) and large early release fraction (LERF) remain Green (acceptable).

The CDF changed from 3.40E-6 to I .32E-5.

The LERF changed from 6.27E-7 to 1 .64E-6.

Based on this information the Facility recommends that no new critical task was created by the incorrect action.

BFN Switchyard and 4160V AC Electrical 0-Ol-57A Unit 0 System Rev. 0156 Page 212 of 213 Illustration 6 (Page 5 of 6)

Limiting Conditions LIMITING CONDITIONS FOR TRANSFER OF 4kV START BUS TO THE SAME CSST AFFECTED BOARD OR BUSS Limiting Unit 1 4kV Unit 2 4kV Unit 3 4kV 4kV Condition Aux Power Aux Power Aux Power Common Start Bus IA Start Bus lB Shutdown UB IC UB 2C UB 3A Sys Sys Sys Bds A & B Start Bus 2A Start Bus 2B Bds Start Buses BK Auto Xfer BK Auto Xfer BK Auto Xfer All 4kV All 4kV All 4kV Normal feed No pre-load No pre-load All SD Bds 1A & lB Must be fed Must be fed Must be fed Boards Boards Boards from USST on Start Bus on Start Bus Aligned 2A & 2B fed from USST from USST from USST Normal feed Normal feed Normal feed (1) 2A 2A Normal from (1) (1) (1) from USST from USST from USST (3) (4) (3) (4) (1)

ACSST (1) (1) (1)

Start Buses BK Auto Xfer BK Auto Xfer BK Auto Xfer All 4kV All 4kV All 4kV Normal feed No pre-load No pre-load All SD Bds 1A & lB Must be fed Must be fed Must be fed Boards Boards Boards from USST on Start Bus on Start Bus Aligned 2A & 2B fed from USST from USST from USST Normal feed Normal feed Normal feed (2) 2A 2A Normal from (2) (2) (2) from USST from USST from USST (5) (6) (5) (6) (1)

B CSST (2) (2) (2)

LIMITING CONDITIONS ASSOCIATED WITH 4kV SHUTDOWN BOARDS AND BUSSES Limiting AFFECTED BOARD OR BUSS Condition UBIA UBIB UB2A UB2B ShutdownBdC Shutdown Bd C on Alt Limit load to 340 amps May exceed 340 amps one time for max feed (SD Bus 1) 100 hrs. (11)

SD Bus 1 BK Auto Xfer to on Alt feed Start Bus (9)

SD Bus I BK Auto Xfer to Auto Xfer Inop Start Bus (8)

SD Bus 2 on BK Auto Xfer to Alt feed Start Bus (7)

SD Bus 2 BK Auto Xfer to Auto Xfer Inop Start Bus (10)

Print Reference Notes (1) 0-45E7150PL2 (3) 1-45E7180PL3 (5) 1-45E7180PL4 (7) 1-45E721 OPL5 (9) 2-45E721 OPL4 (II) 0-45E724-3OPL (2) 0-45E715 OPL 3 (4) 2-45E718 OPL3 (6) 2-45E718 OPL4 (8) 1-45E721 OPL6 (10) 2-45E721 OPL5

033 ITACF 111008027 Ru BKR B Campbell DP Wa Ike r 121211 REVISED PER TACF 111008027 Ri (RETURN TO NORMAL)

REV CHANGE REF PREPARER CHECKER APPROVED DATE SCALE: NONE EXCEPT AS NOTED POWERHOUSE SYSTEM NO.

UNIT 1 575 (202)24 WIRING DIAGRAM 4160V UNIT BDS 1A, 18, 1C SINGLE LINE BROWNS FERRY NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY A

OPL6 - WHEN 4KV UNIT BOARD 2B IS NOT AVAILABLE AS PART OF THE 500KV OFFSITE CIRCUIT TO 4KV SHUTDOWN BUS 1, OR 4KV SHUTDOWN BUS 1 AUTOMATIC TRANSFER LOGIC IS INOPERABLE, BLOCK AUTO TRANSFER OF 4KV UNIT BOARD 1A BY PLACING ITS 43 SWITCH IN MANUAL.