ML18038B361
| ML18038B361 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/17/1995 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | Ebneter S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML18038B359 | List: |
| References | |
| NUDOCS 9508020025 | |
| Download: ML18038B361 (19) | |
Text
Tennessee Valley Authority. Post Office Box 2000. Decatur, Afabama 35609 July 17, 1995 Mr. Stewart D. Ebneter Regional Administrator ATTN:
Branch Chief, Operator Licensing U. S. Nuclear Regulatory Commission Region II, Suite 2900 101 Marietta Street NW Atlanta, Georgia 30323
Dear Sir:
In the Matter of Tennessee Valley Authority Docket Nos.
50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)
NRC WRITTEN EXAMINATION COMMENTS Pursuant to the NRC Operator Licensing Examiner Standards (NUREG-1021),
TVA is submitting post examination comments to the NRC written examination that was administered to BFN personnel on June 23, 1995.
This submittal documents draft comments provided by TVA (Dale E. Hill, BFN Operator Training Manager) on June 29, 1995, to Edwin Lea of your staff.
If you have any questions, please contact Dale E. Hill at (205) 729-3439.
Sincere y, Salas Manager of Site Licensing cc:
See page 2
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July 17, 1995 Enclosure cc:
Mr. Stewart A. Richards (w/o Enclosure)
Chief, Operator Licensing Branch, DLPQ U.S. Nuclear Regulatory Commission MS OWFN 10D-22 Washington, D.C.
20555 Mr. Edwin Lea Chief Examiner U.S. Nuclear Regulatory Commission Region II, Suite 2900 101 Marietta Street, NW
- Atlanta, GA 30323 Mr. Mark S. Lesser, Acting Branch Chief U.S. Nuclear Regulatory Commission
.Region II, Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12 Box 637
- Athens, Alabama 35611
ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1q 2i AND 3 BFN - NRC WRITTEN EXAMINATION COMMENTS SEE ATTACHED
QUESTION 1 (SRO/RO)
Which ONE of the following describes a characteristic of a fill-oilleak from a Rosemont transmitter?
a.
Instrument responds more slowly until the diaphragm contacts the convolution plate, when it fails to function properly.
b.
Instrument responds more rapidly until the diaphragm contacts the convolution plate, when it fails to function properly.
c.
A perceptible oscillation of instrument readings over a period of time using calibration data and check the "as found" vs "as left" data.
d.
A sustained drifting of instrument readings in random directions over a period of time using calibration data and check the "as found" vs "as left" data.
Answer key:
a There are two correct answers to this question:
Answer "a" is correct, however, the instruments slow response would only be noted during a transient condition or by increasing or decreasing the applied pressure during a test calibration.
This is stated on page 18 of 69 of the lesson plan, OPL171.003:
"Slow response to or inability to follow planned plant transients or slow response to either an increasing of decreasing test pressure - most likely detectable during calibration procedures..."
During static conditions this would not be observed.
The instrument would still provide accurate indications until the transient occurred.
Answer "d" is also correct, depending upon where the filloil leak is located.
One Rosemount instrument could respond by exhibiting a sustained drift high, while another Rosemount instrument monitoring the same parameter could respond by exhibiting a sustained driftlow.
Accept answer a or d.
OPL17 L003 Revl.itoll 9 Page 18 ut 69 (6)
Slow response to or inability to follow planned plant transients or slow response to either an increasing or decreasing test pressure - most likelydetectable during calibration procedures as the IM's compare the response of each transmitter to others monitoring the same parameter Inability to respond over the entire design range - most likely detectable during calibration procedures A listing of Rosemount transmitters at'f'ected at TVAare listed in Table 3 These transmitters are not necessary to be replaced because they are low press transmitters and are not suceptible to this type of failure i.
Summary of Characteristics of leaking monitor
'nstrument willcontinue to give accumte reading until the isolating diaphragm makes contact with the convolution plate.
Instrument response willbe slower as fluid is lost.
(3)
(4)
Instrumentation in Main Control Room (MCR) is not accurate enough to give early detection of leaking fluid.
Problems willbe detected via the computer monitoring system with back-up from normal calibration procedures.
Reactor vessel instrumentation description Vessel Level Instrumentation Level instruments are dp cells. The level signal is obtained by comparing:
The pressure exerted by the actual Obj. V.B.7.
height of water in the vessel downcomer (variable leg) to
QUESTION 59 RO (1.0)
Which ONE of the following describes the response of the main turbine that willoccur for BOTH a load reject condition AND a turbine trip?
a.
Load selector automatically runs back to 0%.
b.
EHC speed control network controls turbine speed to prevent turbine from overspeeding.
c.
Generator voltage regulator remains in automatic, d.
Turbine valve position/speed recorder shifts from turbine speed to valve position indication.
Answer key: a This question has two correct answers, a or b.
Answer "a" is correct, the EHC load selector begins running back toward zero load as the result of a load reject signal.
Answer "b" is also correct, the CIVs throttle to control turbine speed at 1800 RPM via the EHC speed control network.
This is stated on page 25 of 63 of lesson plan OPL171.010 (2.c.(6)).
Also on page 11 of 63 (3.a.(1) and (2)) the lesson plan states the purpose of the CIVs is: "To protect the turbine from overspeeding during a generator trip or load reject (load dump).
The overspeed might occur even though the Stop and Control Valves close, should fiashing of moisture to steam occur in the moisture separators."
TP-1 of the EHC Unit shows that following a turbine trip the Intercept Valve Demand receives a zero input value to close the intercept valves.
Accept answer a or b.
OPL171.010 Revision 4
Page 25 of 63 2.
Generator Load Reject a.
Definition INSTRUCTOR NOTFS A greater than 40-percent mismatch between the main generator electrical output and turbine power.
(turbine power > generator power)
Obj. V.B.B.
Comparison is made between stator amps and turbine crossover pressure.
(LP turbine inlet pressure) b.
Cause of a load reject Manual opening of main generator output circuit breakers (PCBs), generator breaker or loss of grid load.
c.
Results of a load reject Obj. V.B.9.
(1)
Turbine Control Valves are tripped closed by the fast acting solenoids.
(2)
Bypass Valves open within their capacity to accept steam that was going to the turbine.
(3)
EHC load selector begins running back toward zero load.
(4)
Control Valves partially reopen when load-mismatch is <40 percent.
(5)
Load selector runback stops when the load-reject condition is cleared.
(6)
CIVs throttle to control turbine speed at 1800 RPH via the EHC speed control network.
3.
(7)
No turbine trip occurs.
Reactor Scrams From The Turbine Refer to reactor protection
- system, OPL171.028.
OPL171.010 Revision 4
Page 11 of 63 (2)
Provides control For rolling, synchronizing and loading of the turbine generator.
b.
Construction INSTRUCTOR NOTES (1)
Valves are welded directly to their respective Stop Valves.
(2)
Each valve is controlled by the EHC system through an operator at the bottom of the valve.
(3)
Hydraulically opera Led open and
- closed, spring loaded closed (4)
Balanced with internal pilot and balance chamber.
During valve operation the internal pilot valve opens first.
(a)
Steam is bled past the stem and out the pilot valve seat to the outlet, reducing balance chamber pressure.
(b)
The valve then opens against a lower 4P.
Combined Intermediate Valves (CIVs Nos.
1 through 6)
TPW a ~
Purpose (1)
To protect the turbine from overspeeding during a generator trip or load reject (load dump).
(2)
The overspeed might occur even though the Stop and Control Valves close, should flashing of moisture to steam occur in the moisture separators.
Pressure drops within the turbine piping and moisture separators due to exposure to condenser vacuum would be the cause of this f1 a shing.
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Question ¹70 RO / 61 SRO (1.0)
An event on Unit 2 has required entry into the EOIs.
Reactor water level is being controlled by RC/L.
Plant conditions are as follows:
RPV pressure RPV level Suppression pool temperature PSC pressure Suppression pool level 200 deg. F 10 feet 200 psig
-175 inches 5 psig Which ONE of the following describes the use of RHR in accordance with EOIs?
One RHR pump in each loop may be started regardless of NPSH and vortex limits.
b.
Each RHR pump started is limited to 10,500 gpm.
C.
d.
Only one RHR pump per loop can be started and each pump is limited to 10,500 gpm.
Both RHR loops are available for unrestricted operation.
Answer key: a This question has three correct answers (a, b, and d).
For obvious reasons, answer "a" is correct, Answer "b" is also correct, due to the fact that the RHR pump during normal conditions is limited to 7000 to 10000 gpm flow. However, with a LPCI signal present the LPCI injection valves willbe wide open and the pump could be expected to deliver 10500 which could be assumed to be max flow on one pump.
Answer "d" is also correct, there is no limitation to the number of pumps that are used on a loop.
The EOI program manual supports this in Section V-G page 19 of 52.
The discussion is to clarify that one Core Spray pump (or RHR) is considered one injection subsystem, for RHR only one pump per loop is needed for an injection subsystem, but does not exclude using both RHR pumps in a loop, and the disregarding of NPSH'and vortex limits still apply, so there would be no restriction to flow.
Question should be deleted from the exam.
C1. ALTERNATE LEVEL CONTROL BASES EOI PROGRAM MANUAL SECTION V-G e
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The operator is given a list ofinjection subsystems which caa be used to inject into the RPV.
No priorityregarding the use ofthe listed subsystems is intended; however those which provide the. largest volume ofwater at the highest RPV pressure are preferred.
The operator should use the most appropriate means available under current plant conditions.
EOI Appendix 6A provides step-by-step guidaace to operate the Condensate system forRPV injection. This is a preferred source ofinjection because itis used for RPV water level control during aormal plant operations, and provides the highest quality water at the highest pressure.
EOI Appendices 6B and 6C, and EOI Appendices 6D and 6E provide guidance to operate LPCI Systems I and II, and Core Spray Systems I and IIrespectively. It is important to note that the individual LPCI and CS injection subsystems each need only one pump operating in order to meet the requirement ofan operating injection subsystem.
For example, CS System I is considered only one operating iajection subsystem with either one pump or both pumps operatiag in the loop. This is an accepted deviation from the requirements ofTechnical Specifications which require two pumps within a loop to be operating.
EOI Appendices 6B and 6C direct the operator to initiate RHRSW Qow through the RHR heat exchangers followingLPCI system initiation. The cooling capacity ofthis injection subsystem should be used as soon as conditions permit.
REVISION 2 PAGE 19 OF 62.
SECTION VN
C
Question ¹89 RO / 80 SRO (1.0)
Fuel is being loaded into the core on Unit 2, when SRMs begin steadily increasing.
Allrods were inserted prior to beginning refueling.
but currently two control rod position indications are inoperable.
Which ONE of the following is the required operator actions per the AOIs?
a.
Scram the reactor, remove fuel bundle from core, ifSRMs continue increasing when bundle clears the top of the core, then evacuate refuel floor.
b.
Scram the reactor, ifSRMs slowly decrease, leave the fuel bundle in the core, disengage and raise the grapple, then move refuel bridge away from the core and discontinue refueling.
c.
Remove fuel bundle from core, ifSRMs continue increasing traverse to the area of the cattle shoot with the refueling bridge, then evacuate the refuel floor.
d.
Remove fuel bundle from core, ifSRMs slowly decrease place the bundle in the spent fuel pool, disengage and raise the grapple, then move refuel bridge away from the bundle.
Answer key: c This question does not have a correct answer.
2-AOI 2 immediate actions for unexpected criticality when moving a control rod requires the rod to be reinserted, and ifthe reactor still cannot be determined to be subcritical then a manual scram should be inserted.
This would be conservative and the prudent action, since the positions of two rods are unknown (rod position indications are inoperable).
With rod position indications inoperable a control rod drift would be undetected, and one possible symptom of this condition might be the reactor unexpectedly going critical, Since the position of two control rods are unknown, the prudent action would be to initiate a manual reactor scram.
However none of the distractors that include a manual scram have correct actions following the scram.
The answer key correct answer is correct for unexpected criticality when inserting a fuel bundle, but does not include a manual scram.
Question should be deleted from the exam.
TITLE:
INADVERTENT CRITICALITY DURING INCORE FUEL MOVEMENTS REV 0008 UNIT 2 2-AOI-79-2 4.1 Immediate Actions (Continued) 4.1.3.2 IF the reactor can be determined to be subcritical AND no radiological hazard is apparent, THEN PLACE the fuel assembly in a spent fuel storage pool
.location with the least possible number of surrounding fuel assemblies, leaving the fuel grapple latched to the fuel assembly handle.
4.1.3.3 IF the reactor CANNOT be determined to be subcritical OR adverse radiological conditions exist, THEN TRAVERSE the refueling bridge and fuel assembly away from the reactor core, preferably to the area of the cattle chute, AHD COHTIHUE at Step 4.1.4.
4.1.4 IF the reactor CANNOT be determined to be subcritical OR adverse radiological conditions exist, THEN EVACUATE the refuel floor.
4.2 Subse uent Actions 4.2.1 NOTIFY the SOS and Reactor Engineer.
4.2.2 IF any EOI entry condition is met, THEN ENTER the appropriate EOIs.
4.2.3 VERIFY all control rods are inserted.
4.2.4 IF criticality is still evident AND at the direction of the
- SOS, THEN PERFORM the following:
4.2.4.1 STOP the CRD pump (if in operation).
4.2.4.2 ISOLATE RWCU System (if in operation).
4.2.4.3 INITIATE SLC (if operable).
419(208)
Page 3 of 5 2-AOI-79-2
TITLE:
INADVERTENT CRITICALITY DURING INCORE UNIT 2 FUEL MOVEMENTS REy P P 08 2-AOI-79-2 3.0 AUTOMATIC ACTIONS 3.1 If annunciator REFUELING ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, Window
- 34) or REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, Window 21)
- alarm, the following will occur:
~ Control Room and Refuel Zone Ventilation Isolates.
~
SGTS initiates.
~ Emergency Control Room Pressurization units start.
~ PCIS Group 6 isolates (partial isolation only for Refueling Zone Exhaust High Radiation).
~ Reactor Zone Ventilation isolates (initiates on Reactor Zone Exhaust High Radiation only).
3.2 If annunciators REACTOR CHANNEL A AUTO SCRAM (2-XA-55-5B, Window 1) and REACTOR CHANNEL B AUTO SCRAM (2-XA-55-5B, Window 2) alarm, a Reactor Scram will occur.
4.0 OPERATOR ACTIONS 4.1 Immediate Actions 4.1.1 IF unexpected criticality is observed following control rod withdrawal, THEN REINSERT the control rod.
4.1.2 IF all control rods CANNOT be fully inserted, THEN MANUALLY SCRAM the reactor.
4.1.3 IF unexpected criticality is observed following the insertion of a fuel assembly, THEN PERFORM the following:
4.1.3.1 VERIFY fuel grapple latched onto the fuel assembly handle AND immediately REMOVE the fuel assembly from the reactor core.
419(208)
Page 2 of 5 2-AOI-79-2
1 Question 99 RO/SRO (1.0)
A plant emergency is in progress.
The control room desires to communicate with an operator in the Reactor Building.
Which ONE of the following is the PRIMARYPortable Radio channel that should be used?
(Assume a Motorola SABER radio is being used.)
a.
Channel 2 b.
Channel 4 c.
Channel 5 d.
Channel 6 Answer key: b This question has two correct answers when considering the use of Motorola Saber radios in the plant. In the Conduct of Operations procedure SSP-12.1, page 47 of 88, paragraph 3.8.4. B states that for emergencies Fl (Channel 1), F2 (Channel 2), F3 (Channel 3), or F4 (Channel 4) are available for communications.
In daily operations of the plant the operations personnel use Fl and or F2 for communications.
During an emergency situation the operations personnel would normally already be on either of these channels, and the Conduct of Operations procedure does allow them to use F2 during an emergency.
Since F2 was one of the distractors and is allowed by the procedure F2 would be an appropriate channel to use during an emergency.
Ifon F2 and there was problems with communications the operators would swap around on the channels until they found one that worked, which would be Fl or F4.
Answer a or b should be accepted.
~P-12.1 Rev 21 Page 47 of 88 3.8A Radios (Continued)
B.
For emergencies, or other plant situations and conditions, portable 2-way radio communications is available via the F1, F2, F3, and F4 repeater systems.
The SABER radios are also capable of simplex or direct radio-to-radio communications on Ch 5, as well as provisions to communicate with Nuclear Security on Ch 6. Ch 6 should be used very judiciously so as not to interfere with plant security functions.
1
~
F1, which is designated Ch 1 on the Motorola SABER Series portable, provides wide area coverage for the turbine building, reactor building, main control room, and service building areas.
2.
F4, which is designated Ch 4 on the Motorola SABERs, provides primary wide area coverage for the reactor building and control bay areas.
3.
F1 and F4 will have self-contained backup battery power supplies to provide coverage during a power loss. Also an F3 and F4 repeat system are installed and accessed from the Saber radios on Ch 3 and Ch 4, respectively.
C.
In regards to plant security, 10 CFR 73.21 prohibits the transmission of safeguard information except on protected frequencies.
Ifprotected frequencies are not used, routine radio transmissions between site security personnel should be limited to message formats that do not disclose facilitysafeguard features or response features, 3.8.5 Other Communication Systems A.
The load dispatcher communication system provides a rapid means of communication for dispatcher requests or transmission of information on plant conditions.
B.
The NRC Emergency Notification System facilitates immediate notification to the NRC of significant incidents and provides continuous and uninterrupted communication between the plant and NRC.
C.
Sound-powered phones are the preferred communication system during surveillance testing, post-maintenance testing, or special operational situations that require coordination through communication, D.
Private Autqmatic Exchange (PAX) System is provided for routine communication.
Calls to the main control room shall be terminated after the fifth ring unless the information is important to plant operation.
3.8.6 Abbreviations and Acronyms Only abbreviations and acronyms obtained from Appendix H of SSP-2.2 should be used in plant communications.
Both written and spoken terms should be prescribed in the lists.