ML030210538
| ML030210538 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/16/2003 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Scalice J Tennessee Valley Authority |
| Shared Package | |
| ML032380265 | List: |
| References | |
| 50-259/02-301, 50-260/02-301, 50-296/02-301 50-259/03-301, 50-260/02-301, 50-296/02-301 | |
| Download: ML030210538 (23) | |
Text
Post-examination Comments (Green Paper)
Licensee Submitted Post-examination Comments BROWNS FERRY EXAM 2002-301 50-259, 50-260, & 50-296 DECEMBER 13, 16-19, 2002
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 Ashok S. Bhatnagar Vice President, Browns Ferry Nuclear Plant December 20,2002 Mr. Luis A. Reyes Regional Administrator Region II U.S. Nuclear Regulatory Commission Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931
Dear Sir:
In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - COMMENTS TO NRC WRITTEN EXAMINATION QUESTIONS Pursuant to the NRC Operator Licensing Examiner Standards (NUREG 1021), Examiner Standard (ES)-402, Attachment 3, TVA submits the attached comments on three questions administered during initial license examinations at BFN on December 13, 2002.
The Enclosure contains comments on questions 27, 44, and 50 in the Reactor Operator examination prepared by your staff.
Pursuant to the ES, these comments were provided in draft form to the NRC Chief Examiner, Mr. Edwin Lea.
There are no new commitments in this submittal.
If you have any questions, please telephone Terry L. Chinn at (256) 729-3439.
Sincerely
.Ashok S. Batnagar PritdnOt n r ecyclet Oaper
Mr. Luis A. Reyes Page 2 December 20, 2002 (w/o Enclosures) cc:
(Via NRC Electronic Distribution)
Mr. Glen M. Tracy, Chief, Operator Licensing Branch Mail Stop O-9-D-24 U.S. Nuclear Regulatory Commission Mail Stop OWFN 0-6D17 Washington, D.C. 20555-0001 Mr. M. E. Ernstes, Chief Operator Licensing and Human Performance Branch U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Edwin Lee, Jr., Inspector Operator Licensing and Human Performance Branch Division of Reactor Safety Region II U.S. Nuclear Regulatory Commission Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931 Mr. S. J. Cahill, Branch Chief, U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
ENCLOSURE 1 Question 27 RO Exam (K/A 223002K3.16)
Unit 3 is in a refueling outage with Shutdown Cooling in operation on RHR Sys I1. A spurious Group I isolation is initiated by the Instrument Techs while performing a surveillance. All isolations occurred as designed.
Which ONE of the following describes the actions to take to allow re-opening 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VLV?
A. Isolation signal has been reset AND either Shutdown Cooling Suction Valve is fully closed.
B. RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton is depressed followed by the group II isolation signal being reset.
C. Either Shutdown Cooling Suction Valve fully closed followed by the RHR SYS IISD CLG INBD INJECT ISOL RESET pushbutton being depressed.
D. RHR SYS IISD CLG INBD INJECT ISOL RESET pushbutton is depressed followed by either Shutdown Cooling Suction Valve being fully closed.
Answer: C Comment In reviewing the question, we believe both "A" and "C" are correct. The basis is as follows:
"A" and "C" are operationally the same correct answer.
In order to manually re-open 3-FCV-74-67 following a Group 2 Isolation signal, at least one of the two Shutdown Cooling Suction Valves must be closed and the Isolation Signal must be reset.
The only method at BFN to reset a Group 2 Isolation Signal to the 3-FCV-74-67, RHR SYS IILPCI INBD INJECT VLV, is to use the RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton and answer "A" implies that this reset method has been performed.
Reference:
3-730E938, Sheet 8, Electrical diagram for Residual Heat Removal System, Circuit B relay logic. OPL171.044, TP-23, Hot License Certification training for the Residual Heat Removal system (RHR), Transparency #23,2-FCV-74-67 Control Circuit.
3-AOI-74-1, Section 4.2, Loss of Shutdown Cooling (Subsequent actions).
Proposed disposition: Accept either "A" or "C" as correct
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OPL1 71.044 Revision 8 10AK63B 2-45E779-22
- (Unit 3 identical On Unit 3 these say "AUTO/OPEN" [3-45E779-22])
and "AUTO/CLOSE" TP-23: 2-FCV-74-67 CONTROL CIRCUIT 3*031595
TITLE: LOSS OF SHUTDOWN COOLING UNIT 3 3-AOI-74-I REV 0009 4.2 Subsequent Actions (continued) (')
4.2.5 IF Shutdown Cooling isolates on low RPV water level or high Drywell press (GROUP 2 ISOL) AND RPV water level needs restoring using LPCI, THEN (Otherwise N/A)
PERFORM the following before reaching -122 inches RPV water level:
4.2.5.1 VERIFY CLOSED the following valves on a Group 2 isolation:
"* RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV, 3-FCV-74-47.
"* RHR SHUTDOWN COOLING SUCT INBD ISOL VLV, 3-FCV-74-48.
"* RHR SYS I LPCI INBD INJECT VALVE, 3-FCV-74-53.
"* RHR SYS II LPCI INBD INJECT VALVE, 3-FCV-74-67.
4.2.5.2 DEPRESS RHR SYS I(II) SD CLG INBD INJECT ISOL RESET, 3-XS-74-126 and 3-XS-74-132 AND VERIFY 3-IL-74-126 and 3-IL-74-132 extinguished.
4.2.5.3 IF the RHR 1oo2 that was in shutdown cooling is needed for RPV water level makeup, THEN PERFORM the following:
4.2.5.3.1 CLOSE RHR PUMP 3A(3B) and 3C(3D) SD COOLING SUCT VLVs, 3-FCV-74-2(25) and 3-FCV-74-13(36).
CAUTION
[NER] Failure to have RHR SHUTDOWN COOLING SHUT OUTBD and INBD ISOL VLVs, 3-FCV-74-47 and 3-FCV-74-48, closed may result in inadvertent draining of the reactor vessel through the RHR PUMP 3A(3B) and 3C(3D) SUPPR POOL SUCT VLVs, 3-FCV-74-1(24) and 3-FCV-74-12(35). [INPO SOER 87-002]
4.2.5.3.2 OPEN RHR PUMP 3A(3B) and 3C(3D) SUPPR POOL SUCT VLVs, 3-FCV-74-1(24) and 3-FCV-74-12 (35) .
Page 3 of 17 3-AOI-74-1
TITLE:LOSS OF SHUTDOWN COOLING UNIT 3 3-AOI-74-1 REV 0009 4.2 Subsequent Actions (continued) (M) 4.2.5.3.3 PLACE RHR SYSTEM I(II) MIN FLOW INHIBIT switch, 3-HS-74-148(149), in NORM.
4.2.5.3.4 VERIFY RHR SYSTEM I(II) MIN FLOW VALVE, 3-FCV-74-7(30),
opens.
4.2.6 IF Primary Containment Integrity is required, THEN (Otherwise N/A)
VERIFY RHR system discharge piping pressure is being maintained >TRM 3.5.4 Limits. REFER TO 3-01-74, Section 8.1.
NOTES:
(1) With the Reactor in Cold Shutdown Condition (Mode 4 or Mode 5), reactor coolant stratification may be indicated by one of the following:
" Reactor pressure above 0 psig with any reactor coolant temperature indication reading at or below 212 0 F.
" Differential temperatures of 50 0 F or greater between either RX VESSEL BOTTOM HEAD (FLANGE DR LINE) 3-TE-56-29 (8) temperatures and RX VESSEL FW NOZZLE N4B END (N4B INBD) (N4B END) (N4D INBD) 3-TE-56-13(14) (15) (16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder, 3-TR-56-4.
" With recirculation pumps and shutdown cooling out of service, a Feedwater sparger temperature of 200 0 F or greater on any RX VESSEL FW NOZZLE (N4B END (N4B INBD) (N4D END) (N4D INBD) 3-TE-56-13(14) (15) (16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder, 3-TR-56-4.
(2) [NER/C] For purposes of thermal stratification monitoring, the bottom head drain line is more representative as long as there is flow in the line.
[GE SIL 251 and 430]
4.2.7 PLOT heatup/cooldown rate as necessary.
REFER TO 3-SR-3.4.9.1(A).
Page 4 of 17 3-AOI-74-1
TITLE:LOSS OF SHUTDOWN COOLING UNIT 3 3-AOI-74-1 REV 0009 4.2 Subsequent Actions (continued) (4) 4.2.8 DIRECT the SRO to ESTIMATE the following times at least once per shift until a method of decay heat removal is restored:
4.2.8.1 DETERMINE the time since shutdown.
4.2.8.2 DETERMINE the current RPV heat-up rate from 3-SR-3.4.9.1(A), or, if reactor coolant stratification is suspected, use Illustration 1. If additional information is required to determine the heat-up rates, contact Reactor Engineer.
4.2.8.3 DETERMINE the reactor coolant temperature or use the last valid reactor coolant temperature available.
4.2.8.4 ESTIMATE the time for reactor coolant temperature to reach 212'F, using data obtained in Steps 4.2.8.1 through 4.2.8.3;,
4.2.8.5 IF the Reactor Vessel head is removed and the cavity is flooded with the fuel pool gates installed, THEN ESTIMATE the time for reactor coolant temperature to reach 125 0 F and 150'F using a plot of. the actual heatup rate or Illustration 1.
4.2.9 IF the loss of Shutdown Cooling is due to inadequate RHRSW flow, THEN (Otherwise N/A)
START the standby RHRSW pump for the appropriate header. REFER TO 0-01-23.
4.2.10 IF the loss of Shutdown Cooling is due to Group 2 PCIS isolation, WHEN conditions which permit resetting Group 2 PCIS isolation are met, THEN (Otherwise N/A)
PERFORM the following:
4.2.10.1 RESET Group 2 isolation by momentarily PLACING PCIS DIV I RESET, 3-HS-64-16A-S32, and PCIS DIV II RESET, 3-HS-16A-S33, in reset.
4.2.10.2 Momentarily DEPRESS RHR SYS I(II) SD CLG INBD INJECT ISOL RESET, 3-XS-74-126 and 3-XS-74-132. VERIFY 3-IL-74-126 and 3-IL-74-132 extinguished.
Page 5 of 17 3-AOI-74-1
Question 44 RO Exam (K/A 264000A1.03)
Diesel Generator 3A is synchronized to 4KV Shut Down Board 3A. The instrumentation readings for the diesel generator are as follows:
voltage: 4160 VAC frequency = 59.8 current = 340 amps vars = 1600 Kvars watts = 2585 KW oil temp = 145 0 F Which ONE of the following actions are required if the diesel is expected to operated for an extended period? (Supply 01-82 illustration #1)
A. The operator must take the voltage regulator control switch to raise to reduce field current.
B. The operator must take the voltage regulator control switch to lower to reduce field current.
C. The operator must take the governor control switch to lower to reduce stator amps.
D. The operator must take the governor control switch to raise to reduce stator amps.
Answer: C Comment In reviewing the question, we believe both "A" and "C" are correct. The basis is as follows:
"A" and "C" both reflect correct responses for these generator conditions. In accordance with 01-82, Standby Diesel Generator System, Step 8.1.12, the required 0.8 power factor may be achieved either by placing the Voltage Regulator Switch to the RAISE position or by placing the Governor Control Switch to the LOWER position in order to reach the diagonal line on Illustration 1, DG KW vs KVAR LOADING.
Since the question only asks for a correct required action (and not the supporting reason), both "A"and "C" are correct.
Note: All four RO candidates selected "A."
Question 44 RO Exam (continued)
Reference:
3-01-82, Standby Diesel Generator System, Section 8.1 , Parallel with System Operation at Panel 9-23.
Proposed disposition: Accept either "A" or "C" as correct
TITLE: STANDBY DIESEL GENERATOR SYSTEM UNIT 3 REV 0072 3-01-82 8.1 Parallel with System Operation at Panel 9-23 (Continued)
NOTE:
Lagging VARS should be maintained when adjusting kW load (rising or lowering) . This may require kW load adjustment to be stopped periodically to allow for adjusting kVAR load. Once desired kW load is achieved, Illustration 1 should be referred to for determination of kVAR loading required to obtain a power factor (pf) of 0.8 lagging. Diesel generator kVAR load should then be adjusted to obtain a 0.8 pf lagging. If system conditions will not permit the kVAR loading required to obtain a 0.8 pf lagging, kVAR I
load should be adjusted to the maximum kVAR lagging the system will allow.
8.1.12 USE the associated Diesel Generator's Governor Control switch and Voltage Regulator control switch to obtain desired kW and kVAR load:
Diesel Instrument Name Instrument No. Panel 3A DG 3A GOVERNOR CONTROL 3-HS-82-3A/3A 3-9-23 DG 3A VOLT REGULATOR CONT 3-HS-82-3A/2A 3B DG 3B GOVERNOR CONTROL 3-HS-82-3B/3A 3-9-23 DG 3B VOLT REGULATOR CONT 3-HS-82-3B/2A 3C DG 3C GOVERNOR CONTROL 3-HS-82-3C/3A 3-9-23 DO 3C VOLT REGULATOR CONT 3-HS-82-3C/2A 3D DG 3D GOVERNOR CONTROL 3-HS-82-3D/3A 3-9-23 DG 3D VOLT REGULATOR CONT 3-HS-82-3D/2A 8.1.13 RECORD time/date loaded on Illustration 2.
8.1.14 MONITOR the offsite sourie that is paralleled with the diesel generator.
8.1.15 IF abnormal voltage or frequency transients are experienced, THEN PERFORM the following:
8.1.15.1 VERIFY OPEN DG 3A (3B, 3C, 3D) output Bkr 1838 (1842, 1832, 1836).
8.1.15.2 PULL UP and RELEASE the associated Diesel Generator control switch in the NORMAL position to initiate the shutdown sequence:
Diesel Handswitch Name Handswitch No. Panel 3A DG 3A CONTROL 3-HS-82-3A/1A 3-9-23 3B DG 3B CONTROL 3-HS-82-3B/IA 3-9-23 3C DG 3C CONTROL 3-HS-82-3C/IA 3-9-23 3D DG 3D CONTROL 3-HS-82-3D/lA 3-9-23 8.1.15.3 REFER TO Section 7.1 and CONTINUE with Shutting down the diesel generator.
8.1.15.4 REFER TO 0-OI-57A and TRANSFER the 4kV shutdown bus to a stable offsite source as deemed appropriate by US.
Page 60 of 89
TITLE: STANDBY DIESEL GENERATOR SYSTEM UNIT 3 3-01-82 ILLUSTRATION 1 REV 0072 (Page 1 of 1)
DIG KW vs KVAR LOADING 3200 - !_ I I-I--------------------
3000- - - - ---
] I I I [ I 4-1 ....
2900 2500 ---- 00 'KW -- - - 1--1AT/1-2400 2200 2000 2000 KIN - 75 PERCENT OF CONTINUOUS RATINGII OII 18001 KW IIA IIIIII 1600 1400
T -- - -- - --
1200 -------00 1 1 1 11 1 1 2200.------------------------- ---- ---- ---- ---
2300 I UI If A ~N DIAGONALI 4--
LINE 1-4--- - - - - II- - -
IA 0.6 pt LAGGINGII OR OUTGOING I goo I 400----------
00 0 200 400 600 800 1000 1200 1400 1600 1300 2000 2200 2400 KVAR (LAGGING OR OUTGOING)
Page 98 of 89
Question 50 RO Exam (K/A 290003G2.3.1 1)
Unit 3 is in a Refueling Outage with fuel movement in progress. The communicator on the phones notifies the Control Room that a fuel bundle was dropped and gas bubbles are visible in the pool. The following indications are received in the Control Room:
FUEL POOL FLOOR AREA RADIATION HIGH in Alarm REFUELING ZONE EXHAUST RADIATION HIGH in Alarm RX BLDG, TURB BLDG, RF ZONE EXH RAD HIGH in Alarm SGTS Running Control Room Ventilation Normal lineup PCIS Group 6 Refueling Zone isolated Which ONE of the following describes the actions that should be taken for these conditions?
A. Stop all fuel moves; Isolate Reactor Zone Ventilation B. Evacuate non-essential personnel from the Refuel floor; Isolate the Control Room Ventilation System C. Stop all fuel moves; Secure SGTS.
D. Evacuate non-essential personnel from the Reactor Bldg; Obtain Operations Manager permission to resume fuel moves.
Answer: B Comment In reviewing the question, we believe both "A" and "B" are correct. The basis is as follows:
There is a high probability that this fuel handling accident would have also resulted in a Reactor Zone Isolation signal. The Reactor Zone HVAC Exhaust Radiation Detectors are physically located on the refuel floor in close proximity to the Refueling Zone HVAC Exhaust Radiation Detectors which was the cause of the REFUELING ZONE EXHAUST RADIATION HIGH alarm. Both Reactor and Refuel Zone HVAC radiation monitors have the same trip set-point. Ifthis setpoint is exceeded, which is evidenced by the fact that the Refueling Zone is isolated, then it could be reasonably concluded that the Reactor Zone should have isolated. The Reactor Zone isolation is an automatic action that should occur if the Reactor Zone Exhaust Radiation Monitor setpoint is exceeded. (see attached drawing for relative location of detectors).
Question 50 RO Exam (continued) (K/A 290003G2.3. 11)
Since the stem of the question does not provide radiation monitor equipment identification numbers (e.g., 3-RM-90-250); actual local dose rates; or Reactor Zone HVAC status, a trained operator would conclude that a high radiation condition exists at the Reactor Zone Radiation detectors. Therefore, verification of a Reactor Zone HVAC Isolation is a correct action.
Additionally, both "A" and "B" include valid immediate Operator required actions based on Abnormal Operating Instruction guidance (see 3-AOI-79-1 attached).
Reference:
3-AOI-79-1, Fuel Damage During Refueling, Section 3.0 (Automatic Actions), Section 4.1.1 and 4.1.2 (Immediate Actions). Mechanical Instrument and Control Drawing, 0-47E600-1, Refuel and Reactor Zone Exhaust Radiation Monitors, (relative location).
Proposed disposition: Accept either "A" or "B" as correct
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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT ABNORMAL OPERATING INSTRUCTION 3-AOI-79-1 FUEL DAMAGE DURING REFUELING REVISION 6 PREPARED BY: Phillip C Chadwell PHONE: 7921 RESPONSIBLE ORGANIZATION: OPERATIONS APPROVED BY: PHILLIP CHADWELL DATE: 08/21/2001 EFFECTIVE DATE: 08/22/2001 LEVEL OF USE: REFERENCE USE QUALITY-RELATED
REVISION LOG Procedure Number: 3-AOI-79-1 Revision Number: 006 Pages Affected: 2, 4 Pagination Pages: NONE Description of Change: IC-007 - BEN PER 01-004639-000 The procedure was revised to replace immediate actions.
Page 2: Replaced the immediate action concerning the stopping of all fuel handling. This action now appears as the first immediate action of this procedure. Replaced the immediate action concerning the evacuation of all non-essential personnel from the Refuel Floor. This action now appears as the second immediate action of this procedure.
REV 0006 TITLE: FUEL DAMAGE DURING REFUELING UNIT 3 3-AOI-79-1 1.0 PURPOSE This instruction provides the symptoms, automatic actions and operator actions for a fuel damage accident.
2.0 SYMPTOMS 2.1 Possible annunciators in alarm:
- FUEL POOL FLOOR AREA RADIATION HIGH (3-XA-55-3A, Window 1).
- AIR PARTICULATE MONITOR RADIATION HIGH (3-XA-55-3A, Window 2).
- RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH (3-XA-55-3A, Window 4).
- REACTOR ZONE EXHAUST RADIATION HIGH (3-XA-55-3A, Window 21).
- RX BLDG AREA RADIATION HIGH (3-XA-55-3A, Window 22).
- REFUELING ZONE EXHAUST RADIATION HIGH (3-XA-55-3A, Window 34).
2.2 Gas bubbles visible, in the Spent Fuel Storage Pool and/or Reactor Cavity, attributed to physical fuel damage.
2.3 Known dropped or physically damaged fuel bundle.
2.4 Portable CAM in alarm.
2.5 Radiation level on the Refuel Floor is greater than 25 mr/hr and cause is unknown.
3.0 AUTOMATIC ACTIONS If annunciator REFUELING ZONE EXHAUST RADIATION HIGH (3-XA-55-3A, Window 34) or REACTOR ZONE EXHAUST RADIATION HIGH (3-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).
Page 1 of 4 3-AOI-79-1
REV 0006 TITLE: FUEL DAMAGE DURING REFUELING UNIT 3 3-AO1-79-1 4.0 OPERATOR ACTIONS 4.1 Immediate Actions 4.1.1 STOP all fuel handling.
4.1.2 EVACUATE all non-essential personnel from Refuel Floor.
4.2 Subsequent Actions NOTE:
The release of iodine is of major concern.
4.2.1 IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s).
4.2.2 VERIFY automatic actions.
4.2.3 NOTIFY RADCON to perform the following:
"* EVALUATE the radiation levels.
"* MAKE recommendation for personnel access.
"* MONITOR around the Reactor Building Equipment Hatch, at levels below the Refuel Floor, for possible spread of the release.
4.2.4 REFER TO EPIP-1 for proper notification.
4.2.5 MONITOR radiation levels, for the affected areas, using the following radiation recorders and indicators:
"* 3-RR-90-1 (points 1 and 2), 3-CONS-90-50A (Address 16),
3-RR-90-142 and 3-RR-90-140 (Panel 3-9-2).
"* 3-RM-90-142, 3-RM-90-140, 3-RM-90-143 and 3-RM-90-141 Detectors A and B (Panel 3-9-10).
"* 3-RI-90-1A and 3-RI-90-2A (Panel 3-9-11).
"* 0-CONS-90-362A (Address 09, 10, 08) for Unit 1, 2, 3-RM-90-250, respectively (Panel 1-9-44).
Page 2 of 4 3-AOI-79-1
REV 0006 TITLE: FUEL DAMAGE DURING REFUELING UNIT 3 3-AOI-79-1 4.2 Subsequent Actions (Continued) 4.2.6 MONITOR portable CAMs & ARMs, if possible.
4.2.7 REQUEST Chemistry to perform 0-SI-4.8.B.2-1 to determine if iodine concentration has risen.
4.2.8 NOTIFY Reactor Engineering Supervisor, or his designee, and OBTAIN recommendation for movement and sipping of the damaged fuel assembly.
4.2.9 OBTAIN Plant Managers approval prior to resuming any fuel transfer operations.
4.2.10 WHEN condition has cleared AND if required, THEN RETURN ventilation systems, including SGTS, to normal. REFER TO 3-OI-30A, 3-OI-30B, 0-OI-30F, 0-01-31, and 0-01-65.
5.0 REFERENCES
5.1 Technical Specifications Section 5.4, Procedures/Section 5.5, Programs and Manuals.
5.2 Final Safety Analysis Report Section 14.6.4, Refueling Accident.
5.3 Offsite Dose Calculation Manual Section 1/2.1.2, Radioactive Gaseous Effluent Monitoring Instrumentation.
5.4 Plant Instructions 3-GOI-100-3, Refueling Operations.
3-ARP-9-3, Annunciator Response Procedures.
EPIP-1, Emergency Plan Classification Logic.
Q-OI-30F, Common and DG Building Ventilation.
3-OI-30A, Refueling Floor Ventilation System.
3-OI-30B, Reactor Building Ventilation System.
Page 3 of 4 3-AOI-79-1
REV 0006 TITLE: FUEL DAMAGE DURING REFUELING UNIT 3 3-AOI-79-1 5.4 Plant Instructions (Continued) 0-01-31, Control Bay and Off-Gas Treatment Building Air Conditioning System.
0-01-65, Standby Gas Treatment System.
SPP-3.1, Corrective Action Program.
SPP-5.8, Special Nuclear Material Control.
OPDP 8, Limiting Conditions for Operation Tracking.
ILLUSTRATIONS/ATTACHMENTS None.
Page 4 of 4 3-AOI-79-1