ML023440122
| ML023440122 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 12/04/2002 |
| From: | Pace P Tennessee Valley Authority |
| To: | Reyes L Division of Nuclear Materials Safety II |
| References | |
| -nr, 50-390/02-301, 50-391/02-301 | |
| Download: ML023440122 (32) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Spnng City, Tennessee 37381-2000 10 CFR 55.40 DEC 0 4 2002 Mr. Luis A. Reyes Regional Administrator, NRC Region II Atlanta Federal Center 61 Forsyth St., Suite 23T85 Atlanta, Georgia 30303
Dear Mr. Reyes:
In the Matter of the
)
Docket No. 50-390 Tennessee Valley Authority
)
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - REACTOR AND SENIOR REACTOR OPERATOR INITIAL EXAMINATIONS 390/2002-301 Beginning on November 26, 2002, license examinations were initiated to a group of reactor operator (RO) and senior reactor operator (SRO) candidates at WBN. Provided in the enclosure to this letter are post-examination comments related to 3 questions on the written examinations. These comments are provided in accordance with Examination Standard (ES) 501, "Initial Post-Examination Activities," of NUREG 1021, "Operator Licensing Examination Standards for Power Reactors."
Should you require additional information regarding this matter, please contact Randy Evans at (423) 365-8989.
Sincerel P. L. Pace Manager, Site Licensing and Industry Affairs Enclosure cc: Page 2 DEC 6 Z)0?
Pented on reycled paper
U.S. Nuclear Regulatory Commission Page 2 DEC 0 42002 Enclosure cc (w/o Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. L. Mark Padovan, Senior Project Manager U,S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 cc (Enclosure):
U.S. Nuclear Regulatory Commission, Region II ATTN: Mr. Michael E. Ernstes Chief, Operator Licensing and Human Performance Branch Sam Nunn Atlanta Federal Center 61 Forsyth St., Suite 23T85 Atlanta, Georgia 30303 U.S. Nuclear Regulatory Commission, Region II ATTN: Mr. Lee R. Miller Sam Nunn Atlanta Federal Center 61 Forsyth St., Suite 23T85 Atlanta, Georgia 30303
Enclosure Comments Related to Examination Questions
FACILITY COMMENTS FOR NRC WRITTEN EXAMINATION WATTS BAR 11/26/2002 NUCLEAR Question #
Comment Associated References SRO 6 Accept two correct answers. Both alternative AOI-2, "Malfunction of C and D are correct. The stem asks for the Reactor Control System",
correct method to realign the rod to its section 3.4 associated bank, thus both C and D alternatives provide correct steps to accomplish realignment.
The differences between the alternatives is simply a matter of how many procedural steps are listed in the alternative. There are no incorrect elements in alternative D.
Watts Bar 2002-301
ý SROFInitT-Exam
- 6. Unit 1 was at 25% power and ramping up when the RO noticed that one of the Bank C control rods is 13 steps below the other rods in Bank C which are at 215 steps.
At 0900, immediately after discovery, power assension was halted.
At 0945, the rod was determined to have an electrical problem which was repaired.
At 1015, the management staff has concurred with realignment of the misaligned rod In accordance with AOI-2, Malfunction of the Reactor Control System.
Which ONE of the following outlines the method of realigment?
A. Record information from Bank Overlap Unit, step counters, and P/A converter. Disconnect lift coil for theaffected rod, reset step counters, select Bank C and insert Bank C control rods.
B. Disconnect lift coil of the affected rod, select Bank C and insert Bank C.
C. Record information from Bank Overlap Unit, step counters, and P/A converter. Disconnect all lift coils in Bank C except the affected rod, reset step counters, select Bank C and withdraw affected control rod.
D. Disconnect all lift coils in Bank C except the affected rod, select Bank C and withdraw the affected rod.
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT ABNORMAL OPERATING INSTRUCTION AOI-2 MALFUNCTION REACTOR CONTROL OF SYSTEM Revision 25 Unit 1 QUALITY RELATED REQUESTED BY:
C. Dale Greer SPONSORING ORGANIZATION:
APPROVED BY:
Operations Benjamin F. McNew, Jr.
Effective Date:
LEVEL OF USE: CONTINUOUS 10/31/02
OPERATOR ACTIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 3.1 Diagnostics 3.0 IF GO TO Subsection
,Continuous Rod Withdrawal/Insertion 3.2 Instrument failure (e.g. T-avg, NIS, PT-1-73) with Rod Control in MAN 3.2 Dropped RCCA 3.3 RCCA Misalignment 3.4 Rod Position Indicator (RPI) Malfunction 3.5 Failure of Control Rods to Move in AUTO 3.6
RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Control rods in MAN position will misaligned RCCA.
DETERMINE if affected bank can be aligned to misaligned RCCA(s) within one hour:
control bank,
"* Misaligned RCCA(s) in control bank.
"* Bank overlap can be maintained during alignment.
The misaligned RCCA is above the affected banks insertion limit.
be used to align affected bank to IF greater than one hour will be required to align RCCA OR misaligned RCCA(s) NOT in THEN
- GO TO NOTE prior to Step 21.
0 Reactor engineering agrees.
NOTE 12.
3.4
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 26 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Rod movement in the inserted direction is permissible if rod movement does not cause further misalignment.
The following aligns misaligned RCCA to affected bank.
ADJUST turbine load to MAINTAIN T-ref and T-avg within 30.
DETERMINE actual RCCA position as follows:
- a.
REQUEST Reactor Engineering to evaluate core anomaly and RCCA position.
- b.
NOTIFY STA to perform 1-SI-0-21, Excore QPTR.
IF Step 22 indicates RPI failure, THEN
- GO TO Section 3.5 Rod Position Indicator (RPI) Malfunction.
3.4 NOTE
- 21.
- 22.
23.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 27 of 49 RCCA Misalignment (Continued)
ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 24 IF two or more RCCAs are misaligned by greater than 12
- steps, THEN
- GO TO AOI-39, Rapid Load Reduction.
REFER TO Tech Specs:
"* 3.1.5, Rod Group Alignment Limits.
"* 3.1.6, Shutdown Bank Insertion Limits.
"* 3.1.7, Control Bank Insertion Limits.
"* 3.2.3, Axial Flux Difference (AFD).
- 26.
INVESTIGATE cause of misalignment.
3.4 25.
I AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 28 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED INITIATE repairs to failed equipment.
NOTIFY Operations Duty Manager of misaligned RCCA prior to recovery.
NOTIFY Reactor Engineering and STA to provide realignment rate considering, as a minimum, the following:
Length of time the RCCA has been misaligned.
Power level at which recovery will be performed.
DO NOT CONTINUE UNTIL initiating problem has been corrected.
3.4
- 27.
- 28.
- 29.
30.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 29 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Computer Points for individual rods are IZM-85-5001 through 5053, and points for banks are U0049 through U0056.
IF any RCCA misaligned by greater than Tech Spec allowed limit, THEN OBTAIN approval from Plant Manager or designate prior to raising reactor power.
DO NOT CONTINUE UNTIL the following agree on realignment that involves a reactor power rise:
Unit SRO.
Shift Manager.
Reactor Engineering.
STA.
3.4 NOTE
- 31.
32.
3.4 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 33.
RECORD value of the following just prior to disconnecting lift coils:
"* Affected bank's step counter(s).
Computer for RCCAs in affected bank.
"* P/A converter for bank (N/A for shutdown bank)
[Aux Inst Room, el 708, 1 -PNL-85-R42].
"* Bank overlap counter (N/A for shutdown bank)
[control rod drive room, el 782, panel 1-L-122].
NOTE Toggle switch up is disconnected. Toggle switch down is connected.
- 34.
DISCONNECT all lift coils in affected bank, EXCEPT for misaligned RCCA.
NOTE Step counter(s) with lift coils disconnected on all rods in the group will not step and need not be reset.
- 35.
PLACE affected step counter(s) to position determined in Step
- 22.
- 36.
PLACE control rods to bank select for affected bank.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 31 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED The following step will cause a CONTROL ROD URGENT FAILURE alarm [86-A].
ALIGN RCCA to affected bank position:
USE rod control to position misaligned RCCA to affected bank position determined in Step 33.
ADJUST turbine load to MAINTAIN T-ref and T-avg within 30.
IF RCCA can NOT be aligned, THEN:
- a. RECONNECT lift coils of affected bank
- b. RESET CONTROL ROD URGENT FAILURE alarm [86-A] using 1-RCAR.
- c. SET affected group step counters to original value.
- d.
RESET control bank P/A converter to its original value USING Attachment 1 if misaligned RCCA in control bank.
- e. COMPLY with Tech Specs:
"* 3.1.5, Rod Group Alignment Limits.
"* 3.1.6, Shutdown Bank Insertion Limits.
"* 3.1.7, Control Bank Insertion Limits.
- f.
ENSURE control rods in MAN.
- g.
NOTIFY Plant Management and Reactor Engineering.
- h.
RETURN TO Instruction in effect.
3.4 NOTE 37.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 32 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 38.
3.4 RECONNECT lift coils of affected bank.
ENSURE the following are reading correct:
Bank overlap counter.
P/A converter display USING.
Group step counters.
Computer points.
RESET CONTROL ROD URGENT FAILURE alarm [86-A] using ROD CONTROL ALARM RESET pushbutton 1-RCAR.
ENSURE control rods in MAN.
RESTORE T-avg and T-ref to within 30F.
39.
40.
- 41.
42.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 33 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED WHEN plant stabilized, THEN PERFORM 1-SI-85-2, Reactivity Control Systems Movable Control Assemblies (Modes 1 and 2) for the affected bank.
PLACE rods in AUTO, if desired.
RETURN TO Instruction in effect.
3.4
- 43.
- 44.
45.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 34 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Steps 46 through 58 are actions to be performed if unit is in Modes 3 through 5. This section places the affected RCCA and bank to the fully inserted position.
STOP rod movement.
OPEN reactor trip breakers.
NOTIFY the following of misaligned RCCA:
"* Unit SRO.
"* SM.
"* Reactor Engineering.
"* STA.
INVESTIGATE cause of misalignment.
INITIATE repairs to failed equipment.
NOTIFY Operations Duty Manager of misaligned RCCA prior to recovery.
DO NOT CONTINUE UNTIL initiating problem has been corrected.
3.4 NOTE
- 46.
- 47.
48.
- 49.
- 50.
51.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 25 Page 35 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED PLACE ROD BANK SELECT to affected bank select position.
INSERT affected bank UNTIL all of the following:
NO rod motion observed on any rods in bank.
"* Misaligned RCCA RPI indicates rod at bottom.
"* Misaligned rod, rod bottom light LIT.
SET affected step counters to "000".
- 55.
EVALUATE effect on the following:
"* Bank overlap.
"* P/A converter.
"* Bank position and required stepping sequence.
3.4 52.
53.
54.
AOI-2 WBN MALFUNCTION OF REACTOR CONTROL SYSTEM Revision 26 Page 36 of 49 RCCA Misalignment (Continued)
ACTION/EXPECTED RESPONSE CONSIDER the following:
"* RESET P/A converter USING.
"* NOTIFY MIG to reset bank overlap counter.
UPDATE computer.
RETURN TO Instruction in effect.
RESPONSE NOT OBTAINED
- END OF SUBSECTION -
3.4
- 56.
- 57.
58.
FACILITY COMMENTS FOR NRC WRITTEN EXAMINATION WATTS BAR NUCLEAR 11/26/2002 Question #
Comment Associated References SRO 32 Delete question from the examination. The Annunciator Response stem does not provide conditions where any Instruction for annunciator alternative is correct. The stated conditions window 92-C, CVCS lesson where letdown is isolated with Charging Flow plan 3-OT-SYS062A Controller, 1-HIC-62-93, rising, will actually excerpt.
result in a rise in Pressurizer level as opposed to the stated condition that Pressurizer level is "dropping slowly". The reason Pressurizer level will rise under these conditions is l-FCV-62-93 does not fully close since it has a minimum stop that prevents charging flow from dropping to less that 35 gpm (RCP seal protection).
Watts Bar 2002-301 SRU InaI Exam
- 32.
Given the following:
-Unit 1 is at 100% power
-Pressurizer level is dropping slowly
-The output of the Pressurizer Level Controller, 1-LIC-68-339, is rising
-The Charging Flow Controller, 1-HIC-62-93A, is rising
-Charging flow dropping.
-Annuciator 92C, PZR LEVEL LO-HTRS OFF & LTDN CLOSED is illuminated Which ONE of the following is the cause of the Pressurizer level decrease?
A. 1-FCV 93, charging flow control valve has developed a diaphram leak.
B. The Charging Flow Controller output, 1--HIC-62-93A, is failing high.
C. The PZR Level Controller, 1-LIC-68-339, is failing high.
D. The Tavg input to the PZR is failing high.
SOURCE I -LS-68-335E/D 1 -LS-68-339D Probable Cause:
- 1.
- 2.
- 3.
4.
Corrective Action:
References:
92-C SETPOINT 17%
Insufficient charging flow Excessive letdown flow RCS leak Instrument malfunction (level or Tavg input)
[1]
CHECK PZR level indication on 1-M-4:
"" 1-LI-68-320
"* 1-LI-68-335A
"* 1-LI-68-339
[2]
IF LETDOWN is NOT in service, THEN ISOLATE CHARGING
[3]
CHECK PZR level and reference level on 1-LR-68-339 [1-M-5].
[4]
IF Malfunction Of Pressurizer Level Control System, THEN GO TO AOI-20, MALFUNCTION OF PRESSURIZER LEVEL CONTROL SYSTEM.
[5]
IF level is low, THEN
[a]
VERIFY All PZR heaters OFF.
[b]
VERIFY Letdown orifice 1-FCV-62-72, -73 & -74, CLOSED.
[c]
VERIFY Letdown isolation 1-FCV-62-69 and -70, CLOSED.
[6]
IF PZR level control system cannot maintain level to program, THEN REFER TO AOI-6, SMALL REACTOR COOLANT SYSTEM LEAK.
[7]
REFER TO Tech Specs.
1-45W600-57-15, 1-47W610-68-5, 1-45W611-68-2, 1-45W760-68-5 AOI-6 AOI-20 WBN Page 31 of 48 ARI-88-94 I
I Rev 13 PZR LEVEL LO-HTRS OFF &
LTDN CLOSED
3-OT-SYS062A Revision 5 Page 36 of 93 X.
LESSON BODY (2)
HS(s) in MCR, switchgear, and locally. Currently operate with valves open and power removed.
(3)
Fail "as is".
(4)
A shunt breaker is installed on each of these valves to prevent spurious actuation. MCR will still have indication of valve position.
ram. FCV-62 CCPs Charging Flow Control (1) Flow setting derived from PZR level control system.
(2)
HIC(s) located in MCR, Aux CR, locally, and Aux Inst RM (R-18).
(3) Fails open.
(4) Local Control Local manual control is via HIC-62-93B. When selected for local a red light beside MCR controller is illuminated.
(5) Manual bypass valves for FCV-62-93 are provided on the discharge of each CCP.
These allow an alternate means of controlling charging and seal injection flow in the event the normal charging flow control valve failed.
INSTRUCTOR NOTES 45W600-62-5 47W610-62-2 47W611-62-4 See SOI-62.01 for discussion of local operation Objective 14 Bypass for FCV-62-93 enters charging line downstream of the charging flow element. If bypass valves were leaking by, a mismatch in normal charging/letdown flows could exist.
3-OT-SYS062A Revision 5 Page 37 of 93 flltn'irx.u.Wt S\\or INSTRUCTOR NOTES (6)
A pneumatic relay is installed on this valve to prevent closing from spurious signals. The t
stop ensures minimum of 35 gpm charging flow. A bypass is provided to allow effective flow control at low RCS pressure.
nn.
FT-62-93A: Provides MCR and local flow indication. Provides control signal for FCV-62-93.
FS-62-93-A/B provides input, with ERFDS, to the "Charging Flow Hi/Lo" annunciation. If either 75 gpm orifice valves is open the low alarm is 55 gpm.
If both orifice valves are closed the low alarm is 47 gpm. This prevents alarm when PD pump is in service.
oo. FT-62-93C: Provides ACR flow indication and alarm of "Charging Flow Lo." Provides control signal for FCV 62-93 (AUX mode) pp. PT-62-92A: Provides local & MCR Charging Header pressure indication.
qq. PT-62-92C: Provides ACR Charging Header pressure indication.
rr.
FCV-62 Charging Hdr/RCP Seal Injection Flow Control (1) Maintains sufficient backpressure in the charging header to ensure adequate flow of seal water to the RCPs.
(2)
HIC(s) located in MCR and ACRt JC LESSON BODY Prevents spurious operation during App. R fire.
Located on local panel L-1 12 47W610-62.2 45W600-62-4 47W610-62-2 47W611-62-4 Normally 8-13 gpm per RCP.
X LESSON BODY
FACILITY COMMENTS FOR NRC WRITTEN EXAMINATION WATTS BAR 11/26/2002 NUCLEAR Question #
Comment Associated References SRO 99 Accept two correct answers. The stem FR-H.2, "Steam Generator specifies that FR-H.2 has been entered and does Overpressure",
not state what specific step the operating crew is performing. Therefore both A and B alternatives are correct. The operator has entered FR-H.2 and will continue with that instruction until transitioned to another procedure, in this instance FR-H.3. This transition does not occur until step 3 of FR-H.2.
As alternative B states, the operator will continue in FR-H.2 to reduce S/G pressure until he reaches step 3. In the case of alternative A, if the operator is at step 3 of FR-H.2 he will then transition to FR-H.3.
Watts Bar 2002-301 SRO Inital Exam
- 99.
Given the following:
- The Unit 1 operating crew is currently executing E-1, 'Loss of Reactor or Secondary Coolant".
- Containment pressure is 1.5 psig and slowly dropping.
- RCS temperature is 570°F.
- S/G Pressures: S/G #1 = 1200 psig; S/G#2 = 1190 psig; S/G#3 = 1230 psig; S/G#4 = 1205 psig.
- SG NR Levels: S/G #1 = 25%; S/G#2 = 30%; S/G#3 = 87%; S/G#4 = 35%.
- FR-H.2, "Steam Generator Overpressure", has been entered.
For the existing plant conditions, the Unit Supervisor should:
A. Direct the operators to NOT release steam from S/G#3 and transition to FR-H.3, "Steam Generator High Level" to control and lower S/G#3's level.
B. Direct the operators to NOT release steam from S/G#3 and continue with FR-H.2 to reduce S/G pressure.
C. Direct the operator to open the PORV on S/G#3 to drop pressure below 1220 psig then transition to FR-H.3, "Steam Generator High Level" to control and lower S/G#3's level.
D. Direct the operator to open the PORV on S/G#3 to drop pressure below 1220 psig and continue with FR-H.2 to reduce S/G pressure.
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS FR-H.2 STEAM GENERATOR OVERPRESSURE Revision 4 Unit I QUALITY RELATED REQUESTED BY:
SPONSORING ORGANIZATION:
APPROVED BY:
Jim Young OPERATIONS J.M. Earles EFFECTIVE DATE:
LEVEL OF USE:
2/21/2002 CONTINUOUS
WBN STEAM GENERATOR OVERPRESSURE I FR-H.2 I
Rev 4 1.0 PURPOSE This Instruction provides actions for an overpressure condition affecting any steam generator where pressure has risen above the highest steamline safety valve setpoint.
2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications S/G pressure greater than or equal to 1220 psig.
2.2 Transitions FR-0, Status Trees, FR-H in YELLOW condition.
3.0 OPERATOR ACTIONS 2 of 6
WBN STEAM GENERATOR OVERPRESSURE FR-H.
S A /
Rev 4 FSt-- IAction/Expected Response RsrieNtObtained
- 1.
IDENTIFY affected SIG(s):
- a. Any S/G pressure greater than or equal to1220 psig.
- 2.
ENSURE MFW isolated to affected S/Gs:
MFW isolation valves CLOSED.
MFW bypass isolation CLOSED.
SMFW reg valves CLOSED.
MFW bypass reg valves CLOSED.
MFWPT A and B TRIPPED.
Standby MFWP STOPPED.
- Cond demin pumps TRIPPED.
- Cond booster pumps TRIPPED.
3.
CHECK affected S/Gs NR level less than 85% [77% ADVI.
- a. IF press in all S/Gs less than 1220 psig, THEN RETURN TO Instruction in effect.
Manually CLOSE valves, and STOP pumps, as necessary.
IF valves can NOT be closed, THEN CLOSE #1 heater outlet valves.
- GO TO FR-H.3, STEAM GENERATOR HIGH LEVEL.
3 of 6
WBN STEAM GENERATOR OVERPRESSURE FR-H, Rev 4 Ste I Action/Expected Response Response Not Obtained CAUTION If affected SIG NR level rises to greater than 85% [77% ADV],
steam should NOT be released from the affected S/Gs.
- 4.
DEPRESSURIZE affected S/Gs:
"* S/G PORVs, OR "o MSIV bypass valves, OR Steam supply to TD AFW pump, OR S/G blowdown valves.
- 5.
ESTABLISH affected S/Gs press control:
- a. CHECK S/G press dropping.
- b. CHECKS/G press less than 1220 psig;
- a. IF S/G press rising or stable, THEN
- GO TO Caution prior to Step 6.
- b. IF press greater than or equal to 1220 psig, THEN
- GO TO Step 3.
- c. MAINTAIN S/G press less than 1220 psig.
- d. RETURN to instruction in effect.
K-4 of 6
S ActionlExected Res onse I Response Not Obtained CAUTION AFW flow should remain isolated to affected S/Gs UNTIL a steam release path is established.
- 6.
ISOLATE AFW flow to affected S/Gs:
- a. IF LCVs will NOT close, THEN Locally CLOSE manual isolation valves.
- b. IF LCVs will NOT close, THEN Locally CLOSE manual isolation valves.
- c. CHECK affected S/Gs AFW flow zero gpm.
- 7.
CHECK T-hot less than 5450F.
COOLDOWN RCS to less than 5450F by dumping steam from Intact S/Gs.
5 of 6
WBN STEAM GENERATOR OVERPRESSURE FR-He2 I
Rev 4 Ste Action/Ex ected Response Resp onse Not Obtained
- 8.
CONTINUE attempt to manually or locally depressurize affected S/Gs:
"* S/G PORVs, OR MSIV bypass valves, OR
"* Steam supply to TD AFW pump, OR
"* S/G blowdown valves.
- 9.
RETURN TO Instruction in effect.
- End -
6 of 6