ML20203K832
| ML20203K832 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/08/1986 |
| From: | Fordham D GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20203K798 | List: |
| References | |
| LO-LP-39211, LO-LP-39211-00, NUDOCS 8608210371 | |
| Download: ML20203K832 (12) | |
Text
Geor8ia Power
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VOGTLE ELECTRIC GENE (IATING PEAN'Ti a e TRAINING LESSON PLAN TITLE:
PLANT SYSTEMS 3/4.7 NUMBER:
LO-LP-39211-00 PROGRAM:
LICENSED OPERATOR TRAINING R[yjgjgy:
0 AUTHOR:
D. FORDHAM DATE:
8/8/86 APPROVED:
g/jjg DATE:
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INSTRUCTOR GUIDELINES:
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LO-LP-39211-00 1.
PURPOSE STATEMENT:
TO TEACH THE STUDENT THE APPLICABILITY AND ACTIONS OF LIMITING CONDITION FOR OPERATION SECTION 3/4.7 PLANT SYSTEMS.
II.
LIST OF OBJECTIVES:
1.
The student will be able to determine if in violation of an LCO if given a list of equipment and a given applicability condition.
2.
The student will be able to give the required action statement from memory if the time limit for action is one hour or less.
3.
The student will be able to look up the required action if given the applicable LCO if action required in more than one hour.
4.
The student will be able to state the values of the following Limiting Condition for Operation.
a.
Safety Valve Setpoints b.
S/G Specific Activity c.
S/G Temperature / Pressure Limits Additional for SRO 5.
The student will be able to explain the bases for each of the LCO's.
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LO-LP-39211-00
REFERENCES:
TECHNICAL SPECIFICATIONS SECTION 3/4.7 PLANT SYSTEMS i
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i LO-LP-39211-00 111.
LESSON OUTLINE:
NOTES I.
PLANT SYSTEM LIMITING CONDITION FOR OPERATION A.
Turbine Cycle 1.
Safety valves a.
Read section 3/4.7.1.1 l
i b.
Bases 1)
Ensures secondary pressure limited to 100% design a)
Design is 1185 psig b) 110% design is 1304 psig c)
During most severe operational transient 2)
Maximum capacity a)
Trip from 100% power b)
CoincidIent with loss of heat sink c) 18,607,220 lbs/hr (120% rated flow) 3)
Minimum of 2 valves /SG a)
Sufficient capacity for power level in Table 3.7-2 4)
Reduced trip setpoint a)
SP = (X) - (Y)(V) x (109)
X.
b)
SP = reduced trip setpoint (Rated Thermal Power) l c)
V = Max i of inoperable valves /
steam line i
d) 109 = Normal high flux setpoint e)
X = Total relieving capacity of all l
safeties / steam line l
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f)
Y = Max relieving capacity of one valve l
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LO-LP-39211-00 Ill.
LESSON OUTLINE:
NOTES 2.
Auxiliary Feedwater System a.
Read section 3/4.7.1.2 b.
Bases i
1)
RCS'can be cooled to less than 350*1' from NOT a)
With loss of all offsite power 2)
Capacities a)
Each motor - 630 gpm at 1220 psig b)
Turbine - 1015 gpm at 1220 psig 3.
Condensate Storage Tank a.
Read section 3/4.7.1.3 b.
Bases 1)
Maintain RCS,at Ho.t Standby for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> a)
Steam discharge to atmosphere b)
Total loss of offsite power 2)
Followed by cooldown to RHR conditions O
3)
Takes into account unusable portion of water 4.
Secondary Activity a.
Read section 3/4.7.1.4 b.
Bases 1)
Ensure offsite doses small fraction of l
10CFR100 i
a)
Main steam line break 2)
Includes a 1 gpm primary / secondary leak l
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r LO-LP-39211-00 ill.
LESSON OUTLINE:
NOTES 5.
Main Steam Line Isolation Valves a.
Read section 3/4.7.1.5 b.
Bases 1)
Ensures only one S/G blows down a)
Steam line break 2)
Required to:
a)
Minimize positive reactivity inserted from cooldown b)
Limit pressure rise in containment B.
Steam Generator Pressure / Temperature Limitation 1.
Read section 3/4.7.2 2.
Bases Stresses in S/G d,o not. exceed fracture tough-a.
ness stress limits 1)
Will prevent brittle fracture C.
Component Cooling Water System 1.
Read section 3/4.7.3 2.
Bases a.
Ensures sufficient cooling capacity 1)
Safety related equipment 2)
Normal and accident condition 3)
Assumed single failure D.
Nuclear Service Cooling Water 1.
Read section 3/4.7.4 e
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LO-LP-39211-00 ill.
LESSON OUTLINE:
NOTES 2.
Bases a.
Ensures sufficient cooling capacity 1)
Safety related equipment i
2)
Normal and accident conditions 3)
Assumed single failure E.
Read section 3/4.7.5 2.
Bases a.
Ensures sufficient cooling capacity 1)
Provide normal cooldown of facility 2)
Mitigate accident conditions 3)
Provides a 30 day supply F.
Control Room Emergency Filt' ration System 1.
Read section 3/4.7.6 2.
Bases a.
Ensures control room:
1)
Air temperature maintained for continuous duty rating of equipment 2)
Remains habitable during all credible accident conditions b.
Limits moisture buildup on HEPA's and adsorbers 1)
On for 31 days and heaters 10 hrs.
c.
Meets 10CFR50 Appendix A Design Criteria 19 1)
Less than 5 Rem W.B. to occupants of control room 7
LO-LP-39211-00 ill.
LESSON OUTLINE:
NOTES G.
Piping Penetration Area Filtration and Exhaust System 1.
Read section 3/4.7.7 2.
Bases a.
Filters radioactivity from leakage following LOCA 1)
Piping penetration room 2)
ECCS equipment b.
Limits buildup of moisture on HEFA's and adsorbers 1)
Heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over 31 day cumulative c.
System was assumed in accident analysis H.
Snubbers 1.
Read section 3/4.7.8 2.
Bases a.
Ensure structural integrity 1)
Safety Related System 3)
During and following seismic event b.
Inspection interval 1)
Varies inversely with number of snubber failures 2)
Determined by number of inoperable snubbers found l
c.
When snubber is found inoperable 1)
Engineering evaluation performed a)
Determine failure mode b)
Equipment adversely affected by failure 8
LO-LP-39211-00
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lli.
LESSON OUTLINE:
NOTES I.
Sealed Source Contamination 1.
Read section 3/4.7.9 l
2.
Bases a.
Limits based on 10CFR70.39(c) (Plutonium) b.
Ensures leakage will not exceed intake values c.
Sealed sources divided into three groups 1)
According to use d.
Surveillances based on probability of damage e.
Sources within shielded mechanism 1)
Portable radiation instruments 2)
Need not be tested unless removed J.
Area Temperature Monitoring 1.
Read section 3/4.7.10 '
2.
Bases a.
Ensures temperatures less than qualification temperature 1)
Safety related equipment 2)
Possibility of degradation to equipment K.
ESF Room Cooler and Safety Related Chiller Systen 1.
Read section 3/4.7.11 l
2.
Bases a.
Ambient air temperature does not exceed continuous duty rating 1)
For equipment cooled by system l
l 9
r LO-LP-39211-00 111.
LESSON OUTLINE-NOTES L.
Reactor Coolant Pump thermal barrier isolation function 1.
Read Section 3/4.7.12 2.
Bases: Prevent RCS spill from breached thermal barrier should a break in ACCW piping downstream of isolation valve occurred.
II.
Summary A.
Plant Cycle 1.
Safety valves 2.
Condensate storage system 4.
Activity 5.
Main Steam Line isolation valves B.
Steam Generator Pressure / Temperature Limitation C.
Component Cooling Water D.
Nuclear Service Cooling Water E.
Control Room Emergency Filtration System G.
Penetration Room Filtration System H.
Snubbers I.
Sealed Source Contamination J.
Area Temperature Monitoring K.
ESF Room Cooler and Safety-Related Chiller System L.
RCP Thermal Barrier Isolation 9
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LO-LP-39211-00 IV.
PRACTICAL EXERCISES IV.
PRACTICAL EXERCISES A.
(NSIC 197971) ON APRIL 7, 1985, FOUR OF THE FIVE MAIN STEAM LINE CODE SAFETY VALVES ASSOCIATED WITH NO. 22 STEAM CENERATOR PREMATURELY LIFTED AND WERE GAGGED SHUT.
ACTION WAS INITIATED TO PLACE THE UNIT IN HOT SHUTDOWN IN ACCORDANCE WITH LCO 3.0.3.
THE CAUSE WAS DETERMINED TO BE INCORRECT LIFT SETTINGS, ALTHOUGH THE ROOT CAUSE COULD NOT BE DETERMINED. BECAUSE THE LIFT SET TESTING REQUIRES STEAM GENERATOR PRESSURE TO BE GREATER THAN 900 PSIG, COOLING DOWN WOULD HAVE PRECLUDED BEING ABLE TO CORRECT THE CAUSE OF THE PROBLEM. THEREFORE, THE C00LDOWN WAS TERMINATED, AND THE VALVES WERE RESET AhT SATISFACTORILY TESTED UNDER THE COGNIZANCE OF THE VENDOR. SINCE THE TECH SPEC BASIS WAS NOT COMPROMISED, AND BECAUSE THE COMMISSION WAS AWARE OF THE SITUATION AND CONCURRED WITH PSE&G'S ACTIONS AT THE TIME OF THE EVENT, THIS EVENT WAS INITIALLY CLASSIFIED AS NON-REPORTABLE. HOWEVER, BECAUSE THE ACTION REQUIREMENTS OF LCO 3.0.3 WERE NOT COMPILED WITH, THE EVENT WAS SUBSEQUENTLY DETERMINED TO BE REPORTABLE IN ACCORDANCE WITH 10 CFR 50.73(A)(2)(1)(B), AND THE LER IS BEING SUBMITTED GREATER THAN THIRTY DAYS FOLLOWING-THE ACTUAL EVENT DATE. A LICENSE CHANGE REQUEST IS BEING SUBMITTED TO REVISE THE TECHNICAL ACTION REQUIREMENTS WHEN MORE THAN THREE (3) SAFETY VALVES ARE INOPERABLE (DUE TO SETPOINT DIFFERENCES) ON ANY STEAM GENERATOR.
1.
Major points for discussion:
a.
LCOs violated:
(action not compiled with) 3.7.1.1 - allows up to 3 safeties / loop inoperable 3.0.3
- action to. place plant in cold shutdown not done B.
(NSIC 188195) ON NOV.18, WITH UNIT 2 AT 100% POWER, 2-FW-P-3B ('B' MOTOR DRIVEN AFP) WAS UNABLE TO DEVELOP FLOW AND WAS DECLARED INOPERABLE. ON NOV. 20. THE 2-FW-P-3B WAS DECLARED INOPERABLE DUE TO AN INOPERABLE LUBE OIL COOLER. LATER THAT DAY, THE 2-FW-P-3B AND 2-FW-P-2 (STEAM DRIVEN AFP) WERE VAPOR BOUND AND WERE DECLARED INOPERABLE. ON DEC. 6, 2-FW-P-3B WAS VAPOR BOUND AND DECLARED INOPERABLE. THESE EVENTS ARE CONTRARY TO TECH SPEC 3.6.B.1 AND 3.6.C AND REPORTABLE PER TECH SPEC 6.6.2.B. (2). A MOTOR DRIVEN AFP REMAINED OPERABLE DURING THE EVENTS. LEAKING CHECK VALVES DOWNSTREAM OF THE AUX. FEED PUMPS ALLOWED HOT WATER FROM THE S.G. MAIN FEED LINES TO SEEP INTO THE PUMP CASING AND VAPOR BIND THE PUMPS. THE LEAKING CHECK VALVES WERE REPAIRED AND THE SYSTEM RETURNED TO OPERATION.
1.
Major points for discussion a.
LCOs violated 3.7.1.2 - requires 3 operable AFW pps b.
Action statement entered when pump would not start.
1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore pump to opdrable status 2)
Check valves replaced l
3)
Pump operability STP run - declared j
operable 11
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'i LO-LP-39211-00 IV.
PRACTICAL EXERCISES C.
(NSIC 193367) ON 2 OCCASIONS ON 1-23-85, TECH SPEC 3.0.4 WAS VIOLATED IN THAT THE WATER LEVEL ABOVE THE REACTOR PRESSURE VESSEL FLANGE WAS DECREASED BELOW 23 FT WHILE 13 SNUBBERS HAD BEEN REMOVED FOR SURVEILLANCE TESTING FROM THE A TRAIN RHR LOOP. REMOVAL OF THE SNUBBERS MADE THE A TRAIN RHR LOOP INOPERABLE FROM A TECH SPECS POINT OF VIEW EVEN THOUGH IT WAS CAPABLE OF PROVIDING WATER TO THE RCS AND THE REFUELING CANAL. TECH SPECS REQUIRE BOTH TRAINS OF RHR TO BE OPERABLE IN MODE 6. WHEN THE WATER LEVEL ABOVE THE TOP OF THE REACTOR PRESSURE VESSEL FLANGE IN LESS THAN 23 FT.
1.
Major points for discussion a.
Which LCO action statements entered?
b.
Would this condition be an LCO violation in a different mode? Why?
c.
What action was required to clear this condition?
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