ML13329A757

Forwards Comments on Reactor Operator Exam & Senior Operator Exam Administered by NRC on 920427,at Montreal,Canada

ML13329A757
Person / Time
Site:	San Onofre
Issue date:	05/08/1992
From:	Morgan H SOUTHERN CALIFORNIA EDISON CO.
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 9205180335
Download: ML13329A757 (55)

Text

ENCLOSURE 1 COMMENTS ON THE REACTOR OPERATOR EXAMINATION

COMMENTS ON RO EXAMINATION (Administered April 27, 1992)

1.

QUESTION 002 (1.00)

Given the following:

Turbine Runback from Underfrequency has reduced power to 60% (Overshoot on turbine controls)

Rod Control System in manual for duration of runback 10.5 degree F Tavg-Tref mismatch WHICH ONE (1) of the.following rod speeds would be observed if the Rod Control Selector switch was taken to the AUTO position?

(Figure 5 of SD-SO1-400, "Rod Control System" is attached)

a.

8.75 inches/min

b.

9.25 inches/min

c.

11.25 inches/min.

d.

13.0 inches/min ANSWER:

c NRC

REFERENCE:

(TV)

1.

SD-SO1-400, Rod Control System, Pg 6 of 52, Figure 5

2.

L.P. 1X1203, Obj 1.3.2, Pg 26 SCE COMMENTS FACILITY DISCUSSION:

The graph removed from system description SD-Sol-400 which was attached to the test does not agree with the formula in system description SD-SO1-400.

Choice "c" is correct using the formula, which is correct.

Answer "d" is correct using the graph. The graphic artist when drawing the graph made a mistake and slipped the endpoint of the graph to the left one unit. The slope for Rod Speed rose from 5 in/min to 15 in/min from 8 to 11 degrees vs. the proper slope from 8 to 12 degrees.

REFERENCE MATERIAL:

L.P. 1X1203, Rod Control System, Obj 1.3.2, Pg 26, does not address this question clearly. Page 12, in lesson plan 1X1203 is the proper location to address COMMENTS.R03, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 2 this question, it discusses the fact that the rod speed is variable between 5" to 15"/min and provides a handout/TP entitled Figure 1-7.

This figure is similar to the graph referenced in the system description but is drawn correctly.

SD-SO1-400, Rod Control System, Pg. 6, properly discussed the correct Tavg program, this is the choice that the NRC chose as correct. The NRC examiner did not check the accuracy of SD-SO1-400, Figure 5, but included it on the examination and this is what led to their being two correct answers.

The reference material provided by SCE was adequate to allow proper addressal of this question and answer.

SCE

REFERENCE:

1.

L.P. 1X1203, Obj 1.3.2, Pgs 12 and 26

2.

L.P. 1X1203, Student Handout Figure 1-7

2.

SD-SO1-400, Rod Control System, Pg 6 and Figure 5 SCE REQUESTED RESOLUTION: Accept both answers c and d COMM1ENTS.RO3, REV 2, May 8, 1992

6.0 LESSON PRESENTATION CONTENT ACTIVITIES b)

Receives analog input from Sends output signal from 5" to 15" the RCPS proportional per minute.

controller -

signal is proportional and to the Show TP Figure 1-6 Rod Drive degree of Tave-Tref Control System.

mismatch. This controls direction of travel and speed of the rods.

C)

Also receives an analog dT/dt is generated by the signal proportional to Differentiator.

P-Pref and rate of change of These have minimal affect on rod nuclear flux from NIS PR speed.

1208.

This affects only Objective #1.3 the speed of the rods.

d)

Produces variable pulsed Show TP Figure 1-7 Rod Speed vs.

output through the master Tave Tref cycler to the slave cycler clutches.

e)

Maximum pulse rate limited to 15 inches (40 pulses)/minute.

12

6.0 LESSON PRESENTATION CONTENT ACTIVITIES

5)

S/D Group Control Relays.

6)

Control bank control relays.

7)

SK1 relays (S/D MGC resistance circuits).

8)

TR and CK1 relays (Control.Group MGC resistance circuits).

III. General System Operations Objective #1.2 A.

Automatic Operation

1.

Tave -

Tret exceeds + 2 F; signal sent to pulser.

2.

Pulser converts analog input to pulsed output voltage signals.

a.

Pulse rate increases with Tave Tref mismatch.

3.

Pulses received by master cycler.

a.

Master cycler distributes pulses among the control group slave cycler clutches.

26

V7 1'

IL 0-E-TE 0F

_USE ONL

NUCLEAR GENERATION SITE SYSTEM DESCRIPTION S-SO1-400 UNIT 1 REVISION 2 PAGE 6 OF 52 ROD CONTROL SYSTEM

2.0 DESCRIPTION

(Continued) 2.1.2 General Control Scheme (Continued)

.1 (Continued)

The rods will start to move in or out as necessary at 5"/minute. Rod speed is maintained at 5"/minute from

+ 2F to + 8*F mismatch. From +/- 8"F mismatch to

+/- 12'F mismatch rod speed increases (2.5"/min./'F) from 5"/minute to 15"/minute which is its upper limit.

Any mismatch greater than +/- 12*F will still result in a rod speed of 15"/minute (Figure 5).

In manual operation the Control Rods receive their In or Out motion signal from the In-Hold-Out Switch on the J-console and their speed signal from manual speed controller HC-413 located in Rack R-1 behind the West Vertical Board.

The speed signal is required to always be set at 15"/minute.

The group or bank to be moved is determined by the position of the Group Selector Switch and the Overlap Control Switch.

Both switches are located on the J-console.

NUCLEAR GENERATION SITE SYSTEM OESCRIPTION S-SO1-400 UNIT I REVISION 1 PAGE 38 OF 52 FIGURE 5:

ROD. CONTROL PROGRAM ROD SPEED 1s (INCHES/MIN.)

14 13 12 11 10 9

8 0 7 6

5 4

3 2

12 11 10 9

8 7

6 5

4 3

2 1

1 2

3 4

5 6

7 8

9 10 11 12 NEGATIVE ERROR (TAVG<TREF)

POSITIVE ERROR (TAVG>TREF)

NUCLEAR GENERATION SITE SYSTEM DESCRIPTION SD-SO-400 UNIT 1 REVISION 2 PAGE 6 OF 52 ROD CONTROL SYSTEM

2.0 DESCRIPTION

(Continued) 2.1.2 General Control Scheme (Continued)

.1 (Continued)

The rods will start to move in or out as necessary at 5"/minute. Rod speed is maintained at 5"/minute from

+ 2*F to + 8*F mismatch. From + 8*F mismatch to

+ 12*F mismatch rod speed increases (2.5"/min./OF) from 5"/minute to 15"/minute which is its upper limit. Any mismatch greater than +/- 12'F will still result in a rod speed of 15"/minute (Figure 5).

In manual operation the Control Rods receive their In or Out motion signal from the In-Hold-Out Switch on the J-console and their speed signal from manual speed controller HC-413 located in Rack R-1 behind the West Vertical Board. The speed signal is required to always be set at 15"/minute.

The group or bank to be moved is determined by the position of the Group Selector Switch and the Overlap Control Switch. Both switches are located on the J-console.

NUCLEAR GENERATION SITE SYSTEM vESCRIPTION Su-SOl-400 UNIT 1 REVISION 1 PAGE 38 OF 52 FIGURE 5:

ROD CONTROL PROGRAM ROD SPEED is (INCHES/MIN.)

14 1A3 2

10 g

8

)

7 6

'3 2

12 11,10 9 8

7 6

5 4

3 2

1 1 2 3

4 5

6 7 8 9 10 11 12 NEGATIVE ERROR (TAVG<TREF)

POSITIVE ERROR (TAVG>TREF)

RO EXAMINATION COMMENTS Page 3

2.

QUESTION 003 (1.00)

Given the following:

Power level 80%

Pressurizer level is at program setpoint.

Tavg recorder pens all indicate approximately 549 degrees F.

Tavg/Tref are matched on recorder TR-405 Rods start stepping in at high speed No turbine runback is in progress WHICH ONE (1) of the following would cause these indications?

a.

TM-407 -

Avg Tave module to Rod Control loss of power.

b.

TE-401A, Loop A hot leg temperature failed high.

c.

TM-405 - Avg Tave summing computer failed low.

d.

PT-415 -

1st Stage pressure failed high.

ANSWER:

a NRC

REFERENCE:

(TV)

1.

SO1-13-3, "Annunciator Response", Pg 13

2.

L.P. 1XC206, Obj 2.2, Dwg RPS-1-7 SCE COMMENTS FACILITY DISCUSSION: The sequence of the information presented in the initiating conditions resulted in confusion for the licensed operator candidates. The indications provided imply the rods start stepping in after all of the other information is provided. It does not state the rods are stepping in during the initial conditions but that they start stepping in. The setup of the question made it unclear as to whether the failure just occurred, with the conditions provided as the starting point or if the failure referenced was already present with the conditions provided.

Due to this confusion the students requested clarificatio n as to which was the correct condition, were the conditions presented time 0, or after the failure.

The response to the question was to consider the conditions time 0. In COMM'ENTS.RO3, REV 2, may 8, 1992

RO EXAMINATION COMMENTS Page 4 addition, the students are not expected to know the failure mode of all components in a given circuit, this caused some to question what the result would be if TM-407 were to loose power.

Answer "a" is correct. If TM-407 loses power, it fails such that Tavg appears high, thus resulting in rods moving in.

Answer "b" is correct. If the conditions are assumed to be at time 0. At time 0, with all systems working properly, the power level provided does not make a difference.

Pressurizer level will be on program as stated in the initial conditions, Tavg recorder pens would all indicate approximately 549 degrees F, and Tavg/Tref would be matched on the recorder TR-405. Turbine runback would be irrelevant at this time, and as stated in the conditions, rods would have just started stepping in at high speed after the failure initiated.

The initial condition of pressurizer level-at program setpoint further led the candidates to believe that the conditions provided in the question marked time 0 vs. some time after the initiation of the failure. Pressurizer level would not be at program setpoint if the failure had occurred prior to the conditions given. Pressurizer level and pressure are normally seen prior to actual temperature changes on the recorders, if the.conditions provided represented a time after time 0, the pressurizer level should have been less then program level, due to RCS temperature decrease, resulting in system shrinkage and an outsurge from the pressurizer.

This question was not reviewed during the examination pre review and had not been previously validated. The.SCE reviewer believes it was added to the examination after the pre-review.

REFERENCE MATERIAL:

L.P. 1XC206, Objective 2.2, Dwg. RPS-1 7, has no application to this question. objective 2.2 states ' Identify the physical location of all indicators, recorders, and controls for the Reactor Control System."

It is not an expectation that an operator know the failure mode of each part of a circuit in the plant, only that he/she can respond to the symptoms as they occur. To know that loss of power to Tavg module TM-407 would result in Tavg appearing high is not an expectation of the lesson referenced, and is not covered by any of the objectives in the referenced lesson plan, 1XC206. Drawing RPS-1-7 does CONMENTS.RO3, REV 2, may 8, 1992

RO EXAMINATION COMMENTS Page 5 not indicate/label TM-407 Tavg module and does not discuss or indicate how it will fail on a loss of power.

None of the references used by the NRC examiner were appropriate for this question, or could have provided the NRC examiner the information needed to correctly answer this question.

The reference material provided by SCE was adequate to allow addressal of this question and answer.

SCE

REFERENCE:

To answer this question takes an integrated knowledge of the entire Reactor Control System and Nuclear Plant Fundamentals. There is no single set of references available that can be used to answer this question

1.

1542 sheet 102C SCE REQUESTED RESOLUTION:

Accept both a and b.

COMMENTS.RO3, REV 2, May 8, 1992

1.0 OBJECTIVES

Upon completion of this lesson, the student will be able to:

1.0 Relate knowledge of system purpose, function and design basis to an overview of the Reactor Control System.

1.1 Describe the purpose of' the Reactor Control System.

1.2 List the functions of the Reactor Control System.

1.3 Describe the design features that enable the Reactor Control System to ensure that the safety limits will not be exceeded during power operation.

2.0 Apply knowledge of component type, location, function, principle of operation, design features and support systems to the ACO tasks of the Reactor Control System.

2.1 List the names and type of components used for inputs to the Reactor Control System.

2.2 Identify the physical location of all indicators, recorders and controls for the Reactor Control System.

2.3 For each RTD/Press. Xmtr. explain how it affects the Reactor Control System.

2.4 Explain how variable Ta, Steam Dump Control and ave Shutdown Margin contribute to overall plant operation.

2.5 Explain how Reactor Control System supports the operation of the Reactor Protection System.

3.0 Apply knowledge of control type, purpose, principle of operation (in normal and abnormal modes), manipulations and actuations to the ACO tasks of the Reactor Control System.

3.1 List and describe the purpose of the Reactor Control System controls.

3.2 Explain how TREF-AVE TAVE control operates.

3.3 List and describe the inputs and outputs for the following Reactor Control System controls:

a.

Loop TAVE Defeat Switch.

b.

Loop AT Defeat Switch.

Page iv of vii

COLD LEG RTD TE TAVG DEVELOPMENT TE HOT LEG RTD RANGE 525-625 OF 401 C 401 A RANGE 525-625 -F TT-401 T + T R/I CONVERTER RANGE TO TC-400A&B TA-401A&B ALARM VARIABLE LOW HIGH ALARM PRESSURE TRIP TAVE HIGH - 557 F CALCULATOR

• LOW

- 533 OF TR-401 -1/2/3 B -

TAVG RECORDER C

525-625 OF BC SW-I SWITQH DEFEAT VG. TAV.

3 AVG.

TAVG.

T ROD CONTROL TM LEAD/LAG TR-405-1/2 COMPUTER 417 Av TAVG TREF RECORDER TO TO LEVEL SHUTDOWN PROGRAM-MARGIN COMPUTER TO TEAM DUMP TA-405 A&B CO TROL TAVG -

TpEF ALARM +/-5 OF TR ROD CONTROL TC-415 FROM 0T-415 RPS-1-7 RJM

1AzT ION OPERATING INSTRUCTION S0I-13-3 ANNUNCIATOR RESPONSE REVISION 1 PAGE 13 OF 84 ATTACHMENT 2 TCN t-;J1 REACTOR PLANT NO. 2 ANNUNCIATOR 1.0 Annunciator Window:

9 REACTOR COOL.

HI TAVG

2.0 Setpoint

557*F LOOP C.

3.0 Initiatino Device:

TA-421A-X 9

4.0 Aoolicability

MODES 1 - 3 5.0 Probable Cause:

(1) Rapid decrease in secondary steam demand.

6.0 Recommended Action:

CAUTION:

With Avg Tv. exceeding its limit of 553.5'F, Avg T; shall be RESTORED to less than 553.5"F within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERIAL POW R to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(Reference #4)

1)

VERIFY conditions and CHECK Loops "A" and "C" Tavg.

Loop "C" Hot Leg and Cold Leg indication.

2)

CHECK feedwater control.

3)

MONITOR Control Rod Motion.

4)

CHECK Loop "C" AT recorder and indication.

5)

If signal is erroneous, REFER to SOI-4-34 to defeat Loop "C" Tavc and variable low pressure trip for Loop "C".

7.0

References:

1)

Westinghouse Reactor Control and Protection System Manual

2)

N1542 sh.

102C, "Reactor Protection"

3)

SO1-4-34. "Reactor Plant Instrumentation Operation"

4)

Letter from B. Carlisle to o.P. Short and R.W. Waldo dated 2/25/91.

Subject:

Administrative Controls for Reduced Tave. SONGS 1.

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RO EXAMINATION COMMENTS Page 6

3. QUESTION 004 (1.00)

WHICH ONE (1) of the following constitutes an "operable low temperature overpressure protection system", for low temperature RCP starts?

a.

One PORV with a lift setting of less than or equal to 360 psig.

b.

Two PORV's with a lift setting of less than or equal to 400 psig.

c.

RHR relief valve RV-206 aligned to the RCS with a lift setting of less than or equal to 515 psig.

d.

A Reactor Coolant System vent of greater than or equal to 1.75 square inches.

ANSWER:

d NRC

REFERENCE:

(TV)

1.

S01-4-3, RCP Operation, Pg 11.of 31, Rev date 3/14/91

2.

LP-1XA203, Rev 2, Objective 1.3.2, Pg 13 SCE COMMENTS FACILITY DISCUSSION:

The inclusion of the statement "for low temperature RCP starts" makes answer "d" an incorrect choice. Had the question stated "which of the following constitutes an operable overpressure protection system?",

and not mentioned a low temperature RCP start, then d would be a correct answer. The requirement for a 1.75 square inch vent for overpressure protection is not related to a RCP start. A 1.75 square inch vent would prohibit a RCP start.

A reference used by the NRC, SO1-4-3, Section B, Pg 11, Precaution 4.3, is a general precaution statement placed in all procedures that relate to low RCS pressure operation.

The procedure precaution makes the following statement:

4.3 When the RCS Temperature is 5 360 F, then at least one of the following overpressure protection systems shall be OPERABLE: (Tech. spec. 3.20.A, Reference 2.1.2) 4.3.1 Two Power Operated Relief Valves (PORVS) with a lift setting of 5 465 psig and RHR relief COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 7 valve RV-206 aligned to the RCS with a lift setting of 5 515 psig (Reference 2.4.6); or 4.3.2 A Reactor Coolant System vent(s) of 1.75 square inches.

The application of 4.3.2, will only be used when the RCS is depressurized. It can not be applied to the condition being addressed by this question, i.e. a low temperature RCP start.

The overpressure protection that would be in place when starting a RCP at low RCS pressure ( 400 psig), would be that referenced in S01-4-3, Step 4.3.1, page 11.

Two PORVs with a lift setting of < 465 psig and RHR relief valve RV 206 aligned to the RCS with a lift setting of < 515 psig.

To start a RCP per S01-4-3, pg. 15, step 6.1.13, the RCS pressure must be 350 -

400 psig. Plant pressure can not be maintained at 350 -

400 psig with a 1.75 in or greater vent.

None of the other choices provided include all of the requirements for OMS and are also not related to the referenced low temperature RCP start.

REFERENCE MATERIAL:

The NRC referenced LP-1XA203, Rev 2, Objective 1.3.2, Pg 13 as one of the source documents for their information. None of the answers provided as possible choices are supported by the referenced lesson plan.

Objective 1.3.2 states "State the interlocks and administrative requirements to start on (sic) RCP."

The referenced page does not have any discussion or implication about anything called a low temperature reactor coolant pump start.

It has no discussion regarding any of the criteria provided in the question and its selection of answers. The LP simply states that the following is needed to start a RCP, lift pump on to upper thrust shoes, #1 seal Op > 275 psi, #1 seal leakoff >.25 gpm, seal injection OP = 20" WG, no reverse OT of 5*F only with To > 520*F,.and no overcurrent relay.

The NRC examination did not properly apply the precaution referenced by S01-4-3, RCP Operation, Pg 11 of 31, Rev date 3/14/91.

The reference material provided by SCE was adequate to allow proper addressal of this question and answer.

COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 8 SCE

REFERENCE:

1.

SO1-4.3, pages 11, 12, 13, and 15.

SCE REQUESTED RESOLUTION:

Delete, question; no correct answers provided.

COMMENTS.RO3, REV 2, May 8, 1992

1.0 OBJECTIVES (Continued) 1.2.3 Describe the physical location of the following major RCP, Seal Injection/Leakoff System components: RCP, oil lift pump, seal water manifold & injection filters, seal manifold filter bypass MOV's, FCV's-1115A-F, PCV-1115A-C, CV-276, RV-2004, CV's 527 and 528, RV-289, CV-291, seal water return filter and heat exchanger, vapor seal head tank.

1.2.4 Describe the functions for all major components of the RCP and Seal Injection system.

1.2.5 Explain the operation of all major components of the RCP and Seal Injection system.

1.2.6 Describe the requirements for Component Cooling Water.(CCW) of the RCP and Seal Injection/Leakoff System. List which RCP and Seal Injection Leakoff System components are cooled by CCW.

1.3 Discuss the name or type, location, function, operation, indication, applicable interlocks, requirements, and applicable normal/abnormal modes for the RCP and Seal Injection system controls.

1.3.1 Describe the physical location of the following RCP and Seal Injection system controls:

RCP oil lift pumps, vapor seal head tank make up, MOV's 18 and 19, FCV's 1115A-F, PCV's-1115A-C, CV-276, CV's 527 and 528 and CV-291 1.3.2 State the interlocks and administrative requirements to start on RCP.

1.3.3 Explain the operating relationship between thermal barrier P and FCV-111SA-F controller response when the controllers are in auto.

1.3.4 Explain how the controllers for FCV-1125A-F would be manually manipulated to maintain proper seal injection flow.

1.4 Discuss the name or type, location, function, point of measurement and applicable alarms/setpoints for the instruments in the RCP and Seal Injection system.

List the names, type of indication (meter/

recorder) and purpose of the instruments used in the RCP and Seal Injection system.

iv

6. 0O1oNPRESENTATION CONTENT ACTIVITIES

.2 Operate opposite of RCP In 4KV room breaker position.

Neon indication

.4 Vibration Monitoring Obj. 1.4.3

.1 Function -

monitoring of equipment condition, possible early warning of impending failure.

.2 Operation -

channel indication and recorder located in rack behind west vertical board

.5 Starting prerequisites/interlocks Obj. 1.3.2, 1.6.2, 1.6.4

.1 Lift pump on, pressure light on Interlock

(> 850 psig to upper thrust shoes)

.2

1. 1 Seal AP > 275 psi Administrative

.3

1. 1 Seal Leak Off >.25 GPM Administrative

.4 Seal Injection AP 20" WC Administrative

.5 NO reverse AT, of > 5F only Interlock W/Tc > 520*F

.6 No overcurrent (186 relay reset)

Interlock 13

NUCLEAR ORGANIZATION OPERATING INSTRUCTION S01-4-3 UNIT 1 PRIMARY PLANT RE'.'ISION 4 PAGE 11 OF 31 TCN

-fD SECTION B: RCP OPERATION 4.0 PRECAUTIONS (Continued) 4.4 When the RCS Temperature is < 360'F, then at least one of the following overpressure protection systems shall be OPERABLE:

(Tech. Spec. 3.20.A, Reference 2.1.2) 4.4.1 Two Power Operated Relief Valves (PORVs) with a lift setting of < 465 psig and RHR relief valve RV-206 aligned to the RCS with a lift setting of < 515 psig (Reference 2.4.5): or NOTES: 1) One PORV blocked OPEN and its block valve OPEN meet the requirement of the 1.75 square inch vent. Two PORVs blocked OPEN and their block valves OPEN provide a 3.5 square inch vent. A removed Safety Valve provides a 5.4 square inch vent.

2) Tech. Spec. 3.20 Action C applies whenever RV-206 is

-inoperable or not aligned to the RCS.

4.4.2 A Reactor Coolant System vent(s) of > 1.75 square inches.

4.5 When the RCS Temperature is : 360'F, then the following conditions also apply:

(Reference 2.4.5) 4.5.1 One Charging Pump Breaker shall be RACKED OUT.

4.5.2 A RCP shall not be started unless the requirements of are satisfied.

4.6 When the RCS Temperature is < 3600F, then the following charging flow limits apply (Reference 2.4.5):

NOTE:

The following limits do not apply when the RCS Temperature is

< 210*F, And the RCS is depressurized and vented through a 1.75 square inch vent(s) (Reference 2.4.5).

4.6.1 If FCV-1112 is not closed and disabled, and is not isolated, then the following limits apply:

.1 FCV-1112, Charging Flow Control Valve, shall be limited to less than 150 gpm flow.

.2 FCV-1115A, P, & C, RCP Seal Injection Flow Control Valves, shall be limited to less than 75 gpm total flow.

.3 FCV-1115D, E, & F. RCP Seal Injection Flow Control Valves, shall be MAINTAINED CLOSED.

NUCLEAR ORGANIZATION OPERATING INSTRUCTION SO1-4-3 UNIT 1 PRIMARY PLANT REVISION 4 PA1 12 OF 31 SECTIONB: RCP OPERATION A

4.0 PRECAUTIONS (Continued) 4.6.2 If FCV-1112 is closed and disabled, or is isolated. then the following limits apply:

.1 Flow through FCV-1115A/D, B/E, C/F, RCP Seal-Injection Flow Control Valves, shall be limited to < 150 gpm.

.2 Only one valve (FCV-1115D,E, or F) shall be used as a flowpath, and the other two valves shall be MAINTAINED CLOSED and DISABLED or ISOLATED 4.7 If the conditions of step 4.6 are not met, then the conditions shall be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (Reference 2.4.5) and Operations Management shall be notified.

4.8 A Reactor Coolant Pump shall not be ctartcd with the RCS pressure

< 400 psig unless: (Tech. Spec. 3.1.2.1) 4.8.1 The Pressurizer water level is < 80% or 4.8.2 The potential for having developed reactor coolant system temperature gradients has been evaluated.

NOTE:

There are several means available for determining that there is not a temperature differential of > 50OF between the secondary and primary systems with < 400 psig primary system pressure. These methods may include but are not necessarily limited to the following:

(Tech. Spec. 3.1.2 Basis) (See Attachment 5 for explanation of each of the following.methods).

1) - Converting steam line nressure indication into maximum temperature or steam generator fluid.

2)

Tagging RCP switches with shutoff temperatures.

3)

Assuring adequate time for temperature gradients to dissipate.

4)

Filling steam generators with water of known temperature.

4.9 In Modes 5 & 6, Reactor Coolant Pumps should be rotated manually or electrically, as appropriate, once per two weeks to prevent damace to bearings, seals, etc. 'Station Technical should be notified to evaluate successive intervals (sequentizr.

2 ~:kperiods) where this action could not or was not performed.

(Reference 2.4.2 and 2.4.7)

NUCLEAR ORGANIZATION OPERATING INSTRUCTION 501-4-3 UNIT 1 PRIMARY PLANT REVISION 4 PAGE 13 OF 31 TCN SECTION B: RCP OPERATION 4.0 PRECAUTIONS (Continued) 4.10 The following starting duty of the RCPs shall be observed:

4.10.1 Only one RCP is to be started at any one time.

4.10.2 If a RCP has failed to achieve full speed after an attempted start, a restart should not be attempted until the motor has been allowed to cool by standing idle for a period of not less than 30 minutes. After a RCP trip, an immediate restart is permitted if the motor has been running under normal load for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.10.,3 For each RCP, the number of starts will be limited to a maximum of 3 within a 2.hour period. A minimum idle period of 30 minutes for cooling will be observed between each attempted start.

When 3 starts or attempted starts have been. made within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period, the fourth attempted start must not be made until the motor has been allowed to cool by standing idle for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

.4.11 Venting of RCP instrumentation should:be performed only with the ROP stopped and FCV-1115 A, B and C in MANUAL.

4.12 During solid system operations, all RCPs shall be momentarily stopped when either RHR pump is started. A RCP restart shall be made only after evaluating the potential temperature gradients.

4.13 To prevent over stressing the thermal barrier heat exchancer's coils when the component cooling water is isolated from the thermal barrier heat exchanger, the maximum allowable thermal barrier heat exchanger pressure shall not exceed the primary side pressure by 225 psig.

4.14 Tests or maintenance activities that might affectiReactor Coolant System pressure shall not *be performed during solid system operation..

4.15 If, during RCP operation, parameters associated with the RCP seem to be degrading-, notification should be made to Station Management and Station Engineering for evaluation. (Reference 2.4.1) 4.16.

When manually rotating a coupled RCP, the requirement to provide a minimum of.25 gpm RCP Seal. leakoff flow may be waived provided the following conditions are met:

4.16.1 The RCS level is at least at the Reactor Vessel Flange, and 4.16.2 The RCP Seal Injection is aligned with the No. 1 RC? Seal leakoff flow-directed to a floor drain, with visible drippage-to the drain.

NUCLEAR ORGANIZATION OPERATING INSTRUCTION 501-4-3 UNIT 1 PRIMARY PLANT REVISION 4 PAGE 15 OF 31 TCN SECTION B:

RCP OPERATION 6.0 INSTRUCTIONS (Continued) 6.1.7 VERIFY RCP motor oil system as follows:

.1 Oil level at mid range in upper and lower oil level sight glasses, and

.2 Oil collection system is intact, and NOTE:

Oil Collection System consists of drain pans, piping, fittings and oil collection tank.

.3 Shows no evidence of oil leakage.

6.1.8 VERIFY the Reactor Coolant Pump Oil SeIhtor Valve is in the UPPER SHOES position.

6.1.9 ESTABLISH VCT pressure for running a RCP per SO-4-7.

6.1.10 PLACE the RCP vibration recorder and the individual vibration modules in service for each RCP.

6.1.11 If any RCP instrumentation is out of service. then NOTIFY the SRO Operations Supervisor prior to starting that RCP.

6.1.12 If the observed Seal Injection System parameters are NOT as expected, then CH.ECK the system for proper alignment per 501-4-36 Attachment 2, RCP Seal Water System Alignment, as a possible cause.

6.1.13 VERIFY RCS pressure is 350 -

400 psig.

CAUTION During solid system operations, all RCPs shall be momentarily stopped when either RHR pump is started.

CAUTION An RCP shall not be started with the RCS pressure

& 400 psig unless Pressurizer water level is < 80%, or the temperature difference between the secondary and primary systems is < 500F. (Tech. Spec. 3.1.2.1)

(Attachment 5 provides an explanation of 4 methods of determining the temperature difference).

6.1.14 If the RCS temperature is < 360 0F, then PERFORM the following:

.1 DETERMINE the Steam Generator to RCS AT per attachment 5.

.2 VERIFY the Steam Generator to RCS AT requirements of are satisfied.

RO EXAMINATION COMMENTS Page 9

4.

QUESTION 005 (1.00)

Given the following:

Plant is in Mode 6 RCP "B" motor is uncoupled from the pump.

Loop "B" is full.

7.

Maintenance is working on the RCP "B" pump.

WHICH ONE (1) of the following prevents leakage of reactor coolant up the RCP shaft?

a.

Pump shaft mates with the top of the thermal barrier assembly.

b.

Seal Leakoff collects any RCS leakage up the shaft and directs it back to VCT.

c.

Seal injection is maintained.

d.

Nozzle dam installation.

ANSWER:

a NRC

REFERENCE:

(TV)

1.

L.P. 1XA203, Objective 1.2.4, Page 17 SCE COMMENTS FACILITY DISCUSSION:

System Description SD-SO1-300, page 21 discusses that the original design of the RCP incorporated a.

maintenance seal.

If the system description is read further it states that work on the pump must be done at mid-loop.

This is because the maintenance seal will no longer prevent leakage as stated in the question.

The current plant design and procedures do not allow the plant conditions stated in the question.

In addition, the answer provided will not accomplish that which is stated, i.e. prevent leakage.

REFERENCE MATERIAL:

The NRC reference to L.P. 1XA203, RCP and Seal Injection.System, Objective 1.2.4, Page 17 does not agree with the question and the answer provided. The statement made in the L.P. is:

COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 10 "Inside diameter of housing contains stellite overlayed seat to mate with similar seat on the outside of the shaft to act as a low pressure valve during seal section maintenance."

The objectives referenced as applicable to this question are 1.1.3 and 1.1.4, not 1.2.4 as referenced by the NRC Answer Key. Objective 1.2.4 states "Describe the functions for all ma]or components of the RCP and Seal Injection system."

The feature briefly mentioned in the L.P. is not considered a major component of the RCP and Seal Injection System.

Objectives 1.1.3 and 1.1.4 reference discussions on how RCS leakage through the seals is controlled for various conditions. The discussion in the lesson plan does not state that the function of this seal design is to prevent leakage of reactor coolant up the RCP shaft during maintenance, it simply states that it acts as low pressure valve, not a low pressure cutoff valve.

Had the information assumed by the examination writer been cross checked with the system description SD-SOl-300, and the plant operating procedures, it would have been known that we can not perform maintenance on the RCP pump or it's seals with the associated loop full, we must be at mid-loop.

The reference material provided by SCE was adequate to allow proper addressal of this question and answer.

SCE

REFERENCE:

1.

SD-S01-300, page 21 SCE REQUESTED RESOLUTION:

Delete question; no correct answer provided.

COMMENTS.RO3, REV 2, May 8, 1992

NUCLEAR GENERATION SITE UNIT 1 RESTS DESCRIPTION SD-So1-300 PAGE 21 OF 49 REACTOR COOLANT PUMP SYSTEM P0

2.0 DESCRIPTION

(Continued) 2.2.4 Reactor Coolant Pump Hydraulic Section (Continued)

THERMAL BARRIER PURPOSE: P Cool reactor coolant flowing up through lower radial bearing and seals if injec tion flow is interrupted.

TYPE:

Crossflow HEAT REMOVAL CRITERIA:

Prevent seal and bearing damage if seal injection flow is interrupted.

TUBE SIDE FLOW:

28 -

33 gpm by procedure DESIGN TEMPERATURE INLET:

RCS 553*F CCW 100'F DESIGN TEMPERATURE OUTLET:

RCS 150'F CCW 128*F DESIGN PRESSURE:

2485 psig DESIGN FEATURES:

Five Penetrations:

Seal Injection Component Cooling Water Inlet Component Cooling Water Outlet

&P High Side Tap AP Low Side Tap Supporting Components and Indications The Thermal Barrier lower housing inside diameter contains a stellite overlayed valve seat that mates with a similar seat on the outside diameter of the impeller and acts as a low pressure valve during seal maintenane with the system depressurized and drained to mid-loop.

The thermal barrier is capable of cooling 6 gpm of reactor coolant flow through the seal during loss of seal injection flow.

VP Instrumentation provides indication in the Control Room on the North Vertical Board and controls Seal Injection and Leak-off valves, FCVs 1115 and PCVs 1115.

1.0 OBJECTIVES (Continued) 1.2.3 Describe the physical location of the following major RCP, Seal Injection/Leakoff System components:

RCP, oil lift pump, seal water manifold & injection filters, seal manifold filter bypass MOV's, FCV's-1115A-F, PCV-1115A-C, CV-276, RV-2004, CV's 527 and 528, RV-289, CV-291, seal water return filter and heat exchanger, vapor seal head tank.

1.;.4 Describe the functions for all major components of the RCP and Seal Injection system.

1.2.5 Explain the operation of all major components of the RCP and Seal Injection system.

1.2.6 Describe the requirements for Component Cooling Water (CCW) of the RCP and Seal Injection/Leakoff System. List which RCP and Seal Injection Leakoff System components are cooled by CCW.

1.3 Discuss the name or type, location, function, operation, indication, applicable interlocks, requirements, and applicable normal/abnormal modes for the RCP and Seal Injection system controls.

1.3.1 Describe the physical location of the following RCP and Seal Injection system controls:

RCP oil lift pumps, vapor seal head tank make up, MOV's 18 and 19, FCV's 1115A-F, PCV's-1115A-C, CV-276, CV's 527 and 528 and CV-291 1.3.2 State the interlocks and administrative requirements to start on RCP.

1.3.3 Explain the operating relationship between thermal barrier P and FCV-1115A-F controller response when the controllers are in auto.

1.3.4 Explain how the controllers for FCV-lll5A-F would be manually manipulated to maintain proper seal injection flow.

1.4 Discuss the name or type, location, function, point of measurement and applicable alarms/setpoints for the instruments in the RCP and Seal Injection system.

1.4.1 List the names, type of indication (meter/

recorder) and purpose of the instruments used in the RCP and Seal Injection system.

(~civ

6.0QSON PRESENTATION CONTENT ACTIVITIES

.5 Thermal Barrier (Item #7)

Obj.

1.1.4, 1.2.2, 1.2.4, 1.2.5

.1 Sandwiched between main flange (Item

10) and pump casing, sealed by gaskets (Item 17).

.2 Heat exchanger -

CCW flows through OBj. 1.2.6, 21 GPM design, 28-33 the tubes to cool RCS leakage up to GPM per procedure CCW Flow, 5 GPM the seals.

of RC flow

.3 This is only necessary if seal Obj. 1.1.3 injection fails.

It is necessary so that hot primary water does not damage the seals

.4 Inside diameter of housing contains Obj. 1.1.3, 1.1.4 stellate overlayed seat to mate with similar seat on the outside of the shaft to act as a low pressure valve during seal section maintenance.

.6 Lower Radial Bearing Obj.

1.1.2, 1.2.3, 1.2.4, 1.2.5 (Item #9)

.1 Two piece split housing and bearing cartridge.

.2 Stellate and carbon

.3 Water cooled and lubricated

.4 Maximum operating design temperature RP #1, Window 66 is 300*F. A high bearing water temp.

alarm set at 175*F.

17

RO EXAMINATION COMMENTS Page 11

5.

QUESTION 006 (1.00)

Given the following:

Make-up to the RCS has increased to 95 gpm, and ONLY the following alarms are received:

"RC' PUMP A NO. 1 SEAL LOW DELTA P" "RC PUMP A NO.

1 SEAL LEAKOFF FLOW OFF NORMAL" "RC PUMP A SEAL WATER HI FLOW" "RC PUMP A NO.

2 SEAL HI FLOW" WHICH ONE (1) of the following has occurred to the A RCP?

a.

1. 1 seal has failed.

b.

1. 1 and #2 seals have failed.

c.

All the seals have failed.

d.

Seal injection has failed.

ANSWER:

a NRC

REFERENCE:

(JB)

1.

Lesson Plan 1XA203, Obj. 7.5, page 14, para. 6.2.2.1.2

2.

SONGS Exam Bank, Review Section 2,. 2 of 15

3.

System Description SD-SO1-300, Reactor Coolant Pump System, Page 27, step 3.2.2

4.

S01-13-4, "Reactor Plant No. 1 Annunciator", Windows 32, 52, 72 SCE COMMENTS FACILITY DISCUSSION:

The conditions/symptoms provided to determine the problem would be indicative of both answers "a" and "b".

The only difference between a number 1 seal failure and number 2 seal failure is the.amount of flow past the number 2 seal. A small increase in flow would take place in a number 1 seal failure, or a large increase in flow which would occur for a number 1 and 2 seal failure.

Procedure SO1-2.1-8, Section C, RCP Seal Trouble, does not differentiate between a failure of the number 1 seal and a combined failure of the number 1 and 2 seals. The procedure simply treats the symptoms. Once number 1 seal has failed, it is hard to determine if number 2 seal has failed also.

COMMENTS.RO3, REV 2, May B, 1992

RO EXAMINATION COMMENTS Page 12 The.only means to identify if the number 2 seal has also failed is to have some idea of the amount.of relative leakage past the number 2 seal.

For either a number 1 seal failure only or a number 1 and 2 seal, the operator would expect to see the "RC PUMP A NO.2 SEAL HI FLOW" alarm, which is typically used to determine a number 2 seal problem.

This alarm can only be accurately used when there is a problem only-with the number 2 seal.

If there is a problem with the number 1 seal, then the alarm can indicate-either a failed number 1 seal or a failure of the number 1 and number 2 seals. Therefore, both answers "a" and "b" should be correct. The question should not have asked for this type of discrimination.

The NRC Referenced SONGS Exam Bank, Review Section 2, 2 of

15. A search of the Examination Bank indicated that this question appears to have come from examination number 2228, review section 2, and was question number 2. The NRC examiner changed the question by adding a 95 GPM makeup and a statement that alarm "RC PUMP A.NO.

2 SEAL HI FLOW" was present. The SONGS Examination had only the following alarms:

"RC Pump A No.

1 Seal Low OP"; "RC Pump A No.

1 Seal Leakoff Flow Off Normal"; and "RC Pump A Seal Water Hi Flow".

The NRC's addition of the additional alarm created the situation where their were two correct answers.

REFERENCE MATERIAL:

The NRC Examination reference L.P.

1XA203, Obj 7.5, page 14, para. 6.2.2.1.2 does not support the conclusion provided by the examination writer.

The NRC Examination references L.P. 1XA203, Obj. 7.5, page 14, para 6.2.2.1.2.

There is no objective 7.5 in this L.P.,

an assumption is made that the examiner made a typographical error, and meant to reference 1.7.5, which states "State the symptoms of a failure of #1 seal for a single RCP. Describe the immediate operator actions and identify the applicable AOI."

The referenced page 14, and paragraph 6.2.2.1.2 are not referenced in the lesson plan as supporting this objective. The LP only discusses that the purpose of the floating ring seal is to limit leakage to 100 GPM @.100 psid, in the event #1 seal is lost. None of the other conditions described in the plant conditions are described by this part of the lesson plan. The only reference in the question that appears to touch on this topic is that RCS makeup has increased to 95 gpm. This flowrate was not the referenced number in the lesson plan and does not discriminate between a number 1 seal failure and a number 1 and 2 seal failure.

COMMENTS.RO3, REV 2, may 8, 1992

RO EXAMINATION COMMENTS Page 13 The initial conditions referenced that there is a "RC PUMP A NO.2 SEAL HI FLOW" alarm annunciated. The same L.P.

referenced by the NRC examiner L.P. 1XA203, page 23, 6.2.4.1.11.5 states "Abnormal -

High level may be indication of failure of #2 seal or both #1 and #2 seals".

High level in the Vapor Seal Head Tank referenced above is indicated by a "RC PUMP A NO.2 SEAL HI FLOW" alarm. This section of the L.P. does reference the appropriate objective of 1.7.5.

This section supports again that there is not enough information provided to discriminate between a #1 seal failure only or a #1 and #2 seal' failure.

The NRC Referenced SONGS Exam Bank, Review Section 2, 2 of 15, was inappropriately modified. See previous discussion.

The reference material provided by SCE was adequate to allow addressal of this question and answer.

SCE

REFERENCE:

1.

SD-SO1-300, Reactor Coolant Pump System, page 27, step 3.2.2.

2.

501-2.1-8 page 21, step 7.

3.

S01-2.1-8 background document, section G, page 4 of 6.

4.

SONGS Exam Bank, Review Section 2, 2 of 15

5.

L. P. 1XA203,, Pg 23.

SCE REQUESTED RESOLUTION:

Accept "a" and "b" as correct answers.

COMMENTS.RO3, REV 2, may 8, 1992

NUCLEAR GENERATION SITE SYSTEM DESCRIPTION SD-SO1-300 UNIT 1 REVISION 2 PAGE 27 OF 49 REACTOR COOLANT PUMP SYSTEM 3.0 OPERATION (Continued) 3.2 Abnormal or Emergency Operations 3.2.1 Loss of Seal Water Flow Should seal water supply be interrupted, the Reactor Coolant Pumps can still be operated for short periods of time due to the thermal barrier.

If seal injection flow is lost, the thermal barrier is capable of cooling 'normal seal leak-off flow indefinitely.

Operating Instructions limit the amount of time Reactor Coolant Pumps are operated without seal injection flow because they are prone to failure when operating on unfiltered water.

3.2.2 Failure of Reactor Coolant Pump Seal The.Reactor Coolant Pumps seals are designed to provide redundancy.in controlling leakage. If the Number 1 Seal were to fail, the Number 2 Seal is designed to take over as the main barrier to seal leakage:

The Number 2 Seal is not designed to be operated in this mode for long periods and the Operating Instruction directs the operator to quickly reduce pow6r and stop the RCP within 30 minutes of Number 1 Seal failure. If the Number 2 Seal should fail subsequent to the failure of Number 1 Seal, the injection flow and Number 2 Seal leak-off flow would increase until the differential pressure between the floating ring seal and Number 2 Seal leak-off line equals approximately 2050 psid.

This pressure drop occurs at a flow rate of approximately 100 gpm. The design of the RCP thermal barrier is such that the maximum allowed No.

1 Seal flow rate is limited to 6 gpm to ensure adequate heat removal is achieved to prevent damage to temperature sensitive seal components such as internal 0-rings.

As a result, pump operation is limited to 30 minutes after Number 1 seal failure.

3.2.3 Transfer to Cold Leg Injection and Recirculation After a LOCA or other transient has resulted in safety injection lowering RWST level to near or below -21%, safety injection will be terminated to provide recirculation of the spilled coolant from the containment sump back to the RCS and containment sprays. This is done by resetting the SI; stopping the Feed and SI pumps; placing seal supply.flow controllers in manual and opening MOVs 18, 19, 356, 357 and 358; resetting charging pump lockout and starting the second charging pump; closing FCV-1112 and CV-304, then adjusting seal supply flow to obtain desired value. If instrument air is not available, control of FCV-1115 D, E and F is transferred to the West Vertical Board (FCV-1115 A, B and C fail OPEN). Recirculation Flow to each seal is indicated in the Main Control Room on FI-2114B, FI-2114C and FI-3114A. Once there is adequate recirculation volume, the Recirculation System is placed in service.

(See SD-SO1-580, "Safety Injection and Recirculation System".)

NUCLEAR ORGANIZATION OPERATING INSTRUCTION S01-2.1-8 UNIT 1 ABNORMAL REVISION 2 PAGE 21 OF 34 TCN 9.

C. REACTOR COOLANT PUMP SEAL TROUBLE OPERATOR ACTIONS STEP

.,ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 7

TAKE action as applicable per the following table: (Continued)

RCP SEAL TROUBLE ACTION TABLE SEAL INDICATI6N SPECIFIED ACTION SEAL WATER SUPPLY FLOW is E)

INCREASE SEAL WATER INJECTION WATER FLOW AS REQUIRED TO MAINTAIN

>12 GPM AND THERMAL BARRIER IP NORMAL TEMPERATURE OF THE RCP WATER BEARING.

is NmvAr7frAY INDICATE FAILED NO.1 SEAL)

2) GO TO STEP 8.

NO.2 SEAL LEAKOFF FLOW HIGH

1) ENSURE CLOSED, MAKEUP VALVE TO VAPOR SEAL HEAD TANK FOR AFFECTED RCP.

(VAPOR SEAL HEAD TANK HIGH LEVEL)

2) MONITOR AND LOG THE FOLLOWING PARAMETERS ONCE PER SHIFT ON ATTACHMENT 1 & 2, FOR ANY AFFECTED REACTOR COOLANT PUMP,

a.

SEAL RETURN FLOW.

b. SEAL SUPPLY FLOW.

c.

SEAL SUPPLY TEMPERATURE.

d.

SEAL LEAKOFF TEMPERATURE.

e.

AT BETWEEN SEAL SUPPLY AND SEAL RETURN.

f.

WATER BEARING TEMPERATURE.

.9.

PUMP VIBRATION.

h.

INCREASE IN REACTOR COOLANT DRAIN TANK IN LEAKAGE PLUS NO. 1 SEAL

3)

NOTIFY OPERATIONS MANAGEMENT AND STATION TECHNICAL FOR AN EVALUATION OFSEAL PERFORM.ANCE.

4)IF THE AT B ETWEEN SEAL SUPPLY AND SEAL LEAKOFF EXCEEDS 50*F OR WATER z

EARING TEMPERATURE INCREASES BY -% 5F OR IF NO. I SEAL LEAKOFF PLUS L

THE INCREASE IN REACTOR COOLANT DAAIN TANK IN LEAKAGE IS :- 6 GPM, THEN PERFORM THE FOLLOWING:

BEGIN LOAD REDUCTION TO < 104 POWER.

b.

AFTER LOAD REDUCTION, STOP THE AFFECTED REACTOR COOLANT PUMP.

5)

IF THE AT BETEEN SEAL SUPPLY AND SEAL LEAKOFF IS INCREASING AT MEATER THAN SF/MINUTE THEN PERFORM THE FOLLOWING, PRIOR TO EXCEEDING A MAXIMUM AT OF 100'F: (M~?ENCE 4.4)

a.

IMMEDIATELY TRIP THE REACTOR.

b.

STOP THE AFFECTED REACTOR COOLANT PUMP.

C.

IMPLEMENT SO1-1.0-10, REACTOR TRIP OR SAFETY INJECTION.

6)

GO TO STEP 9.

RCS LEAKAGE TO CONTAINMENT HAS

1) GO TO STEP 8.

INCREASED, AN RCP HAS ERRATIC SEAL INDICATION AND VAPOR SEAL HEAD TANK LO LEVTFALARM IS ACTIVE. (INDICATIVE OF NO.3 SEAL FAILURE)

.MACKJsuuhD DOCUMNTFOR REACTOR COOLANT PULP TROUDLC OPERATING INSTRUCTION S01-2.1-8, REV. 2, TCN 2-1 A note is provided as a reminder that closing pCV-1115A, 8 or C places the affected pump on the No.2 seal which is a limited duty seal.

This should be taken into account when performing the evaluation in the following step.

If seal water cannot be restored to the affected PCP, Step 8 provides direction for the SRO Operations Supervisor to initiate an evaluation for continued plant operation. This evaluation should include as a minimum the impact on continued plant operation, the actions required to isolate and repair any faulty components and any radiological hazards that may be involved.

If continued operation is safe, Step 9 provides direction to continue to monitor the affected RCP for a time period as established by Station Technical. If continued operation is not safe, thena unit shutdown is initiated. When reactor power is less than 10%,

then the affected RCP is stopped and the operator is directed to Step 12.

Step 10 is entered only if seal injection has been restored and directs the operator to verify that PCV-1115A, 8 and C are open. Any PCV found closed is opened. Step 11 gives direction to verify that No.1 seal leakoff flow is less than 5 gpm.

If the No.1 seal leakoff flow for any pump is greater than or equal to 5 gpm, then its associated PCV is placed in automatic. Step 12 provides direction to initiate a LCOAR/EDMR if required.

Step 13 directs the operator to initiate maintenance to correct the problem.

Step 14 directs the operator to resume normal plant operation.at the direction of the Shift Superintendent and then exits this instruction.

Sectin Reactor Coolant Pump Seal Trouble The action steps provided in this section deal with Reactor Coolant Pump seal problems.

Step 1 directs the operator to determine which pump is experiencing seal trouble. Step 2 directs the operator to verify No. 1 seal leakoff flow is less than 5 gpm.

If leakoff flow is greater than 5 gpm, then action is taken to place the seal leakoff isolation valve in automatic. This will allow the valve to close on low thermal barrier differential pressure.

Step 3 directs the operator to determine if the seal water supply flow controller or control valve to the affected RCP has failed.

Seal flow greater than 12 gpm and the high seal water supply alarm are indicative of this problem. If the seal water supply flow control valve is operating properly, the operator is directed to Step 5.

Step 4 provides direction to place the flow controller in manual and adjust seal supply flow to establish a thermal barrier AP of 20"WC.

Step 5 directs the operator to verify that. the seal supply temperature is high, greater than 1400F. If seal supply temperature is less than 1400F, the operator is directed to Step 7. Step 6 provides action to be taken in an effort to lower the seal supply temperature. Step 7 directs the operator t' the Action Table for RCP seal trouble.

This table was developed to lis: the various combinations of seal problem indications that may.e encountered and the actions to be taken for each-and/or a kickout to a subsequent step to continue on with this instruction. The table is laid *out to address problems with the No. 1 seal first, then the io. 2 seal, and then the No. 3 seal.P 4 OF 6 PAGE 4 OF 6

1.0 OBJECTIVES' (Continued) 1.6.3 Describe how changing seal return header pressure will affect the RCP Seal Injection/Leakoff System. Include changes in seal leakoff flows and temperatures.

1.6.4 Describe the operating precautions and limitations of the RCP and Seal Injection system per S01-4-3.

1.6.5 Describe the monitoring requirements for the RCP and Seal Injection System per S01-12.1-4.

1.6.6 Describe the use of procedure 501-4-3 in starting RCP's and aligning Seal Injection for the different RCS conditions as stated in the procedure.

1.7 Discuss the abnormal condition response for an event including the RCP or Seal Injection system including recognition of symptoms event identification, immediate operator actions, applicable procedure use and' a general recovery sequence.

1.7.1 Describe the conditions specified by S01-4-3 that are necessary for opening CV-276 (#1 seal bypass) 1.7.2 Describe the manipulation and automatic operation of PCV's 1115A-C when a high seal leakoff flow condition exists.

Include setpoint and conditions causing PCV's-1115A-C to close.

1.7.3 State the symptoms, immediate operator actions, applicable AOI and general recovery sequence for a complete or partial loss of RCP seal water supply.

1.7.4 Assuming FCV-1115A/D controller fails in automatic causing both FCV-1115A and D to shut, describe symptoms in the control room, immediate operator actions, applicable AOI and the general recovery sequence for this event.

1.7.5 State the symptoms of a failure of the #1 seal for a single RCP. Describe immediate operator actions and identify the applicable AOI.

1.7.6 Describe how additional seal leakoff flow may be provided, per 501-4-3, when establishing seal injection and leakoff for initial RCP starting.

vi

6.$SON PRESENTATION CONTENT ACTIVITIES

.6 Trips Obj. 1.4.4

.1 Overcurrent(3) 50,51, ITH relays All in same case.

.2 Undervoltage on bus

.3 Coastdown (40% voltage if Frequency 5 58 cycles and 11A04 or 11B04 closed.

.4 Undercurrent 137 160 AMPS 1 see (Sheared Shaft Event) Placed inservice 30 sec. after RCP bks closes.

.5 SI

.6 Loss of Voltage Auto Transfer Discuss principle of mechanical seals.

.7 Over current 150-1 1760 AMPS 3 sec. TD (Seized Rotor Event) placed inservice 30 sec. after RCP bks close.

6.2.2 Shaft Seal Section Obj. 1.2.5

.1 Floating ring seal Show TP #1, SEAL, Student Handout Page 6 (Item #63)

.1 Breakdown bushing Obj. 1.1.2, 1.1.3

.2 In case #1 seal is lost, controls leakage 100 GPM @ 2100 psid

.3 14 stationary and 14 floating rings.

14

2 (22 q

REVIEW SECTION 2 PAGE 2 OF 15 (0.5) 2 You are the control room operator when the following alarms are received:

"RC Pump A No. 1 Seal Low AP" "RC Pump A No.1 Seal Lakoff Flow Off Normal" "RC Pump A Seal Water Hi Flow" Based on these indications, what most likely has occurred?

a.

1. 1 seal has failed for the A RCP.

b.

1. 1 and #2 seals have failed on the A RCP.

c.

The whole seal package for the A RCP has failed.

d.

Seal injection has failed to the A RCP.

ANSWER:

a.

1. 1 seal has failed for the A RCP.

REFERENCE:

1XA203, Obj. 7.5

6 @

CONTENT ACTIVITIES

.11 Vapor Seal Head Tank and Makeup Obj. 1.1.4, 1.2.2, 1.2.3, 1.2.4, Valves 1.2.5, 1.3.1, 1.4.1, 1.4.2, 1.4.3, 1.4.4, 1.5.2, 1.7.5

.1 Location -

containment

.2 Function -

accept leakoff from

1. 2 seal and direct it to RCS drain tank -

to provide necessary head to assure lubrication of #3 seal

.3 Control-N.V.B. for makeup valves HCV 427 A, B, C supplied by #1 seal leak off

.4 Indication -

High Level Alarm -

Hi -

RP #1, Windows 72, 73, 74 LC405,415,425 -

Low level

+ 5" Center Line alarm -

LC404,414,424 Lo -

RP #1, Windows 44, 45, 46

-5" Center Line

.5 Abnormal -

High level may be indication of failure of #2 seal or both #1 and #2 seals.

Low level may be indication of

1. 3 seal deterioration.

.6 Power supplies -

utility bus 23

RO EXAMINATION COMMENTS Page 14

6. QUESTION 009 (1.00)

Given the following:

Plant has been shutdown for 6 days.

RCS temperature is 145 degrees F.

RCS pressure is 340 psig.

Pressurizer level is 62%.

S/G's have been in wet layup for 2 days.

Tag on Loop "A" RCP control switch has 190 degrees F written on it.

SRO Ops Supervisor directs the RO to start the Loop "A" RCP.

WHICH ONE (1) of the following allows the start of the Loop "A" RCP.

a.

Pressure is adequate for D/P across the #1 Seal.

b.

Secondary temperature can be determined from steam pressure.

c.

Adequate time for temperature gradients to dissipate has passed.

d.

RCP shutoff temperature is within allowable value.

ANSWER:

d NRC

REFERENCE:

(TV)

1.

S01-4-3, Reactor Coolant Pump Operation, Attachment 5

2.

L.P. 1XA203, Obj. 1.6.6, Pg 28 SCE COMMENTS FACILITY DISCUSSION:

If the procedure referenced by the NRC is used it specifically prohibits starting an RCP with the conditions given in the question. In addition, the initial conditions given are not achievable with San Onofre procedures and requirements. The plant must be solid to achieve a pressure of 340 psig at a temperature of 145 degrees F. The 62% level is misleading.

This is a very complex scenario that requires the use of procedural attachments and curves. An operator should not be required to memorize curves and attachments, only know COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 15 how to use them and the information contained in them. This would be a good open reference style question.

It requires taking information using the procedures to perform calculations as directed and then taking the calculated information and conditions and applying this to a curve.

There is no way that one could expect operators to have the curve memorized or the other details required to-properly answer this question had it even provided a correct answer.

S01-4-3, Page 11, Step 4.4 states: "When RCS Temperature is 5 360 OF, then, a RCP shall not be started unless requirements of Attachment 4 are satisfied."

Based on, the stated conditions would not allow a RCP start, the conditions place the plant in the part of the attachment that is titled "RCP START FORBIDDEN".

Per S01-4 3, Page 15, a RCP could not be started for the following reasons:

a.

Step 6.1.13 requires 350 -

400 psig, initial condition provided in the question was RCS pressure of 340 psig.

b.

The Caution prior to Step 6.1.14 discusses the fact that if pressurizer level is less than 80% there are no temperature restrictions, it states:

"A RCP shall not be started with RCS pressure 5 400 psig unless Pressurizer water level is < 80% OR the temperature difference between the secondary and primary systems is < 50*F."

This OR situation is further discussed in S01-4-3, RCP Operation, Page 12, steps 4.7, 4.7.1 and 4.7.2.

This caution discriminates against the answer selected (d) as the correct answer on the answer key.

c.

Step 6.1.14.2 would not allow a RCP per Attachment 4 discussed above.

REFERENCE MATERIAL:

The NRC reference to L.P. 1XA203, Obj 1.6.6, Pg 28 does not support the answers provided by the NRC Examination.

Page 28 of L.P. 1XA203 does reference Obj. 1.6.6.

The L.P.

tells the instructor to utilize procedure S01-4-3, Reactor Coolant Pump Operations, to discuss: 1. Precautions and limitations, 2. System Alignment, 3. Monitoring, 4. Seal bypass, 5. Abnormal conditions, and 6. Initial RCP starts.

The lesson plan does not discuss any of the criteria addressed by the question.

COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 16 From the above discussion it can be seen that if the NRC examination writer had properly read and interpreted S01-4 3, the procedure he referenced, then the problem discussed would not have existed.

The reference material provided by SCE was adequate to allow addressal of this question and answer.

SCE

REFERENCE:

1.

S01-4-3, Reactor Coolant Pump Operation, Pages 11, and 12; also pg 15, Steps 6.1.13, 6.1.14.2, Caution prior to step 6.1.14, and attachment 4.

2.

Technical Specification 3.1.2.1.

SCE REQUESTED RESOLUTION:

Delete question; no correct answer available.

COMMENTS.RO3, REV 2, May 8, 1992

NUCLEAR GENERATION SITE OPERATING INSTRUCTION S01-4-3 UNIT 1 PRIMARY PLANT REVN 4,

PAGE 11 OF 31 SECTION B:

RCP OPERATION 4.0 PRECAUTIONS (Continued) 4.3 When the RCS Temperature is < 3600F, then at least one of the following overpressure protection systems shall be OPERABLE:

(Tech. Spec. 3.20.A, Reference 2.1.2) 4.3.1 Two Power Operated Relief Valves (PORVs) with a lift setting of 1 465 psig and RHR relief valve RV-206 aligned to the RCS with a lift setting of 1 515 psig (Reference 2.4.6); 2r NOTES: 1) One PORV blocked OPEN and its block valve OPEN meet the requirement of the 1.75 square inch vent. Two PORVs blocked OPEN and their block valves OPEN provide a 3.5 square inch vent. A removed Safety Valve provides a 5.4 square inch vent.

2) Tech. Spec. 3.20 Action C applies whenever RV-206 is.

inoperable or not aligned to the RCS.

4.3.2 A Reactor Coolant System vent(s) of 2 1.75 square inches.

4.4 Wh=n the RCS Temperature is I 360'F, than the following conditions also apply:

(Reference 2.4.6) 4.4.1 One Charging Pump Breaker shall be RACKED OUT.

4.4.2 A RCP shall not be started unless the requirements of are satisfied.

4.5 Whe the RCS Temperature is < 3600F, then the following charging flow limits apply (Reference 2.4.6):

NOTE: The following limits do not apply when the RCS Temperature is

< 210'F, And the RCS is depressurized and vented through.a 1.75 square inch vent(s) (Reference 2.4.6).

4.5.1 If FCV-1112 is not closed and disabled, And is not isolated, 1h the following limits apply:

.1 FCV-1112, Charging Flow Control Valve, shall be limited to less than 150 gpm flow.

.2 FCV-1115A, B, & C, RCP Seal Injection Flow Control Valves, shall be limited to less than 75 gpm total flow.

.3 FCV-1115D, E, & F, RCP Seal Injection Flow Control Valves, shall be MAINTAINED CLOSED.

NUCLEAR GENERATION SITE-OPERATING INSTRUCTION S01-4-3 UNIT 1 PRIMARY PLANT REVISIO PAGE 12 OF 31 TCN SECTION B:

RCP OPERATION 4.0 PRECAUTIONS (Continued) 4.5.2 11 FCV-1112 is closed and disabled, or is isolated, then the following limits apply:

.1 Flow through FCV-1115A/D, B/E, C/F, RCP Seal Injection Flow Control Valves, shall be limited to < 150 gpm.

.2 Only one valve (FCV-1115D,E, or F) shall be used as a flowpath, and the other two valves shall be MAINTAINED CLOSED and DISABLED pr ISOLATED 4.6 If the conditions of step 4.5 are not met, thb the conditions shall be restored within I hour (Reference 2.4.6) And Operations Management shall be notified.

4.7 A Reactor Coolant Pump shall not be started with the RCS pressure 400 psig unless: (Tech. Spec. 3.1.2.1) 4.7.1 The Pressurizer water level is < 80% or 4.7.2 The potential for having developed reactor coolant.system temperature gradients has been evaluated.

NOTE:

There are several means available for determining that there is not a.temperature differential of > 50'F between the secondary and primary systems with < 400 psig primary system pressure. These methods may include but are not necessarily limited to'the following:

(Tech. Spec. 3.1.2 Basis) (See Attachment 5 for explanation of each of the following methods).

1)

Converting steam line pressure indication into maximum temperature of steam generator fluid.

2)

Tagging RCP switches with shutoff temperatures.

3)

Assuring adequate time for temperature gradients to dissipate.

4)

Filling steam generators with water of known temperature.'

NUCLEAR GENERATION SITE OPERATING INSTRUCTION S01-4-3 UNIT 1 PRIMARY PLANT REVISION PAGE 15 OF 31 TCN 1 SECTION B:

RCP OPERATION 6.0 INSTRUCTIONS (Continued) 6.1.7 VERIFY RCP motor oil system as follows:

.1 Oil level at mid range in upper and lower oil level sight glasses, and

.2 Oil collection system is intact, ADA NOTE:

Oil Collection System consists of drain pans, piping, fittings and oil collection tank.

.3 Shows no evidence of oil leakage.

6.1.8 VERIFY the Reactor Coolant Pump Oil Selector Valve is in the UPPER SHOES position.

6.1.9 ESTABLISH VCT pressure for running a RCP per SO1-4-7.

6.1.10 PLACE the RCP vibration recorder and the individual vibration modules in service for each RCP.

6.1.11 If any RCP instrumentation is out of service, thiD NOTIFY the SRO Operations Supervisor prior to starting that RCP.

6.1.12 if the observed Seal Injection System parameters are NOT TM as expected, then CHECK the system for proper alignment per S01-4-36 Attachment. #2, RCP Seal Water System Alignment, as a possible cause.

6.1.13 VERIFY RCS pressure is 350 - 400 psig.

CAUTION During solid system operations, all RCPs shall be momentarily stopped when either RHR pump is started.

CAUTION An RCP shall not be started with the RCS pressure mnsoon

< 400 psig unless Pressurizer water level is < 80%, or the temperature difference between the secondary and primary systems is < 50*F. (Tech. Spec. 3.1.2.1)

(Attathment -5 provides an explanation of 4 methods of determining the temperature difference).

6.1.14 if the RCS temperature is 1 360*F, then PERFORM the following:

.1 DETERMINE the Steam Generator to RCS AT per attachment 5.

.2 VERIFY the Steam Generator to RCS AT requirements of are satisfied.

NUCLEAR GENERATION SITE OPERATING INSTRUCTION 501-4-3 UNIT 1 PRIMARY PLANT REVISION 4 PAGE 29 OF 31 ATTACHME TCN RCP START ALLOtlABLE STEAN GENERATOR TO RCS DELTA-T Af' T

!gcx 4 a.-.S per us& or-A41.

RCP START FORBIDDEN

e.

35 A.

38 585 PRESSURIZER LEUEL AND 27.56F LINIT 25 8

km RCP START 0M IF PRESSURIZER LEUEL (584 Z

44 e

18*

SOLID ICP START 0 118 126 14e 161 igg 200 228 248 261 288 308 328 32a DECREES F INITIAL RCS TEMPERATURE ATTACHMENT 4 PAGE 1 OF 1

NUCLEAR GENERATION SITE OPERATING INSTRUCTION S01-4-3 UNIT 1 PRIMARY PLANT REVISION 4 PAGE 30 OF 31 ATTACH E 5 TCN DETERMINING TEMPERATURE GRADIENTS

1) Converting Steam Line Pressure Indication into Eauivalent Temoerature: The maximum temperature of the fluid in the steam generator can be converted from the steam line pressure. For example, if steam line pressure is ambient, it must be assumed that the temperature of the fluid in the steam generator is 212'F unless another method would support a lower temperature. For higher pressures, the temperature must be assumed to be equal to the saturation temperature at that pressure. This is one method that allows a quick conservative determination to be made regarding the acceptability of starting an RCP. (Section B Reference 2.4.4)

2) Tagqing Reactor Coolant Pump Switches with Shutoff Temperatures: During cooldown to cold shutdown, the reactor coolant pumps are normally kept running until the RCS and steam generator metal temperatures have had time to equalize after reaching the desired cooldown temperature. Once this condition has been established (about 1 hour*), yellow caution tags with final RCS temperature are placed at each switch. This is an indication of steam generator temperature when the pumps were last run. The only way for a > 50'F temperature gradient to develop under these conditions is to heat the RCS to a higher temperature (which could also heat the steam generators), or to cool the RCS > 500F below the temperature at which the pumps were, stopped.

Before starting an RCP the operator would merely check the tagged temperature on the.

RCP switch and determine whether any.recent RCS temperature excursions exceeded the tagged temperature by > 50*F, or if the current RCS temperature is more than 500F lower than the tagged temperature.

By doing this it can be assured that the differential temperature is < 50'F and a pump can be safely started. (Section B Reference 2.4.4)

3) Assuring Adeouate Time for Temoerature Gradients to Dissioate: During extended periods (on the order of at least one week*) of shutdown without any temperature transient in the primary and secondary, the RCS and the secondary side temperatures will equalize. Under these circumstances of extended periods with no temperature transients, an RCP can be safely started. It should also be noted that an RCP can be stopped at any time under water solid conditions and restarted within a short period of time if the cooldown rate has not been too rapid to create a > 50'F temperature differential. For example, when going onto RHR, the RCP's are stopped when an RHR pump is started. If the cooldown rate at this time is 50*F per hour, then the pumps could be restarted within one hour without the possibility of having developed temperature gradients. If more than an-hour elapsed since the pumps were secured, Vh=n an alternate means would be required to determine the maximum temperature gradient. (Section BReference 2.4.4)

These time periods are not based on analysis done to determine RCS and steam generator metal temperatures, rather the time periods are based on conservative engineering judgement.

ATTACHMENT 5 PAGE 1 OF 2

1.0 OBJECTIVES (Continued) 1.6.3 Describe how changing seal return header pressure will affect the RCP Seal Injection/Leakoff System. Include changes in seal leakoff flows and temperatures.

1.6.4 Describe the operating precautions And limitations of the RCP and Seal Injection system per S01-4-3.

1.6.5 Describe the monitoring requirements for the RCP and Seal Injection Systemper So-12.l-4.

1.6.6 Describe the use of procedure SOl-4-3 in starting RCP's and aligning Seal Injection for the different.RCS conditions as stated in the procedure.

1.7 Discuss the abnormal condition response for an event including the RCP or Seal Injection system including recognition of symptoms event identification, immediate operator actions, applicable procedure use and a general recovery sequence.

1.7.1 Describe the conditions specified by S01-4-3 that are necessary for opening CV-276 (#1 seal bypass) 1.7.2 Describe the manipulation and automatic operation of PCV's 1115A-C when a high seal leakoff flow condition exists. Include setpoint and conditions causing PCV's-1115A-C to close.,

1.7.3 State the symptoms, immediate operator actions, applicable AOI and general recovery sequence for a complete or partial loss of RCP seal water supply.

1.7.4 Assuming FCV-1115A/D controller fails in automatic causing both FCV-1115A and D to shut, describe symptoms in the control room, immediate operator actions, applicable AOn and the general recovery sequence for this event.

1.7.5 State the symptoms of a failure of the #1 seal for a single RCP.

Describe immediate operator actions and identify the applicable AOl.

1.7.6 Describe how additional seal leakoff flow may be provided, pDer S-4-3, when establishing seal injection and leakoff for initial RCP

/)11 vi

6.QESSON PRESENTATION CONTENT ACTIVITIES 6.2.5 System Operation

.1 Normal System Operation

.1 Refer to SO-4-3, Reactor-Coolant Pump Operations

.1 Precautions and limitations Obj. 1.6.1, 1.6.2, 1.6.3, 1.6.4, 1.6.5, 1.6.6, 1.6.7, 1.7.1, 1.7.6

.2 System alignment

.3 Monitoring

.4 Seal bypass

.5 Abnormal conditions

.6 Initial RCP Start

.2 Abnormal System Operation Obj. 1.3.5, 1.7.2, 1.7.3, 1.7.4, 1.7.5

.1 Refer to S01-2.1-7, Loss of Reactor Coolant Pump Motor Cooling S01-2.1-8, Reactor Coolant Pump Seal Trouble

.1 High seal leakoff flow

.2 Complete or partial loss of RCP seal water supply 28

H. During MO0E 5 with reactor coolant loops not filled, the following specifications shall apply:

1. Two RESIDUAL HEAT REMOVAL (RHR) TRAINS shall be OPERABLE* and at least one RHR TRAIN shall be in operation".

2. With less than the above required RHR TRAINS OPERABLE, immediately initiate corrective action to return the required RHR TRAINS to operable status as soon as possible.

3. With no RHR TRAIN in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR TRAIN to operation.

I. A reactor coolant pump shall not be started with the RCS pressure j 400 psig unless:

1. the pressurizer water level is less than 80%, or

2. the potential for having developed reactor coolant system temperature gradients has been evaluated. -t U=:l One pressurizer safety valve is sufficient to prevent over pressurizing when the reactor is subcritical, since its relieving capacity is greater than that required by the sum of the available heat sources, i.e., residual heat, pump energy and pressurizer heaters.

Prior to reducing boron concentration by dilution with make up water either a reactor coolant pump or a residual heat removal pump is specified to be in operation in order to provide effective mixing. During boron injection, the operation of a pump, although desirable, is not essential.

The boron is injected into an inlet leg of the reactor coolant loop. Thermal circulation which exists whenever there is residual heat in the core and the reactor coolant systee is filled and vented, will cause the boron to flow to the core.

One RHR TRAIN may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR TRAIN is operable and in operation.

The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

SAN ONOFRE - UNIT 1 3.1-6 AMENDMENT NO:

77, 102, 130

RO EXAMINATION COMMENTS Page 17

7.

QUESTION 012 (1.00)

WHICH ONE (1) of the following occurs at "DIESEL GENERATOR BREAKER CLOSED +2 (TWO) SECONDS" after receipt of a SIS/LOP signal?'

a.

TCV-601 A & B, CCW to RHR Heat exchangers close.

b.

Sequencer output to Load Group "B".

c.

Feedwater pump recirc to condenser closes.

d.

CCW pumps start.

ANSWER:

c NRC

REFERENCE:

(TV) 1..

L.P. 1XC207, Obj 3.8.3, Pg 20

2.

S01-4-46, Safety Injection System Operation"

3.

Drawings 5149178 through 5149182 SCE COMMENTS FACILITY DISCUSSION:

This question is requesting detailed circuitry knowledge that is unexpected of an operator without the.use of references. It is true that allowed to use references, the operator would be able to go to the prints and determine which loads come off of which load groups for the sequencer. However, there are far too many loads to expect that the operator would have each of these load groups memorized.

To have memorized the individual actions that occur for each load group and the start times of each load group associated with the sequencer during a SIS, SIS/LOP, or a LOP is not an expected knowledge at San Onofre. To know what is going to occur to the plant and how each component is expected to operate, i.e should a pump start or stop, or should a valve open or close during these sequencer signals is the expected knowledge.

The wording for answer "c" states that "Feedwater pump recirc to condenser closes" at +2 seconds. The sequencer signal to the valve is generated at that time.

COMMENTS.RO3, REV 2, May 8, 1992

RO EXAMINATION COMMENTS Page 18 The Feedwater pump recirc to condenser valves are also operated based on the position of SI Valves HV853A and HV853B.. HV853A and HV853B receive a signal to operate at time zero (0).

As these valves move off of their closed seat the feedwater recirc valves (CV36 and 37) are operated.

This contact is in parallel with the sequencer contacts, the sequencer contacts are a backup to the limit switches.

associated with SI Valves HV853A and HV853B.

As can be seen if all is working properly the feedwater pump recirc to condenser valves will start to close slightly after time zero, and not at DG Breaker Close +2 seconds.

REFERENCE MATERIAL:

The objective referenced by the NRC examiner 3.8.3 states "Explain how the following components respond to SIS, LOP, or SISLOP:

SISLOP 4KV and 480V Lockouts". The referenced objective is inappropriate, there is no expectation that the operator know what occurs at each given load block of the Sequencer. There is no objective that requires the operator to know this. The referenced portion of the lesson plan, does not reference itself to objective 3.8.3. All that is required is that the operator knows what happens to each component, i.e. end result for plant alignment upon the various sequencer initiation signals. L.P. 1XC207, does break each sequencer into the respective load groups by time just to make the flow of the lesson easier for operator understanding. It is easier to talk in blocks of information, building the entire picture and then summarizing at the end.

If the lesson plan is used in it's entirety, and the.person using the lesson plan understands the systems at San Onofre, it would have informed the question writer that the wording of the question would not have provided the desired response.

Two pages prior to that referenced by the NRC Examiner, pg 18, the reason for the inaccuracy of the question wording is provided. In the content part of the lesson the response for time 0 seconds states Main Feedwater to S/G's isolates, and Feedwater pump's HVs reposition to SI position. From lessons that the students received earlier in their training program, they understand that when these valves reposition, whether initiated by an SI or manual operation of the valves, the recirc valves to the condenser reposition.

The lesson plan does state on the referenced page, pg 20, in the content section, that at time D.G. breaker closed +2 seconds, that the feedwater pump recirc to condenser closes.

The Instructor activity section directly across from this COMMENTS.R03, REV 2, may a, 1992

RO EXAMINATION COMMENTS Page 19 statement clarifies that this is only an initiation signal with the statement "Sequencer output to load group "C".

The students would have already discussed the HV's closure before and be instructed that this is only another signal to close the recirc valves, and that they should already be closed.

Procedure 501-4-46, is not titled "Safety Injection System Operation."

The title of S01-4-46, is "Sequencer Operation."

This procedure does not support the NRC examination question. The procedure simply discusses the following: 1) De-energizing Sequencer Amber Lights, 2)

Energizing Sequencer Amber Lights, 3) De-energizing the Sequencer, and 4) Energizing the sequencer.

Reference to Dwgs 5149178 through 5149182 are the elementaries associated with the Sequencer Load Blocks.

They only support that a signal is provided to feedwater recirc valves at the time specified.

The reference material provided by SCE was adequate to allow proper addressal of this question and answer. However, it would take knowledge of system interrelations not covered in the specific lesson plan referenced.

SCE'

REFERENCE:

1.

L.P. 1XC207, Obj 3.8.3, Pgs 18 and 20

2.

501-4-46, Safety Injection System Operation"

3.

Drawings 5149178 through 5149182

4.

1543 sheet 1

5.

1542 sheet 33 SCE REQUESTED RESOLUTION:

Delete question; no correct answer available.

COMMENTS.RO3, REV 2, may 8, 1992