ML20210C791

Forwards Comments Re 860930 Reactor & Senior Reactor Operator Written Exams Administered by CA Casto.Exams Reviewed by Listed Instructors from Training Unit

ML20210C791
Person / Time
Site:	Robinson
Issue date:	10/02/1986
From:	Morgan R CAROLINA POWER & LIGHT CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20210C684	List: ML20210C682 ML20210C735 ML20210C791
References
NO-86R251, NUDOCS 8702090459
Download: ML20210C791 (52)

Text

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H. B. ROBINSON SEG PLANT POST OFFICE 80X 790 HARTSVILLE, SOUTH CAROLINA 29550 l

I October 2,1986 i

File: NTS-4208(F)

Serial: N0-86R251 Mr. John Munro, Acting Chief Operator Licensing Section U. S. Nuclear Regtlatory Commission, Region II 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 l

SUBJECT:

Reactor Operator and Senior Reacter Operator Written Examinations

Dear Mr. Munro:

On September 30, 1986, Mr. C. A. Casto of the ViNRC administered written reactor operator and senior reactor operator examinctiois at H. B. Robinson.

These examinations were developed by_Mr. K. L. Parkinson and Mr. F. W. Victor of Sonalysts. Enclosed, please find coments for each examination.

Instructors from the H. B. Robinson Training Unit reviewed the examinations for coments. They included:

R. S. Allen W. T. Hensley T. R. Byron D. A. Neal J. E. Gethen V. L. Smith If you have any questions, please contact me or Mr. C. A. Bethea.

Verv truly yours,

/

R. I. Morgan General Manager H. D. Robinson S.E.G. Plant-RSA:eaw Enclosures (5) 8702090459 870112 PDR ADOCK 05000261 V

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" To Serial:

N0-86R251 NRC EXAMINATION COMMENTS H. B. ROBINSON R0 CLASS 86-1 EXAMINATION DATE - SEPTEMBER 30, 1986 1.

Section 1

~

Question 1.01, Part a Recommend also accept Critical Heat Flux divided by Actual Heat Flux.

Reference:

HBR HTT&FF (General Physics, Page 126).

Question 1.04, Part a Recommend accepting a plus or minus 4 BTU per pound mass tolerance since interpretation is' required.

Question 1.07, Parts a and b Recommend tolerance of plus or minus one second to allow for rounding off of numbers.

Question 1.13 Recommend accepting a plus or minus four degree tolerance in answer since a plus or minus two degree tolerance is used in each of the constituent parts of the calculation.

Recommend alternate answer also be accepted based on reasonable assumed value of Core Exit Thermal Couple Temperature for the calculation of subcooling since normal operating conditions were specified.

Question 1.15 Delete word directly from answer and add statement " reasonably worded answers accepted".

Reference:

HBR HTT&FF (General Physics, Pages 313 and 314.)

2.

Section 2 Question 2.08 Sections a and c, recommend tolerance of plus or minus 25 PSIG.

Section b, recommend any value of pressure between 600 PSIG and 700 PSIG based on the minimum and maximum pressure available in the SI Accumulators.

Reference:

Technical Specification, Page 3.3-2,50-002, Section 3.7, Page 20.

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Question 2.09, Part a

-Add to acceptable answers "overcrank or start failure".

Refe.rence: SD-005,-Page 15.

Question 2.13 Recommend also accepting as one of the two choices, pressure.

Reference:

S0-003, Page 11.

Question 2.14, Parts a-and b Part a, recommend also accept five percent runback at 30 second intervals until the initiating signal is clear.

Part b, recommend also accept overpower delta T and overtemperature delta T' runbacks from the Reactor Protection System.

Reference:

S0-011, Pages 27 and 28.

Question 2.15.

Recommend accepting MCC-5 alone, as question only asked for power supply and not path from the power supply.

Question 2.16, Part b Recommend delete Part 1 of question as three-way valves do not f ail open or close.

Part 2, acceptable answer is, VCT.

Reference:

A0P-017, Page 9.

I.

Section 3 Question 3.03 Number 1 answer, recommend accepting Spray Actuation or P Signal, since both terms are synonymous.

Reference:

S0-006, Page 16.

Question 3.04 Answer given is correct but also recommend that points not be taken off for additional information given relating to three out of four power ranges less than P10 "10 percent".

Question 3.05, Part b Question could have two answers depending on which NI fails.

NI-41, 42, c.nd 43 no change since these NI's do not input power nismatch circuitry.

NI-44 decrease since a power mismatch exists between nuclear power and turbine

. power.

Reference:

OST-005, Page 7; Rod Control Handcut 1, Page 11.

2 of 4

e Question 3.07 C

Power range-answer for coincidences is one out of four, not two out of three.

Reference:

S0-011, Page 26.

i Question 3.08 l

Correct answer should be "RHR Valves 750 and 751 (Loop 2 hot leg isolation valves) can not be open unless RCS pressure is less than 465 PSIG".

(This protects-against RHR System overpressure. Other interlocks protect against RWST overpressurization.)

Reference:

SD-003, Page 12; HBR Drawing 5379-1484 (RHR) Note 6.

Question 3.09 The letdown isolation valves will automatically reopen when pressurizer level increases above the setpoint. Therefore, recommending changing the last part of the answer to read " level will oscillate (increase and decrease) around the setpoint, reactor will trip on low RCS pressure (or over temperature delta T)".

Reference:

50-021, Page 17.

Alternate answer, since no operator actions was not specified in the question, therefore, candidate may have answered with no trip due to operator action, i.e., energizing heaters, taking control of pressurizer level, taking channel 459 out of service, etc.

Question 3.10 Correct answer should be "all positions".

(The alarm comes from i delta pressure switch 1608A/B and is set at 2.5 PSID.

It does not matter which position the switch is in.)

Reference:

APP-008-45.

Question 3.11 Load Bistable - answer should be " provide arming signal for steam dump valves". Turbine trip or rapid load rejection was not required by the question.

Temperature Bistable - Delete "or modulation open". Modulation is performed by the valve positioner not the bistable.

Reference:

50-025, Page 14 and 15.

Question 3.13, Part b Recommend delete Part b as could be answered true or false. True if no turbine trip condition exists; fals2 if turbine trip exists.

Reference:

S0-033. Page 20.

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Question 3.15, Part b Delete pressurizer low level as this does not close the ORIFICE isolation valves or initiate an SI signal. Also recommer.d accepting loss of electrical power, loss of air, and phase A isolation signal.

Reference:

50-021, Page 17; Drawing 5379-685, Sheet 1.

4.

Section 4 Question 4.01, Part b Recomn.end also accept "by checking safe. guards status panel lights (pink and blue)'. - (Supplement A only checked if unable to verify from status panel.)

Reference:

OMM-022, Page 10.

Question 4.05

" Prior to starting pump repair" is worth.5 points in the answer key, but this information is given in the stem of the question and should not be required in the answer. Recommend accepting:

"the remaining SI pumps must be tested to assure that they are operable" for one point.

Question 4.11 Recomment also accepting "if_ vacuum is approaching the low vacuum trip, then reduca ti rbine load".

Reference:

A0P-012, Page 4.

Question 4.12, Part a For the TSC and E0F, recommend accepting a general description of the location. Also E0F/TSC/ Training Building are housed in the same building.

Question 4.12, Part b Requi es Technical Specification shift organization. Answer key references PLP-00,7.

Recommend also accepting:

a.

one shift foreman (SR0 license);

b.

one tenior control operator (SR0 license); c.

two control operators (Ro licenses); d.

two additional shift members; e.

one STA.

Reference:

Technical Specification, Page 6.2-1.

i Question 4.13 Recommend eliminating loss of off site power from answer key since loss of off site power is not a prerequisite for implementation of DSP-001.

Reference:

OSP-001, Pages 4 and 29.

END OF R0 EXAMINATION COMMENTS 4 of 4

g 8

, To Serial:

N0-86R251 NRC EXAMINATION COMMENTS H. B. ROBINSON SR0 CLASS 86-1 EXAMINATION DATE - SEPTEMBER 30, 1986 1.

Section 5 Question 5.06 Question gave a thumb rule of minus 10 pcm per ppm for boron calculation, then used minus 9 pcm per ppm in key. Correct answer should be 48.3 ppm (46-51).

Question 5.09, Part a.l.

Recommend also accepting delta T, as delta T is the parameter used as input to RIL computer.

Reference:

50-001, Page 29.

Question 5.10 Recommend deleting Part a due to insufficient information with respect to core reactivity. Higher count rate may indicate closer to critical cgndition due to subcritical multiplication.

Reference:

Reactor Theory Handout, Session 41.

2.

Section 6 Question 6.01, Part b Recommend also accept for answer number 2 in place of overcra.1k, " start failure" as these terms are synonymous.

Question 6.02, Part b Recommend deleting answer Number 4 " pressurizer low level" as prcssurizer low level does not input safety injection or letdown ORIFICE clot.ure signals.

Recommend also accepting loss of electrical power, loss of a.r, and phase A isolation signal.

Reference:

S0-021, Page 17; Orawing 5379-685, Sheet 1.

Question 6.03, Part A Correct answer should be turbine first stage pressure and output of the nonlinear gain unit.

Reference:

SD-007, Pages 23 and 30.

1 of 3

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' Question 6.03, Part b' Number. 3 correct answer is 0 steps per minute.

Reference:

50-007, Pages 25 and 26.

Question'6.06

' Recommend accept any of the two below:

1.

Rod drop NIS 2.

Rod drop /bettom from RPI 3.

Overtemperature delta T signal

4. - Overpower delta T signal

Reference:

~SD-Oll, Pages 27 and 28.

Question 6.08 Part d, correct answer should be 16 percent.

Reference:

Pt ecautions, limitations, and Setpoints Number.l. "PLS-001",

Page 16.

(This number was changed approximately.the same time the-books were sent to the NRC. All personnel were made aware of this change due to the safety significance of this change.)

Question 6.09, Part_a Recomend also accepting any reasonably acceptable source of water from the site since the question did not specify borated water source.

Question 6.12 The letdown isolation valves will automatically reopen when pressurizer level increased above the setpoint. Therefore, recommend changing the last part of the answer to read " level would oscillate (increase and decrease) around the setpoint, react]r will trip on low RCS pressure (or over temperature delta T).

Reference:

SD-021, Page 17.

Alternate answer:

"No operator actions" was not specified in the question, therefore, candidate may have answered with no trip due to operator action (energized heaters, take control of pressurizer level, channel 459 out of service,etc.).

Question 6.13, ? art b Recomend also acceptable answers component coolant water tank level increase andcomponentcoolantwatervalve626(CCWreturn)goesshut.

Reference:

APP-001-17, Page 22.

2 of 3

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3.

Section 7 Question 7.02, Part a Recommend accepting only " path one" and " critical safety function status tree" as answer.

Reference:

General Comment.

Question 7.02, Part c Recommend deleting less than 1200 degrees F and less than 350 degree F from answer.

Reference:

General Comment.

Question 7.05 Recommend delete this question.

Re'ference: General Comment.

Question 7.14 Recommend eliminating loss of off site power from the answer key since loss of off site power is not a prerequisite for implementation of OSP-001.

Reference:

OSP-001, Pages 4 and 29.

4.

Section 8 Question 8.05, Part f Recommend also accepting foreman as he is normally the workers immediate supervisor.

Reference:

OMM-005, Page 8.

Question 8.08, Parts a and b Recommend deleting time requirements, since time requirements are not asked or implied by the question. Also recommend accepting any reasonably worded responses.

Question 3.12, Parts a and b Reconmend also accepting possible answer Number 1 in addition to the given answer, since the RC foreman or higher line management should be notified if an individuals dose:

a.

exceeds regulatory limits, or b.

exceeds 5 rem in a calendar year.

Reference:

OP-004, Page 10.

END OF SR0 EXAMINATION COMMENTS 3 of 3 L.

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= Enclosure 3 To Serial: NO-86R251

GENERAL COMMENT

Please find enclosed the purpose of OMM-022 (Emergency Operating Procedures User's Guide). This states that operator memorization of E0P steps are no longer required. The purpose of this General Comment is to highlight this important concept.

Reference:

OMM-022, Page 3, Section 1.

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' To Serial:

N0-86R251 1

REFERENCE FOR

GENERAL COMMENT

1.0 PURPOSE The purpose of this procedure is to provide instructions for the use of Emergency Operating Procedures (EOP). Memorization, by the Operator, of E0P steps is no longer required. The need for Operator memorization of the "Immediate Action" E0P steps has been eliminated by the development of symptom-based E0P's and the " Flow Path" format of the' initial recovery actions. The Control Room copies of PATH-1 and PATH-2 and the CSFST's are board mounted and kept readily available to the Operators. This results91n timely, consistent response, by the Operators, to events requiring EOP use without dependence on memorization. To successfully use the E0P's, the Operator should be generally familiar with each E0P withia the E0P Network and the " Rules of Usage" contained in this procedure.

2.0 REFERENCES

2.1 NUREG - 0660, NRC Action Plan Developed as a Result of the TMI-2 Accident.

2.2 NUREG -0737, Clarification of the TMI Action Plan Requirements.

2.3 OMM-013, Emergency Operating Procedures Writer's Guide.

3.0 RESPONSIBILITIES 3.1 The Shift Foreman is responsible for insuring that the Emergency Operating Procedures are used correctly and understood by the personnel on shift.

OMM-022 REV. 0 1 of 1

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' Enclosure 5 To S: rial: NO-86R251'Section II Part A, Chapter 3

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1 In practice, if the heat flux is increased, the transition frcm j

nucleate boiling to film boiling occurs suddenly and the temperature difference increases rapidly, as shown by the dashed line on Figure 3-9.

The point of transition from nuclear boiling to film boiling is called the point of departhre from nuclear boiling, commonly written DNB.

The

'l heat flux associated with DNB is commonly called the critical heat flux.

In many applications,. tile critical heat flux is an important parameter, q

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Fc,r exaatple, in a reactor, if the critical heat flux is" exceeded _ and J

DNB occurs at any location in the core, the temperature difference

.l required to transfer the heat being produced from the surface of the 4

fuel red to the reactor coolant increases greatly.

Since the reactor coolant temperature is fixed, this means that the temperature of the surface of the fuel rod increases greatly.

If, as could be the case, tre temperature increase associated with the transition from nuclear y

I boiling to film' boiling causes the fuel rod cladding to exceed its design limits, a failure will occur.

Cladding failure resulting from

}

DNB will be discussed later in this text.

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126

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Ssetion III Part A, Chtpter 2 F

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Then, using the belou equation, for a 4-inch pipe uith cross-sectional 2

area of 0.08840 ft ;

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k - v,Ap b

in = (21.6 ft/sec)(0.08840 ft'E)(62.4 lb]ft )

h=119.2lb}sec A venturi meter is another device used to measure flow.

It consists

,f" of specially constructed converging and diverging pipe sections as shown in Figure 2-9.

A venturi meter is etnalyzed using Bernoulli's equation and the continuity equation.

The resulting relationships for the aver-L age veloc'.ty in the venturi nozzle v and the mass flow rate in are as'follows:

7_.

2g AP v

=C 2-

• ' 2 (2-26)

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f 2g aP

(

m = C,A,a 1

A )l (2-27) p f

( 1 L

where v,, = average velocity (ft/sec)

AP = differential pressure (Ib,/ft.,

2 a

p

= density (Ib,/ft ).

L A = cross-sectional area (ft')

5 = mass flow rate (Ib,/se:)

7 L.

C. - venturi factor (no units) g

= acceleration due to gradty = 32.2 ft/sec8 The venturi factor C is added to account for friction losses.

The venturi factor equals about 0.98 for most venturi meters.

313

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S;ction III Prrt A, Ch;:ptcr 2

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An orifice meter is another device used to measure flow.

It oper -

ates in a manner similar to that of a venturi meter.

Rather than having gradually converging and diverging sections, however, the changes are sharp.

A typical orifice meter is shown in Figure 2-10.

The relation-ships for the average velocity in the orifice v and the mass flow j

2 rate 6 are the same as for a venturi meter.

However, the orifice co-efficient C is used in place of the venturi factor C.

The orifice 9

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coefficient ranges from about 0.5 to 0.97, depending on the Reynolds number of the flow and the size of the orifice.

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Liquid L 1*

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Figure 2 9 Typical Venturi Meter m

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ce ~

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b.

Each accumulator is pressuriced to at least 600 psig and 3

3 contains at least 825 ft and no more than 841 ft of water with a boron concentration of at least 1950 ppm.

No accumulator may be. isolated.

c.

Three safety injections pumps are operable.

d.

Two residual heat removal pumps are operable.

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e.

Two residual heat exchangers are operable.

f.

All essential features including valves, interlocks, and piping associated with the above components are operable.

g.

During conditions of operation with reactor coolant pressure in excess of 1000 psig the A.C. control power shall be removed from. the following motor operated valves with the valve in the specified position:

x Valves Position MOV 862 A&B open MOV 864 A&B Open MOV 865 A,B,&C open MOV 878 A&B Open MOV 863 A&B Closed MOV 866 A&B Closed h.

During conditions of operation with reactor coolant pressure in excess of 1000 psig, the air supply to air operated valves 605 and 758 shall be shut off with valves in the closed position.

i.

Power operation with less than t,hree loops in service is prohibited.

3.3-2 Acendment No. 97 Rev. 92

pt-W 1-W 4 u S 2.08

=

3.0 INSTRUMENTATION AND CONTROLS (Continued)

-k..

1 CONTROLLER NO.

WINDOW NAME WINDOW NO.

FIC-658 SI PMP C00LINC WTR LO FLOW APP-002-37 Breakers CONT SRY PMP A/B MOTOR' APP-002-41 OVERLOAD TIC-934 A & B BOR INJ TK HEADER HI TEMP APP-002-43 PT-929/PT-93 L ACCUM TK C HI/LO PRESS APP-002-44 Breskers SI PMPS MOTOR OVRLOAD/ TRIP APP-002-45 PC-951B, 9535, CONTAINMENT HI OR HI-HI APP-003-36 955B, 951A, 953A, PRESS 950, 952, 954, 955A TS-A2 SAFETY INJ PUMP AREA TEMP APP-004-24 HI 3.7 Relief Vatve Setpoints VALVE NO.

LOCATION SETPOINT SI-857A & B Baron Injection Header 1750 psig t 52 psig SI-858A, B, C Safety Injection 700 psig i*21 psig Accumulator A, B, i: C SI-859 High Head Safety Injection 1750 psig 2 52 psig Test Line SI-871 Containment Spray Pump 200 psig 2 6 psig Suction SI-872 Spray Additive Tank 270 psig i 8 psig 3.8 Instrument Setpoints INSTRUMENT NO.

FUNCTION SETPOINT PCV-937 Accumulator Nitrogen Supply 675 psig 110 psi Regulator Actumulator Level High Alarm 78%(+0, -2):

Low Alarm 64%(+2, -0)%

's.

i SD-002 Rev. 3 Page 20 of 34 l

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<J 4.2-cos d 15 OF 24 Diesel Generators u.

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2.0 INSTRUMENTATION AND CONTROLS (Continued)

LIGHT NAME CONTROLLER -NO.

Local Control (W)

Local / Remote Switch Lube Oil Temp Hi (W)

OTHA Lube Oil Temp Lo (W)

OTLA Coolant Temp Hi (R)

CTHS Coolant Te=p Lo (W)

CTLA 10 Second Overcrank (R)

Speed Switch Overspeed (R)

Overspeed Trip S tarting Air Pressure Low (W)

APLA Crrnkcase Pressure Hi (R)

Crankcase Pressure Switen Day Tank Level Hi (W)

LSHH-1963A for "A" Diesel LSHH-1963B for "B" Diesel Day Tank Level Lo (W)

LSLL-1963A for "A" Diesel LSLL-1963B for "B" Diesel Lube 011 Press. Low (R)

OPLS Service Water Pressure Low (W)

WPLA Low Coolant Pressure (R)

CPLS NOTE:

(W) denotes alarm function only (R) denotes diesel engine trip 2.4 Indications 2.4.1 RTGB Gauges on the RTGB indicate diesel generator output voltage and amperes on each generator when it is running.

Indicating lights on start switch panels indicate whether s.

diesel is stopped, starting or running.

There are also indicating lights for generator output breaker position.

MdNcc Fod.

2,/3,

m 5.0 OPERATION (Continued)

The system is pressurized by isolating it from the Refueling Water 4

Storage Tank and using HCV-142 and PCV-145 (Chemical & Volume Control System) which is set at approximately Reactor Coolant System

. pressure. The loop suction valves (RER-750 and RER-751) are opened when the system is pressurized. Flow through the Residual Heat Reroval to Chemical and Volume Control System letdown line will warm Reutdual Heat Removal System and equalize bor'on. During this evolution the Residual Heat Removal Pumps are alternately started

.and stopped to minimize the temperature difference between the two loops.

When boron and temperature are close to that of the Reactor Coolant System, the system is put into operation by opening discharge valves to the Reactor Coolant System cold legs.

![

The race of heat removal from the reactor coolant is controlled by

. regulating the reactor coolant flow rate through the Residual Heat Removal Heat Ex9 angers with HCV-758, from the control board. Flow h

costreller'FCV-605 in the Residual Heat Removal System operates a coatrol valve in the bypass line around the Residual Heat Removal Hest Exchangers to maintain the Residual Heat Removal System design flow rate at a constant value. As the reactor coolant camperature decreases, the flow through the Residual Heat Removal Heat Exchangers is reduced using HCV-758 and the flow through the by-pass line is automatically increased.

During Post Fire Repairs the Valves HCV-758 and FCV-605 have been providad with an alternate means of motive power, that is nitrogen.

The required Tie-in equipment and materials have been stored on site for use during Post Fire Repairs.

, s_

l SD-003 Rev. 2 Page 11 of 14

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3.0 INSTRLHENTiTION AND CONTROLS (Continued) 3.1.8.1 Load Reference Reduction The Load Reference Reduction is caused when any of the following occur:

a.

Rod Drop Signal (NIS) - when 1 out of.4 Power Range Channels in Train "A" senses a 5% change in 5 seconds. This causes the Turbine to reduce load at the rate of 200%/ minute for a duration of 9 seconds (this equals to a,30% Load Reduction).

The Power Range Channel that caused the runback is reset or bypassed at it's respective Power R.inge A Drawer and has to be reset or bypassed before another runback can occur.

b.

Overtemperature AT - when 2 out of 3 Reactor Coolant Loop AT's 4

(T - T ) exceeds the calculated setpoint. This causes a h

cyclic reduction in Turbine Load at the rate of 200%/ minute.

The cyclic duration is that of a 1.'i second runback (which equates to 5% Load Reduction) and a 30 second wait, if the Overtemperature AT condition still exists, the 1.5 second runback and the 30 second wait repeats. This cyclic reduction continues until the Overtemperature AT :endition is corrected.

The setpoint is identical to the Reactor Trip Setpoint except the K term becomes 1.1265 (Refer to 3.1.5.5).

g c.

Overpower AT - when 2 out of 3 Reac.:or Coolant Loop AT's (Th~

T,) exceeds the calculated setpoint. Inis causes a cyclic reduction in Turbine Load at the rate ef 200%/ minute. The cyclic duration is that of a 1.5 second runback (which equates to 5% Load Reduction) and a 30 second wait, if the Overpower AT condition still exists, the 1.5 second zunback and the 30 second wait repeats. This cyclic reduction continues until the Overpower AT condition is corrected.

The Setpoint is identical to the Reactor Trip Setpoint except the K term becomes 1.04 (Refer to 3.1.5.6).

4 SD-011 Rev. 2 Page 27 of 35 l

I fEFG42iMK FrJR 2, l4 3.0 INSTRUMENTATION AND CONTROLS (Continued) 3.1.8.2 Load Limit Reduction The Load Limit Reduction is caused when 1 out of 4 Power Range Channels (NIS) in Train "B" senses a 5% change in 5 seconds or any Rod Bottom Signal (20 Steps) from the. Rod Position Indication (RPI)

Circuitry is received and both Turbine First Stage Pressure Elements are above the preseure which correlates to 70% Turbine Load. This.

causes the Turbine to reduce load at the rate of 200%/ minute until

~

at least one of the two Turbine First Stage Pressure Elements is below the pressure which correlates to 70% Turbine Load.

Should the initiating NIS and/or RPI signal be reset or bypassed before the 70% Turbine Load Pressure is reached, the runback will stop.

Th.re is no Lead Limic Reduction if Turbine First Stage Pressure is less than 70% Turbine Load.

3.2 Annunciators and Alarms 3.2.1 First Out Reactor Trips Annunciator All Reactor Trip Signals have a First Out Annunciator. Other alarms on the RTGB will also alarm frem the same bistables and are covered in their respective System Description. When a Reactor Trip Signal is received, the appropriate First Out Annunciator will come in Blinking. Any other trip-signal that occurs will come in Solid (not Blinking). The operator should note which alarm (Blinking) caused the trip before depressing the Reset Pushbutton. Any light that is solid will reset (Go out) when the Actuation Signal is cleared.

i SD-Oli Rev. 2 Page 28 of 35 l

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4.0 GDERAL (continued) 4.2 A sustained loss of Instrteent' Air could lead to a Plant shutdown and/or loss of control of major system valves.

Major system valves that could be affected are:

Valve Number and Descriction Failed Position 1.

KV-l'.3A (Boric Acid to Blender)

Fall Open 2.

FCV-ll4A (Primary Water to Blender)

Fail Closed 3.

FCV-113B (Blended Flcw to Charging Pump Suction)

Fail Closed 4.

FCV-ll4B (VCr Makeup)

Fail Closed 5.

LCV-115A (VCr to Holdup Tank Diversicn)

Fail to VC2 6.

LCV-115B (Emergency Makeup Tb Charging Pump Suction)

Fall Closed 7.

CVC-310 A&B (Charging Flow to Loops 1&2)

Fail Open 8.

CVC-311 (Auxiliary Pressurizer Spray)

Fail Closed 9.

LCV-4601.1B (Letdown Line Stop)

Fall Closed

10. CVC-200 A. B, &C (Letdown Orifice Isolation)

Fall Closed

11. CVC-204 A1B (Letdcwn Line Isolation)

Fail Closed

12. ECV-137 (Excess Letdown Flow)

Fail Closed

13. BCV-121 (Charging Flcw Centrol Valve)

Fail Open

14. Charging Pump Speed Control Full Speed v.

ACP-017 Rev. 2 Page 9 of 10

f.I. p (4, fi2( 3.o3

+*

m r ' ~.

\\ ~

5.0 OPERATION (Continued)

NOTE "B'! and '!C" Charging Pumps will be tripped if a SI signal is present and power is being supplied by diesel generators.

NOTE "B" and "C" Component Cooling Pumps will be tripped if a SI and a spray signal are present and power is being supplied by diesel generators.

NOTE "B" and "C" Component Cocling Pumps rill not start automatically on low pressure if the diesel generator,is supplying its respective bus.

NOTE "A" Component Cooling Pump will start anytime on low pressure if power is available.

"A" pump is en the DS Bus.

5.4 Actions That Can Be Initiated By Other Signals 5.4.1 Steam Line Isolation - As previously noted, a spray ar.tuation (P-signal) will closu all three main steam isolation valves.

Th..s action will also occur if we have a high steam line flow coincident with low steam line pressure or low T,y,,,g,.

(Does not occur on manual spray actuation)

They can be shut individually from the RTCB by their control swit'ch or by the steam lina isolation pushbuttons.

5.4.2 Feedwater Isolation - As previously noted, a SI ac:uat. ion will cause a complete feedwater isolation. A reactor trip with auctioneered T ay,,,g, less than $54*F will shut the main feedwater regulatir.g valves. A high-high steam generator level (2/3 @ 75%) will shut its respective main feedwater ragulating valves and bypasses and trip both main feedwater pumps.

The Feedwater isolation signal must be reset manually if it was caused by an SI signal or steam generator high-high level before normal operation can resume.

SD-006 Kev. 7 Page 16 of 13

Agetoace PA 3. of pMa CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON SEG PLANT PLANT OPERATING MANUAL VOLUME 3 PART 9 '

OPERATICNS SURVEILLANCE TEST PRCOEDURE OST-005 NUCLEAR INSTRUMENTATION POWER RANGE (BI-WEEKLY)

(POWER LEVEL ABOVE P-8)

A REVISION 3 Effective Date 2-01-85 RECOMMENDED BY:

dL

/-2 I-h Unit 2 - Operating Supervisor Date APPROVED BY:

N.[

} -2 0 -$ $

..,-4 Manager - Opera:1cns InElaintenance Date N

CO \\~RO _'_EJ CO?Y r+ /

Page 1 of 14

6&retua.d.

F<.%

4 vs Sr,ction 7.1 380f Paga 1 cf 6

-s' 7.0 PROCEDURE N-41 N-42 N-43 N-44 7.L Power Range Channel Tests N-41[N-42/N-43/N-44 7.1.1 Power Range Channel being tested.

7.1.2 Remove the computer point for the channel being tested fran Scan'(N-41* F0049A, N-42: N0050A, N-43: N005LA, N-44: N0052A).

7.1.3 Verify the ROD BA?IK SELECTOR. Switch,on the RTGB,

-)l 1s in MANUAL when testing Power Ragge Channel N-44.

i 7.1.4 Place the DROPPEP ROD MODE switch, on Power Range A 6

drawer, in the BYPASS position.

7.1.5 Verify the DROPPED ROD BYPASS indicator, on Power

~

Range A drawer, is illuminated.

l 7.1.6 Verify the NIS ROD DROP BTPASS status lignt, on the t

t RIGB, is illumint.ted for the channel being tested.

7.1.7 Verity the NIS TFIP 3TPASS annunciatce, on the RTGB, La illuminated.

.i 7.1.8 Place OVERPOWER 4T and OVERTEMPERATURE 4T Reactor Trip Bistables, associated only with the Power Range j

channel being tested, in the TRIPPED mode.

7.1.8.1, 'N-41 Associated with Protection Channel No. I i

Location:

l i

Rack No. 1 BS-412C-1 r

1 Rack No. 1 BS-412B-1 l

N-42 Associated with Protection Channel No. II 7.1.8.2 Location:

Rack No. 11 BS-422C-1 Rack No. LL BS-422B-1 7.1.8.3 N-43 Associated y,ich Protection Channel No. III I

Location:

Rack No. 14 BS-432C-1 Rack No. 14 BS-432B-1 OST-005 Rev. 1 Page 7 of 14 9

e

,-_m,

.e -o--

e.

- _ ~ _

,. e

Receuwe rat-summing unit.

This circuit compares the rate of change of nuclear power to the rate of change of turbine load. When the rates of change are equal, the circuit provides no input signal to the summing unit allowing the temperature error signal to make the fine adjustments to Tavg. The input signals to this channel are reactor nuclear power (Qn) and turbine load (Qtu)- Qn is provided by an input from power range channel N-M.

Qtu is developed from impulse turbine first-stage pressure. The rates of change between Qn and Qtu are compared in the rate comparator circuit. When the rates of. change differ, an error signal j

is produced and sent to a non-linear gain unit.

The non-linear gain unit converts the power mismatch signal to a temperature signal.

As shown in RDCNT-TP-1.4, when the power mismatch signal is greater than 1%, it is multiplied by 1.5 F/% mis-match to produce the temperature error signal.

When the power mismatch signal is equal to or less than 1%, it is multiplied by 0.3 F/%

mismatch to produce the temperature error signal. The output of this unit is then sent to the variable gain unit.

s The variable gain unit compensates for reactivity changes.

This is needed because the reactivity changes at low power levels have smaller effects on nuclear power than at higher power levels.

As shown in RDCNT-TP-1.4, the temperature signal input is multiplied by 2 when turbine load is equal to or less than 50%, is multiplied by 1/Qtu when turbine load is between 50% and 100%, and is multiplic.d by I when turbine load is greater than 100%. This adjusted temcerature signal is j

then sent to the summing unit.

1 The summing unit adds the three input signals and produces a total temperature error signal, T, for use in the rod speed programmer and E

the IN and OUT red direction bistables. As shown in ROCNT-TP-i.5, the rod speed programmer produces current signals between 10 and 50 milliamperes for use by the pulser unit in developing rod speeds between 3 and 72 steps per minute, respectively. As the temperature error signal increases from -2.5 to +0.5 F, the programmer produces no output, and i

the rods are held in position. This is called the dead band. At +0.5 F temperature error signal, the programmer will produce. a 10 milliampere il sf $2

(MAENld b*05 3.0 INSTRUMENTATION AND CONTROLS (Continued)

An Automatic Rod Withdrawcl is blocked by:

a.

1 out of 4 Power Range Channels above 103%, this block can be bypassed.

b.

1 out of 2 Intermediate Range Channels above 20T, this block can be bypassed.

I out. of 3 Overtemperature AT's above calculated setpoint, this c.

calculated setpoint-is less than the Overtemperature AT Trip.

d.

1 out of 3 Overpower AT's above calculated setpoint, this calculated setpoint is less than the Overpower AT Trip.

e.

Turbine First Stage Pressure less than 15%.

f.

Dropped Rod L.

Rod Bottom Signal (Red Position Indication) 2.

I out of 4 Power Range Channels Rod Drop Signal (Train "A" and Train "B")

(

A Manual Rod Withdrawal is biccked by:

a.

I out of 4 Power Range Channels above 103%, this block can be

bypassed, b.

1 out of 2 Intermediate Range Channels above 20%, this block can be bypassed.

2 out of 3 Overtemperature AT's above calculated setpoint, this c.

calculated setpoint is less than the Overtemperature AT Trip, d.

2 out of 3 Overpower AT's above calculated setpoint, this calculated setpoint is less than the Overpower AT Trip.

3.1.8 Turbine Runbacks The Turbine Runbacks are designed to reduce Reactor Load and thus avoid an unnecessary trip. The Turbine Runbacks are caused by Rod Drop Signals or High AT Signals and are initiated by either a Load Reference Reduction or a Load Limit Reduction.

SD-011 Rev. 2 Page 26 of 35 l I

fErseence FM 3 08

~

,r -\\

5.0 OPERATION (Continued)

N.

Valves RHR-750 and RHR-751 cannot be opened unless Valyes SI-862A and B and SI-863A and B in the Safety Injection System are closed.

~

Below 210 psig, in the Residual Heat Removal 3tstem, SI-862A and B and SI-863A and B may be opened. This is to avoid depressurizing the Reactor Coolant System to the Refueling Water Storage Tank and/or overpressurizing the low pressure portions of the Safety Injection System.

Valve RER-750 is interlocked with Reactor Coolant System pressure so that it may not-be opened above 465 psig. This protects against system overpressure downstream of the RHR pumps and still allows Residual Heat Removal System operation to be initiated before the Reactor Coolant Pumps have been stopped at 325 psig.

5.3 Plant Hertuo On a plant start-up ::he Residual Heat Removal System will be removed

(,

from service after a bubble has been formed in the Pressurizer.

Basically the Residual Heat Removal System has to be isolated from the Reactor Coolant System and cooled down alternating Residual Heat Removal Pumps. Then the system can be depressurized by reducing letdown pressure in Chemical and Volume Control System and opening HCV-142. -The Residu 1 Heat Removal System is then isolated from letdown and lined up for automatic Low Head Safety Injection.

S.4 Refueling Operations During refueling the Residual Heat Removal System is used to maintain Reactor Coolant System camperature within required limits.

The Res'idual Hea: Removal System can be used to fast fill the Refueling Canal.

SD-003 Rev. 2 Page 12 of 14

r FfGLGMcG FoQ Y]

i.

2.0 COMPONENT DESCRIPTION (Cuatinued) 2.21 Discharge Pulsation Dampeners The discharge pulsation dampeners are spherical type vessels installed in each of the charging pump discharge lines.

Their internal baffles reduce discharge pressure pulsations.

3.0 INSTRUMENTATION AND CONTROLS All valves discussed in this section are operated from th.2 RTG3 unless otherwise specified. The failed position for the air operated valves can be found in the system drawing. All instrumentation that provides local indication only can be found in the system drawing.

3.1 Valves

~

3.1.1 Letdown Stop Valves (LCV-460A and 460B)

{' '

Both valves are controlled by one three position switch.

The switch positions are: OPEN, CLOSE and AUTOMATIC. These valves are closed automatically on a pressurizer low level alarm and will automatically re-open when the low level alarm clears. T*tese valves are' located in "A" reactor coolant pump bay.

3.1.2 Letdown Orifice Isolation Valves (CVC-200A, 200B,*200C)

Three air operated valves are provided to determine which letdown orifices are in service. One orifice will pass 45 spm and the other two will pass 60 gpm each when the RCS is at normal pressure and letdown pressure is adjusted to approximately 300 psig. Care should be taken not to exceed design flow race through the deminerall:ers.

These valves are located next to the letdown orifices in the Regenerative Heat Exchanger cubicle. These valves will close on a Phase "A" Contain=ent Isolation Signal (T signal).

5D-021 Rev. 4 Page 17 of 55

APP-008-45 REFE% Enc 4 rox 3* g AMRM

'C SERVICE WIR STRAINER A/B PRESS HI

• • • NO REFLASH ***

AUTOMATIC ACTIONS.

1.

None Applicable CAUSE 1.

. Strainer Screen has not Backwashed 2

Motor Fault 3.

Failure of Scraper "

4 Strainer Screen did not Flush 5.

Controller Failure OBSERVATIONS 1.

Service Water Strainer Differential Pressure ACTIONS 1.

Flush Screen.

2.

Check Flush Discharge valves.

(

DEVICE /SETPOINTS l.

DPS'-1608A/2.5 psid 2.

DPS-1608B/2.5 paid POSSIBLE PLANT EFFECTS 1.

Loss of affected Service Water Supply Line REFERENCES 1.

AOP-032, Loss of Service Water 2.

Tech. Specs. 3.3.4 APP-008 Rev. 2 Page 48 of 51 l

J

' :-l (Ie m ci r-*A. 3.1I PAGE

/f;,

TITLE REV.

PROC. NO.

14 26 Main S eam System 0 ~

J'e--55 OF C

2.0 INSTRUMENTATION AND CONTROL (Continued) trip.

The TAVG control position places the steam. dump in automatic controllin order that it can function'following-a loss of load or turbine trip at which time the dump is' controlled by Tavg.

The steam pressure control 'osition p

places the dur.>. system under the control of the steam. header pressure.

A three position switch is provided to allow bypass of the low Tavg interlock during a planned, controlled plant cooldown.

This switch has the'following pos cions:

OFF RESET TAVG FYPASS*

P DN BYPASS TAVG INTERLOCK (spring return to ON) 2.10.2 Sudden Loss cf Load Bistables These bistabJes provide the arming signal for the steam dump valves in tre event of a turbine trip or rapid load rejection.

An arming signal opens a solenoid valve in the nitrogen supply (condenser dump valves) but does not open the valves.

The arming signal for PORV's shifts its controls fron normal to the steam dump conttoller.

Setpoints are:

A.

Inpulse Unit Time Constant of Loss of Load Interlock Channel.

(PM-447C) 120 second.

B.

Sudden Loss of Load 31 stables W

fEF6t.geJc6 Fok be] $

r.'

PAGE l

TITLE REV.

PR OC. NO.

(

..l5, op _2.1 Main Steam System 2

SD-025 2.0 INSTRUMENTATION AND CONTROL (Continued)

Controller NL Arms Se tpoint (PM-447A)

• 3 Condenser Dumps 15% of full load (PM-4473)

• 2 Condenser Dumps 35%'of full 1 cad (PM-447D)

• • 3 PORV's 70% of full load

• NOTE': The condenser dumps can only be armed if the condenser is available, i.e., at least one circulating water pump running and suf ficieat vacuum in condenser.

• • NOTE:

The PORV's can be armed only if the turbine is not tripped.

2.10.3 Temperature Bistab les and Contre 11e 3 Temperature bistables are provided to trip open the valves if t, hey are armed. The trip open feature is accomplishec.

by a three way solenoid that bypasses the positioner and applies nitrogen directly to the condenser dump valves or instrument air directly to the PORV's.

The bistables that are allowed to trip open the valves are determined by whether or not the turbine is tripped.

If the signal is not sufficient to trip the bistables, the valves may be nodulated by thu turbine trip or load rejection controller.

A.

Turbine not tripped (Load Rejecti'on)

The bistables and controller for a ic.d rejection receive their signals from auctioneered Taveragu and first stage pressure (PT-446).

L.

High (Tavg - Tref) three valves trip open (Bistab le TC-408F) 12'.1 F ja

PAGE TITLE REV.

PROC, NO.-

40-o1-7 Turbina.and Controls 0

20 80

- 0. ~

OF 1.0 GENERAL DESCRIPTION (Continued) kEF6taaJc1E Tbit 3.I3 i

Two oil coolers are pr uided, being connected by a tanden- '

f operated three-way valve to switch from one cooler. to the other, as desired.

When changing over, follow instructions given on the valve handwheel.

The oil inlet to each cooler is connected through a crossover pipe and interchange ' valve.

.This interchange valve s,hould be open at all: times. during normal operation to permit the inactive cooler to be filled with oil. When it is desired to clean one of the' coolers, the interchange valva shculd be closed until the cleaning.

has been completed.

The interchange valve should then be opened. A sight flow is installed in each of the oil cooler vent lines,to the reservoir which indicater. when the coolers

~

are filled with oil.

Funda=entally, the overspeed trip valve, protective trip de-vices and the interface emergency trip. valve are hydraulically operated utill:ing orificed auto-stop oil s' applied to.the oil header by the main oil punp discharge.

Relief valves are provided in this line to limit the oil pressure to 90 - 100 psig.

The auto-stop oil header is linked to the high pressure fluid emergency trip header by means of the diaphragm-operated interface emergency trip valve.

The operation of any turbine trip device, with the resultant loss of auto-stop. oil pressure will open the interf ace emergency trip salve, releasing to e

pb

~

Repeteac w

3. lY

-~x f

t 2.0 COMPONENT DESCRIPTION (Continued) 2.21 Discharge Pulsation Dampeners The discharge pulsation dampeners are spherical type vessels installed in each of the charging pump discharge lines. Their internal baffles reduce discharge pressure pulsations.

3.0 INSTRUMENTATION AND CONTROLS All valves discussed in this section are operated from the RTGB unless otherwise specified. The failed position for the air operated valves can be found in the system drawing. All instrumentation that provides local indication only'can be found in the system drawing.

3.1 Valves

(

3.1.1 Letdown Stop Valves (LCV-460A and 460B)

Both valves are controlled by one three position switch.

The switch positions are: OPEN, CLOSE and AUTOMATIC.- These valves are closed automatically on a pressuriser low level alarm and will automatically re-open when the low level alarm clears. These valves are located in "A" reactor coolant pump bay.

3.1.2 Letdown Orifice Isolation Valvec (CVC-200A. 200B. 200C)

Three air operated valves are provided to determine which letdown orifices are in service. One orifice will pass 45 gpm and the other two will pass 60 gpm each when the RCS is at normal pressure and letdown pressure is adjusted to approximately 300 psig. Care should be taken not to exceed desit;n flow rate through the demineralizers.

Th'ese valves are located nest to the letdown orifices in the Regenerative Heat Exchan,ger cubicle. These valves will close on a Phase "A" Containment Isolation Signal (T signal).

SD-021 Rev. 4 Page 17 of $5

hEFEA.Eelc6 M

• O 5.0 PROCEDURE (Continued) f Brackets are used to indicate values that should be used if adverse containment conditions are present. Adverse containment conditions exist when containment pressure is greater than 4 psig.

Example: Verify RCS Subcooling - GREATER THAN 25*F (45'F]

5.1.2 Foldouts Several E0Ps require the use of a. FOLDOUT. The applicable FOLDOUT is intended to be available (visible) when the E0P is being performed. The FOLDOUT page contains several pieces of information or actions which are applicable at any step in the E0P being used.

The mest important of these actions are procedure transitions which allow immediate response to new symptoms as they appear. The placement of these transitions on the FOL."OUT allows prompt response to the appearance of subsequent symptoms. The contents of each FOLDOUT specify actions which must be taken when the symptoms

~'

associated with that action are recognized.

5.1.3 Supplements The EPP Supplements are used to provide information too lengthy to include on a PATH, EPP or FRP. Specific uses are described in the following paragraphs.

In PATH 1, the Operator is expected to be able to verify proper SI,,

Phase A, Containment Ventilation Isolation, etc., valve alignments l

from the Safeguards Status Panel (pink & blue lights).

If he is unable to verify proper alignment from this panel, he can then refe to Supplement "A".

Supplement "A" contains a list of all valves and their required position which actuate on a Safety Injection actuation. Using Supplement "A" and other available indications thu Operator can verify proper valve alignment independent from the Safeguards Status Panel.

OMM-022 Rev. O Page 10 of 25

~

PAIEIAL IDSS T diOMVRULM

~

(qFEfL5uc6 p(C Yell 2.0 ADIM ATIC ACTIONS (continued)

(....

~

2.4 cmdanser Cleaning System screens will go into backwash at 36 inches E 0 2

DP.

0 2.5 Generator lockout will occur if 5'wharat Hood temperature is above 225 F for five minutes.

3.0 GPERATOR ACTIONS 3.1 m aadiate Actions 3.1.1 STAIE standby Circulating Purg.

3.1.2 VERIFY' standby Vacuum Pung is running.

3.1.3 VERIFY Condenser Vacuum Breaker Valves are closed.

3.1.4 R Condenser vacuum is approaching the low vacuum trip point (20 2 in. Eg),

M.RIDOCE Turbine Generator load.

3.2 Subsequent Actiens 3.2.1 14 vacuum cannot be maintained above trip point M VERIFY Turbine trips LM) FOLIG PATE-1 if Reactor trips (above P-7).

3.2.2 E Condenser DP is high, M INITIATE backwash of the Condenser Cleaning System Screens MD MAINTAIN DP less than 36 inches.

3.2.3 R Icad capability is reduced, M NCTfIFY Icad Dispatcher.

3.2.4 VERIFY there is adequate circulating water flow through the Main Condensers.

3.2.5 E there is indicati:,n of air binding M VENT the Water Boxes.

ACP-012 Rev. O Page 4 of 6

ENG FoA y. / 2.

e y

6.2 ORCANIZATION h_.

Offsite 6.2.1

-The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

6.2.2 Definitions a) Personnel reporting to the'Cener.21 Hanager - Robinson Plant shall be identified in Section 6 of the Technical Specifications as the plant staff.

b) Personnel reporting to the Managar - Control and Administration

~

shall be. identified in Section 6 of the Technical Specifications as the C&A staff.

Facility Staff 6.2.3 The Robinson Fuclear Project organizatioa shall be as shown in Figure 6.2'-2 and:

a) The shift complement during hot operations shall consist of at least one Shift Foreman holding a Srnier Reactor Operator's License, one Senior Control Operato. holding a Senior Reactor Operator's License, two Control Operators each holding a Reactor

~

~'

Operator's License, two additional shift members, and one Shift Technical Advisor.

The limitations on the use of overtime applies to the HBR2 Shift Foremein, Senior-Control Operators, Control Operators, and Shif t Enginee rs.

These limitations apply j

only when HBR2 Reactor Coolant lystem is greater than 200 F or when fuel is being moved within the Reactor Pressure Vessel.

i These Limitations may be applied to other key " safety" personnel i

as warranted by the plant conditions and other circumstances at b

the discretion of the Plant Cenitral Manager.

'6.2-1 Amendment No. 95 I

Rev. 39 i

F ECT=ewce Po a. 4. I 3 1.0 SYMPTOMS l

1.1 Notification of a fire, AND 1.2 Inability to restore Plant control using.the Emergency Operating Procedures (EOPs).

2.0 AUTOMATIC ACTIONS 21 Due to the potential of fire-induced

damage, no automatic actions should be relied on to occur.

All required operations must be the result of positive operator actions.

3.0 IMMEDIATE ACTIONS 3.1 REPORT to the Shift Foreman in an approp'riate area and receive instructions.

4.0 SUBSEQUENT ACTIONS

(_

4.1 VERIFY the Fire Brigade has responded to control the fire emergency.

I DSP-001 Rev. O Page 4 of 34

.,---------,-,,,,,,,-,_,--.-.-.-.----,--,,--,-.-------,,,..,..,,.-..-r-.,

,_..--r.,

_.-9

T.

RCMce for gfj 6.0 GENERAL f

This procedure is used to safely bring the reactor plant to a

hot shutdown condition subsequent to a severe fire.

The proc'edure will only be used if the extent of the fire induced damage precl'udes the use of the Emergency operating Procedure,s Network to safely control the plant.

The procedure utilizes the Dedication Shu*.down(DS)

System components and manual (local) ope. rations to achieve a safe hot shutdown condition.

This procedure presupposes a loss of off-site power as a lir_iting

(_

condition.

In the event that off-site power is not /

lost or is recovered during this procedure the components operated in this procedure have their normal power removed to prevent spurious operations.

i i

(

,s DSP-001 Rev. O Page 29 of 34

,r

T

$6F6t&UC6 Fdt.

$. OQ PAGE TITLE RE F R OC. N J.

SL)-ost 22_ op 4_2_

Reactor Coolant Syste=

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e

(,,.

2.0 INSTRUMENTATION (Continued) f a.

AT/Tavg Control The AT/Tavg control signals suppit AT deviation and Tavg deviation alar =s along with meters (TI's) on the RTGB for each temperature sensed.

Switches on the RIGH allows one of the loo'p signhls to be defeated if it is not operable, IhehighestTavgisselected by an auctioneering circuit and is used for feedvater isolation, steam dump, pressurizer level program, auto rod control and is recorded on TR-408.

The highest aT is selected by an auctioneering circuit and is used to calculate rod insertion limits.

The rod insertion limits are displayed on TR-409.

\\s.

b.

AT/Tavg Protection Protection Tavg is used for hign/ low Tavg alarms along with reters (TI's) on the RTGB for each temp-erature sensed.

It also supplies a signal in cal-culating Over-temperature AT (OI AT) and overpower AT (OP AT) setpoints.

Through n two out of three (2/3) matrix it supplies a signal to steam break protection and the steam dump interlock.

Protection AT is used for high AT alarm.

The loop AT is compared to its OP aT and OT aT setpoints for a reactor trip function.

A switch on the RTGB allows the operator to sele:t one of the AT's to be recorded on TR 412.

pb

Y QEfeupcs M

G P 2-2.0 COMPONENT DESCRIPTION (Continued) 2.21 Discharge Pulsation Dampeners The discharge pulsation dampeners are spherical type vessels installed in each of the charging pump discharge lines. Their internal baffles reduce discharge pressure pulsations.

3.0 IFSTRUMENTATION AND C'NTROLS O

~

All valves discussed in this section are operated from the RTGB unless otherwise specified. The failed position for the air operated valves can be found in the system drawing. All instrumentation that provides local indication only can be found in the system drawing.

31 Valves 3.1.1 Letdown Stop Valves (LCV-460A and 4603)

,{'

Both valves are controlled by one three position switch. The switch positions are: OPEN, CLOSE and AUTOMATIC. These valves are closed automatically on a pressurizer low level alarm and will itutomatically re-open when the low level alarm clears. These valves are located in "A" reactor coolant pump bay.

3.1.2 Letdown Orifice Isolation valves (CVC-200A, 2003, 200C)

Ihree air operated valves are provided to determine which letdown orifices are in service. One orifice will pass 45 gpm and the other two will pass 60 gpm each when the RCS is at nor=al pressure and letdown pressure is adjusted to approxi=acely 300 psig. Care should be taken not to exceed design flow race through the deminerali:ers.

These valves are located next to the letdown orifices in the Regenerative Heat Exchanger cubicle. These valves will close on a Phase "A" Containment Isolation Signal (T signal).

I l

t i

SD-021 Rev. 4 Page 17 of 33

hP6Wc6

% d* O C

(

5.0 OPERATION (Continued)

~~

Turbine lond and nuclear power are compared in a rate comparator (impulse-lag unit) with the resultant deviation modified in accordance with the rate of change of the deviation.

The impulse-lag unit is designed to speed up system response to a power. mismatch.

Since the Tavg channel provides fine control during steady-state operation, the power mismatch channel should not produce a steady-state error signal.

This is provided for by the derivative action in the impulse-lag unit, which causes 'the output of this unit to go to zero during steady-state operation although the nuclear power and turbine load signals may not match exactly.

A non-linetr gain unit, placed at the output of the impulse-lag unit, converts the power mismatch signal to a temperature error in addition to varying the effect of this channel, with larger load changes having a larger effect.

The low and high gains are 0.3*F and 1.5*F per percent of power mismatch, respectively.

The low gain is uaed when the percent of power mismatch is one (1) percent or less.

The high gain is used above one (1) percent.

Increasing the gain for high mismatches further increases the initial output of the channel

~

thereby initiating rod motion more quickly.

Since reactivity changes at low power levels have a smaller effect on the rate of change of thermal power levels than reactivity changes at high power levels, a variable-gain unit is provided at the output of the non-linear gain unit.

This variable gain unit imposes a high gain of 2.0 for a, turbine power level of 0-50%.

The gain then decreases inversely proportional to a low of 1.0 at a turbine power level of 100%.

This variable gain enables the mismatch channel to provide adequate control at low power levels, as well as stable operation at high power levels.

SD-007 Rev. 1 Page 23 of 30 i

7

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SD-007 Rev. 1 Page 30 of 30

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5.0 OPERATION (Continded)

NOTE Pref is normal steady state pressure.

The error signal must be in excess of !100 psi (the control band of the pressurizer pressure control system) for the dead band unit to

_ j have on output. The output is in the form of a ramp function, incretsing with increasing deviations.

The output of this channel will'be added to (subtracted from) the output of the other two channels, Tavg deviation and power mismatch, in the summing unit.

5.2.4 Summing Unit The summing unit adds the inputs from the two channels,' (average temperatore and power mismatch) and produces a total temperature, error sy,nal which in turn feeds the red speed program.

C' 5.2.5 Rod Direction /Soeed Program The rod speed program converts the temperature error to rod motion withis the following constraints:

5.2.5.1 Deadband is the amount of input signal needed before any automatic rod motion will occur.

In this case, Tavg must be at least 0.5*F abo've Tref before the automatic control system will initiate control rod inward motion. The deadband for outward rod motion is -2.5*F.

In other words, Tavg must be at least 2.5'F less than Tref before the system will call for outward rod motion.

Lockup is the amount of signal needed, after automatic rod motion has commenced. before the automatic signal is cleared.

The setpoint for i

lockup is 0.5*F of the deadband setpoint.

'4 hen this point is l

reached, rod motion will cease. Therefore, when rod outward moeion has commenced due to low Tavg, it will cease when Tavg is 2.0*F'6elow Tref. Rod inward motion will cease when Tavg equals Tref.

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I SD-007 Rev. 1 Page 25 of 30 L.

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5.0 OPERATION (Continued) 1 5.2.5.2 Minimum Rod Speed The minimum rod speed, fixed by the speed controller is 8 steps / minute (5 inches / minute) and is used for an error signal of up to 13*F.

The minim.? rod speed used is based upon stability considerations.

5.2.5.3 Maximum Rod Speed The maximum rod speed, fixed by the speed controller, is 72 steps / minute (45 inches / minute).

The maximum roa speed used is based upon stability considerations and rod drive limitations, with the latter providing the ultimate restraint in this case. Maximum rod speed will be called for when the error signal reaches tS*F.

This error may not be visible on the Tavg-Tref instrunencation because the error signal also receives ar. input from the' power mismatch e.ircuit.

5.2.5.4 Proportional Rod Speed The speed gain in the proportional region is 32 steps / minute /*F.

Therefore the speed increases from 8 to 72 steps per minute for an error signal increase of 3-5'F.

This gain is selectad to be consistent with the need for rapid rod motion to limit transient overshoot while at the same time preventing overi:ompensation and oscillation.

'5.3 Dropped Rod Retrieval i

A dropped control rod from any bank'can be repositioned using the rod control system. This is accomplished by opening the lift coil disconnect switches for all rods in the affected bank except for the dropped rod. Then, using the individual bank controi, the rod can be returned to the correct position. Retrieval of.t drcpped control rod should always be done in accordance with the appropriate Abnormal Operating Procedure.

SD-007 Rev. L Page 26 of 30

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3.0 INSTRUMENTATION AND CONTROLS (Continued)

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t 3.1.8.1 Load Reference Reduction The Load Reference Reduction is caused when any of the following occur:

a.

Rod Drop Signal (NIS) - when 1 out of 4 Power Range Channels in Train "A" senses a 5% change in 5 seconds. This causes the Turbine to reduce load at the rate of 200%/ minute for a duration of 9 seconds (this equals to a 30% Load Reduction).

The Power Range Channel that caused the runback is reset or bypassed at it's respective Power Range A Drawer and has to be reset or bypassed before another runback can occur.

b.

Overtemperature AT - when 2 out of 3 Reactor Coolant Loop AT's (T - T ) exceeds the calculated setpoint. This causes a h

cyclic reduction in Turbine Load at the rate of 200%/ minute.

The cyclic duration is that of a 1.5 second runback (which ecuates to 5% Load Reduction) and a 30 second wait, if the Overtemperature AT condition still exists, the 1.5 second runback and the 30 second wait repeats. This cyclic reduction continues until the Overtemperature AT condition is corrected.

The setpoint is identical to the Reactor Trip Setpoint except the K term becomes 1.1265 (Refer to 3.1.5.5).

1 c.

Overpower 6T - when 2 out of 3 Reactor Coolant Loop AT's (T -

3 l

T,) exceeds the calculated setpoint. This causes a cyclic reduction in Turbine Load at the race of 200%/ minute. The cyclic duration is that of a 1.5 second runback (which equates l

to 5% Load Reduction) and a 30 second wait, if the -Overpower AT l

condition still exists, the 1.5 second runback and the 30 second wait repeats. This cyclic reduction continues until the Overpower AT condition is corrected.

i The Setpoint is identical to 'the Reactor Trip Setpoint except the K term becomes 1.04 (Refer to 3.1.5.6).

4 SD-011 Rev. 2 Page 27 of 35 l

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6.06 3.0 INSTRUMENTATION AND CONTROLS (Continued) 3.1.8.2 Load Limit Reduction The Load Limit Reduction is caused when 1 out of 4 Power Range Channels (NIS) in Train "B" senses a 5% change in 5 seconds or any Rod Bottom Signal (20 Steps) from the Rod Position Indication (RPI)

Circuitry is r.eceived and both Turbine First Stage Pressure Elements are above the pressure which correlates to 70% Turbine Load. This causes the Turbine to reduce load at the rate of 200%/ minute until at least one of the two Turbine First Stage. Pressure Elements is below the pressure which correlates to 70% Turbine Load.

Should the initiating NIS and/or RPI signal be reset or bypassed before the 70% Turbine Load Pressure is reached, the runback will stop.

There is no Load Limit Reduction if Turbine First Stage Pressure is less than 70% Turbine Load.

3.2 Annunciators and Alarms 3.2.1 First Out Reactor Trips Annunciator All Reactor Trip Signals have a First Out Annunciator. Other alarms on the RTGB si.L1 a.'so alarm from the same bistables and are covered in their respect:.ve System Description. When a Reactor Trip Signal -

is received, the appropriate First Out Annunciator will come in Blinking. Any other trip-signal that occurs will come in Solid (not Blinking). The operator should note which alarm (3 linking) caused the trip befere capressing the Reset Pushbutton. Any light that is solid will ruset (Go Out) when the Actuation Signal is cleared.

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SD-Oll Rev. 2 Page 28 of 35 l

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PAGE TITLE REV.

PROC.NO.

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PLS-1 RF. ACTOR CONTROL AND PROTECTION SYSTEM 26 PLS-1.

l'2.0 Setpoints for Uprating (2300 MWT)

(Continued) 12.1.2.5 Low Pressurizer Pressure (PC-455C, PC-456C, PC-457C)

(PM-455A, PM-456A, PM-457A) trip setpoint 1844 psig lead time constant 10 see lag-time constant 1 see 12.1.2.6 Loss of Primary Coolant Flow (FC-414, FC-415, FC-416)

(FC-424, FC-425, FC-426)

(FC-434, FC-435, FC-436) low flow (flow channel is calibrated a 91" full load flow and temperature such that 100% = nor=al flow) low frequency 58.2 cps low voltage 75% of nor=al undervoltage time delay

.3 seconds 12.1.2.7 Loss of Feedwater 1.

Low-low steam generator water level (LC-474A, LC-475A, LC-47eA)

(LC-484A, LC-485A, LC-486A) 15% of span (LC-494A, LC-495A, LC-496A) 2.

Coincident low level and steam /feedwater flow miscacch.

l low level 20% of span (LC-474B, LC-4753)

(LC-484B, LC-485B) 1 (LC-494B, LC-495B) l 6

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(FC-488A, FC-4883)

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i, 2.0 CCMPONENT DESCRIPTION (Continued) 2.21 Discharge Pulsation Dampeners The discharge pulsation dampeners are spherical type vessels installed in each of the charging pump discharge lines.

Their internal baffles reduce discharge pressure pulsations.

3.0 INSTRUMENTATION AND CONTROLS All valves discussed in this section are operated from the RTG3 unless otherwise specified. The failed position for the air operated valves can be found in the. system drawing. All instrumentation that provides local indication only can be found in the system drawing.

3.1 Valves 3.1.1 Letdown Stop Valves (LCV-460A and 460B)

Both valves are controlled by one three position switch.

The switch positions are: OPEN, CLOSE and AUTOMATIC.

These valves are closed automatically on a pressurizer low level alarm and will automatically re-open when the low level alarm clears. These valves are l'ocated in "A" reactor coolant pump bay.

3.1.2 Letdown orifice Isolation valves (CVC-200A, 200B, 200C)

Three air operated valves are provided to determine which letdown orifices are in service. One orifice will pass 45 gpm and the other two will pass 60 gpm each when the RCS is at nor=al pressure and letdown pressure is adjusted to approximately 300 psig. Care should be taken not to exceed design flow rate through the demineralizers.

These valves are located next to the letdown orifices in the Regenerative Heat Exchanger cubicle. These valves will close on a Phase "A" Contain=ent Isolation Signal (T signal).

I SD-021 Rev. 4 Page l'7 of 53 A

APP-001-17 f

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RCP THER BAR COOL WTR HI FLOW AUTOMATIC ACTIONS 1.

Therm Barrier Outlet Isol, FCV-626 CLOSED CAUSE 1.

Reactor Coolant leakage to Component. Cooling 2.

CCW Flow to Thermal Barrier excessive OBSERVATIONS 1.

Reactor Coolant Pump Tempeeatures (TRA-7, TRA-15, and TRA-23) 2.

FI-630, FI-633, and FI-636 (Local in CV) 3.

Check flow meter FIC-626 in Pipe Alley.

ACTIONS 1.

Refer to AOP-018, Reactor Coolant Pump Abnormal Conditions.

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DEVICE /SETPOINTS L

1.

FIC-626/100gpm POSSIBLE PLANT EFFECTS 1.

Loss of RCP(s) 2.

Power Reduction 3.

Plant Shutdown REFERENCES 1.

AOP-018, Reactor Coolant Pump Abnormal Conditions 2.

PATH-1, E0P Network 3.

Tech. Specs. 3.1.1

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.Rev. 3 Page 22 of 56

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1.0 SYMPTOMS 1.1 Notification of a fire, AND 1.2 Inability to restore Plant control using the Emergency Operating Procedures (EOPs).

2.0 AUTOMATIC ACTIONS 2.1 Due' to,the potential of fire-induced

dauage, no automatic actions should be relied, on to occur.

All required operations must be the result of positive operator actions.

3.0 IMMEDIATE ACTIONS 3.1 REPORT to,the Shift Foreman in an appropriate area and receive instructions.

4.0 SUBSEQUENT ACTIONS

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4.1 VERIFY the Fire Brigade has responded to control the f, ire emergency.

P DSP-001 Rev. 0

.Page 4 of 34 l

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6.0 GENERAL l

This procedure is used to safely bring the reactor plant to a

hot shutdown condition subsequent to a severe fire.

The precedure will only be used if the extent of the fire induced damage precludes the use of the Emergency Operating Procedures Network to safely control the plant.

The procedure utilizes the Dedication Shutdown (DS)

System components and manual (local) operations to achieve a safe hot shutdown condition.

This procedure presupposes a loss of off-site power as a limiting

(

condition.

In the event, that off-site power is not lost or is recovered during this procedure the components operated in this procedure have their normal power removed to prevent spurious operations.

L DSP-001 Rev. O Page 29 of 34 l

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5.0 CENERAL (Continued)

Operations retains the option of refusing to issue a clearance to any individual that they believe does not fully understand the importance and responsibilities of holding a clearance.

5.2 Ceneral Precautions 5.2.1 Anyone at anytime coming to or leaving a job on which a clearance is in effect must report t'o the man holding the~ clearance.

5.2.2 Do not issue a clearance to a person who is not capable of performing the work to be done or one who is not fully capable of handling the clearance sr.fely.

Equipment will not be operated for any purpose while the equipment 5.2.3 is under clearance. No breaker is to be closed or valve turned if there is a MEN AT WORK ree tag on the equipment.

5.2.4 The person holding the clearaace will always cancel the clea-ance

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if he is available.

If the work is completed on the equipment and the person holding.the clearance is not available, his immediate supervisor may cancel the clearance.

5.2.5 Two or more persons may hold separate clearances on the same equipment with a separate set of tags issued for each clearance.

However, it is important e. hat only one set of MEN AT WORK red tags be removed fcr each clearance cancelled.

6.0 PROCEDURE 6.1 Station and Line Clearances 6.1.1 Station Clearances 6.1.1.1 Station Clearances are issued when it is possible for the person taking the clearance to persorally check all feeds and determine that they have been propeely opened, cleared and tagged.

l OMM-005 Rev. 5 Page 8 of 27

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10.0 PROCEDURE (Continued) l l

10.4.10 Submit the Personnel Exposure Investigatiori Form to the Plant Dosimetry Supervision or R&CS Dosimetry Supervision for approval.-

10.4.11 Submit the Personnel Exposure Investigation Form to the Radiation Control Supervision at the plant or the R&CS Dosimetry Supervision for review.

10.4.11 Forward the original Personnel Exposure Investigation Form to the R&CS Dosimetry Supervision, and place a copy in,the individual's exposure history file.

10.4.13 Record TLD results on the Dose Record Fcem, if any, for the exposure period.

10.4.14 If the TLD badge has been read, end the exposure period on RIMS using Change Employee Status, Option 6, End Exposure Period.

10.4.15 Update the exposure period record for the individual using Periodic Dose and TLD Reader, Option 1, Update Period Record, in accordanca with DP-006, " Updating Dose Records," or equivalent plant procedure.

10.4.15.1 Update the field for investigation reason with the appropriate code.

10.5 Possible Exposures in Excess of Regulatory Limits / Administrative Limits 10.5.1 Advise the RC Foreman or higher line management, as soon as possible, whenever it appears that an individual's dose may have exceeded a regulatory limit or 5 REM in a calendar year.

10.5.1.1 The RC Foreman shall ensure the Radiation Control Supervisor and the Manager - Environmental & Radiation Control are appraised of I

the situation as soon as possible.

10.5.2 Immediately, notify the USNRC pursuant to 10CFR20.403, Nuclear Operations Department Procedure No. 7.13, and Plar.t Emergency Procedure PEP-2.1, if any individual received or may have received 25 rems or more to the whole body; 150 rems or more to the skin of the whole body; or 375 rems or more to the extremities.

DP-004 Rev. 5 Page 10 of 15

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