ML20136J522

Comments on Reactor & Senior Reactor Operator Exams on 850910 to Clarify Candidate Responses

ML20136J522
Person / Time
Site:	Nine Mile Point
Issue date:	10/04/1985
From:	Lempges T NIAGARA MOHAWK POWER CORP.
To:	Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20136J510	List: IR 05000220/1985014 ML20136J505 ML20136J522
References
NMP-13742, NUDOCS 8601130479
Download: ML20136J522 (16)

Text

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NMP-13742 NIAGARA MOHAWK POWER CORPORATION /300 ERIE BOULEVARD WEST, SYRACUSE. N.Y.13202/ TELEPHONE (315) 4741511 October 4,1985 Mr. Robert Keller United States Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406

Dear Mr. Keller:

1.

This letter addresses comments and recommendations from the Niagara Mohawk Power Corporation Training Department regarding the RO/SRO NRC exam,ination conducted September 10, 1985.

2.

The below listed comments are divided int. two sections, the first section addresses the RO exam, and the second section addresses the SRO exam.

These comments are being provided to help clarify the responses candidates may have provided to the questions asked.

RO EXAMINATION COMMENTS SECTION 1 1.12 There may be confusion 20 te whether the question asks how the coefficient affects reactivity or how the coeffi-cient-itself is aft'ected.

Alternate answer could reGd:

1.

Moderator / Voids

-adds negative reactivity 2.

Voids

-adds negative reactivity 3.

Voids

-adds positive reactivity 4.

Doppler

-adds negative reactivity SECTION 2 2.03 Second choice in the answer key should read:

"Lo-Lo reactor water level" vice "Lo-Lo-Lo reactor water level".

(Simulator System Manual, Chapters 11 & 12) 8601130479 860103 PDR ADOCK 05000220 g

PDR

'k NMP-13742 Page 2 SECTION 2 (CONTINUED) 2. 0'5 a Does the'canhidate need to list the

. components of ADS with their power supplies, or just list the power

- supplies?. Question' only asked for power supplies.

2.09 For clarity, the answer key should read as=follows:

" pressure' difference between the two loops and the core bottom" (Simulator Systems Manual, Chapter.17) 2,.12 Part B-is worth (.5) on test, but is given a total of (.75) points on the

- answer key.

r, SECTION 3-3.05a' Candi'date'may list m urces vice Voltages:

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120-Volts AC is regulated from PB 167 24 Volt DC from 24 volt batteries.

3.' 0 9 '

Does the< candidate-need to include setpoints?

This was not specifically asked for on the test.

Some-questions vividly flag this.re-g, quirement.

3. l lb-In OP-48,,the MG set startup procedure, Section D.4, the % scram is only necessary-if you can't parallel, ie. phase angle

>l5* on synch scope, otherwise no inter-ruption power is observed.

This is also reflected in_Section G.4 when removing MG set 131/141 from service.

3.12e The description of the valves in part (e) is unclear.

NMPC Training refers to these air-operated valves as " Containment Spray c"

Discharge' Valves" which fail open upon loss of air.

We do not think the candidate will identify the valves by the description written on the exam.

The Training Staff verified the answer on the key after consulting system prints:

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NMP-13742 Page 3 i

-SECTION 4 4.03b Candidates should realize that the vessel must be drained and MSIV's opened by the time we reach the " heating range".

4.07b-The bottom portion.of the answer key is not.part of~the answer.

Suggest removing this un-needed section from the answer key.

SRO EXAMINATION COMMENTS SECTION 5 5.05b The answer key addresses this transient solely on feedwater heating effects.

We

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believe that correctly stated answers based upon fuel temperature changes during the transient should also be considered.

5.07a All of'the candidates believe that the examiner's clarification solicited the reactivity curve as the answer.

We be-lieve that an appropriate discussion of the curve should constitute an acceptable answer.

An examination of the attached NMP.1 APHLGR curves will reveal that some of the curves (3.17b) only marginally follow the verbal description of the question.

Consequently, this information is.not taught in the NMP 1 license class.nor is it in-cluded in the reference submitted in the answer key.

-SECTION 6 6.01a2 The MSL flow restrictors do not generate a containment isolation.

They do generate an automatic MSIV closure and also serve as input to the FWC system.

6.07a Request that " Power available on power board 102/103"-be accepted in lieu of " Power is available to the Core Spray pumps and ADS logic"..since the ADS actuating circuit and Core Spray pumps do in fact have the same power ' supply.

This is consistent with the answer key reference. a

+-

NMP-13742 Page 4 e

SECTION 6 (CONTINUED) 6.09b Recommend that " Rod block alarm"'be accepted as well as " Rod block with-drawal alarm".

6.10

'The sum of water available to the Emer-gency Cooling System from the Emergency Condenser Make-up tanks and the Condensate Storage tanks will provide 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

SECTION 7 7.06 From the wording of this question, the candidates interpreted the question as asking for the subsequent operator actions if the entry condition (s) still existed.

They provided the five (5) conditions required for injecting SBLCS (liquid poison), thinking that plant conditions where such that one (1) of the five (5) conditions have already been met, and now, what action must be taken.

7.09 This question was interpreted to mean after the Recirculation pumps had al-ready been secured.

For clarification purposes, the setpoint is 20% from full open(or 80% open) in the plant".

SECTION 8 8.06b The question submitted the following data:

Identified Unidentified Leakage Leakage Shift 1 11,700 Gal 2,100 Gal Shift 2 12,000 Gal 2,800 Gal Shift 3 12,200 Gal 2,200 Cal The answer key states that unidentified leakage technical specifications were not violated, since 5 gal / min X

60 min /hr X 24 hr/ day 7,200 Gal,

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and unidentified leakage total lor the day was 7,100 gallons.

However, the candidates properly interpreted t h'i s as a technical specification violation, since 2,800 Gal on Shift 2 translates to:

2800 gal X

1 hr 5.83 gal / min for that 8 hr 60 min 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift s

_4_

NMP-13742 Page 5 e

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SECTION 8 (CONTINUED) 8.09 Because of the nature of the question, the candidate required the logic elementary for IRM's/APRM's to adequately analyze the ques-tion to arrive at the right answer.

3.

Also, enclosed.are copies of documents to support comments for the ~SRO section of the examination.

Very truly yours, j?"'L-9 T.

Lempges Vice President-Nuclear Generation

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AVERAGEPt.ANAREXPOSURE(Gis/ST)

Figure 3.1.7c Maxinus Allowable Average Planar LHGR Appilcable to 80ff8277 j

i Fuel as descrjbed in Reference 8.

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1.

Main Steam Line Flow Instrumentation The steam flow in each main steam line is monitored by flow sensors mounted on each main steam line flow restrictor. The differential pressure developed by the flow of steam through the venturi-type flow restrictor is utilized by the flow sensors to produce electrical signals for both control and indication purposes. Figure 21-6 shows a simplified drawing of the flow instrumentation for one of the main steam lines. The other main steam line has identical instrumentation.

The main steam line flow restrictor ha's two pressure sensing penetrations:

one high pressure tap and one low pressure tap. The pressure sensing lines are directed through the p'rimary containment to one flow transmitter, for indication, and four differential pressure switches. An excess flow check valve is mounted in each sensing line to limit the flow of reactor coolant if a break develops in the line. The excess flow check valves are designed to offer a high resistance to flow, but a low resistance to pressure changes, thereby maintaining a sensitive instrumentation response to varying steam flows.

QMQ The four differential pressure switches for each steam line provide steam j

flow Inputs to the reactor protection system which will Initiate automatic closing of the MSIVs if flow exceeds 120% of rated.

[.0 73,gio,,,,,,miet,,4,,,iop,,,,i c,1c,i,ign ipropo,, ion,itoin,,,,,,a

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Ng differential pressure and sends'it to a square root extraction circuit. The l

output from the square root extractor is proportional to the main steam line flow and provides steam flow indication at F panel in the control room. The steam flow signal is also directed to the reactor feedwater control syste'm j

here it is utilized in ghe automatic control of reactor vessel water 1,evel.

2.

Main Steam Line Radiation Detectors Each steam line is equipped with two gamma sensitive ion chambers which provide indication of a gross cladding failure. The output of these detectors provide signals to the reactor protection system which will initiate MSIV closure if radiation levels exceed 5 times normal full power background level. These monitors actuate an annunciator alarm at the high level trip setpoint indicating high radiation. A low level trip set point of 1 mr/hr will also sound the same alarm indicating system failure.

3.

MSIV Closure Trip Each MSIV has a limit switch which will actuate if that valve is more than 10% travel from its fully opened position (less than 90% opened). These limit switches are connected to the RPS which will initiate a reactor scram. The purpose of this scram is to reduce the effects of a power excursion that would occur as a result of the pressure transient accompanying a MSIV closure.

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CONTROL FUNCTIONS AND INTERLOCKS The control functions of the automatic depressurization system that will be

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discussed in this section are:

ADS Actuation Contro!

Individual Relief Valve Control 1.

ADS Actuation Control ADS will actuate upon coincident receipt of high drywell pressure and

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reactor vessel low-low-low level provided power is available to operate the core spray pumps.

Unlike most system actuating circuits, actuation is achieved by energizing a relay instead of de-energizing. The actuating l

circuit receives 120 VAC from the same powerboard that supplies the core t

i spray pumps (powerboard 102 and 103), thus ADS will not auto-initiate if power is not available.

Figure 15-3 shows a simplified drawing of the automatic initiation logic circuits of channel 11 with channel 12 data shown Qg inside parentheses. Each channel has two sensors which close contacts upon 1,,.

V a high drywell pressure and two which close contact upon low-low-low p

reactor vessel level. As soon as pressure and level sensor contacts are closed, power is applied to relay K207 and the time delay contact or (TDC).

Relay K207 will energize and seal itself in. The TDC wi!! energize but will (o7A

== cl== its catact uatii afto 120 ecods ho posed. if during the 120

' second timing period power to the TDC is interrupted the timer will

~l restart. Once the timer has completed its 120 second cycle it will close its

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contacts. One set of the TDC contact seals itself in around the sensor A

contacts and will continue to apply an initiation signal regardless of the input sensor, thus the blow dgwn will continue until completed. Although the TDC seals in around the sensors, it is not sealed in around the reset switch; the operator may stop the blow down by depressing both reset buttons on F panel.

2.

Individual Relief Valve Control The lower half of the figure 15-3 is a simplified drawing of the individual relief valve control circuit. The valve is opened by energizing the control, solenold. In series with the contro! solenoid is a resistor which is bypassed by a cutout relay. As energizing power is first applied to the solenoid, the cutout contacts are closed thus bypassing the resistor and allowing full current to flow through the solenoid. Once the solenoid actuates the cutout relay contacts open which places the resistor in the current path thereby reducing the flow through the control solenoid. This configuration reduces the chances of solenoid burnout.

Three conditions will energize the control solenoid thereby opening the valve. First, placing the control switch in the open position closes the CS-O contact which applies power directly to the control solenoid. Placing the control switch in the AUTO position opens the CS-O contacts and closes the CS-A contacts.

In the auto mode the valves may be opened by a high pressure in the reactor vessel. Contacts 63-1 are closed at approximately 25 psig below the relief valve setpoint. As reactor pressure reaches the upper

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limit of the operating band (1090,1095 and 1100 prig), the 63-2 contacts i 5-5 e

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The pressurized coolant is condensed in the tube bundle of the condenser by boiling the shell water at a lower pressure. The steam from the shell water is vented to the atmosphere.

Radiation monitors are placed on the vent to detect any tube bundle leak.

The level control valves allow make-up water to drain from the elevated 40,000 gallon make-up storage tanks to the condensers to maintain the 6 inch level tolerance.

The condensers combined with the respective make-up tanks can provide continuous. cooling for eight hours.

Normally, water will be supplied to hf@y the tanks through the condensate transfer system from the two 200.000 y

gallon condensate storage tanks.

Thus, approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of fcontinuouscoolingispossible. The electric and diesel driven fire pumps f are also available to supply the make-up tanks and thus the condenser l

shells, k

vents are provided at the high points in the steam lines to purge non-condensible gases from the system. One vent is a continuous vent that connects to the main steam line beyond the outside isolation valve. The other vent is a motor-operated system that directs flow to the torus. Its purpose is to enable venting any non-condensible gases from the reactor coolant system during post accident vessel and/or containment isolation conditions.

Drains have been provided on the steam lines to drain off condensate formed during standby condition.

Isolation valves are provided on the drains and vents.

A flow deteccor is located on each of the steam lines at an elbow inside the containment.

The detector initiates closure of isolation valves in the event of a line break resulting in approximately 300 percent of. rated flow. This is equivalent to rupture of 30 percent of the cross-sectional area of the pipe.

Local temperature detectors are provided to detect I

small leaks and provide alarm funct, tons.

1.

Tables Table 14-1 includes Emergency Cooling System ratings and

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normal system parameters.

TABLE 14-1 BCS RATINGS AND NORMAL SYSTEM PARAMETERS P_ARAMTTRR RATING l

Design Heat Removal 1.9 x 108 Btu /hr Capacity (1 Loop)

Maximum Cooling 48 Hours Time Available I

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b 3.0 AUTOMATIC STATION RESPONSE

1). Core Rod Display " Fails" to indicate " Full In" for all control s

rods.

2)

Alarm computer and annunciators identify cause of Rx Scram.

3)

Rx power fails to decrease, or drops, but not as in a cormal scram condition as verified by incore nuclear instrumentation.

4.0 IMMEDIATE OPERA'IVR ACTIONS a), Monitor the reactor neutron flux and indication of local areas of high reactivity by observing full core display and LPRM indicating meters.

b)

Place the reactor mode switch in shut-down.(This action will generate an additional Rx scram signal).

c)

Follow procedure No. N1-SOP-16. "Rx Scram" in parallel with this

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procedure.

d)

If procedure symptoms still exist, trip recirculation pumps. g Not Isolate Pumps.

NOTE:

Reactor high pressure (1135 + 10 psig) or Io-lo level (5" indicated) should cause an ATWS automatic recire

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pump trip.

.p e)

Fully insert remaining control rods using " Emergency Rod In**

switch. Select rods at position 48 by column, but alternate rows such that two (2) adjacent rods are not inserted one after another.-

f)

If procedure symptoms still exist, reset the Reactor Protection System (RPS) trip. Manually scram the reactor and if rod motion is observed, alternately reset the RPS trip and manually scram until rods are fully inserted.

g)

If procedure symptoms still exist, individually scram rods from "M" panel.

h)

If procedure symptoms still exist, isolate and vent scram air header locally.

NOTE:

Scram discharge volume vent and drain valves cannot be opened if air header is isolated.

5.0 SUBSEQUENT OPERATUR ACTIONS gg a)

If conditions are 4till abnormal and neutron instrumentation indicates increasing reactor power. initiate Stand-by Liquid Control System per Operating Procedure N1-OP-12.

g (Con $im) On N&

N1-SOP-32

-8 July 1982 o

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4, 5.0 SUBSEQUENT OPERATOR ACTIONS (Continued) 4 b)

IF i

2 or more adjacent control rods cannot be or inserted below 1

30 or more control rods

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AND Reactor water level cannot be maintained or Suppression pool water temperature cannot be maintained

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THEN

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~ Initiate Stand-by Liquid Control System per Operating Procedure N1-OP-12 c)

NOTE: Whenever the Standby tiquid Control System is initiated as above, maintain the system in operation until the entire tank volume is

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injected into the reactor.

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NOTE: Do not restart the recircul: tion pumps.

Liquid Poison must be injected within 5 tc 10 minutes following the initial transient.

d)

Initiate Containment Spray System in the torus cooling mode as required to maintain torus temperature.

1._

.e)

Keep all plant parameters under observation.

f)

Proceed to cold shutdown condition at a cooldown rate I

not to e>ceed 100*F/hr per Operating Procedure N1-OP-43.

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N1-SOP-32

-9 November 1980

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ATTACHMENT 3 NRC Resolutions of Comments on Written Examinations j

The following represents the NRC resolution to those comments made by the facility as a result of the current exam review policy.

1 Only those comments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not listed.

ie: typo errors, relative acceptable terms, minor set point changes.

1.12 Comment accepted.

Candidates will receive full credit for proper interpretation.

2.05a The candidates will not be required to list the components of ADS and their power supplies.

3.09 Comment accepted. The setpoints will not be required since they were not asked for in the question.

3.12.e Comment not accepted.

Proper reference material to substantiate the comment was not provided.

5.07.a Comment accepted.

Parts a. and b. will be deleted from the exam.

7.06 Comment not accepted.

The question was sufficiently phrased to elicit the response on the answer key.

8.06 Comment accepted.

Examiner was not aware of shift requirements.

8.09 Comment accepted. This question has been deleted from the exam.

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