ML18005A460

Submits Comments Re NRC Written Senior Reactor Operator & Reactor Operator Exams Received on 880425

ML18005A460
Person / Time
Site:	Harris
Issue date:	05/02/1988
From:	Watson R CAROLINA POWER & LIGHT CO.
To:	Bill Dean NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML18005A458	List: ML18005A457 ML18005A459 ML18005A460
References
CON-NRC-624 NUDOCS 8807050529
Download: ML18005A460 (72)

Text

CPS L CaroIina Power 6 Light! Company Harris Training Unit Post Office Box 165 New Hill, North Carolina 27562 May 2, 1988 Mr. William Dean US NRC - Region II 101 Marietta St. NW Atlanta, GA 30323

SUBJECT:

SRO/RO NRC Written Exam Comments NRC-624 Dear Mr. Dean'.

On April 25, 1988, Shearon Harris Nuclear Power Plant received NRC written SRO and RO examinations. The examination coianents are submitted by CP&L. Copies of reference material are included where indicated.

Should you need any explanations or additional reference material, please do not hesitate to contact the SHNPP Manager - Training, Mr. A. W. Powell, at (919)362-2618.

R. . Watson Vice President-Harris Nuclear Project HWS/raw Attachments cc'Dr. J. N. Grace (NRC)

Mr. T. Guilfoil (Sonalyst, Inc.)

Mr. G. F. Maxwell (NC-SHNPP) co@.

8807050529 880627 PDR ADOCK 05000400 V PNU

APRIL 25 1988 NRC EXAMS RO EXAM

GENERAL COMMENT

S

1. Several questions in Sections 2.0 and 3.0 required the memorization of factual information or lists not required for safe operation of plant.

Example are:

a. Conditions that activate amber lights on the LFDCP (Question 2.08)

b. Conditions that result in modulation of lIA-648 (Question 2.15)

c. Setpoints for RHR miniflow valves (Question 3.06)

d. Frequency of Rod Drive MG sets (Question 3.18)

2. Two questions in Section 4.0 (Question 4.01 and 4.11) required the operator to reproduce long lists of symptoms without allowing him to use the most likely to be observed (i.e., RMS response is disallowed as a symptom of a LOCA). A more operationally oriented approach would be to list several symptoms in the questions and require a diagnosis of the potential failures.

SRO EXAM

GENERAL COMMENT

S

1. The emphasis on Technical Specification use in Section 8.0 is commendable and reflects an emphasis on testing information important to plant operation. Additionally, questions in Section 7.0 were relatively clear and straight forward. No comments or recommendations are made for any questions in Section 7.0.

2. Some questions did not provide sufficient information or vere worded in such a confusing manner that the information could not be readily extracted to enable the examinee to provide the desired ansver. Examples of this were questions 5.04, 5.09,. 6.04 & 6.20.

3. Some questions vere not screened for applicability to SHNPP ~ Examples of these are questions 6.01, 6.02, 6.07 6 6.20.

4. More detailed comments are noted on the following pages.

There are two primary effects that cause differential boron worth (DBW) to change as the core ages.

a. List the TWO effects and their relative impact on DBW (increase or decrease).

b. State what the total resultant effect is on DBW over corelife (increase or decrease).

ANSWER 1.04

a. 1. Boron concentration decreases over core life which INCREASES DBW (or decreasing boron concentration decreases the amount of spectrum hardening which INCREASES DBW). (0.5)

2. Fission products build up decreases DBW (0.50)

b. INCREASES over core life. (0.5)

CP&L COMMENT: 1.04 The answer to part a states that the decrease of C over core life INCREASES DBW with spectrum hardening included parenthetical y as an alternate response. RT-LP-3.11, p. 11 (Attachment 1-1) and RT Theory Manual, p. 12-17 Competition also explains why FP buildup decreases DBW.

The Core Data Report (Attachment 1-3) expresses the reasons differently.

~De letion of the BPBA, causes DBW (the reciprocal of inverse Boron Worth) to decrease and is predominant between BOL and HOL. ~Burnu of fuel causes DBW to increase and is predominant between MOL and EOL.

The answer to part b states DBW INCREASES over core life. RT-LP-3.11, pp. 9-11 (Attachment 1-1) uses Exercise B to show the change is very small.

The point values for the subparts of the question are not specified.

RECOMMENDATION: 1,04 For part a, accept competition as alternate explanation for effects of the CB changes and the FP buildup. Also accept the two alternate primary effects mentioned by the Core Data Report, BPRA burnup and fuel burnup.

For part b, accept "no significant change" as an alternate answer.

Point values should be 1.00 pts. for part a - 0.50 pts ~ for each of the two responses. Within each response the effect should be worth 0.25 pts ~ and the impact worth 0.25 pts. Part b should be worth 0.50 pts.

A reactor has been shut down from 100 percent power and cooled down to 140 degrees F over 5 days. During the cooldown, boron concentration was increased by 100 ppm. Given the following absolute values of reactivity which ONE of the answers below would be the value of the shutdown margin?

Rods 6918 pcm Temperature = 500 pcm Boron 1040 pcm (100 ppm increase)

Power Defect 1575 pcm

a. minus 3803 pcm

b. minus 4803 pcm

c. minus 5883 pcm

d. minus 6883 pcm A <SWER 1.09

REFERENCE:

SHEARON HARRIS RT-LP"3.13 p. 7 Westinghouse, Reactor Core Control For Large Pressuriaed Water Reactorsg 1983'.

7-21 thru 7-23.

192002K113 3.5 ~ . ~ (KA'S)

CP6L COMMENT: 1.09 The question ignores the Tech Spec definition of SDM (Attachment 1-4), which states the most reactive rod is assumed to be stuck out. If this assumption is made, SDM is decreased by the worth of the most reactive rod (2050 pcm at EOL) to -3833 pcm. The worth of the most reactive rod is given in the Core Data Report, p. 6.6 (Attachment 1-5). This is approximately the same as answer a. Additionally, the question is somewhat confusing since it does not mention the effects of Xenon and Samarium.

RECOMMENDATION: 1 ~ 09 Accept either answer a or c

Cl What is the primary reason for arranging symmetrical control rods in groups?

ANSWER: 1.17 (1.00)

To prevent the formation of abnormally high flux peaks.

REFERENCE:

SHEARON HARRIS REACTOR THEORY MANUAL p. 13-31 Westinghouse, Reactor Core Control for Large PWRs, 1983, p. 6"28 192005K108 2.7 ... (KA'S)

CP&L COMMENT 1.17 The question is vague. It implies that the reason for the syametric arrangement be addressed in the response as well as the reason for the division into banks. The reference to symmetry might well illict a response concerning radial flux distribution or QPTR.

instead of symmetry or division into groups. It then states bank overlap provides "a more uniform differential control rod worth and more uniform radial neutron flux distribution. The alternative is potentially a perturbed flux distribution which could cause "high power peaks...resulting in fuel damage".

RECOMMENDATION: 1.17 Accept any one of four possible answers in addition to the one given by the key.')

More uniform differential control rod worth

2) More uniform radial flux distribution

3) Preveht unacceptable power peaks

4) Prevent fuel damage

I ~" s ~

What are the indications of a cavitating RCP?

ANSWER 1. 18 (1.00)

1. Erratic or low flow indication

2. Pump motor current fluctuating

3. Ercessive pump vibration

4. Abnormal noise (0.25 each)

REFERENCE:

SHEARON HARRIS FF"LP"3.2 p.. 15; FFM File 12.3 No. p. 3-33 Westinghouse, Thermal-Hydraulic P~inciples and Applications to the PWR, Vol. 2, 1982, p. 10-54.

193008K117 2.9 ... (KA'S)

CP&L COMMENT: 1.18 The question does not state the number of required responses. The fourth response, "abnormal noise", is not a directly observable indication in the control room. The references cited, FF-LP"3.2 (Attachment 1-7) and FF Manual (Attachment 1-8), mention noise as a general indication. It is not applicahle to the RCPs cited in the questions.

RECOMMENDATION: 1. 18 Delete the fourth response. Require two of the remaining three responses for full credit and adjust the point values appropriately.

If the control rods are NOT maintained above the rod insertion limits during routine reactor operations at power, which ONE of the following is most likely already outside specification limits?

a. Local Power Density (KW/ft)

b. Departure from Nucleate Boiling Ratio (DNBR)

c. Axial Flux Difference (AFD)

d. Quadrant Power Tilt Ratio (QPTR)

ANSWER 1.20 (1.00)

C ~

REFERENCE:

SHEARON HARRIS REACTOR THEORY MANUAL p. 13"31 Westinghouse, Reactor Core Control for Large PWRs, 1983, p. 6-32 3.4 ...(KA'S) '92005K115 CP&L COMMENT: 1.20 Two of the potential answers, a and c, are interrelated. This is expressed in the reference cited (Attachment 1-9). The Tech Spec basis for AFD (Attachment 1-10) cites FQ(Z) as the basis -for maintaining AFD within limits.

RECOMMENDATION: 1.20 Accept either answer a or c.

I (1.00)

The plant has experienced a loss-of-coolant accident (LOCA) with degraded safety injection flow. The reactor coolant pumps are manually tripped and the resulting phase separation causes the upper portion of the core to uncover.

(Core is only slightly uncovered). Which ONE of the following describes Excore Source Range (BF3) neutron level indication relative to indication just prior to partial core uncovery?

a. Significantly less than actual neutron level.

b. Significantly greater than actual neutron level.

c. Essentially unchanged.

d. Impossible to estimate with the given core conditions.

ANSWER 1.24 (1.00)

C~

REFERENCE:

SHEARON HARRIS MCD-LP-2.6 p. 7,8 Westinghouse, Mitigating Core Damage, 1984, p. 9,8 191002K117 3.3 ... (KA'S)

CP&L COMMENT: lo24 The question does not give any information concerning water level in that the downcomer. The reference cited, MCD-LP-2.6 (Attachment l-ll), states downcomer level is the "most significant effect" in SR response. The omission of the status of the downcomer level may cause answer d to be chosen, "Impossible to estimate with given core conditions."

RECOMMENDATION: 1.24 Accept either answer d or c

SELECT the one statement below that is correct if the Power Range instruments have been adjusted to 100X based on a calculated calorimetric.

a. If the feedwater temperature used in the calorimetric calculation vas HIGHER than actual feedwater temperature, actual power will be LESS than indicated power.

b. If the reactor coolant pump heat input used in the calorimetric calculation is OMITTED, actual power will be LESS than indicated power.

c. If the steam flov used in the calorimetric calculation vas LOWER than actual steam flow, actual power will be LESS than indicated power.

d. If the steam pressure used in the calorimetric calculation is LOWER than actual steam pressure, actual power will be LESS than indicated pover.

ANSWER 5 .03 & 1.26 (1.00) b.

REFERENCE:

NUS, Vol 4, pp 2.2-4 Surry 1-PT-35 SHNPP: HT-LP-3.2, L.Oo 1.1 ~ 5 GP-L'P-3+5 TS 3.3.1 OST 1004 2o6/3.1 3.1/3.4 01500K504 193007K108 ...(KA'S)

CP&L COMMENT: 5.03 & 1.26 The effect of steam pressure changes on a calorimetric vill depend on the pressure involved. At typical main steam pressures (950-1100 psia) a decrease in pressure leads to an increase in enthalpy (See Attachment 5-1). This increase causes hh across the steam generator to increase leading to an increase in calculated power. Thus ve adjust our indicated power so that it is now greater than actual pover. This means ansver d is also true (actual power less than indicated).

RECOMMENDATION: 5 .03 & 1.26 Accept either ansver b or d.

WHICH one of the following situations will the insertion of control rods cause Delta I to become MORE positive?

a. Burnout of Xenon in the top of the core with rods initially fully withdrawn.

b. Positive MTC during a reactor startup.

c. Band D control rods inserted toward the core midplane.

d. Excessively negative MTC at EOL.

ANSWER, 5.04 & 1.25 (1.00) a.

REFERENCE t SHNPP RT LP 3 ~ 14~ LoO 1 1o3~ 1 ~ 1 ~ 11 HBR RXTH-H0-1 Session (CAF) 3.2/3.5 192005K114 ~ ~ ~ (KA'S)

CP&L COMMENT: 5 o04 & 1.25 This question apparently addresses the issue of axial flux/xenon oscillations induced by control rod insertion. When control rod insertion takes place from 100X power with all rods initially withdrawn, for example, a flux/xenon oscillation is induced. (See Attachment 5-2 paragraphs 5 ' and 5.4). If started from or near the fully withdrawn condition, initially a rod insertion will cause hl to become more negative (See Attachment 5-3) ~ However, as the effects of the rod insertion progress, a flux/xenon oscillation begins and eventually hl becomes more positive than it was initially (See Attachment 5-4). Regardless of which situation proposed by choices a, b, c, or d is occurring AI will become more positive eventually. Since the wording of the question is not specific as to when the positive AI was to occur (immediately, or at any time in the future) any of the choices are correct given an insertion of control rods.

CP&L COMMENT: 5 .04 & 1.25 (Continued)

If the question was intended to imply an iaxnediate increase in hI then either choices a or c are plausible. At SHNPP we have a 4 rod D bank (See Attachment 5-5). For D bank rod insertion on SHNPP Cycle 1 from fully withdrawn down to about 160 steps 41 is controllable with rods (i.e. when rods are inserted, AI becomes more negative). When bank D is inserted past about 160 steps down to 113 steps when the 8 rod bank C beings insertion, the effect on AI is no longer so predictable. In some instances bank D rod insertion in this range has either had no effect or has caused hl to become more positive. The AI movement in the more positive direction is usually accompanied by hl moving more positive initially prior to rod movement. The insertion of D bank in this case sometimes serves to accelerate this more positive hI trend suggesting some combination of choices a and c is occurring. Attachment 5-2 paragraph 5.3.9 addresses the erratic behavior associated with relatively deep insertions of bank D, but does not take into account the observed effect noted above in the 160 to 113 step range.

RECOMMENDATION: 5.04 & le25 The preferred corrective action would be to delete this question due to its vague wording. A less desirable alternate corrective action should be to accept either answer a or c.

List the FOUR conditions that will trip the Emergency Diesel Generator during EMERGENCY operations, in addition to the Emergency Stop Pushbuttons.

ANSWER 2.07 (1.00)

(0.25 each)

1. . Engine Overspeed

2. DG differential (87 relay)

3. Emergency bus differential

4. Emergency voltage regulator shutdown pushbutton REFERENCE SHEARON HARRIS SD-155.01 p. 11 3 ' 06400K402 ~ (KA S)

CP&L COMMENT: 2.01 The fourth condition generates the same signal as the "loss of potential transformers" for the generator. Tech Spec Surveillance Requirement (Attachment 2-1) actually refers to the "Loss of generator potential transformer circuit" as an emergency trip.

COMMENT 2.0 Accept "Loss of potential transformers" as a correct response.

(2.00)

a. With the Residual Heat Removal System (RHRS) in a normal lineup and the reactor plant operating at 100X power describe how the following are isolated:

1. The RCS hot leg supply to the RHR pumps. (0.5)

2. The RHR pump discharge to the RCS cold legs. (0.5)

b. What is the design basis for the size (flow rate) of the relief valves

& 45) located between the isolation valves in the lines leading (1RH"?

from the RCS loops to. the suction of the RHR pumps'1.00)

ANSWER 2.07 (2.00)

a. 1. Two motor operated valves (0.25) in series. (0.25)

2. Two check valves (0 ~ 25) in series ~ (0.25)

b. Each relief valve is sized to pass the combined flow of three charging pumps/SI pumps (1.0) (operating against relief valve set pressure of 450 psig.)

REFERENCE:

SHEARON HARRIS SD-11 p. 4 and 7 3.6 005000K109 ...(KA'S)

CP&L COMMENT: 2.07 Part b.

1RH-7 & 45 are located after the isolation valves and not between the isolation valves as stated in the question. The locations are clarified in RHR-LP"3 .0, p. 11 (Attachment 2-2) .

RECOMMENDATION: 2.07 Part b.

Accept answers relating to either the reliefs between isolation valves OR those after isolation valves or delete part b of question 2.07 2485 + 75 psig 'ETWEEN:

< 1 gpm thermal expansion AFTER: 450 + 13.5 psig 900 gpm discharge of all CSIP's enthrottled with L/0 isolated

List FIVE conditions that activate amber lights at both the Local Fire Detection Control Panel (LFDCP) and the Main Fire Detection Information Center (MFDIC), as well as actuate an audible alarm distinct from the fire alarms (fire horn) ~

ANSWER 2.08 (1.50)

1. Loss of a detection circuit.

2~ Loss of an activation circuit.

3~ Loss of an alarm circuit.

4~ Water not flowing 5 seconds after deluge activated.

5 ~ Operation of water flow detection device.

6. Loss of supervisory air pressure.

7 ~ Operation of a Fire Protection System valve away from normal'Any 5 at 0.3 each)

REFERENCE:

SHEARON HARRIS SD-149 p. 18,19 SHEARON HARRIS L.O. 1.1.4 FP-LP-3.0 File No. 4.14 p. 4 086000K403 086000K604 ...(KA'S)

CPSL COMMENT: 2o08 Conditions that activate amber lights at fire detection panels and in the information center are beyond the scope of the Control Operator's position and not listed as part of the CO's responsibility in OMM&01 (Attachment 2-3).

This responsibility is primarily that of the "Shift Technical Aide Fire Protection", as addressed in FPP-001 (Attachment 2-4). This person is always present and is part of the shift compliment. It is necessary from time to time for Control Room personnel to ~caela information to the Shift Technical Aide Fire Protection, but at no time are Control Room personnel responsible for interpreting information at the various panels regarding conditions other than true fire alarms as addressed in FPP-002 (Attachment 2<<5) ~ If a condition other than a fire alarm is present as represented by an amber light (trouble alarm), the information regarding the amber light is relayed to the Shift Technical Aide << Fire Protection and he investigates. Lesson Objective 1.1.4 of FP-LP-3.0 (Attachment 2-6) requires that the Control Operator be able to relate the integrated response of the Main Fire Detection Information Center to a Local Fire Detection Panel. The only real information a Control Operator needs to be able to interpret is that associated with a true fire alarm and not that associated with a trouble (amber light and non-fire alarm) condition in the fire protection system.

RECOMMENDATION: 2.08 DELETE question 2.08

a. List THREE components that have their Component Cooling Water supply isolated on a phase A signal. (1.5)

b. List the TWO loads supplied by each Component Cooling Water essential loop.

(1.0)

ANSWER 2.09 (2.50)

a. 1. The Cross Failed Fuel detector.

2. The Sample System Heat Exchanger (0.5 each ans.)

3. The Excess Letdown Heat Exchanger

b. 1. One RHR Heat Exchanger

2. One RHR Pump Oil Cooler (0.5 each ans.)

REFERENCE:

SHEARON HARRIS SD-145 p. 5 and 16 3~3 008000K102 ...(KA'S)

CP&L COMMENT: 2.09 Part a.

The reference cited, SD-145, p. 16 (Attachment 2-7) is incorrect for isolation signals. The Excess Letdovn HX and the HCDT HX CCW val~ves 100-376 and lCC-200) are isolated on a Phase A signal as shown in the OMM-004, Phase A Verification Form (Attachment 2-8). The GFFD Valve (1CC-304 and 305) and Sample Panel CCW valves (1CC-114 and 115) are isolated on a Safety Injection Signal directly as shown in the OMM"004, SI Verification Form (Attachment 2"9) and on Lo CCW Surge Tank level as shown in AOP-014 (Attachment 2-10). CCW-LP-3.0, p. 20 (Attachment 2"11) summarizes the isolation signals.

Part b:

The only CCW cooled cooler associated with the RHR pumps is the "Seal" cooler, not the "oilfg cooler. The reference cited, SD-145, p. 5 (Attachment 2-12) it ~correctl states the function in the preceding paragraph CCW-LP.-3.0, p. 0 (Attachment 2-13) correctly identifies the function as cooling the RHR Seal Wate~ HX.

RECOMMENDATION: 2.09 (Continued)

Part a:

Delete the GFFD and Sample System HX's from the answer key. Accept instead, the RCDT HX and Excess Letdown Hx, ignore any third answer, and adjust point values appropriately.

Part b:

Accept RHR Pumps "Seal" cooler instead of RHR pump "oil" cooler on the answer key.

Which ONE of the following would result in the modulation of the Instrument and Service Air Crosstie valve (1IA-648):

a. Instrument air pressure is 80 psig and Service air pressure is 92 psig.

b. Instrument air pressure is 88 psig and Service air pressure is 92 psig.

c. Instrument air pressure is 98 psig and Service air pressure is 78 psig.

d. Instrument air pressure is 93 psig and Service air pressure is 88 psig.

ANSWER 2.15 (1.00) d.

REFERENCE:

SHEARON HARRIS SD-151 p.10 SHEARON HARRIS ISA-LP-3.0 File No. 5.5 p. 8 3.2 078000K402 ...(KA'S)

CP&L COMMENT: 2.15 The question represents too high a degree of required recall for the Control Operator. This question requires a memorization of the logic associated with 1IA-648 given in SD-151 (Attachment 2"14), information which is readily available in the form of a logic diagram. A more REALISTIC degree of. recall would be that in AOP-017 under Section 2.0, AUTOMATIC ACTIONS (Attachment 2-15 ). Here, priority is placed on Instrument Air Header Pressure and open (closed) positions of 1SA-6 and lIA-648.

RECOMMENDATION: 2.15 Delete the question.

8 Answer EACH of the following with regard to the Emergency Service Water System.'.

LIST two (2) design features of the ESW system that prevent the escape of radioactivity from containment via the ESW header during a loss of coolant accident.

b. A valve interlock prevents opening the ESW pump backup suction supply valves while the preferred supply valves are still open. STATE the purpose of this interlock.

ANSWER 2. 19 (1 ~ 50)

a. 1. The ESW booster pumps start on an SI signal.

2. The containment air cooler orifice bypass valves close. (0.5 each)

b. To prevent sluicing water from the auxiliary reservoir (preferred source) to the main reservoir (backup source). (0.5)

REFERENCE:

SHNPP ESWS LP 3 ~ Oy p 13'7 19' Oo 1 1 6g 1 ~ 1 ~ 3y 1 1 ~ 5 076000K402 076000K119 ...(KA'S)

CP&L COMMENT: 2.19 Part b:

This valve interlock no longer exists following FCR's E-1031, E-1044 and I-3545 which determinated 1SW-1, 2, 3, and 4. 1SW-1 through 4 are manually operated and required to be locked in position.

These yales are shown as manual valves on ESW-TP-1.0 (Attachment 2-16), and in ESW-LP-3.0, pp. 14 & 15 (Attachment 2"17) ~ Additionally, the valve lineup in OP-137 (Attachment 2-18) specifies these valves as "locked open" or "locked closed", a designation that could only be applied to manual valves.

RECOMMENDATION: 2+19 Part b:

Delete this part of question 2.19

a. List the THREE conditions that will satisfy the RHR System interlocks and allow the RHRS hot leg suction valves (RH-1; RH-2, RH-39, RH-40) to be opened. (1,0)

b. What condition will automatically OPEN and what condition will automatically CLOSE the RHRS miniflow valves (RH-31 and RH-69)7 (0.5)

ANSWER 3.07 (1.50)

a. 1. RCS pressure < 363 psig +/-5 psig.

2. RHR discharge to CSIP suction valves (RH-25/RH"63) shut.

3. Suction from RWST must be shut. (0.33 each ansi'

b. Automatically OPEN when RHRS flow is between 725 and 775 gpm.

Automatically CLOSE when RHRS flow is between 1375 and 1425 gpm.

(0.25 each ans.)

REFERENCE:

SHEARON HARRIS RHRS-LP-3.0 File No. 2.2 p. 20,21 005000K407 3 ' ~ ~ (KA S)

CP&L COMMENT: 3.07 The answers to part b are stated as between particular sets of valves. It is unclear whether these valves represent a range of acceptable answer or exact valves are required. The REFERENCE cited, RHR-LP-3 0, p. 21 (Attachment if the 3-1) gives valves for auto opening (746 gpm) and auto closure (1402 gpm).

These numbers are valid for 350'F. SD-111, p. 12 (Attachment 3"2) states second set of valves valid at 68'F: auto open at 713 gpm and auto close at 1339 gpm. The question does not specify the operating temperature.

RECOMMENDATION: 3 e07 Accept a range for auto opening of 750 gpm (+ 50 gpm), and a range for auto closure of 1375 gpm (+ 50 gpm)

The following pertain to indications on the Reactor Vessel Level Indicating Sys'ame

a. What will the upper range indication show when a RCP is running in the associated Loop?

b. How does dynamic head indication change as reactor power is increased from

'0 - 100X?

c. Is OPERABILITY of the Reactor Vessel Level Indicating System required by Technical Specification in Mode 1?

ANSWER 3.16 (1.50)

a. Upper range will indicate minimum level.

b. Dynamic head will read higher than 100X.

c. Yes (accident Monitoring Instrumentation) (0.5 each)

REFERENCE SHEARON HARRIS ICCM-LP-3.0 File No. 10.16 p. 14, 15,21 and 27 016000A302 016000K101 2 ' 3' ~ ~ ~ (KA S)

CPGL COMMENT: 3.16 The answer for part a is stated as "minimum level". The wording "offscale low" is an equivalent description and is used in ICCM-LP-3.0, p. 14 & 15 (Attachment 3"3). Figure 7.10 in SD-106 (Attachment 3-4) shows the expected indictions if RCPs are running.

Part h oi the question ask hoe the RVLIS dynamic head indiction ~chan es. The answer given in the key makes no reference to changing values. Instead it gives the expected indication for power operations.

RECOMMENDATION: 3.16 For part a accept "offscale low" as equivalent wording'elete part b.

(2.00)

a. List the TWO types of power (voltage, phase, frequency) supplied to the DC Hold Cabinet AND state the source for each type. (1.2)

b. List the functions of the 125 VDC and 70 VDC power outputs from the DC Hold Cabinet.

(0.8)

ANSWER 3 ~ 18 (2.00)

a. l. 260 VAC, 3 phase, 58.3Hz (0.3)

From the rod drive MG sets. (0.3)

2. 120 VAC, 1 phase, 58.3Hz (0.3)

From the rod drive MG sets. (0.3)

b. 125 VDC for latching 70 VDC for holding rods (0.3)

(0 ') (0.2 for correct association)

REFERENCE SHEARON HARRIS SD-104 p. 8; RODCS"LP-3.0 File NO. 10.6 p.31 001050C007 3.2 ...(KA'S)

CP&L COMMENT: 3.18 different values for the frequency of the The two REFERENCES give two Bets. SD-104 gives 58.3Hz, (Attachment 3-5) gives a the value given in the key.

value at 58.5Hz.

RODCS-LP-3 ', MG

p. 30 RECOMMENDATION: 3.18 Accept either of two values for frequency - 58 'Hz or 58 'Hz.

Which ONE of the following statements correctly describes the operation of the Main Steam Line isolation logic?

a. Any ESFAS signal which isolates the MSIVs will also isolate the steam supplies to the turbine driven auxiliary feedwater pump.

b. A low steam line pressure signal in one channel of 2/3 main steam lines will initiate an isolation signal.

c. A trip signal to an MSIV causes redundant solenoid valves to energize and bleed air from pilot valves'.

A retentive memory in the isolation logic prevents the MSIVs from being reset with the actuation signal still present.

ANSWER 3.20 (1.00) a.

REFERENCE:

SHNPP: SD-126.01, p. 11, 29 ESFAS-LP-3.0, p. 14-15, L.O. 1.1.5 039000K405 ...(KA'S)

CP&L COMMENT: 3.20 Answer a is incorrect per the OMM-004, Main Steamline Isolation Checklists (Attachment 3-6). Further documentation to support this can be obtained from Control Wiring Diagram (CWD) 2166"B-401 sheets 1974 and 1975 (Attachment 3-7). These drawings show only the control switches close these valves.

Answer d is correct per the Logic Diagram CAR-1364"871, Westinghouse Logic Diagram 108D831 Sheet 8 (Attachment 3-8) the MSIS cannot be reset if an actuation signal is present. The reset signal is generated by taking the switches to the RESET position, but the signal is reactuated as soon as the switches are relesed and the MSIS present. This type of reset, is explained in PSPR-TP-15.7 and 15.9 (Attachment 3"9).

Recoamendation.'hange answer on key from a to d.

List SIX indications, other than annunciators or radiation monitors that are symptoms of excessive RCS leakage, as listed in AOP-016, Ezcessive Primary Plant Leakage.

ANSWER 4.01

l. Increased frequency of RCS makeup

2. Increased Containment Pressure

3. Increasing xeactor vessel cavity sump level or pump operation

4. Increased reactox'oolant drain tank temperature

5. Increase in PRT parameters

6. Reactor vessel flange leak-off temperature increasing

7. PORV discharge tempexature indication increasing

8. Pressurizer Safety Valve discharge line temperature increasing

9. Increasing Containment Temperature

REFERENCE:

SHEARON HARRIS AOP-16, p. 3,4 000028A106 3.3 ...(KA'S)

CP&L COMMENTS: '4.01 AOP-016 also includes the symptom "Notification to control room of leakage by plant personnel" . (Attachment 4-1)

RECOMMENDATION: 4.01 Accepts the additional symptom x'eferenced as one of the required responses.

Require only four or five of the symptoms and adjust the point values appxopriately.

Abnormal Procedure AOP-002, Emergency Boration, lists five available paths to deliver boric acid to the suction of the charging pumps. If the normal path (through the blender) and the preferred Emergency Boration path (through 1CS-278) are not available, list the THREE remaining paths.

ANSWER 4.02 (1.50)

a. 1. From the RWST (or through LCV-1158, 115D)

2. Into the top of the VCT (or through FCV-113A and FCV-114A)

3. Bypass the Boric Acid Blender (or through FCV"113A and 1CS-287)

(0.25 each ans.)

b. 1. Seal water supply lines to RCPs.

2. Auxiliary spray to the pressurizer.,

(0.25 each ans./0.25 correct order)

REFERENCE SHEARON HARRIS AOP-LP-3.2 File No. 16.12 p. 7,8 004000K104 , 004000K117 00400K609 3.4 4,4

...(KA'S)

CP&L COMMENT: 4 o02 The unavailability of the normal path (through the blender) could be a result of a failure of FCV-113A. This failure should also make the flowpath to the top of the VCT unavailable. This would leave only two available options'.

1. From the RWST (LCV-115B, 115D)

2. Bypass blender LCV"113A, 1CS-287 RECOMMENDATION: 4.02 Accept the two flowpaths above as complete answer provided assumption made that flow path THROUGH the blender 's not available.'djust point values appropriately.

State the THREE criteria that determine when adverse containment parameters should be monitored, including setpoints where applicable.

ANSWER 4 ~ 04 (1.50)

Containment pressure (0.25) greater than or equal to 3 psig (Hi-1) (0.25) or Containment radiation (0.25) greater than or equal to 100,000 R/hr (0.25) or Integrated containment radiation dose (0.25) greater than 1,000,000 R (determined by TSC staff) (0.25)

REFERENCE:

SHEARON HARRIS EOP-LP"3.16, File No. 16.4 p. 60 000011G011 4.3 ...(KA'S)

CPSL COMMENTS: 4.04 The EOP Users Guide (Attachment 4-2) states the parameters must be "greater than" the values listed instead of "greater than or equal to" the values as stated in the Answer Key.

RECOMMENDATION: 4+04 Accept answers stated as either "greater than" or "greater than equal to".

. 8 Answer the following questions concerning procedure PLP-702, Independent Verifications

a. Attachment 7.1 to PLP-702 lists systems, subsystems and components which require independent verification. Under what conditions, as specified in PLP-702, would a system, subsystem or component NOT listed in Attachment 7.1 require independent verification?

b. When may the Shift Foreman waive the requirements for independent verification?

c. How does a qualified person outside the Shearon Harris organization receive approval to perform independent verification on plant systems or equipment?

ANSWER 4 ~ 09 (1.50)

a. When installing and removing temporary jumpers and lifting electrical leads (0.5)

b. If a component will be frequently cycled during a shift (in which case final position is independently verified). (0.5)

c. Must receive written approval (0.25) by the manager responsible for the procedure in use. (0.25 )

REFERENCE:

SHEARON HARRIS PLP-702 p. 5 to 7 194000K)01 3.6 ...(KA'S)

CP&L COMMENT: 4 '9 Objecti've 1.1.8 for LP-PP-3.0 (Attachment 4"3) requires the operator "STATE Independent Verification." The questions asked, however, are administrative in nature and therefore fall under the responsibility of the Shift Foreman.

The Control Operator would not be responsible for making the judgements required by any part of the question. Part "b" even cites the Shift Foreman as waiving the requirements.

RECOMMENDATION: 4 ~ 09 Delete the question

(2.00)

a. List FOUR indications, other than annunciators, of a Partial Loss of Condenser Vacuum, as listed in AOP-012, Partial Loss of Condenser Vacuum. (Do not include circul;.ting water flow and pressure nor condenser vacuum.)

(1.00)

b. If one of the three running Circulating Water Pumps were to trip resulting in the standby vacuum pump automatically starting, what TWO immediate operator actions are required per AOP-012, Partial Loss of Condenser Vacuum? (1.0)

ANSWER 4.11 (2.00)

a. 1. Condensate Pump discharge temperature increasing.

2. Increasing Turbine Exhaust Hood temperature

3. Abnormal Gland Seal Steam pressure.

4. Increase in Turbine vibration. (0.25 each ans.)

b. 1. Verify tripped Circulating Water Pump Discharge valve closes. (0.5)

2. Reduce Turbine load. (0.5)

REFERENCE:

SHEARON HARRIS AOP-012 p. 3,4 000051G010 2.6 ...(KA'S)

CPLL COMMENT: 4.11 AOP-016 also includes the symptom "Condenser Vacuum Breaker Valves not closed" (Attachment 4-4).

RECOMMENDATION: 4.11 Accept the additional symptom reference as one of the required responses.

Require. only three symptoms and adjust the point values appropriately.

Emergency Procedure, Path-2 (Path-2 Guide) directs the operator to adjust the ruptured SG PORV controller setpoint to 8.8 (1145 psig) and shut the MSIV and MSIV Bypass valves'hat are TWO reasons for requiring this action?

ANSWER 4 .12 (1.50)

To isolate flow from the ruptured SG (0.5) which will effectively minimize any release of radioactivity from the ruptured SG (0.5 ) and allo~ the establishment of a differential pressure between the ruptured SG and non-ruptured SG. (0.5)

REFERENCE:

SHEARON HARRIS EOP-LP-3.2 File No. 16.4 p. 13, 14 000038K306 4 .2 o..(KA'S)

CP6L COMMENT: 4+12 The answers given is appropriate for closure of the MSIVs and MSIV Bypasses only as cited in the reference, EOP-LP-3.2, pp. 13614 (ATTACHMENT 4-5). These answers are not appropriate for the adjustment the SG PORV setpoint. The basis of this action is to decrease the probability of lifting the PORV while minimizin an challen e to the code safet valves. The KNOWLEDGE Section for Step 3 in Step Description Table for E-3 Attachment 4-6) gives this additional basis.

RECOMMENDATIONS: 4 '2 Accept the two reasons below with the associated point values.'.

Isolate flow from the ruptured SG (0.50) to minimize any release of radioactivity (0.25) and all establishment of a differential pressure between the ruptured and intact SGs (0.25).

2. Minimize challenging of code safeties (0.50)

~ ~ \ 'I ~ ~

(2.00)

Answer the following questions concerning EOP usage.

a. What indication is used in an EOP procedure to inform the operator that a task must be completed before proceeding to a subsequent step?

(0.5)

b. What is the, operator required to do if a response not achieved contingency action is required, but CANNOT be successfully completed, and no additional contingency actions are listed? (0.5)

c. What operator actions are required if, during performance of steps in PATH-l, a MAGENTA terminus on a CSF Status is encountered? (1.0)

ANSWER 4 ~ 16

a. An associated NOTE or the stop will state the task must be completed prior to proceeding.

(0.5)

b. Return to next step or sub-step on the left side. (0.5)

c. Monitor all remaining trees for a RED terminus (0.5) and if not encountered, suspend any PATH in progress and perform the applicable FRP for the MAGENTA terminus. (0.5)

CP&L COMMENT: 4.16 The answer to part c is valid only after the initial actions of PATH-1 are complete. Once the operator is directed to implement FRP's, the convention applies as described in the EOPs User's Guide, p. 11 (Attachment 4-7).

Additionally the different parts of the question have different point values.

RECOMMENDATION: 4.16 Accept for part c the additional response that "the FRP would be entered after completion of the PATH-1 immediate actions." Adjust the point value consistent with the key such that each response is worth 0.50 pts. for a total of 1.50 pts.

'SELECT the one statement below that is correct if the Power Range instruments have been adjusted to 100X based on a calculated calorimetric.

a. If the feedwater temperature used in the calorimetric calculation was HIGHER than actual feedvater temperature, actual power will be LESS than indicated power.

b. If the reactor coolant pump heat input used in the calorimetric calculation is OMITTED, actual power vill be LESS than indicated power.

c. If the steam flow used in the calorimetric calculation vas LOWER than actual steam flov, actual pover will be LESS than indicated power.

d. If the steam pressure used in the calorimetric calculation is LOWER than actual steam pressure, actual'power will be LESS than indicated power.

ANSWER 5.03 & 1.26 (1.00) b.

REFERENCE:

NUS, Vol 4, pp 2.2-4 Surry 1-PT-35 SHNPP: HT-LP-3.2, L.O. 1.1.5 GP"LP-3.5 TS 3.3.1 OST 1004 2.6/3.1 3.1/3.4 01500K504 193007K108 ... (KA'S)

CP&L COMMENT: 5.03 & 1.26 The effect of steam pressure changes on a calorimetric vill depend on the pressure involved. At typical main steam pressures (950-1100 psia) a decrease in pressure leads to an increase in enthalpy (see Attachment 5-1). This increase causes hh across the steam generator to increase leading to an increase in calculated power. Thus we adjust our indicated pover so. that it is nov greater than actual power. This means answer d is also true (actual pover less than indicated) ~

RECOMMENDATION: 5 '3 & 1.26 Accept answers b or d.

WHICH one of the following situations will the insertion of control rods cause Delta I to become MORE positive?

a. Burnout of Xenon in the top of the core with rods initially fully withdrawn.

b. Positive MTC during a reactor startup.

c. Band D control rods inserted toward the core midplane.

d. Excessively negative MTC at EOL.

ANSWER 5.04 & 1 ~ 25 (1.00) a.

REFERENCE:

SHNPP RT LP 3 ~ 14~ LeOo 1 1 ~ 3~ le loll HBR RXTH"HO-1 Session [CAF]

3.2/3.5 192005K114 ..0(KA'S)

CP&L COMMENT: 5.04 & 1.25 This question apparently addresses the issue of axial flux/xenon oscillations induced by control rod insertion. When control rod insertion takes place from 100Z power with all rods initially withdrawn, for example, a flux/xenon oscillation is induced. (See Attachment 5-2 paragraphs 5.1 and 5.4). If started from at or near the fully withdrawn condition, initially a rod insertion will cause hl to become more negative (See Attachment 5"3) ~

However, as the effects of the rod insertion progress, a flux/xenon oscillation begins and eventually hl becomes more positive than it was initially (See Attachment 5-4). Regardless of which situation proposed by choices a, b, c, or d is occurring hl will become more positive eventually.

Since the wording of the question is not specific as to when the positive hl was to occur (immediately, or at any time in the future) any of the choices are correct given an insertion of control rods.

CP&L COMMENT: 5.04 & 1.25 (Continued)

If the question was intended to imply an iaxnediate increase in hl then either choices a or c are plausible. At SHNPP we have a 4 rod D bank (See Attachment 5"5). For D bank rod insertion on SHNPP Cycle 1 from fully withdrawn down to about 160 steps AI is controllable with rods (i.e. when rods are inserted, AI becomes more negative). When bank D is inserted" past about 160 steps down to 113 steps when the 8 rod bank C beings insertion, the effect on hl is no longer so predictable. In some instances bank D rod insertion in this range has either had no effect or caused hl to become more positive. The AI movement in the more positive direction is usually accompanied by hl moving more positive initially prior to rod movement. The insertion of D bank in this case sometimes serves to accelerate this more positive AI trend suggesting 'some combination of choices a and c is occurring. Attachment 5-2 paragraph 5.3.9 addresses the erratic behavior associated with relatively deep insertions of bank D, but does not take into account the observed effect noted above in the 160 to 113 step range.

RECOMMENDATION: 5+04 & 1.25 The preferred corrective action would be to delete this question due to its vague wording. A less desirable alternate corrective action would be to accept either answers a or c.

(2.00)

INDICATE whether EACH of the following fuel loading situations would result in a 1/M plot that was CONSERVATIVE (under predicts criticality) or NONCONSERVATIVE (over predicts criticality).

a. Detector located too far from core (source).

b. Detector located too near core (source).

c. Loading core from center (source) towards detector.

d. Loading highest worth assemblies first; lowest worth last.

ANSWER 5.09 (2.00)

a. NONCONSERVATIVE (0.5 each) bo NONCONSERVATIVE c NONCONSERVATIVE

d. CONSERVATIVE

REFERENCE:

Westinghouse Nuclear Training Operations, pp. I"4.19 - 21 SHNPP: RT-LP-3 .7, L.O. 1 .1 .6 AOP-LP-3.7, L.O. 1.1.1.A 4.0/4.3 3.4/4.1 2.9/3.1 000036A202 000036K103 192008K106 ~ ~ ~ (KA S)

CPSL COMMENT: 5.09 It is not clear in part a, b, 6 c what the meaning of "(Source)" is. Is the core the source'? Is the source located in the same place as the core7 Our reactor theory tezt discusses 1/M plot accuracy in terms of a geometric relationship between the detector, source, and core (See Attachment 5-6). It is not clear, particularly for part a and b, ezactly what that geometry is.

We teach the students that good conservative 1/M plots are obtained by having the detector closer to the core than the source (See Attachment 5-7). Part a implies that the core is far away so regardless of source position to infer that the 1/M plot is nonconservative. However with the it is'easonable core close by, if the source is farther away then a conservative plot would result. In this instance (core nearer than source to detector), a rapid increase in count rate occurs for additional fuel loaded near the detector.

As fuel is loaded further and further away, the detector will see a reduced increase in neutron fl.uz with each additional fuel assembly yielding a conservative 1/M plot.

RECOMMENDATION: 5+09 Since source position is not specified and it makes a difference as to whether or not the 1/M plot is conservative, delete part b of this question.

a. STATE the primary factor at BOL that causes redistribution of the axial flux as power is increased. (0.5)

b. DESCRIBE how the axial flux will shift as power is REDUCED from full to zero power at EOL. STATE the main cause of this behavior. (1.00)

ANSWER 5.16 (1.50)

a. Density changes of the moderator with core height. (O.s)

b. Flux will shift significantly towards the top of the core. (0.5)

This is due to uneven fuel burnup (higher density fuel at the top). (O.s)

REFERENCE:

Westinghouse Reactor Core Control, pp. 3-51 to 3-53 SHNPP:

CP&L COMMENT: 5.16 The main cause of the axial flux shift with reduced power at EOL could be attributed to both fuel burnup and moderator temperature (See Attachment 5-8). The disapperance of the moderator temperature coefficient induced peak low in the core (See Attachment 5.-9) and the continuing presence of the fuel burnup induced peak high in the core are the two effects involved. The attached theory text (Attachment 5-8) makes no suggestions as to a main cause',

only that two causes exist.

RECOMMENDATION: 5 '6 In part b accept either moderator temperature defect reduction or fuel burnup.

(2.so)

a. The plant is currently in Mode 5 with one train of RHR in operation.

Assume a nominal RHR flow of 4000 gpm and a reduction in temperature of 8 deg F across the RHR heat exchanger. The reactor engineer informs you that his calculated decay heat load is 0.3Z of rated power. With the above plant conditions, STATE whether you CAN or CANNOT control the heat load . SHOW YOUR WORK and state any assumptions. (1.5)

b. LIST two (2) actions that can be taken if the RHR system can not handle the heat load. (1.0)

ANSWER 5.18 (2.so)

a. m = 4000 gpm x 60 min/hr x 1 cu. ft/7.48 gal x 1 lb./.0166 cu. ft

= 1.93 x 10E6 lbs/hr (+/- 10,000 lbs/hr) (o.s)

Q

= mc(delta-T)

= 1.93 x 10E6 lbs/hr x 1 BTU/lb-deg F x 8 deg F

= 1.544 x 10E7 BTU/hr / 3.413 z 10E6 BTU/hr/MW

= 4.52 MW (0.5)

Z = 4.52 MW/2775 MW (o.25)

= 0.16X (Since 0.16 X < 0.3Z) Cannot maintain heat load. (0.25)

NOTE: ECF will be applied and comparable solutions accepted.

b. Increase mc by starting a second RHR pump (0.5 each)

Increase delta-T by increasing CCW flow

REFERENCE:

BVPS Thermodynamics Manual Chapter 3 BVPS System Description Chapter 10 SHNPP: HT-LP-3.1, L.O. 1.1.2.3 2+2/2+3 2.5/2.7 2.4/2.4 191006K1'03 191006K104 191006K108 ~ ~ ~ (KA S)

CPSL COMMENT: 5 .18 (Continued)

In part b an equally acceptable way to increase RHR cooling effectiveness'is to raise Emergency Service Water flow rate. Emergency Service Water is the cooling medium for the Component Cooling Water Heat (see Attachment 5-10)

Exchangers. Increasing Emergency Service Water flow increases heat transfer out of the Component Cooling Water System which in turn increases heat transfer out of the Residual Heat Removal System.

RECOMMENDATION: 5 . 18 In part b add "increase Emergency Service Water flow" as an alternate acceptable answer.

NRC UESTION: 6.01 & 3.20 (1.00)

WHICH one of the following statements correctly describes the operation of the Main Steam Line isolation logic?

a. Any ESFAS signal which isolates the MSIVs will also isolate the steam supplies to the turbine driven auxiliary feedwater pump.

b. A low steam line pressure signai in one channel of 2/3 main steam'lines will initiate an isolation signal.

C ~ A trip signal to an MSIV causes redundant solenoid valves to energize and bleed air from the MSIV pilot valves.

d. A retentive memory in the isolation logic prevents the MSIVs from being reset with the actuation signal still present.

ANSWER 6.01 & 3.20 (1.00)

REFERENCE:

SHNPP: SD-126.01, p. 11, 29 ESFAS-LP-3.0, p. 14-15, L.O. 1.1.5 3.7/3.7 03900K405 ...(KA's)

CP&L COMMENT: '6.01 & 3 '0 Answer "a" in incorrect per Attachment 6-1 (OMM-004, Page 57 and 58). 1MS70 and 72 (TDAFW pump steam isolation valves) do not isolate on a Main Steam Isolation Signal. Further documentation to support this can be obtained from Control Wiring Diagram (CWD) 2166-8-401 Sheets 1974 and 1975 "d" is correct per the Logic Diagram CAR-1364-871, Westinghouse Logic

'nswer Diagram 108D831 Sheet 8 (Attachment 6"2) ~ The ISIS cannot be reset if an actuation signal is still present.

RECOMMENDATION! F 01 & 3 '0 Change answer on 'key from uau to udu

WHICH one of the following statements correctly describes the operation of the reactor trip breaker shunt trip coils?

a. They provide the primary mechanism for tripping the reactor in response to automatic and manual trip signals.

b. They deenergize in response to a reactor trip signal thereby operating a

'lever which strikes the breaker trip bar to open the breaker.

c. They are ONLY on the main trip breakers and not on the bypass breakers.

d. They energize ONLY in response to automatic reactor trip signals.

ANSWER 6.02 (1. 00.)

REFERENCE:

SHNPP: SD-103, p. 11 RPS-LP-3.0, L.O. 1.1.8 3.>/4.2 001000K603 ~ ~ ~ (KA S)

CP&L COMMENT: 6 '2 Answer "c" is not correct as supported by RPS-TP-22.0 and Logic Diagram CAR-1364-871, Westinghouse Logic Diagram 108D831 Sheet 2 (Attachments 6-3.1 & 6-3.2). Our bypass breakers do have shunt trip (ST) coils'he shunt trip coils energize at the same time the VV coils deenergize on the Reactor Trip Signal.

RECOMMENDATION: 6o02 Delete this question since none of the four choices are correct.

(F 00)

The pl'ant is operating normally at 100Z power with all control systems in AUTOMATIC. A normal load reduction to 90X power is initiated, but the.

controlling feedwater flow transmitter for the "A" steam generator remains stuck at the 100Z value. SELECT the one (1) statement below which correctly describes the effects of this malfunction if NO ACTION is taken to correct the problem.

a. Steam generator level will stabilize at a level sufficiently LESS than the original level to offset the flow error.

b. Steam generator level will stabilize at a level sufficiently MORE than the original level to offset the flow error.

c. Steam generator level will remain stable at 66Z because of the constant level program regardless of power level.

d. Steam generator level will oscillate around the 66X program setpoint as flow and level errors rise and fall.

ANSWER 6.04 (1.00)

REFERENCE:

SHNPP: SGWLC-LP-3.0, p. 5-6, L.O. 1.1.4; SD-126.02, p. 9 3.4/3.4 059000K104 ~ ~ ~ (KA s)

CP&L COMMENT: 6 e04 Due to the small feed flow deviation involved in this transient, answer "d" can also be justified per SGWLC-LP-3.0 pages 7 and 8 (Attachment 6-4).

Flow error = 10Z Valve lift = 2X = 10Z ( ~ 2X valve lift/X flow)

Level error = 3.3X valve lift/X level deviation

.'. When flow decreases, level will decrease producing some level error (Dominant Signal). Level will rise due to the corrective signal. When level returns to normal (66X) then the Level Error Signal will be gone and feed flow will cause valve to go closed again.

RECOMMENDATION: 6+04 Also accept "d" for this slight transient due to answer "a" containing the word "sufficiently".

Answer EACH of the following with regard to the Emergency Service Water System.'.

LIST two (2) design features of the ESW system that prevent the escape of radioactivity from containment via the ESW header during a loss of coolant accident.

b. A valve interlock prevents opening the ESW pump backup suction supply valves while the preferred supply valves are still open. STATE the purpose of this interlock.

ANSWER 6.07 (1.50)

a. 1. The ESW booster pumps start on an SI signal.

2. The containment air cooler orifice bypass valves close. (0.5 each)

b. To prevent sluicing water from the auxiliary reservoir (preferred source) to the main reservoir (backup source) ~ (0.5)

REFERENCE SHNPP ESWS LP 3 ~ Oy po 13'7 19'oOo 1 ~ 1 ~ 6y 1 ~ 1 ~ 3y 1 ~ 1 ~ 5 3.6/3.7 2.9/3.2 07600K119 076000K402 ~ ~ ~ (KA S)

CP&L COMMENT: 6.07 (1.50)

Part b. of this question is no longer applicable to SHNPP per FCR-E-1031, 1044, and 3545. This can be shown per. ESWS-TP-1.0 (Attachment 6"5) ESWS-LP-3.0 Pages 14 and 15 (Attachment 6-.6), and OP-139 Page 49 (Attachment 6-7).

The motors for these valves have been removed as well as their MCB Control Switches. Valves are now manually operated with position indication on the MCB.

RECOMMENDATION: 6.07 (1.50)

Delete Part b.

The pressurizer protection circuits generate several signals that feed the reactor protection or safeguards initiation circuits. LIST the five (5) protection signals - INCLVDING SETPOINTS " generated by pressurizer pressure.

ANSWER 6.08 (2.50)

1. Lou pressure reactor trip [0.4] - 1960 psig [0.1]

2. Lov pressure SI [0.4] 1850 psig [0.1]

3. P-ll permissive bistable [0.4] " 2000 psig [0.1]

4. High pressure reactor trip [0.4] - 2385 psig [0.1]

5. Over temperature delta-T [0.4] - variable [0.1]

REFERENCE SHNPP: SD-100.03, p. 12 PZRPC LP 3 '~ p 12 13~ L 0 ~ 1 ~ 1 4~ lan 5 3.9/4.1 3.9/4.1 3.8/4.1 010000K101 "'10000K102 010000K403 .0.(KA'S)

CPhL COMMENT: 6.08 For OThT setpoint, accept Tech Spec value of 109X (+ penalties) as well as variable (See Attachment 6-8)

RECOMMENDATION: 6+08 Accept 109X (+ penalties) as well as variable

~ .

(2.00)

Answer EACH of the folloving with regard to 118 volt AC Uninterruptable Instrument Panel 1DP-1A"Sl:

a. LIST the normal, backup and bypass power sources for this instrument panel. INCLUDE the bus designation.

b. .TRUE or FALSE:

If the ESF inverter (7.5 KVA Channel I) vere to malfunction',

pover to the instrument panel would automatically transfer to the backup source. (0.5)

ANSWER 6 ~ 13 (2.00)

a. Normal - 480 V AC Emerg Bus lA3-AA Backup 125 V DC Emerg Bus DPlA-SA Bypass 480 V AC Emerg Bus lA3-SA (MCC 1A21-SA; PP lA211-SA) (0.5 each)

b. False

REFERENCE:

SHNPP: SD"156, p. 11, 27 120VUPS LP 3 'p p 7 8g L 0 F 1 4y 1 1 ~ 7 3.1/3.5 2.7/3.2 062000K410 063000K102 ~ ~ ~ (KA S)

CPhL COMMENT: 6.13 Normal supply for the S-I inverter is from 480 VAC MCC-1A21-SA vhich gets its power from 480 VAC Bus lA3-SA. (See Attachment 6-10).

RECOMMENDATION: 6.13 Accept MCC-lA21-SA as veil as 480 VAC Bus 1A3-SA.

(2.00)

STATE what actions must be taken and conditions/interlocks met to trip the Emergency Diesel Generators (EDGs) from EACH of the following locations. BE SPECIFICl

a. Diesel Engine Control Panel (DECP)

b. Auxiliary Control Panel (ACP)

ANSWER 6.20 (2.00)

a. 1. The MCSS must be in LOCAL (0.25)

2. Simultaneously (0.25) depress the EMERGENCY STOP (0.25) and the EMERGENCY STOP THINK pushbuttons (0.33)

b. Simultaneously (0.33) depress the EMERGENCY STOP (0.33) and the ACP TRANSFER CONTROL pushbuttons (0.33)

REFERENCE:

SHNPP SACP LP 3 ~ 0 ~ pe 16 ~ L 0~ 1 ~ 1~2 3.9/4.2 064000K402 ...(KA'S)

CP&L COMMENT: 6 .20 a) The answer key assumes the diesel was started and is running on an SI or UV signal. Whether the Diesel is running on a normal or emergency start is not stated in the question. If candidate assumes the Diesel is running via a normal start from OP-155, the normal engine stop pushbutton will also stop the Diesel. (See Attachment 6-11).

b) Answer states "ACP TRANSFER CONTROL PUSHBUTTON DEPRESSED". There is no such pushbutton. If the key svitch on the transfer panels is in the transfer position and the transfer svitch on the ACP has been actuated, then 6 transfer relays (latching type relays) will be rolled to the "TRANSFER (LOCAL)" position. This will enable an emergency shutdown of the diesel from the ACP if the operator places the Diesel Emergency Shutdovn Switch in the TRIP position. (See Attachment 6-11) ~

RECOMMENDATION: 6 ~ 20 a) Also accept normal engine stop pushbutton as an alternate answer b) Accept the following:

Transfer relays in the TRANSFER (LOCAL) POSITION AND The Diesel Emergency Shutdown Switch on the ACP is in the TRIP position.

Unit 1 has a Tavg of 250 deg F and is in the process of raising temperature to the normal operating range for plant startup. Twelve hours ago, RHR Heat Exchanger A was declared INOPERABLE. The maintenance supervisor now reports that the suction valve from the Containment Sump to RHR Pump B is INOPERABLE. Upon review, you concur . From the following statements, SELECT the one that correctly describes the allowances and/or limitations imposed by the Technical Specifications that apply in this situation.

NOTE: APPLICABLE TS ARE ENCLOSED FOR REFERENCE

a. Suspend all operations involving reductions in Reactor Coolant System (RCS) boron concentration and immediately initiate corrective action to return loop to operation.

b. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, action shall be initiated to place the unit in at least COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. Restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg less than 350 deg F by use of alternate heat removal methods.

d. Restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ANSWER 8.03 (1.00)

REFERENCE:

SHNPP: TS 3/4.5.3 TS-LP-3.0, L.O. 1.1.7 3.5/4.2 3.6/4.2 006000G005 006000G011 ~ ~ ~ (KA S)

CP6L COMMENT: 8.03 Based on given data both Trains of RHR are inoperable.. One due to an inoperable heat exchanger, second due to an inoperable containment sump suction valve.

Upon review of action statement (See Attachment 8.1) for the above condition, it was determined that no action statement addresses the inoperable containment sump suction valves. None of the actions satisfy the exact situation, therefore T.S.3.0.3

'3 applies'ECOMMENDATION:

8 Answer selection b. states the actions required per T.S.3,0 ' and should therefore be considered the correct answer for this question.

~ q

~ ~

NRC UESTION: 8.05 (1.00)

The reactor is operating at 20Z power, normal operating temperature with all systems in AUTOMATIC. WHICH one of the following situations does NOT have an associated 1-hour Technical Specification action items

a. One shutdown rod is found to be partially inserted.

b. One of three Overpower Delta T indications has failed.

c. One isolation valve on an RCS accumulator is found closed.

d. The RWST solution temperature is 35 deg F.

ANSWER: 8.05 (1.00) c (Requires IMMEDIATE action)

REFERENCE:

SHNPP: TS 3.1.3.5 TS 3.3.1, Table 3.3 '

TS 3.5.1 TS 3.5.4 TS-LP-3.0, L.O. 1.1.7 3.5/4.2 3 '/4.2 006000G005 006000G011 ..;(KA'S)

CP&L COMMENT: 8+05 Per answer key, item c (Immediate T.S) is correct. However, item b is also correct.

1) If only indication is failed, then the channel is not inoperable and therefore no T.S. apply

2) If the channel is declared inoperable due to indication failure then the applicable T.S has a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> requirement. (See Attachments 8.2 and 8.3)

In both cases above (items 1 6 2) the associated action is NOT a 1 hour T.S.

RECOMMENDATION: 8 +05 Accept either items b or c as correct. (Question asked for only one).

The Control Operator has just satisfactorily completed an operations surveillance test and submitted it to you, as Shift Foreman, for disposition. STATE the three (3) actions per OMM-001, "Conduct of Operations" you are required to take with regard to the completed test ANSWER 8.09 (1.50)

1. Review it (for completeness and accuracy).

2. Sign and date the procedure.

3. Route the completed test to the Operating Supervisor/designee (and ISI, as required, for review and Document Control for retention). -

(0.5 each)

REFERENCE:

SHNPP: OMM-001, p. 64 PP LP 3 0~ L Oo lolo4 2.5/3.4 194001A103 ...(KA'S)

CP&L COMMENT: 8.09 OMM-001 "Operations-Conduct of Operations" (see Attachment 8.5) listed four actions required vice three as stated in the answer key:

1. Review

2. Sign and date

3. Ensure entry is made on the Control Room Surveillance Test Schedule to document completion

4. Route the completed test RECOMMENDATION: 8.09 Add "Ensure entry made on surveillance test schedule" to answer key as acceptable alternate answer.

The unit is operating normally at full power with only one significant inoperable component - the 1B CSIP - which is not expected to be repaired for three days'hile performing a periodic surveillance test on the lA emergency diesel generator, it trips unexpectedly and is declared inoperable at 11:00 a.m. The EDG is repaired, satisfactorily tested and restored to operability at 8:00 p.m., that evening. LIST all the LCO compensatory actions that were required to have been completed as a result of this equipment failure.

INCLUDE the time/day by which each must be completed.

NOTE: APPLICABLE TS ARE ENCLOSED FOR REFERENCE ANSWER 8.12 (2.00)

1. Demonstrate operability of offsite sources by 12:00 noon, same day

2. Verify operability of all redundant components by l!00 p.m., same day.

3. Demonstrate operability of offsite AC sources by 8:00 p.m., same day.

4." . Test the 1B EDG by 11:00 a.m., next day.

(0.5 each)

REFERENCE:

SHNPP: TS TS-LP"3 SACP-LP-3

',',

3.8. 1.1 L.O. 1.1.7 L.O. 1.) ~ 7 CP&L COMMENT: 8o12 All compensatory actions on the answer key are correct, however one additional action was not included on the Answer Key.

Per T.S. 3.8.11 action d.l, (see Attachment 8.4) with the B CSIP inoperable (given), the action to verify all components on the 1B Safety bus within two hours could not be satisfied. Based on this, the action to place the unit in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> applies.

RECOMMENDATION: 8+12 Add to answer key as another acceptable response Unit in Hot Standby by 7:00 p.m., same day"

ENCLOSURE 4 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Carolina Power and Light Company Facility Licensee Docket No.: 50-400.

Operating Tests administered at: , Shearon Harris Nuclear Power Plant Operating Tests Given On: April 26-28, 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed:

(1) Out of a total of 11 simulator scenarios run during this time period, there were five simulator lockups, each resulting in 15-20 minute delays in continuing the simulator scenarios.

(2) There was no capability to simulate radiation monitor response as the Radiator Honitoring panels were inoperable.