ML20006D585

Forwards Comments & Refs Re post-exam Review of Senior Reactor Operator & Reactor Operator Exams Administered on 891030

ML20006D585
Person / Time
Site:	Catawba
Issue date:	11/02/1989
From:	Owen T DUKE POWER CO.
To:	Baldwin R NRC
Shared Package
ML20006D582	List: ML20006D585
References
NUDOCS 9002140078
Download: ML20006D585 (22)

Text

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.- 1) uke frukt Company

' CatawbiNuckar Station ,

O' :0. I (803) 5 3000 5 l'O Bat 256 ~

Cloter. S.c 29710 :

ENCLOSURE 3 lDUKEPOWER

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• .Hovember[2,-1989 TO: .R. S. Baldwin, Chief Examiner

SUBJECT:

'NRC Exam Comments-

'FROM: Catawba Nuclear Station The following attached- coments and references have been prepared as

a. result of:our post exam review of the CNS NRC administered SRO and RO exams on' 10/30/89. 'Please take into account these comments-in.your grading of the examinees' exams to provide an accurate evaluation of

'their individual. responses. We are pleased at the high quality exam that'was administered with entirely objective questions. We commend

-.you in your efforts. Thank you for the service provided, w-Tony B.m Dwen,: Station Manager Catawba Nuclear Station 0

CTK/sjw- '

.. Attachments.

cc: W. H. Barron J. W. Cox G.:E. Spurlin-

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&  ;.:tUESTION 2.02: (1.00)H

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Which ONE of; the ~ following is NOT ari-interlock between the GTA breaker.

and the FTA= incoming feeder from ETA?-

La); If;GTA is closed, FTA cannot.be closed.~.

'b) ~ ~If GTA remainsLclosed with power lavailable, the blackout C: sequencer will close FTA then close the, ETA feeder breaker.:

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.c) ~If'GTA.is closed, FTA can;be closed only'if-the-diesel g- Ei.; generator output breaker is closed.

m 4 d)-~ -If FTA is closed and.GTA is then closed, FTA will trip.

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-COMMENTt2.02 s

LThis question has two correct-answers. The question asks which one

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an' interlock. That means 3 selections must be interlocks.-

Selection "a": 18:ancinterlock per the_ attached.--Selection "b" is

!-i 1NOT an' interlock.:aa the: answer key incidates. Selection "c" is:also cNOT an~ interlock since "a".-applies.'at all times..

• [ RECOMMENDATION 2,02 Delete this question'from the exam.

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c. An Auto Transfer will occur. After a loss of power on the Nomal Loadcenter, the Normal Feeder- to the McC's is tripped and the Alternate Loadcenter feeder will close after a time delay.

4. An interlock exists.between the GTA(B) breaker and the FTA(B) incoming feeder from ETA (B).'

a. If GTA(B) is closed, the-FTA(B); feeder from ETA (B).

cannot be closed.

b,- If the FTA(B), feeder from ETA (B) is closed and GTA(B)'

is then closed, the.FTA(B)Jincoming' feeder from ETA (B) will trip.

c. To closelGTA(B)-with the FTA(B): feeder from' ETA (B) closed,;the D/G must be~ running and intsync with.the-D/G output breaker open. .If the D/G is shutdown or

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supplying the Essential bustwith,the FTA(B) alternate breaker closed, then GTA(B) cannot' be closed.:

, 5. There is also an interlock between the-breaker on ETA (B) to.FTA(B) and the FTA(B) breaker from ETA (B).

a. The breaker on FTA(B) must be closed before the '

breaker on ETA (B) can be closed.

b. When'the incoming breaker on FTA(B) opene, the breaker on ETA (B) will open.

, 6. ~ The blackout bus will be loaded on by the sequencer during a blackout if the blackout is due to a loss of the 4160V transformer (GTA opens). If GTA remains closed with power available, the sequencer will not load the B/0 bus. The sequencer will close the bkr on FTA, then close the ETA feeder to FTA.

'7.- To swap FTA(B) power supplies, refer to OP/1/A/6350/05

" Alternate AC Power Sources".

8. The FTA(B) buses-are located in=the TB 594 level, behind the control room.

-9. GTA(B) is beside ATC(D) in the TB basement.

10. The FTA(B) and GTA(B) breakers are the same as the 13.8KV breakers and will be covered with 13.8KV system.

INSTRUCTOR NOTE: Stop at this point and ask the following questions to ensure student's understanding of material:

1. List the power sources available to the Essential Switchgear?

2. When would the sequencer not load on the B/O bus?

3. What type of loads are powered from the Essential Switchgear.

O. 13.8KV Power System (ISS/RO/SRO #16, 16a & 17)

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. QUESTION 3.15 (1.00)

Which ONE of the following operating limits is based on minimizing the probability of brittle fracture?

a) Steam generator temperature limits b) Cooldown rate limit c) Minimuni tcmperature for criticality d) 1.ow Tave limit ANSWER 3.15 (1.00) b)

REFERENCE 3-OP-CN-ilo-PTS Objective 11 002000G005 ..(Y# s)

COMMENT 3,15 A pre-exam review comment was made to change distractor "a", since "a" is also correct. It was requested to char.ge "a" to "PZR Spray Nozzle AT limits". .As is, there are three correct answers "a", "b" and "c". See attached.

RECOMMENDATION'3.15

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3.15 Oha for 3 PLANTSYSTEMS BASES STEAM GENERATOR POWER OPERATED RELIEF VALVES (Continued) concurrent with loss of offsite power. This assumes that the PORV on the ruptured steam generator is unavailable, and that the other two are used to cool- the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator.

Concurrent with the requirement that a specific number of PORVs be OPERABLE is the requirement that the associated PORV block valves upstream be open or OPERABLE.

Should an associated PORY block valve be closed and inoperable, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERA 8LE status.

Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are based on a steam generator RT to prevent brittle fracture. NOT of 60'F and are sufficient r

CATAWBA - UNITS 1 & 2 B 3/4 7-2a Amendment No. 63 (Unit 1)

Amendment No. 62 (Unit 2)

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$FERENc6 3. /r O 2 ef y Pfoe 4

1 REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

, 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4 3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and -

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods prov'ided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200*F/h, respectively, and

5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the vessel are determined in accordance with the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the NRC Branch Technical Position MTEB 5-2, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code.

I CATAWBA - UNITS 1 & 2 B 3/4 4-7 L

REACTIVITY. CONTROL Sy Ms FUEdE @& dp fog BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC usociated with a core condition of all-rods inserted (most positive MDC) to en all rods withdrawn condition and, a conversion for the rate of change of moe rator density with temperature at RATED THERMAL POWER conditions. Thp value of the MDC was then transformed into tbg limiting MTC value -4.1 x 10 ok/iv'F . Tfte HTC value of -3.2 x 10 Ak/k/*F represents a conservative v0 ue (with corrections for burnup and soluble boron) at a corw conditfon of 300 ppm equilibriumboronconcentrationand.jsobtainedby$4kingthSWcorrectionsto the limiting MTC value of -4.1 x 10 Ak/k/*F. l The Surveillance Requirements for measureme s of the HTC et we beginning and near the end of the fuel cycle are adequate to confirm that the 'HTC remains within its limits since this coefficient changes slowly due pe bcicatly to the reduction in boron concentration associated with fuel burnup..

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be rado critical with the Reactor Coolant System average temperature less then E!ilF. This limitation is requii'od to ensu' re: (1) the moderator tempe"atuto coefficient is within it analyzed temperature range, (2) the trip inst m entation is within its normal operating range, (3) the P-12 interlock is above its setpoint, (4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RT tuperature.

g 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The componer.ts required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.-

With the coolant average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN CATAWBA - UNITS 1&2 B 3/4 1-2 Amendment No. 14(Unit 1)

Amendment No. 6 (Unit 2)

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. QUESTION 3.26 :(l.00):

. Given the following conditions:

1 - ;The'CA' system' automatically" started on S/G ?BLO-LO level

-< Trains A and B CA were reset-to manual control

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'- 'The valves were repositioned to control' level for S/G A,.C, and D-

.S/G 'A' level decreased to the LO-LO setpoint, 4

3 c; >U . Which ONE~of'the following correctly describes CA system response for the

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given monditions?

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a) The< turbine-CA pump starts.but the discharge valves do not open.

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. b )' - CA train'A discharge valves go to the fully open position..

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c) 'CA train.B. discharge valves go to the fully open position.

g 'd) _

System alignment is not affected.

n- ANSWER: 3.26- . ( 1. 00 )'

4T  : d).

REFERENCE:

LOP-CN-HO-CA Objective 4 4 .061000K406 ..(KA's)-

h COMMENT 3.26

A pre-exam review comment was made that this question is not correct.

2 Per the attached there is NOT a correct answer in question 3.26. The

' correct answer'la "the CAPT starts and its discharge valves go to the (fully.open position."

, -RECOMMENDATION 3.26 q

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,. l QUESTION 5.02.- ( 1. 00 )',

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QUESTION 5.50 (1.00)

There are FOUR (4) activity limits that an item must meet to be considered

" clean" or " unconditionally released" from a Restricted Control Area according to the Health Physics (Radiation Protection) Manual - SD3.8.8.

FILL IN THE BLANKS (Value below based upon "DPM per 100 square em" of sample).

a) Loose surface contamination shall not exceed DPM Beta / Gamma.

b) Total surface contamination (fixed & loose) shall not exceed DPM Beta / Gamma, c) Alpha radioactive contamination shall not exceed DPM (loose).

d) Alpha radioactive contamination shall not exceed DPM (fixed).

ANSWER 5.50 (1.00) a) 1000 b) 5000 c) 20 d) 100 (4 parto 0 0.25 ca.)

REFERENCE T&Q CO-80010, Element #8; Catawba HP Manual; SD 3.8.B'and 3.8.10 k/a 2.8 / 3.4 194001K103 ..(KA's)

COMMENT 5.50 Per attached Station Directives 3.8.8, 3.8.3, and 3.8.10 the answer to "d" is 0 not 100.

RECOMMENDATION 5.50 Change the answer key for."d".to 0.

L f r P E ReNCE S~ TO A. / ep f]!

CatawbaNucle:rStatjoh. ective 3.8.3 .' Pess 3 13 1

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4.3 Catawba Nuclear station uses the following working limits for j contamination: -

l 4.3.1 Inside the RCA Loose: 1000 dps/100 co' ST I 20 dps/100 co' s Total: 15,000 dpe/100 co' (or 450 cpe above background)  ;

on contact ()") ST l 300 dpe/100 ca' s  ;

4.3.2 Outside the RCA (Unconditional Release)

Loose: 1000 dpe/100 co8 ST l

. 20 dpn/100 co' s J Totalt. 5000 dpa/100 ca' (or 150 cpn above background) {

using a thin window pancake probe on contact  ;

()")ST, no not' observable count above i background e i 1

4.3.3 Contaminated Storage and Issue Areas (Hot Tool Crib, Not Calibration Lab, Not Instrument Shop)  !

Loose and Fixed: 2 mRea/hr on contact ()") ST j NOTE: Tools from the Hot Tool Crib, Hot Cal. Lab, or Hot Instrument Shop are identifiably marked j with red paint and are not releasable from the RCA. These tools shall be wiped down and .

begged by the individual using then prior to ,

Health Physics surveying and tagging them, ne ,

individual then transports the tools to the Contaminated Storage and Issue area. 4 l Contaminated Storage and Issue personnel may. '

restock and reissue tools below this limit.

Tools 2 mRom/hr or greater shall be decontaminated, disposed of or stored in the Contaminated Parts Warehourt as appropriate. '

4.4 Frisking and Monitoring Devices for Equipment and Personnel Whole Body Friskers (PCH-18) will be used as the primary method of frisking personnel before leaving the RCA. Both hand held friskers -

(RM14) gd hand and foot monitors will be used as'an alternate method.- Hand-held friskers and hand and foot monitor will be placed at various locations-inside the RCA/RCZ and personnel should ,

used these frisking devices anytime a potential for the spread of.

contamination exists. All personnel are required to use these <

instruments per this directive. Enclosures 6.1 through 6.6 are examples of signs placed throughout the RCA/RCZ that explain .

appropriate use of the frisking devices.

~

-.. . .. ____-__-_-___________________1

k5Y EWCE SOf}

1 i I.

CNS Directive 3.8.8 h (- ( O P. fa g

3.8' Duke Power Company Fundamentals of Radiation Protection 3.9 Code of Federal Regulations, Title 10 Parts 20 and 30.

4.0 ADDITIONAL INFORMATION 4.1 General hardling of radioactive and/or contaminated material.

4.1.1 Tools and equipment used in the RCA shall be surveyed for contamination prior to removal from the RCA, per Station

-Directive 3.4.3.

4.1.2 Wrap or bag materials as required when removing them from -

" contaminated RCE's" and contact Health Physics for appropriate survey and tagging as necessary.

4.1.3 Personnel should use caution when collecting and disposing of radioactive liquid and solid waste to assure they are disposed of in designated containers / location (s).

4.1.4 Contaminated materials left unattended shall be bagged or wrapped and have the Health Physica yellow tag attached

• with the appropriate information.

4.2 Minimise the generation of excess waste by using common sense work practices such as:

4. 2.1' Only take necesserv items into contaminated areas.

, 4.2.2 Remove packing materials and containers prior to entering the area.

4.2.3 Utilise the Aux. Blds. Tool Roos and Contaminated Warehouse.

4.2.4 Keep known contaminated items separate from clean items as much as possible.

4.3 Working limits for contamination control 4.3.1 Working Limits Inside the RCA:

Loose - 1000 dpa/100ce* Si, 20 dps/100ca 8 e Total - 15000 dps/100 ca 8(or 450 cpe above background) on contact ()") ST, 300 dpe/100cas, L

I, 4.3.2 Working Limits Outside the RCA (Unconditional Release)

Loose - 1000 dpe/100ca 8 ST, 20 dpe/100ca8 e Total - 5000 dpe/100 ca (or 8 150 cpm above background) using thin window pancake .nrobe on contact (i") 4

$T, no not observable count above background =.

i

(

CNS Stati n Dir:ctivo 6 b b [ d[

.8.10 (TS) Paga 2 of 2 -

i i

j

3.0 REFERENCES

.3.1 HP/0/B/1000/02 Taking, Counting, and Recording Surveys 3.2 'HP/0/B/1006/01 Shipment of Radioactive Material l

3.3 System Health Physics Manual 3.4 S.D. 3.7.3 Vehicle Access to Protected Area l 4.0 ADDITIONAL INFORMATION 4.1 Radiation and contamination limits for vehicles and materials,

, excluding radioactive shipments, leaving the Restricted Area are i 8

LOOSE: 1000 dpm/100 cm beta gamma (6-T),

20 dps/100 cm 8 alpha (e).

-TOTAL: Beta gamma:- <150 ccpm as measured with a thin window pancake GM detector at one half inch from the surface (Eberline HP 210 or 260 connected to a RM 14 equivalent) -

which corresponds to 5000 dpm/100 cm8 .

No not observable indication ~above background alpha (c.).

4.2 Upon indication of possible radiological contaminants, minimize the  ;

spread of contamination by: *

- Controlling suspected items / area, '

- Controlling personnel, and

- Requesting Health Physics support.

5.0 PROCEDURE 5.1 Health Physics shall survey all vehicles that have been in an area or areas listed in Section 2.1 of this Directive prior to the ';

vehicle's release from the Restricted Araa. .

5.2 Health Physics shall notify Security prior to the release of the .

vehicle from the Restricted Area. "

-6.0 ENCLOSURES-None 9

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g  ; -QUESTION

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6.05-. (1.00) 7 gyz kis ' iWhichh0NE'ofthe following conditibns would cause a P-4,+ Reactor Trip,

y.a u '

' signal to be generated?,

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p <

'[ 'a)l .RTA'and f BYB open,

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m, 4 M.. i . REFERENCE-

.I.

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K 012000A307 ' ..(KA's)-

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' COMMENT 6.05 D.

m,

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R . .' RECOMMENDATION 6.05 '

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..See' reconenendation 3.10.

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11 iQUESTIONj6.ON ( 1. 00)' .;

4

, 3 Which ONE of the following operating limits:is based on minimizing the; 'i

- probability of brittle ~ fracture of the reactor vessel?-

']

c- 1

.a). -Steam generator' temperature limits.

y;,

, . b)- .Cooldown rate: limit?  ::1 o9 7'

v c)$ Minimum tenperattire ior! criticality- ~'

ms d.

d)i Low:Tave limit , ,

, a, !

• . . . . > .E

4- .. ANSWER.

' 6'. 07 ?  : ( l'. 00)1 0 1,

0-

. b ) '.-

r-;

REFERENCE' pl -;

.- ; m

. OP-CN-HO-PTS Objective.11. ,

B s

002000G005. . . ( KA 'si).

-COMMENT 6.07 8 '

See consnent' 3.15  ;

. RECOMMENDATION-6-.D7 4"Q g

~ See reconsnendation 3.15 '

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O O QUESTION 6.17- (1.00)

Given the following conditions:

-- The CA system automatically started on S/G 'B' LO-LO level

-. Trains A and B CA were reset to manual control The' valves were repositioned to control level for S/G A, C, and D S/G 'A' level decreases to LO-LO level-Which ONE of the following correctly describes CA system response for the given conditions?

a) The. turbine CA pump starts but the discharge valves do not open, b) CA train A discharge valves go to the. fully open position.

c) CA train B discharge valves go to the fully open position, d) System alignment is not affected.

ANSWER- 6'.17 (1.00) d)

REFERENCE OP-CH-HO-CA Objective 4 061000K406 ..(KA's)

COMMENT 6.17 See comment 3.26 RECOMMENDATION 6.17 See recommendation 3.2E l-l

, i

• i

\

ENCLOSURE 4 l i

NRC RESOLUTION OF COMMENTS

a. RO/(SRO) Exam'CommentResolution f (1) Question ~2.02(5.02) .

NRC Resolution: Coment accepted. The question will be deleted from the examination and point values adjusted accordingly.

(2) Question 3.10(6.05)

NRC Resolution: Coment not accepted. Using diagram CN-IC-IPX-3 provided by the facility, it can be seen that Answer "C" is the only i correct answer for this question. Answer "C" represents the trip _

logic for a train "A" P-4 signal. The answer key will be changed to  !

reflect "C" as the only correct answer. i.

(3) Question 3,15(6.07)

NRC Resolution: Coment accepted. The question will be deleted ,

from the examination and point value adjusted accordingly.

l (4) Question 3.26(6.17) ,

NRC Resolution: Coment accepted. The question will be deleted

,- from the examination and point value adjusted accordingly. .

i

b. SR0 Exam Coment Resolution j l

l Question 5.50(d) l

! NRC Resolution: Coment accepted. The answer key will be changed as recommended by the facility.

7 1

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= ,.

i ENCLOSURE 5 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Duke Power Company Facility Docket Nos : 50-413 and 50-414 Operating Tests. Administered On: October 31 and November 1, 1989

.This fom is used only to report observaticus. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).

These observations do not affect NRC certification or approval of the simulation facility other than to provide infomation which may be used in future evaluations, No licensee action is required in response to these observations.

During the conduct of the simulator portions of the operating tests, the following items were observed:

No discrepancies were noted.