Text

Written Exam Impact of SRO Question Deletions on Exam Validity
	ML042730577
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Site:	Oyster Creek
Issue date:	09/24/2004
From:	Caruso J; NRC/RGN-I/DRS/OSB
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ML031620284	List: ML033100410; ML042730577; ML042810448; ML050870512;
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E M 0 1 BWR SRBExamination Outline Form ES-401-1 Abilities Categories cf, @q9

Notes for post-exam distribution:

1.

2.

3. SRO-12 dejeted

4.

SRO-23 deleted.

5.

SRO-25 deleted WA Change on SRO-7 from 295004/2.1.33 to 2.2.1 1 (Knowledge of the process for controlling temporary changes)

KIA Change on SRO-3 from 295024EA2.02 to 295028/EK2.04 (Knowledge of the interrelationship between High Drywell Temperature and Drywell Ventilation)

6.

It should be noted that all SRO applicants took the RO, 75 question written exam. The RO exam had two questions in Generic Knowledge and Abilities Category 3 "Radiation Control". These were Questions 64 and 65. Questions RO-2, RO-4, RO-23 and RO-31 also had Radiation Control aspects. Specifically, Question RO-23 (Tier 1, Group 1) has two knowledge aspects. One is related to High off Site release rates and the second is related to the biological effects of radioisotope Ingestion. The first one is testing RO SGTS and AOG system knowledge, while the second is really testing the SROs ability to make plant alignment changes based on the effects that the release would have (on the thyroid). This second part has a very close correlation to the SRO Generic Knowledge Category 3.

[In addition, for information, the Operating Test covered the Radiological aspect as follows: two JPMs with Radiation Control aspects; Admin 4s Approve Radioactive Discharge Permit" and PlantlSimulator

!5, "Bypass Isolation Interlock for Torus Vent Valves and Prepare to Vent the Torus". Scenario ##4 had a malfunction "RBHVAC Rad Ventiliation monitor fails upscale, RBHVAC trips, SGTS fails to staff in which the operators were required to manually start SGTS.]

Therefore, even though the SRO Written Exam had its Generic Knowledge and Abilities Category 3 "Radiation Control" question deleted (SRO-25) there were adequate opportunities in the remainder of the exam to evaluate the SROs in this generic category.

Note: 1.

2.

3.

4.

5.

. 6.*

7.

8.

9.

Ensure that at least two topics from every K/A category are sampled within each tier of the RO outline (Le., the Tier Totals in each K/A category shall not be less than two). Refer to Section D.1.c for additional guidance regarding SRO sampling.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by k1 from that specified in the table based on NRC revisions. The final RO exam must total 38875 points and the SRO-only exam must total 25 points.

Select topics from many systems and evolutions; avoid selecting more than two*

tttfee WA topics from a given system or evolution unless they relate o plant-specific Driorities.

,A$* &

tings, and point totals 19 of 34 NUREG-1021, Draft Revision 9

Emergency anc I!

I BWRE normal C K A 3

1 I:

11 BAPE # I Name I Safety Function 295006 SCRAM I1

~

295001 Partial or Corn lete Loss of Forced core FIOW Circulation ?j 4

I 295003 Partial or Complete Loss ofAC16 295021 Loss of Shutdown Cooling 14 295005 Main Turbine Generator Trip i 3

295016 Control Room Abandonment I 7 29501 8 Partial or Total Loss of CCW/8 I

I 29501 9 Partial or Total Loss of Inst. Air / 8 295023 Refueling Acc Cooling Mode I8 ll I

~~

11 295025 High Reactor Pressure / 3 I 295026 Suppression Pool High Water Temp. / 5 295027 High Containment II Temperature I 5 lination Outline Form ES-401-1 utions - Tier l/Gmup 1 (RO ISRO).

WA TopIc(s)

AA2. Abiljty to determine andor interpret the foltowrng as they a ly to PARTIAL OR COMPLETE FLOW CIRCULATION :

(CFR: 41.10 / 43.5 / 45.13)

LOSS 4K: FORCED CORE AA2.01 Powedflow map8 AA2. Ability to determine andlor interpret the following as the ap 1ytoSCRAM.

(d!

81.10 / 43.5 /45.13)

AA2.02 Control rod position.

2.2.27 Knowled e of the reheling process.

(CFR: 43.6 / 45.f3)

EA2.02 D w e l l kmerature.

3 of 34 NUREG-1021, Draft Revision 9

ES-401 I

I I

I 11

\\

2.4.20 Knowle e of operational implications of 2 OP warnings I cautions I and Group Pbint Total:

I 1 l a ll

ES-40 I ES-401 Emergency an(

UAPE # I Name / Safety Function 295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure 13 295008 High Reactor Water Level I 2 295009 Low Reactor Water Level 12 29501 0 High Drywell Pressure I 5

29501 1 High Containment Temp I 5

295012 High D Temperature / P" 295013 High Suppression Pool Temo. I 5 295014 Inadvertent Reactivity Addition / 1 29501 5 Incomplete SCRAM / 1 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont.

Isolation I5 & 7 295022 Loss of CRD PumDs / 1 295029 High Suppression Pool Wtr Lvl 15 camination Outline Form ES-401-1 i lutions -Tier l/Group 2 (RO I SRO)

KIA Topic@)

2.4.31 Knowledge of annunciators alarms and indications / and use of the response instructions.

(CFR. 41.10 / 45.3)

AA2. Ability to determine andlor inteIpret the followln as they app& to LOW REACTOR WATER LEVEL (CFR: 41.lo / 43.5 I45.13)

AA2.03 Reactor water cleanup blowdown rate................... 2.9 2.9 2.4.44 Ability to recognize abnormal indications for s stem operating which are entry !eve1 conditions KIjmergenc Operatkg Procedures [related to High Drywely tern erature (13%:

41.1d/43.5/45.13) meters AA2.03 Ability to determine andlor interpret fhe cause of reactiyity adetion as it applies to madvertent reachvity ad&bon (CFR: 41.10/43.5 /45.13) 5 of 34 NUREG-I 021, Draft Revision 9

ES-401 295032 High Secondary Containment Area Temperature /

5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation I9 295035 Seconda Containment High Differential geswre / 5 295036 Seconda Containment High SumpfArea %ate, Level I 5 500000 High CTMT Hydrogen Conc. I5 WA Category Point Totals:

Group Point Total:

7 I

4

ES-401 Ih IR 3.8 11 ES-401 1

11 System # I Name 203000 RHWLPCI:

Injection Mode I

205000 Shutdown 11 Cooling 207000 Isolation 209002 HPCS 21 1000 SLC 21 2000 RPS 215003 IRM 21 5004 Source Range 217000 RClC 11 218000 ADS 223002 PCIS/Nuclear Steam SUDDIV Shutoff Level Control 11 261 000 SGTS Plant Systf BWR sR(FEx is - Tie F

O

~

ES-401-I nination Outline p 1 (RO I SRO)

WA Topic(s)

A2. Ability to (a) predict the impacts of thefollowm on MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correc those abnormal conditions or operations:

(CFR 41.5 / 45.6)

A2.04 Up scale or down scale trips the IFJT~RMEDIATE RANGE control, or mitigate i l e consequences of 7 of34 NUREG-I 021 Draft Revision 9

ES-401 (Cm.43.5 145.12145.13)

S 2.?.25 Knowledge of bases in TS for limiting conditions of operations and safe limits [Related to AC Electrical 3.7 Distrr % ution 2.4.48 Ability to interpret control room indications to verify the status and operation of system / and understand how operator action s and directives afect lant and s stem conditions. Related to EDGs]

(EFR: 43.5 145.12f 3.8 Group Point Total: fl

ES-401 System # I Name I

201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 21 5001 Traversing In-core Probe 11 21 5002 RBM 21 6000 Nuclear Boiler Inst.

21 9000 RHWLPCI:

TorudPool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHWLPCI: CTMT 230000 RHWLPCI:

Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSlV Leakage II Control 241 000 Reactor/Turbine 245000 Main Turbine Gen. I 11 Aux.

256000 Reactor Condensate 9of34 NUREG-1 021, Draft Revision 9

ES-401 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000. Radiation Monitorina 286000 Fire Protection I 288000 Plant Ventilation 290002 Reactor Vessel lnternals WA Category Point Totals:

A2. Abili to a) redictthe the FIRE PROTECTIO~I SYSTEM

and (b) based on those predictions, useprocedures to correct, control, or mitlgate the consequences of those abnonnal conditions or operations

(CFR: 41.5 /45.6)

A2.03 A.C. distribution failure:

Plant-Specific impactso 7 the 6s o owin on I

3.0 2.2.32 Knowledge of the effects of alterations on core configuration.

3.3 (CFR: 43.6)

I Group Point Total:

ES-401 11 of34 NUREG-1021, Draft Revision 9

ES-40 1 b

ES-401

~

WA #

2.1.14 2.1.

2.1.

Date of Exam:

I I

Topic RO I SRC 1

I IR IR

~~

Knowledge of system status criteria which 1

requires notification of plant personnel (CFR 43.5/45.12)

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-53 Facility:

Description:

Knowledge of the reasons for Emergency Depressurization as it applies to High Secondary Containment Radiation Levels Question Source:

Bank #

Modified Bank #

X (Note changes or attached parent}

New Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehensive or Analysis 55.43 I O CFR Part 55 Content:

55.41 X

Comments: Changed KIA based on oversampiing of SGTS. Modified 1996 Quad Cities.

QUESTION #4 In procedure 205.0, Reactor Refueling, the operator is directed to IMMEDIATELY EVACUATE the area if refuel floor radiation levels begin to increase following a drop of a fuel assembly from the refueling equipment.

What is the basis for evacuating?

A.

The direct radiation from the dropped fuel bundle could cause unplanned radiation exposure.

B.

The dropped fuel bundle may breach the cavity seal and drain the cavity.

C.

The dropped bundle may release radioactive gasses that could will cause unplanned radiation exposure.

D.

The dropped bundle may create a criticality event in the reactor.

ANSWER:

C U(PLANATI0N:

This is the classic fuel handling accident in which the fission product gasses contained within the fuel cladding are released and bubble up through the water. Personnel remaining over the cavity can be exposed to doses approximating 1 OCFRI 00 limits. The approximately 25 feet of water provide adequate shielding from direct radiation and allow time for action to restore level.

Criticality is avoided by refueling interlocks and refueling patterns, though answer D could be considered a plausible distractor for someone unfamiliar with refueling interlocks and core design.

TECHNICAL REFRENCE(S):

RG 1.25: Procedure 205.0 DU 5.0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective:

(As available)

Examination Outline Cross-reference:

Level RO SRO -

1 Tier #

2 Group #

K/A #

295034/EKl. O T Importance Rating 3.8 WA Topic

Description:

Knowledge of the operational implications of Personnel Protection as it applies to Secondary Containment High Radiation Question Source:

Bank #

Modified Bank #

New X

(Note changes or attached parent)

Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehensive or Analysis 55.43 10 CFR Part 55 Content:

55.41 X

Comments:

QUESTION #23 Given the following plant conditions:

0 AOG is in service 0

Stack Effluent HI alarm 0

Reactor is at 100% power Main Steam Line Radiation Monitors all at approximately 550 mrhr Reactor Bldg Vent Radiation at 8 mrhr RCS activity at 90% of TS limit 6" IC isolated for maintenance Significant/visible packing leak from "A" IC outboard steam isolation valve 0

0 0

0 a

NO leaks in the "A" IC tube bundle What action(s) would result in having the greatest reduction in the thyroid damage for the public?

A.

Close "A" IC outboard steam isolation valve B.

Reduce reactor power until stack effluent HI alarm clears C.

Start SGTS and shutdown Reactor Building HVAC D.

Close "A" IC vent valve ANSWER:

C EXPLAN AT1 0 N :

Starting SGTS is the only action that will remove radioactive iodine being released from the steam leak. The AOG will remove all iodine from the off gas regardless of reactor power so reducing power will not result in a reduction in iodine.

TECHNICAL REFERENCE(S):

(Attach if not previously provided)

Proposed references to be provided to applicants during examination:

None Learning Objective:

(As available)

Examination Outline Cross-reference:

Level RO SRO Tier #

i Group #

KIA #

295038/EKlx 1

Importance Rating 2.5 WA Topic

Description:

Knowledge of the operational implications of the biological effects of radioactive ingestion as it applies to Off Gas Release rate.

Question Source:

Bank #

Question Cognitive Level:

Memory or Fundamental Knowledge -

Modified Bank #

New X

(Note changes or attached parent)

Comprehensive or Analysis X

QUESTION #31 A reactor startup is in progress with the following conditions:

0 The previous shift had increased power from 60% to 80%.

Your shift just increased power from 80% to 90% power when all operable Main Steam Line Rad Monitors are verified to read greater than 550 mr/hr but less than 800 rnrhr.

Within one minute, area radiation monitors in the vicinity of the Moisture Separators experience at least a doubling of their readings.

The Off Gas monitors are observed to be unaffected.

What actions are required to mitigate the consequences of these abnormal conditions?

A.

SCRAM the reactor in accordance with ABN-1, Reactor Scram and CLOSE MSlVs and IC Vents.

B.

Reduce Hydrogen Injection flow to between 5 and 6 scfm and monitor the effect on Main Steam Line Activity for the next 10 minutes.

C.

Direct Chemistry to sample Off Gas (Technical Specification 4.6.E) and request guidance from Reactor Engineering.

D.

Start plant shutdown in accordance with Procedure 203, Plant Shutdown ANSWER:

B EXPLANATION:

Since the radiation increase is not observed in the off Gas System it can be concluded that the radiation is from an isotope with a very short half life. Hydrogen Injection produces, almost exclusively, N16 with a half life of approximately 7 seconds. Per "Plant Startup" procedure Hydrogen Injection is increased in 10% power increments and, given that there has been no increase in Off Gas activity it is reasonable to assume the increase is attributed to hydrogen Injection. The reference procedure has the operator first reduce Hydrogen Injection before taking other action. If the MSL radiation decreases in response to reducing Hydrogen Injection then only monitoring of Off Gas is required. The other actions are specified if the HI MSL condition is attributed to fuel failure.

TECHNICAL REFERENCE(S):

ABN-26 Rev 0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

ABN-26 Learning Objective:

(As available}

Examination Outline Cross-reference:

Level RO SRO -

2 Tier #

2 Group #

WA #

272000/A2.0r Importance Rating 2.6

QUESTION # 64 You are about to sign on to an RWP to operate a valve in the Radwaste Building and notice there is no requirement to wear a respirator. You recall from your tour this morning that the area you will enter is posted as a High Airborne Activity Area. You ask the HP tech about this and are told it is consistent with ALARA to NOT wear a respirator.

Why would NOT wearing a respirator in a High Airborne Activity Area be consistent with ALARA?

A.

The respirator is only effective on particulates so there will be NO difference in TEDE.

B.

The respirator will limit your vision and may be a safety hazard in "tight" spaces.

C.

Wearing a respirator may increase your stay time and actually increase TEDE.

D.

The Airborne Activity is short lived and the respirator will have NO impact on ALARA.

ANSWER:

C EXPIAN ATION:

TEDE includes internal and external dose. If the HP calculates TEDE to be lower without a respirator then that is consistent with ALARA.

TECHNICAL REFERENCE(S):

(Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective:

(As available)

Examination Outline Cross-reference:

Level RO SRO Tier #

3 3

Group #

Importance Rating 2.5 WA #

2.3.2 WA Topic

Description:

Knowledge of the facility AMRA program.

Question Source:

Bank #

Modified Bank #

New X

(Note changes or attached parent)

Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehensive or Analysis 10 CFR Part 55 Content:

55.41 X

55.43 x

Comments: This goes beyond what is taught in GET.

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