Text

Draft - Outlines (Folder 2)
	ML053330521
Person / Time
Site:	Hope Creek
Issue date:	06/02/2005
From:	Suzanne Dennis; Operations Branch I
To:	Reid J; Public Service Enterprise Group
	Conte R
Shared Package
ML050960050	List: ML053330006; ML053330521; ML053570219; ML053570230; ML053570233;
References
	ES-201, ES-201-2 IR-05-301
	Download: ML053330521 (193)
	v • d • e

Es-201 Examination Outline Quality Checklist Form ES-201-2

t.

11 I d. Assess whether the justifications for deselected or rejected WA statements are appropriate. I@&[

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a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, 11 I

and major transients.

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b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

3.

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a. Verify that the systems walk-throughoutline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control mom and in-plant tasks distributed among the safety functions as specified on the form (2) task repetition from the last two NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of m o r modified tasks meets or exceeds the minimums specmed on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria 11 a. Author Date W 0 5

b. Facility Reviewer (*)

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v C. NRC Chief Examiner (#)

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d. NRC Supervisor IT, 7 r,, h 7 &-&GA-Note:

Independent NRC reviewer initial items in Column "c"; chief examiner concurrence required.

ES-201 Examination Outline Quality Checklist Form ES-201-2 11 I d.Assess whether the justifications for deselected or rejected WA statements are appropriate.

Facility: Hope Creek - SRO Only Exam Date of Examination:

11/28/05 on the form.

b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is repeated from the last two NRC licensing examinations

c. Determine if there are enough different outlines to test the prqected number and mix of applicants and ensure that no items are duplicated on subsequent days.

a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam sections.

b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.

c. Ensure that WA importance ratings (except for plant-specific priorities) are at least 2.5.

d. Check for duplication and overlap among exam sections.

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4.

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e. Check the entire exam for balance of coverage.

' f. Assess whether the exam fits the appropriate job level (RO or SRO).

Item

1.

W R

I T

Task Description

a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401.

b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.1 of ES-401 and whether all WA categories are appropriately sampled.

c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

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2.

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3.

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a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients.

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b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

c. To the extent possible, assess whether the outfine(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

~~

a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks distributed among the safety functions as specified on the form (2) task repetitii from the last tvro NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of new or modified tasks meets or exceeds the miniiwns specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria P r i n p q g n a t u r @,&&+-\\

a. Author Michael L. Brown I

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b. Facility Reviewer r) 9 I

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C. NRCChief Examiner(#) S w f i f%d,4/

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d. NRC Supervisor K tx cw2<- iw Initials E

Date 6/2/05 %

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Note:

Independent NRC reviewer initial items in Column 'c";;hief examiner concurrence required.

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ES-401 BWR Examination Outline Form ES-401-1 Facility: Hope Creek - SRO Only Exam Date of Exam: 11/28/05 I

I II Note:

1, Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the 'Tier Totals" in each WA category shall not be less than two).

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 i 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5.

Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

8.

On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

2.

3.

9.

ES-401 2

Form ES-401-1 WA Topic(s)

~

ES-401 WAPE # I Name I Safety Function K

K K

1 2

3 295001 Partial or Complete Loss of Forced 0 0 0 Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 0 0 0 295004 Partial or Total Loss of DC Pwr I 6 0 0 0 IR 295028 High Drywell Temperature 15 AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41.IO/ 4 3 2 AA2.04 - Ability to determine and interpret the following as they apply to Partial or Total loss of 45.12)

DC power:(CFR: 41.10 J 43.5 / 45.13) - System Lineups WA Randomly Rejected AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41. 10/ 43.2,'

45.12)

~~

295030 Low Suppression Pool Wtr L i / 5-1 0 I 0 I 3.8 1

3.3 1

3.8 1

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~~

BWR Examination Outline Form ES-401-i AA2.02 - Ability to determine and interpret the following as they apply to Partial or Total loss of Instrument Air:(CFR: 41.10/43,5/ 45.13) - Status of safety-related instrument air system loads (see AK2.1 - AK2.19) 3.7 1

EG2.4.50 - Ability to verify system alarm setpoints and operate controls identified in the EA2.01 - Ability to determine and interpret the following as they apply to Low alarm response manual. (CFR 45.3) 3.3 1

4.2 1

295031 Reactor Low Water Level I2 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 600000 Plant Fire On Site / 8 WA Category Totals:

Form ES-401-1 ES-401 4

ES-401 E/APE # / Name / Safety Function 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5

~~

29501 1 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure I 5 BWR Examination Outline Form ES-401-1 3nergency and Abnormal Plant Evolutions - Tier l/Group 2 (RO / SRO)

A G

WA Topic(s)

IR n

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0 0

I I

I 1 0 AA2.02 - Ability to determine and interpret the following as they apply to Low Reactor Water 3.7 1

0 1 AG2.4.6 - Knowledge of symptom based EOP mitigation strategies (CFR: 41.10 / 43.5 /

4.0 1

Level (CFR: 41.lo/ 43.5 I45.13) - Steam flow/ feed flow mismatch I

I 45.13)

I I

0 0 0 0 0 0 0

0 0

1 EG2.4.6 - Knowledge of symptom based EOP mitigation strategies (CFR: 41.10 / 43.5 /

4.0 1

45.13)

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ES-401 6

Form ES-401-1 ES-401 BWR Examination Outline Plant Systems - Tier 2/Group 1 (RO I SRO)

System # I Name K

K K

K A

A A

A G

WA Topic(s) 1 2

3 4

5 6

1 2

3 4

203000RHWLPCI: Injection 0 0 0 0 0 0 0 0 0 0 0 Mode 205000 Shutdown Cooling 0

0 0

206000 HPCl 0 0 0 0 0 0 0 0 0 0 I

G2.1.14 - Knowledge of system status criteria which require the notification of plant personnel. (CFR: 43.5 145.12) 207000 Isolation (Emergency) 0 0 0 0 0 0 0 0 0 0 0 Condenser 209001 LPCS 0 0 0 0 0 0 0 1 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the LPCS and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.51 43.51 45.31 45.13) - Valve closures 209002 HPCS 0

0 0

21 1000 SLC 0

0 0

21 2000 RPS 0 0 0 0 0 0 0 - 0 0 0 0 21 5003 IRM 0

0 0

215004SourceRangeMonitor 0 0 0 0 0 0 0 0 0 0 0 215005 APRM I LPRM 0 0 0 0 0 0 0 1 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the APRMI LPRM and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.51 43.51 45.31 45.13) - Upscale or downscale trips.

217000 RClC 0

0 0

21 8000 ADS 0

0 0

223002 PClSlNuclearSteam 0 0 0 0 0 0 0 0 0 0 0 Supply Shutoff 239002 SRVS 0

0 0

259002 Reactor Water Level 0 0 0 0 0 0 0 0 0 0 1 G2.1.33 - Ability to recognize indications for system operating parameters which are Control entry-level conditions for technical specifications. (CFR: 43.2 I 43.3 / 45.3)

Form ES-401-1 IR 3.3 1

3.2 1

3.7 1

4.0 1

261 000 SGTS 0

0 262001 AC Electrical 0

0 Distribution 262002 UPS (AC/DC) 0 0

263000 DC Electrical 0

0 Distribution 264000 EDGs 0

0 300000 Instrument Air Water WA Category Point Totals:

I 3

2 Group Point Total:

5

ES-401 8

Form ES-401-1 ES-401 System # I Name

~~

201001 CRD Hydraulic 201002 RMCS 201 003 Control Rod and Drive Mechanism 201004 RSCS 201005 RClS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPlS 21 5001 Traversing In-core Probe 21 5002 RBM 216000 Nuclear Boiler lnst 219000 RHWLPCI TorusIPool Cooling Mode 223001 Primary CTMT and Aux 226001 RHWLPCI CTMT Spray Mode 230000 RHWLPCI Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSlV Leakage Control 241000 Reactornurbine Pressure Reaulator BWR Examination Outline Form ES-401-1

?ant Systems - Tier ZGroup 2 (RO I SRO)

G KIA Topic(s)

IR 1

G2 1 28 - Knowledge of the purpose and function of major system components and controls 3 3 1

1 G2 1 23 -Ability to perform specific system and integrated plant procedures during all modes 4 0 1

of plant operation (CFR 45 2 145 6)

I I

I 0

0 I

1 I

0 0

0 t

o

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Hope Creek Date of Examination:

11/28/05 Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Describe activity to be performed Type Code*

s, A, N R, N D, s, A R, N, A Check Drywell to Torus D/P during power operations per Daily Surveillance Log Procedure Change - Make a change to a procedure for Emergent work Rod Worth Minimizer Operability -

Enter and exit a High Radiation Area for a valve lineup. 2.3.10

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

Form ES-401-3

4.

Emergency Procedures/

Plan Facility: Hope Creek - SRO Only Exam Date of Exam:

1 1/28/05 of pre-and post-maintenance operability requirements 2.4.22 2.4.36 Knowledge of the bases for prioritizing safety functions during abnormaVemergency operations (CFR: 43.5 / 45.1 2)

Knowledge of chemistryhealth physics tasks during emergency operations (CFR: 43.5)

I Subtotal I

I SRO-On1

l 1 4.4 2.9 1

2 3.3 1

3.5 1

2 3.1 1

4.0 I

2.8 I

1

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

Form ES-401-3 RO Facility: Hope Category SRO-Onlv

1.

Conduct of Operations IR 3.1

2.

Equipment Control IR 1

3.

Radiation Control 2.3.1 2.3.1 0 2.3.

reek - RO WA #

Knowledge of 10 CFR 20 and related facility radiation control requirements (CFR: 41.12 / 43.4. 45.9 / 45.10).

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure (CFR: 43.4 / 45.1 0) 2.1.21 2.1.14 2.1.33 2.1.

2.1.

Subtotal 2.2.1 2.2.34 2.2.

Subtotal Kam Date of Exam:

1 1 /28/05 Topic Ability to obtain and verify controlled procedure copy (CFR: 45.1 0 /

45.1 3)

Knowledge of system status criteria which require the notification of plant personnel (CFR: 43.5 / 45.1 2)

Ability to recognize indications for system operating parameters which are entry-level condition for Technical Specifications (CFR: 43.2 / 43.3

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Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

(CFR: 45.1)

Knowledge of the process for determining the internal and external effects on core reactivity (CFR: 43.6)

Subtotal I

2.6 I

I 2.9 I

1 I

4.

Emergency Procedures/

Plan I

I I

2.4.27 2.4.39 2.4.31 Knowledge of fire in the plant procedure (CFR: 41.10 /43.5 /45.13)

Knowledge of the RO's responsibilities in emergency plan implementation (CFR: 45.1 1)

Knowledge of annunciators alarms and indications / and use of the response instructions. (CFR: 41.1 0 / 45.3) 3.0 3.3 I 2.4.

i I

I I

I 1

1

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ES-401 BWR Examination Outline Form ES-401-1 Note:

1.

Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the Tier Totals" in each WA category shall not be less than two).

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 *1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

e.

Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

8.

On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

2.

3.

1

9. For Tier 3, select topics from Section 2 of the WA catalog, and enter the K/A numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1IGroup 1 (RO I SRO)

E/APE # I Name 1 Safety Function K

K K

A A

G WA Topic(s)

IR 1

2 3

1 2

295001 Partialor CompleteLossof Forced 0 0 0 0 0 0 Core Flow Circulation I 1 & 4 295003 Partial or Complete Loss of AC I 6 0 0 0 0 0 1 AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41.101 43.U 3.8 295004 Partial or Total LOSS of DC Pwr I 6

0 0 0 0 1 0 AA2.04 - Ability to determine and interpret the following as they apply Io Partial or Total loss of 3.3 45.12)

DC power:(CFR: 41.10343.5 /45.13) - System Lineups 295005 Main Turbine Generator Trip I 3 0 0 0 0 0 0 WARandomlyRejected 295006 SCRAM I 1

0 0 0 0 0 1 AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41.101 43.U 3.8 45.12) 295016 Control Room Abandonment / 7 29501 8 Partial or Total Loss of CCW I 8

0 0

295019 Partial or Total Loss of Inst. Air I 8 0 0 0 0 1 0 AA2.02 - Ability to determine and interpret the following as they apply to Partial or Total 3.7 loss of Instrument Air:(CFR: 41,10/43,5/ 45.13) - Status of safety-related instrument air system loads (see AK2.1 - AK2.19) 295021 Loss of Shutdown Cooling 1 4 295023 Refueling Acc 18 0

0 0

0 295024 High Drywell Pressure I 5

295025 High Reactor Pressure I 3 295026 Suppression Pool High Water 0

0 0

0 Temp. 1 5 295027 High Containment TemperatureI5 0 0 0 0 0 0 295028 High Drywell Temperature I 5 0 0 0 0 0 1 EG2.4.50 - Ability to verify system alarm setpoints and operate controls identified in the 3.3 295030 Low Suppression Pool Wtr Lvl I 5 0 0 0 0 1 0 EA2.01 - Ability to determine and interpret the following as they apply to Low 4.2 alarm response manual. (CFR 45.3)

SuDpression Pool Water level (CFR:41.lo/ 43.5I 45.13) - SuoDression Pool level 1

1 1

1

I I

295031 Reactor Low Water Level I 2 0

295037 SCRAM Condition Present 0

and Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 0

1 0 600000 Plant Fire On Site / 8 AA2.13 - Ability to determine and interpret the following as they apply to Plant Fire On 3.8 1

Site: (CFR:41.lo/ 43.5/ 45.13) - Need for emergency plant shutdown I o 4 3 WA Category Totals:

I Group Point Total:

7 O 1 O 1 I

II

ES-401 4

Form ES-401-1 A

A G

WA Topic@)

IR 1

2 ES-401 UAPE # / Name / Safety Function K

K K

295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure 13 295010 High Drywell Pressure / 5 0

0 0

0 0 0 0

0 0

295034 Secondary Containment 295035 Secondary Containment High o

o a

Differential Pressure / 5 0

0 0

0 0 1 0 AA2.02 - Ability to determine and interpret the following as they apply to Low Reactor Water Level (CFR: 41.lo/ 43.5 I45.13) - Steam flow/ feed flow mismatch 3.7 1

0 0 1 AG2.4.6 - Knowledge of symptom based EOP mitigation strategies (CFR: 41.10 / 43.5 /

45.13) 4.0 1

0 0

0 0 0 0 0 1 EG2.4.6 - Knowledge of symptom based EOP mitigation strategies (CFR: 41.10 143.5 /

45.13) 0 0 0 0 0 0 4.0 1

0 0 0 0

0 0

I I

O 1 O 1 O1

ES-401 6

Form ES-401-1 ES-401 BWR Examination Outline Plant Systems - Tier YGroup 1 (RO I SRO)

System # I Name K

K K

K A

A A

A G

WA Topic(s)

IR 1

2 3

4 5

6 1

2 3

4 203000RHWLPCI: Injection 0 0 0 0

0 0

0 0 0 0 0 Mode 205000 Shutdown Cooling 0

0 0

206000 HPCl 0 0 0 0 0 0 0 0 0 0 1 G2.1.14 - Knowledge of system status criteria which require the notification of plant 3.3 personnel. (CFR: 43.5 / 45.12) 207000 Isolation (Emergency) 0 0 0 0 0 0 0 0 0 0 0 Condenser 209001 LPCS 0 0 0 0 0 0 0

1 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the LPCS and (b) based 3.2 on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.51 43.5/ 45.3/ 45.13) - Valve closures 209002 HPCS 0

0 0

21 1000 SLC 0

0 0

212000 RPS 0 0 0 0 0 0 0 0 0 0 0

~

215003 IRM 0

0 0

215004 Source Range Monitor 0 0 0 0 0 0 0 0 0 0 0 215005 APRM I LPRM 0 0 0 0 0 0 0 1 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the APRMI LPRM and 3.7 (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.5/ 43.51 45.31 45.13) - Upscale or downscale trips.

217000 RClC 0

0 0

21 8000 ADS 0

0 0

223002 PCIS/NuclearSteam 0 0 0 0 0 0 0 0 0 0 0 Supply Shutoff 239002 SRVs 0

0 0

259002 Reactor Water Level 0 0 0 0 0 0 0 0 0 0 1 G2.1.33 - Ability to recognize indications for system operating parameters which are 4.0 Control entrv-level conditions for technical sDecifications. (CFR: 43.2 / 43.3 I 45.3) 1 1

1 1

261000 SGTS 0

0 0

262001 AC Electrical 0

0 0

Distribution 262002 UPS (ACIDC) 0 0

0 0

0 263000 DC Electrical 0

0 0

Distribution 264000 EDGs 0 0 0 0 0 0 0 1 0 0 0 A2.08 - Ability to (a) predict the impacts of the following on the EDGs and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.W 43.5/ 45.3/ 45.13) - Initiation of emergency generator room fire protection system.

300000 Instrument Air 0

0 0

400000ComponentCooling 0 0 0 0 0 0 0 0 0 0 0 Water 3.7 1

WA Category Point Totals:

3 2 Group Point Total:

5

BWR Examination Outline Form ES-401 Plant Systems - Tier ZGroup 2 (RO / SRO)

IR System # I Name K

K K

K A

A A

A G

WA Topic(s) 201001 CRD Hydraulic 1

2 3

4 5

6 1

2 3

4 0

0 0 0 0

0 I G2 1 28 - Knowledge 01 the purpose and function of major system components and controls 3 3 (CFR 41 7) 201 002 RMCS 0

0 0

201003 Control Rod and Drive 0

0 0

Mechanism 201 004 RSCS 0

0 0

201 005 RClS 0

0 0

2010% RWM 0

0 0

202001 Recirculation 0

0 0

0 0 0

0 0 1

G2 1 23 -Ability to perform specific system and integrated plant procedures during all modes 4 0 of plant operation (CFR 45 2 145 6) 202002 Recirculation Flow Control 0

0 0

204000 RWCU 0

0 0

214OOO RPlS 0

0 0

21 5001 Traversing In-core Probe 0

0 0

21 5002 RBM 0

0 0

216000 Nuclear Boiler lnst 0

0 0

219000RHRRPCl TorudPoolCooling 0

0 0

0 0 0 0

0 0

0 0 Mode 223001 Primary CTMT and Aux 0

0 0

226001 RHWLPCI CTMTSprayMode 0

0 0

230000RHWLPCl TorudPoolSpray 0

0 0

0 0 0 0 0 0

0 0 Mode

~~

233000 Fuel Pool Cwling/Cleanup 0

0 0

234000 Fuel Handling Equipment 0

0 0

239001 Main and Reheat Steam 0

0 0

239003 MSlV Leakage Control 0

0 0

241000 Reactornurbine Pressure 0

0 0

Reaulator ES-401-1 1

1

245000 Main Turbine Gen. I Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel lnternals WA Category Point Totals:

3.8 A2.05 - Ability to (a) predict the impacts of the following on the Main Turbine Gen. / Aux and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.51 43.5/ 45.3 45.13) - Generator trip I

1 Group Point Total:

1 3

0 1

0 0

0

ES-401 BWR Examination Outline Form ES-401-1 Note:

1.

Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each WA category shall not be less than two).

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 *1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

Systems/evolutions within each group are identiied on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

e.

Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

8. On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

2.

3.

9. For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

Form ES-401-1 ES-401 2

' 295023 Refueling Acc / 8 I

ES-401

~,

I 295024 High Drywell Pressure / 5 I

' 295025 High Reactor Pressure / 3 WAPE # / Name / Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 BWR Examination Outline sency and Abnormal Plant Evolutions - Tier l/Group 1 (RO / SRO)

A A

G WA Topic($

IR 0 0 0 AK1.03 - Knowledge of the operational implications of the following concepts as they 3.6 1

2 apply to the Partial or Complete Loss of Forced Core Flow Circulation: Thermal Limits

(CFR: 41.8 to 41.10 / 45.3) or Complete Loss of AC : Whether a partial or complete loss of A.C. Power has occurred:(CFR: 41.10 /43.5/ 45.13)

Partial or Total Loss of DC Pwr : Load shedding Plant Specific:(CFR: 41.5/41.10/45.6 0

1 0 AA2.05 - Ability to determine and interpret the following as they apply to Partial 3.9 0 0 0 AK3.01 - Knowledge of the reasons for the following responses as they apply to 2.6 145.13) 0 0 1 AG2.1.2 - Knowledge of operator responsibilities during all modes of plant operation 3.0 0 0 0 AK1.03 - Knowledge of the operational implications of the fallowing concepts as they 3.7 0 0 1 AG2.1.30 - Ability to locate and operate components, including local controls. (CFR:

3.9 0

1 0 AA2.04 - Ability to determine and interpret the following as they apply to Partial 2.9 AA1.03 - Ability to operate and I or monitor the following as they apply to Partial or (CFR: 41.10 145.13) apply to the SCRAM: Reactivity Control:(CFR: 41.8 to 41.10 /45.3) 41.7 145.7) or Total Loss of CCW System Flow:(CFR: 41.10/43.5/ 45.13) 1 0 0 Total Loss of Inst. Air: Instrument Air Compressor Power supplies:(CFR:

41.71 45.5/45.6) 3.0 0 1 0 AA2.05 - Ability to determine and interpret the following as they apply to Loss of 3.4 0 0 0 AK2.03 - Knowledge of the interrelations between Refueling Accidents and the 3.4 1 0 0 EA1.03 - Ability to operate and/ or monitor the following as they apply to High Drywell 4.0 1 0 0 EA1.02 - Ability to operate and I or monitor the following as they apply to High 3.8 Shutdown Cooling: Reactor Vessel Metal Temperature (CFR: 41.1 0 /43.5/45.13) following: Radiation Monitoring equipment (CFR41.7 /45.7/ 45.8)

Pressure: LPCS - Plant specific (CFR41.7/ 45.51 45.6)

Reactor Pressure : Reactornurbine pressure regulating system :(CFR: 41.7/45.5/

45.61 295004 Partial or Total Loss of DC Pwr / 6 1

1 1

1 295005 Main Turbine Generator Trip / 3 295006 SCRAM / 1 29501 6 Control Room Abandonment / 7 295018 Partial or Total Loss of CCW I 8 295019 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling / 4

295026 Suppression Pool High Water Temp. I 5 295027 High Containment Temperature / 5 295028 High Drywell Temperature / 5 295030 Low Suppression Pool Wtr Lvl I 5 0

295031 Reactor Low Water Level / 2 1

EG2.1.23 - Ability to perform specific system and integrated plant procedures during 3.9 1

all modes of plant operation. (CFR: 45.2 145.6)

~~

~

295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown I 1 295038 High Off-site Release Rate I 9 0

600000 Plant Fire On Site I 8 0

0 0

1 EG2.1.30 - Ability to locate and operate components, including local controls. (CFR:

3.9 1

41.7 145.7)

, 295005 Main Turbine Generator Trip 13 0

I KIA Category Totals:

0 EK3.07 - Knowledge of the reasons for the following responses as they apply to Low 3.5 1

Suppression Pool Wtr Lvl: NPSH considerations for ECCS pumps:(CFR:

41.5141.1 0145.61 45.1 3)

I I

0 0 EK2.10 - Knowledge of the interrelations between Reactor Low Water Level and the 4.0 1

foliowing: Redundant reactivity control: Plant specific (CFR: 41.7/45.7/45.8) 0 0

0 0 EA1.02 - Ability to operate and I or monitor the following as they apply to SCRAM 3.8 1

Condition Present and Reactor Power Above APRM Downscale or Unknown: RRCS:

Plant Specific (CFR: 41.7145.51 45.6) 0 EK3.02 - Knowledge of the reasons for the following responses as they apply to High AK1.01 - Knowledge of the operational implications of the following concepts as they 3.9 1

Off-site Release Rate: System Isolations :(CFR: 41.5/41.10/45.6/ 45.13) apply to the Plant Fire On Site: Fire Classifications by type (CFR: 41.8 to 41.10 145.3) 0 2.5 1

0 0 AK2.04 Knowledge of the interrelations between MAIN TURBINE 3.3 1

GENERATOR TRJP and the following: Main generator protection (CFR: 41.7 145.8)

I I,

I I

3 4 Group Point Total:

20

4 Form ES-401-1 ES-401 OAPE # / Name / Safety Function 295009 Low Reactor Water Level 12 Area Radiation Levels I 9

295034 Secondary Containment Ventilation High Radiation I 9 0 0 0 1 0 0 EA1.01 - Ability to operate and/ or monitor the following as they apply to Secondary Containment Ventilation High Radiation: Area radiation monitoring system: (CFR41.7145.5l45.6) 295035 Secondary Containment High 0

0 0

0 Differential Pressure I 5 1

3.8 0

ES-401 6

Form ES-401-1 Plant Systems - Tier 2IGroup 1 (RO I SRO)

System # / Name K

K K

K A

A A

A G

WA Topic@)

IR 203000 RHWLPCI: Injection 0 0 0 0 0 0 0 0 0 0 1 G2.2.25 - Knowledge of bases in technical specifications for limiting conditions 2.5 1

2 3

4 5

6 1

2 3

4 Mode for operations and safely limits (CFR: 43.2) 205000 Shutdown Cooling 0 0 0 0 0 0 0 0 1 0 0 A3.03 - Ability to monitor automatic operations of the Shutdown Cooling 3.5 System(RHR Shutdown Cooling Mode) including: lights and alarms (CFR:41.7/45.5) 206000 HPCI 0 0 0 0 0 0 0 1 0 0 0 A2.07 - Ability to (a) predict the impacts of the following on the HPCl and (b) 3.4 based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: Low suppression pool level: BWRQ, 3, 4 (CFR:41.5/43.5/45.3145.13) 206000 HPCl 0 0 0 0 1 0 0 0 0 0 0 K5.05 - Knowledge of the operational implications of the following concepts as 3.3 they apply to the HPCI: Turbine speed control: BWR-2,3,4 (CFR:41.5/ 45.7) 207000 lsolation(Emergency) 0 0 0 0 0 0 0 0 0 0 0 Condenser 209001 LPCS 0 1 0 0 0 0 0 0 0 0 0 K2.01 - Knowledge of electrical power supplies to the following: Pump power 3.0 (CFR41.7) 209002 HPCS 0

0 0

21 1000 SLC 0 0 0 1 0 0 0 0 0 0 0 K4.04 - Knowledge of SLC design feature(s) and or interlock(sj which provide 3.8 212000 RPS 0 0 1 0 0 0 0 0 0 0 0 K3.11 -Knowledge of the effect that a loss or malfunction of the RPS will have 3.0 215003 IRM 0 0 0 1 0 0 0 0 0 0 0 K4.04 - Knowledge of the IRM design feature(s) and or interlock(s) which 2.9 for the following: Indication of fault in explosive valve firing circuits (CFR41.7) on the following: Recirculation system (CFR41.7/45.6) provide for the following: Varying system sensitivity levels using range switches (CFR41.7) 215003 IRM 0

1 0 0 0 0 0 0 0 0 0 K2.01 - Knowledge of electrical power supplies to the following: IRM Channels1 2.5 215004 Source Range Monitor 1 0 0 0 0 0 0 0 0 0 0 K1.02-Knowledge of the physical connections and/or cause-effect 3.4 detectors (CFR41.7) relationships between Source Range Monitor and the following: Reactor Manual Control (CFFt41.2 to 41.9145.7 to 45.8)

I 1

1 1

1

~~

I 0 215005 APRM I LPRM 3.2 G2.1.28 - Knowledge of the purposes and function of major system components and controls (CFR: 41.7)

K1.05 - Knowledge of the physical connections and/or cause-effect relationships between RClC and the following: Residual Heat Removal System (CFR:41.2 to 41.9/ 45.7 to 45.8)

A2.04 - Ability to (a) predict the impacts of the following on the ADS and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal operation: ADS failure to initiate (CFR 41.5143.5/45.3/45.13)

A4.02 - Ability to manually operate and/or monitor in the control room: Manually initiate the system (CFR:41.7/45.5 to 45.8)

A4.06-Ability to manually operate and/or monitor in the control room: Reactor water level (CFR: 41.7/45.5 to 45.8)

K3.06 - Knowledge of the effect that a loss or malfunction of the Reactor Water Level Control will have on the following: Main Turbine (CFR:41.7/45.6)

K3.02 - Knowledge of the effect that a loss or malfunction of the SGTS will have on the following: Off-site release rate (CFR:41.7/45.6)

K4.03 - Knowledge of AC Electrical distribution design feature(s) and or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers (CFR41.7)

K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the UPS (ACIDC): DC electrical power (CFR:41.7/45.7)

Al.01 - Ability to predict and/or monitor changes in parameters associated with operating the DC Electrical distribution controls including: Battery chargingdischarging rate (CFR:41.5/45.5)

G2.1.14 - Knowledge of system status criteria which require notification of plant personnel (CFR: 43.5 I45.12)

A3.02 - Ability to monitor automatic operations of the Instrument Air including:

Air temperature (CFR 41.7/45.5)

K5.01 - Knowledge of the operational implications of the following concepts as they apply to the Instrument Air: Air Compressors (CFR:41.5/ 45.7)

K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the Commnent Coolina Water: Valves (CFR:4 1.5145.5)

I 2.6 1 -

1 -

1 I1 21 7000 RClC 21 8000 ADS 4.1 a

3.9 223002 PCIS/Nuclear Steam 239002 SRVs 3.9 I

1 -

1 2.8 259002 Reactor Water Level 261 000 SGTS 3.6 3.1 262001 AC Electrical 0

Distribution 262002 UPS (ACIDC) 0 263000 DC Electrical 0

Distribution 2.8 2.5 2.5 264000 EDGs 300000 Instrument Air 2.9 300000 Instrument Air I o

~

2.5 2.7 400000 Component Cooling Water

9 Form ES-401-1 ES-401

239003 MSlV Leakage Control 0

0 241000 Reactormurbine Pressure 0

0 Regulator 245000 Main Turbine Gen. / Aux.

1 0

256000 Reactor Condensate 0

0 259001 Reactor Feedwater 0

0 268000 Radwaste 0

0 271 000 Off gas 0

0 272000 Radiation Monitoring 0

0 286000 Fire Protection 0

1 288000 Plant Ventilation 0

0 290001 Secondary CTMT 0

0 290003 Control Room HVAC 0

0 290002 Reactor Vessel lnternals 1

0 226001 RHWLPCI: CTMT Spray Mode 0

0 lolo 223001 Primary CTMT and Aux.

ll I WA Category Point Totals I 2 1 1

~

II-I I

0 0

0 0 K1.02 - Knowledge of the physical connections and/or cause effect relationships 2.5 1

between Main Turbine Generator / Aux and the lollowing: Condensate system (CFFt41.2 to 41 9 145.7 to 45.8) 0 0

0 0

A2.01 - Ability to (a) predict the impacts of the following on the Radwaste and (b) 2.9 1

based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal operation: System rupture (CFR41.9 43.9 45.3 45.13) 0 0

K3.01 -Knowledge of the effect that a loss or malfunction of 3.6 1

the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: Secondary containment (CFR: 41.7 I45.4) 1 0

Group Point Total:

12

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

Form ES-401-3 I Facility: Hope Category

1.

Conduct of Operations

2.

Equipment Control

3.

Radiation Control Ir i

WA ##

Topic 2.1.21 2.1.14 2.1.33 2.1.

~

Ability to obtain and verify controlled procedure copy (CFR: 45.1 0 /

45.1 3)

Knowledge of system status criteria which require the notification of plant personnel (CFR: 43.5 I 45.1 2)

Ability to recognize indications for system operating parameters which are entry-level condition for Technical Specifications (CFR: 43.2 I 43.3 145.3) 2.2.1 2.2.34 2.2.

Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

(CFR: 45.1)

Knowledge of the process for determining the internal and external effects on core reactivity (CFR: 43.6) fl 2.3.1 2.3.1 0 2.3.

Knowledge of 10 CFR 20 and related facility radiation control requirements (CFR: 41.1 2 143.4. 45.9 I 45.1 0).

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure (CFR: 43.4 145.10)

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

Form ES-401-3 2.1.7 2.1.34 e Creek - SRO Only Exam Date of Exam:

I Category Ability to evaluate plant performance and make operational judgments based on operating characteristics / reactor behavior / and instrument interpretation. (CFR: 43.5 / 45.1 2 / 45.1 3)

Ability to maintain primary and secondary plant chemistry within allowable limits (CFR: 41.10 / 43.5 / 45.12)

1.

Conduct of Operations 2.2.20 2.2.21

2.

Equipment Control Knowledge of the process for managing troubleshooting activities (CFR: 43.5 / 45.1 3)

Knowledge of pre-and post-maintenance operability requirements (CFR: 43.2)

3.

Radiation Control

4.

Emergency Procedures/

Plan 2.3.4 WA# I Knowledge of radiation exposure limits and contamination control /

including permissible levels in excess of those authorized. (CFR: 43.4

/ 45.1 0)

Topic 2.4.22 2.4.36 2.4.

Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations (CFR: 43.5 / 45.1 2)

Knowledge of chemistry/health physics tasks during emergency operations (CFR: 43.5) 11 Tier 3 Point Total 11/28/05 RO SRO-Onlv I

I

ES-401 Record of Reiected WAS Form ES-401-4 i 295027 EK2.01 Tier I Group Tier 1/

Group 1 RO exam Tier 2/

Group 1 RO exam

~ 262002, Al.02 Tier 21 Group 2 RO exam Tier 2/

Group 2 RO exam Tier 3 RO exam

' 21 5002, A4.04

' 223001

, A4.02 Randomly Selected WA

259002, Al.06 G2.2.3 Reason for Rejection WA is for a Mark Ill containment and Hope Creek has a Mark I containment Hope Creek does not have ( W C l ) Feedwater Coolant Injection Not applicable to Hope Creek Not applicable to Hope Creek Not applicable to Hope Creek Not applicable to Hope Creek - Not a Multi-unit facility II I

I

ES-401 Written Examination Quality Checklist Form ES-401-6

8.

9.

RefBrences/handouts provided do not give away answers or aid in the elimination of distractors.

Question content conforms with specific WA statements in the previously approved examination outline and is appropriate for the tier to which they are assigned; deviations are justified.

Facility:

Date of Exam:

Exam Level: RO d R 0 Id I

flo n?9

10.

11. The exam contains the required number of one-point, multiple choice items; the total is correct and agrees with the value on the cover sheet.

Question psychometric quality and format meet the guidelines in ES Appendix B.

WY

&/fl ill F7

a. Author

b. Facility Reviewer (*)

c. NRC Chief Examiner (#)

d. NRC Regional Supervisor n

Note:

The facility reviewer's initialdsignature are not applicable for NRCdeveloped examinations.

Independent NRC reviewer initial items in Column 'c"; chief examiner concurrence required.

Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 p r k ~ 1 Group#

1 1 Imaortance 3.6 Partial or Complete Loss of Forced Core Flow Circulation I 1 8 4 AK1.03 Knowledge of the operational implications of the following concepts as they apply to the Partial or Complete Loss of Forced Core Flow Circulation Thermal Limits :(CFR: 41.8 to 41.10 145.3)

Question Given: Hope Creek was at 100% power when the '6" Recirc pump developed excessive vibration and needed to be tripped.

WHICH ONE of the following actions is REQUIRED to be taken in accordance with either HC.OP-AB.RPV-O003(Q),

Recirculation System or HC.OP-I0.U-0006, Power Changes during Operation?

A The MAPLHGR limits must be reduced.

B The discharge valve HV-F031 B must be closed and maintained closed.

The MCPR safely limit must be reduced.

C Speed control for the operating pump must be placed in Master Manual Control.

D Answer A

References HC.OP-AB.RPV-0003 (Q), Rev. 9, Recirculation System Justification References during Exam None A. is CORRECT per IOP-6, step. 5.3.7

6.

is INCORRECT per RPV-3, Condition A, step A.3 which states that the discharge valve must be closed for approximately 5 minutes and then re-opened.

C. is INCORRECT per IOP-6. step 5.3.7 which states that the MCPR safety limit must be raised D. is INCORRECT per IOP-6, step 5.3.7 which states that speed control must be placed in Local Manual Control HC.OP-IO.ZZ-0006,Rev. 33, Power Changes during Operation

_ _ _ ~ _

Question Source Mod

@I Memory Level 0

Comprehension Level Question History:

SXD review - 7/21/05 - LOD 1.75 perbaps rewrite to make more difficult, removed 'initially" from in front of at 100%

power in stem.

Exam -Cross-Ref Hope Creek RO Exam - Nov 2005 p#

1 Group#

1 I Imvortance 3.9 295003 Partial or Complete Loss of AC I 6 AA2.05 Ability to determine and interpret the following as they apply to Partial or Complete Loss of AC Whether a partial or complete loss of A.C. Power has occuned:(CFR: 41.10 143.51 45.13)

Question Given the following conditions:

.The plant is in Operational Condition 5 with the Electrical Distribution System aligned in the Normal lineup.

.An internal short on Transformer 1 BX-501 causes a sudden pressure fault on the transformer.

Which one of the following describes the resulting availability of power for the Safe Shutdown Systems?

A Power is lost permanently to both 4.16KV switchgear 10A401 and 10A403. 13 KV breakers BS 2-3 and BS 1 2 stay closed. B and D Diesel Generators start but their output breakers DO NOT CLOSE.

B Power is lost momentarily to both 4.16KV switchgear 10A402 and 10A404. 13 KV breakers BS 2-3 and BS 1-2 trip open. Power is restored when the B and D Diesel generators output breakers close.

Power is lost momentarily to both 4.1 6KV switchgear 1 OA402 and 10A404. 13 KV breakers BS 2-3 and BS 1 -

2 trip open. Power is restored when the alternate supply breaker from Transformer 1AX501 closes. B and D diesel generators START but their output breakers DO NOT CLOSE.

~

~~

~

Power is lost momentarily to both 4.16KV switchgear 10A402 and 10A404. 13 KV breakers BS 2-3 and BS 1-2 trip open. Power is restored when the alternate supply breaker from Transformer 1AX501 closes. B and D diesel generators DO NOT START.

~

~~

Answer References Hope Creek Question (276871 - Modified Drawing E-0001 and 066-01 : Class 1 E AC Power Distribution page 32 of 93 NOHOlEAC00 CLASS 1E AC POWER DISTRIBUTION, Justification References during Exam Drawing E-0001 Justification:

Correct answer. 13 Kv Breakers BS 2-3 and BS 1-2 trip open. Bus section 2 is de-energized, Bus section 1 remains energized. The bus infeed breaker swap to the AX501 feed. The loads remain energized. Because one infeed is always available, the Diesels do not start.

A - INCORRECT - Power is not permanently lost to both 4.16KV switchgears. Power is restored when the bus infeed breaker swaps to the AX501 feed.

B - INCORRECT - Power is not restored from the B 8 D Diesel Generators C - INCORRECT - The B & D Diesel Generators DO NOT START Question Source Mod Memory Zmel Comprehension Level

Question History:

SXD Review 7/21/05 - OK

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref 3

Tier#

Group#

1 Importance 2.6 295004 Partial or Total Loss of DC Pwr / 6 AK3.01 Knowledge of the reasons for the following responses as they apply to Partial or Total Loss of DC Pwr Load shedding Plant Specific:(CFR: 41.5/41.10 I 45.6 145.13)

Question With the plant at 100% power, the plant loses power to 125V DC Class 1 E switchgear 10D410.

If the plant were to experience a LOCA, how will Load shedding and control of non-1 E loads be affected:

Load shedding of Non-1 E loads that get control power from 10D410...

A will still occur and these loads can be still be operated from the Control Room (ie. Load shedding and control will not be affected)

B will still occur, however, these loads can NOT be operated from the Control Room.

will not occur, however, these loads can still be operated from the Control Room.

C

~

~~~~

will not occur and these loads can NOT be operated from the Control Room.

D References INPO Question 23597 (somewhat)

Hope Creek Lesson Plan NOH01 EAC00-02, CLASS 1 E AC POWER DISTRIBUTION p34 talks about load shedding of non-1 E loads on a LOCA talks about 125V DC supplying breaker control power Answer NOH01 DCELEC-00, DC ELECTRICAL DISTRIBUTION p.22 Justification References during Exam None A. - CORRECT - loss of DC control power will result in the loss of all remote control to the affected breaker.

6. - INCRRECT - load will NOT auto trip on a Load Shed Signal and CANNOT be operated from the Control Room.

C. - INCORRECT - load will NOT auto trip on a Load Shed signal D. - INCORRECT - load CANNOT be operated from the Control Room Question Source Mod E) Memory Level 0

Comprehension Level

Question History:

SXD Review - 7/21 - Question Stem confusing -

7/27 - Rewrote Question Stem - re-submitted 8/2 JD - Weak Question, doesnt address WA - WA Q about DC load manual shedding to consewe battery life 8/3 - rewrote question again.

295005 AG2.1.2 Question Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 I Tier#

1 Group#

I Importance 3

Main Turbine Generator Trip / 3 Knowledge of operator responsibilities during all modes of plant operation (CFR: 41.10 I45.13)

Due to a main turbine vibration problem with a generator load of 110 MWe, a manual turbine trip is performed.

Which of the following describes when the operator is REQUIRED (Maximum Time Limit) to open the generator Output Breakers for the given conditions? (Assume they have not already tripped on reverse power.)

A Immediately B

Within 15 seconds of the turbine trip Within 60 seconds of the turbine trip C

Within 90 seconds of the turbine trip D

Answer References Hope Creek Question - 053470 HC.OP-SO.AC4OOl(Cl) - Rev. 48, MAIN TURBINE OPERATION - P&L 3.1.1 5 Justification References during Exam None Correct Answer:."15 seconds' -Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power.

The following distractors are incorrect as follows:

'immediately" - Procedure caution calls for operator actions within 15 seconds

'60 seconds' - Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power.

'90 seconds'-Procedure caution calls for operator actions within 15 seconds of the turbine trip at low power. Only when above 150 MWe is the time extended to 90 seconds.

Question Source Bank Memory Level 0

Comprehension Level Question History:

SXD Review - 7/21 - Had question about lower power -

7/27 - verified power level ok per IOP-4 p.15

/Question 5 1 Exam - Cross-Re f Hope Creek RO Exam - Nov 2005 I

0 SRO Importance 3.7 295006 SCRAM I 1 AK1.03 Knowledge of the operational implications of the fallowing concepts as they apply to the SCRAM Reactivity Control:(CFR:

41.8 to 41.10 145.3)

~~

Question A reactor scram has just occurred and the crew is executing HC.OP-AB.=-0000, REACTOR SCRAM.

Which of the following is the reason that step S-8 directs the operator to RESET the scram (SB) if conditions permit AND INSERT a Half-Scram (if Required)?

A B

To reduce the potential for CRD pump runout and reduce the amount of time for the HCU accumulators to recharge.

To restore the CRD hydraulic system to normal for insert and withdrawal capability if rods are found at the 02 or beyond position.

To reestablish the normal primary vessel boundaries by isolating the CRD HCU from the Scram discharge volume (SDV) and closing the SDV vent and drain valves.

To prevent excessive discharge of hot radioactive water to the Reactor Building Equipment Drain Sump.

Answer References Hope Creek Question - Q56128 NOHOlAB0000-01, Reactor Scram AB-0000 p.14 Justification References during Exam None Justification:

A - INCORRECT - To reestablish the normal primary vessel boundaries by isolating the CRD HCU from the scram discharge volume (SDV) and closing the SDV vent and drain valves. Incorrect - the Scram reset will open the vents and drains B - CORRECT - To restore the CRD hydraulic system to normal for insert and withdrawal capability if rods are found at the 02 or beyond position. Correct.

C - INCORRECT -To reduce the potential for CRD pump runout and reduce the amount of time for the HCU accumulators to recharge. Incorrect - system flow restricting orifice limit pump runout to 200 gpm D - INCORRECT - To prevent excessive discharge of hot radioactive water to the Reactor Building Equipment Drain Sump. Incorrect - resetting scram will send water to the Rx Bldg Equipment Drain Sump Question Source Memory Level Comprehension Level Question History:

Submitted 7/22 SXD Reviewed 7/23 - for Distactor C - asked is this verified?

Exam-Cross-Ref O S R O I 295016 AG2.1.30 Hope Creek RO Exam - Nov 2005 I Tier#

Group#

I Importance 3.9

~-

Control Roam Abandonment I 7 Ability to locate and operate components, including local controls. (CFR: 41.7 I 45.7)

Question Remote Shutdown Panel Transfer Switch 'B" has been placed in the EMERGENCY position.

Which of the following lists the SRVs that can be operated at the Remote Shutdown Panel (10C399) AND describes the status of their controls in the Control Room (CR)?

A A, B, C, D, & E. CR controls still function normally B

A, B. C, 0,

& E. CR controls are disabled F, H, & M. CR controls still function normally C

F, H & M. CR controls are disabled.

D Answer References Hope Creek Question - (262205, HC.OP-IO.ZZ-0008, Section 5.1, Attachment #1, Step 8.2.9 NOH01 MSTEAMC-02, MAIN STEAM SYSTEM, Obj R3d Justification References during Exam None D - CORRECT - F, H & M. CR controls are disabled. Only SRVs M, F & H can be controlled from the RSP and when the transfer switches are in EMERGENCY, the CR functions are disabled.

A - INCORRECT - A, B, C, D & E are the ADS valves not the valves that can be controlled from the RSP.

B - INCORRECT - A, 6, C, D & E CANNOT be controlled from the RSP.

C - INCORRECT - CR controls are disabled when RSP transfer switch B has been placed in the EMERGENCY position.

Question Source Bank E] Memory Level Comprehension Level Question History:

SXD Review - 7/21 - LOO 1.75 evaluate Revising

MRO SRO 29501 8 Partial or Total Loss of CCW / 8 I Tier#

Group#

1 Importance 2.9 AA2.04 Ability to determine and interpret the following as they apply to Partial or Total Loss of CCW :(CFR: 41.10/43.5/ 45.13)

System Flow

~~

Question With the plant operating at 100% power, power is lost to one of the two Operating SACS pumps due to a breaker fault.

After completing all immediate and subsequent actions of HC.OP-AB.COOL-0002, SAFETY/TURBINE AUXILIARIES COOLING SYSTEM there (1) flow in both loops of SACS. The most restrictive LCO for this condition requires the plant to be placed in cold shutdown within (2)

Complete the blank statements from the list below:

A (1)iS (2) 31.5 days (30 days + 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months + 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

~

B (1)is not (2)4 days (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months + 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

(1)is (2) 4 days (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months + 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

(1)is not (2) 31.5 days (30 days + 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months + 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

Answer A

INPO Question 25996 Hope Creek Procedure HC.OP-AB.COOL-0002, References SAFETYflURBlNE AUXILIARIES COOLING SYSTEM.,p. 9-1 3 Justification References during Exam Tech Specs - 3.7.1.1 -> 3.7.1.3 A - CORRECT - Per AB.COOL-0002, p. 9 when the Operating SACS pump trips, TACS will receive an AUTO SWAP signal, this section of the procedure ensures that the Standy SACS pump auto starts and then also ensures that if the Associated SACS pump tripped due to Low Delta P it is restarted. Thus at the end of the procedure, flow has been restored to both loops. Per TS 3.7.1.1 inop pump must be returned to service within 30 days or Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B - INCORRECT - flow will be restored to both loops C - INCORRECT - TS action must be taken within 31.5 days. This answer is plausible if the student mis-reads TS and determines that either one of the 2 Diesel Generators or Service Water pumps are inoperable per the *** Note.

D - INCORRECT - flow will be restored to both loops.

Question Source Mod 0

Memory Level Comprehension Level Question History:

SXD Review 7/21 - Maybe SRO level question, maybe a direct lookup 7/27 - I don't think it's a direct lookup - Look up Lesson Plan Objective

huestion 8 I Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 p i e r #

1 Group#

]

295019 Partial or Total Loss of Inst. Air I 8 AA1.03 Ability to operate and/or monitor the following as they apply to Partial or Total Loss of Inst. Air Instrument Air Compressor Power supplies:(CFR:

41.7145.5I45.6)

Question Given the following conditions:

Hope Creek is starting up from a Refueling outage, the plant is currently in OPCON 3 with temperature at 240°F and with the Instrument Air pressure at 105 psig and the InstrurnenVIService Air Systems aligned as follows:

Compressor Control Mode Status 00K107 MAN Running 10K107 MAN OFF 10K100 AUTO OFF A Maintenance Worker accidentally bumps into 7.2KV Bus 10A120 causing it's input breaker to open and the bus to d e energize.

Assuming no operator actions, which of the following correctly states the expected response of the Instrument/ Service Air systems?

~

A B

Service Air compressor 10K107 de-energizes, Instrument Air header pressure remains at 105 psig.

Service Air compressor 00K107 de-energizes, Instrument Air header pressure drops to 92 psig, when Service Air Compressor 10K107 starts and returns pressure to -95 psig.

Service Air compressor 00K107 de-energizes, Instrument Air header pressure drops to 85 psig when Emergency Air Compressor 10K100 starts and returns pressure to -105 psig.

Service Air compressor 00K107 de-energizes, Instrument Air header pressure drops to 85 psig when Emergency Air Compressor 10K100 starts and returns pressure to -95 psig.

NOH01 SERAIR-01, SERVICE AIR SYSTEM, p.47-48 NOH01 INSAIR-01, INSTRUMENT AIR SYSTEM, p15,42 Answer References Justification References during Exam None A. INCORRECT - Power to SAC 10K107 is from 7.2 KV bus 10A110, not 10A120

6.

INCORRECT - SAC 10K107 will not start at 92 psig because it's in MAN control.

C. INCORRECT - ElAC lOKlOO will auto start at 85 psig. however, it unloads at 100 psig, thereby making in not capable of raise pressure to 105 psig.

0.

CORRECT - Loss of Power to 10A120 causes a loss of Power to SAC 00K107, Instrument Air header pressure drops to 85 psig, when ElAC lOKlOO starts and brings pressure back to some value e 100 psig.

Question Source New 0

Memory Level Comprehension Level Question History:

SXD reviewed 7/25 - minor editorial changes to stem and distractor B - changed 105 psig to 95 psig.

IQuestion 9 I Exam-Cross-Ref Hope Creek RO Exam 1

1 Importance 3.4 Nov 2005 295021 Loss of Shutdown Cooling I 4 AA2.05 Ability to determine and interpret the fdlowing as they apply to Loss of Shutdown Cooling Reactor Vessel Metal Temperature (CFR: 41.10 143.5145.1 3)

Question Given the following conditions:

.The reactor has been shutdown for 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months following lo00 EFPD of operation.

.The plant is in Op Cond 4 with coolant temperature at 140°F.

.A total loss of Shutdown Cooling occurred at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months .

.All efforts to restore heat removal from the RPV have failed.

Assuming no additional operator action, when will the plant reach OPCON 3?

A 1245 B

1307 1 330 C

1352 D

References Hope Creek Question - 061328, HC.OP-AB.RPV-0009, Figure land Technical Specification Table 1.2 Answer Justification References during Exam Figure 1 of HC.OP-AB.RPV-0009 Justification

.1307-correct-Operational Condition 3 is achieved when the Reactor temperature reaches 200°F. The 140°F curve of Figure 1 intersects the 90-hour line between the 1.OOO and 1.250 hour0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months lines. 1307 is the only option that is between 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and fifteen minutes following the loss of SDC.

.1245. incorrect-Value obtained by using the 180°F curve.

.1330. -incorrect-Value obtained by using the 120°F curve.

.1352. -incorrect-Value obtained by using the 100°F curve.

Questiorr Source Bank Memory Level E] Comprehension Level Question History:

SXD Review 7121 - OK

Exam-Cross-Ref (euestion 10 I 295023 Refueling Acc I 8 AK2.03 Knowledge of the interrelations between Refueling Accidents and the following Hope Creek RO Exam - Nov 2005 Radiation Monitoring equipment (CFR41.7 145.71 45.8)

Question Given the following conditions:

-The plant is in a refueling outage with a fuel move in progress.

-The 'A' Refuel Floor Radiation Monitor has failed downscale. No actions have been taken to address this failure.

-At time 0000 a fuel bundle is dropped and radiation levels on the refuel floor start to slowly rise.

-At time 0005 the B Refuel Floor Radiation Monitor reaches its Hi Trip Setpoint.

-At time 0010 the C Refuel Floor Radiation Monitor reaches its Hi Trip Setpoint.

Under these conditions, an automatic trip of the Reactor Building Ventilation Exhaust (RBVE) fans due to Hi Refuel Floor Radiation levels:

A will occur at time 0010.

B is effectively disabled due to the 'A' Refuel Floor Radiation Monitor being failed downscale will m u r at time 0005.

c will NOT occur until at least 1 Reactor Building exhaust radiation monitor senses high radiation.

D Answer A

References lNPO Question 25978 NOH04000221C-01, RADIATION MONITORING SYSTEM p. 29 Justification References during Exam None A. - CORRECT - Per lesson plan p.29 item g. Automatic actions on a Refuel Floor Exhaust RM-23A HIGH radiation intensity level (any two of the three) - RBVE fans trip.

B - INCORRECT - still have 33 monitors available C - INCORRECT - since A channel is failed downscale, need 2/3 to get actuation. Therefore won't get actuation when B channel gets high signal.

D - INCORRECT -will get a trip of RBVE fans on either Hi Refuel Floor Rad levels and RBVE rad levels.

Question Source Memory Level 0

Comprehension Level Question History:

SXD Review 7/21 - Changed Distractor D to make it clearer

IQueslion 11 1 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 rTier#

1 Group#

1 I ImDortance 4

295024

~

High Drywell Pressure I 5 EA1.03 Ability to operate and/ or monitor the following as they apply to High Drywell Pressure LPCS

~~

Question The A Core Spray pump is in full flow test mode in accordance with HC-0P.IS.BE-0001, Core Spray Pumps A and C lnservice Test. A steam leak in the drywell has caused the following conditions:

Reactor was scrammed and all rods inserted.

RPV level reached -60 inches and is now rising with HPCI.

Drywell pressure is 3.0 psig rising.

RPV pressure is 800 psig lowering.

Offsite power remains available to the 4KV buses.

Based on the above conditions, which one of the following is the correct response of the Core Spray system?

A "A' Core Spray pump continues to run in full flow test, all others are operating in min flow.

g ALL Core Spray pumps are operating on min flow.

ALL Core Spray pumps are tripped and ALL pumps will start when RPV pressure lowers to 461 psig.

C ALL Core Spray pumps are injecting.

D References INPO Question 24762 NOHOlCSSYSO-01, CORE SPRAY SYSTEM Answer Justification References during Exam None A - INCORRECT - Core spray full flow test valve closes upon Receipt of a CSS initiation signal.

B - CORRECT - Core Spray received a start signal at DW pressure 5 1.68 psig. This caused all Core Spray pumps to start, however, RPV pressure is > 461 psig so upstream injection valves are closed and pumps are operating on their mini-flow valves. Core Spray test valve auto closed upon receipt of a CSS initiation signal.

C - INCORRECT - Core Spray pumps receive a start signal with pressure > 1.68 psig.

D - INCORRECT - Core Spray pumps upstream injection valves don't open until RPV pressure is e 461 psig.

'initiation" pump start signal is reached., A Core Spray running, no trip signal to any CS pumps and no loss of power.,

No Core Spray 'initiation" pump start signal is reached., Correct, > 2 psig signal closes full flow test valve Question Source Mod 0

Memory Level E] Comprehension Level Question History:

SXD review 7/21 - OK 8/2 JD - Minor editorial change to "A' distractor - Incorporated

Hope Creek RO Exam - Nov 2005 RO 1

1 P

0 SRO Importance 3.9 EG2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 45.2 /

45.6)

Question While implementing HC.OP-EO.ZZ-101, Step RC/L-5. the following plant conditions exist:

Suppression Pool Temperature is 230°F.

-RPV Water Level is -100 inches and rising.

.RHR Pump AP202 is injecting at 5,000 gpm.

.RHR Pump BP202 is injecting at 10,000 gpm.

.Loop A Core Spray is injecting at 1,000 gpm.

.Loop 6 Core Spray is injecting at 2,200 gpm.

.Suppression Chamber pressure is 5.0 psig.

.Suppression Chamber water level is 0 inches.

Which one of the following describes an action to be taken to ensure proper NPSH requirements are met?

- - I'm not sure about Supp. Chamber water level at 0" is that acceptable?

A Reduce "6" Core Spray Loop flow.

B Secure "A' RHR Pump.

Secure 'A' Core Spray Loop.

C Reduce "8" RHR Pump flow.

D References Hope Creek Question - a53616 HC.OP-EO.=-101, Reactor Pressure Vessel Control BWR Owners Group EPGs/SAG Appendix 6 - Section 5 -

Cautions Answer Justijication References during Exam None JUSTIFICATION:

CORRECT - Reduce 'B" RHR Pump flow. 'B" RHR pump is the pump for the given conditions of this question operating the closest to EOP Caution 2 NPSH limitations. The RHR flow limit is 6000 gpm.

.INCORRECT - Reduce "B" Core Spray Loop flow. No NPSH limitation concerns and RPV level is at -100". need to maintain "B' Core Spray Loop flow.

.INCORRECT - Secure "A" RHR Pump.No NPSH limitation concerns and RPV level is at -loo", need to maintain 'A' RHR Pump in service.

.INCORRECT - Secure "A' Core Spray Loop. 'A" Core Spray Loop flow is within NPSH limits.

Question Source Bank 0

Memory Level Comprehension Level Question History:

SXD review 7/21 - OK

(euestion 14 I Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 I=#

Group#

1 Importance 3.9 295028 High Dlywell Temperature I 5 EG2.1.30 Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)

Question Given the following conditions:

.A Large Break LOCA has occurred in the Drywell concurrent with a LOP

.Only "C' EDG is running

.All control rods are fully inserted

.Drywell pressure is 25 psig and rising

.Drywell temperature is 310 F and rising

.Reactor pressure is 25 psig and steady Suppression Pool Level is 80 inches and rising

."C' RHR Pump has been injecting LPCl flow for 3 minutes

.All RPV level indicators have failed upscale Based on the above conditions, which one of the following actions is REQUIRED?

A Stop LPCl injection because adequate core cooling is assured B

Continue LPCl injection because Drywell Spray is required Continue LPCl injection because adequate core cooling is not assured C

Stop LPCl injection because Drywell Spray is required D

Answer References Hope Creek Question - (256161, EOP-Caution 1. LP 0302-000.00H-00134-13 Obj 8 HC.OP-E0.U-0206, RPV Flooding HC.OP-EO.U-0101, RPV Control Justification References during Exam None CORRECT - Continue LPCl injection because adequate core cooling is not assured. Adequate core coaling is not assured because RPV level indication are failed upscale and the criteria in RPV flooding for RPV Pressure > 50 psig above Supp Pool pressure has not been met. 'C" RHR pump is not capable of injecting enough water to accurately reflect RPV water level indication off scale high.EOP-206 should be entered and LPCl injection should be continued.

INCORRECT - Stop LPCl injection because adequate core cooling is assured. Adequate core cooling is not assured because RPV level indication is in the unreliable region of EOP Caution 1. LPCl should be continued.

INCORRECT - Stop LPCl injection because Drywell Spray is required. "C' RHR pump can not be used to spray the Drywell.

INCORRECT - Continue LPCl injection because Drywell Spray is required. 'C' RHR pump can not be used to spray the Drywell.

Question Source Bank 0

Memory Level E] Comprehension Level Question History:

SXD review 71 21 - OK JD 8/2 - WA - Locate & Operate - asked to write question to J. Munro about Locate & Operate question.

IQuestion 15 I Exam-Cross-Ref Hope Creek RO Exam - Nov 2005

[Tier#

1 Group#

1 i L

1 l O S R 0 I Importance 3.5 295030 Low Suppression Pool Wtr Lvl I 5 EK3.07 Knowledge of the reasons for the following responses as they apply to Low Suppression Pool Wtr Lvl NPSH considerations for ECCS pumps:(CFR:

41.5/41.10/45.W 45.13)

Question The plant has experienced a transient and the following is observed:

- Suppression Chamber Overpressure: 9 psig

- Suppression Pool temperature: 240 degrees F

- Suppression Pool level at 74.5"

- Reactor pressure: 1000 psig

- RHR 'A" pump flow: 10,OOO gpm

- Core Spray '6" pump Flow: 1500 gpm

- All other low pressure ECCS pump are NOT in service.

Use the attached curves to determine if Net Positive Suction Head (NPSH) requirements are being met.

A There is sufficient NPSH for the "B" Core Spray Pump ONLY.

B There is sufficient NPSH for the 'A" RHR pump ONLY.

There is sufficient NPSH for both the "A" RHR pump and the 'B" Core Spray Pump.

C There is NOT sufficient NPSH for either the 'A" RHR pump or the "6' Core Spray pump.

D Answer A

References INPO Question 14383 EOP CAUTION 2 Justification References during Exam EOP Caution 2 Using EOP Caution 2 and realizing that being above the curve is the area of Unacceptable operation:

The limiting temperature for CS pump at 5 psig and 1500 gpm = 232°F The limiting temperature for CS pump at 10 psig and 1500 gpm = 244°F Interpolating for 9 psig gives a Temperature limit of -242°F for 9 psig. Since given temperature = 24OoF this puts the B CS pump in the area of ACCEPTABLE operation.

The limiting temperature for RHR pump at 10 psig is 235"F, since Given temperature is 240°F this puts the pump in the region of UNACCEPTABLE operation.

This makes ONLY Answer A CORRECT.

Question Source Mod 0

Memory Level Comprehension Level Question &tory:

SXD reviewed 7/22 - OK

(euestion 16 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 295031 Em. 10 Question Importance 4

Reactor Low Water Level I 2 Knowledge of the interrelations between Reactor Low Water Level and the fdiowing Redundant reactivity control Given the following:

-The plant is operating at 100% power.

-A transient results in a scram setpoint being exceeded.

-The Reactor Protection System fails to automatically scram the Reactor.

Without operator action, which of the following describes how the Control Rods will be automatically inserted to shutdown the Reactor via the ARI system?

A RPV level less than or equal to minus 38 (-38) inches will immediately ENERGIZE the ARI valves to depressurize the scram air header.

B RPV level less than or equal to minus 38 (-38) inches will immediately DE-ENERGIZE the ARI valves to depressurize the scram air header.

RPV pressure greater than or equal to 1037 psig will immediately ENERGIZE the ARI valves to depressurize the scram air header.

RPV pressure greater than or equal to 1037 psig will immediately DE-ENERGIZE the ARI valves to depressurize the scram air header.

Answer A

References INPO Question 22776 SYSTEM (ARCS), p.8 NOH01 RRCS00-00, REDUNDANT REACTIVITY CONTROL Justification References during Exam None A - CORRECT - with RPV level < -38" the ARI valves are energized to depressurize the scram air header resulting in rod insertion.

8. - INCORRECT - valves are Energized to actuate, not de-energized.

C. - INCORRECT - ARI pressure setpoint is 1071 psig, not 1037 psig D - INCORRECT - ARI pressure setpoint is 1071 psig, not 1037 psig.

Question Source Mod Memory Level u

Comprehension Level Question History:

SXD review - 7/21 - Add (via the ARI system) to the end of the stem. Removed 'control rod insertion will begin within 15...) from all distractors

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref r

T 1

Importance 3.8 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown I 1 EA1.02 Ability to operate and / or monitor the following as they apply to SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown RRCS Question The plant was operating at 98% power when a transient occurred. Following the transient all SRVs opened. 2 minutes later, Reactor pressure is stable with 6 SRVs open. No operator actions have been taken.

Which of the following is correct for these conditions?

Both Recirculation Pumps A

have tripped B

are running normally.

are running at minimum speed C

are currently running but will trip in 1.9 minutes when a time delay times out.

D References INPO Question 23485 NOH01 RRCSOO-00, REDUNDANT REACTIVITY CONTROL Answer A

SYSTEM (RRCS), p.13 Justification References during Exam None A. CORRECT - Following the transient, all SRVs opened. Reactor pressure has to be greater than 1071 psig for all valves to open. Reactor pressure greater than 1071 psig causes both Recirc Pumps to trip. A is the only correct answer.

6. INCORRECT - plausible because may not have hit a trip condition.

C. INCORRECT - plausible because recirc pumps have runbacks, operator may incorrectly believe a runback condition has been met.

D - INCORRECT - plausible because a 3.9 minute timer does exist on RRCS, however it is for SLC initiation, not Recirc pump trip.

Question Source Mod 0

Memory Level E] Comprehension Level Question History:

SXD Review 7/21 - removed # of SRVs from stem. Removed off from Distractor A

[Question 18 1 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 rTier#

1 Group#

1 I Imuortance 3.9

~

295038 High Off-site Release Rate / 9 EK3.02 Knowledge of the reasons for the following responses as they apply to High Off-Site Release Rate System Isolations (CFR:41.8 to 41.10l45.3)

Question HC.OP-EO.ZZ-O103/4, Reactor Building & Rad Release Control, step RR-5, directs isolation of all primary systems discharging into areas outside Primary Containment or Reactor Building, except those systems required to assure adequate core cooling andor shutdown the reactor.

In accordance with the EOP Bases document, HC.OP-EO.ZZ-103/4. Reactor Building & Rad Release Control, these systems are specifically exempted from isolation, because:

A systems operated for RPV control are given a higher priority than stopping a rad release.

B isolation of a EOP support system requires an upgrade of the Emergency Classification.

they are required to support alternate reactor depressurization methods.

additional radiological consequences from them are unlikely.

Answer A

References INPO Question 25837 BWROG, EPGdSAGs Appendix B, section 9 Radioactivity Release control HC.OP-EO.U-103/4. Reactor Building & Rad Release Control Bases Document - p. 13 & 14 Justification References during Exam None Per EOP Bases document 103/104:

The objectives of RPV Control, Primary Containment Control, and the EPG contingencies are given higher priority than the objectives of Radioactivity Release Control. Systems that must be operated to perform other steps of the EPGs are therefore not isolated in this step.

A - CORRECT matches bases document B - INCORRECT - Not in accordance with bases document C - INCORRECT - not in accordance with bases document D - INCORRECT - not in accordance with bases document Question Source Bank Memory Level Comprehension Level Question History:

SXD review 7122 - Minor editorial changes (added procedure)

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref 19 (Question I

rTrer#

1 Group#

1 I 0

SRO 1

I Importance 2.5 AK1.01 Knowledge of the operational implications of the following concepts as they apply to the Plant Fire On Site Fire Classifications by type (CFR: 41.8 to 41.10 /45.3),

Question A fire occurs in the Upper Cable Spreading Room (Control Equipment Mezzanine Room 5403).

The installed fire protection system automatically actuates.

The room must be entered to determine if the fire has been extinguished.

(1) What is the classification of the fire that is expected in this area?

AND (2) What safety hazard, from the automatic system actuation, should be considered prior to operators entering the Cable Spreading Room?'

A Class C - Suffocation from oxygen depletion due to the discharge of C02 in the area B

Class B - Suffocation from oxygen depletion due to the discharge of halon in the area Class C - Suffocation from oxygen depletion due to the discharge of halon in the area c

Class B - Suffocation from oxygen depletion due to the discharge of C02 in the area 0

Answer A

References INPO Question 24855 NOH01 FIRPRO-02, FIRE PROTECTION, p.55. p. 63 and p.85 Justification References during Exam None A-CORRECT - Class C fire due to electrical equipment in area, Suffocation due to discharge of C02 B - INCORRECT - not a Class B fire and no halon in that room C - INCORRECT - not expecting to get a halon discharge in that room D - INCORRECT - not a class B fire Question Source Mod Memory Level 0

Comprehension Level Question History:

SXD review - 7/21 - Changed water to Halon

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref 20

[Question I

p2-1 Importance 3.3 295005 Main Turbine Generator Trip I 3 AK2.04 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following:

Main generator protection (CFR: 41.77458)

Question Given the following conditions:

.The plant is operating at 20% power

.A main generator load reject has just occurred

.A fault in the control circuit causes a powerlload unbalance trip during the load reject Which of the following is the immediate expected response of the Turbine Control Valves (TCVs) and the Reactor Protection System (RPS)?

A TCVs throttle close, RPS trips

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B TCVs throttle close, RPS does not trip TCVs fast close, RPS trips C

TCVs fast close, RPS does not trip D

Hope Creek Question - 061 307, HC.OP-AB.BOP-0002 Additional Information /Automatic actions and notes NOH01 MNTURB-02, MAIN TURBINE CONSTRUCTION AND COMPONENTS, p. 66 Answer References Justification References during Exam None CORRECT - TCVs fast close, RPS does not trip. The load reject causes the TCVs to fast close. The fast closure does not initiate a RPS trip because turbine load is <W%. Since power is within the capacity of the BPVs, no pressure transient will trip RPS.

INCORRECT - TCVs throttle close, RPS does trip. The load reject causes the TCVs to fast close. The fast closure does not initiate a RPS trip because turbine load is <W%. Since power is within the capacity of the BPVs, no pressure transient will trip RPS.

INCORRECT - TCVs fast close, RPS does trip. The fast closure does not initiate a RPS trip because turbine load is

<30%. Since power is within the capacity of the BPVs, no pressure transient will trip RPS.

INCORRECT - TCVs throttle close, RPS does not trip. The load reject causes the TCVs to fast close Question Source Bank 0

Memory Level E] Comprehension Level Question History:

SXD review - 7/21 - OK

(euestwn 21 I Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 m#

Group#

I ImDortance 3.2 295002 Loss of Main Condenser Vac I 3 AA2.02 Ability to determine and interpret the following as they apply to Loss of Main Condenser Vacuum's Reactor Power Plant Specifc:(CFR:

41. 1 Ol43.W 45.1 3)

Question Given the following:

.All four Circulating Water Pumps are in operation

.Plant is operating at 100% power Circulating Water System Inlet temperature is 80°F

.Indicated Main Condenser pressure is 2.75 in HgA Assume the remaining Circulating Pumps' Discharge Valves are reopened fully, NO rise in basin temperature and no other operator actions are taken.

What is the expected condenser backpressure and what is the expected change in reactor power following the removal of Circulating Water Pump AP501 from service?

A 3.5 in HgA, reactor power increases (ie. greater than 2%)

B 3.5 in HgA, reactor power stays the same (ie. Doesn't change more than 2%)

4.15 in HgA, reactor power increases (ie. Greater than 2%)

c 4.15 in HgA, reactor power stays the same (ie. Doesn't change more than 2%)

D Answer References Hope Creek Question - Q55132 HC.OP-SO.DA-0001, Rev. 35, Attachment 5 Justification References during Exam from HC.OP-SO.DA-0001 A - INCORRECT - Reactor power should not change with a decrease in vacuum. If anything reactor power may go down a little bit due to increased condenser temperature and reduced condenser subcooling B-CORRECT-3.5 inHgA. If CW inlet temp does not change, then the condenser vacuum rises vertically on the graph until it reaches the line for three pump operation 42 80 degF. Since the inital back-pressure of 2.75 indicates 100 percent CF. Reactor power should remain the same C - INCORRECT 4.15 - 3 pump ops at 70 percent CF.

D - INCORRECT 4.15 - 3 pumps ops at 70% CF

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Question Source Bank 0

Memory Level Comprehension Level Question History:

SXD review 7/21 - OK JD W2 - WA asking for Reactor Power 8/3 - initially was going to change question to add reactor power change, decided to ask Steve on Monday 8/4 Re-wrote questions

295008 Exam - Cross - Re f Hope Creek RO Exam - Nov 2005 I Tier#

Group#

1 importance 3.4 High Reactor Water Level I 2 AK3.06 Knowledge of the reasons for the following responses as they apply to High Reactor Water Level RClC Turbine Trip Question During a transient, the RO started the RClC system for reactor water level control using the appropriate operating procedure. The RO became distracted and allowed level to rise above the High Reactor Water level at 58" after which it lowered below the Low Reactor Water level at -38".

Which of the following describes the reason for, and expected response of RClC during the reactor water level transient?

A B

The RClC Trip and Throttle Valve (HV-4282) will close on High Water Level and RClC will automatically restart on Low Reactor Water Level.

The RClC Trip and Throttle Valve (HV-4282) will close on High Water Level and RClC will have to be reset and manually started on Low Reactor Water Level.

The RClC Steam Supply Valve (F045) will close on High Water Level and RClC will automatically restart on Low Reactor Water Level.

The RClC Steam Supply Valve (F045) will close on High Water Level and RClC will have to be reset and manually started on Low Reactor Water Level.

Answer References NOH01 RCICOO-02. REACTOR CORE ISOLATION COOLING SYSTEM, p22-23 Justification References during Exam None A - INCORRECT - Trip and Throttle valve does not close on Level 8 6 - INCORRECT - Trip and Throttle valve does not close on Level 8 C - CORRECT - Steam supply valve will close and RClC will auto restart at Level 2 D - INCORRECT - RClC will auto restart at Level 2 Question Source Mod El Memory Level 0

Comprehension Level Question History:

SXD review 7/22 - LOD = 1 - re-write question BM - re-wrote question

Hope Creek RO Exam - Nov 2005 Exam -Cross-Ref 23

[Question I

rTier#

1 Group#

2 I Imvortance 3

295009 Low Reactor Water Level I 2 AK1.02 Knowledge of the operational implications of the following concepts as they apply to the Low Reactor Water Level Recirculation pump net positive suction head

~

Question The plant is currently at 27% power. Plans for the shift are to continue the startup and power ascension. A malfunction in the Feedwater Control System has resulted in the following:

- RPV level is 25 inches and trending down

- Total Feedwater flow is 2.5 mlbhr and steady

- 3 Circ Water pumps are running

- Condenser Vacuum is 3.8" HgA and degrading Assume no operator actions have been taken. Which of the following statements is correct regarding the Reactor Recirculation system response based on these CURRENT plant conditions?

A B

Speed Limiter 1 (30% flow) is actuated to ensure Recirculation Pump net positive suction head protection based on RPV level.

Speed Limiter 2 (45% flow) is actuated to ensure Recirculation Pump net positive suction head protection based on RPV level.

Speed Limiter 2 (45% Row) is actuated to bring Condenser Vacuum back to normal.

C Speed Limiter 1 (30% flow) is actuated to bring Condenser Vacuum back to normal.

D References New Question NOH01 RECIRC-02. Reactor Recirculation System, P. 53-55 Answer A

Justification References during Exam None A - CORRECT - Total FW flow is -17% which is c 20%, this causes a Speed Limiter #1 runback to ensure Recirc Pump NPSH 6 - INCORRECT - Speed Limiter 1 is actuated, not Speed Limiter 2 C - INCORRECT - Speed Limiter 1 is actuated, not Speed Limiter 2 D - INCORRECT - Condenser vacuum is rising but still within normal limits. Must be > 4.5" to cause a Recirc pump runback.

Question Source New 0

Memory Level Comprehension Level Question History:

SXD Review 7/21 - minor editoral comments

(Question 24 El-295029 EK2.07 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 Importance 3.1 High Suppression Pool Wtr LvlI5 Knowledge of the interrelations High Suppression Pool Wtr Lvl and the following Drywell/ containment water level:(CFR: 41.7 145.7145.8)

Question An Override step in HC.OP-E0.U-0202, Emergency Depressurization, directs the operator to open the Inboard MSL Drain Valve (AB-HV-FOl6) when Containment water level is expected to exceed 48 feet.

Which one of the following describes the reason for this action?

Opening the Inboard Main Steamline Drain Valve A

B maintains the availability of the Main Steamline drain path for reactor vessel pressure control if required.

ensures as much heat energy as possible is rejected to the Main Condenser to minimize the dynamic loading on Containment.

maintains Containment water level below the SRV solenoids by establishing a drain path from the reactor vessel to the Main Condenser.

ensures the SRV Tail Pipe Level Limit is not exceeded prior to emergency depressurization.

D Answer A

References INPO Question 21944 HC.OP-EO.ZZ-0202 flowchart HC.OP-EO.=-0202, Emergency Depressurization Bases, p.5 BWROG EPGISAG'S App. B - P 326 Justification References during Exam None A - CORRECT - per the BW ROG guidelines - If primary containment water level rises above the elevation of the SRV solenoids, the SRVs may no longer be operable. Other methods must then be used to control RPV pressure and prevent repressurization. Opening the inboard main steam line drain valve preserves the main steam line drains for future use.

B - INCORRECT but plausible, while Opening AB-HV-FOl6 does not reject any heat to the Main Condenser it could reject heat to the condenser if the FO19 and F021 were open.

C - INCORRECT but plausible, while opening AB-HV-FO16 does not necessarily maintain CNMT water level below the SRV solenoids, it may open a drain path to the main condenser.

D - INCORRECT but plausible, while opening AB-HV-FO16 does not drain water from the steam lines, it could if both FO19 and F021 were open.

Question Source Mod Memory Level 0

Comprehension Level Question History:

SXD reviewed 7/22 - OK

Exam-Cross-Ref rI(i-n Group#

2 I 0

SRO Importance 500000 High CTMT Hydrogen Conc. I 5 EK2.02 Knowledge of the interrelations between High CTMT Hydrogen Conc. And the following Question Given the following conditions:

Hope Creek has experienced a transient and the following conditions are present:

Drywell H2 concentration is reading 1.5% by volume Drywell 0 2 concentration is reading 5.5% by volume Drywell Pressure is 2.0 psig and stable Reactor water level is +lo" and rising slowly (lowest level - 0")

Hope Creek RO Exam - Nov 2005 Containment oxygen monitoring systems(CFR:

41.7 1 45.7 145.8)

Assuming no other operator actions have occurred, what is the status of the 0 2 monitors?

Assuming the above readings are correct and Containment venting cannot be performed, what actions shall be taken with regards to the H2 Recombiners in accordance with HC.OP-EO.=-0102, Primary Containment Control?

A g

02 monitors are OPERABLE and the H2 Recombiners should be placed in service.

02 monitors are INOPERABLE because of Containment Isolation, however the H2 Recombiners should be placed in service.

0 2 monitors are INOPERABLE because of a Containment Isolation, however the H2 Recombiners should NOT be placed in service 0 2 monitors are OPERABLE and the H2 Recombiners should NOT be placed in service.

D NOH01 H202AN-01, Hydrogen Oxygen Analyzer System - p. 17 NOH01 H2RECM-00, CONTAINMENT HYDROGEN Answer References RECOMBINER SYSTEM, p.8 CONTROL, step PCIH-1 HC.OP-EO.U-O102(Q)-FC, PRIMARY CONTAINMENT Justification References during Exam None A. INCORRECT - 0 2 monitors are INOPERABLE due to a Contaiment Isolation on High Drywell Pressure.

B. CORRECT - H2 Recombiners should be placed in service due to High H2 concentration per EOP 102, concentration > 0.5% and c 2%

C. INCORRECT - H2 Recombiners should be placed in service due to High H2 Concentration per EOP 102 D. INCORRECT - 0 2 monitors are INOP Question Source New 0

Memory Level El Comprehension Level Question History:

SXD review - 7/21 - OK

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref puestion 29 I

[Tier%

Group#

I Importance 3.5 205000 Shutdown Cooling A3.03 Ability to monitor automatic operations of the Shutdown Cooling System(RHR Shutdown Cooling Mode) including lights and alarms (CFR:41.7/45.5)

Question Given the following Plant conditions:

Hope Creek is in OPCON 3 Cooling down to for a Refueling Outage, "A" Shutdown Cooling is being placed in service and is currently in the following status:

"A' RHR fill and vent has been completed. However, the F007A - RHR Pump mini-flow valve's breaker was inadverntantly left closed.

"A" RHR Loop has been warmed up.

Both Reactor Recirc Pumps have been secured.

The RO is lining up 'A' RHR system for Shutdown cooling and valves are currently lined up as follows:

FOO9 - Shutdown Cooling INBD ISLN MOV - Open F008 - Shutdown Cooling OUTBD ISLN MOV - Open AP202 RHR PUMP - Running F015A - RHR Loop A Ret to Recirc - Throttled Open F007A - "A" RHR pump mini-flow - Closed F024A - "A" RHR Full Flow test valve - Closed F027A - "A' Torus Spray lnj valve - Closed To reduce an RCS cooldown the RO throttles closed on F015A when the following alarm is received.

'RHR A S/D CLG & MIN FL VLV OPEN" alarm is received in the control room.

Assuming NO Operator actions are taken, which of the following conditions will result:

~

A B

F008 and FOO9 will Auto close once the Mini-flow valve F007A gets full Open.

FOO8 and FOO9 will Auto Close on Low RPV level 3 (+12.5")

No Auto Actions will occur, this is an expected alarm for the above conditions.

C F008 and FOO9 will Auto Close on Low RPV level 1 (-129")

D

~~

~

NOH01 RHRSYSC-03, RESIDUAL HEAT REMOVAL SYSTEM,

p. 30 References Answer Justification References during Exam None A - INCORRECT - FOO8 and 9 will NOT Auto Close based on mini-flow valve position B - CORRECT - Having the Mini-flow valve open and taking suction from Reactor vessel will cause Reactor Vessel to lower, when vessel level reaches Low RPV Level 3, F008 and 009 will Auto Close.

C - INCORRECT - Reactor vessel will lower due to Mini-flow open and taking suction Reactor vessel.

D - INCORRECT - F008 and FOO9 will auto close on Low RPV level 3 and level should not get to Low RPV level 1.

Question Source New 0

Memory Level Comprehension Level

Question History:

SXD review 7/27 - OK

Hope Creek A 0 Exam - Nov 2005 Exam-Cross-Ref

[Tier#

2 Group#

1 J Imuortance 3.3 206000 HPCl K5.05 Knowledge of the operational implications of the following concepts as they apply to the HPCl Turbine speed control Question Given the following conditions:

.The HPCI system running in automatic at rated flow.

.The flow element providing feedback to the flow controller begins to fail downscale, slowly.

How will actual HPCl turbine speed and system flow respond?

A Turbine speed will increase and flow will increase

~~

~

g Turbine speed will decrease and flow will decrease Turbine speed will decrease and flow will remain at rated C

Turbine speed will increase and flow will remain at rated D

Answer A

References Hope Creek Question Q56448 NOHOlHPC100-02, HIGH PRESSURE COOLANT INJECTION SYSTEM, p.30 Justification References during Exam None Correct answer:turbi'ne speed will increase and flow will increase The following distractors are incorrect as follows:

.turbine speed will increase and flow will remain at rated-Incorrect-As flow feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise

.turbine speed will decrease and flow will decreaselncorrect-As flow feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise

.turbine speed will decrease and flow will remain at rated-Incorrect-As feedback lowers, controller will raise turbine speed and, with it actual flow rate will raise Question Source Bank 0

Memory Level Comprehension Level Question History:

SXD review - 7/21 - OK

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref 32 buestion I

rn &#

2 GrouD#

1 I 1

I Importance 3

~

l O S R 0 I 209001 LPCS K2.0 1 Knowledge of electrical power supplies to the following Pump power (CFR41.7)

Question Hope Creek has experienced a transient and a partial loss of Offsite power.

Current conditions are as follows:

Red Lion Transmission Line (61x50) is DE-ENERGIZED with a ground fault on it 500KV Circuit Breaker BS1-3(61x) failed to OPEN 13.8KV Circuit Breaker BS1-2 failed to OPEN Reactor has SCRAMMED and all rods are INSERTED Reactor water level is -135" and rising slowly Drywell Pressure is 1.35# and lowering slowly (Max. Pressure -1.5#)

"C' CS pump NORMAUEMERGENCY TAKEOVER switch is(was) in the EMERGENCY position A and B Diesel Generators FAILED TO START Based on the above conditions, what is the status of the Core Spray Pumps?

A All Core Spray Pumps are running B

A, B, and D Core Spray Pumps are running Only C Core Spray Pump is running C

Only D Core Spray Pump is running D

Answer A

References NOHOICSSYSO-01, CORE SPRAY SYSTEM p.16 NOH01 EAC00-02, CLASS 1 E AC POWER DISTRIBUTION 066-01: Class 1 E AC Power Distribution (Training drawing) 027-01: Core Spray System (Training Drawing)

Justification References during Exam 1.13.8KV Ring Bus - [AV1593E.vsd]

A: INCORRECT - 'C" CS pump will not have started because it's Takeover switch is in the EMERGENCY Position 6: CORRECT - The Loss of the Red Lion Line and the Circuit breaker faults will have caused a loss of Bus Section 1OX and Station Sewice XFMR 16x501, however 1AX501 will still be energized from Offsite power, therefore power to 10A402 and 10A404 will auto transfer to 1AX501 causing all of the 4.16KV buses to be energized. As stated above "C" CS pump will not have started, leaving A, B and D CS pumps running.

C: INCORRECT - 'C" CS pump will not have started because it's Takeover switch is in the EMERGENCY Position D: INCORRECT - A and B Diesel Generators failing to start will not cause their respective buses to be de-energized because they will have received power from 1AX501 Question Source New 0

Memory Level Comprehension Level Question History:

SXD reviewed 7/22 - give students 5OOKV switchyard print

Question 33 I Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 MRO b# Group#

I 0

SRO Importance 3.8 K4.04 Knowledge of SLC design feature(s) and or interlock(s) which provide for the following Indication of fault in explosive valve firing circuits (CFR41.7)

Question Hope Creek was operating at full power when an instrument air line break caused the outboard MSlVs to go closed. The following then occurred:

- The reactor failed to scram and attempts to drive rods were unsuccessful.

- The Shift Supervisor ordered SLC injection.

- Both SLC pump AP208 and BP208 START pushbuttons have been depressed.

- SLC pump control bezel start pushbuttons are backlit RED.

- The squib valve continuity lights are lit.

- Pump discharge pressure is 1395 psig.

- Reactor Pressure is currently 1025 psig.

Based on these indications which of the following correctly describes the status of the SLC system?

A SQUIB valves are closed, with SLC pumps running therefore, SLC is NOT injecting.

B SQUIB valves are OPEN, with SLC pumps running, therefore SLC is injecting SQUIB valves are OPEN, however, the SLC pumps are NOT running, therefore SLC is NOT injecting C

SQUIB valves are closed AND SLC pumps are NOT running, therefore, SLC is NOT injecting D

References INPO Question 20790 NOH01 SLCSYS-00, STANDBY LIQUID CONTROL SYSTEMS, p.27-29 Answer A

Justification References during Exam None A - CORRECT - the pump control bezel start pushbuttons backlit RED, along with pump discharge pressure of 1395 psig indicate the pumps are running. Squib valve continuity lights being lit, indicate valves are closed, therefore no injection is occurring.

B - INCORRECT - Squib valves are closed C - INCORRECT - Squib valves are closed D - INCORRECT - SLC pumps are running Question Source Mod 0

Memory Level Comprehension Level Question History:

SXD review 7/21 - Minor editorial changes

IQuestion 34 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 I Tier#

2 Group#

1 Importance 3

212000 RPS K3.11 Knowledge of the effect that a loss or malfunction of the RPS will have on the following Recirculation system (CFR41.7/45.6)

Question Given the following:

.The Reactor is initially at 20% power

.The Main Turbine is synchronized to the grid and loaded

.The RX RECIRC PUMPS RPS TRIP BYP alarm (Cl-E3) is NOT illuminated

.A loss of "6" RPS Bus has occurred What is the operational effect of a fast closure of all Turbine Control Valves during this condition?

A EOC-RPT trip of Recirculation Pump A and NO trip of Recirculation Pump B B

EOC-RPT trip of both Recirculation Pumps EOC-RPT trip of Recirculation Pump B and NO trip of Recirculation Pump A C

Both Recirculation Pumps running with half-scram inserted D

Hope Creek Question - 061263, HC.OP-AB.U.IC-0003 discussion section step 2 NOH01 RECIRC-02, Reactor Recirculation System, p.37 and p.

69 Answer References Justification References during Exam None Justification:

.EOC-RPT trip of both Recirculation Pumps - Correct, loss of RPS bus power, at any reactor power level, in conjunction with the cited Turbine Control Valve fast closure will result in EOC-RPT trip of both Recirculation Pumps.

This occurs due to a loss of the automatic bypass for EOC-RPT when less than about 30% power (first stage pressure less than 135.7 psig). The keylock bypass of the EOC-RPT trip is removed with the Main Turbine loaded.

The RX RECIRC PUMPS RPS TRIP BYP alarm is cleared when the RECIRC PUMP TRIP A/B SYSTEM DISABLE switch is placed in the NORM position. This defeats the bypass of the RPT trips.

.EOC-RPT trip of Recirculation Pump A and NO trip of Recirculation Pump 6 - Incorrect, both pumps will trip.

.EOC-RPT trip of Recirculation Pump B and NO trip of Recirculation Pump A - Incorrect, both pumps will trip.

.Both Recirculation Pumps running with half-scram inserted - Incorrect, both pumps will trip.

Question Source Bank 0

Memory Level Comprehension Level Question History:

SXD Review 7/21 - OK

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref 36 IQuestion I

[Tier#

2 Group#

1 I Imvortance 2.5 215003 IRM K2.01 Knowledge of electrical power supplies to the following IRM Channels/ detectors (CFR41.7)

-~

Questio n A Loss of 24VDC occurs for 1AD307 DC Distribution Panel.

Which of the following describes the effect on Nl's:

SRM IRM APRM A

nochange fails low no change B

fails low no change no change fails low fails low no change C

fails low fails low fails low D

-~

NOHOlDCELEC-00, DC ELECTRICAL DISTRIBUTION, p.38 NOH01 IRMSYS-01, Intermediate Range Monitoring System, p26 Simplified Training prints for SRM, IRM and APRMs Answer References

~

Justification References during Exam None A - INCORRECT - SRM's are powered from 24VDC and would fail downscale B - INCORRECT - IRM's are powered from 24VDC and would fail downscale C - CORRECT - SRM's and IRMs are powered from 24 VDC and would fail downscale, APRM's are powered from 120 VAC panels and would remain unchanged D - INCORRECT - APRM's are powered from 120 VAC and would not fail downscale Question Source New Memory Level 0

Comprehension Level Question History:

SXD Review - 7/21 - LOD 1.O - rewrite question to make it more difficult W3 - Re-wrote question

('SRO I

importance 3.9 223002 PClSlNuclear Steam Supply Shutoff A4.02 Ability to manually operate and/or monitor in the control room Manually initiate the system (CFR:41.7/45.5 to 45.8)

Question Select the action(s) that will close all the NS4 outboard isolation valves other than the MSIVs.

"6" and "C" NS4 logic channels are deenergized.

A B

'B" NS4 logic manual initiation collar is armed and pushbutton is depressed.

"A' and "D" NS4 logic channels are deenergized.

C

'0" NS4 logic manual initiation collar is armed and pushbutton is depressed.

D Answer References Hope Creek Question - Q53931 NOHOlNSSSSO-00, NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM (NSSSS) - p.10, p.13 Training Print 045-01: Nuclear Steam Supply Shutoff System Justification References during &am None IAW 621-1090-0062 and HC.OP-SOSM-0001 -

A - INCORRECT - this will cause a full group one [MSIV] isolation [e.g. MSIV's will close]

6 - INCORRECT - this will cause no isolation C - INCORRECT - this will cause a full NS4 isolation and the MSIV's will close D - CORRECT - 'D" NSSSS logic manual initiation collar is armed and push-button is depressed.-Correct Question Source Bank Memory Level 0

Comprehension Level Question History:

SXD review 7/21 - Minor editorial change - LOO 1.5 - evaluate making question more difficult

Exam - Cross-Re f 42 kuestion I

w r #

Group#

1 Importunce 3.9

_ _ _ _ _ _ _ _ ~ ~ _ _

239002 SRVs A4.06 Ability to manually operate and/or monitor in the control room Hope Creek RO Exam Nov 2005 Reactor water level (CFR:

41.7145.5 to 45.8)

Question The plant is operating at 100% power, with the following:

- Reactor water level is 35 inches

- An SRV inadvertently opens With NO operator action, which one of the following describes Reactor Water level response?

Reactor Water level will:

~

A lower and then return to 35 inches B

lower and remain below 35 inches rise and then return to 35 inches C

rise and remain above 35 inches D

Hope Creek Question ID - 22077 NOH01 FWCONTC-02, FEEDWATER CONTROL SYSTEM, p.11 Answer References Justification References during Exam None A - INCORRECT - lower and then retum to 35 inches (see answer C) 6 - INCORRECT - lower and remain below 35 inches (see answer C)

C - CORRECT - rise and then retum to 35 inches. RPV Swells up on the RPV pressure reduction when the SRV initially opens. RPV level returns to 35 inches due to DFCS setpoint of 35 inches.

D - INCORRECT - rise and remain above 35 inches Question Source Bank

@ Memory Levei Comprehension Level Question History:

SXD review 7/21 - Minor Editorial changes

Exam-Cross-Ref Hope Creek RO Exam - Nov 2005

( O S R O 1 262001 K4.03

-. Importance 3.1 AC Electrical Distribution Knowledge of AC Electrical distribution design feature(s) and or interlock(s) which provide for the following Interlocks between automatic bus transfer and breakers (CFR:41.7)

~-

Question With the plant in a normal electrical lineup for 100% power, the TRIP pushbutton is pressed for breaker 52-40201.

Normal Feed Breaker for 10A402 on Control Room panel 100351 E.

Which choice below describes the response of the 10A402 Bus and "B" EDG?

A The Alternate Feed Breaker, 52-40208 will close energizing Bus 10A402,"B" EDG will not be running.

B Bus 10A402 will be de-energized. The '8' EDG will NOT be running.

Bus 10A402 will be de-energized. The "B" EDG will be running with its output breaker open.

C The "B" EDG will start and its output breaker will close energizing Bus 10A402.

D Answer References Hope Creek Question - (253557, NOH01 EACOO-02, CLASS 1 E AC POWER DISTRIBUTION, p.27 Justification References during Exam None CORRECT - Bus 10A402 will be de-energized. The "B' EDG will NOT be running. The automatic transfer to the alternate feed and the start of the Diesel will not m u r if the normal breaker is manually tripped.

INCORRECT - The Alternate Feed Breaker, 52-40208 will close energizing Bus 10A402:B" EDG Lockout will prevent the EDG start and output breaker closure. The automatic transfer to the alternate feed will not occur if the normal breaker is manually tripped.

INCORRECT - Bus 10A402 will be de-energized.The 'B" EDG will be running with its output breaker open. The automatic start of the Diesel will not occur if the normal breaker is manually tripped.

INCORRECT - The 'B" EDG will start and its output breaker will close energizing Bus 10A402.The automatic start of the Diesel will not occur if the normal breaker is manually opened Question Source Bank Memory Level 0

Comprehension Level Question History:

SXD review 7/21 - minor editorial changes

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref rTier#

2 Group#

1 I Importance 2.8 262002 UPS (AC/DC)

K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the UPS (AC/DC)

DC electrical power (CFR:41.7/45.7)

Question Hope Creek is at 100% power with the following lineup on 120V Class 1 E Cyberex 20KVA Inverter 1 AD481 CB 125V DC Power Breaker Closed CB-201 - 480V AC Normal Power Breaker Closed 26-301 - 480V AC Backup Power Breaker Open Auctioneered Bypass Switch is in the BYPASS 1 Position Manual Bypass Switch is in the NORMAL Position An Operator inadvertently opens the CB-21 (Battery Output from Auctioneered Circuit).

What effect will that have on Class 1 E Instrument Distribution Panel 1AJ481?

A B

Class 1 E Panel 1AJ481 will be de-energized due to Auctioneered Bypass Switch being in the BYPASS 1 Position.

Class 1 E Panel 1 AJ481 will be energized from 480V AC Backup Power.

Class 1 E Panel 1 AJ481 will be energized from 480V AC Normal Power.

C Class 1 E Panel 1AJ481 will be de-energized due to CB-301 - 480V AC Backup Power Breaker being Open.

D Answer References NOHOlEAC00-02, CLASS 1E AC POWER DISTRIBUTION, p.

60-62 Justijication References during Exam Figures 6 and 8 of NOH01 EAC00-02, AV2114D.vsd and AV2114F.vsd A - INCORRECT - Auctioneer Bypass - Allows bypassing of one of the two Auctioneer Diodes (either diode can perform the design function) since either diode can perform the design function, bypassing diode 1 will have NO EFFECT.

6. - INCORRECT -Breaker CB-301 is given as OPEN and there are NO auto closures for this breaker.

C. - CORRECT - Power is normally supplied to 120V AC Distribution Panels from the Normal AC Power source ->

Rectified to DC and then inverted back to AC. Since backup DC Power is lost, normal AC Power will still be availaMe and the Distribution Panel will be powered as it normally is.

D. - INCORRECT - Panel 1AJ481 is not de-energized.

Question Source New Question History:

SXD review - 7/22 - OK 0

Memory Level Comprehension LRvel

Al.01 Ability to predict and/or monitor changes in parameters Battery associated with operating the DC Electrical distribution controls including (CFR:41.5/45.5) chargingdischarging rate Question "Control Room annunciator D3-F2 '1 25VDC SYSTEM TROUBLE" is alarming. Upon investigation the Operator determines that Digital Point D4631 '125VDC BATTERY CHARGER 1AD413" is in alarm and Battery Charger 1AD414 is INOP. On panel 10C650 the Operator reports the following:

125VDC Switchgear 1 OD41 0:

- Bus Voltage is reading 126 VDC

- Bus Current is reading 380 Amps The following is indicated on the 125VDC Battery Charger, 1AD413, control panel:

- DC Voltmeter is reading 126 VDC

- DC Ammeter is reading 360 Amps

- Timer switch is at 0

- FLOAT light is lit

- EQUALIZING light is off

- AC PW R ON light is lit

- DC Under Voltage light if off

- DC Over Voltage light is off

- Hi Voltage Shutdown light is off

- Insufficient Charging Current light is on WITH NO OPERATOR ACTION, which one of the following describes the expected 10D410 bus voltage trend and the reason for that trend?

The bus voltage will...

A B

lower because the bus load exceeds the charger's capacity.

rise because an equalizing charge is being provided.

rise because a malfunction of the float charge is indicated.

C lower because AC power is NOT being supplied to the charger.

Answer A

References INPO Question 24538 NOH01 DCELEC-00, DC ELECTRICAL DISTRIBUTION, p25-26,

p. 19-20 Justification References during Exam None A - CORRECT - with Switchgear Load > Charger Oulput voltage will lower over time B - INCORRECT - Equalizing Carge is NOT being provided with Timer switch at 0.

C - INCORRECT - Float charge is malfunctioning because charge voltage should be > bus voltage, however this will cause voltage to lower, not rise over time.

D - INCORRECT - AC on and float equalize lights indicate charger has AC power

Question Source Mod 0

Memory Level Comprehension Level Question History:

SXD review - 7/22 - OK

~

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref (iZuestion 49 1 wr#

Group#

1 Importance 2.9 300000 Instrument Air A3.02 Ability to monitor automatic operations of the Instrument Air including 41.7145.5)

Air temperature (CFR Quest ion Hope Creek is at 100%

Instrument Air status is as follows:

00K107, Service Air Compressor - Disassembled for Compressor work 10K107, Service Air Compressor - Tripped due to Low Lube Oil Pressure - currently being investigated 10K100, Emergency Instrument Air Compressor - Running Instrument Air Pressure - 90 psig stable A SACS/TACS AUTO ISOLATION alarm is received on low pressure.

The Operators take the Mode Switch to shutdown and stablize the plant at a Reactor level of +35" (lowest level = +lo").

Assuming no operator actions are taken and Instrument Air loads after the trip equal Instrument Air loads before the trip, what effect will this have on the Instrument Air system.

~

A It will have no effect on the Instrument Air System, instrument air pressure should be - equal to pre-trip value.

B Discharge air temperature will increase until the Air Compressor trips on Discharge Air Temperature high, instrument air pressure will be lower than pre-trip value.

Cooling water supply flow will decrease until the Air Compressor trips on Low Cooling Water Supply pressure, instrument air pressure will be lower than pretrip value.

~~~~

~

Reactor water level dropping to 10" causes the Air Compressor to trip on Low RPV Level, instrument air pressure will be lower than pre-trip value.

Answer A

References NOH01 INSAIR-01, INSTRUMENT AIR SYSTEM. P.13-14 Justification References during Exam None A - CORRECT - Since ElAC is running and it is coded by RACS and trips on low RPV level of -38', a loss of TACS should have no effect on EAlC and instrument air pressure should remain constant.

B - INCORRECT - ElAC is cooled by TACS, plausible distractor, if candidate thinks cooling water is isolated to compressor, discharge air temperature would increase and may cause compressor trip.

C - INCORRECT - ElAC is cooled by TACS, plausible distractor, if candidate thinks cooling water is isolated to compressor, cooling water supply flow would decrease and may cause compressor trip.

D - INCORRECT - RPV level must drop to -38' to cause ElAC to trip.

Question Source New Memory Level

&?I Comprehension Level Question History:

SXD Review 7/21 - LOD - 1.O - rewrite to make more difficult W4 - Re-wrote question.

IQuestion 50 Exam-Cross-Ref Hope Creek RO Exam - N O ~

2005

[Tier#

Group#

1 Importance 2.5

_ _ _. _ _ ~ _

30oooO Instrument Air K5.01 Knowledge of the operational implications of the following concepts as they apply to the Instrument Air Air compressors (CFR:41.5/ 45.7)

Question Which statement below describes the operation of the Emergency Instrument Air Compressor if the controls are aligned for the Auto Mode.

A B

The cornpressor auto starts at 100 psig in the Emergency Instrument air receiver and loads at 100 psig and unloads at 110 psig, once auto started the compressor will run continously.

The compressor auto starts at 70 psig in the Emergency Instrument air receiver and loads at 70 psig and unloads at 85 psig. compressor will auto stop if running unloaded for 45 minutes.

The compressor auto starts at 85 psig in the Emergency Instrument air receiver and loads at 85 psig and unloads at 100 psig, compressor will auto stop if running unloaded for 45 minutes..

The compressor runs CONTINUOUSLY and maintains85-100 psig by loading and unloading.

D Answer References Hope Creek Question - Q54114 Justification References during Exam None CORRECT: The compressor auto starts at 85 psig in the Emergency Instrument air receiver and loads at 85 psig and unloads at 100 psig.

INCORRECT: The compressor starts and loads at 70 psig and unloads at 85 psig. No: 85 to 100 INCORRECT: The compressor auto starts at 100 psig and maintains pressure between 100 and 1 10 psig. Wrong values. No: 85 to 100 INCORRECT: The compressor runs continuously and maintains85-100 psig by loading and unloading. Does not run continuously. on an AUTO Start if it runs unloaded for 45 mins the compressor will STOP Questwn Source Bank Memory Level Comprehension Level Questwn History:

SXD review 7/21 - OK MB - 7/28 - Need to add references

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref p r #

2 Group#

1 1 PI Importance 3.3 223002 PClSlNuclear Steam Supply Shutoff K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT Nuclear boiler instrumentation (CFR: 41.7 I ISOLATION SYSTEM NUCLEAR STEAM SUPPLY SHUT-OFF 45.7)

Question While operating RHR in shutdown cooling, reactor water level transmitter LT-NOBOA fails downscale.

SELECT the response of the RHR shutdown cooling supply valves, HV-F008 and HV-FOOS.

A 60th RHR shutdown cooling supply valves will automatically close.

B Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will close if low level is sensed by LT-N0806.

Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will close if LT-N080C fails downscale.

~~~

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Neither RHR shutdown cooling supply valve will change position automatically.

D Hope Creek Question - (253932 NOH01 RHRSYSC-03, RESIDUAL HEAT REMOVAL SYSTEM, P.30 Answer References Justification References during Exam None

- 60th RHR shutdown cooling supply valves will automatically close. -Incorrect - the trip must occur in both channes

'a* and 'b'/"c" and "d" to cause any isolation

- Neither RHR shutdown cooling supply valve will change position automatically. Correct - the trip must occur in both channel 'A' and "B" to cause an isolation

- Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will close if Level 3 is sensed in the '6" NSSSS logic. -Incorrect - the trip must occur in both channels to cause any isolation. Only one would close and only when the second signal is received.

- Only one of the RHR shutdown cooling supply valves automatically close and the second RHR shutdown cooling supply valve will close if Level 3 is sensed in the 'C" NSSSS logic. -Incorrect - the trip must occur in both channels to cause any isolation Question Source Bank E] Memory Level Comprehension Level Question History:

SXD review 7/21 - OK

Hope Creek RO Exam - Nov 2005 Exam-Cross-Re f

[Tier#

2 Group#

2 1 Importance 2.9

~

201006 RWM K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the RWM Rod Position indication Question There is a Control Rod with an inoperable notch position reed switch. When looking at the Rod Worth Minimizer display screen for that rod, how would it's position be indicated?

A RWM would display a suggested substitute position B

RWM would display a default value of *--*

RWM would display the last known good position C

RWM would display a default value of '00" D

Answer A

References INPO Question 1885 Justification References during Exam None A - CORRECT - per Lesson Plan - If a control rod is moved to a position with a failed reed switch, the RWM program wil1:a)Allow a single notch insert or withdraw permissive to allow the control rod to be moved to verify its actual position. b)Suggest to the operator a substitute position, which is its calculated inferred position.

B - INCORRECT - See "A' C - INCORRECT - See 'A" D - INCORRECT - See 'A' NOHOlRODMIN-01, ROD WORTH MINIMIZER p.15 Question Source Mod

~

Memory Level Comprehension Level Question History:

SXD review - 7/22 - Had questions talk to Archie about what would be displayed. Perhaps change imp notch position to a given position (ie. 12). If you pull rod from 10 to 12 and position 12's reed switch is INOP is 12 displayed.

Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 p e r #

Group#

I Importance 3.8 Mode 21 9Ooo RHWLPCI: TorudPool Cooling K4.03 Knowledge of RHWLPCI Torus/Pool Cooling Mode design feature(s) and or interlocks which provide for the following Unintentional reduction in vessel injection flow during accident conditions Question You have the following plant conditions:

o Drywell pressure 3.2 psig o Dlywell temperature 170°F o Suppression Pool pressure 1.a psig o Reactor water level

+ 25 inches The plant has scrammed on high Drywall pressure and the actions of both Primary Containment Control and RPV Control are being carried out.

The RHR system was in a normal lineup at the beginning of the transient and all automatic actions occurred as designed.

The CRS orders Suppression Pooling Cooling started on the '"A"RHR Loop. Which of the following switch manipulations will have to be performed in order to start Suppression Pool Cooling On the "A"" RHR Loop IAW HC.OP-SO.BC-0001, RHR System Operation?

o Suppression Pool temperature 96°F A

AUTO OP OVRD must be pressed on BC-HV-F017A, RHR LOOP A LPCl INJ MOV before valve can be closed. Once valve is closed then BC-HV-F024A, RHR LOOP A TEST RET MOV can be opened by depressing it's OPEN pushbutton.

BC-HV-F017A, RHR LOOP A LPCl INJ MOV must be closed by depressing it's closed pushbutton. Once F017A is closed then BC-HV-F024A. RHR LOOP A TEST RET MOV can be opened by depressing it's OPEN pushbutton.

B AUTO OP OVRD must be pressed for BC-HV-F017A, RHR LOOP A LPCl INJ MOV prior to depressing it's CLOSED pushbutton. Once F017A is closed then AUTO CL OVRD must be pressed for BC-HV-F024A, RHR LOOP A TEST RET MOV prior to depressing it's OPEN pushbutton.

AUTO CL OVRD must be pressed on BC-HV-FO17A. RHR LOOP A LPCl INJ MOV before valve can be closed. Once valve is closed then AUTO OP OVRD must be pressed on BC-HV-F024A, RHR LOOP A TEST RET MOV prior to opening F024A.

References INPO Question 2069 HC.OP-SO.BC-0001 (a) - Rev. 40, RESIDUAL HEAT REMOVAL SYSTEM OPERATION, p. 23, Note 5.5.5 Answer Justification References during Exam None A - INCORRECT - AUTO CL OVRD must be pressed on F024A before valve can be opened with LPCl initiation signal present.

6. INCORRECT - must depress AUTO OP OVRD for F017A prior to closing F017A with LPCl signal present C. CORRECT - per Procedure Note 5.5.5 - If a LPCl Initiation signal is present, the AUTO OP OVRD must be pressed on BC-HV-F017A(B) RHR LOOP A(B,C,D) LPCl INJ MOV, before the valve can be closed. The AUTO CL OVRD must be pressed on BC-HV-F024A(B) RHR LOOP A(B) TEST RET MOV, and BC-HV-F017A(B) must be closed before BC-HV-F024A(B) can be opened.

D. INCORRECT - Must Depress AUTO OP OVRD on F017A not AUTO CL OVRD Question Source Mod 0

Memory Level 0

Comprehension Level

Question History:

SXD Review 7/22 - verify pushbutton labels are correct

Hope Creek RO Exam - Nov 2005 Exam-Cross-Ref IQuestion 57 I Tier#

Group#

I Imuortance 4.2 23900 1 Main and Reheat Steam A3.01 Ability to monitor automatic operations of the Main and Reheat Isolation of main steam system including system (CFR:41.7/45.5)

Question The plant is shutting down for a refueling outage.

Current plant conditions are as follows:

Mode Switch - STARTUP Reactor Power - 4%

Reactor Pressure - 1000 psig Reactor Level - 35" "A' RFP running Both Recirc pumps running Condenser vacuum - 3.5" abs 3 Circ Water Pumps running All MSIV's open An event occurs:

3 Minutes later plant conditions are as follows:

Mode Switch - SHUTDOWN Reactor Power - All Rods inserted Reactor pressure - 700 psig decreasing Reactor Level - (-50" lowering)

Condenser Vacuum - 23" abs Degrading Based on the above conditions and assuming no operator actions, what is the status of the MSIV's and explain the reason for that status.

A MSIV's all OPEN - No automatic closure signal exists B

MSIV's all CLOSED - due to 1 Automatic Closure signal - Low Reactor Pressure MSIV's all CLOSED - due to 1 Automatic Closure signal - Low Condenser Vacuum C

MSIV's all CLOSED - due to 2 Automatic Closure signals - Low Reactor Pressure and Low Condenser Vacuum D

Answer References NOH01 MSTEAMC-02, MAIN STEAM SYSTEM p.24 Justification References during Exam Figure of NSSS A - INCORRECT - Condenser Vacuum of > 21.5" will cause MSIV's to close. Plausible distractor - this isolation can be bypass with a keylock switch.

6 - INCORRECT - Low Reactor Pressure MSlV closure signal is bypassed when Mode Switch is NOT in RUN C - CORRECT - Low Condenser vacuum setpoint of 21.5" has been reached and limit has not been bypassed.

D - INCORRECT - Low Reactor Pressure MSlV closure signal is bypassed when Mode Switch is NOT in RUN Question Source New 0

Memory Level Comprehension Level

Question History:

SXD Review 7/21 - LOD 1.O re-write question 6J4 - Wrote new question

Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 272000 K5.01 Importance 3.2 Radiation Monitoring Knowledge of the operational implications of the following concepts as they apply to the Radiation Monitoring Hydrogen injection operation's effect on process radiation indications Question "The plant was operating at full power with indicated H2 injection flow at 10 SCFM, when FE-601 (Flow Input to Hydrogen Flow Controller - FIC-601) fails LOW (ie. A LOW flow is INPUT into FIC-601).

Which of the following describes the expected result?

The Hydrogen Flow Control Valves will..."

A open rapidly resulting in rising Main Steam Line radiation levels.

B fail as is with no adverse consequences open rapidly resulting in a "LOW Recirc dissolved Oxygen level" alarm.

C close rapidly resulting in 'HIGH Recirc dissolved Oxygen level' alarm.

D References INPO Question 8753 NOH01 HWC100-01, HYDROGEN WATER CHEMISTRY M-101-0 sht 1 & 2 Answer A

INJECTION SYSTEM, p. 12 Justification References during Exam None A - CORRECT - FIC-601 attempts to maintain a certain H2 flow to the Secondary Condensate Pumps, when this flow input fails LOW - FIC-601 will attempt to raise H2 flow by opening the H2 Flow Control valves, opening these valves will result in Rising Main Steam Line Radiation Levels.

B - INCORRECT - H2 FCV's will open C - INCORRECT -While the FCV's will open rapidly, there is NO Low Recirc Dissolved Oxygen Level alarm.

D - INCORRECT - FCVs will open.

Question Source Mod 0

Memory Level Comprehension Level Question History:

SXD review - 7/27 - Minor editorial changes

~~

Exam-Cross-Re f LQuestion 68 L Tier f 3

Group #

I Importance 3.4

~

2.1.33 Generic Ability to recognize indications for system operating parameters which are entry level condition for Technical Specifications (CFR: 43.2 143.3 145.3)

Hope Creek RO Exam - Nov 2005 Question During Plant startup the following conditions are observed:

TIME RPV Pressure 0700 172 psig 0715 191 psig 0730 211 psig 0745 233psig 0800 373psig Which one of the following is the latest time at which heatup must be secured in order to prevent exceeding the Technical Specification limit for heatup at the CURRENT heat up rate?

B 0815 0830 C

0845 D

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Answer References Hope Creek Question - (356983 Steam Tables Tech Spec Justification References during Exam Steam Tables Justification 172 psig = 186.7psia=376F 191psig=205.7 psia = 384°F 21 1 psig - 225.7 psia = 392F 233 psig = 247.7 psia = 400F 373 psig = 387.7 psia = 442F-This gives a 42F change in 15 mins. Current heatup rate is 42F every 15 min (168 degreedhr). 0815 - Correct-At this rate we must terminate the HIU by 0815 to keep from exceeding the allowable heatup, we would be at 484°F (this would be 100 degreedhr).

Question Source

=ank 0

Memory Level Comprehension Level Question History:

SXD review 7/21 - OK

IQuestion 69 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 Question The plant is shutdown with '6' RHR in shutdown cooling, OPCON 4. Inservice stroke time testing needs to be performed on the discharge valve of the 'A' recirculation pump prior to commencing startup.

What precautionsllimitiations exist to allow/prevent this evolution to take place?

A As long as RPV vessel level is pegged high on all Narrow Range instruments, Shutdown cooling may be secured and the recirculation discharge valve stroked without potential problem of loss of decay heat removal and vessel stratification.

System Operating procedures for both Recirculation system and RHR system prohibit the opening of Recirculation pump discharge valves while RHR is in Shutdown Cooling, to prevent potential core bypass flow and vessel stratification.

B This evolution can only be performed after the '8' Recirc pump is placed in service and establishment of forced circulation through the vessel is assured.

_ _ _ ~

~

Prior to stroking the discharge valve on ' A Recirculation pump, the suction valve must be verified closed, and the suction valve's power supply breaker open.

Answer References Hope Creek Question - Q56375 HC.OP-I0.U-0002 section 3.2.5 Justification References during Exam None Justification:IAW HC.OP-I0.U-0002 section 3.2.5

,'Prior to stroking the discharge valve on 'A' Recirculation pump, the suction valve must be verified closed, and the supply breaker opened."-Correct

."This evolution can only be performed after the '6' Recirc pump is placed in service and establishment of forced circulation through the vessel is assured.'- Incorrect-The 'can only' distractor is wrong because the word "only " is used, along with the combination of RHR and Recirc pump combinations would still require the suction valve closed while stoking the valve

,'System Operating procedures for both Recirculation system and RHR system prohibit the opening of Recirculation pump discharge valves while RHR is in Shutdown Coding, to prevent potential core bypass flow and vessel stratification.' - Incorrect-The 'SOP' distractor is wrong because the IO allows this condition and applicable exception to the SO guidance

.'As long as RPV vessel level is pegged high on all Narrow Range instruments, Shutdown cooling may be secured and the recirculation discharge valve stroked without potential problem of loss of decay heat removal and vessel stratification." - Incorrect-The 'RPV vessel level' is wrong because minimum level for natural circulation is +80" which is well above the Narrow Range detector capability to read, and does not assure the appropriate level.

Question Source Bank Memory Level 0

Comprehension Level

Question History:

SXD review 7/21 - OK

IQuestion 71 Exam-Cross-Ref Hope Creek RO Exam - Nov 2005 VI I Tier#

3 Grouu#

1 1

I Importance 2 6

~

l O S R 0 1 2.3.1 Generic Knowledge of 10 CFR 20 and related facility radiation control requirements (CFR: 41.12 143.4. 45.91 45.10).

~

Question Radiation Protection technicians have surveyed the Refuel Floor Reactor Head Laydown Area during an outage and obtained the following results:

.Highest Area Dose Rate one foot from any source in the room: 72 mrlhr

.Airborne Concentration: 0.1 5 DAC

,Smear Results: 750 dpm/l00 cm2 gamma Based on these results the area should be posted as a:

I.

II.

111.

IV.

V.

A Radiation Area High Radiation Area Very High Radiation Area Contaminated Area Airborne Radioactivity Area I, and V B

I, IV, andV I and IV C

II and IV D

Answer A

References Hope Creek Question - Q76884 (Modified slightly)

NC.NA-AP.ZZ-0024, rev 13, p.23 Justification References during Exam None A - CORRECT - Airborne rad area > 10% or.10 DAC B - INCORRECT - Not a Contaminated Area - must be > 1000 dpm/lOcm2 C - INCORRECT - Not a Contaminated area and it is an Airborne Area D - INCORRECT - Not a High Radiation Area - must be > 100mrhr Question Source Bank Memory Level Comprehension Level Question History:

SXD review 7/21 - LOD 1.5 - evaluate writing a more difficult question Changed question out with another HC bank question that seems more difficult

IQuestion 75 p q Exam-Cross-Ref Hope Creek SRO Exam - Nov 200 (MSRO I 2.4.31 Question I Tier 4' 3

Group #

I Imuortance 3.3 Generic Knowledge of annunciators alarms and indications / and use of the response instructions. (CFR: 41.10 I 45.3)

The plant is in Mode 5 with the Spent Fuel Pool/ Cavity Gates installed. Preparations are underway to flood the reactor cavity when the following annunciators are received:

D1A5 - FUEL POOL LEVEL HllLO D1D5 - FUEL POOL COOLING SYS TROUBLE The Operator reports that Spent Fuel Pool level and Skimmer Surge tank level are -21.5 and lowering.

Radiation levels in the Spent Fuel pool are rising.

Which procedure should be entered and what is the preferred method to keep the fuel bundles covered?

- - Hope Creek - is it acceptable to directly enter AB or would you prefer I state per the ARP.

A HC.OP-ABCOOL-0004, FUEL POOL COOLING should be entered and Spent Fuel Pool level should be raised using the Fuel Pool Cooling System.

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g HC.OP-AB.COOL-0004, FUEL POOL COOLING should be entered and Spent Fuel Pool level should be raised using the RHR system.

HC.OP-EO.U-0103/4, REACTOR BUILDING AND RADIOACTIVE RELEASE CONTROL, should be entered and Spent Fuel pool Level should be raised using the Fuel Pool Cooling System.

HC.OP-EO.ZZ-0103/4. REACTOR BUILDING AND RADIOACTIVE RELEASE CONTROL, should be entered and Spent Fuel pool Level should be raised using the RHR System.

References HC.OP-ARZ-0013, Overhead Annunciator Window Box D1, p.

36 and 89 NOH01 FPCCOO-03, Fuel Pool Cooling and Cleanup System, p.17 Answer Justification References during Exam None A - INCORRECT - Cannot use Fuel Pool Cooling for makeup because Fuel Pool Cooling pumps trip on Low skimmer surge tank level of 22'.

B - CORRECT - Based on the Annunciator Response Procedure for Low Fuel Level - AB.COOL-0004 should be entered and level raised using the RHR system because the Fuel Pool Cooling System is not available.

C - INCORRECT - even though radiation levels are rising, based on the Annunciator response procedure AB.COOL-0004 should be entered.

D. INCORRECT - even though radiation levels are rising, based on the Annunciator response procedure ABCOOL-0004 should be entered.

Question Source Mod 0

Memory Level E) Comprehension Level Question History:

SXD review 7/27 - Not SRO level - re-write Rewrote question - 7/29 - somewhat based on INPO Question 22362 JD - W - Why are C.D plausible MB - #3 - Changed AB.CONT-0005, Irradiated Fuel damage to EO.U-0103/4 since Radiation levels in the Reactor Building are rising and operator may be concerned about reactor building release.

Hope Creek SRO Exam - Nov 200 Exam-Cross-Ref

[Tier#

Group#

1 3.8 295003 Partial or Complete Loss of AC / 6 AG2.1.32 Ability to explain and apply system limits and precautions (CFR 41.10/43.2/ 45.12)

Quest ion Hope Creek was operating at 30% power when a Station Blackout (loss of all onsite and offsite power) occurred causing a Reactor Scram.

Current plant conditions are as follows:

D w e l l temperature - 300°F decreasing slowly RPV pressure - 273 psig decreasing slowly Reactor Power - all rods fully inserted Reactor level - (-100' decreasing)

RClC - tagged out and disassembled HPCl - tripped on overspeed and will not restart

'A' EDG - tagged out for maintenance "B' EDG - running unloaded - output breaker failed open on anti-pump circuitry "C" EDG -tripped on Bus differential overcurrent

'D' EDG - failure to start - low air pressure -20 psig Based on these conditions, the Control Room Supervisor shall:

A direct the NE0 to reset the Bus differential overcurrent on the 'C' EDG and restart the "C" EDG.

~.

B direct the RO to depress the TRIP pushbutton on the "6' EDG output breaker and verify output breaker closes enter procedure HC.OP-EO.=-0202, Emergency Depressurization based on high Drywell temperature.

C enter procedure HC.OP-E0.U-0202, Emergency Depressurization before Reactor Water Level decreases to -

129" HC.OP-AB.U-0135, Station Blackout// Loss of Offsite Power//

Diesel Generator Malfunction p. 2 Answer References Justification References during Exam None A - INCORRECT - bus differential current should not be reset without electrical maintenance determining and correcting the cause.

B - CORRECT - per HC.0P.AB.U-0135, Station Blackout p. 18 step 5.16 - The Anti-pump circuitry on the DIG output breaker could cause the output breaker to fail open. To load the D/G under this condition the operator must depress the TRIP push-button (even though the breaker is already tripped) to reset the logic. When the TRIP push-button is released, then the breaker will close and the D/G will load.

C - INCORRECT - Emergency Depressurization procedure should not be entered until DW temperature exceeds 340°F and current drywell temperature is decreasing.

D - INCORRECT - Emergency Depressurization procedure should not be entered until is less than -129' but before level decreases to -185" Question Source New 0

Memory Level E] Comprehension Level Question History:

SXD review 7/27 - Not SRO level - re-write 8/2 - re-wrote question

leuestion 80 1 Exam - Cross-Re f Hope Creek SRO Exam - Nov 200 b r #

Group#

1

@! SRO Zrnportance 3.3 295028 High Drywell Temperature I 5 EG2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR 45.3)

Question Given the following conditions:

.A small steam leak has occured in the drywell causing a reactor scram

.Two control rods are at position 4PV level

+30 inches

.RPV pressure 920 psig

.Suppression pool level 75 inches

.Suppression pool temperature 80 "F

.Drywell pressure 3 psig

.Average drywell temperature Suppression chamber pressure 3 psig Which of the following describes the next operator action(s) in accordance with the Emergency Operating Procedures?

06 330 "F and rising at 1°F per minute A

Shutdown the Reactor Recirculation Pumps and Drywell Cooling Fans and initiate one loop of drywell spray.

B Verify all injection into the RPV except SLC, CRD and RClC is terminated and prevented and then emergency depressurize the reactor.

Rapidly depressurize the reactor using the main turbine bypass valves.

c Initiate suppression chamber sprays and commence a normal reactor cooldown. (Less than 90 F per hour)

D Answer References Hope Creek Question Q56045 HC.OP-EO.=-0102 Bases, step DWn-5 Justification References during Exam None A - INCORRECT -.Shutdown the Reactor Recirculation Pumps and Drywell Cooling Fans and initiate one loop of drywell spray.-incorrect-Cannot DW Spray since outside of DWT-P curve.

B - CORRECT -.Verify all injection into the RPV except SLC, CRD and RClC is terminated and prevented and then emergency depressurize the reactor.-correct-EOP-0202 step ED-3 C - INCORRECT -.Rapidly depressurize the reactor using the main turbine bypass valves.-incorrect-EOP-1O1A prevents use of BPVs in this situation D - INCORRECT -,Initiate suppression chamber sprays and commence a normal reactor cooldown. (Less than 90 F per hour)-incorrect-must stabilize pressure until SID under all conditions without Boron Question Source Bank Memory Level

@I Comprehension Level Question History:

SXD review - 7/29 - OK

Hope Creek SRO Exam - Nov 200 Exam-Cross-Ref 83 I Tier#

Group#

1 Importance 3.7

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295009 Low Reactor Water Level / 2 AA2.02 Ability to determine and interpret the following as they apply to Low Reactor Water Level (CFR: 41.101 43.5 145.13)

Steam flow/ feed flow mismatch Question Hope Creek is operating at 75% power reducing power to remove the 6A Feedwater Heater from service due to a problem on the Bleeder trip valve with the following conditions:

Feedwater control is in 3 element control A Steam Flow indicates - 2.0 E6 Ibshr B Steam Flow indicates - 2.0 E6 Ibshr C Steam Flow indicates - 2.0 E6 lbslhr D Steam Flow indicates - 2.0 E6 Ibshr FW flow (N001A) indicates - 4.0 E6 Ibshr FW flow (N001B) indicates - 4.0 E6 lbshr Reactor Water level - 33" stable Reactor Pressure - 1000 psig stable Generator MW - 750 MW An event occurs.

1 Minute after event initiation the following conditions are observed:

A Steam flow indicates - 1.7 E6 Ibshr B Steam flow indicates - 2.0 E6 Ibshr C Steam flow indicates - 2.0 E6 Ibdhr D Steam flow indicates - 2.0 E6 lbslhr FW flow (NO01 A) indicates - 3.9 E6 Ibdhr FW flow (N0016) indicates - 3.9 E6 Ibdhr Reactor Water level - 38' and lowering slowly Reactor Pressure - 990 psig stable Generator MW - 710 MW Based on the above conditions, what event has happened and what procedure shall you direct the operators to respond to the event?

- Hope Creek to supply valid numbers (Steam Flow/Feed flow or 1'11 do it when I go back down there.

A B

"A" Steam line's input to Total Steam flow has partially failed causing Steam flow/Feed flow mismatch, go to procedure HC.OP-AR.ZZ-0007 window F-1, 'DFCS ALARMRBL" "A" Main Stop Valve has failed closed, go to procedure HC.OP-AB.BOP-0002. MAIN TURBINE A Safety has opened on the 'A" steam line, go to procedure HC.OP-AB.RPV-0006,SAFETY RELIEF VALVE C

The 6A Feedwater heater bleeder trip valve has failed, go to procedure HC.OP-AB.U-0001, TRANSIENT PLANT CONDITIONS Answer References HC.OP-ABRPV-0006, Safety Relief Valve p.1 NOH01 MSTEAMC-02. MAIN STEAM SYSTEM, Justification References during Exam None

A - INCORRECT - While ' A steam line's input to Total Steam flow could cause the difference in indicated Steam Flow, it would not cause Generator MW to decrease.

B - INCORRECT - While "A' Main stop valve failing closed would cause a decrease in MW. it would not cause Reactor pressure to decrease, it would increase.

C - CORRECT - A safety on "A" steam line would cause, " A 3 steam line flow to decrease, MW to decrease and Reactor Pressure to decrease.

D - INCORRECT - 6As bleeder trip valve going closed would cause MW to go up not down.

Question Source New 0

Memory Level Comprehension Level Question History:

SXD review 7/27 - LOD 1 - re-write 8/1 - re-wrote question - MB

Hope Creek SRO Exam - Nov 200 Exam-Cross-Ref 95 Group#

I 2.9

-~ -

Importance

~

_ _ _ ~

2.1.34 Generic Ability to maintain primary and secondary plant chemistry within allowable limits (CFR: 41.10 I43.5 I45.12)

~

Question The plant was operating at 20% power. Plant Chemistry reported to the Main Control Room the following chemistry parameters:

- Reactor pH 8.8

- Reactor Water conductivity

- Reactor Water chlorides 11 micromhos/cm 150 ppb Six hours later with the plant in OPCON 2, Chemistry reports the following:

- Reactor pH 6.5

- Reactor Water conductivity

- Reactor Water chlorides 0.9 micromhos/crn 150 ppb Which one of the following actions is appropriate for these plant conditions?

A Be in OPCON 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and OPCON 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

B Return to OPCON 1 where chemistry would be back in spec.

Stay in OPCON 2 and restore chlorides to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in OPCON 3 within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and OPCON 4 within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore Chlorides to within spec within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months or perform an engineering evaluation.

D References INPO Question 24577 UFSAR 5.2.3.2.2.2 and UFSAR Table 5.2-8 Answer A

Justification References during Exam UFSAR 5.2.3.2.2.2 and Table 5.2-8 A - CORRECT - per ACTION a. - with conductivity exceeding lOmmho/cm be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B - INCORRECT - plausible because based on given conditions for OPCON 2, plant chemistry would be in spec if plant returned to OPCON 1.

C - INCORRECT - plausible if only look at Action b.

D - INCORRECT - plausible if only look at Action c.2 Question Source Mod 0

Memory Level E] Comprehension &vel Question History:

SXD review 7/27 - Talk to licensee ensure correct answer is correct and once conductivity is.z limit. they exit the condition and can retum to power.

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buestion 98 J Exam-Cross-Ref Hope Creek SRO Exam - Nov 200 p i e r #

3 Group#

3.1

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@ SRO Importance 2.3.4 Generic Knowledge of radiation exposure limits and contamination control / including permissible levels in excess of those authorized. (CFR: 43.4 / 45.10)

Question Lowest authorization necessary to receive increase a worker's dose control level to 3500mrem/yr per NC.NA-AP.U-0024, Radiation Protection Program, is the responsibility of the A

Radiation Protection Supervisor B

Radiation Protection Manager Plant Manager C

~~

~

Emergency Director R

References JNPO Question 19298 NC.NA-AP.U-0024, RADIATION PROTECTION PROGRAM - p.

27 Answer Justification References during Exam None A - INCORRECT - Radiation Protection Supervisor may only increase dose level to 3000 mrem/yr 8 - CORRECT - Radiation Protection Manager may raise dose control level to 4000 mrem/yr C - INCORRECT - While Plant Manager may raise level to 4750 mrem/yr, he is not the Lowest aulhorization necessary.

0. - INCORRECT - While ED may approve Emergency Doses, he is also not the Lowest authorization necessary.

Question Source Mod hd Memory Level 0

Comprehension Level Question History:

SXD review 7/27 - too easy - LOD 1

Exam-Cross-Ref Hope Creek SRO Exam - Nov 200 3

Group#

lmuortance 4

~~

2.4.22 Generic Knowledge of the bases for prioritizing safety functions during abnormallemergency operations (CFR: 43.5 / 45.12)

Question During an ATWS, automatic initiation of the Automatic Depressurization System (ADS) is inhibited to prevent which one of the following?

A A power excursion due to low pressure ECCS injection

~~

B Large irregular neutron flux oscillations Exceeding 1 10°F Suppression Pool Temperature before boron injection C

Causing a Pressurized Thermal Shock to the Reactor Vessel D

Answer A

References INPO Question 24595 HC.OP-EO.ZZ-OlOlA, ATWS - RPV CONTROL, P. 18 Justification References during Exam None A - CORRECT - Per EOP lOlA bases - Further, rapid and uncontrolled injection of large amounts of relatively cold, unborated water from low pressure injection systems may occur as RPV pressure decreases to and below the shutoff heads of these pumps. Such an Occurrence would quickly dilute in-core boron concentration and reduce reactor coolant temperature. When the reactor is not shutdown, or when the shutdown margin is small, sufficient positive reactivity might be added in this way to cause a reactor power excursion large enough to severely damage the core.

B - INCORRECT - ADS initiation would NOT cause flux oscillation but rather a rapid reduction in core power due to voids C - INCORRECT - This may or may NOT be true but it is NOT the reason for inhibiting ADS D - INCORRECT - While an ADS actuation will cause a Thermal Shock to the vessel, the vessel will be de-pressurized so you will not have a PTS concern Question Source Mod E] Memory Level 0

Comprehension Level Question History:

SXD review 7/27 - OK

buestion 12 I Exam-Cross-Ref 295025 High Reactor Pressure 13 EAl.02 Ability to operate and I or monitor the following as they apply to High Reactor Pressure Hope Creek RO Exam - Nov 2005 Reactorflurbine pressure regulating system :(CFR:

41.7145.5145.6)

Question Given the following conditions:

.The plant is operating at power 91 % power.

.The steam pressure input signal to the "A" EHC regulator fails downscale.

.No operator actions are taken.

Which of the following is the response of Reactor pressure to the conditions above'?

- - NEED DIFFERENT QUESTION DUE TO DIGITAL EHC ***

A Pressure rises 3 psig and stabilizes.

B Pressure lowers to the MSlV isolation setpoint.

Pressure lowers 3 psig and stabilizes.

C

~

Pressure rises to the scram setpoint D

Answer A

References Hope Creek Question - (261768, HC.OP-AB. RPV-0005 Automatic Actions Justification References during Exam None CORRECT - Pressure rises 3 psig and stabilizes. This failure causes the TCVs to throttle closed, raising RPV pressure. When the actual pressure increase overcomes the 3 psi bias on the "6" pressure regulator, the "6' regulator will re-open the TCVs and maintain pressure 3 psig higher than the 'A" regulator.

INCORRECT - Pressure lowers 3 psig and stabilizes. This failure causes the TCVs to throttle closed, raising RPV pressure.

INCORRECT - Pressure lowers to the MSlV isolation setpoint. This failure causes the TCVs to throttle closed, raising RPV pressure.

INCORRECT - Pressure rises to the scram setpoint. When the actual pressure increase overcomes the 3 psi bias on the "6' pressure regulator, the '6" regulator will re-open the TCVs and maintain pressure 3 psig higher than the 'A' regulator.

Question Source Bank Question History:

0 Memory Level 0

Comprehension Level

ES-201 Examination Outline Quality Checklist Form ES-201-2 Initials a I b' I c#

Task Description VG d 57 Wd9 d 5-4

a. Verify that the outline(s) fit@) the appropriate model, in accordance with ES-401.

b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.l of ES-401 and whether all WA categories are appropriately sampled.

c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

/ '

Facility:

Hope Creek Date of Examination:

11/28/05 Item

1.

W R

I T

T E

N

2.

S I

M U

L A

T 0

R

3.

W

/

T L

4.

G E

N E

R A

L Michael L. Brown /

a. Author

b. Facility Reviewer (*)

c. NRC Chief Examiner (#)

Steven Dennis /

d. NRC Supervisor Note:

v

\\

Rich Conte I

Independent NRC reviewer initial it&in Column "c"; chief examiner concurrence required.

~~~

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m 9 39

d. Assess whether the justifications for deselected or rejected WA statements are appropriate.

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients.

b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative mg ~~,

and quantitative criteria specified on Form.,

a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks distributed among the safety functions as specified on the form (2) task repetition from the last two NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria (1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified

ES-301 Transient and Event Checklist Form ES-301-5 I Facility: Hope Creek Date of Exam:

11/28/05 Operatinq Test No.:

Scenarios Instructions:

1.

Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must do one scenario, including at least two instrument or component (VC) malfunctions and one major transient, in the ATC position.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-1 basis.

2.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirement.

Author:

+P=

NRC Reviewer:

Appendix D Scenario Out1 i ne Form ES-D-1 Facility: Hope Creek Scenario No.:

1 Op-Test No.:

Examiners:

Operators:

)I Initial Conditions: 4% power, Reactor Startup in prowess, B EHC pump blocked for maintenance 11 Turnover:

A Reactor Startup is in proqress with IOP-3 completed up to step 5.3.21. Reactor power is approximately 4%. RClC is beinq operated for HC.OP-ST.BD-0002, RClC Pump Valve and flow test and should be completed within the next hour. 1 BP116 EHC pump is taqaed out for maintenance and will be out of service until a new pressure compensator arrives tomorrow.

Event Malf.

I No. I NO.

II I

Event Type*

R (CRS)

R (RO)

I (ALL) c (RO)

C(CRS)

M (ALL)

C (BOP)

C (CRS)

C (BOP)

C (CRS)

Event DescriDtion Withdraw control rods until 4 bypass valves open CRD Flow controller fails downscale in AUTO Loss of B MG Set Control Rod 22-35 inadvertently scrams (TS)

Steam Leak from RClC piping (

RClC isolation valves fail to close L

I1 *

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Out1 i ne Form ES-D-1 11 Facility: Hope Creek Scenario No.:

2 Op-Test No.:

11 Examiners:

Operators:

Initial Conditions: 80% power, middle of cvcle. A control rod sequence exchanqe has iust been performed and power is beincl raised back to 100%. The load dispatcher has requested a temporary hold at 80% power 11 Turnover:

SLC pump AP-208 has been taaqed out for a motor replacement and is expected back in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

OPRM system is INOP due to an existina 10CFR21 issue. The OPRM system is still functional but is considered INOP Der Tech Specs. No other equipment is Out of Service.

Maintain power at 80% until contacted bv Load Dispatcher, then raise power to 100°/~.

HC.OP-ST.BE-0002. Core Sprav Pump Loop A Full Flow Test is in proaress and completed up to step 5.23. (pump testinq). Complete HC.OP-ST.BE-0002 as soon as shift turnover is complete.

I Event Malf.

II 9

Event Type*

N (BOP)

I (BOP)

I (CRS)

R (RO)

FRO)

(CRS)

I (BOP) 3%

M (ALL)

I (RO) u2zL M(ALL)

I (BOP)

~

Event Description Perform HC.OP-ST.BE-0002, Core Spray Pump Loop A Full flow test. Core Spray Loop A discharge flow instrument fails during full flow test Power Increase Recirc Flow Inadvertent HPCl initiation S

\\

l l

a l

n I

o s

s FRVS fails to start T/nlc

- 1 t t t.

480 Volt Unit Substation 1 OB1 30 trips Broken SRV tailpipe, SRV Fails open, PSP function lost, E required ( w m e m & & - d e a d O G u d c)r-Ln

)

fl 1

?

RHR Spray Logic failure [

L o o f 3)

L (N)orrnal, (R)eactivity, (I)nstrurnent, (C)ornponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Event No.

1 2

3 4

5 6

Facility: Hope Creek Scenario No.:

3 (Spare)

Op-Test No.:

ll Malf.

Event Event No.

Type*

Description N (BOP)

N (CRS)

R (RO)

R (CRS)

I (RO)

A APRM fails I (CRS)

C (BOP)

C (CRS)

C (RO)

M (ALL)

C (BOP)

Start the 3rd RFP IAW HC.OP-SO.AE-0001 Commence load increase after starting RFP Drywell Chiller Compressor fails B Recirc Pump high vibration. Operator will trip Loss of off-site power due to storm A D/G m r e a k e r failure to Auto Close rcuTpvr 1 Examiners:

Operators:

M (ALL)

Initial Conditions:

oDerators are PreparinQ to place the third RFP in service IAW HC.OP-SO.AE-0001. Severe weather is predicted for the upcomincr shift.

Plant is at 80% power. middle of life, returnina to power after a mini-outaae, the Recirc suction piping leak. Small enough that crew can control parameters for loss of all high pressure feed

~

I Turnover:

ll Start the 3d RFP and increase power to 100%

I

Scenario ends with Emergency Depressurization and level restored above TAF.

lr -

I *

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

ES-301 Administrative Topics Outline Form ES-301-1 Conduct of Operations Facility:

Hope Creek Date of Examination:

1 1 /28/05 Examination Level : RO Administrative Topic (see Note)

Conduct of Operations N

Operating Test Number:

~~

Describe activity to be performed Check Drywell to Torus D/P during power operations per Daily Surveillance Log Procedure Change - Make a change to a procedure for Emergent work Rod Worth Minimizer Operability -

Equipment Control Radiation Control Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Enter and exit a High Radiation Area for a valve lineup. 2.3.10

Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 5 3 for ROs; I 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 2 l), (A)lternate Path (P)revious 2 exams (I 1 ; randomly selected)

I

ES-301 Control Room / In-Plant Systems Form ES-301-2 System / JPM Title

a. LPRM / Bypass failed LPRM (SE001) ( a1,:,- r;lsLcd@4 Facility:

HoDe Creek Date of Examination:

11/28/05 Exam Level (circle one): RO Control Room Systems* (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

Operating Test No.:

Type Code*

Safety Function D, S 7

b. FRVS / Manually start FRVS system (GU001)

c. Recirc Flow Control System / Reset a Recirc MG Set Scoop Tube Lockup (Alt. Path - Recirc speed inexplicably rises following reset)

(BS002)

d. Transfer station loads prior to shutting down Generator pc$)

e. PClS / Restart RWCU following Group Isolation

f. Swap FW Level Control, Single to 3-Element $:

g. HPCl-Startup HPCl in the CST to CST mode (BJ002)

D, s 9

M, A, S 1

6 S, N e >

N,&,

E 5

N, S 2

D, S, A 4

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h. Main Steam System - Main Steam System recovery following a N, s, L 3

Group 1 isolation.

In-Plant Systems* (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

I II 11 i. Main Steam1 Closing MSlV from Outside the Control Room

j. A/C Electrical / Startup a 20KVA Inverter

k. Control Rod Drive / Isolate a CRD HCU (BF006)

N, la 6

D, R 1

NUREG-1021, Revision 9 Type Codes (A)lternate path (4)

(C)ontrol room (D)irect from bank (4)

(E)mergency or abnormal in-plant (1)

(L)ow-Power (1)

(N)ew or (M)odified from bank including 1(A) (7)

(P)revious 2 exams (0)

(R)CA (3)

(S)imulator Criteria for RO / SRO-l I SRO-U 4-6 14-6 I 2-3

< 9 1 s 0 1 s 4 1 / 1 / 1 5 1 1 2 112 1 2 2 f 2 212 1

< 3 / 5 3 / s 2 (randomly selected) t 11.2 1 / 2 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Hope Creek Date of Examination:

11/28/05 Examination Level : SRO Operating Test Number:

(I Administrative Topic Conduct of Operations Radiation Control Emergency Plan Type Code*

Describe activity to be performed Determine SRWIRM overlap per procedure and Tech Specs Procedure Change - Make a change to a procedure for Emergent work Review and approve a clearance prior to maintenance Enter and exit a High Radiation Area for a valve lineup. 2.3.10 Classify an Emergency Event - 2.4.41 - May be done after a scenario using the simulator.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (I 3 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 2 l), (A)lternate Path (P)revious 2 exams ( 5 1 ; randomly selected)

ES-301 Control Room / In-Plant Systems Form ES-301-2 Outline

i. Main Steam/ Closing MSlV from Outside the Control Room N, R N, R D, R

j. N C Electrical / Startup a 20KVA Inverter

k. Control Rod Drive / Remove an HCU from service (BF006)

Facility:

HoDe Creek Date of Examination:

1 1 /28/05 Exam Level (circle one): SRO Control Room Systems" (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

Operating Test No.:

System / JPM Title 3

6 1

a. LPRM / Bypass failed LPRM (SE001)

b. FRVS / Manually start FRVS system (GUOOl)

Type Codes (A)lternate path (4)

(C)ontrol room (D)irect from bank (4)

(E)mergency or abnormal in-plant (1)

(L)ow-Power (1)

(N)ew or (M)oc'ified from bank including 1(A) (7)

(P)revio:@s 2 exams (0)

( W A (3)

(S)imulator

c. Recirc Flow Control System / Reset a Recirc MG Set Scoop Tube Lockup (Alt. Path - Recirc speed inexplicably rises following reset)

(88002)

Criteria for RO / SRO-I I SRO-U 4-6 14-6 1 2-3 5 9 1 s 8 1 s 4 1 / 1 / 1 z l / s l f 2 1 2 2 / z 2 / r 1 I 3 / 5 3 / I 2 (randomly selected) z 1 1 2 1 1 2 1

d. Transfer station loads prior to shutting down Generator

e. PClS / Restart RWCU following Group Isolation 11 f. HPCl - Startup HPCl in the CST to CST mode (BJ002)

~~~~

Type Code*

M, A, S Safety Function 7

9 1

ES-401 Written Examination Quality Checklist Form ES-401-6

a. Author

b. Facility Reviewer (*)

d. NRC Regional Supervisor 1 $J A

L-

c. NRC Chief Examiner (#)

C&VL.J DLdJU t

Note:

The facility reviewers initialskignature are not applicable for NRC-developed examinations.

Independent NRC reviewer initial items in Column c; chief examiner concurrence required.

FS-Ml Examinatian Outline Qualitv Checklist Form ES-201-2 Task Description Item

1.

w

a. Verify that the outline($ fit(s) the appropriate model, in accordance with ES-401.

Initials a

b' c#

WB \\i'n,f) bdL In 11 d.-Assess whether the justifications for deselected or rejected WA statements are appropriate.

R I

b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.l of ES-401 and whether all WA categories are appropriately sampled.

W& 14 0

a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks distributed among the safety functions as specified on the form (2) task repetiion from the last two NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of new or modified tasks meets or exceeds the minimums Specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria

~~

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2.

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients.

M U

L A

T 0

R 1 7

a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam sections.

b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

v#

5,o >q on the form.

b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is repeated from the last two NRC licensing examinations Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days.

c.

11 a. Author I

i

\\I

/

I I

Michael L. Brown I

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b.

c.

Assess whether the 10 CFR 55.41143 and 55.45 sampling is appropriate.

Ensure that WA importance ratings (except for plant-specific priorities) are at least 2.5.

Check the entire exam for balance of coverage.

f.

Assess whether the exam fits the appropriate job level (RO or SRO).

d. Check for duplication and overlap among exam sections.

e.

b. Facility Reviewer (*)

I.

c. NRC Chief Examiner (#)

d. NRC Supervisor ( ff) Rich Conte /

Steven Dennis /

J - //

flb 5) '> 3 fll$

>.h 5 ?

d

$0 5 0 n\\B 5 3 5 3 qfi j-0 5,

I Note:

Independent NRC reviewer initial items in Column "c"; chief examiner concurrence required.

ES-401 BWR Examination Outline Form ES-401-1 Note:

1.

Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each WA category shall not be less than two).

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 i 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

e.

Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

8.

On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

2.

3.

9.

ES-401 2

Form ES-401-1 WA Topic(s)

AK1.03 - Knowledge of the operational implications of the following concepts as they apply to the Partial or Complete Loss of Forced Core Flow Circulation: Thermal Limits

(CFR: 41.8 to 41.10 145.3) or Complete Loss of AC : Whether a partial or complete loss of A.C. Power has occurred:(CFR: 41.10 143.51 45.13)

AK3.01 - Knowledge of the reasons for the following responses as they apply to Partial or Total Loss of DC Pwr : Load shedding Plant Specific:(CFR: 41.5/41.10 I45.6

/45.13)

AG2.1.2 - Knowledge of operator responsibilities during all modes of plant operation AA2.05 - Ability to determine and interpret the following as they apply to Partial ES-401 Emeroe IR 3.6 3.9 2.6 3.0

/APE # /Name /Safety Function K

K K

A 1

2 3

1 295001 Partial or Complete Loss of Forced 1 0 0 0 Core Flow Circulation I 1 & 4 295003 Partial or Complete Loss of AC I 6 0 0 0 0 AK1.03 - Knowledge of the operational implications of the fallowing concepts as they apply to the SCRAM: Reactivity Control:(CFR: 41.8 to 41.10 /45.3) 295004 Partial or Total Loss of DC Pwr I 6 0 0 1 0 295005 Main Turbine Generator Trip I 3

295006 SCRAM I 1 1

0 0

0 29501 6 Control Room Abandonment I 7

0 0

295018 Partial or Total Loss of CCW I 8 0

0 0

0 3.7 295021 Loss of Shutdown Cooling I 4 AA2.04 - Ability to determine and interpret the following as they apply to Partial or Total Loss of CCW System Flow:(CFR: 41.10/43.5/45.13)

AA1.03 - Ability to operate and I or monitor the following as they apply to Partial or Total Loss of Inst. Air: Instrument Air Compressor Power supplies:(CFR:

41.7145.5I45.6)

AA2.05 - Ability to determine and interpret the following as they apply to Loss of AK2.03 - Knowledge of the interrelations between Refueling Accidents and the Shutdown Cooling: Reactor Vessel Metal Temperature (CFR: 41.10 143.5I45.13) following: Radiation Monitoring equipment (CFR41.7 145.7I 45.8)

EA1.03 - Ability to operate and/ or monitor the following as they apply to High Drywell Pressure: LPCS - Plant specific (CFR41.7/ 45.V 45.6)

EAl.02 - Ability to operate and / or monitor the following as they apply to High Reactor Pressure : Reactor/Turbine pressure regulating system :(CFR: 41.7145.5I I O1 O1 295023 Refueling Acc 18 2.9 3.0 3.4 3.4 4.0 3.8 295024 High Drywell Pressure 15 295025 High Reactor Pressure 13 AG2.1.30 - Ability to locate and operate components, including local controls. (CFR:

1 3.9 I 45.6) 1

-401 -1

=

295026 Suppression Pool High Water 0

0 0

0 Temp. 15 295027 High Containment Temperature 15 0 0 0 0 0

0 0

0 295028 High Drywell Temperature 15 295030 Low Suppression Pool Wtr LvlI5 0 0 1 0 0 0 EK3.02 - Knowledge of the reasons for the following responses as they apply to High 3.9 1

0 0 AK1.01 - Knowledge of the operational implications of the following concepts as they 2.5 1

Off-site Release Rate: System Isolations :(CFR: 41.5/41.10/45.6/ 45.13) apply to the Plant Fire On Site: Fire Classifications by type (CFR: 41.8 to 41.10 /45.3)

I

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0 0 AK2.04 Knowledge of the interrelations between MAIN TURBINE 3.3 1

GENERATOR TRIP and the following: Main generator protection (CFR: 41.7 145.8)

I I

t I

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3 4 Group Point Total:

20

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ES-401 4

Form ES-401-1 Fori ES-401-1 ES-401 BWR Examination Outline WAPE # / Name / Safety Function

295034 Secondary Containment 0 0 0 1 0 0 EA1.01 - Ability to operate andl or monitor the following as they apply to Secondary Ventilation High Radiation / 9 Containment Ventilation High Radiation: Area radiation monitoring system:(CFR41.7145.5145.6) 295035 Secondary Containment High 0

0 0

0 Differential Pressure 15 295036 Secondary Containment High 1 0 0 0 0 0 EK1.01 - Knowledge of the operational implications of the following concepts as they SumpIArea Water Level 15 apply to the Secondary Containment High Sump1 Area Water Level: Radiation releases (CFFt41.8 to 41.10145.3) 500000 High CTMT Hydrogen Conc. 15 0 1 0 0 0 0 EK2.02 - Knowledge of the interrelations between High CTMT Hydrogen Conc. And the following: Containment oxygen monitoring systems (CFR: 41.7 145.7 145.8) 3.8 2.9 3.1 KIA Category Point Totals:

2 2 1 1 1 0 Group Point Total:

ES-401 6

Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO I SRO)

System # / Name K

K K

K A

A A

A G

WA Topic@)

1 2

3 4

5 6

1 2

3 4

203000 RHWLPCI: Injection 0 0 0 0 0 0 1 0 0 0 0 Al.04 Ability to predict and/or monitor changes in parameters associated with Mode operating the RHWLPCI: INJECTION MODE (PLANT SPECIFIC) controls including: System Pressure (CFR: 41.5 / 45.5) 205000 Shutdown Cooling 0 0 0 0 0 0 0 0 1 0 0 A3.03 - Ability to monitor automatic operations of the Shutdown Cooling System(RHR Shutdown Cooling Mode) including: lights and alarms (CFR:41.7/45.5) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: Low suppression pool level: BWR-2, 3, 4 (CFR:41.5/43.5/45.3/45.13) they apply to the HPCI: Turbine speed control: BWR-2,3,4 (CFR41.5145.7) 206000 HPCl 0 0 0 0 0 0 0 1 0 0 0 A2.07 - Ability to (a) predict the impacts of the following on the HPCl and (b) 206000 HPCl 0 0 0 0 1 0 0 0 0 0 0 K5.05 - Knowledge of the operational implications of the following concepts as 207000 Isolation (Emergency) 0 0 0 0 0 0 0 0 0 0 0 Condenser 209001 LPCS 0

1 0 0 0 0 0 0 0 0 0 K2.01 - Knowledge of electrical power supplies to the following: Pump power (CFR41.7) 209002 HPCS 0

0 0

21 1000 SLC 0 0 0 1

0 0 0 0 0 0 0 K4.04 - Knowledge of SLC design feature@) and or interlock(s) which provide for the following: Indication of fault in explosive valve firing circuits (CFR41.7) 212000 RPS 0 0 1 0 0 0 0 0 0 0 0 K3.11 - Knowledgeof theeffect thata lossor malfunctionof the RPS will have on the following: Recirculation system (CFR41.7/45.6) 215003 IRM 0 0 0 1 0 0 0 0 0 0 0 K4.04 - Knowledge of the IRM design feature(s) and or interlock(s) which provide for the following: Varying system sensitivity levels using range switches (CFR41.7) 215003 IRM 0

1 0 0 0 0 0 0 0 0 0 K2.01 - Knowledge of electrical power supplies to the following: IRM Channels/

I detectors (CFR41.7) lES401 IR 2.5 3.5 3.4 3.3 3.0 3.8 3.0 2.9 2.5 Form ES-401-1 BWR Examination Outline 1

1 -

1 1

1 -

K1.02-Knowledge of the physical connections andlor cause-effect relationships between Source Range Monitor and the following: Reactor Manual Control (CFR:41.2 to 41.9/45.7 to 45.8) components and controls (CFR: 41.7) relationships between RClC and the following: Condensate storage and transfer system (CFFt41.2 to 41.9/ 45.7 to 45.8)

G2.1.28 - Knowledge of the purposes and function of major system K1.01 - Knowledge of the physical connections and/or cause-effect G2.1.28 - Knowledge of the purpose and function of major system components A4.02 - Ability to manually operate and/or monitor in the control room: Manually A4.06-Ability to manually operate and/or monitor in the control room: Reactor K3.06 - Knowledge of the effect that a loss or malfunction of the Reactor Water K3.02 - Knowledge of the effect that a loss or malfunction of the SGTS will K4.03 - Knowledge of AC Electrical distribution design feature@) and or and controls.

initiate the system (CFR:41.7/45.5 to 45.8) water level (CFR: 41.7/45.5 to 45.8)

Level Control will have on the following: Main Turbine (CFR:41.7/45.6) have on the following: Off-site release rate (CFR:41.7/45.6) interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers (CFFk41.7) have on the UPS (AC/DC): DC electrical power (CFR:41.7/45.7) operating the DC Electrical distribution controls including: Battery charginddischarging rate (CFR:41.5/45.5) 1 GENERATORS (DIESEUJET) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Consequences of operating under/over excited (CFR:

41.5 I45.6)

K6.02 - Knowledge of the effect that a loss or malfunction of the following will Al.01 - Ability to predict and/or monitor changes in parameters associated with A2.04 - Ability to (a) predict the impacts of the following on the EMERGENCY A3.02 - Ability to monitor automatic operations of the Instrument Air including:

K5.02 - Knowledge of the operational implications of the following concepts as Air temperature (CFR 41.7/45.5) they apply to A.C. ELECTRICAL DISTRIBUTION: Breaker Control (CFR: 41.5

~

3.4 1

3.2 1

3.5 1

3.2 1

3.9 1

2.8 1

3.6 1

3.1 1

2.8 1

2.5 1

2.9 1

2.6 1

K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the Component Cooling Water: Valves (CFR:41.5/45.5)

Al.03 - Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including: RPS status 1CFR. A I 5 145.51 K6.04 Knowledge of the effect that a loss or malfunction of ISOLATION SYSTEWNUCLEAR STEAM SUPPLY SHUT-OFF :Nuclear boiler instrumentation (CFR: 41.7 / 45.7)

' the following will have on the PRIMARY CONTAINMENT I

2.7 3.4 3.3 1

1 1

Group Point Total:

26

ES-401 9

Form ES-401-1 System # I Name 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201 004 RSCS 201005 RClS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control K

1 0

0 0

0 K

K 3

4 I

0 0

0 1

0 0

204000 RWCU 214000 RPlS 21 5001 Traversing In-core Probe 21 5002 RBM 216000 Nuclear Boiler Inst.

219000 RHWLPCI: TorudPool Cooling Mode 0

0 0

0 ES-401 BWR Examination Outline Form ES-401-1 223001 Primary CTMT and Aux.

226001 RHWLPCI. CTMT Spray Mode 230000 RHWLPCI. TorudPool Spray Mode 233000 Fuel Pool CoolinqlCleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam Plant Systems - Tier ZGroup 2 (RO I SRO)

A I A l A l G l WA Topic($

0 1

0 0

O I O 1 O 1 O I 1

1 1

1 0

0 0

0 K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the RWM: Rod Position indication - Plant Specific 0

0 0

0 1

0 0

0 A2.07 -Ability to (a) predict the impacts of the following on the Recirculation flow control and (b) based on those predications, use procedures to correct, control, or mitigate the conseqences of those abnormal operation: Loss of feedwater singal inputs: Plant specific (CFR:41.5/43.5/45.3/45.13) interlocks which provide for the following: Unintentional reduction in vessel injection O I O 1 O 1 O 1

~

0 0

0 1

0 0

A3 01 - Ability to monitor automatic operations of the Main and Reheat system includina Isolation of main steam svstem (CFR 41.7/45 5) z

'+

+ I

~ 4.2 I 1

r ol N

x 0

0 0

~

r 0

0 -

0 0

0 -

r 0

0 0

0 0 -

0 0

0

~

0 0

W

.- b -

C r"

C 0

m 73 m

a k w C

L

% e a

0 8

3 N

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

Form ES-401-3 Topic I Facility: Hope Category

1.

Conduct of Operations

2.

Equipment Control

3.

Radiation Control RO SRO-Onlv I

reek - RO Exam Date of Exam:

1 1 /28/05 2.1.21 WA #

IR IR Ability to obtain and verify controlled procedure copy (CFR: 45.10 /

3.1 1

Knowledge of system status criteria which require the notification of 2.5 1

I 45.13)

I I

I Knowledge of 10 CFR 20 and related facility radiation control requirements (CFR: 41.1 2 / 43.4. 45.9 / 45.1 0).

2.1.14 2.6 1

2.1.33 I 3.4 I

I Ability to recognize indications for system operating parameters which are entry-level condition for Technical Specifications (CFR: 43.2 / 43.3 Subtotal 2.2.1 2.2.34 2.2.

Subtotal Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

(CFR: 45.1)

Knowledge of the process for determining the internal and external effects on core reactivity (CFR: 43.6) 2.3.1 2.3.10 2.3.

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure (CFR: 43.4 / 45.1 0) I

4.

Emergency Procedures/

Plan 2.4.27 2.4.39 2.4.31 2.4.

Knowledge of fire in the plant procedure (CFR: 41.10 / 43.5 / 45.13)

Knowledge of the RO's responsibilities in emergency plan 3.3 1

Knowledge of annunciators alarms and indications / and use of the 3.0 1

implementation (CFR: 45.1 1) response instructions. (CFR: 41.1 0 / 45.3) 3.3 1

~

ES-401 BWR Examination Outline Form ES-401-1

2.

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 *1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

SystemsJevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

e.

Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

3.

I

9. For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

Form ES-401-1 ES-401 2

ES-401 BWR Examination Outline Form ES-401-1 WAPE # / Name / Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 0

0 295004 Partial or Total Loss of DC Pwr / 6 0

K K

K A

1 2

3 1

0 0 0 0 0 0 0 0 0 295006 SCRAM / 1 0

1 0

0 nergency and Abnormal Plant Evolutions - Tier l/Group 1 (RO / SRO)

A G

WA Topic(s)

IR i

~~

1 AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41.lo/ 43.a 45.12) 3.8 0 AA2 04 - Ability to determine and interpret the following as they apply to Partial or Total loss of 3.3 DC power:(CFR: 4 1. IO J 43.5 I 45.13) - System Lineups 0 WA Randomly Rejected 1 AG2.1.32 - Ability to explain and apply system limits and precautions (CFR 41.lo/ 43.a 45.12) 3.8 01 01 295016 Control Room Abandonment I7 29501 8 Partial or Total Loss of CCW l 8 29501 9 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling / 4 0

0 0

0 0 0 0 0 0 1 G2.4.30 - Knowledge of which events related to system operationslstatus should be 3.6 reported to outside agencies loss of Instrument Air:(CFR: 41.10/43,5/ 45.13) - Status of safety-related instrument air system loads (see AK2.1 - AK2.19) 0 0 0 0 1 0 AA2.02 - Ability to determine and interpret the following as they apply to Partial or Total 3.7 0

0 0

0 295023 Refueling Acc / 8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 295027 High Containment Temperature I 5 295028 High Drywell Temperature / 5 0

0 0

0 0 0 0 0 0

0 0

0 0 0 0 0 0 0 0

1 1 EG2.4.50 - Ability to verify system alarm setpoints and operate controls identified in the 3.3 alarm response manual. (CFR 45.3)

EA2.01 - Ability to determine and interpret the following as they apply to Low SuDDression Pool Water level (CFR:41.lo/ 43.5/ 45.13) - Sumression Pool level 0

4.2 295030 Low Suppression Pool Wtr Lvl / 5 0

0 0 0 1

1 1

' 4 Group Point Total:

I 7

ES-401 4

Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1IGroup 2 (RO I SRO)

E/APE # I Name I Safety Function K

K K

A A

G WA Topic@)

IR 1

2 3

1 2

295002 Lossof MainCondenserVacI3 0 0 0 0 0 0 0

0 0

0 295007 High Reactor Pressure I 3 295008 High Reactor Water Level I 2 295009 Low Reactor Water Level I 2 0 0 0 0 I 0 AA2.02 - Ability to determine and interpret the following as they apply to Low Reactor Water 3.7 295010 High Drywell Pressure I 5 0 0 0 0 0 1 AG2.4.6 - Knowledge of symptom based EOP mitigation strategies (CFR: 41.10 143.5 I 4.0 29501 1 High Containment Temp I 5 295012 High Drywell Temperature / 5 0 0 0 0 1 0 AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH DRYWELL 3.8 Level (CFR: 41.lo/ 43.5 145.13) - Steam flow/ feed flow mismatch 45.13) 0 0

0 0

TEMPERATURE : Drywell temperature (CFR: 41.10 143.5 145.13) 295013High Suppression PoolTemp./5 0 0 0 0 0 0 295014 Inadvertent ReactivityAdditionI 1 0 0 0 0 0 0 0

0 0

0 295020 Inadvertent Cont. Isolation15 & 7 0 0 0 0 0 0 0

0 0

0 295029HighSuppression Pool WtrLvll5 0 0 0 0 0 0 Area Temperature I 5 29501 5 Incomplete SCRAM I 1 295017 High Off-site Release Rate I 9 295022 Loss of CRD Pumps I 1 295032 High Secondary Containment 0

0 0

0 295033 High Secondary Containment 0

0 0

0 295034 Secondary Containment 0

0 0

0 Area Radiation Levels I 9 Ventilation Hiqh Radiation I 9 1

1 1

295035 Secondary Containment High 295036 Secondary Containment High Differential Pressure / 5 Sump/Area Water Level / 5 500000HighCTMTHydrogenConc.15 0 0 0

' WA Category Point Totals:

0 0

0 0 0 2

1 Group Point Total:

3

ES-401 6

Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K

K K

K A

A A

A G

K/A Topic(s)

IR 1

2 3

4 5

6 1

2 3

4 203000 RHR/LPCI: Injection 0 0 0 0 0 0 0 0 0 0 0 Mod8 205000 Shutdown Cooling 0

0 0

206000 HPCl 0 0 0 0 0 0 0 0 0 0 1 (32.1.14 - Knowledge of system status criteria which require the notification of plant 3.3 personnel. (CFR: 43.5 / 45.12) 207000 Isolation (Emergency) 0 0 0 0 0 0 0 0 0 0 0 Condenser 209001 LPCS 0 0 0 0 0 0 0 I 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the LPCS and (b) based 3.2 on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.5/ 43.5/ 45.3 45.13) - Valve closures 209002 HPCS 0

0 0

21 1000 SLC 0

0 0

21 2000 RPS 0

0 0

215003 IRM 0

0 0

215004SourceRangeMonitor 0 0 0 0 0 0 0 0 0 0 0 21 5005 APRM / LPRM 0 0 0 0 0 0 0

1 0 0 0 A2.02 - Ability to (a) predict the impacts of the following on the APRW LPRM and 3.7 (b) based on those predictions, use procedures to correct control or mitigate the consequences of those abnormal operation (CFR: 41.V 43.5/ 45.3/ 45.13) - Upscale or downscale trips.

217000 RClC 0

0 0

21 8000 ADS 0

0 0

223002 PCIS/NuclearSteam 0 0 0 0 0 0 0 0 0 0 0 Supply Shutoff 239002 SRVs 0

0 0

259002 Reactor Water Level 0 0 0 0 0 0 0 0 0 0 1 G2.1.33 - Ability to recognize indications for system operating parameters which are 4.0 Control entrv-level conditions for technical wecifications. fCFR: 43.2 / 43.3 / 45.3) 1 1

1 1

0 0

0 W

0 0

0 d

(0 N

0 0

0 -

0 0 -

0

0 0

-xnv ' u a ~

auwni u!ew OOOSPZ

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

Form ES-401-3 1 1/28/05 Facility: Hope Creek - SRO Only Exam Date of Exam:

I I

SUMMARY

New Scenario.

PREPARED BY:

M. L. Brown NRC Operations Examiner 9/29/05 DATE FACILITY REVIEWER:

Nuclear Operations Training Supervisor -

Hope Creek DATE APPROVED BY:

NRC Chief Examiner DATE

NRC-001 REV-060 Enabling Objectives A.

The crew must demonstrate the ability to operate effectively as a team while completing a series of CREW CRITICAL TASKS, which measure the crews ability to safely operate the plant during normal, abnormal, and emergency plant conditions.

(Crew critical tasks within this examination scenario guide are identified with an *.)

A.

B.

C.

D.

E.

F.

G.

E SRV fails Open Withdraw Control Rods until 1 bypass valve is open CRD Flow Controller fails downscale in AUTO Loss of B MG Set Control Rod 22-35 inadvertently scrams (TS)

Steam Leak from RCIC piping RCIC isolation valves fail to close The scenario begins with a Reactor Startup in service with IOP-3 completed up to step 5.3.28.

Reactor power is approximately 4%. RClC is being operated for HC.OP-ST.BD-0002, RClC Pump Valve and Flow Test and should be completed within the next hour. 1 BP116 EHC pump is tagged out for maintenance and will be out of service until a new pressure compensator arrives tomorrow. After the operators have raised power using the control rods, the CRD flow controller fails downscale in AUTO and the operators are forced to take MANUAL control of the flow controller. Once the operators have stabilized the plant, the B MG Set trips causing a ?h scram and RWCU isolation. The operators will have to restore power to the B RPS from the alternate source. Once power has been restored to the B RPS bus, Control Rod 22-35 inadvertently partially scrams (stops at position 12) and its Accumulator is INOP. With Reactor Pressure < 900 psig and Charging Header pressure < 940 psig, this will require a Scram. After the crew has stabilized from the scram, a steam leak develops on RCIC, RClC isolation valves fail to close causing RClC room temperature to increase. Crew should enter EOP-103 based on high room temperature and place FRVS in service and attempt to shutdown RCIC. Crew should discover that the RClC isolation valves cant be closed. Crew should scram the reactor based upon RClC room temperature approaching safe operating limit and enter EOP-101. When temperature exceeds Max Safe Operation limit in 2 areas, crew should enter EOP-202, Emergency Depressurization. When the crew goes to Emergency Depressurize, E ADS valve will not open, requiring the BOP to open another SRV. Scenario will end after 5 SRVs have been opened.

NRC-00 1 Page 2 of 17 Rev.: 060

NRC-001 REV460 1

2 Initial INITIALIZE the simulator to 4% power, MOL EVENT ACTION:

COMMAND:

PURPOSE:

EVENT ACTION:

COMMAND:

PURPOSE:

Initial I Description I

COMPLETE Attachment 2 Simulator Ready-for-TrainingExamination Checklist of NC.TQ-DG.ZZ-0002(Z).

3 Initial E

EVENT ACTION:

COMMAND:

PURPOSE:

ET # 1 Description I

NRC-00 1 Page 3 of 17 Rev.: 060

NRC-OO 1 REV-060 Initial Description Delay Ramp Trigger Init Val NONE I

I I

I Final Val

=

Initial Description Delay Ramp Trigger NONE NONE NONE Init Val I

I 1 NONE A

I I

Init Val Final Val INSTALL INSTALL INSTALL INSTALL TAGGED OPEN 100%

ON I

I I

Final Val

=-i NRC-001 Page 4 of 17 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlantIStudent Response Comments Crew assumes the watch at 0

RO/PO Ensure HPCl High Level Rx Pressure should be step 5.3.28 of 1O.ZZ-0003, and Trip Reset allowed to continue continues plant startup per 0

RO monitors RPV parameters to increasing to 500 PSig procedure ensure proper RX startup 0

CRS directs PO to begin pre-warming of SJAE IAW HC.OP-SO.CG-0001, Condenser Air Removal Operation o PO stops raising Throttle pressure and allows at least 1 Turbine Bypass Valve to OPEN o CRS starting of the 2d Secondary Condensate Pump IAW procedure 0

At - 500psig CRS Flow controller fails 0

RO recognizes:

downscale in AUTO:

After the PO stops raising Throttle pressure and allows at least 1 Turbine Bypass valve to

OPEN, OR at the discretion of the Lead

Examiner, Failure of CRD flow controller and reports failure to CRS TRIGGER RT-1.

0 CRS directs RO to place CRD flow controller in MANUAL and attempt to open the flow control valve 0

RO places CRD flow controller in MANUAL RO Depresses OPEN on CRD Flow Controller to establish normal CRD parameters NRC-00 1 Page 5 of 17 Rev.: 060

NRC-OO 1 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Expected PlanUStudent Response Comments Event / Instructor Activity 0

RO/PO contact personnel outside of Control Room to investigate failure of the CRD Flow Control Valve in AUTO.

Loss of B MG Set Once the crew has returned CRD parameters to normal OR At the discretion of the Lead Examiner Crew recognizes trip of B MG Set TRIGGER - RT2 RO responds to Annunciators RO recognizes that NO actual scram condition exists and DOES NOT scram the reactor CRS enters AB-IC-0003 Determines Normal CANNOT be restored Directs RO to Transfer power to Alternate power supply RO verifies Alternate Power is avai la ble RO Transfers Power to Alternate Power Supply by Positioning the RPS MG SET TRANSFER SWITCH to the Alternate Position.

RO Resets the Y2 Scram by:

Improper operation of this will cause a Scram o Turning the key for the Affected RPS channel to the RESET position o Turning the key back to the NORMAL position o Verify the scram is reset NRC-00 1 Page 6 of 17 Rev.: 060

NRC-00 1 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlantIStudent Response Comments BOP verifies MSIVs are still OPEN CRS directs that the tripped NSSS logic be reset o RO presses the NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM TRIP LOGIC B RESET Pb.

o RO Verifies MSlV TRIP LOGIC TRIPPED light goes off CRS Directs restoration of RWCU system 0

RO restores RWCU system IAW applicable SOP.

Control Rod 22-35 partially 0

RO references ARP for C6-E3, scrams Rod Drift Once RWCU system has been RO recognizes that Rod 22-35 returned to service has inserted to position 12 OR At the discretion of the Lead Examiner TRIGGER - RT3 0

CRS enters AB.IC-0001, Control Rod CT-1 0

CRS determines that Reactor

- Have CRS address Pressure < 900 psig AND tech specs required for Charging Water Header pressure this malfunction after e 940 psig AND Control Rod the scenario is over.

Scram Accumulator 22-35 is INOPERABLE and that a Manual Scram is required CRS directs the RO to LOCK Mode Switch in SHUTDOWN 0

RO LOCKS the Mode Switch in the SHUTDOWN Position 0

CRS enters AB-0000 0

RO verifies all Control Rods fully inserted NRC-001 Page 7 of 17 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlanUStudent Response Comments CRS Enter EO-101 if RPV level drops below +12.5 0

RO Inserts SRMs and IRMs

=, BOP verifies H2 Injection system tripped 0

CRS directs tripping the Main turbine AND verifying Generator Lockout (MA) at 0 Mwe 0

BOP trips the Main turbine 0

BOP Ensures Generator Lockout 0

CRS directs RO to maintain RPV level between +12.5 AND +54 using feedwater 0

RO Verifies feedwater aligns to startup level control 0

CRS directs BOP to maintain pressure within a band

=, BOP verifies EHC is controlling RPV pressure below 1037 psig 0

BOP maintains Condenser vacuum using SJAEs 0

BOP controls plant cooldown/

depressurization using Main Turbine Bypass valves RClC Steam Leak 0

BOP responds to annunciator D3-A2, RCIC/RHR B Area Leak Temp Hi Once the crew has stabilized the plant after the Scram OR At the discretion of the Lead Examiner TRIGGER - RT3 0

BOP dispatches an NE0 to investigate High temperature NRC-001 Page 8 of 17 Rev.: 060

NRC-OOI REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Planus tuden t Response Comments BOP determines RCIC has NOT 0

PERFORMS the following in Tripped and quick succession:

o DETERMINES the Channel that initiated this alarm from the Digital Points OR NUMAC Monitor 1 OC621-Z5 (1 SKXR-11502)

OR 1 OC64O-Z7 (1 SKXR-1 1503).

a 0

BOP CHECKS Page 13 and Page 16 of This attachment for Channel(s) in respective monitor which initiated alarm.

0 CHECK area cooling 0

MAXIMIZE area cooling

IF continued operation is not required, SHUT DOWN RCIC.

IF the turbine is injecting water into Reactor Vessel AND IF it is desirable to continue the operation, PLACE CHANNEL B(D)

ISOLATION BYPASS SWITCH in BYPASS (Local Panel P621 (P640)).

0 CRS refers to HC.OP-EO.ZZ-01 03(Q) Reactor Building Control CRS directs RO/BOP to monitor and Control Reactor Building Temps NRC-001 Page 9 of 17 Rev.: 060

~

V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity 5

5 0

3 5 minutes after dispatched, 0

NE0 reports steam leak inside RClC room 0

Expected PlanUStudent Response BOP reports RClC pump room temperature > Column 1 - Max Normal Op Temp CRS determines FRVS is NOT in service CRS directs BOP to verify proper operation of RBVS and Emergency Area Cooling System BOP determines RBVS and Emergency Area Cooling systems are operating properly CRS directs BOP to start all available RBVS fans BOP starts all available RBVS fans CRS directs isolation of RClC RO trips RClC and attempts to shut RClC isolation valves RO determines that RClC tripped, however, RClC isolation valves failed to close RO requests assistance from NEO/ WCC to close RClC isolation valve BOP reports RClC pump room temperature > Max Safe Op Temp CRS determines RCS is discharging into the Reactor building CRS directs a Recirc Runback and a Reactor Scram CRS enter EO-1 01 concurrently with this procedure NRC-001 REV-060 Comments If asked NE0 reports steam is still leaking in RClC room.

Note - Reactor is already scrammed, so CRS may not direct this action NRC-001 Page 10 of 17 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments e

e e

CT-2 3

Note - Failure of E ADS to open was input as an initial condition e

CT-3 e

Termination Requirement:

When 5 SRV are OPEN OR At Lead Examiner Discretion e

e BOP reports 2d area has exceeded its Max Safe Op Limit CRS determines Emergency Depressurization is required and enters EO-0202 CRS determines the following:

o Reactor is shutdown from all conditions without boron o DW pressure is e 1.68 psig o Supp Pool level > 0 CRS orders 5 ADS valves to be Opened BOP Places all 5 ADS valve hand switches to OPEN BOP recognizes PSV-F013E failed to OPEN CRS directs BOP to open non-ADS SRVs until a total of 5 SRVs are open BOP opens an additional SRV until a total of 5 SRVs are open CRS determines an Alert Classification is required IAW ECG Section 3.2.2.b (Valid High Drywell Pressure).

Also have the CRS address the Tech Specs for the failed rod.

- Need to check on ECG classification **

NRC-001 Page 1 1 of 17 Rev.: 060

NRC-001 REV-060 A.

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NC.TQ-DG.ZZ-0002 Conduct of Simulator Training.

NUREG 1021 Examiner Standards JTA Listing Probabilistic Risk Assessment Technical Specifications Emergency Plan (ECG)

Alarm Response Procedures (Various)

SH.OP-AS.ZZ-000 1 Operations Standards SH.OP-AP.ZZ-0101 Post Transient Response Requirements SH.OP-AP.ZZ-0108 Operability Assessment and Equipment Control Program HC.OP-IO.ZZ-0003 Startup from Cold Shutdown to Rated Power HC.OP--.IC-0003 REACTOR PROTECTION SYSTEM HC.OP--.IC-0001 Control Rod HC.OP-AJ3.ZZ-000 Reactor Scram HC.OP-AE3.ZZ-0001 Transient Plant Conditions HC.OP-EO.ZZ-0101 RPV Control HC.OP-EO.ZZ-0 10 1 A ATWS-RPV Control HC.OP-EO.ZZ-0102 Primary Containment Control HC.OP-EO.ZZ-0202 Emergency RPV Depressurization HC.RE-IO.ZZ-0001 Core Operations Guidelines NRC-OO 1 Page 12 of 17 Rev.: 060

NRC-001 REV-060 NRC-001/ 00

1.

Recognize that Reactor Pressure < 900 psig AND Charging Water Header pressure < 940 psig AND Control Rod Scram Accumulator 22-35 is INOPERABLE and Manually Scram within two minutes WA 201001 Control Rod Drive Hydraulic System A2.04 Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE HYDRAULIC SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ?Scram conditions(CFR: 41.5 / 45.6) RO 3.81 SRO 3.9 Control Rod Drive system is malfunctioning at a low reactor pressure. The reactor must be scrammed immediately to insure that all control rods are successfully inserted prior to pressure dropping below the point where the rods would insert. Two minutes is deemed adequate time to recognize the condition and implement the Immediate Operator Actions of AB.IC.OOO1.

2.

WA 295032 High Secondary Containment Area Temperature EK3 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE EK3.01 Emergencyhormal depressurization RO 3.5 SRO 3.8 EA2 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE EA2.01 Area temperature RO 3.8 SRO 3.8 Crew actuatesfive SRVS within two minutes of RCIC room temperature exceeding 250 degrees by Control Room indication (SPDSKRIDS).

The steam leak in the HPCI room is now affecting a second area. The reactor must be depressurized to place it in its lowest energy state due to the potential for multiple inoperable safety systems, to reduce the driving head for the leak, and to reject decay heat to the suppression pool rather than the Reactor Building.

The term Crew actuates five SRVs takes into account the F013D failure, which is already inserted. Two minutes is deemed adequate time to recognize the condition and implement EOP-202 and AB.ZZ-0001 Att.

13.

3.

WHEN the PSV-FO13D SRV fails to open, THEN before RPVpressure drops below 50 psig, the Crew ensures a fiph SRV is opened to achieve five open SRVs.

WA 239002 RelieYSafety Valves A4 Ability to manually operate and/or monitor in the control room:

A4.01 SRVs RO 4.4 SRO 4.4 The Minimum Number of SRVs required for Emergency Depressurization (MNSRED) is five. The MNSRED is utilized to assure the RPV will depressurize and remain depressurized when Emergency Depressurization is required. When the PSV-F013D fails to open, the Crew needs to open an additional SRV to achieve MNSRED. This is directed by both EOP-202 and AB.ZZ-0001. SRVs are designed to open with a minimum differential pressure of 50 psid between the reactor vessel and the suppression chamber. Below this d/p, they may not open. If the Crew does not attemp; to open the fifth SRV before this minimum d/p is lost, they cannot validate its operation. This would prevent them from detecting the failure and pursuing the use of the Alternate Depressurization Systems in EOP-202.

NRC-00 1 Page 13 of 17 Rev.: 060

NRC-001/ 00 NRC-001 REV-060 HOPE CREEK NRC - PRA RELATIONSHIPS EVALUATION FORM EVENTS LEADING TO CORE DAMAGE y/N EVENT TRANSIENTS:

Turbine Trip Y/N EVENT SPECIAL INITIATORS:

Loss of ssw Y

Loss of Feedwater MSIV Closure Loss of Condenser Vacuum Inadvertent Open SRV Loss Of Offsite Power Station Black Out Loss of SACS Loss of RACS Loss of Instrument Air Y

ATWS Y

LOCA COMPONENT/TRAIN/SY STEM UNAVAILABILITY THAT INCREASES CORE DAMAGE FREQUENCY Y/N COMPONENT, SYSTEM, OR TRAIN Y/N COMPONENT, SYSTEM, OR TRAIN HPCI RCIC One SRV EDG A One SSW Pump / Loop EDG B Circulating Water System - 4 pumps TACS Class 1E 120VAC Bus - A Train Class 1E 120VAC Bus - D Train OPERATOR ACTIONS IMPORTANT IN PREVENTING CORE DAMAGE Y/N OPERATOR ACTION Y

Manual RPV Emergency Depressurization when required Manual RPV Depressurization during ATWS Initiation of RHR for Decay Heat Removal Y

Initiation of Containment Venting Restore Offsite power within 45 minutes SACS / SSW restoration after total loss of both systems Avoiding Loss of Feedwater during transient Recovery of the Main Condenser Complete this evaluation form for each Exam.

NRC-00 1 Page 14 of 17 Rev.: 060

NRC-001 REV-060 RxPower: 100%

MWe: I136 (May vary slightly):

Work Week: Any Risk Color: Green SMD: None River Temp: 65 Activities Completed Last Shift:

None Major Activities Next 12 Hours:

None Protected Equipment:

None Tagged Equipment:

None NRC-00 1 Page 15 of 17 Rev.: 060

NRC EXAMINATION SCENARIO GUIDE REVIEWNALIDATION Note: This form is used as guidance for an examination team to conduct a review for the proposed exam scenario(s). Attach a separate copy of this form to each scenario reviewed.

SELF-CHECK

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NRC-001 REVIEWER:

The scenario has clearly stated objectives in the scenario.

The initial conditions are realistic, equipment and/or Instrumentation may be out of service, but it does not cue crew into expected events.

Each event description consists of 0

0 0

0 The event termination point The point in the scenario when it is to be initiated The malfunction(s) that are entered to initiate the event The symptomskues that will be visible to the crew The expected operator actions (by shift position)

The use of non-mechanistic failures (e.g. pipe break) should be limited to one or a credible preceding event has occurred.

The events are valid with regard to physics and thermodynamics.

Sequencinghiming of events is reasonable (e.g. the crew has time to respond to the malfunctions in an appropriate time frame and implements procedures and/or corrective actions).

Sequencindtiming of events is reasonable, and allows for the examination team to obtain complete evaluation results commensurate with the scenario objectives.

If time compression techniques are used, scenario summary clearly so indicates.

The simulator modeling is not altered.

All crew competencies can be evaluated.

Appropriate reference materials are available (SOERs, LERs, etc.)

If the sampling plan indicates that the scenario was used for training during the requalification cycle, evaluate the need to modify or replace the scenario.

Proper critical task methodology used IAW NRC procedures.

NRC-001 Page 16 of 17 Rev.: 06

NRC EXAMINATION SCENARIO GUIDE VALIDATION (cont)

Note: The following criteria list scenario traits that are numerical in nature. A second set of numbers indicates a range to be met for a set of two scenarios. Therefore, to complete this part of the review, the set of scenarios must be available. The section below should be completed once per scenario set.

NRC:

001 SELF-CHECK 1

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Total malfunctions inserted: 4-W10-14 Malfunctions that occur after EOP entry: 1-413-6 Abnormal Events: 1-2/2-3 Major Transients: 1-2/2-3 EOPs used beyond primary scram response EOP: 1-3/3-5 EOP Contingency Procedures used: 0-311-3 Approximate scenario run time: 45-60 minutes (one scenario may approach 90 minutes)

EOP run time: 40-70% of scenario run time Crew Critical Tasks: 2-5/54 Technical Specifications are exercised during the test Events used in the two scenarios are not repeated The scenario sets for the exam week do not contain duplicate scenarios NRC:

Comments:

NRC-00 1 Page 17 of 17 Rev.: 06

SCENARIO TITLE:

Electrical ATWS/ SRV fails/ Small Break LOCA SCENARIO NUMBER:

NRC-002 EFFECTIVE DATE:

EXPECTED DURATION:

1.O Hours REVISION NUMBER:

00 PROGRAM:

1-1 L.O. REQUAL (x(

INITIAL LICENSE

-1 OTHER REVISION

SUMMARY

New Scenario.

PREPARED BY:

M. L. Brown 9/29/05 NRC Operations Examiner DATE FACILITY REVIEWER:

Nuclear Operations Training Supervisor -

DATE Hope Creek APPROVED BY:

NRC Chief Examiner DATE

NRC-001 REV-060 Enabling Objectives A.

The crew must demonstrate the ability to operate effectively as a team while completing a series of CREW CRITICAL TASKS, which measure the crew's ability to safely operate the plant during normal, abnormal, and emergency plant conditions.

(Crew critical tasks within this examination scenario guide are identified with an " *.")

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Perform Core Spray Full Flow test Power increase using Recirc Flow Inadvertent HPCI initiation FRVS fails to start 10B130 trips EHC pump trips, Electrical ATWS SRV fails OPEN, Broken SRV tailpipe Small Break LOCA, PSP function lost RHR Spray Logic Failure The plant is operating at 80% power, Middle Of Cycle with SLC Pump AP-208 tagged out for a motor replacement and is expected back within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Core Spray Loop A operability PT will be performed. When the test return valve is opened, the loop flow instrumentation will fail to respond. Core Spray A should be declared Inoperable.

When power has been increased by -lo%, HPCl will inadvertently initiate. The crew will respond per AB.RPV-0001, Reactor Power, and terminate HPCl operation. A scram on high flux may occur if HPCl is not terminated. HPCl will be declared Inoperable and Tech Specs addressed.

- Talk to Steve ** Power to the Main Stack Rad Monitor will fail. The power loss results in Group 6 isolation, Secondary Containment Isolation but FRVS fails to auto start. The crew must manually start FRVS to maintain Secondary Containment integrity.

480VAC Unit Substation 1 OB1 30 will trip. This results in loss of power to the running Stator Cooling pump. The standby pump fails to auto start and must be manually started to prevent a turbine trip. The Unit Substation loss also results in a loss of Recirc Pump 28 due to oil pump B2 tripping and oil Pump B3 fails to start automatically or manually. Recirc MG Set 2B oil pressure drops below the trip setpoint but fails to trip. The MG set must be manually tripped. This places the plant in Region B - Immediate Exit region of the power to flow map. Recirculation flow must be increased or control rods must be inserted to exit Region B.

The EHC pressure regulator will fail resulting in opening of Turbine Control and Bypass valves to the Max Combined Flow limit (1 10%). Steam line pressure drops to the MSlV isolation setpoint and the MSlVs close. The reactor fails to auto scram and manual scram also will fail. RRCS will fail to auto initiate on high RPV pressure. The rods can be inserted by manually initiating RRCS.

NRC-001 Page 2 of 21 Rev.: 060

NRC-001 REV-060 RPV pressure will spike high due to the MSlV closure and RPS failure. SRVs will lift on high pressure. SRV F will fail to reclose when rods are inserted and RPV pressure lowers. The tail pipe on SRV F will rupture in the suppression chamber airspace shortly after the valve sticks open resulting in rapidly rising containment pressure and temperature.

Feed flow is lost to the RPV due to MSlV closure. HPCl may be manually started to restore RPV level but will fail shortly after being started.

If containment spray is attempted the B RHR Spray logic will fail and F016A will not OPEN.

Suppression chamber pressure will rise above the safe value for Pressure Suppression Pressure requiring emergency depressurization. Low pressure ECCS and Condensate must be operated during depressurization to prevent uncontrolled injection.

When the reactor has been depressurized, the containment spray will be repaired and can be placed in service then the scenario can be terminated requiring the BOP to open another SRV.

Scenario will end after 5 SRVs have been opened.

NRC-001 Page 3 of 21 Rev.: 060

NRC-001 REV-OM Initial INITIALIZE the simulator to 80% power, MOL Inifial Description 1

COMPLETE Attachment 2 Simulator Ready-for-TrainingExamination Checklist of NC.TQ-DG.ZZ-0002(2).

I initin1 I ET # I Description

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2 EVENT ACTION:

COMMAND:

PURPOSE:

EVENT ACTION:

COMMAND:

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EVENT ACTION:

COMMAND:

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NRC-001 Page 4 of 21 Rev.: 060

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SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Crew assumes the watch and starts performing HC.OP-ST.BE-0002 TRIGGER RT-1 Expected PlanUStudent Response 0

BOP observes proper Core Spray pump A suction pressure 0

BOP Ensures pump suction valve (HV-F001 A) is Open o BOP Sends an NE0 to pump to check pump out prior to start BOP Starts A Core Spray Pump while monitoring pump discharge pressure and confirms discharge pressure rises to > 300 psig in less than or equal to 5.0 seconds 0

BOP Records time Core Spray pump was started 0

BOP ensures the following:

0 Core Spray Division I Room Cooler fan has started

Service Water Outlet valve is Open (NE0 to report)

BOP - Throttles open Core Spray Full Flow Test Byp Valve, HV-FOl5A to obtain 2 4625 gpm flow.

0 BOP Determines that Core Flow indicator E21 -FI-RGOlA fails to indicate actual flow and reports instrument malfunction 0

CRS directs Core Spray A to be shutdown and returned to standby lineup CRS directs I&C to investigate CRS refers to Tech Spec 3.5.1 and determines Action A (7 day) applies NRC-001 REV-060 Comments

- Need to get a copy of this procedure from Archie **

NRC-00 1 Page 6 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlantIStudent Response 3

3 Raise Power usinn Recirc 3

Flow 7

Once Core Spray has been returned to a standby alignment and Tech Spec call has been made OR At the discretion of the Lead Examiner Have Load Dispatcher contact crew to raise power 0

3 BOP - shuts down Core Spray pump A and returns Core Spray to standby alignment Close E21 -FO15A Close E21 -F031 A CRS - directs RO/BOP to raise power to 100% using IOP-0006.

RO monitors plant for proper operation RO refers to HC.OP-SO.BB-0002 regarding MG set critical vibration and flow instability points Comments RO - raises reactor power by increasing Recirc Flow per lop-0006 at a rate not to exceed 1 Yo/minute RO slowly turns the Recirc pump Master Speed Control potentiometer in the clockwise direction.

- Ask Archie if Hope Creek typically operates with the Master Speed controller or Not **

RO monitors the following for proper operation Recirc speed increases Recirc loop flow increases Reactor power increases NRC-001 Page 7 of 21 Rev.: 060

NRC-OOI REV -060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments INADVERTENT HPCl INITIATION Once power has been raised OR At the discretion of the Lead Examiner by - 10%

TRIGGER RT-2 BOP verifies #4 STEAM LEAD DRAIN (HV-1018A) is CLOSED when the #4 CONTROL VALVE indicates off its open seat RO - diagnoses and reports inadvertent HPCl initiation

- Ask Archie if RO or CRS should direct entry into AB **

CRS directs entry into AB.RPV-0001 RO verifies Reactor level > -38 Drywell pressure < 1.68#

RO presses and holds the HPCl TURB TRIP PB RO observes the following close FD-W-4880 FD-FV-4879 RO adjusts FIC-R600 HPCl Flow controller to 0 gpm RO place FIC-R600 in MANUAL RO - PRESS FIC-R600 DECREASE Pb for approximately 7 seconds.

RO RELEASE the HPCI TURB TRIP PB.

RO VERIFY the FD-FV-4879 remains shut.

BOP reduces reactor power with Reactor recirculation flow as necessary to prevent a reactor scram CRS contacts I&C to investigate HPCl failure NRC-001 Page 8 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments CRS refers to Tech Spec 3.5.1.

Determines Action D applies (Verify RClC OPERABLE and restore HPCl to Operable within 14 days)

CRS refers to ECG and

- Check on determine reportability reportability requirements (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for IOSS of requirements **

single train)

BOP recognizes and reports loss

- Talk to Steve and of power to stack rad monitor see if he wants to do this malfunction Power Loss to Main Stack Rad Monitor When CRS has determined reportability requirements OR At Lead Examiner discretion TRIGGER RT-3 0

BOP refers to ARPs and verifies auto actions BOPKRS determines FRVS did not start as required 0

CRS directs starting of FRVS BOP starts FRVS Place FRVS A and B control switches to ON and verifies negative pressure restored by observing pressure indication and alarm clearing BOP determines Group 6 has isolated a BOP determines that Secondary Containment has isolated BOP dispatches N E 0 to investigate power loss NRC-OO 1 Page 9 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlanUStudent Response Comments 480V Unit Substation 1 OB1 30 trips Once Tech Specs and ODCM have been addressed OR At the discretion of the Lead Examiner TRIGGER RT-4 0

CRS directs I&C to investigate and determine if stack rad monitor can be transferred to alternate power supply 0

CRS refers to Tech Spec 3.3.6.1 (PCIS Instrumentation) determine Function 2c is inoperable, determine actions A

& 6 apply (single channel), then C and F after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 0

CRS - Refer to TRM 3.4 (post accident monitoring), determine Function 5 is inoperable and Condition A applies 0

CRS - Refer to ODCM 7.3.2 (gaseous effluent monitoring),

determine function 1 is inoperable, Conditions A & B apply (grab samples) and notify Chemistry

- Check with Archie to ensure pump should be tripped

- Check with Steve and either have 1 B2P120 either fail to start or have it tagged out for maintenance Crew responds to loss of 10B130 a RO diagnose failure of Recirc Pump B to trip and manually trip Place B Recirc pump MG-Set supply breaker to OFF 0

CRS direct entry to AB.RPV-0003 a RO - inserts rods to clear APRM Upscale Alarms NRC-00 1 Page 10 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Comments Event / Instructor Activity Expected Plant/Student Response RO - ENSURE that the Recirc MG Drive Motor Breaker has TRIPPED for the tripped Pump.

RO - CLOSE HV-F031A(B) for approximately 5 minutes, THEN RE-OPEN HV-F031A(B).

[C D-97681 RO - IMPLEMENT the following:

DL.ZZ-0026 Att. 3n (as required)

CRS - DIRECT the Reactor Engineer to develop a Rod Sequence to achieve an 80% Rod Line.

Requirements for Single Loop operations.

pow er/flo w map DL.ZZ-0026 Att. 3~

CRS - IMPLEMENT 10-6 CRS determines region of operation on CRS directs actions to exit Region B RO either Raises Recirc flow with Recirc pump A or inserts control rods to exit Region B CRS refers to Tech Spec 3.4.1 and COLR for SLO, determine APLHGR limit and APRM setpoints must be modified within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per IOP-6)

CRS contacts I&C to determine cause of failure and to adjust setpoints as required CRS refers to IOP-6 and determines all appropriate actions have been taken in accordance with Section 5.3

- See if anything else needs to be done for single loop operation NRC-001 Page 11 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlanVStudent Response Comments EHC Failure/ Electrical ATWW T SRV fails to close After CRS has notified I&C to adjust setpoints OR At the discretion of the Lead Examiner TRIGGER RT-5 CT-1

- See if any other actions need to be taken for loss of 10B130 and where guidance may be obtained crew diagnoses failure Of EHC Malfunctions to be pressure control inserted Pressure regulator fails high (or #4 control valve fails open, something to drag pressure down)

Auto Scram defeat Manual Scram defeat F SRV fails to close after opening 0

RO - recognizes MSlV closure and failure to auto scram CRS directs manual scram and entry into EO.ZZ-0101 RO - manually scrams the reactor by depressing manual scram pushbuttons 0

RO - recognizes the failure of the manual scram and places Mode Switch to SHUTDOWN CRS - directs RRCS to be initiated if not already completed by RO 0

CRS enters EO.ZZ-0101 A if rods are not yet inserted RO manually initiates RRCS 0

Place RRCS keylock in Trip Place RRCS CS in Trip RO reports when all rods are inserted NRC-001 Page 12 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlantlStudent Response Comments 0

BOP recognizes MSlV closure and ensures pressure is controlled by the SRVs as required.

0 CRS exits EO.ZZ-0101 A after control rods are inserted, returns RO/BOP diagnoses the failure of F SRV to close to EO.ZZ-0101 F SRV Fails to C l o d P C l

- Talk to Archie with f ai I ure plant tripped and 1 SRV open is HPCl needed or will plant depressurize Once Rods are inserted OR down to CS/RHR entry conditions At Lead Examiner Discretion TRIGGER RT-6 Note F SRV failure is inserted as an Initial condition 0

RO attempts to close F SRV 0

CRS directs actions of AB.RPV-0006 for SRV failure while continuing in EO.ZZ-0101 0

CRS performs actions for EO.ZZ-01 02 as appropriate 0

IAW AB>RPV-0006, RO reduces Recirc pump speed to minimum 0

RO cycles F SRV control switch several times to attempt to close the SRV 0

BOP ensures MSIVs and HV-F016 and HV-FO19 are closed to attempt to control cooldown 0

RO starts suppression pool cooling IAW AB-0001 0

BOP breaks condenser vacuum as follows 0

Verify Main Turbine < 1200 rpm

- See if RO has to do anything for this OPEN HV-1972 A/B/C NRC-00 1 Page 13 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlanVStudent Response Comments 0

RO attempts to start HPCl to control RPV level 0

RO - diagnoses and reports failure of HPCl to operate 0

RO/BOP - diagnose and report rapidly rising containment

- Talk to Archie/ Steve and see what is best method to fail HPCl T SRV Tailpipe Rupture, CNMT Sprav Failure/ Small Break LOCA pressures

- Talk to Archie - Goal it push operators to ED Once RO has initiated Suppression Pool Cooling OR At Lead Examiner Discretion TRIGGER RT-7

- Talk to Steve - not sure why we are putting LOCA in here. See what he says. How do you expect operators to react HV-F027A, B failed closed

Maybe override PBs dont work 0

RO/BOP determine leak is in suppression chamber based on higher suppression chamber pressure and/or vacuum breaker operation suppression chamber spray per 0

CRS directs initiation of drywell spray per EO.ZZ-0102 0

RO/BOP Attempt to initiate spray per SO.BC-0001 0

RO - Press BC-HV-F027B RHR LOOP B SUPP CHAMBER SPRAY HDR ISLN MOV AUTO CL OVRD PB 0

CRS directs initiation of EO.ZZ-0102 0

RO - Attempts to OPEN HV-F027B wont open F027B NRC-00 1 Page 14 of 21 Rev.: 060

NRC-001 REV-060 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments CT 0

RO - diagnoses Drywell and Suppression pool spray cant be initiated exceeded, determine Emergency Depressurization is required.

CRS - Directs entry to EOZZ-0202 when appropriate CRS/RO determines Reactor is shutdown under all conditions without boron 0

CRS/RO determines Drywell pressure > 1.68 psig 0

CRS directs RO to prevent injection from Core Spray and LPCl pumps not required to assure adequate core cooling 0

RO overrides Core Spray and RHR pumps not required for core cooling to Off CRS/BOP determines Suppression Pool level > 0 CRS directs RO/BOP to OPEN 5 ADS valve and Defeat PClG isolation interlocks if necessary 0

CRS -When PSP limits

- When would it be necessary to defeat PClG interlocks -What are we talking about here.

RO opens 5 ADS valves BOP operate Condensate system to prevent uncontrolled injection 0

- Talk to Steve about stopping point NRC-00 1 Page 15 of 21 Rev.: 060

NRC-OO 1 REV-060 A.

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NC.TQ-DG.ZZ-0002 Conduct of Simulator Training.

NUREG 1021 Examiner Standards JTA Listing Probabilistic Risk Assessment Technical Specifications Emergency Plan (ECG)

Alarm Response Procedures (Various)

SH.OP-AS.ZZ-000 1 Operations Standards SH.OP-AP.ZZ-0101 Post Transient Response Requirements SH.OP-AP.ZZ-0108 Operability Assessment and Equipment Control Program HC.OP-IO.ZZ-0003 Startup from Cold Shutdown to Rated Power HC.OP-AB.IC-0003 REACTOR PROTECTION SYSTEM HC.OP-AE3.IC-0001 Control Rod HC.OP-AB.ZZ-OOO Reactor Scram HC.OP-AB.RPV-0001 Reactor Power HC.OP-E0.Z-0101 RPV Control HC.OP-EO.ZZ-0101A ATWS-RPV Control HC.OP-EO.ZZ-0102 Primary Containment Control HC.OP-EO.ZZ-0202 Emergency RFY Depressurization HC.RE-IO.=-0001 Core Operations Guidelines HC.OP-IO.=-0006, POWER CHANGES DURING OPERATION NRC-OOI Page 16 of 21 Rev.: 060

NRC-001 REV-060 NRC-002 / 00

1.

WA 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EAl. Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

EA1.O1 Reactor Protection System RO 4.6 SRO 4.6 EA1.03 ARI/RPT/ATWS RO 4.1 SRO 4.1 Before Reactor Water Level reaches LVL I, the Crew manually actuates RPS and/or ARI to shutdown the reactor.

RPS has failed to scram the reactor both manually, and automatically. The RPV LVL 3 scram setpoint was chosen to ensure there is adequate protection for the fuel during transient analyses associated with coolant inventory decrease events. With no feedwater being supplied to the vessel and the reactor at power, water level will rapidly lower until the reactor is shutdown and steaming is reduced to decay heat levels.

Additionally, ARI is failed and will not automatically scram the reactor at -38. Operator action is required to shutdown the reactor. The need to manually initiate ARI or RPS by LVL 1 was chosen because it represents an acceptable level of performance considering the rate of RPV water level reduction in this scenario and the time required to implement the scram hard card. Also, if the plant is not scrammed by LVL 1, the subsequent shrink will reduce level to below TAF.

2.

WA 295024 High Drywell Pressure EA2 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

EA2.04 Suppression chamber pressure RO 3.9 SRO 3.9 WA 223001 Primary Containment Systems and Auxiliaries A2. Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions of operations:

A2.02 Steam bypass of the suppressions pool RO 3.9 SRO 4.1 Crew actuatescfive SRVs before Suppression Chamber pressure exceeds 33 psig.

If suppression chamber pressure cannot be maintained below the pressure suppression pressure, EOPs direct actions to emergency depressurize the reactor. A LOCA condition while in the action required region of the Pressure Suppression Pressure curve, could cause design containment limits to be exceeded.

Based upon the rate of pressure increase in this scenario, the upper limit of 33 psig is established to give the operator time to evaluate conditions and direct emergency depressurization actions.

NRC-00 1 Page 17 of 21 Rev.: 060

NRC-001 REV-060 NRC-002 / 00 HOPE CREEK NRC - PRA RELATIONSHIPS EVALUATION FORM EVENTS LEADING TO CORE DAMAGE y/N EVENT TRANSIENTS:

Turbine Trip Y

Loss of Feedwater Y/N EVENT SPECIAL INITIATORS:

Loss of ssw Loss of SACS MSIV Closure Loss of Condenser Vacuum Loss of RACS Loss of Instrument Air Inadvertent Open SRV Loss Of Offsite Power Y

ATWS Station Black Out Y

LOCA COMPONENT/TRAIN/SY STEM UNAVAILABILITY THAT INCREASES CORE DAMAGE FREQUENCY y/N COMPONENT, SYSTEM, OR TRAIN HPCJ RCIC One SRV One SSW Pump 1 Loop Circulating Water System ' 4 pumps y/N COMPONENT, SYSTEM, OR TRAIN Class 1E 120VAC Bus - A Train Class IE I20VAC Bus - D Train EDG A EDG B TACS OPERATOR ACTIONS IMPORTANT IN PREVENTING CORE DAMAGE y/N OPERATOR ACTION Y

Manual RPV Emergency Depressurization when required Manual RPV Depressurization during ATWS Initiation of RHR for Decay Heat Removal Y

Initiation of Containment Venting Restore Offsite power within 45 minutes SACS / SSW restoration after total loss of both systems Avoiding Loss of Feedwater during transient Recovery of the Main Condenser Complete this evaluation form for each Examination.

NRC-001 Page 18 of 21 Rev.: 060

NRC-001 REV-060 Rx Power: 80%

MWe: (May vary slightly):

Work Week: Any Risk Color: Green SMD: None River Temp: 65 Activities Completed Last Shift:

Power lowered to 80% and Control Rod Sequence Exchange performed Major Activities Next 12 Hours:

Maintain power at 80% until contacted by the Load Dispatcher, then return to 100% power Complete HC.OP-ST.BE-0002, Core Spray Pump Loop A Full Flow Test. Currently in progress and completed up to step 5.23 (pump testing).

Protected Equipment:

None Tagged Equipment:

SLC Pump AP-208 is tagged out for pump rebuild and is expected back within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPRM System is INOPERABLE due to an existing 10CFR21 issue. The OPRM System is still functional but is considered INOPERABLE per Technical Specifications.

No other equipment is Out of Service NRC-OOI Page 19 of 21 Rev.: 060

NRC EXAMINATION SCENARIO GUIDE REVIEWNALIDATION Note: This form is used as guidance for an examination team to conduct a review for the proposed exam scenario(s). Attach a separate copy of this form to each scenario reviewed.

SELF-CHECK NRC-002 REVIEWER:

1.

2.

The scenario has clearly stated objectives in the scenario.

The initial conditions are realistic, equipment and/or Instrumentation may be out of service, but it does not cue crew into expected events.

Each event description consists of:

3.

0 0

The event termination point The point in the scenario when it is to be initiated The malfunction(s) that are entered to initiate the event The symptomskues that will be visible to the crew The expected operator actions (by shift position)

4. The use of non-mechanistic failures (e+ pipe break) should be limited to one or a credible preceding event has occurred.

The events are valid with regard to physics and thermodynamics.

Sequencingltiming of events is reasonable (e.g. the crew has time to respond to the malfunctions in an appropriate time frame and implements procedures and/or corrective actions).

Sequencingltiming of events is reasonable, and allows for the examination team to obtain complete evaluation results commensurate with the scenario objectives.

If time compression techniques are used, scenario summary clearly so indicates.

The simulator modeling is not altered.

All crew competencies can be evaluated.

Appropriate reference materials are available (SOERs, LERs, etc.)

If the sampling plan indicates that the scenario was used for training during the requalification cycle, evaluate the need to modify or replace the scenario.

Proper critical task methodology used IAW NRC procedures.

5.

6.

7.

8.

9.

10.

1 I.

12-

13.

NRC-00 I Page 20 of 21 Rev.: 06

NRC EXAMINATION SCENARIO GUIDE VALIDATION (cont)

Note: The following criteria list scenario traits that are numerical in nature. A second set of numbers indicates a range to be met for a set of two scenarios. Therefore, to complete this part of the review, the set of scenarios must be available. The section below should be completed once per scenario set.

NRC:

002 NRC:

SELF-CHECK

1.

Total malfunctions inserted: 4-8/10- 14

2.

3.

Abnormal Events: 1-212-3 Malfunctions that occur after EOP entry: 1-413-6

4.

Major Transients: 1-212-3

5.

6.

EOPs used beyond primary scram response EOP: 1-313-5 EOP Contingency Procedures used: 0-3/1-3

7.

Approximate scenario run time: 45-60 minutes (one scenario may approach 90 minutes)

8.

EOP run time: 40-70% of scenario run time

9.

Crew Critical Tasks: 2-5/58

10. Technical Specifications are exercised during the test 1 1. Events used in the two scenarios are not repeated

12. The scenario sets for the exam week do not contain duplicate scenarios Comments:

NRC-001 Page 21 of 21 Rev.: 06

SCENARIO TITLE:

APRM Failure/ Recirc Pump Hi Vibd LOP SCENARIO NUMBER:

NRC-003 EFFECTIVE DATE:

EXPECTED DURATION:

1.0 Hours REVISION NUMBER:

00 PROGRAM:

-1 L.O. REQUAL r X-I INITIALLICENSE REVISION

SUMMARY

New Scenario.

PREPARED BY:

M. L. Brown 9/29/05 NRC Operations Examiner DATE FACILITY REVIEWER:

Nuclear Operations Training Supervisor -

Hope Creek DATE APPROVED BY:

NRC Chief Examiner DATE

NRC-001 REV-06 Enabling Objectives A.

The crew must demonstrate the ability to operate effectively as a team while completing a series of CREW CRlTICAL TASKS, which measure the crews ability to safely operate the plant during normal, abnormal, and emergency plant conditions.

(Crew critical tasks within this examination scenario guide are identified with an *.)

A.

B.

C.

D.

E.

F.

G.

H.

start 3rd RFP Load increase after RFP start A APRM Fails Drywell Chiller Compressor fails B Recirculation Pump High Vibration Loss of Offsite Power A EDG Output breaker fails to Auto close Recirc Suction pipe leak The plant is operating at 80% power, Middle Of Cycle returning to power after a mini-outage.

The Operators are at Step 5.4.48 of IOP-3, preparing to start the 3rd RFP per HC.OP-SO.AE-0001, Feedwater System Operation.

After starting the 3 RFP, the operators are to raise power to 100% by increasing recirc flow.

While raising power the A APRM fails causing the crew to enter Abnormal Procedure HC.OP-AB.IC-0004,NEUTRON MONITORING and bypass the APRM.

After the APRM is bypassed, the A Drywell Chiller Compressor will fail. Drywell pressure will rise causing the operators to enter AB.CONT-0001, Drywell Pressure. When pressure rises to >

0.75 psig, Operators will vent drywell. Once drywell pressure is lowering, B Recirc pump vibrations will increase causing the operators to enter AB. RPV-0003, Recirculation system.

Operators will reduce recirc pump speed in an attempt to clear the vibration alarm. Vibration will continue to increase and cause the operators to trip the B Recirc pump on high vibration. After tripping the Recirc pump the operators will have to insert rods to exit Region 2 on the Power flow map. As rods are being inserted a Loss of Offsite power occurs with the A EDG output breaker failing to close. In addition, HPCl and RClC fail to auto start. The loss of offsite power will cause a scram to occur, shortly after the scram occurs a small recirc suction pipe leak occurs, the operators will be forced to either restart HPCl and feed the reactor or Emergency Depressurize.

The scenario will end once the operators either stabilize level or Emergency Depressurize.

NRC-001 Page 2 of 17 Rev.: 060

NRC-001 REV-06 1

2

~~~

~

INITIALIZE the simulator to 80% power, MOL EVENT ACTION:

COMMAND:

PURPOSE:

EVENT ACTION:

COMMAND:

PURPOSE:

Zniriuf I Description COMPLETE Attachment 2 Simulator Ready-for-TrainingExamination Checklist of NC.TQ-DG.ZZ-0002(2).

3 1 Iniriuf I ET # I Description I

EVENT ACTION:

COMMAND:

PURPOSE:

NRC-001 Page 3 of 17 Rev.: 060

NRC-001 REV-06 tniriul I

I I

I Description Delay Ramp Trigger Init Val Final Val NONE ON NRC-00 1 Page 4 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments Crew assumes the watch and starts performing HC.OP-SO.AE-0001 section 5.6.1 (Note - Feedpump should be on recirc)

B o p OPENS HV-1769CP RFP c Discharge Stop Check valve 0

BOP closes HV-l772C, RFPT C Steam Low Pressure supply stop valve below seat drain o BOP opens HV-1751 C, RFPT C Low pressure steam isolation valve BOP depresses the SEL push-button as required to select DEMAND on the in-service RFPT(s) whose demand will be matched BOP Presses SEL push-button for the C RFPT to select SPEED CTRLR DMND decrease buttons as necessary to equalize demand signals while Monitoring:

BOP Presses Increase or RFPT Discharge Pressure RFPTDEMAND 0

RFPT FLOW Power Increase NRC-001 BOP matches Flow and speed and transfers RFPT C Speed Control to automatic by depressing the N M push-button and observing A illuminates BOP reports to CRS that 3rd RFP has been placed in service CRS directs RO/BOP to coordinate Power increase to 90% at < l%/minute using IOP-0003 Page 5 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Comments Event / Instructor Activity Expected PlantlStudent Response 2

2 2

A APRM Fails 2

After Power has been raised 5%

OR At the discretion of the Lead Examiner TRIGGER RT-1 RO/BOP coordinate raising power RO slowly increases Recirc pump speed BOP monitors RFP speed to ensure proper response RO diagnoses and reports A APRM has failed UPSCALE Should get a Half scram 0

CRS acknowledges report and enters HC. IO-AB. IC-0004, Neutron Monitoring 0

RO stops all Control Rod Should not be any Withdrawals control rod withdrawals 0

RO bypasses the A APRM in progress RO ensures all RPS trip conditions are clear RO turns the A RPS Trip logic key to reset and returns it to the normal position o RO verifies that RPS is reset CRS refers to Tech Specs 3.3.1 Should only be an INFO only LCO - only required to have 2 OPERABLE NRC-00 1 Page 6 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/Student Response Comments TURBINE BLDG CHILLER 0

BOP diagnoses/ observes the A ** Goal here is to have COMPRESSOR A FAILS Turbine Building Chiller trips Drywell pressure Once the CRS has addressed Tech Specs OR drywell vent.

At the discretion of the Lead Examiner increase to the point where the operators need to perform a TRIGGER RT-2 0

RO/BOP observe Drywell temperature/ pressure rising 0

CRS directs entry into AB.CONT-0001, Drywell Pressure No drywell cooling in progress RO - TERMINATE Drywell Inerting.

Turbine Bldg. Chilled water system is NOT operating properly RO MAXIMIZE Drywell Cooling by 0

All Drywell Fan Cooling Coils are ENSURING:

Open.

0 All Drywell Fans are running in Fast Speed.

0 Turbine Bldg. Chill Water system is operating properly.

RO - PERFORM the following:

0 Check Reactor Recirc. Pump 0

Check SRV Tailpipe Seals.

Temperatures.

- Talk to Archie/Steve what is the best/ quickest way to raise Drywell pressure so crew will vent containment NRC-001 Page 7 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlanUStudent Response B Recirc Pump Hiah Vibration Recirc pump vibration Once Drywell vent has been initiated OR At the discretion of the Lead Examiner RO diagnoses/ observes rising B TRIGGER RT-3 CRS directs entry into AB.RPV-0003, Recirculation System RO PRIOR to reducing Recirc Pump

Speed, PERFORM the following:

ENSURE the following controllers are in MANUAL SIC-R621A PUMP A SPD CONT SIC-R621B PUMP B SPD CONT RO RECORD affected pump speed:

0 B Recirc Initial Pump Speed RO MAINTAIN the affected Pump ALERT limit

[REFER to Table 21 clear as follows:

INTERMITTENTLY PRESS SIC-R621A(B) PUMP A(B) SPD CONT DECREASE push button on the affected Recirculation Pump.

required by Reactor Engineering lnst ructions.

INSERT Control Rods as Comments NRC-001 Page 8 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected PlantlS tuden t Response Comments CT-1 R O E ALERT limit cannot be maintained clear {REFER to Table 21 AND the affected Recirculation Pump Speed has been lowered by

>20% (below the value logged in Step K.l.B), THEN REMOVE the affected Recirc Pump from service IAW HC.OP-SO.BB-0002, Single Loop Operation.

RO removes pump from service IAW SO. BB-0002 CRS - IMPLEMENT 10-6 Requirements for Single Loop operations.

power/flow map CRS determines region of operation on CRS directs actions to exit Region B 0

RO either Raises Recirc flow with Recirc pump A or inserts control rods to exit Region B 0

CRS refers to Tech Spec 3.4.1 and COLR for SLO, determine APLHGR limit and APRM setpoints must be modified within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per IOP-6)

CRS contacts I&C to determine cause of failure and to adjust setpoints as required 0

CRS refers to IOP-6 and determines all appropriate actions have been taken in accordance with Section 5.3

- See if anything else needs to be done for single loop operation LOSS OF OFFSITE POWER Crew diagnoses LOSS Of Offsite. Malfunctions to be After CRS has notified I&C to power inserted adjust setpoints OR At the discretion of the Lead Examiner TRIGGER RT-4 Loss of Offsite power A EDG Output breaker fails to Auto Close HPCI fails to auto start RCIC fails to auto start NRC-001 Page 9 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/S tuden t Response Comments RECIRC PIPING LEAK Once LOP loads have sequencedon OR At the discretion of the Lead Examiner TRIGGER RT-5 CT-2 CRS directs entry into HC.OP-AB-0000 and HC.OP-AB.ZZ-01 35 RO locks Mode Switch in Shutdown BOP observes failure of A EDG to Auto close and Closes EDG output breaker action

- Ask Archie if it is expected to have level drop

< 12.5 and force entry into EO-101

- Check with Archie if this is the appropriate RO verifies the Scram RO inserts SRMs and IRMs AND selects IRMs on the Recorders BOP verifies H2 injection system tripped BOP Trips the Main turbine and verifies Generator lockout is 0 Mwe RO maintains level between

+12.5 and 54 RO starts RClC RO observes RPV level decreasing, and Drywell temperature/ pressure increasing CRS directs entry into EO-0101 and EO-0102 based on Drywell pressure > 1.68 psig and level <

12.5 may not observe the Crew may elect to auto start HPCl and control level, if this occurs they failure of HPCl to auto start RO starts HPCl and controls level NRC-00 1 Page 10 of 17 Rev.: 060

NRC-001 REV-06 V.

SCENARIO GUIDE SEQUENCE AND EXPECTED RESPONSE Event / Instructor Activity Expected Plant/ Student Response Comments After HPCl has been started and level is being controlled, scenario may be terminated.

NRC-001 0

Once Scenario has been terminated have the SRO classify the event.

a 0

Page 1 1 of 17

- Talk to Steve about stopping point Rev.: 060

NRC-001 REV-06 A.

B.

C.

D.

E.

F.

G.

H.

I.

J.

K.

L.

M.

N.

0.

P.

Q.

R.

S.

T.

U.

V.

NC.TQ-DG.ZZ-0002 Conduct of Simulator Training.

NUREG 1021 Examiner Standards JTA Listing Probabilistic Risk Assessment Technical Specifications Emergency Plan (ECG)

Alarm Response Procedures (Various)

SH.OP-AS.ZZ-000 1 Operations Standards SH.OP-AE'.ZZ-0101 Post Transient Response Requirements SH.OP-M.ZZ-0108 Operability Assessment and Equipment Control Program HC.OP-IO.ZZ-0003 Startup from Cold Shutdown to Rated Power HC.OP-AB.IC-0003 REACTOR PROTECTION SYSTEM HC.OP-AB.IC-0001 Control Rod HC.OP-AB.ZZ-000 Reactor Scram HC.OP-AB.RPV-0001 Reactor Power HC.OP-EO.ZZ-0101 RPV Control HC.OP-EO.ZZ-0 10 I A ATWS-RPV Control HC.OP-EO.ZZ-0102 Primary Containment Control HC.OP-EO.ZZ-0202 Emergency RPV Depressurization HC.RE-IO.ZZ-0001 Core Operations Guidelines HC.OP-IO.ZZ-0006, POWER CHANGES DURING OPERATION NRC-OO 1 Page 12 of 17 Rev.: 060

NRC-001 REV-06 NRC-003 / 00

1.

WA 202001 Recirculation System A2 Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.17 Loss of seal cooling water RO 3.1 SRO 3.2 This action is listed as a Retainment Override in the Abnormal Procedure, a time limit of 2 minutes is deemed adequate for the operator to recognize the condition and take the appropriate action.. The basis of this action is to prevent pump damage and potential piping damage due to vibration. Damage to the pump casing is a degradation of a Reactor Coolant System boundary.

CREW secures B Reactor Recirc pump within two minutes of Vibration reaching the DANGER limit IAW guidance in AB.RPV-0003..

2.

WA 206000 High Pressure Coolant Injection System A3 Ability to monitor the operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including:

A3.03 System lineup RO 3.9 SRO 3.8 WA 295031 Reactor Low Water Level EA1. Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL:

EA1.02 High Pressure Coolant Injection RO: 4.5 SRO 4.5 Before RPV water level reaches -161 and without Emergency Depressurizing, CREW manually places HPCI in service and injects with HPCI to maintain Reactor water level above -1 61.

HPCI has failed to automatically start. HPCI is the only High Pressure injection system available with adequate capacity to maintain RPV water level. If RPV water level is allowed to drop below -161, the fuel will be uncovered and the fuel cladding will be challenged. This would escalate the event to a General Emergency. HC.OP-AB.ZZ-0001 Attachment 6 has the necessary guidance to step the operator through manually initiating HPCI in the injection mode. The rate of level drop in this scenario is very slow and provides more than adequate time to execute the guidance an restore RPV level with HPCI.

NRC-001 Page 13 of 17 Rev.: 060

NRC-002 / 00 NRC-001 REV-06 HOPE CREEK NRC - PRA RELATIONSHIPS EVALUATION FORM EVENTS LEADING TO CORE DAMAGE y/N EVENT TRANSIENTS:

Turbine Trip Y

Loss of Feedwater EVENT SPECIAL INITIATORS:

Loss of ssw Loss of SACS MSIV Closure Loss of RACS Loss of Condenser Vacuum Inadvertent Open SRV Loss of Instrument Air Loss Of Offsite Power Station Black Out COMPONENT/TRAIN/SYSTEM UNAVAILABILITY THAT INCREASES CORE DAMAGE FREQUENCY Y/N COMPONENT, SYSTEM. OR TRAIN HPCI RCIC One SRV Y

ATWS Y

LOCA One SSW Pump / Loop Circulating Water System - 4 pumps y/N COMPONENT, SYSTEM, OR TRAIN Class 1E 120VAC Bus - A Train Class 1E 120VAC Bus - D Train EDG A EDG B TACS OPERATOR ACTIONS IMPORTANT IN PREVENTING CORE DAMAGE y/N OPERATOR ACTION Y

Manual RPV Emergency Depressurization when required Manual RPV Depressurization during ATWS Initiation of RHR for Decay Heat Removal Initiation of Containment Venting Restore Offsite power within 45 minutes SACS / SSW restoration after total loss of both systems Avoiding Loss of Feedwater during transient Y

Recovery of the Main Condenser Complete this evaluation form for each Examination.

NRC-001 Page 14 of 17 Rev.: 060

NRC-001 REV-06 Rx Power: 80%

W e : (May vary slightly):

Work Week: Any Risk Color: Green SMD: None River Temp: 65 Activities Completed Last Shift:

Power lowered to 80% and Control Rod Sequence Exchange performed Major Activities Next 12 Hours:

Maintain power at 80% until contacted by the Load Dispatcher, then return to 100% power Complete HC.OP-ST.BE-0002, Core Spray Pump Loop A Full Flow Test. Currently in progress and comple (pump testing).

d up D step 5.23 Protected Equipment:

None Tagged Equipment:

SLC Pump AP-208 is tagged out for pump rebuild and is expected back within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPRM System is INOPERABLE due to an existing 10CFR21 issue. The OPRM System is still functional but is considered INOPERABLE per Technical Specifications.

No other equipment is Out of Service NRC-001 Page 15 of 17 Rev.: 060

NRC EXAMINATION SCENARIO GUIDE REVIEWNALIDATION Note: This form is used as guidance for an examination team to conduct a review for the proposed exam scenario(s). Attach a separate copy of this form to each scenario reviewed.

SELF-CHECK NRC-002 REVIEWER:

1.

2.

The scenario has clearly stated objectives in the scenario.

The initial conditions are realistic, equipment and/or Instrumentation may be out of service, but it does not cue crew into expected events.

Each event description consists of:

3.

0 0

The event termination point The point in the scenario when it is to be initiated The malfunction(s) that are entered to initiate the event The symptomskues that will be visible to the crew The expected operator actions (by shift position)

4. The use of non-mechanistic failures (e.g. pipe break) should be limited to one or a credible preceding event has occurred.

The events are valid with regard to physics and thermodynamics.

Sequencingltiming of events is reasonable (e.g. the crew has time to respond to the malfunctions in an appropriate time frame and implements procedures and/or corrective actions).

Sequencingltiming of events is reasonable, and allows for the examination team to obtain complete evaluation results commensurate with the scenario objectives.

If time compression techniques are used, scenario summary clearly so indicates.

The simulator modeling is not altered.

All crew competencies can be evaluated.

Appropriate reference materials are available (SOERs, LERs, etc.)

If the sampling plan indicates that the scenario was used for training during the requalification cycle, evaluate the need to modify or replace the scenario.

Proper critical task methodology used IAW NRC procedures.

5.

6.

7.

8.

9.

10.

11.

12.

13.

NRC-00 1 Page 16 of 17 Rev.: 06

NRC EXAMINATION SCENARIO GUIDE VALIDATION (cont)

NRC Examination Validation:

Note: The following criteria list scenario traits that are numerical in nature. A second set of numbers indicates a range to be met for a set of two scenarios. Therefore, to complete this part of the review, the set of scenarios must be available. The section below should be completed once per scenario set.

NRC:

002 SELF-CHECK

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

Total malfunctions inserted: 4-8110- 14 Malfunctions that occur after EOP entry: 1-4/3-6 Abnormal Events: 1-2/2-3 Major Transients: 1-2/2-3 EOPs used beyond primary scram response EOP: 1-3/3-5 EOP Contingency Procedures used: 0-3/1-3 Approximate scenario run time: 45-60 minutes (one scenario may approach 90 minutes)

EOP run time: 40-70% of scenario run time Crew Critical Tasks: 2-5/5-8 Technical Specifications are exercised during the test Events used in the two scenarios are not repeated The scenario sets for the exam week do not contain duplicate scenarios NRC:

Comments:

NRC-001 Page 17 of 17 Rev.: 06

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