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2009 Initial Examination 05000259/2008/301, 05000260/2008/301 and 05000296/2008/301 Draft SRO Written Examination
	ML091130442
Person / Time
Site:	Browns Ferry
Issue date:	02/25/2008
From:	; Tennessee Valley Authority
To:	; NRC/RGN-II
References
	50-259/08-301, 50-260/08-301, 50-296/08-301
	Download: ML091130442 (191)
	v • d • e

{{#Wiki_filter:ES-401 Written Examination Quality Checklist Form ES-401-6 I Facility: Browns Ferry Date of Exam: 2/25/2008 Exam Level: RO D SRO [8J Initial Item Description a b* c#

1. Questions and answers are technically accurate and applicable to the facility. '()W9 ~.

2. a. NRC KlAs are referenced for all questions.

b. Facility learning objectives are referenced as available. ~ a

3. SRO questions are appropriate in accordance with Section D.2.d of ES -401 (~ 0L-

4. The sampling process was random and systematic (If more than 4 RO or 2 SRO questions were repeated from the last 2 NRC licensing exam, consult the NRR OL program office). (:RW- (Z{L

5. Question duplication from the license screening/audit exam was controlled as indicated below (check the item that applies) and appears appropriate:

           ~ the audit exam was systematically and randomly developed; or

_ the audit exam was completed before the license exam was started; or

           ~ the examinations were developed independently; or
           ~ the licensee certifies that there is no duplication; or

_other (explain) ~~ e~ Bank Modified New

- -

6. Bank use meets limits (no more than 75 percent 4

from the bank, at least 10 percent new, and the rest new or modified); enter the actual RO / SRO -only question distribution(s) at right. 3

                                                                           \'2. tfo    I(::.OJt>     -7Z.fSJb 18 I~ f21..

7. Between 50 and 60 percent of the questions on the RO Memory CIA exam are written at the comprehension /analysis level; the SRO exam may exceed 60 percent if the randomly SRO SRO

                                                                          .3 jlZOJo l'Z/8b% ~

selected KAs support the higher cognitive levels; enter the actual RO / SRO question distribution(s) at right. fli-References/handouts provided do not give away answers (~ 8. or aid in the elimination of distractors. gIL

9. Question content conforms with specific KIA statements in the previously approved

                                                                                                                 ~ l&-

examination outline and is appropriate for the Tier to which they are assigned; deviations are justified

10. Question psychometric quality and format meet the guidelines in ES Appendix B. (~ C2L

(~ 0L

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

~~~

Date

                                                                                                          " .-,.
                                                                                                  ""--

a. Author Robert M. Spadoni 2/15/2008

b. Facility Reviewer (*) Randv E. Knioht -eC~~ ./"

2/15/2008 U "'l

c. NRC Chief Examiner (#)

d. NRC Regional Supervisor Note:

The facility reviewer's initials/signature are not applicable for NRC-developed examinations.

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

ES-401, Page 29 of 33

Name: _____________________________ 0610 NRC SRO Exam Form: 0 Version: 0

1. SRO 295006AA2.01 001lCIAlT1G1IRPSI1295006AA2.01l4.6/SRO ONLYINEW 11127/07 RMS Given the following plant conditions:

Unit 1 was at 100% power with RPS 'A' on Alternate due to problems with the MG set.
A fault on '1 B' 480V RMOV Board causes a loss of the board.
The following conditions are observed:

                     - Multiple control rods failed to insert on the scram.
                     - Reactor pressure is 960 psig and slowly rising with 1 SRVopen.

Which ONE of the following describes the approximate value of reactor power, the appropriate actions and the basis for the actions? I Reactor power is (1) . Maintain RPV level between ______-->,;;{2::..<,)_____ Boron mixing would be provided by ______......(3"-")_________ (1 ) (2) (3) A. less than 5%. (+)2 and (+)51 inches. forced circulation. B. greater than 5%. (-)50 and (-)100 inches. forced circulation. C. less than 5%. (+)2 and (+)51 inches. natural circulation. D!' greater than 5%. (-)50 and (-)100 inches. natural circulation. Sunday, February 17, 2008 3:28:26 PM 1

0610 NRC SRO Exam KIA Statement: 295006 SCRAM I 1 AA2.01 - Ability to determine andlor interpret the following as they apply to SCRAM: Reactor power. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the reactor power level using alternate methods and select the appropriate procedure to address those conditions. Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17,20083:28:26 PM 2

0610 NRC SRO Exam REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly, the candidate must: Determine the relationship between MSRV steam flow and reactor power as follows: Total rated steam flow through 13 MSRVs = 84% divided by 13 MSRVs=6.5%. The fact that the MSIVs are closed along with the value and trend of reactor pressure should provide sufficient information to conclude the reactor power is above 5 percent which is the breakpoint for C5 for lower water level. The highest allowable water level in this condition is -50 inches. In RC/Q, 5 percent power is the breakpoint for removing recirc pumps (forced ciruclation) from service. This value is selected since boron injection into the lower head will be sufficiently mixed inside the core by natural convention currents from the steam flow out of the reactor. At power levels below 5%, recirc pumps are needed to mix boron. D

correct:

A* incorrect: Reactor power is above 5%. However, if power was below 5%, the level band along with the assumption of forced circulation would be correct. B

incorrect: Based on the assumption of forced circulation. Recirc pumps must be tripped if power is above 5%.

C

incorrect: Reactor power is above 5%. However, the level band is correct for the conditon of less than 5%. However, natural circulation is not sufficient to mix the boron with power level below 5%.

Sunday, February 17, 2008 3:28:26 PM 3

OPL171.009 Revision 10 Page 15 of 62 (6) The worst over pressure transient is: (a) 3-second closure of all MSIVs neglecting the direct scram (valve position scram) .

        . (b)     Results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable over pressure limit of 1375 psig bottom head pressure.

(7) To meet operational design, the analysis of the plant isolation transient (generator load reject without bypass valves) shows that 12 of the 13 valves limitpeak pressure to a value well below the limit of 1375 psig.

b. The total safety I relief valve capacity has been established to meet the over pressure protection criteria of the ASME code.

(1) There are 13 Safety I Relief valves. (a) Each SRV has a capacity of 905,OOOlb/hr @ 1135psig. This gives a total capacity ~ 84.1 % (79.5% EPU) design steam flow at the reference pressure. (b) Valve leakage is detected by Obj. V.B.6 a temperature element and an Obj. V.CA acoustic monitor on each tailpipe. However, only the acoustic monitor will generate an alarm on panel 9-3.

CS, LEVEUPOWER CONTROL BASES EOI PROGRAM MANUAL SECTION O~V-K . I DISCUSSION: STEP CS-ll (Continued) I These interrelationships between RPV water level, natural circulation core flow, and reactor power have been observed in BWRs with RPV water level in or near the normal operating band. Computer analysis and scale model tests have confinned the continued validity of these fundamental thermal hydraulics and reactor physics principles for RPV water levels at and below the elevation of the steam separators. The interrelationships between RPV water level, natural circulation core flow, and reactor power are graphically illustrated in the following figures: JO Jo CORE REACTOR flOW 20 POWER RPV Water lttYel Dnd Natural (Thermol) 20

I of Rated flow bnd Reoc:lor CirclJlolton Core now Relationship Pu**r Rekallonshlp "of Roled 10 10

                 -10            -5        o           +5                                     10           20          JO RPV WATER lEV£l                                                        CORE flDI/

(lI: of Ratod) feet Above SI!Ip"rotor Skir'l JO REACTOR POW[R (Thermo I) 20 RPV Wo\tr Level and R_tor

                                        " of Roted                  Power R.lGlionship 10
                                                           -10        -~        0         +5 RPII WATER LEVEL f ""I AbOYe Seporalor Skin REVISION 0                                                 PAGE 31 OF 110                                          SECTION O-V-K

0610 NRC SRO Exam

2. SRO 295021AA2.07 00 IIMEM/Tl/GIIRHRlSDCIB31129502lAA2.071ISRO ONLYINEW 113112008 Unit ilsTn Mode 5 with a fuel shuffle in progress. Alternate Decay Heat Removal (ADHR) is in service. Maintenance activities required installation of Jet Pump Plugs.

Technical Specification Required Action 3.9.7.C.1 states:

         "Verify reactor coolant circulation by an alternate method."
                                                    ~"'--~

Which ONE of the following conditions would satisfy this action statement?' !~~~i-1 <~9'-~ A. 01 Verify RPV level is above (+) 70 inches. B. Verify a Recirc Pump is in service at 480 RPM. C. Verify ADHR is in service with a minimum of one pump. D. Verify vessel bottom head to feedwater nozzle L\T is within 5°F. KIA Statement: 295021 Loss of Shutdown Cooling /4 AA2.07 - Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor recirculation flow KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the actions necessary to establish coolant circulation flow following a Loss of Shutdown Cooling with specific emphasis placed on refueling activities.

References:

2-AOI-74-1 and 2-01-68 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of information and apply this to the complicance of technical specification. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (7) Fuel handling facilities and procedures. This unique situation requires the SRO to take plant conditions and satisfy a Tech Spec action statement. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:26 PM 4

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must: Demonstrate knowledge of the specific reactor conditions and use this information to determine the appropriate implmentation of a Technical Specification Action statement. The distractors provided are all indicative of reactor coolant circulation. However, the conditions in the stem disallow the use of these other conditions. A - correct: because the coolant circulation is provided by natural ciruculation due to density differences in the water as it is alternately heated and cooled. B - incorrect: based on the initial conditions; which is the jet pump plugs are installed. This eliminates NOT only RHR SDC, but also the Recirculation System's common discharge. C - incorrect: based on the system characteristics of ADHR. The suction and discharge of this system is in the fuel pool which will not provide direct coolant circulation. ADHR is the heat removal mechanism which satisfies other TS actions in this condition, such as fuel pool temperature, but not coolant circulation by alternate methods. D - incorrect: This is an indication of thermal stratification, not coolant circulation by alternate methods. Sunday, February 17, 2008 3:28:26 PM 5

BFN Loss of Shutdown Cooling 2-AOI-74-1 Unit2 Rev. 0032 Page 15 of 31 4.2 Subsequent Actions (continued) [12.10] WHEN time permits after RHR pump is started, THEN VERIFY RHR Pump Breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged. 0 [12.11] SLOWLY THROTTLE RHR HX 2A(2C)(2B)(2D) RHRSW OUTLET VALVE, 2-FCV-23-34(40)(46)(52), to obtain desired cooldown rate. 0 [13] IF necessary, RAISE RWCU flow rate to maximum AND maximize RWCU blowdown as required to maintain reactor coolant temperatures less than 200°F on all indications. REFER TO 2-01-69. o CAUTION Accurate coolant temperatures will NOT be available if all forced circulation is lost. [14] [NER/C] IF forced circulation has been lost AND vessel cavity is less than 80 inches, THEN (Otherwise N/A) PERFORM the following: o [14.1] RAISE RPV water level to 80 inches as indicated on RX WTR LEVEL FLOOD-UP, 2-U-3-55. 0 [14.2] MAINTAIN RPV water level between +70 inches to

                       +90 inches as indicated on RX WTR LEVEL FLOOD-UP, 2-U-3-55.                                       0

[14.3] RAISE monitoring frequency of reactor coolant temperature and pressure, using multiple indications. 0 [15] IF the affected loop of RHR cannot be placed back in Shutdown Cooling, THEN RESTORE power to affected breakers per 2-POI-74-2 if applicable (Otherwise N/A) PLACE the alternate loop of RHR in Shutdown Cooling. REFER TO 2-01-74. (Otherwise N/A) o

0610 NRC SRO Exam

3. SRO 295024G2.1.28 001lCIAlTlGl1BASIS11295024G2.1.2811SRO ONLYINEW 113112008 An ATWS and LOCA have occurred on Unit 2, resulting in the following plant conditions: '

Suppression Chamber pressure is 51 psig and rising 1 psig every 5 minutes.
Drywell temperature is 325°F and rising 1°F every 15 minutes.
Suppression Pool level is 19 ft. and steady.

Which ONE of the following describes the required actions and the basis for those actions? -::r:-~) A\?f~'~' ;:> C~~~ce? DW>JL~l-t?J A. Initiate Drywell Sprays ts teFminaieJemperatureand -pressl:lf8-Fise. Do NOT vent the Suppression Chamber irrespective of off-site release rates, since this is an unfiltered release path. B. Initiate Drywell Sprays to-termtnatetemlJeratlJre*-and~presstJTe~rtse. Vent the Suppression Chamber irrespective of off-site release rates, to ensure long term containment capabilities are maintained. C.oI Do NOT intiate Drywell sprays .Gl:le~tetRe possiblity.QUmmegiat&*eGRte*iflfflent failure. Vent the Suppression Chamber irrespective of off-site release rates, to ensure long term containment capabilities are maintained. Do NOT intiate Drywell sprays dble"tothepossiblity*gf,imffledtate'contatnmeni- ffitl.l:lre. Do NOT vent the Suppression Chamber irrespective of off-site release rates, since this is an unfiltered release path. Sunday, February 17, 2008 3:28:26 PM 6

0610 NRC SRO Exam KIA Statement: 295024 High Drywell Pressure I 5 2.1.28 - Conduct of Operations Knowledge of the purpose and function of major system components and controls. KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate knowledge of the basis for primary containment high pressure limits.

References:

FSAR limitations described in EOIPM Section O-V-D page 55 of 244 Level of Knowledge Justification: This question is rated as CIA due to the necessity to recall basis for actions along with processing information to determine the correct course of actions. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (1) CoCnditions and limitations in the facility license. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:26 PM 7

0610 NRC SRO Exam REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly, the candidate must: Demonstrate knowledge of the bases considered for determination of the Primary Containment pressure limit and the specific limitation applicable to BFN. This limitation is provided in greater detail in the BFN Updated Final Safety Analysis Report (UFSAR), however the bases associated with the distractors are outlined in the EOI Program Manual and not in the UFSAR. The limitation is defined as the lesser of four possible limitations. Plausibility for the distractors is assured based on their potential applicability, under slightly different conditions. C - correct: A - incorrect: Suppression Pool water level above 18 feet will prevent the equalization of pressure from the SC air space and the DW airspace because the vacuum breakers are covered. This condition will lead to immediate containment failure. Emergency venting of the contaiment is an unfiltered release path, however a short duration release that maintains the contaiment equipment in tact is preferred over complete contaiment failure. B - incorrect: Suppression Pool water level above 18 feet will prevent the equalization of pressure from the SC air space and the DW airspace because the vacuum breakers are covered. This condition will lead to immediate containment failure. Emergency venting is correct. D - incorrect: this choice is incorrect since emergency venting is required prior to exceeding the containment design pressure. Amplification: EOIPM Section O-V-D page 55 of 244 is attached for reference. Sunday, February 17, 2008 3:28:26 PM 8

 ',EOI-2, PRIMARY CONTAINMENT CONTROL BASES                                    EOI PROGRAM MANUAL SECTION O-V-O

I DISCUSSION: STEP PCIP-7 This decision step has the operator evaluate the present status of suppression pool water level to determine if suppression chamber to drywell vacuum breakers are operable.

Suppression chamber to drywell vacuum breakers will not function as designed if any portion of the valve is covered with water. Suppression pool water level <A.5> corresponds to elevation of the bottom of the vacuum breakers less vacuum breaker opening pressure in feet of water. Ifvacuum breakers are submerged, they are inoperable. Drywell spray operation with vacuum breakers inoperable (i.e., with no vacuum relief capability) may cause primary containment differential pressure capability to be exceeded because of the inability to equalize pressure through vacuum breakers. If suppression pool water level is below <A.5>, suppression chamber to drywell vacuum breakers are not submerged and will function as designed. The operator continues at Step PC/P-S. Ir"suppression pool water level is at or above <A.S>, drywell spray operation is not permitted and the operator is directed to Step PC/P-ll. REVISION 0 PAGE 45 OF 244 SECTION O-V-O

EOl-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-O

( I DISCUSSION: STEP PCIP~16 TIlls action step directs the operator to emergency vent the suppression chamber as specified in EOI Appendix 13. EOI Appendix 13 provides step-by-step guidance on how to emergency vent I the suppression chamber and defeat isolation interlocks, irrespective of the resultant release rate. Because this step is prioritized with the miniature before decision step PCIP-14 symbol, this

      . action must be performed before suppression chamber pressure exceeds <A.61>, Primary Containment Pressure Limit.

Emergency venting of the suppression chamber is performed to: 1) assure that integrity of primary containment is maintained, and 2) prevent core damage that might be caused by inability to vent the RPV as necessary to permit injection of water to cool the core. The consequences of not emergency venting may result in a catastrophic loss of primary containment integrity with a subsequent uncontrolled release of radioactivity much greater than that which might otherwise occur. Direction is given to vent only until suppression chamber pressure can be reduced and maintained below Primary Containment Pressure Limit. This action minimizes offsite radioactivity release rate while still assuring primary containment integrity. Venting via the suppression chamber is preferred to take advantage of the scrubbing effect (Le. vented gas from primary containment, that exits from downcomers and passes through the suppression pool water volume). The scrubbing effect minimizes the amount of radioactivity released .

REVISION 0 PAGE 61 OF 244 SECTION O-V-O

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0610 NRC SRO Exam

4. SRO 295026EA2.01 001lCIA/Tl/GlIE0I-211295026EA2.01//SRO ONLYINEW 113112008 Unit 2 is at 15% power following a refueling outage.

( ~ \ t

Surveillance 2-§)R-:3.~1.8, "HPCI Main and Booster Pump Set Developed Head and Flow Rate T at Rated Pressure" is in progress.
The Unit Operato announces "SUPPR POOL AVERAGE TEMP HIGH" Alarm and bulk pool te perature is 96°F and rising at approximately 1°F every 3 minutes. \ ' . ceJ

                                      '(\(""",,.11 y' .

1,;;t;;P.s I I

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RHR SysteiiiJjs in Suppression Pool Cooling. .? h "'-' v\'.""'""'G (j' Which ONE of the following describes the mlnimum required actions in accordance with Emergency Operating Instructions and Technical Specifications?

                                                                                          ------- 9SF ~ loS" REFERENCE PROVIDED A. Maximize RHRSW flow on the inservice RHR Heat Exchangers.

NO Tech Spec Actions required at this time. B." Place additional RHR Pumps in suppression pool cooling. N~Tech Spec actions required at this time. C. Place additional RHR Pumps in suppression pool cooling. IS Verify suppression pool temp ren iains below 110°F and reduce suppression pool temp to less than 95°F within 24 hours. D. Maximize RHRSW flow on the inservice RHR Heat Exchangers. Verify suppression pool temp ~ below 110°F and reduce suppression pool temp to less than 95°F within 24 hours. Sunday, February 17, 2008 3:28:26 PM 9

0610 NRC SRO Exam KiA Statement: 295026 Suppression Pool High Water Temp. / 5 EA2.01 - Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature KiA Justification: This question satisfies the KIA statement by requiring the candidate to interpret plant indications of Suppression Pool temperature and determine the appropriate procedures required to address those conditions.

References:

2-EOI-2, U2 Tech Spec Section 3.6.2.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 10

0610 NRC SRO Exam REFERENCE PROVIDED: U2 Tech Spec Section 3.6.2.1 ( Ptausibility Analysis: In order to answer this question correctly, the candidate must: Determine the actions of EOI-2 to m~imize all availalbe SP cooling. This reference is not provided. The term "maximize available SP cooling" is used to ensure that all the function, of the available system, is in service. This is accomplished by maximizing the number of RHR pumps along with the maximum available RHRSW flow. 1b~J~~tJbPt only_one loop is in service reflects rul~.s regarding operation in SP cooling mode as

               ~,~- .~- "-' ,,~.,~. _  ,  """.a.,,~ **..* ,~,",' "'"ii~>-~'- ,~,. -,~ ,-." ." -      .. -      -.- - " ..-' .~ - - . ,--1 outlined in NR"GTflformatlon Notice_87~tO. This is based on the fact that the LPCI           """~""-+-"~1<.,.,~

function may be jeopardized during a LOOP LOCA; due to water hammer damaging the system. Only while executing the EOls is it acceptable to operate both loops of RHR for SE?-cooling. Also the SRO must determine the applicability of the 95 of and 105 of Tech Spec limit based on given plant conditions. Normally, the EOI entry cqndition and the TS limit are the same, ie 95 of. However, the stem provides information about the HPCI testing which satisifies the TS LCO 3.6.2.1.b. Once the heat addition is stopped and pool temp is above 95 of, then the LCO is NOT met and action is required; which is to maintain SP temp below 110°F and reduce to less than 95°F within 24 hours. B - correct: A - incorrect: Additional RHR pumps are required to comply with the EOI required actions. The Tech Spec action is correct. C - incorrect: The LCO is met and NO Tech Spec action is required. The use of RHR is correct in these conditions. D - incorrect: Additional RHR pumps are required to comply with the EOI required actions. Also, the LCO is met and NO action is required. Sunday, February 17, 2008 3:28:27 PM 11

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(REFERENCES PROVIDED TO ( CANDIDATE

Suppression Pool Average Temperature 3.6.2.1 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LCO 3.6.2.1 Suppression pool average temperature shall be:

a. s 95°F when any OPERABLE intermediate range monitor (IRM) channel is > 70/125 divisions of full scale on Range 7 and no testing that adds heat to the suppression pool is being performed;

b. s 105°F when any OPERABLE IRM channel is> 70/125 divisions of full scale on Range 7 and testing that adds heat to the suppression pool is being performed; and

c. s 110°F when all OPERABLE IRM channels are s 70/125 divisions of full scale on Range 7.

APPLICABILITY: MODES 1,2, and 3. BFN-UNIT 2 3.6-24 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool A.1 Verify suppression pool Once per hour average temperature average temperature

   > 95°F but ~ 110°F.             ~110°F.

AND AND Any OPERABLE IRM A.2 Restore suppression pool 24 hours channel> 70/125 average temperature to divisions of full scale on ~ 95°F. Range 7. AND Not performing testing that adds heat to the suppression pool. S. Required Action and B.1 Reduce THERMAL 12 hours associated Completion POWER until all Time of Condition A not OPERABLE IRM met. channels are ~ 70/125 divisions of full scale on Range 7. (continued) BFN-UNIT 2 3.6-25 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Suppression pool C.1 Suspend all testing that Immediately average temperature adds heat to the

   > 105°F.                        suppression pool.

AND Any OPERABLE IRM channel> 70/125 divisions of full scale on Range 7. AND Performing testing that adds heat to the suppression pool. D. Suppression pool 0.1 Place the reactor mode Immediately average temperature switch in the shutdown

   > 110°F but::; 120°F.           position.

AND 0.2 Verify suppression pool Once per 30 average temperature minutes

120°F.

AND 0.3 Be in MODE 4. 36 hours (continued) BFN-UNIT 2 3.6-26 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Suppression pool E.1 Depressurize the reactor 12 hours average temperature vessel to < 200 psig.

     > 120°F.

AND E.2 Be in MODE 4. 36 hours ( BFN-UNIT 2 3.6-27 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature 24 hours is within the applicable limits. 5 minutes when performing testing that adds heat to the suppression pool BFN-UNIT 2 3.6-28 Amendment No. 253

0610 NRC SRO Exam

5. SRO 295037EA2.05 001lC/A/Tl/G1IEOI-l//295037EA2.05//SRO ONLYINEW 11129/07 RMS Unit 3 is operating at reduced load due to grid instabilities.

            *   '3C' Diesel Generator (DIG) is out of service for emergent repair of the Fuel Oil Priming Pump.

A total Loss of Offsite Power occurs.
RPS Circuit Protector 'A l' trips and cannot be reset.
ALL systems function a§ designed.

Which ONE of the following statements describes the available indications used to determine reactor power as it relates to EOI u~ge and the basis for this conclusion (assume no operator actions)? A'!' Control/od positiollrllay ~E:lusedtQEI~_!~Jmine-PQwerJeyel as iot relates to EOI RClO.(hPRMs are NOTavailable due to loss of power supply. ;

                      -----~ - --------.... o ___ ~-~-~ __
                                                              ---~--"'" -

B. APRM power must be used to determine power level as it relates to EOI RC/O. Control rod position indications are NOT available due to loss of power supply C. Control rod position may be used to determine power level as it relates to EOI RC/O. However, APRM indications are needed to ensure the appropriate level band is selected in C5. D. APRM power may be used to determine power level as it relates to EOI RC/O. However, control rod positions are needed to ensure the reactor is shutdown under all condition without boron. Sunday. February 17. 2008 3:28:27 PM 12

0610 NRC SRO Exam KIA Statement: 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown I 1 EA2.05 - Ability to determine andlor interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Control r09 position. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions to restore control rod position indication during an ATWS.

References:

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17,20083:28:27 PM 13

0610 NRC SRO Exam REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly, the candidate must: Understand the following concepts as it relates to the conditions in the stem: No RPS power is available which eliminates the use of APRMs for power level determination. Rod Position indication is available; as it is powered from D/G backed Unit Preferred or Batteries. and based on the conditions stated in the stem, sufficient power is available to use the RPIS function. Subcritical under all conditions is sufficient to detemine the power and level control functions of the EOls. A - correct: B- incorrect: APRMS are NOT available because of the loss of power. In addition, control rod position indication IS available. C - incorrect: Although APRM power is used to determine the appropriate level band in C5, "Level/Power Control", since Control Rod positions are available, it is inappropriate to enter EOI C5, "Level/PowE?r Control." D - incorrect: APRMS are NOT available because of the loss of power. ( Sunday, February 17, 2008 3:28:27 PM 14

OPL 171.028 Revision 17 Appendix C Page 38 of 50 From 480v RMOV 1 B, (2B-Unit 2) (3B- Unit 3) FLYWHEEL FLYWHEEL START START I-- BREAKER 1-- I .,---------, (\I I 480VAC 480VAC eJ-RMOVBdA RMOVBd B I I I I I I I On U2 only, this I I transformer is I I shared with the I I /1 I I Unit Preferred I I To Unit Preferred - ---' System RPS MG SetA Alternate Supply RPS MG SetB (Unit 2 Only) Fused Disconnect OUTPUT

      )    BREAKER OUTPUT BREAKER          )

( A1 Circuit Protectors Open On: High Voltage :s 132 V Low Voltage 2:108.5V Low Frequency 2: 56 Hz ( C1 ( B1 A2 C2 B2 Breaker 902 Breaker 952 Batt Bd 1 (2,3) Batt Bd 1 (2,3) Pnl9 Pnl9 NOR ALT ALT INTERLOCKED UiU) o~"O .- Ul 0 r:: C OJ OJ Ul 0

                                   ~Ul
                               <Du U 0*-      <l:

ill S"E

                                   ~o
                                    >.n  '"

CiJ: U'l> Om

"-~

RPS BUS A DISTRIBUTION PNL RPS BUS B DISTRIBUTION PNL TP-l: RPS POWER SUPPLIES

OPL 171.029 Revision 12 Page 31 of 63 INSTRUCTOR NOTES

d. A RWM insert block and any select block (not Obj. V.B.9 bypassed by a rod selected and driving Obj. V.D, E. 10 signal) will override EMERG ROD IN function.

E. Relationships to Other Systems Obj. V.B.7 Obj. V.C.1.a/b

1. Control Rod Drive Hydraulic System

a. Stabilizing Valves and HCU Directional Control Valves

b. RMCS operates directional control valves (40A-D) and stabilizing valves in the CRD Hydraulic system.

2. Control Rod Drive Mechanism

3. 120VAC Unit Preferred

a. Select power and RPIS power are fed from Obj. V.B.8.f Unit Preferred. A loss of the 120 VAC Unit Obj. V.B.7.e Preferred power source results in: Obj. V.C.1.a (1) Rod Position Indication is lost. (RPIS)

(2) Loss of power to the CRD select modules (select power) (3) Loss of Unit Non-Preferred (Panel 9-9, Cabinet 5)

b. Refer to AOI-57-4, While RPIS and process AOI-57-4 computer are inoperable, control rod movement may only be accomplished by manual reactor scram.

c. The Unit Preferred loads should automatically transfer to the alternate 120VAC Unit Preferred power source.

4. 120VAC Unit Non-Preferred. A loss of unit Non Obj. V.B.7.f Preferred will result in: Obj. V.B.8.f Obj. V.C.1.b

a. Loss of power to the rod withdrawal block relays (rod block) and PLC Timer

L RCl0-1 WHILE EXECUTING THE FOLLOWING STEPS: 0)----- l. STOPSORON IN.J um.ESS P.EOOIRED BY OTHER PROCE{)URES

2. ErXITRCQ AND

           "!'HE RXiS SU<SCRITIC.A.L EXIT ReiO 1>.1+0 ENTER AOl-10Ci-1, REACTOR SGRAI.{

L RCl0-2 NOTES r1'

    ~~j THE REACTOR \'IIlU_ REMAIN SUI"ICRIilC.A.l 'WITHOUT BORON UNDER.All CONDfflONS WHEN:
              .,  IINY 19 CONTROL RODS ARE AT NOTCH 0.2, V...,1H
                  ,All OTHER CONTRCXL RODS FULL'!' INSERTED OR
              .,  .All OONTRCXLRODS ~OHE.AREINSERTEO TO OR BEYOND POSITION 00 OR

DETERMINED SYREAClOR ENGlNEERING

I Cl, AI. TERNATE LEVel CONTROL.

                      ) -__- - - - I.......

( ..

                                           ---------i......---....

Cl-3, Cl-15, C1-19, C l - 2 7 ! ~.~.~~

                                             #1 AMBIENT TEMPMA Y AFfeCT RPV WATER LVlINCICAT10N
                                                   .4NOTRENO L

C4-1 I C5, i.EVEI.iI'OWER "'-. j

                             +              RCA.-1 MONtTOR ANOCOIfTROL RI'V WATER l Vl L

CONTROl / C5-1 VERIFY AS REQUIRED:

PCIS ISOLATIONS (GROUPS 1,2, AND 3)
ECCS
RCle L RCfl-2 WHILE EXECUTING THE FOLLOWING STEPS:

ElUTRCllNIl} ENTER c.l, RPV fLOOC1NG L RCIL-3

0610 NRC SRO Exam

6. SRO 295038G2.2.22 001lCIA/TlGlITRM 3.7.211295038G2.2.221ISRO ONLYINEW 113112008 Unit 1 is operaing at 59% power with one Recirc Pump out of service. Actual Minimum Critical Power Ratio (MCPR) is determined to be 1.10.

r:~UI~ Which ONE of the following describes the)lppmpriate operator actions and the possible I consequences of this condition?

      ..R.estere-eom~nce          with-the-ReaGter=-Ger-e-8afety Limit..l1L insert all operable control rods within 2 hours.       Possi~onsequences         of this condition are that high offsite releases may result during~~nsient ue to _ _ _ _ _""'{2""'l_ _ _ _ __

(1 ) (2) A'I AND; perforations in the fuel cladding safety barrier. B. OR; perforations in the fuel (1;ladding safety barrier. C. AND; a failure of the Reactor Coolant System safety barrier./>( D. OR; a failure of the Reactor Coolant System safety barrier. ~ J KIA Statement: 295038 High Off-site Release Rate 2.2.22 - Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA Justification: This question satisfies the KIA statement by requiring the candidate to identify design features to reduce off-site release rates which are related to Tech Spec LCOs and their bases.

References:

Reference U1 TSB for Reactor Coolant Safety Limits, specifically 2.1.1.2 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam REFERENCE PROVIDED: None Pl'ausibility Analysis: Sunday, February 17, 2008 3:28:27 PM 15

0610 NRC SRO Exam In order to answer this question correctly, the candidate must: Demonstrate understanding, application, required action and basis for Reactor Core Safety limits as they relate to high offsite release rates and the mechanism of elevated release. Specific knowledge needed: Recall from memory the appropriate Safety Limit (SL) for the conditions of above 785 psig and greater than 10% core flow during SLO which is 1.11 minimum. Determine the Safety limit is NOT met, the appropriate actions to take, and combine that with the basis for the Safety Limit (see below).

       "The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

Exceeding an SL may cause fuel damage and create a potential VIOLA TlONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 3). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours. The 2-hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

2. 1. 1 Reactor Core SLs 2.1.1.2. With the reactor steam dome pressure above 785 psig and core flow above 10% rated core flow:

MCPR shall be greater than or equal to 1.09 for two recirculation loop operation or greater than or equal to 1.11 for single loop operation. With any SL violation, the following actions shall be completed within 2 hours: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods." Safety limit violated, action required and posible consequences to not restoring safety limit. A - correct: B - incorrect: because the restoration of the Safety Limit is only part of the required action, even when compliance with the safety limit may be restored. This condition also requires a reactor shutdown. C - incorrect: correct action but wrong safety limit. The second part of this choice is the basis for SL 2.1.2, NOT the one in the stem. D - incorrect: because the restoration of the Safety Limit is only part of the required action, even when compliance with the safety limit may be restored. This condition also requires a reactor shutdown. The second part of this choice is the basis for SL 2.1.2, Sunday, February 17, 2008 3:28:27 PM 16

SLs 2.0 ( 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

                      < 10% rated core flow:

THERMAL POWER shall be::; 25% RTP. 2.1.1.2 With the reactor steam dome pressure ~ 785 psig and core flow

                      ~ 10% rated core flow:

MCPR shall be ~ 1.09 for two recirculation loop operation or ~ 1.11 for single loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be::; 1325 psig. 2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hours: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods. BFN-UNIT 1 2.0-1 Amendment No. ~267 February 06, 2007

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning -of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties. The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of bOiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis. (continued) BFN-UNIT 1 B 2.0-5 Revision 0

0610 NRC SRO Exam

7. SRO 600000G2.2.24 00l/CIA/SSI/EECWI1600000G2.2.241ISRO ONLYINEW 12/1812007 RMS All Units are in Mode 1.

Corrective Maintenance is in progress for 'C' Fire Pump Control Room handswitch. Delays in parts delivery will result in inability to replace the handswitch for another 130 days. "------. Which ONE of the following describes the required compensatory actions? Establish fire Watches in _ _-->.(....:...1./-)___ . PORC review of compensatory actions will be required (2) REFERENCE

        ~._._. __.. _L.. _ ...

PROVIDED (1) A. Fire Area 16. once. B .. Fire Area 16. twice. C. Fire Areas 21,22,23 and 24. once. D..... Fire Areas 21,22,23 and 24. twice. KIA Statement: 600000 Plant Fire On-site 2.2.24 - Equipment Control Ability to analyze the affect of maintenance activities on LCO status. KIA Justification: This question satisfies the KIA statement by requiring the dmdidate to demonstrate knowledge of the affect of maintenance activities on Appendix R related components including LCO actions and compensatory measures required.

References:

Manual #: Fire Protection Report, Vol. 1 PLANT: BFN UNIT(s):1/2/3 PAGE 462 ,. .- - . of 915 n_ TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 0 Level of Knowledge Justification: This question is rated as CIA due to the requirement to determine the action required as compensatory for loss of APP R SSD equipment. SRO Level Justification: T~~stion satisfies the requirementsof 10 CFR 55.43(b) (4) Radiation hazards that maY/-BnsE\dunng normal and abnormal Situations, Including maintenance activities and various contamination conditions. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 17

0610 NRC SRO Exam Plausibility Analysis: In order to answer this question correctly, the candidate must: Recognize that in accordance with Fire Protection Report Volume 1, Appendix R Safe Shutdown Program, the function that is not available is the control room function Determine which of the given conditions qualifies as an Equivalent Shutdown Capability as defined by the Fire Protection Report Volume 1, Appendix R Safe Shutdown Program. A fire watch in specified fire areas qualifies as an Equivalent Shutdown Capability, and since the the equivalent shutdown capability is the fire watch, then PORC review is required every 60 days. Manual #: Fire Protection Report, Vol. 1 PLANT: BFN UNIT(s):1/2/3 PAGE 462 of 915 TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 0 (EXCERPT) COMPENSATORY MEASURES: A. Restore the equipment function in 7 days or provide equivalent shutdown capability by one of the following methods:

1) A temporary alteration in accordance with plant procedures that allows the equipment to perform its intended function, or

2) A fire watch in accordance with the site impairment program in the affected areas/zones as specified in Section III.

Note: Fire watch requirements in the Turbine Building (FA #25) and Control Building (FA #16) may be evaluated on a case by case basis due to the large size of these areas. For example, fire watches in the Turbine Building can be limited to within 20 feet of the south wall (near M-Line wall on EL 565' and 586') or the Intake Pumping Station due to the location of the RHRSW power cables in the areas. No Safe Shutdown circuits are located in any other location within the Turbine building. Control Building areas, even though not separated by fire resistive barriers, provide substantial protection against the spread of fire due to installed fire suppression systems and concrete floor slabs and walls. The potential of fire spread between control building compartments and the turbine building compartments has been evaluated in Section 3.0 of the IPEEE Fire Induced Vulnerability Evaluations for Units 1-3 (Reference A 16 Section 3). These evaluations may be reviewed to determine the extent of fire watches.

3) If equivalent shutdown capability is used, restore the equipment function in 60 days or provide an engineering evaluation and a change to this program that provides an alternate method to perform the Appendix R function, otherwise provide PORC review and Plant Manager Approval of the equivalent shutdown capability to ensure its adequacy. This review shall be conducted every 60 days until an alternate method is in place. Site Engineering may be contacted for assistance in determining what constitutes equivalent shutdown capability. An example would be the use of the spare SD BD battery charger in lieu of one of Sunday, February 17, 2008 3:28:27 PM 18

0610 NRC SRO Exam 3-CHGA-248-0003EB). D - correct: A - incorrect: Fire area 16 is the incorrect location. The fire watch is required in the area of the PROTECTED train or component, not the AFFECTED train or component. B - incorrect: Fire area 16 is the incorrect location. The fire watch is required in the area of the PROTECTED train or component, not the AFFECTED train or component. The second part of the answer is correct to ensure the adequacy of an Equivalent Shutdown Capability if the component will not be restored for 130 days. C - incorrect: The fire areas are correct, however the second part of the answer is incorrect based on the duration required for an Equivalent Shutdown Capability of 130 days. Two PORC reviews will be required. Sunday, February 17, 2008 3:28:27 PM 19

Manual #: Fire Protection Report PLANT: BFN UNIT (s) : 1/2/3 PAGE 462 of915 Vol. 1 TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 0 The listed compensatory measure in the Unit 1, 2 & 3 tables due to equipment degredation or the compensatory measures due to lack of spatial separation per 9.3.11.G.l.a of the Fire Protection Plan may be removed/revised if:

the affected unit is brought to COLD SHUTDOWN, or
an engineering analysis is performed, this program is changed and the Safe Shutdown Instructions are changed to provide an alternative shutdown path or
a different compensatory measure or combination of measures is established (e.g., additional administrative controls, operator briefings, temporary procedures, interim strategies, operator manual actions, temporary fire barriers, temporary detection or suppression systems). An engineering analysis of the alternative measure should incorporate risk insights regarding the location, quantity and type of combustible material in the fire area; the presence of ignition sources and their likelihood of occurrence; the automatic fire suppression and fire detection capability in the fire area; the manual suppression capability in the fire area; and the human error probability where applicable. (Reference 5)

Compensatory Measure A will be documented and tracked in accordance with Attachment A of this instruction. COMPENSATORY MEASURES A. Restore the equipment function in 7 days or provide equivalent shutdown ( capability by one of the following methods.

1) A temporary alteration in accordance with plant procedures that allows the equipment to perform its intended function, or

2) A fire watch in accordance with the site impairment program in the affected areas/zones as specified in Section III.

Note: Fire watch requirements in the Turbine Building (FA #25) and Control Building (FA #16) may be evaluated on a case by case basis due to the large size of these areas. For example, fire watches in the Turbine Building can be limited to within 20 feet of the south wall (near M-Line wall on EL 565' and 586') or the Intake Pumping Station due to the location of the RHRSW power cables in the areas. No Safe Shutdown circuits are located in any other location within the Turbine building. Control Building areas, even though not separated by fire resistive barriers, provide substantial protection against the spread of fire due to installed fire suppression systems and concrete floor slabs and walls. The potential of fire spread between control building compartments and the turbine building compartments has been evaluated in Section 3.0 of the IPEEE Fire Induced Vulnerability Evaluations for Units 1-3 (Reference A16 Section 3). These evaluations may be reviewed to determine the extent of fire watches.

Manual #: Fire Protection Report PLANT: BFN UNIT(s):1/2/3 PAGE 463 of915 Vol. 1 TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 0

3) A temporary change to the SSI's which provide safe shutdown without the required function.

If equivalent shutdown capability is used, restore the equipment function in 60 days or provide an engineering evaluation and a change to this program that provides an alternate method to perform the Appendix R function, otherwise provide PORC review and Plant Manager Approval of the equivalent shutdown capability to ensure its adequacy. This review shall be conducte~yery 60 d~until an alternate m~_ is in place. Site Engineering may be~ for assistance in determining what constitutes equivalent shutdown capability. An example would be the use of the spare SHDN BD battery charger in lieu of one of the permanent SHDN BD chargers (i.e., 0-CHGA-248-0000A, B, C, D or 3-CHGA-248-0003EB).

(REFERENCES PROVIDED TO c CANDIDATE

                                                                                                                                     "=-',"

Manual #: Fire Protection Report PLANT: BFN UNIT(s): 1/2/3 PAGE 472 of915 Vol. 1 TITLE: Appendix R Safe Shutdown P!ogram__ SECTION: 4 REV: 0 SECTION III - REQUIRED SAFE SHUTDOWN EQUIPMENT - UNIT 0 EQUIPMENT DESCRIPTION UNIT(S) APPENDIXR COMPENSATORY AREA I ZONES AFFECTED FUNCTION MEASURES SYSTEM 026 - HIGH PRESSURE FIRE PROTECTION O-PMP-026-000 1 FIRE PUMP A 0 VERIFY START FROM A 3-3 MCR 0 RESET START A 3-3 LOCALL Y FOLLOWING LOOP, THEN RESTART FROM BOARD OR MCR 0-PMP-026-0003 FIRE PUMP C 0 ISOLATE AND START A 16 FROM 4KV SHDN BD C 0 VERIFY START FROM A 21,22,23,24 MCR 0 RESET START A 21,22,23,24 LOCALL Y FOLLOWING LOOP, THEN RESTART FROM BOARD OR MCR O-PMP-026-0118 DSL FIRE PUMP 0 START FROM MCR A 1-1, 1-2, 1-3, 1-4, 1-5, 1-6,2-1,2-2, 2-3,2-4,2-5,2-6,3-1,3-2,3-4,4,5, 6,7,8,9, 10, 11, 12, 13, 14, 15, 17, 18, 19,20,25-1, 25-II

0610 NRC SRO Exam

8. SRO 295002G2.4.6 001lCIAlEOI/INTRO/51295002G2.4.61ISRO ONLYINEW 11129107 RMS Unit 3 is operating at full power when a condensate leak results in the following plant conditions:

The reactor scrams and condensate pumps are removed from service.
Water level is restored and maintained with HPCI and RCIC.
A cooldown is initiated using Main Turbine Bypass Valves.

The unit operator reports a(fO;~~i~g?Yondenser vacuum.

                                           -~---~"~'-A--"' o_~ .-"

Which one of the following describes the reason fb-rJQwe!iog,condenser vacuum and the aGtions FOE}l::H-r:eEl-te-l::J-#l+ze-the ma in'-e0flEl*HRseuQ!-tlle.-cocldowr.l4~~* f' Y-I7-u"d..(.Nt£ ~..fo IN-(l'v'vtR..-1kR. (.£7)re~~pVl Condenser vacuum is lowering due to a loss of _____(..;.1.... ) _ __ Direct the ~j7dLv',,\ -k {If'-'vh' (2) A. hotwell level. r,,-..~.>!-~,:,;-j~{"'i; B B. the Steam Jet Air Ejector. restoration oftheSteamJ,etAiLEjedor. ;7 J: s "~rv' 10 ()- / z.A C. hotwell level. ~lTffient*afld~st-aFt**6Hhe*G0fldenser*"V,aUl;lm~l2.um ps. (;.D1: _ t;.,- /()O-/2.A? D." the Steam Jet Air Ejector. alignrnentand-staFt-of*the-.(;oRGleRser*V*aclJum.. gumps; Sunday, February 17, 2008 3:28:27 PM 20

0610 NRC SRO Exam KIA Statement: ( 295002 Loss of Main Condenser Vac I 3 2.4.6 - Emergency Procedures I Plan Knowledge symptom based EOP mitigation strategies. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate mitigation strategy to restore the main condenser to service while executing Emergency Operating Instructions.

References:

EOIPM, 1-EOI-1 Flowchart EOIPM SECTION O-V-C REV 1 page 40 basis for step RC/P 10: If the Main Condenser can be made availalble THEN stabilize pressure below below 1073 psig with the main turbine bypass valves (appendi>;. 88) Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55,43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 21

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: The leak in the condensate system which necessitated removing the condensate pumps from service resulted in an isolation of the operating SJAE from low condensate flow. This will result in lowering vacuum. Although the leak in the condensate system will cause lower hotwell level, the makeup from the CST would maintain sufficient level to preclude losing vacuum. To continue a cooldown to the condenser, vacuum must be restored. Since use of SJAEs is prevented with no condensate flow, the only remaining alternative is to use the Condenser Vacuum Pumps. D is correct: A is incorrect: Hotwell level will not result in lowering vacuum until makeup from the CST is no longer available. In addition, restoration of the Steam Jet Air Ejector is not possible without condensate flow. B is incorrect: The cause of lowering vacuum is correct. However, restoration of the Steam Jet Air Ejector is not possible without condensate flow. C is incorrect: Hotwell level will not result in lowering vacuum until makeup from the CST is no longer available. However, starting the COndenser Vacuum Pump will restore condenser vacuum. Sunday, February 17, 2008 3:28:27 PM 22

OPL 171.030 Revision 16 Page 14 of 71 INSTRUCTOR NOTES

d. Normal flow rate of dry gases at air Hydrogen Water ejector at 130°F and 1 atmosphere Chemistry (HWC) pressure with Normal Water Chemistry changes the rate of (NWC) water decomposition (1) Between 200 and 300 SCFM (2) Off-gas composition The following values are by DESIGN.

Actual flowrates are much lower now. 154 SCFM H2 (from reactor Off-gas flow water decomposition) instruments measure following the 77 SCFM 02 (from reactor recombiner and only measure the water decomposition) remaining air inleakage 18.5 SCFM air inleakage 46 SCFM water vapor 295.5 SCFM TOTAL

5. Interlocks and Valve Controls TP-4

a. Start signal permissive logic Obj. V.B.3/ V.C.1.

HS-90-155 in AUTO after RESET. Obj. V.D.3.a All other start permissive signals have been removed.

b. Additional logic required to place individual SJAE in service manually (1) Inlet and outlet condensate valves for the intercondenser must be open.

(2) Condensate pressure from the Obj V.B.3N.C.1 intercondenser must be ~ 60 psig. Steam isolation valves will not open (and will close if already open) if condensate pressure is less than 60 psig.

OPL 171.030 Revision 16 Page 35 of 71 INSTRUCTOR NOTES (4) Normal power 480 DIG Aux Bd A Alternate power 480 control bay vent Bd A the fan will auto swap to alternate power if the normal supply is lost (5) Standby fan auto starts on low flow When both flow provided its handswitch is in auto and switches are picked standby fan is selected. Fans cannot up, an alarm will be be stopped from control room while a received in the low flow condition exists. Each fan control room has a low flow switch. (6) If standby fan is not selected it will not auto start (7) Fan is prevented from auto starting due to 480 load shed logic. Fan can be manual started

23. Condenser vacuum pump Obj. V.B.9N.C.7

a. Location: Just above SJAE Room in Turbine Obj. V.D.2.c Building near precooler

b. Flow (rated) 3000 SCFM at 15" Hg 1850 SCFM AT 7"HG ABS

c. Removes air from condenser shell and establishes sufficient vacuum for condenser operation

d. Requires sealing water for pump to operate

e. Discharges to stack via the 1.75 minute holdup volume

f. Seal water pump gets a start signal from mech vacuum pump C.S.

g. Procedurally, the vacuum pumps should not be used at greater than 5% power.

h. Condenser vacuum pump cannot be started Obj. V.D.3.b unless:

( (1 ) Seal water pump is running, and

OPL 171.030 Revision 16 Page 36 of 71 INSTRUCTOR NOTES (2) Condenser vacuum pump suction is < 26" Hg.

i. Condenser vacuum pump will auto trip if:

(1 ) Condenser vacuum is pump suction ~ 26" Hg, or Press switch is located between pump and (2) Main Steam line radiation is ~ 3x suction valve. Pump NFLB, or will trip on low vacuum if suction valve is shut (3) Seal water pump trips, or (4) Condenser vacuum is ~ 22" Hg and reactor pressure is ~ 600 psig, or (5) Undervoltage is present

j. Vacuum pump suction valve will auto close if:

(1) Main Steam line radiation is ~ 3x NFLB, or (2) Condenser vacuum is ~ 22" Hg and reactor pressure ~ 600 psig.

24. Steam packing exhauster and condenser

a. Location: Turbine Building

b. Condenses turbine seal steam and removes non-condensibles

c. Cooling water supplied by Condensate System

(

0610 NRC SRO Exam

9. SRO 295009G2.4.31 001lCINTlG2/EOII1295009G2.4.3111SRO ONLYINEW 113112008 Unit 2 is operating at full power with RCIC out of service for schedule maintenance. All other systems are normal. The following conditions are noted:

Loss of Offsite Power
The UO notes the following conditions

                      -  All control rods inserted
                      -  Drywell Pressure 2.6 psig
                      -  Reactor Water Level -4Q inches and slowly lowering
                      -  Reactor pressure is 425' psig steady
                      -  Battery Bd 2. VOLTS, 2-EI-57-37 indicates 255, volts
 ."

All systems respond as expected except for the following annunciators are in alarm:
BAT BD 2 BKR TRIP OUT / FUSE BLOWN OR GROUND 2-EA-57-117
250V REACTOR MOV BD 4f\' UV 2-EA-57-94 Which one of the following describes the status of HPCI system and any required actions?

HPCI is currently __________(I.-:.1.J-)_ _ _ _ _ __ The proper course of action would be to _ _ _ _ _ _ _ _....l(-=:2'-/...)_ _ _ _ _ _ _ _ __ A. .(1) available for pressure control but NOT level control due to high drywell pressure. (2) vent the drywell to allow HPCI to be used for level control. B. (1) available for level control but NOT pressure control due to high drywell pressure. (2) monitor HPCI system operating parameters while injecting to the vessel. G.!' (1) NOT available for injection OR pressure control due to a loss of the power supply. (2) transfer '2A' 250VDC RMOV board to the alternate power supply. I. D. (1) NOT available for injection OR pressure control due to a loss of the power supply. (2) bypass the load shed signal to the battery charger. Sunday, February 17, 2008 3:28:27 PM 23

0610 NRC SRO Exam KIA Statement: 295009 Low Reactor Water Level I 2 2:4.31 - Emergency Procedures I Plan Knowledge of annunciators alarms and

      .indications, and use of the response instructions.

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions to control reactor water level while executing Emergency Operating Instructions.

References:

2-ARP-9-8C windows 4 and 7. Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome and determine remedial actionss. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 24

0610 NRC SRO Exam REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following:

1. HPCI is unavailable due to undervoltage on the 2A 250 RMOV board.

2. Normal power is available from Battery Board 2, although a load shed has removed the battery charger from service.

3. HPCI operation is not possible in AUTO or MANUAL, Injection or Pressure control.

(ARP)

4. Lowering OW pressure will allow reset of the load shed signal, however this is not the cause of the HPCI failure. This will restore the battery charger to supply Battery Board 2, but the 2A 250V RMOV Board is still de-energized.

C - correct: A - incorrect: HPCI is unavailable until 2A 250V RMOV board is restored. However, even if HPCI was available, OW pressure does not have to be lowered to use HPCI for level control. This is only required to use HPCI for pressure control. B - incorrect: HPCI is unavailable until 2A 250V RMOV board is restored. D - incorrect: HPCI is unavailable due to undervoltage on the 2A 250 RMOV board. However, resetting the load shed signal will not correct the cause of the HPCI failure. This will restore the battery charger to supply Battery Board 2, but the 2A 250V RMOV Board is still de-energized. Sunday, February 17, 2008 3:28:27 PM 25

BFN Panel 9-8 2-ARP-9-8C Unit 2 2-XA-55-8C Rev. 0013 p~g~!fJf 46 Sensor/Trip Point: 250V REACTOR 72N-BA Normal supply overcurrent. MOVBD2A 72E-BA Alternate supply overcurrent. UV 27EX Normal supply undervoltage. 2-EA-57-94 27B MOV bd undervoltage (7sec TDDO) (Page 1 of 1) Sensor 250V RMOV Bd 2A, E1621', R-14 Q-LlNE, Shutdown Bd Rm C Location: Probable A. Loss of normal supply (250V Battery Bd 2, Pnl 3, Bkr 302). Cause: B. Transfer of power supply to alternate breaker (manually only). C. Overcurrent on normal or alternate supply to the board. D. Fuse failure. E. Sensor malfunction. Automatic None. Action: Operator A. VERIFY conditions of alarm: Action:

Loss of HPCI indicating lights on Panel 2-9-3. o
Loss of backup scram valve lights on Panel 2-9-5. o B. DISPATCH Personnel to MOV board to check for abnormal conditions: undervoltage, breaker tripped, etc. o NOTE

[1I/C] If the mechanism resets (hear click and feel resistance when pushing in), this indicates that an overcurrent condition tripped the breaker. C. IF Normal or Alternate feeder breaker tripped, THEN Manually DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device. [NERIC 11-8-92-069] 0 D. VERIFY Bkr 302 closed at Battery Bd Room 2, Panel 3, E1593'. 0 E. REFER TO TS Section 3.B.7. 0 F. REFER TO 0-01-57D to re-energize or transfer the board. 0 G. REFER TO appropriate 01 for recovery or realignment of equipment. 0

References:

45N620-11 2-45E712-1 45N714-7 TS Section 3.B.7.

BFN Panel 9-8 2-ARP-9-8C Unit2 2-XA-55-8C Rev. 0013 PCige 110f 46 SensorlTrip Point: BAT BD 2 Breaker tripout. BKR TRIPOUT/FUSE Ground on Battery BLOWN OR GROUND Bd 2. 2-EA-S7-117 Cleared Fuse BLUE BAR .~ (Page 1 of 1) Sensor Battery Bd Rm 2, EI 593' Location: Probable A. Breaker trip on overload or fault (thermal and magnetic trip) on Panels 1-14. Cause: B. 1100-series breakers (Heineman) on Panel 11 annunicate in OFF or TRIP. C. A ground fault exists on 250V DC Battery Bd 2. D. Fuse failure to instrumentation circuit, Pnl 1. E. Fuse 0-FU2-039-0007 cleared. F. Blown light bulb in ground detector indicator photo cell. Automatic A. None Action: Operator A. CHECK Battery Bd 2 VOLTS, 2-EI-57-37, and AMPS, 2-EI-57-3B, Action: on Panel 2-9-B to determine if load is on battery or charger. 0 B. CHECK Battery Bd 2 for abnormal conditions: 0

Loss of volt and amp indication. 0
Breaker position. 0
Ground indication. 0
Fuse failure. 0
Check ground detector indicator photo cell light bulb (requires electrical maintenance). 0 C. NOTIFY Unit 3 Unit Supervisor. 0 D. CHECK condition of equipment fed from tripped breaker and attempt reclosure as directed by Unit Supervisor. 0 E. REFER TO TS Section 3.B.7. 0 F. REFER TO 0-GOI-300-2 for grounds and location of D-C grounds. 0 G. If alarm is invalid, THEN REFER TO 0-01-55. 0 H. While alarm is sealed in, MONITOR other breakers on the Battery Board until alarm is reset. 0

References:

2-45E620-11 0-45E702-1, -3 0-45E702-2 TS Section 3.B.7

OPL171.037 Revision 10 Page 49 of 70 i ~

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a !<< i i g TP-4 ECCS Division I & II DC Separation Logic

OPL171.037 Revision 10 Page 50 of 70 ( Unit 2, Division I Unit 2, Division II 81 I 82 I I I I I 83 I I I I I I I I 250'1-00 RMOV fID:2B : 25ov.DC RMOVBO 2C 25O'I...IJC RMOV BD 2A CORE SPRAY "CBE SfBA:t: r- OIVI LOGIC ~ OIVil LOGIC PANEL 9-32 PANEL 9-33 HPCI DIVI LOGIG .t:IW PANEL9-'J2 DIVIILOGIC

>--

REOlJlNOANT ~ PANEL 9--39. SHUTOFF TO VAlJIES, TURBINE CST VALVE AUXILIARIES AOO OIVI RELIEF

                                         .ADS                    8DI DIVI RELIEF r-- DIVII    RELIEF
                                     ~
 ~

VALVES, LOGIC VALVES VALVES. LOGIC PANEL 9-30 PANEL 9-33 RHR RHR

 ~     i5iViLOGIC                                            ~

DIV II LOGIC PANEL 9-32 PANELs..33 PCISDIV I PCISOIVII r- INBOARD ISLN OUTBOARD VALVES. PANEL ~ ISOLATION 9-42& 25-32 VALVES. PANEL 943&25-32 EOCS

 >--    ATU INVERTERS                                                  ~
                                                             >--   ATU INVERTERS RCIC OIVI LOGIC 0-   PANEL 25-31.                                                 RCIG OUTBOARD                                               ~

OIVIILOGIG ISOLATION PANEL 9..33. VALVE OTHER OTHER I--- B.Aa<.UP SCRAM VALVES PANEL 9-17

                                                             -    BACKUP SCRAM VALVES PANEL 9w.15 TP-5: Unit 2 Distribution Logic (U-1 and U-3 Similar)

0610 NRC SRO Exam

10. SRO SOOOOOEA2.02 00 lICIA/T! G2/CONT/PRII2IS00000EA2.02/3.S/SRO ONLY/NEW 113112008 Unit 2 has experienced a LOCA with the following plant conditions: " -J::-.s

RPV level dropped below Top of Active Fuel (TAF) .~~a[r~co~~;:B~~;~~1f j~~O~

Severe Accident Management Guidelines (SAMG) entry was required. ) '::J

Reactor water level has been raised to (+)32 inches.
Venting per EOI-2 has reduced Drywell pressure to 3 psig. /

           ...                  *f              L~j'if;';.~   - r [fo "

Which ONE of the following describes how monitoring of the containment H2 and 02 concentrations is accomplished and what methods of control are used during this condition? Monitoring is accomplished by _ _ _ _ _ _ _->.(. . :. 1~)_ _ _ _ _ _ _ . Control is accomplished by use of the ______("""2'-'-.)_ _ _ _ __ A." (1) the use of the Containment H2/02 Monitors. Nitrogen Makeup System. B. (1) the use of the Containment H2/02 monitors. Containment Atmosphere Dilution (CAD) system. C. (1) Chemistry Department sampling due to PCIS isolation of the H2/02 monitors. (2) Nitrogen Makeup System. D. (1) Chemistry Department sampling due to PCIS isolation of the H2/02 monitors. Containment Atmosphere Dilution (CAD) system. I

Sunday, February 17, 2008 3:28:27 PM 26

0610 NRC SRO Exam KIA Statement: 500000 High CTMT Hydrogen Conc. 15 EA2.02 - Ability to determine and lor interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Oxygen monitoring system availability. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the operability of the Hydrogen-Oxygen Monitoring system.

References:

EOI-2, "Primary Containment Control". Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5)* Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 20083:28:27 PM 27

0610 NRC SRO Exam REFERENCE PROVIDED: None piausibility Analysis: In order to answer this question correctly, the candidate must determine the following: Without references, the candidate must recognize that EOI-2, "Primary Containment Control" actions are in progress based on current plant conditions. Although the H2/02 monitors isolated due to a Group 6 PCIS isolation signal, provisions are made on EOI-2 to bypass that isolation and restore the monitors to service. In addition, recent modifications have procedurally removed the CAD system from available containment atmosphere control options unless Severe Accident Management Guidelines are in progress. This leaves only the normal Nitrogen Makeup System available for atmosphere control since SAMG entry was avoided by level recovery. A - correct: B - incorrect: Although the Containment H2/02 Monitors are available, the atmosphere control is NOT from the CAD System. C - incorrect: Grab samples taken by Chemistry are only necessary if the Containment H2/02 Monitors are unavailable. However, the Nitrogen Makeup System is used to control primary containment atmosphere. D - incorrect: Grab samples taken by Chemistry are only necessary if the Containment H2/02 Monitors are unavailable. In addition, the atmosphere control is NOT from the CAD System. Sunday, February 17, 2008 3:28:27 PM 28

WHILE EXECUTING THE FOLLOWING STEPS: iF THEN RfilSET ""NAL YZER iSOtATJON.

       *H2{Q2 ",';N .*"l..YZERS .ARE IN STANDS';!"     IF NECESSARY Ao\lD
                                                   ~ . puce "'NAt YZtR lSOLAllON £YPAS;S K"EYLo.OK SWf!'CH£S TO. BYf>ASS 2.SIl:'~eT     [JIN OR SUf-iPRCHMSR ANO MOME.NT~""RIL*"'* PULL OUT SELECT SWlTCH HANDLE. TO sr;O\Rr SAMPLE PuMPS ti2 .A.ND 0.2 MoNrrORlNO SYSTEM IS INOPERASLE L

( P'C/H-1 L

OPL 171.032 Revision 12 Page 10 of 48 IX. Introduction The Containment Inerting and Purge Systems provide for inerting and make-up of nitrogen to the primary containment during normal operations, and purging primary containment when access is desired. The Containment Atmosphere Dilution System is used for the dilution of the primary containment atmosphere following a LOCA, to ensure oxygen concentrations remain less than 5% and hydrogen concentrations remain less than 4%. Review BFN plant work expectations as they apply to this lesson plan.

OPL 171.032 Revision 12 Page 22 of 48 INSTRUCTOR NOTES B. Containment Atmosphere Dilution (CAD) System TP-3

1. In normal operation, the primary containment atmosphere is maintained at less than 3.5% oxygen levels by the Nitrogen Storage and Supply System.

2. Following a loss-of-coolant accident, hydrogen is Obj. V.E.9 produced in the containment from zirc-water reactions, and H2 and 02 are produced from radiolytic decomposition of water. If uncontrolled the concentration of oxygen and hydrogen could get to a combustible gas mixture.

3. The CAD System is used in the post LOCA condition to prevent the containment atmosphere from reaching a combustible mixture, by diluting the atmosphere with nitrogen and venting as necessary.

The CAD System also serves as a backup Obj. V.B.7 pneumatic supply to the Drywell Control Air System.

4. System Design Basis

a. CAD is shared system, and can supply TP-3

( nitrogen to any of the three units.

b. CAD can supply nitrogen at rate sufficient to maintain the oxygen concentrations of both the drywell and torus below 5% by volume, (based on calculated generation rates) and the hydrogen concentrations to less than 4%

by volume.

c. The containment atmospheric dilution portion TP-3 of CAD system is an engineered safety feature. The system is remote-manually operated from the control room and must be able to start within ten hours following a LOCA.

5. Basic Flow Path

a. Nitrogen is supplied from one of the storage Obj. V.B.8 tanks through its associated ambient Obj. V.C.4 vaporizer, and electric heaters.

b. Flow is then directed to the drywell and/or Obj. V.E.10 torus (via flow unit specific control valves). Obj. VD.3.a

0610 NRC SRO Exam

11. SRO 203000A2.16 001lCIA/T2G1IEOI C-lI/203000A2.161ISRO ONLYINEW 113112008 Unit 2 is operating at 100% power with the following conditions:

A leaking Safety Relief Valve (SRV) is causing heating of the suppression pool at about 1 of every 4 hours.
Two Residual Heat Removal (RHR) Pumps and associated RHR Service Water (RHRSW) Pumps are capable of maintaining temperature below 95 of with 8 hours of operation per day.

Which ONE of the following describes the appropriate actions and the reason for that determination? Use _ _ _...... (1'-")_ _ _ to minimize the potential for loss of injection function during a LOCA concurrent with Loss of Offsite Power due to _ _ _ _--'(=2J-)_ _ _ __

A"! two RHR Pumps in one loop; water hammer. B. ALL available RHR pumps; water hammer. C. two RHR Pumps in one loop; the time required for valve realignment. D. ALL available RHR pumps; the time required for valve realignment. (

                                                               ?/"                 (I ol~        - 70.                        I)()

Sunday, February 17, 2008 3:28:27 PM 29

0610 NRC SRO Exam KIA Statement: 203000 RHR/LPCI: Injection Mode 1\2.16 - Ability to (a) predict the impacts of the following on the RHRlLPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accident. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions and their bases to control RH R in LPCI mode following a Loss of Coolant Accident.

References:

3-EOI-1 Flowchart, 3-EOI-C1 Flowchart, EOIPM Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during 110rmal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 30

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly, the candidate must determine the following:

                         ><<"""",**,."",,,'H"'__ '''''_''''':'~"~-':'____ "'~

p~r~t\fRf"I~furmation Notice 87:JD?[NRC/C] concerns exist, with RHR in Suppression FtoQCCOOTIngrvfoae:lfiann:nCA and a LOSP could result in severe water hammer following the RHR pump restart when DIG power becomes available. With the SP Cooling valves re-aligned, the line may drain down while the DIGs start and re-energize the RHR pumps. To minimize the potential of this condition occurring, the following guidelines should be used to try and maintain the system below the PSA Risk Assessment goals:

1. RHR in suppression pool cooling should be minimized.

2. Two Loops of RHR in suppression pool cooling should be minimized.

3. Use two pumps per loop, if needed, to maximize suppression pool cooling in order to minimize total time spent in suppression pool cooling.

A - correct: B - incorrect: Although the basis is correct, there is no requirement to operate more than one loop of RHR in SP Cooling. C - incorrect: Although the RHR lineup is correct, the basis for the lineup is incorrect. The RHR system has a blue light on Panel 9-3 that will de-energize if the SP Cooling valve is opened too far such that LPCI injection is delayed pending valve re-alignment. Although this limitation is a concern, it is NOT based on the potential for a LOSS of injection, but a DELAY in injection. D - incorrect: There is no requirement to operate more than one loop of RHR in SP Cooling. In addition, the RHR system has a blue light on Panel 9-3 that will de-energize if the SP Cooling valve is opened too far such that LPCI injection is delayed pending valve re-alignment. Although this limitation is a concern, it is NOT based on the potential for a LOSS of injection, but a DELAY in injection. Sunday, February 17, 2008 3:28:27 PM 31

CAUTION

     #2       FUMP l\IPSH AND VORTEX LIMITS L

L SP/T-1 ( NO L Of'E.U'm..Ii.tl A"V.II.iLABLE SUPPR Pt COOLING USINO ONiY RHR PUMPS NOT REQUIRED TO ASiSUREADEOOJ....rE CORE Cor'JlING 8'",' 00l'mNUOUS INJ (APPX ~7Al L

OPL 171.044 Revision 15 Page 59 of 159 INSTRUCTOR NOTES

e. Status indications and Controls (Panel 9-3 unless indicated)

(1) RHR SYS I AUTO-INIT LOCKOUT - Amber lights for A & C RHR SYS II AUTO-INIT LOCKOUT - Amber lights for B & D Indicates that RHR pump has been stopped with initiation signal present and is locked out for automatic re-start (2) Unit 1 RHR SYS I INITIATION - Amber light 1-IL-74-10AK74A (qty 2) 1-IL-74-10AK74B (a) Unit 2 ONLY due to shared power supplies. (b) Alerted operator that availability of RHRlCS could be affected if accident signal received. (3) FCV-74-59 LOCA CLOSURE TIME - Blue light Normally on Indicates that valve too far open and may not be closed in time to prevent part of injection flow to RPV being diverted from LPCI. (4) RHRSW PUMP INITIATE LOCKOUT - Amber lights (A1/A3/C1/C3/B1/B3/D1/D3) Indicates that RHRSW pump has been stopped with automatic initiation signal present and will not automatically re-start. (5) CNMT SPRAY/CLG VLV INTLK LOGIC A- IL-74-125 Amber light CNMT SPRAY/CLG VLV INTLK LOGIC B- IL-74-131 Amber light TP-24/25 CNMT SPRAY/CLG VLV INTLK LOGIC A XS-74-125 Reset Pushbutton CNMT SPRAY/CLG VLV INTLK LOGIC B XS-74-131 Reset Pushbutton Note: When light is "on", it indicates that an LPCI has been generated and that a closure signal sent to the Containment Cooling valves (System 1: 74- TP-34/35/36/371 57/58/59/60/61; System 2: 74-71/72/73n4n5). 38/39/40/41/42/43

SSINS No.: 6835 IN 87-10 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 February 11, 1987 Information Notice No. 87-10: POTENTIAL FOR WATER HAMMER DURING RESTART OF RESIDUAL HEAT REMOVAL PUMPS Addressees: All boiling water reactor (BWR) facilities holding an operating license or a construction permit. Purpose: This information notice is to alert addressees of the potential for water hammer in the residual heat removal (RHR) system of BWRs during a design basis loss of coolant accident (LOCA) coincident with a loss of offsite power (LOOP) if the RHR system is aligned to suppression pool cooling. Recipients are expected to review the information for applicability to their facilities and consider actions, if appropriate, to preclude similar problems occurring at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required. Description of Circumstances: On December 11, 1986, the Susquehanna nuclear power plant reported that based on results of an ongoing review of the potential effects of water hammer events, the RHR system could be susceptible to water hammer loads that would exceed the allowable stresses in the RHR system and piping. The specific condition of concern involves a design-basis LOCA coincident with a LOOP, while one or one RHR loops are in the suppression pool cooling mode. During the power loss and subsequent valve realignment, portions of the RHR system will void because of the drain down to the suppression pool as a result of elevation differences. A water hammer may occur in those RHR loops that were in the suppression pool cooling mode when the RHR pumps restart after the diesel generators reenergize the buses. The core spray system also may be subject to such a water hammer if it is lined up in the suppression pool mixing mode full flow test. The Susquehanna design basis for LOCA/LOOP assumes that the suppression pool cooling flow path valves are initially closed in the standby lineup. The potential duration factor used in the consideration of the coincident LOCA/LOOP with the RHR in suppression pool cooling mode was one percent, or roughly 90 hours per year. 8702100126 IN 87-10 February 11, 1987 Page 2 of 2 Contrary to the design basis assumption, a licensee review of operating history found that the worst case RHR system usage factor approached 25% during cycles in which significant safety relief valve weeping was experienced.

For interim corrective action, the licensee has modified operating procedures to allow, at a time, only one loop of RHR to operate in suppression pool cooling. In addition, the licensee will revise plant procedures to address the restart of an RHR pump if it trips while operating in the suppression pool cooling mode. The core spray system is currently prohibited from being operated in the suppression pool mixing mode, except for required surveillance testing. Discussion: The NRC discussed the potential for this general type of event in Engineering Evaluation No. AEODjE309, "The Potential for Water Hammer During the Restart of RHR Pumps at BWR Nuclear Power Plants," dated April 1983. In the type of scenario discussed in AEODjE309, the line most likely to drain and experience a water hammer is the drywell spray line because it has the largest elevation difference between it and the suppression pool. RHR system pipes less than 33 feet above the suppression pool will not usually drain because atmospheric pressure will support a column of water that high. A water hammer in the drywell spray line could endanger RHR system integrity, and thus jeopardize all modes of RHR including low-pressure coolant injection. The analysis performed by the licensee of the Susquehanna nuclear power plant goes beyond AEODjE309 in that detailed site-specific computer modeling was performed which shows that piping system integrity could be challenged. Besides Susquehanna, other plants may have high usage factors for suppression pool cooling mode and large elevations differences in the RHR system, making those plants potentially subject to water hammer in the RHR system. No specific action or written response is required by this information ( notice. If you have questions about this matter, please contact the Regional Administrator of the appropriate NRC regional office or this office. Edward L. Jordan Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contact: Eric Weiss, IE (301) 492-9005 George Lanik, IE (301) 492-9007

Attachment:

List of Recently Issued IE Information Notices

0610 NRC SRO Exam

12. SRO 215003G2.1.14 001/C/A/GOIlOO-1A/IRMIB10/215003G2.1.141ISRO ONLYIBANK 12/1/07 RMS Given the following plant conditions:

Unit 3 is performing a startup and heatup in accordance with 3-GOI-100-1A, "UNIT STARTUP."
All IRMs are OPERABLE on Range 7 when a half-scram occurs on RPS 'A'.

                       ~ct:t+3B     ' i               .

The Reactivity Manager reports the cause of the half-scram was due to a momentary Upscale Trip on IR~ 'G;' but,the IRM is currently reading normally.
No IRM Range Switches were being manipulated at the time the half-scram was received.

Which ONE of the following describes the required actions AFTER stopping Control Rod withdrawal? Fir§t, place IRM 'G' in BYPASS (1) Next, reset the half-scram and (2) And then, enter _ _ _ _~(3.::::...)'--_ _ _ __

                 /'\(

A. ((1) and notify the System Engineer. I (2 monitor IRM 'G' for 15 minutes. l)) LCO 3.3.1.1.A.1 on IRM 'G'. I B..... (1) and notify the System Engineer. monitor IRM 'G' fo~ 15 minutes.

          ..     (3) an INFORMATION ONLY LCO on IRM 'G'.
         ,

c. ) AFTER obtaining Plant Manager's permission.

(2) continue with the heatup. (3) LCO 3.3.1.1.A.1 on IRM 'G'. D. (1) AFTER obtaining Plant Manager's permission. ! yvtlT f f/",/f/YTU / continue with the heatup. an INFORMATION ONLY LCD on IRM 'G'. Sunday, February 17, 2008 3:28:27 PM 32

0610 NRC SRO Exam KIA Statement: 2150031RM 2.1.14 - Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions and notifications due to an IRM failure.

References:

3-01-92A, 3-ARP-9-5A (33), OPDP-8, TS 3.3.1.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 33

0610 NRC SRO Exam REFERENCE PROVIDED: U3 Tech Spec Section 3.3.1.1 Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following:

1. 3-ARP-9-5A (33) directs that IRM 'G' be placed in BYPASS in accordance with 3-01-92A.

2. Plant Manager's permission is NOT required since the bypass is directed by an approved p-rocedure~----

3. 3-ARP-9-5A (33) directs that the scram/half-scram be reset.

4. 3-01-92A directs that the IRM be monitored for 15 minutes following a spike which results in a scram trip.

5. System Engineering must be notified to obtain concurrence prior to returning the IRM to service.

6. Tech Spec 3.3.1.1 requires 3 OPERABLE channels per trip system. IRM 'G' is not required to be OPERABLE since all other IRMs assigned to trip system "A" are OPERABLE.

7. OPDP-8 directs an INFORMATION ONLY LCO be entered in the LCO Tracking Log.

B - correct: IS.. - incorrect: LCO 3.3.1.1 .A 1 is NOT appropriate. This is plausible since the first and ( second part of the answer are correct. C - incorrect: Plant Manager's permission is NOT required to bypass IRM 'G' and LCO 3.3.1.1.A 1 is NOT appropriate. In addition, continuing with the heatup is not appropriate until the IRM is monitored for 15 minutes. This is plausible based on the subtle difference between an IRM that is "noisy" as opposed to an IRM that "spikes". D - incorrect: Plant Manager's permission is NOT required to bypass IRM 'G'. In addition, continuing with the heatup is not appropriate until the IRM is monitored for 15 minutes. This is plausible based on the subtle difference between an IRM that is "noisy" as opposed to an IRM that "spikes". Sunday, February 17, 2008 3:28:27 PM 34

BFN Intermediate Range Monitors 3-01-92A Unit 3 Rev. 0014 Page 8 of 15 3.0 PRECAUTIONS AND LIMITATIONS (continued) L. [11fF] An IRM or SRM may be bypassed in the following conditions:

1. STOP control rod withdrawal and PLACE the channel in bypass when the SRM or IRM first gets noisy.

2. STOP control rod withdrawal and PLACE the channel in bypass immediately upon receipt of a single event large noise spike.

These conditions bypass the instrument for an operability assessment based on whether the noise is transitory or sustained. Transitory noise is considered a one time occurrence that does not repeat itself and the channel can be removed from bypass and restored to service. Sustained noise is when the duration exceeds 15 minutes and may result in signal build up until a trip signal is reached. If a trip or high flux signal was generated, the channel is required to be observed for at least 15 minutes before returning the instrument to service with concurrence from System Engineering. When the initial assessment and recognition of the magnitude of the event has been determined, then control rod withdrawal may be resumed where it has been left off as long as the minimum number of SRM and IRM channels operable are within the Technical Specification limits. [1I-B-91-04O] /-<:;;:'/Y7 M. [QA/C] SPP-10A requires approval of the Plant Manag~~forhi~~d;;~'~~~rior to any planned' operation with IRMs bypas~~9 unless bypassrng~iSspecrfically allowed within approved procedures. [ISE-NPS-92-R01]

BFN Panel 9-5 3-ARP-9-5A Unit 3 3-XA-55-5A Rev. 0037 Page 43 of 45 ( SensorlTrip Point: IRM Relay K-16 A. HI-HI;;:: 116.4 on 125 scale CHA, C, E, G B. INOP. HI-HIIINOP 1. Hi voltage low.

2. Module unplugged.

3. Function switch NOT in RED BAR f33 OPERATE.

(Page 1 of 1) 4. Loss of +/- 24 VDC to monitor Sensor Control Room Panel 3-9-12. Location: Probable A. Flux level at or above setpoint. Cause: B. One or more inoperable conditions exist. C. SI or SR in progress. D. Malfunction of sensor. E. Control rod drop accident. Automatic A. Half-scram if one sensor actuates (except with Rx Mode Sw. in RUN). Action: B. Reactor scram if one sensor per channel actuates, (except with Rx Mode Sw. in RUN). Operator A. STOP any reactivity changes. 0 Action: B. VERIFY alarm by multiple indications. 0 C. RANGE initiating channel or BYPASS initiating channel. REFER TO 3-01-92A~. 0 D. With SRO permission, RESET Half Scram. REFER TO 3-01-99 E. IF alarm is from a control rod drop, THEN REFER TO 3-AOI-85-1. 0 F. [NRC/C]IF one or more IRM recorder reading is downscale, THEN CHECK for loss of +/- 24 VDC power. 0 G. NOTIFY Instrument Maintenance that functional tests of any monitors indicating an INOP condition, including a downscale reading, are required before the instrument can be considered operable. [NRC IE item 86-40-03] 0 H. NOTIFY Reactor Engineer. 0 I. REFER TO Tech 'Spec Table 3.3.1.1-1, TRM Tables 3.3.4-1 and 3.3.5-1. 0

References:

3-45E620-6 197R114-16 GEK 3-730E915-10 3-730E915RF-12 3-01-92A 3-AOI-85-1 Technical Specifications Technical Requirements Manual-TRM

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours channels inoperable. OR A. 2 -------------NOT E------------- Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Place associated trip 12 hours system in trip. (continued) BFN-UNIT 3 3.3-1 Amendment No. 212, 213, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNElS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 ,.; 120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 ,.; 120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High. 2 3(b) G SR 3.3.1.1.1 ,.;15% RTP (Setdown) SR .3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16

b. Row Biased Simulated F SR 3.3.1 .1.1 ,.;0.66W Thermal Power - High SR 3.3.1.1.2 +66% RTP SR 3.3.1.1.7 and,.; 120%

SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16

c. Neutron Flux - High 3(b) F SR 3.3.1.1.1 ,.; 120% RTP SR 3.3.1.1.2 SR3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [.66 W + 66% - .66 L\ Wj RTP when reset for single loop operation per LCO 3.4.1. "Recirculation Loops Operating," BFN-UNIT 3 3.3-7 Amendment No. 216 December 23, 1998

TVANSTANDARD OPDP-8 DEPARTMENT LIMITING CONDITIONS FOR OPERATION TRACKING Rev. 1 PROCEDURE Page 50f24 3.3 TS LCO Evaluations 3.3.1 General Guidelines A. When equipment identified in TS is made or becomes inoperable, existing plant/unit conditions may require LCOs be entered. B. LCOs are entered if, for existing plant/unit conditions, TS require action(s) to be taken. C. LCOs are exited when the equipment is returned to operable status or when the plant/unit is put into a condition where TS no longer require action(s) to be taken. TS 3.0.6 and 5.7.2.18 (WBN)/5;5.11 (BFN) addresses LCO entry relative to support systems and supported systems. D. TS action requirements may change as the aggregate of inoperable equipment changes. Determination ofTS action(s) and the quantity of LCOs to be entered are based on the aggregate of inoperable systems, equipment, and components. E. Multiple LCOs shall be entered and logged if equipment is inoperable or removed from service and more than one TS LCO action is required to be taken. F. If equipment is identified or made inoperable that does not apply to an LCO based on the current planfconditions, an "Information Only" LCO should be entered into the Unit Log or LCO Tracking Log, as appropriate. The "Information Only" LCO entry should contain information similar to an "Active" LCO with possibly the exception ofthe LCO expiration date not being required. During plant shutdown/outages, it is not required to utilize INFORMATION ONLY LCOs for conditions that are applicable only in other modes which are controlled by other plant instructions (i.e., general operating instructions, surveillance instructions etc.). 3.3.2 Trackinq of Inoperable Equipment A. When equipment identified in TS is discovered or determined to be inoperable, LCO action is assessed by the affected unit's US. This assessment is based on existing and planned near-term plant/unit conditions and evolutions. The assessment shall be in accordance with Appendix B (WBN) or Appendix C (BFN) "Safety Function Determination Program (SFDP)" in accordance with the WBN TS 5.7.2.18 orBFN TS 5.5.11. B. The affected US(s) shall identify if the unit requires an entry into an LCO based on the current plant conditions. C. If TS require that actions be taken, LCOs associated with the inoperable equipment shall be entered into the Unit Log and/or LCO Tracking Log(s) maintained by the affected unit's US/designee. (

(REFERENCES PROVIDED TO ( . CANDIDATE I i,

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours channels inoperable. OR A. 2 -------------NOT E------------- Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Place associated trip 12 hours system in trip. (continued) BFN-UNIT3 3.3-1 Amendment No. 212, 213, 221 September 27,1999

RPS Instrumentation 3.3.1.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. -------------NOT E------------- B.1 Place channel in one trip 6 hours Not applicable for system in trip. Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR

   ----------------------------------

B.2 Place one trip system in 6 hours One or more Functions trip. with one or more required channels inoperable in both trip systems. C. One or more Functions C.1 Restore RPS trip 1 hour with RPS trip capability capability. not maintained. D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C not met. channel. E. As required by Required E.1 Reduce THERMAL 4 hours Action D.1 and POWER to < 30% RTP. referenced in TablE? 3.3.1.1-1. F. As required by Required F.1 Be in MODE 2. 6 hours Action D.1 and referenced in Table 3.3.1.1-1. (continued) BFN-UNIT 3 3.3-2 Amendment No. 212, 213, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required G.1 Be in MODE 3. 12 hours Action 0.1 and referenced in Table 3.3.1.1-1. H. As required by Required H.1 Initiate action to fully Immediately Action 0.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1. containing one or more fuel assemblies. I. As required by Required 1.1 Initiate alternate method 12 hours Action 0.1 and to detect and suppress referenced in Table thermal hydraulic 3.3.1.1-1. instability oscillations. J. Required Action and J.1 Be in MODE 2 4 hours associated Completion , Time of Condition I not met. BFN-UNIT 3 3.3-3 Amendment No. 212, 213, 221,231 September 13, 2001

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

N()TES----------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.

SURVEILLANCE ' FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.1.1.2 --------------------------N()TE---------------------M--- Not required to be performed until 12 hours after THERMAL P()WER ~ 25% RTP. Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is

2% RTP while operating at ~ 25% RTP.

SR 3.3.1.1.3 --------------------------N()TE--------~---------------- Not required to be performed when entering M()OE 2 from M()OE 1 until 12 hours after entering M()OE 2. Perform CHANNEL FUNCTI()NAL TEST. 7 days (continued) BFN-UNIT 3 3.3-4 Amendment No. 213 September 03, 1998

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6 --------------------------NOTE------------------------- Only required to be met during entry into MODE 2 from MODE 1. Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. 1000 MWDIT average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92.days SR 3.3.1.1.9 -------------------------NOTES----------~-------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL CALIBRATION. 92 days (continued) BFN-UNIT 3 3.3-5 Amendment No. 213 September 03, 1998

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.10 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.11 (Deleted) SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 --------------------------NOTE------------------------- Neutron detectors are excluded.

              ------------------------------------------------------------

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST. SR 3.3.1.1.15 Verify Turbine Stop Valve - Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is ~ 30% RTP. SR 3.3.1.1.16 --------------------------NOTE------------------------- For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

              ------------------------------------------------------------

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is ~ 25% and recirculation drive flow is < 60% of rated recirculation drive flow. BFN-UNIT 3 3.3-6 Amendment No. 212, 213, 215, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION 0.1

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 <; 120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 <;120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High, 2 3(b) G SR 3.3.1.1.1 <; 15% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16

b. Row Biased Simulated F SR 3.3.1.1.1 <;0.66 W Thermal Power - High SR 3.3.1.1.2 +66% RTP SR 3.3.1.1.7 and <; 120%

SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16

c. Neutron Flux - High F SR 3.3.1.1.1 <; 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [.66 W + 66% - .66 t. W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." BFN-UNIT 3 3.3-7 Amendment No. 216 December 23, 1998

0610 NRC SRO Exam

13. SRO 259002G2.1.23 001lCINT2G1ICAUTION1I31259002G2.1.23/4.0/SRO ONLY/MOD 113112008 Given the following Unit 2 conditions:

Reactor pressure: 10 psig
Drywell temperature: 250°F
Secondary Containment temperatures:

74-95F 220°F 74-95C & D 245°F 69-835A thru D 260°F 69-29F, G & H 200°F

Reactor water level indications: A Ii ?

If!" -+r~~.

            ><LI-3-58A & B
       ~LI-3-g§t.~~2A Erratic/

(-)1S0,inches ~ J W~ ~

            --:::4-1-3-53, 60 & 206            0 i,nqhes
             )<LI-3-55                         0 inches Which ONE of the following describes the required action and the basis for that action?

Enter _ _ _ _('-'1.....) _ _ _ _ due to _ _ _ _ _ _.....(0.:; ;:2... ) _ _ _ _ _ _ in the reference legs of LI-3-58A & B, as a result of low reactor pressure. REFERENCE PROVIDED B." 2-~~~.~~'~~~~,I~~~i,~::::1.:",,,~ ___, flashing steam O.(-'2~EOI-C-1, "Alternate L~vel Control;" ) noncondensible gases coming out of

           -:::----"-'--,, "-"-_. """-"'"                  solutiorr---

D. 2-EOI-C-4, "RPV Flooding;" noncondensible gases coming out of solution Sunday, February 17, 2008 3:28:27 PM 35

0610 NRC SRO Exam KIA Statement: 259002 Reactor Water Level Control 2.1.23 - Conduct of Operations Ability to perform specific system and integrated plant procedures during all modes of plant operation. KIA Justification: This question satisfies the KIA statement by requiring the candidate to' use specific plant conditions to determine the appropriate procedure and its bases to control reactor water level.

References:

2-EOI-C-4 Flowchart, EOIPM Level of Knowledge Justification: This question is rated as CIA due to the re.quirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 36

0610 NRC SRO Exam REFERENCE PROVIDED: EOI Caution 1, Curve 8, Table 6, PIP 95-64 Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following:

1. RPV pressure vs. temperature is in the UNSAFE area of Curve 8, RPV Saturation Temp.

2. Perfdrm corrections to L,I-3-5f ~ 62A using PIP 95-64 to determine their proximity to TAF;.

3. Recognize Ll3-53, 60, 206 & 55 are indicating below their Minimum Indicated Level for the current temperatures in the Secondary Containment.

4. Recognize that erratic indication on LI-3-58A & 58B is indicative of reference leg flashing.

5. Recognize the difference between "flashing" and "notching" with respect to indicated instrument response.

B - correct: A - incorrect: Flashing steam in the reference legs of LI-3-58A1B is indication of a loss of RPV level instrumentation since Curve 8, RPV Saturation Temp is currently unsafe. However, the correct action is to enter 3-EOI-C4, RPV Flooding. This is plausible because LI-3-52 and 62A are approaching

                         .IAE: which would indicate a need totransition to 3-EOI-C1, "Alternate Level Control."

C - incorrect: Noncondensible gases are indicative of "Notching" instruments caused

         .,               by rapid depressurization. In addition, the incorrect EOI flowchart is entered. This is plausible because LI-3-52 and 62A are approaching IAI=,. which would indicate a need to transition to 3~EOI-C1, "Alternate Level Control."

D - incorrect: Non-condensible gases are indicative of "Notching" instruments caused by rapid depressurization. However, the correct EOI flowchart is entered due to a loss of RPV level instrumentation. Sunday, February 17, 2008 3:28:27 PM 37

CAUTIONS CAUTION #1 .. A.N RP\/ "l':ATER L\!L INS-:"RU ~.AE1,,"::- ~~'\4Y BE uSED TO DE~ERJlllNE (JH TREND L\lL ONLY \;'VHEh 17 REA.DS .lt3;JVE

   ~l';; E ?,,*H~. dl\'\L M IN DH..:-..t(!' ED L\/L .lISSCK::J..:J.:-;"ED V~"TTH THE H:IGh'ESi r..-tll,..:x: a'll\' OR 8C: RUN -:'Er\~p" II- IF Dt':'EMPS, OR            se .4REA -EMPS CTABLE 6), AS APPLICABLE, ARE OUTSIDE THE SAFE REGION OF CURVE S,
   -'1' E ASSOCI/FED INS',RUMEN" /,.v"y 8E UKREUABLE DUE -'-0 BOILING IN                                            ~HE  RUN.

M!NHvlUM f.1AX DW RUNTD"lP ~..lAX SC INSTRUMENT INDICATED WROM XR-64-50 RUN TEMP L\!L OR Il..64.. 52AB) (FROM TABLE 6) ON SC,A.LE NiA BELO.V 15~1

                                                                       .. '145                              Ni.A.                        151 102(1()

EMERGENCY t---:.....::----4-----N::,*,I.:...A.~-----+--:..:..::....:..::...::..:::..:~--l _*155 Tej +1iQ. -14(1 2{}1 TO 2S{!

                                                                       .0                              N/,A,                         251 TO 3(}C
                                                                       -120                                 Ni.A.                        3G110 353 ON SCALE                                 NlA                          BELOW'15G U .. 3 .. 5D                                                                                                                            '151 TO .2G{l NORMAL U..3.. 206                                                                                                 NI.A,
                                           -Gl'(} 4S'G LI-.3-2S3                                                                                                  NiA                          .251 TO 3(H}

LI-3-2Cg.~\, B,C, D 301103St} U-3.. :;.2 POST Li-3.--21\ ACCIDENT ON SCALE. N!A t,¥!\

                                        -268, TO .;.32 Baovv 100                                       NIP"
                                                                           +tS                  'IX TO 150                                     ~¥A S!IUTDOV...N       I-_ _~-=ZJ=-'_ _ _+          __-!..=-.!..:=-=:.::....

151 T02()*c) _ _ _+ ________ Ni!:>" .....j FLOODUP .30 .20'1 T02S;} t,J//\

                                                                                               .251 TO 3(!-u                                   N'A 33'1 TO 35{)                                    NiA 351 TO 4;:}(i-                                 Nip.,

CURVE 8 RPV SATURATION TEMP 4*00 3fH} Ii- 300 ~- v', z

> 340 Ir 0

Z uJ 320

> 300 Ci'.

1-" Q'j ~ 200 Ir ~ 200

z 4.

240 uJ f-220 200

{) 50 100 150 200 250 RPV PRESS {pSIG)

                                                  *CONSTANT ABOVE 250 PSiG TABLE 6 SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT                  SC TEMP ELEMENTS AND LOCA.TIONS El621         El593                 ElS65        RWCU HXRlYl 174-95F)  (74-9SC AND D)      (69-83514. THRU D)  (S9-Z!lF. G. H)

LI-3~58A ~F QF NfA ~F LI-3-S88 (.'F ~ NfA N/A L1-3*5J ~ UF N1A of U-J-$O ilF ilF NlA N!A L1-3-2fi6 ilF ~ NfA uF LI-3-253 <f of NfA NiA U-3-52 ilF uF uF NlA LI-3-62A ilF "F uF NfA U-3-55 of c'F NlA NiA L1-3-2C!-8A. B ilF ilF NfA up LI-3-20ac. u ('I' ilF NlA N/A

OPL 171.201 Revision 7 Page 33 of 117 INSTRUCTOR NOTES

d. It is important to note that the information SER 03-05 presented in Caution #1 is not just a simple accommodation for inaccuracies in RPV water level indication which occur when plant conditions are different from those for which the instruments are calibrated. Rather, the caution defines conditions under which the displayed value and the indicated trend of RPV water level cannot be relied upon.

e. Part B of Caution #1 identifies the limiting SER 03-05 conditions beyond which water in instrument legs may boil. Water in the RPV water level instrument legs is maintained in a liquid state by cooling action of the surrounding atmosphere and pressure in the reactor vessel. Water in the instrument legs will boil, however, if its temperature exceeds saturation temperature for the existing RPV pressure.

f. Boiling is a concern in both horizontal and SER 03-05 vertical reference and variable instrument leg runs. Boil-off from reference leg water inventory reduces the reference head of water, decreases dP sensed by the instrument, and results in an erroneously high indicated RPV water level. Boiling in the instrument's variable leg exerts increased pressure on the variable leg side of the dP cell. This effect results in a lower sensed dP and an erroneously high indicated RPV water level.

g. Part B of Caution #1 references the RPV SER 03-05 Saturation Temperature Curve (Curve 8) The RPV Saturation Temperature Curve is generic, based simply on the properties of water. The axis for RPV pressure is plotted from atmospheric pressure to the pressure setpoint of the lowest lifting MSRV. Note that the temperature axis of the RPV Saturation Temperature Curve is not simply drywell temperature. Depending upon the relative location of instrument reference legs and variable legs, indications from monitors near instrument runs must be considered.

OPL 171.201 Revision 7 Page 34 of 117 INSTRUCTOR NOTES

h. Because BFN does not have the capability of directly reading temperature indications near instrument runs located in secondary containment, the RPV Saturation Temperature Curve (Curve 8) is supplemented with Table 6, Secondary Containment Instrument Runs. Table 6 identifies the temperature elements and general locations for the instrument runs to each RPV water level instrument.

L Caution 1 part B says instruments "may be unreliable" if Curve 8 is exceeded. This means instruments may continue to be used until and unless erratic indication is observed since momentary excursions (expected in some post LOCA situations) into curve 8 unsafe region will not result in boiling. If, however, indications of boiling are observed then that instrument is unusable until the instrument lines can be cooled and refilled. J. Part A of Caution #1 allows the operator to The instrument determine if each indicated RPV water level will indicate high range is reliable by being above the Minimum by the amount of Indicated Level for each of a series of this offset instrument run temperature ranges. throughout its Engineering calculations have determined range. that when indicated RPV water level is above the Minimum Indicated Level, the operator is assured that actual RPV water level is above the instrument variable leg tap, and trends are valid.

k. The Minimum Indicated Level is defined to be SER 03-05 the highest RPV water level instrument indication which results from off-calibration instrument run temperature conditions when RPV water level is actually at the elevation of the instrument variable leg tap. Separate levels are provided for each RPV water level instrument.

iREFERENCES

PROVIDED TO

( CANDIDATE

( CURVE 8 RPV SATURATION TEMP 400 380 tS60 co

              ~     340
              ~     320

Ii

              !     300
              !     280
              ~

260 m240 220

   "' .... ""                                   TABLES SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT                 SC TEMP El...ElMENTS ANOLOCATIONS eL621         eLS93                  eL565          R.WCUHXRM (14-95F), (74-Q5C ANOO)       .iDh ...."'". THRlJ 0>> (69-29F,G. H)
  -            LI..:w&A.           *Of          Of                     KIA                !If N/A LI~                  Of          Of                     KIA LI.a..03 ~-          Of_         Of                     KIA
                          -                                                               !If
      ?        LI-3-00 LI..a..206 -

F Of Of Of KIA NlA. N/A

                                                                                          !If
 -             LI..a..253 LI..a..52 LI..a..e2A LI..a..55 Of
                                    !IF.

Of Of

                                                !If IlF IlF Of
                                                                       ~

jllf iOf

                                                                     \1:NA /'
                                                                              \1 J

N/A N/A NfA N/A LJ..3..2.08A,B IlF Of NlA OF lI-3--208C, 0 !If Of NlA N/A

3-Ll-3-52 & 62 CORRECTION CURVES _ r

       -150" TAF     ,,-162" -                                         -162"      = TAF (RED LINE)
                                                         -185"     = MSCRWL (GREEN LINE)
       -175"                                             -200"     = MZIRWL (BLUE LINE)
                                                         -215"     = TWO-THIRDS CORE HEIGHT (BLACK LINE)
 ....J          <

W r\

W r-.. r-.r-,

 ....J -200"
                                                    ""I"-

0 1\ 1-01"- W ~ ACTUAL I- r-.. t-.

<<

() "" r-I"-r-

                                                                                          ~,...

i'- LEVEL r-. 0 -225 I' l"- t-. ~

                                                                                                                     ~       -162" Z                                 l'
                  "      1\                  r-,r-,

f-.. l-

I' r- -185"

       -250" I"-""r-                                                                 "I...L
                                                                                                                       '" -200"
                                                                                                                       ~
                                                                                                                       '" -215"
       -268"       J o              100 200                         300 400            500 600       700 800   900 1000 1100 REACTOR PRESSURE (PSIG)                                             PIP-95-64 REV. 12

CAUTIONS CAUTION #1

AN Rf'V WATER l VllNSTRUMENT MI\Y BE USED TO DETERMINE OR TRENtllVl ONlY \lVHENIT READS ABOVE THE MINIMUMINOICATED LVl. ASSOCIATED 'WITH THE HIGHEST MAX OW OR SC RUN TEMP .
IF OW TEMPS, OR SC AREA TStilPS (fA8L.£: 6), AS APPLlCA8L.E, ARE OUTSIDE THE SAFE REGION Of" CURVE 8, THE ASSOCIATEO lNSTRUMENi MAY 6EUNREUA8L.EOUE TO SOILING IN THE RUN.. "

I'-<.-~* .* MINIMUM MAX DW RUN TEMP MAXSC INSTRUMENT RANGE INDICATED (FROM XR-tJ4..50 RUN TEMP LVL OR TI..64-52AS', (FROM TABLE 6) ON SCALE N/A BElOW 100

                                                   *145             NfA                  151 TO 200 EMERGENCY                                   N/A U-.3-5M,B                                        -140                                  201 TO 2.00
                         -15510 +00
                                                   -130              N/A                 251TO~
                                                   -120             NiA                  30110350 U-.3-53    ~

ON SCALE N/A BElOW 100 U-3-00 .'~ .+i? NlA 151 TO 200 NORMAL U-3-2re -*...>>~15) N/A 201 TO 400 OT0400 LI-.3-253 +'20 N!A 251TO~ U-.3-2:OfI.A, B, C, D +30 N/A 301 TOsOO POST j,/ U-3-52 ACCIDENT ON SCALE N/A NlA U-3-62A

          ,              -268 TO +32                .

( +10

                                                     +15 BELOW 100 100 TO 100 NlA NtA SHUTOOWN                    +20         15110200                    NiA U-3-55                  FlOODUP                  (t.3Ql        20110200                    NiA OTO +400                   +40                                     NiA 25110300
                                                     +00         30110350                    NtA
                                                     +65-        351 TO 400                  NiA

0610 NRC SRO Exam

14. SRO 205000A2.06 001lCIA/T2G1I1-AOI-74-1II205000A2.061ISRO ONLYINEW 1211812007 RMS Given the following plant conditions:

(

Unit 1 is in Mode 4 at 19.5 of following a shutdown 18 hours earlier for refueling.
RPV Level is (+)80 inches and§teC)dy.
BOTH 'Recirc Variable Frequency Drives (VFD) are tagged for maintenance
RHRLoop !lis in Shutdown Cooling when a loss of RPS 'B' occurs.

Determine which ONE of the following describes the effects on temperature monitoring and decay heat r~moval (assume no other operator actions have occurred)? Then, determine the course of actions to restore a primary decay heat removal system .

          .
                                                                                                                                                        \t         I r Temperature monitoring is _ _ _ _ _----'('-'1.....) _ _ _ _ _ _ . AI~[I§l!~_<!~~~y heat r~!!lQygl    is prov,ided by                        (2)                                                                                The proper course of action to restore a primary decay heat removal system is to _ _ _ _ _---->,..;:(3'-'-)_ _ _ _ __

A'! (1) avai@able due to natural*::cor<W~n,circulation. ~/ (i?i~~.~~ent

             ".....          ..     ~

losse~;;~)

                               ..... ... ........
                                        -~

transfer RPS 'B' to alternate, reset PCIS, align RHR Loop" and start a RHR pump. B. (1) avail@lJkble due to natural G@;f.l¥e.Ct~efl* circulation (2) the Alternate Decay Heat Removal System.)

                                                    ,           . - -',' ... >,,~~--,,~"~""'"'-'"'. ~--,.""""""""-"'~--,,<:-.,,-"'---~",/

transfer RPS 'B' to' alternate, reset PCIS, align RHR Loop" and start a RHR pump. C. (1) NOT~",~il.c@cl~le since forced circulation is unClvailable. v' (2)Jb~ Reactor Water Cleanup Syst~m./

                        "'.~.--"- "~--                               -",-,."

Remove Clearance Tag(s), place Recirc VFD in service, and start a Recirc Pump. D. (1) NOT avai.~ble since forced circulation is unavailable. v,/ (~Ltb.sLAlt~mClte Decay Heat Removal System. v (3) Remove Clearance Tag(s), place Recirc VFD in service, and start a Recirc Pump. Sunday, February 17, 2008 3:28:27 PM 38

0610 NRC SRO Exam KIA Statement: 205000 Shutdown Cooling A2.06 - Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use proced~....co.rrect,_c.QJJJ:[~I, or mitigate the consequences of those abnormal conditions0DC/RH~__ p~_~~_t~i~~ KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate procedure to mitigate and/or recover Shutdown Cooling following an event resulting in a SDC/RHR pump trip.

References:

1-AOI-74-1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 39

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: Based on current conditions, whether natural circulation is available or Whether forced circulation is required. Natural circulation is accomplished only if RPV level is above the top of the moisture seperators to complete a path for circulation through the active fuel region and back into the downcomer region outside the core shroud. An RPV level of (+) 80 inches given in the stem allows natural circulation. In addition, the candidate must determine the amount of decay heat removal necessary based on current conditions. With reactor coolant temperature below saturation, ambient losses are sufficient 18 hours after shutdown as long as some circulation of coolant is available. Based on the answers determined for parts one and two, a proper course of action must be determined to restore a primary method of decay heat removal. The logical choice is to restore SD Cooling by taking the appropriate steps to restore RPS, reset the PCIS isolation, and re-align the RHR system. A - correct: B - incorrect: Although (1) and (3) are correct, the Alternate Decay Heat Removal lineup also requires RPS power and PCIS reset. In addition, this lineup is designed as a backup to fuel pool cooling decay heat removal, not directly from fuel loaded in the reactor vessel. C - incorrect: Temperature monitoring IS available via natural convection circulation. Since part (1) is incorrect, part (3) is not correct. Restarting Recirc Pumps is not required to restore temperature monitoring. However, although not required, the RWCU system is capable of removing decay heat in excess of ambient losses, so this lineup would be appropriate if Parts (1) and (3) were corrrect. D - incorrect: Temperature monitoring IS available via natural convection circulation. Since part (1) is incorrect, part (3) is not correct. Restarting Recirc Pumps is not required to restore temperature monitoring. The Alternate Decay Heat Removal lineup also requires RPS power and PCIS reset. In addition, this lineup is designed as a backup to fuel pool cooling decay heat removal, not directly from fuel loaded in the reactor vessel. Sunday, February 17, 2008 3:28:27 PM 40

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 6 of 28 4.0 OPERA TOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions CAUTIONS

1) Reactor vessel stratification may occur until Shutdown Cooling is restored or a Reactor Recirculation Pump is placed in service.

2) Loss of Shutdown Cooling during the first 24 hours is most critical due to massive decay heat and limitations on the RHRSW piping. If Shutdown Cooling is lost during the first 24 hours post reactor shutdown, priorities shall be placed on the recovery of shutdown cooling in an expeditious manner [BFN PER 02-003140:-000].

NOTE The following systems, if available, may be used as alternate methods of decay heat removal. REFER TO the applicable Tec Spec Bases B 3.4.7, B 3.4.8, B 3.9.7, B 3.9.8 ADHR System- (0-01-72) Fuel Pool Cooling System- (1-01-78) RWCU System- (1-01-69) Ambient losses with natural or forced circulation [1] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s). (Otherwise N/A) o [2] NOTIFY the Shift Manager. o [3] IF Refueling is in progress, THEN NOTIFY the Refueling Floor SRO. (Otherwise N/A) o [4] REVIEW EPIP.;1, Emergency Plan Classification Logic, for entry conditions. (Otherwise N/A) o

0610 NRC SRO Exam

15. SRO 212000A2.12 001lCIA/T2Glll/212000A2.121ISRO ONLYIMOD 113112008 Unit 2 is at 40% power following a refueling outage. , " . ," j. .....'11 r-.e..<.,f-6:;..i!:;:; --n.\.e ..,L ~. ~.. e T~§-~the_MajD~T'.~biIJH_BtQP-\l:~llle_(T£.\L}j\umb~L~~e'l~~~\~ J2Lobt~~\-With~t~ -ecV!~~+ -fP limit switch~l:ta.S.sDciate'QwithB()'fH"f~:em;torProtection'Sy:stem~(RPS)-trip-~ste-r:tls. \ C W I ~~es

      . . . . Ihelimit.swttq,h.assoGiate~with.this~\1~e:;wiIl*Ne:r,,~tfil3wt:le\'-stb.e.,'ol~'hv~~)§.gIQ§:~cj .. ALL \ tvIJ2.

other vaI\Tes'ftft'lQtion'nnrmatiy. -'- ....... Determine which ONE of the following describes the effect on the RPS trip function, associated with the main turbine trip, in this condition and the required remedial action. The SCRAM function due to Main Turbine trip above 30% power (1) Remove fuses for RPS A and B, associated with TSV number 4, (2) REFERENCE PROVIDED A. is maintained; within 1 hour. B. is NOT maintained; within 1 hour. C~ is maintained;. within 4 hours. ( D. is NOT maintained; within 4 hours. Sunday, February 17, 2008 3:28:27 PM 41

0610 NRC SRO Exam KIA Statement: 212000 RPS A2.12 - Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Main turbine stop control valve closure. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate status and required actions for a Main turbine stop control valve closure logic failure.

References:

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam ( MODIFIED FROM OPL 171.028 #113 Sunday, February 17, 2008 3:28:27 PM 42

0610 NRC SRO Exam REFERENCE PROVIDED: Unit 2 TS .3.3.1.1 Table and 01-99 Trip Function Table Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following:

1. Recognize that the limit switch has failed in the non-conservative direction for the

                                                                -~
            . given plant conditions. (i.e; >30% power)

2. Based on Item 1 above, recognize that the TCV/SV Closure Scram capability is still maintained as long as power remains above 30%.

3. Once the trip function capability has been determined, identify the correct TS Action statement to address this situation.'

C - correct: A - incorrect: TCV/SV Closure Scram capability is maintained. However, the actions in the distractor match the required actions in the LCO if the TCV/SV Closure Scram capability was NOT ~aintained. B - incorrect: TCV/SV Closure Scram capability IS maintained. However, the actions in the distractor match the required actions in the LCO if the TCV/SV Closure Scram capability was NOT maintained. D - incorrect: TCV/SV Closure Scram capability IS maintained. However, the actions ( in the distractor match the required actions in the LCO if the TCV/SV Closure Scram capability WAS maintained. Sunday, February 17, 2008 3:28:27 PM 43

(REFERENCES PROVIDED TO i*** \ CANDIDATE

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours channels inoperable. OR A.2 -------------NOT E------------- Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Place associated trip 12 hours system in trip. (continued) BFN-UNIT 2 3.3-1 Amendment No. 258 March 05, 1999

RPS Instrumentation 3.3.1.1 ( \ ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. ------------- NOTE ------------- B.1 Place channel in one trip 6. hours Not applicapJefor system in trip. FUlJctiens~2.a, 2.b, 2.c, 2.d, or 2.f. OR

                                                                         ,
     ----------------------------------

B.2 Place one trip system in 6 hours One or more Functions trip.

       "   I                   ",

with on.e or more required channels inoperable ih both trip systems. C. One or more Functions C.1 Restore RPS trip 1_,b.our with RPS trip capability capability. not maintained. ( D. Required Action and 0.1 Enter the Condition Immediately associated Completion .- referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C not met. - channel. E. As required by Required E.1 Reduce THERMAL 4 hours Action 0.1 and POWER to < 30% RTP. referenced in , Table 3.3.1.1-1. F. As required by Required F.1 Be in MODE 2. 6 hours Action 0.1 and referenced in Table 3.3.1.1-1. (continued) BFN-UNIT 2 3.3-2 Amendment No. 258 March OS, 1999

RPS Instrumentation 3.3.1.1 A CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required G.1 Be in MODE 3. 12 hours Action 0.1 and referenced in Table 3.3.1.1-1. H. As required by Required H.1 Initiate action to fully Immediately Action 0.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1. containing one or more fuel assemblies. I. As required by Required 1.1 Initiate alternate method 12 hours Action 0.1 and to detect and suppress referenced in Table thermal hydraulic 3.3.1.1-1. instability oscillations. J. Required Action and J.1 Be in Mode 2. 4 hours associated Completion Time of Condition I not met. ( BFN-UNIT 2 3.3-3 Amendment No. ~ 273 July 26, 2001

RPS Instrumentation 3.3.1.1 ( SURVEILLANCE REQUIREMENTS

 ----------------------------------------------------N()TES----------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.1.1.2 --------------------------N()TE------------------------- Not required to be performed until 12 hours after THERMAL P()WER;::: 25% RTP. ( Verify the absolute difference between the average power range monitor (APRM) 7 days channels and the calculated power is s; 2% RTP while operating at;::: 25% RTP. SR 3.3.1.1.3 --------------------------N()TE------------------------- Not required to be performed when entering M()DE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTI()NAL TEST. 7 days (continued) BFN-UNIT 2 3.3-4 Amendment No. 253

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6 --------------------------NOTE------------------------- Only required to be met during entry into MODE 2 from MODE 1.

              ------------------------------------------------------------

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. 1000 MWDfT average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 -------------------------NOTES------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

               ------------------------------------------------------------

Perform CHANNEL CALIBRATION. 92 days (continued) BFN-UNIT 2 3.3-5 Amendment No. 253

RPS Instrumentation 3.3.1.1 ( SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.10 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.11 (Deleted) SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 --------------------------NOTE------------------------- Neutron detectors are excluded.

                ------------------------------------------------------------

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST. SR 3.3.1.1.15 Verify Turbine Stop Valve - Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is;;:: 30% RTP. SR 3.3.1.1.16 --------------------------NOTE------------------------- For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

                 ------------------------------------------------------------

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is ;;:: 25% and recirculation drive flow is < 60% of rated recirculation drive flow. BFN-UNIT 2 3.3-6 Amendment No. 258 March 05, 1999

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 :;; 120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 :;; 120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

(~~ a. Neutron Flux - High, (Setdown) 2 3(b) G SR SR 3.3.1.1.1 3.3.1.1.6

15% RTP SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16

b. Flow Biased Simulated 3(b) F SR 3.3.1.1.1 :;;0.66W Thermal Power - High SR 3.3.1.1.2 +66% RTP SR 3.3.1.1.7 and:;; 120%

SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16

c. Neutron Flux - High 3(b) F SR 3.3.1.1.1 :;; 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [.66 W + 66% - .66 il W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating. NOTE: This page is applicable after commencing Cycle 11 operation. BFN-UNIT 2 3.3-7 Amendment No. 256 December 23, 1998

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION 0.1

2. Average Power Range Monitors (continued)

d. Inop 1,2 G SR 3.3.1.1.16 NA

e. 2-0ut-Of-4 Voter 1,2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.16

f. OPRM Upscale SR 3.3.1.1.1 NA SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17

3. Reactor Vessel Steam Dome 1,2 2 G SR 3.3.1.1.1 ~ 1090 psig Pressure - High(d) SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14

4. Reactor Vessel Water Level - 1,2 2 G SR 3.3.1.1.1 ~ 528 inches Low, LeveI3(d) SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14

5. Main Steam Isolation Valve - 8 F SR 3.3.1.1.8 ~ 10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14

6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 ~2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Resistance Temperature 1,2 2 G SR 3.3.1.1.8 ~50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 ~50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable. Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable. The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report. BFN-UNIT 2 3.3-8 Amendment No. 253, 254, 258, 260, 296 September 14, 2006

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 30f3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION 0.1

7. Scram Discharge Volume Water Level - High (continued)

b. Float Switch 1,2 2 G SR 3.3.1.1.8 ::;46 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 ::;46 gallons SR 3.3.1.1.13 SR 3.3.1.1.14

@Urbine Stop Valve - Closure ~ 30% RTP E SR 3.3.1.1.8 ::; 10% closed SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Control Valve Fast ~30% RTP 2 E SR 3.3.1.1.8 ~550 psig Closure, Trip Oil Pressure - SR 3.3.1.1.13 Low(d) SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.14 5(a) H SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1,2 G SR 3.3.1.1.8 NA SR 3.3.1.1.14 5(a) H SR 3.3.1.1.8 NA SR 3.3.1.1.14

12. RPS Channel Test Switches 1,2 2 G SR 3.3.1.1.4 NA 5(a) 2 H SR 3.3.1.1.4 NA 13.Deleted (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable. Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable. The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report. BFN-UNIT 2 3.3-9 Amendment No. 258, 276, 296 September 14, 2006

BFN Reactor Protection System 2-01-99 Unit2 Rev. 0073 Paae 67 of 77 Illustration 3 (Page 1 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-FCV-1-14(LS-4) 2-FU1-1-15D 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3A) 03A EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A1 Channel See 2-FCV-1-15 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-14(LS-3) 2-FU1-1-15E 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3B) 03B EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B1 Channel See 2-FCV-1-15 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-15(LS-4) 2-FU1-1-15D 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3A) 03A EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A1 Channel See 2-FCV-1-14 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-15(LS-3) 2-FU1-1-15E 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3B) 03B EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B1 Channel See 2-FCV-1-14 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-26(LS-4) 2-FU1-1-27A 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3E) 03E EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A1 Channel See 2-FCV-1-27 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-26(LS-3) 2-FU1-1-27B 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3D) 03D EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B2 Channel See 2-FCV-1-27 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-27(LS-4) 2-FU1-1-27A 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3E) 03E EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A1 Channel See 2-FCV-1-26 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-27(LS-3) 2-FU1-1-27B 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3D) 03D EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B2 Channel See 2-FCV-1-26 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions. -------

BFN Reactor Protection System 2-01-99 Unit2 Rev. 0073 Paae 68 of 77 Illustration 3 (Page 2 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-FCV-1-37(LS-4) 2-FU1-1-38A 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3C) 03C EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A2 Channel See 2-FCV-1-38 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-37(LS-3) 2-FU1-1-38B 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3F) 03F EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B1 Channel See 2-FCV-1-38 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-38(LS-4 ) 2-FU1-1-38A 2-RLY-099-05AKO 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3C) 3C EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A2 Channel See 2-FCV-1-37 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-38(LS-3) 2-FU1-1-38B 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3F) 03F EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B1 Channel See 2-FCV-1-37 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-51 (LS-4) 2-FU1-1-52A 2-RLY-099-05AK 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3G) 03G EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A2 Channel See 2-FCV-1-S2 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: S NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-S1 (LS-3) 2-FU1-1-52B 2-RLY-099-05AK 9-17 2-730E91S-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (SAF3H) 03H EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B2 Channel See 2-FCV-1-S2 (LS-3) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: S NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-52(LS-4) 2-FU1-1-52A 2-RLY-099-05AKO 9-15 2-730E915-9 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3G) 3G EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). A2 Channel See 2-FCV-1-51 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. 2-FCV-1-52(LS-3) 2-FU1-1-52B 2-RLY-099-05AK 9-17 2-730E915-10 NONE EACH MSIV HAS SEPARATE LIMIT SWITCHES ON MSIV (5AF3H) 03H EACH VALVE FOR RPS A(LS-4) AND RPS B(LS-3). B2 Channel See 2-FCV-1-51 (LS-4) BOTH LS-3 AND LS-4 CAN BE DE-ENERGIZED AND Function: 5 NO 1/2 SCRAM IS RECEIVED. NOTE: Device Function corresj:londs to the TS Table 3.3.1.1 Functions.

o BFN Reactor Protection System 2-01-99 Unit 2 Rev. 0073 Page 69 of 77 Illustration 3 (Page 3 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-PIS-1-81A 2-FU1-1-81AA 2-RLY-099-05AK 9-17 2-730E915-10 2-XA-55-5B-16 WHEN FUSE PULLED ALARM WILL TURB FIRST STAGE (5AF9D) 09D 2-45E763-10(RPT) TURB CV FAST CLOSURE RESET AND 2-RLY-099-05AK09D PRESS BYPASS 2-45E671-44 TURB SV CLOSURE DE-ENERGIZES. THIS ENABLES RX B2 Channel SCRAM/RPT TRIP LOGIC SCRAM AND RPT TRIP BY Function: 8 & 9 BYPASS SIMULATING >30% POWER. I 2-PIS-1-81B 2-FU1-1-81 BA 2-RLY-099-05AK 9-15 2-730E915-9 2-XA-55-5B-16 WHEN FUSE PULLED ALARM WILL TURB FIRST STAGE (5AF9C) 09C 2-45E763-10(RPT) TURB CV FAST CLOSURE RESET AND 2-RLY-099-05AK09C PRESS BYPASS 2-45E671-32 TURB SV CLOSURE DE-ENERGIZES. THIS ENABLES RX A2 Channel SCRAM/RPT TRIP LOGIC SCRAM AND RPT TRIP BY Function: 8 & 9 BYPASS SIMULATING >30% POWER. 2-PIS-1-91A 2-FU1-1-91AA 2-RLY-099-05AK 9-17 2-730E915-10 2-XA-55-5B-16 WHEN FUSE PULLED ALARM WILL TURB FIRST STAGE (5AF9B) 09B 2-45E763-9(RPT) TURB CV FAST CLOSURE RESET AND 2-RLY-099-05AK09B PRESS BYPASS 2-45E671-38 TURB SV CLOSURE DE-ENERGIZES. THIS ENABLES RX B1 Channel SCRAM/RPT TRIP LOGIC SCRAM AND RPT TRIP BY Function: 8 & 9 BYPASS SIMULATING >30% POWER. 2-PIS-1-91 B 2-FU1-1-91 BA 2-RLY-099-05AK 9-15 2-730E915-9 2-XA-55-5B-16 WHEN FUSE PULLED ALARM WILL TURB FIRST STAGE (5AF9A) 09A 2-45E763-9(RPT) TURB CV FAST CLOSURE RESET AND 2-RLY-099-05AK09A PRESS BYPASS 2-45E671-26 TURB SV CLOSURE DE-ENERGIZES. THIS ENABLES RX A1 Channel SCRAM/RPT TRIP LOGIC SCRAM AND RPT TRIP BY Function: 8 & 9 BYPASS SIMULATING >30% POWER. NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

*.                                  BFN Unit2 Reactor Protection System                 2-01-99 Rev. 0073 Page 70 of 77 Illustration 3 (Page 4 of 11)

Actions to Place RPS Instruments in Tripped CondiUons (TS Table 3.3.1.1-1) CAUTION This table was written for the removal of one fuse. If removing two fuses research the logic carefully for effects on RPS and RPTs. If two fuses are removed, one fuse in an "A" channel and one fuse in a B channel, and a Rx scram and turbine trip occur while these fuses are removed, they are required to be reinstalled in order to reset the scram and RPT logic. DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS TURB STOP VLV#1 A1 CHANNEL 2-FU1-1-74A 2-RLY-099-05AK10A 9-15 2-730E915-9 NONE BOTH CIRCUITS FOR ANY SINGLE TURB 2-FCV-1-74 (5AF10A) STOP VLV CAN BE DE-ENERGIZED AND B1 CHANNEL 2-FU1-1-74B 2-RLY -099-05AK1 OB 9-17 2-730E915-10 NO 1/2 SCRAM OR RPT LOGIC ACTUATION Function: 8 (5AF10B) 2-RLY-099-05AK10K 2-45E763-9(RPT) WILL OCCUR. TURB STOP VLV #2 A1 CHANNEL 2-FU1-1-78A 2-RLY-099-05AK10E 9-15 2-730E915-9 NONE BOTH CIRCUITS FOR ANY SINGLE TURB 2-FCV-1-78 (5AF10E) 2-RLY-099-05AK10J 2-45E763-9(RPT) STOP VLV CAN BE DE-ENERGIZED AND B2 CHANNEL 2-FU1-1-78B 2-RLY-099-05AK10D 9-17 2-730E915-10 NO 1/2 SCRAM OR RPT LOGIC ACTUATION Function: 8 (5AF10D) WILL OCCUR. TURB STOP VLV #3 A2 CHANNEL 2-FU1-1-84A 2-RLY-099-05AK10C 9-15 2-730E915-9 NONE BOTH CIRCUITS FOR ANY SINGLE TURB 2-FCV-1-84 (5AF10C) 2-RLY-099-05AK10L 2-45E763-10(RPT) STOP VLV CAN BE DE-ENERGIZED AND B1 CHANNEL 2-FU1-1-84B 2-RLY-099-05AK10F 9-17 2-730E915-10 , NO 1/2 SCRAM OR RPT LOGIC ACTUATION Function: 8 (5AF10F) WILL OCCUR. TURB STOP VLV#4 -A2 CHANNEL 2-FU1-1-88A 2-RLY-099-05AK10G 9-15 2-730E915-9 NONE BOTH CIRCUITS FOR ANY SINGLE TURB ~ 2-FCV-1-88 (5AF10G) I

                       - B2 CHANNEL 2-FU1-1-88B       2-RLY-099-05AK10H      9-17     2-730E915-10 STOPVLV CAN BE DE-ENERGIZED AND NO 1/2 SCRAM OR RPT LOGIC ACTUATION Function: 8        . (5AF10H)                    2-RLY-099-05AK10M               2-45E763-10(RPT)             WILL OCCUR.

NOTE: Device FunctioncQr~gnds tothe TS Table 3.3.1.1 Functions.

BFN Reactor Protection System 2-01-99 Unit2 Rev. 0073 Page 71 of 77 Illustration 3 (Page 5 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-PIS-3-22AA 2-FU1-3-22AA 2-RLY-099-05AK05A 9-15 2-730E915-9 2-XA-55-4A-9 ALARMS AND 1/2 SCRAM IN A CHANNEL RX HIGH PRESS (5AF5A) 2-45E671-26 RX VESSEL PRESSURE HIGH A1 CHANNEL HALF SCRAM 2-XA-55-5B-1 REACTOR CHANNEL A AUTO Function: 3 SCRAM 2-PIS-3-22BB 2-FU1-3-22BA 2-RLY-099-05AK05B 9-17 2-730E915-10 2-XA-55-4A-9 ALARMS AND 1/2 SCRAM IN B CHANNEL RX HIGH PRESS (5AF5B) 2-45E671-38 RX VESSEL PRESSURE HIGH B1 CHANNEL HALF SCRAM 2-XA-55-5B-2 REACTOR CHANNEL B AUTO Function: 3 SCRAM 2-PIS-3-22C 2-FU1-3-22CA 2-RLY-099-05AK05C 9-15 2-730E915-9 2-XA-55-4A-9 ALARMS AND 1/2 SCRAM IN A CHANNEL RX HIGH PRESS (5AF5C) 2-45E671-32 RX VESSEL PRESSURE HIGH A2CHANNEL HALF SCRAM 2-XA-55-5B-1 REACTOR CHANNEL A AUTO Function: 3 SCRAM 2-PIS-3-22D 2-FU1-3-22DA 2-RLY-099-05AK05D 9-17 2-730E915-10 2-XA-55-4A-9 ALARMS AND 1/2 SCRAM IN B CHANNEL RX HIGH PRESS (5AF5D) 2-45E671-44 RX VESSEL PRESSURE HIGH B2 CHANNEL HALF SCRAM 2-XA-55-5B-2 REACTOR CHANNEL B AUTO Function: 3 SCRAM NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

BFN .. Reactor Protection System 2-01-99 Unit2 Rev. 0073 Page 72 of 77 Illustration 3 (Page 6 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-LlS-3-203A 2-FU 1-3-203AA 2-RLY-099-05AK06A 9-15 2-730E915-9 ALARMS AND 1/2 SCRAM IN CHANNEL A 2-XA-55-4A-2 RXWATER (5AF6A) 2-RLY-099-5A-K25A 2-730E927-7 NO PCIS DEVICES ACTUATE. 2-RLY-064-16AK5A 2-45E671-26 RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK6A HALF SCRAM (Level 3) 2-XA-55-58-1 A1CHANNEL REACTOR CHANNEL A AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-2038 2-FU1-3-2038A 2-RLY-099-05AK068 9-17 2-730E915-10 ALARMS AND 1/2 SCRAM IN CHANNEL 8 2-RLY-099-5A-K258 2-730E927-8 2-XA-55-4A-2 NO PCIS DEVICES ACTUATE. RXWATER (5AF68) 2-45E671-38 RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK58 2-RLY-064-16AK68 HALF SCRAM (Level 3) 2-XA-55-582 81 CHANNEL REACTOR CHANNEL 8 AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-203C 2-FU 1-3-203CA 2-RLY-099-05AK06C 9-15 2-730E915-9 ALARMS AND 1/2 SCRAM IN CHANNEL A 2-730E927-7 2-XA-55-4A-2 NO PCIS DEVICES ACTUATE. RXWATER (5AF6C) 2-RLY-099-5A-K25C RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK5C 2-45E671-32 HALF SCRAM (Level 3) 2-RLY-064-16AK6C 2-XA-55-58-1 A2CHANNEL REACTOR CHANNEL A AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-203D 2-FU 1-3-203DA 2-RLY-099-05AK06D 9-17 2-730E915-10 ALARMS AND 1/2 SCRAM IN CHANNEL 8 2-XA-55-4A-2 RXWATER (5AF6D) 2-RLY-099-5A-K25D 2-730E927-8 NO PC IS DEVICES ACTUATE. 2-45E671-44 RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK5D HALF SCRAM I (Level 3) 2-RLY-064-16AK6D 2-XA-55-582 82 CHANNEL REACTOR CHANNEL 8 AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

BFN Reactor Protection System 2-01-99 Unit 2 Rev. 0073 Page 73 of 77 Illustration 3 (Page 7 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-PS-47-142 2-FU1-1-75CA 2-RLY-099-05AKOSA 9-15 2-730E915-9 2-XA-55-4A-15 ALARMS AND IF >30% POWER, 1/2 SCRAM 2-FCV-1-75 (5AFSA) 2-RLY-099-05AKOSE 2-45E763-9 TURB CONTROL VLV FAST CHANNEL A CONTROL CLOSURE HALF SCRAM 1/2 LOGIC PICKED UP IN RPT DIV I, VLVFAST BUT CLOSURE 2-XA-55-5B-1 NO TRIP UNLESS PS-47-144(2-FCV-1-S0) A1 Channel REACTOR CHANNEL A ALSO PICKED UP. Function: 9 AUTO SCRAM 2-PS-47-144 2-FU1-1-S0CA 2-RLY-099-05AKOSB 9-17 2-730E915-10 2-XA-55-4A-15 ALARMS AND IF >30% POWER, 1/2 SCRAM 2-FCV-1-80 (5AFSB) 2-RLY-099-05AKOSF 2-45E763-9 TURB CONTROL VLV FAST CHANNEL B 1/2 LOGIC PICKED UP IN RPT CONTROL CLOSURE HALF SCRAM DIVI, VLVFAST BUT CLOSURE 2-XA-55-5B-2 NO TRIP UNLESS PS-47-142(2-FCV-1-75) B1 Channel REACTOR CHANNEL B ALSO PICKED UP. AUTO SCRAM Function: 9 2-PS-47-146 2-FU1-1-S5CA 2-RLY-099-05AKOSC 9-15 2-730E915-9 2-XA-55-4A-15 ALARMS AND IF >30% POWER, 1/2 SCRAM 2-FCV-1-85 (5AFSC) 2-RLY-099-05AKOSG 2-45E763-10 TURB CONTROL VLV FAST CHANNEL A 1/2 LOGIC PICKED UP IN RPT CONTROL CLOSURE HALF SCRAM DIVII, VLVFAST BUT CLOSURE 2-XA-55-5B-1 NO TRIP UNLESS PS-47-14S(2-FCV-1-89) A2 Channel REACTOR CHANNEL A ALSO PICKED UP. AUTO SCRAM Function: 9 2-PS-47-14S 2-FU1-1-SSCA 2-RLY-099-05AKOSD 9-17 2-730E915-10 2-XA-55-4A-15 ALARMS AND IF >30% POWER, 1/2 SCRAM 2-FCV-1-S9 (5AFSD) 2-RLY-099-05AKOSH 2-45E763-10 TURB CONTROL VLV FAST CHANNEL B 1/2 LOGIC PICKED UP IN RPT CONTROL CLOSURE HALF SCRAM DIVII, VLVFAST BUT CLOSURE 2-XA-55-5B-2 NO TRIP UNLESS PS-47-146(2-FCV-1-S5) B2 Channel REACTOR CHANNEL B ALSO PICKED UP. AUTO SCRAM Function: 9 NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

BFN Reactor Protection System 2-01-99 Unit 2 Rev. 0073 Page 74 of 77 Illustration 3 (Page 8 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-PIS-64-56A 2-FU1-85-5A1K4A 2-RLY-099-05AK04A 9-15 2-730E915-9 2-XA-55-4A-8 ALARMS AND 1/2 SCRAM IN HI DRYWELLL PRESS (5AF4A) 2-RLY-099-5A-K24A 2-45E671-27 DRYWELL PRESSURE CHANNELA. A1 Channel 2-RLY-064-16AK5A 2-730E927-7 HIGH HALF SCRAM NO PCIS DEVICES ACTUATE. 2-XA-55-5B-1 REACTOR CHANNEL A AUTO SCRAM Function: 6 2-PIS-64-56B 2-FU 1-85-5A1K4B 2-RLY-099-05AK04B 9-17 2-730E915-10 2-XA-55-4A-8 ALARMS AND 1/2 SCRAM IN HI DRYWELLL PRESS (5AF4B) 2-RLY-099-5A-K24B 2-45E671-39 DRYWELL PRESSURE CHANNEL B. B1 Channel 2-RLY-064-16AK5B 2-730E927-8 HIGH HALF SCRAM NO PCIS DEVICES ACTUATE. 2-XA-55-5B-2 REACTOR CHANNEL B Function: 6 AUTO SCRAM 2-PIS-64-56C 2-FU1-85-5A1K4C 2-RLY-099-05AK04C 9-15 2-730E915-9 2-XA-55-4A-8 ALARMS AND 1/2 SCRAM IN HI DRYWELLL PRESS (5AF4C) 2-RLY-099-5A-K24C 2-45E671-33 DRYWELL PRESSURE CHANNEL A. A2 Channel 2-RLY-064-16AK5C 2-730E927-7 HIGH HALF SCRAM NO PCIS DEVICES ACTUATE. 2-XA-55-5B-1 REACTOR CHANNEL A AUTO SCRAM Function: 6 2-PIS-64-56D 2-FU1-85-5A1K4D 2-RLY-099-05AK04D 9-17 2-730E915-10 2-XA-55-4A-8 ALARMS AND 1/2 SCRAM IN HI DRYWELLL PRESS (5AF4D) 2-RLY-099-5A-K24D 2-45E671-45 DRYWELL PRESSURE CHANNEL B. B2 Channel 2-RLY-064-16AK5D 2-730E927-8 HIGH HALF SCRAM NO PCIS DEVICES ACTUATE. 2-XA-55-5B-2 REACTOR CHANNEL B Function: 6 AUTO SCRAM NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

BFN Reactor Protection System 2-01-99 Unit2 Rev. 0073 Page 75 of 77 . Illustration 3 (Page 9 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-LS-85-45A 2-FU1-85-45AA 2-RLY-099-05 9-15 2-730E915-9 2-XA-55-4A-1 ALARMS AND 1/2 SCRAM IN CHANNEL A. SDV HIGH LEVEL (5AF1A) AK01A WEST CRD DISCH VOL WTR A1 Channel LVL HIGH HALF SCRAM 2-XA-55-5B-1 REACTOR CHANNEL A Function: 7a AUTO SCRAM 2-LS-85-45B 2-FU1-85-45BA 2-RLY-099-05 9-17 2-730E915-10 2-XA-55-4A-1 ALARMS AND 1/2 SCRAM IN CHANNEL B. SDV HIGH LEVEL (5AF1B) AK01B WEST CRD DISCH VOL WTR B1 Channel LVL HIGH HALF SCRAM 2-XA-55-5B-2 REACTOR CHANNEL B Function: 7a AUTO SCRAM NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions.

BFN Reactor Protection System 2-01-99 Unit 2 Rev. 0073 Page 76 of 77 Illustration 3 (Page 10 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-LS-85-45C 2-FU1-85-45CA 2-RLY-099-05AK1 C 9-15 2-730E915-9 2-XA-55-4A-1 ALARMS AND 1/2 SCRAM IN CHANNEL A. SDV HIGH LEVEL (5AF1C) WEST CRD DISCH VOL WTR A2 Channel LVL HIGH HALF SCRAM 2-XA-55-58-1 REACTOR CHANNEL A AUTO Function: 7b SCRAM 2-LS-85-45D 2-FU1-85-45DA 2-RLY-099-05AK01D 9-17 2-730E915-10 2-XA-55-4A-1 ALARMS AND 1/2 SCRAM IN CHANNEL 8. SDV HIGH LEVEL (5AF1 D) WEST CRD DISCH VOL WTR 82 Channel LVL HIGH HALF SCRAM 2-XA-55-58-2 REACTOR CHANNEL 8 AUTO Function: 7b SCRAM 2-LS-85-45E 2-FU1-85-45EA 71X-85-45E 9-15 2-730E915-11 2-XA-55-4A-29 ALARMS AND 1/2 SCRAM IN CHANNEL A. SDV HIGH LEVEL EAST CRD DISCH VOL WTR A1 Channel LVL HIGH HALF SCRAM 2-XA-55-58-1 REACTOR CHANNEL A AUTO SCRAM Function: 7b 2-LS-85-45F 2-FU1-85-45FA 71X-85-45F 9-17 2-730E915-12 2-XA-55-4A-29 ALARMS AND 1/2 SCRAM IN CHANNEL 8. SDV HIGH LEVEL EAST CRD DISCH VOL WTR 81 Channel LVL HIGH HALF SCRAM 2-XA-55-58-2 REACTOR CHANNEL 8 AUTO Function: 7b SCRAM

BFN" Reactor Protection System 2-01-99 Unit 2 Rev. 0073 Page 77 of 77 Illustration 3 (Page 11 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-LS-85-45G 2-FU 1-85-45GA 71X-85-45G3 9-15 2-730E915-11 2-XA-55-4A-29 ALARMS AND 1/2 SCRAM IN CHANNEL A. SDV HIGH LEVEL EAST CRD DISCH VOL WfR A2 Channel LVL HIGH HALF SCRAM 2-XA-55-5B-1 REACTOR CHANNEL A AUTO SCRAM Function: 7a 2-LS-85-45H 2-FU1-85-45HA 71X-85-45H3 9-17 2-730E915-12 2-XA-55-4A-29 ALARMS AND 1/2 SCRAM IN CHANNEL B. SDV HIGH LEVEL EAST CRD DISCH VOL WfR B2 Channel LVL HIGH HALF SCRAM 2-XA-55-5B-2 REACTOR CHANNEL B AUTO Function: 7a SCRAM NOTE: Device Function corresponds to the TS Table 3.3.1.1 Functions. NUCLEAR TRIP INSERT REMARKS INSTRUMENTATION IRMS PLACE MODE SWITCH FOR AFFECTED DRAWER REACTOR MODE SWITCH NOT IN RUN: IN POSITION OTHER THAN OPERATE IRMS RESULTS IN HALF SCRAM OF RPS A (ACEG) OR RPS B (BDFH). REACTOR MODE SWITCH IN RUN: WILL NOT RESULT IN A HALF SCRAM. Function: 1 APRMs PLACE KEYLOCK SWITCH FOR AFFECTED ONE CHANNEL WILL RESULT IN AN APRM HIGHIINOP AND A ROD BLOCK. NO RPS CHANNEL IN "INOP" POSITION OUTPUT FROM THE 2/4 VOTERS UNTIL 2 OF THESE CONDITIONS ARE RECEIVED, THEN THE 214 VOTERS WILL INITIATE A FULL SCRAM .. Function: 2 NOTE: Nuclear Instrumentation Function corresponds to the TS Table 3.3.1.1 Functions.

0610 NRC SRO Exam

16. SRO 268000A2.0 1 0021elAlT2G210PL 171.08411268000A2.0 l//SRO ONLY/NEW 2117/08 Given the following plant conditions:

BFN is in the process of discharging the Waste Sample Tank to the river in accordance with an approved Discharge Permit.
The discharge has been in progress for 90 minutes when Security reports that water is bubbling up from the ground in the vacinity of the SGT Building.
0-RR-90-130 (Radwaste Effluent Radiation Monitor) is currently reading the same as the initial background radiation level prior to commencing the discharge.

Which ONE of the following describes the appropriate action and the basis for this action? Terminate the discharge and (1) . Enter procedure (2) in order to determine _____(~3".J..)_ _ _ _ __ A. (1) have Radcon survey the area of the leak. (2) EPIP-13. (3) an Offsite Dose Assessment. B. (1) have Chemistry sample the spilled water.. (2) EPIP-13. (3) an Offsite Dose Assessment. C. (1) have Radcon survey the area of the leak. (2) EPIP-1. (3) if the Effluent Concentration Limits have been exceeded. D~ (1) have Chemistry sample the spilled water. (2) EPIP-1. (3) if the Effluent Concentration Limits have been exceeded. Sunday, February 17, 20084:52:33 PM 44

0610 NRC SRO Exam KIA Statement: 268000 Radwaste A2.01 - Ability to (a) predict the impacts of the following on the RADWASTE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System rupture. KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate knowledge of the consequences of a rupture affecting the Radwaste system.

References:

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. 0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: Without references, determine the appropriate notification requirements, procedure entry and basis for a rupture in the radwaste system. D is correct. A is incorrect. A radcon survey is not required in accordance with EPIP-13 until directed by either EPIP-2 through EPIP-5. B is incorrect. A chemistry sample is appripriate but entrance into EPIP-13 is not required until an Emergency Classification has been entered. This may or may not occur depending on the results of the chemistry sample. C is incorrect. A radcon survey is not required in accordance with EPIP-13 until directed by either EPIP-2 through EPIP-5. In addition, determination of ECls is by chemistry sample, not radiation surveys. Sunday, February 17, 20084:52:33 PM 45

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE TECHNICAL BASIS EPIP-1 ( UNUSUAL EVENT

EAL: liquid release rate exceeds 20 times ECl as determined by chemistry sample AND Release duration exceeds or will exceed 60 minutes.

OPERATING CONDITION: All BASIS: Liquid release rates are determined using Surveillance Instructions which utilize liquid samples rather than instrument readings for activity determination. Effluent Concentration Limits (ECl) are those annual concentrations given in 10CFR20 Appendix B, Table 2, Column 2. 10 times ECl is equivalent to the instantaneous ODCM limit. Unplanned radioactivity releases that exceed 20 times ECl (2 times ODCM limit) and continue for 60 minutes or longer represent an uncontrolled situation and potential degradation in the level of safety of the plant. The release should not be averaged over 60 minutes. For example, a release of 40 times ECl for 30 minutes does not meet the requirements of this event classification. The 60 minute time period allows sufficient time to isolate any release after exceeding ECL. Greater than 60 minutes represents inability to isolate or control the release. The Site Emergency Director should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. The Chemistry Department ( determines the magnitude of the release by sample procedure for any release as required by initiating procedures (Le., SI, ARP, AOI, EOI). The sample results are reported to the Site Emergency Director as a fraction or multiple of ECL. Escalation to Alert is based on release in excess of 2000 times ECl for greater than 15 minutes.

REFERENCES:

Reg Guide 1.101 Rev. 3, (NUMARC-AU1 example-2) EDMS l63 010206 800 10CFR20 NOTES: CURVESITABLES: PAGE 142 OF 201 REVISION 42

I BROWNS FERRY DOSE ASSESSMENT EPIP-13

1.0 INTRODUCTION

1.1 Purpose The purpose of this procedure is to describe actions and responsibilities of Radiation Protection (RP) personnel during an assessment of environmental radiological conditions at Browns Ferry. EPIP-13 will be initiated when the RP Shift Supervisor or designee requests or requires information regarding dose assessment. EPIP-13 contains instructions for RP regarding methods for Projecting Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (CDE) from airborne radioactivity releases. The method for projecting TEDE and/or CDE from airborne radioactivity releases may be requested by Operations to support the emergency classification process and/or protective action recommendations. The use of this method should only be utilized in the absence of more sophisticated dose models, when the Central Emergency Control Center (CECC) is not activated. This procedure will normally be initiated by way of EPIP-1, "Emergency Classification Procedure".

2.0 REFERENCES

2.1 Industry Documents A. NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" B. 10 CFR 50.47, Code of Federal Regulations C. 10 CFR 72.75, Code of Federal Regulations D. Central Emergency Control Center Emergency Plan Implementing Procedure - 8 2.2 Plant Instructions A. TVA Radiological Emergency Plan

8. EPIP - 1, "Emergency Classification Procedure" C. EPIP - 2, "Notification of Unusual Event" D. EPIP - 4, "Site Area Emergency" E. EPIP - 5, "General Emergency" PAGE 1 OF 17 REVISION 0013

0610 NRC SRO Exam

17. SRO 271000G2.4.36 001lCIA/T2G2/EPIP-1I!271000G2.4.361ISRO ONLYINEW 12/3/07 RMS Given the following plant conditions:

A transient has occurred on Unit 1 resulting in the following annunciators in alarm:

                -   STACK GAS RADIATION HI (1-RA-90-147B)
                -   STACK GAS RADIATION HIGH-HIGH (1-RA-90-147A)
                -   OG PRETREATMENT RADIATION HIGH (1-RA-90-157A)
                -   RX BLDG,TURB BLDG, RF ZONE EXH RADIATION HIGH (1-RA-90-250A)

Which ONE of the following describes the required operator action? Declare a/an _ _----1,..;(1.....) ___ and notify _ _=(2"...)__ to implement Emergency Plan Implementing Procedure, EPIP-13, for dose assessment. REFERENCE PROVIDED (1) (2) A. Notification of Unusual Event; CECC B. Alert; CECC C!' Notification of Unusual Event; Radco~ D. Alert; Radcon Sunday, February 17, 2008 3:28:27 PM 46

0610 NRC SRO Exam KIA Statement: 271000 Off-gas 2.4.36 - Emergency Procedures I Plan Knowledge of chemistry I health physics tasks during emergency operations. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine that Chemistry and/or Radcon support is required.

References:

ARPs for listed annunciators Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. 0610 NRC SRO Exam REFERENCE PROVIDED: EPIP-1 Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: f: Recognize that a valid Off-Gas pretreatment radiation high alarm requires an Unusual Event declaration in accordance with EAL 1.4-U.

2. Recognize that the Central Emergency Control Center (CECC) is not manned during a NOUE, therefore Radcon will implement EPIP-13 for dose assessment.

C* - correct: A.'- incorrect: The classification is correct however, the CECC will not be manned until an ALERT or higher classification is determined. B - incorrect: Conditions are not met to upgrade to an ALERT classification and the CECC will not be manned until that point. D - incorrect: Conditions are not met to upgrade to an ALERT classification however, Radcon IS responsible for implementation of EPIP-13 during a NOUE. Sunday, February 17, 2008 3:28:27 PM 47

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A. OPERATING CONDITION: Mode 1 or 2 or 3 Reactor moderator temperature can NOT be maintained below 2120 F whenever Technical Specifications require Mode 4 conditions or during

                                                                                                 ;Do operations in Mode 5.                                    r m
                                                                                              .*.::q.

OPERATING CONDITION: Mode 4 or 5 OPERATING CONDITION: Mode 1 or 2 or 3 o m z m>

                                                                                                 ......
                                                                                                 =0 .
                                                                                                 >>r-m S

m

~.

m z o

-<

PAGE 22 OF 201 REVISION 42

IBROWNS FERRY DOSE ASSESSMENT EPIP-13

1.0 INTRODUCTION

1.1 Purpose The purpose of this procedure is to describe actions and responsibilities of Radiation Protection (RP) personnel during an assessment of environmental radiological conditions at Browns Ferry. EPIP-13 will be initiated when the RP Shift Supervisor or designee requests or requires information regarding dose assessment. EPIP-13 contains instructions for RP regarding methods for Projecting Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (CDE) from airborne radioactivity releases. The method for projecting TEDE and/or CDE from airborne radioactivity releases may be requested by Operations to support the emergency classification process and/or protective action recommendations. The use of this method should only be utilized in the absence of more sophisticated dose models, when the Central Emergency Control Center (CECC) is not activated. This procedure will normally be initiated by way of EPIP-1, "Emergency Classification Procedure".

2.0 REFERENCES

2.1 Industry Documents A. NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" B. 10 CFR 50.47, Code of Federal Regulations C. 10 CFR 72.75, Code of Federal Regulations D. Central Emergency Control Center Emergency Plan Implementing Procedure - 8 2.2 Plant Instructions A. TVA Radiological Emergency Plan B. EPIP - 1, "Emergency Classification Procedure" C. EPIP - 2, "Notification of Unusual Event" D. EPIP - 4, "Site Area Emergency" E. EPIP - 5, "General Emergency" PAGE 1 OF 17 REVISION 0013

(REFERENCES PROVIDED TO ( CANDIDATE

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 42 PREPARED BY: RANDY WALDREP PHONE: 2038 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: TONY ELMS DATE: 04/06/2007 EFFECTIVE DATE: 04/06/2007 LEVEL OF USE: REFERENCE USE QUALITY-RELATED

0610 NRC SRO Exam

18. SRO 288000A2.03 001lCIA/T2G2/0I-3011288000A2.031ISRO ONLYINEW 12/15/2007 RMS Given the following Unit 3 conditions:

(

Unit 3 was at 100% power
A Loss of Coolant Accident occurred resulting in the following plant indications:
Reactor water level is (-)130 inches and steady with RCIC injecting.
Reactor pressure is 750 psig and lowering slowly.
Drywell pressure is 5.0 psig and rising slowly.
Reactor Zone exhaust radiation is 65 mR/hr.
Refuel Zone exhaust radiation is 4 -mR/hr.
SGT trains 'A', 'B' and 'C' are running.

Which ONE of the following describes the status of Reactor and Refuel Zone ventilation and the corrective actions required for these conditions? Reactor and Refuel Zone ventilation systems are (1) Performing 3-EOI Appendix 8E and restarting ventilation per 3-EOI Appendix &F _ _. .>.:(2: . .<)_ _ A'!' isolated. is required to maintain the availability of the RCIC system. B. isolated. is NOT required to maintain the availability of the main condenser. C. NOT isolated. is required to maintain the availability of the RCIC system. D. NOT isolated. is NOT required to maintain the availability of the main condenser.

0'1\

~\

Sunday, February 17, 2008 3:28:27 PM 48

0610 NRC SRO Exam KIA Statement: 288000 Plant Ventilation A2.03 - Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accident: Plant-Specific.

      'KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine ventilation system status and take the appropriate corrective actions.

References:

3-01-30A and B. 3-EOI Appendix 8E & F Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 49

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: Whether or not a Group 6 isolation has occurred based on current conditions. If an isolation has occurred, whether performing EOI Appendicies 8E and 8F is required to restore Reactor and Refuel Zone ventilation and the basis for that conclusion. If an isolation has NOT occurred, whether performing EOI Appendicies 8E and 8F will maintain Reactor and Refuel Zone ventilation in service and the basis for that conclusion. A - correct: B - incorrect: Ventilation has isolated on a Group 6 isolation. However, performing EOI Appendicies 8E and 8F ARE required to prevent a Group I isolation signal due to steam tunnel high temperature, thus preventing the main condenser from being utilized as the primary heat sink. C - incorrect: Although ventilation has isolated on a Group 6 isolation signal, performing EOI Appendicies 8E and 8F would be appropriate to maintain RCIC availability due to high temperature in the reactor building had the Group 6 isolation NOT occurred. D - incorrect: Although ventilation has isolated on a Group 6 isolation signal, performing EOI Appendicies 8E and 8F would be appropriate to prevent a Group I isolation signal due to steam tunnel high temperature, thus preventing the main condenser from being utilized as the primary heat sink. Sunday, February 17, 2008 3:28:27 PM 50

EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION ON-E I DISCUSSION: PURPOSE: EOI-3 The purpose of EOI-3, Secondary Containment Control, is to protect equipment in secondary containment, limit I radioactivity release to secondary containment, and either: 1) maintain secondary containment integrity, or 2) limit radioactivity release from secondary containment. The purpose of this procedure relates directly to the basic functions perfonned by secondary containment structures:

1. Containing fission products that may leak from primary containment.

2. Minimizing ground level releases of airborne radioactive material by discharge through an elevated release point.

3. Shielding personnel from radiation that penetrates primary containment.

4. Providing a protected environment for key equipment important to safety.

The procedure's purpose is accomplished through concurrent control of secondary containment temperatures, water levels, and radiation levels. (~ *

REVISION 1 PAGE 7 OF 73 SECTION O-V-E

                                                                                                   .. * *...* - - . - -_ _ _ _ _ _ _ _ _-11

EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-E ~ ......D.I.SC.U.S.S.IO . N.:

                                 . . .S.C.C.-l. . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..

This retainment override step applies throughout performance ofEOI-3, Secondary Containment Control. The first and second retainment override statements direct the operator to verify both isolation of Reactor Zone and/or Refuel Zone Ventilation, and initiation of SGTS, when high ventilation exhaust radiation isolation setpoint for the respective system has been reached. Confirming isolation of Reactor Zone and/or Refuel Zone Ventilation after receipt ofa high radiation isolation signal halts any further release of radioactivity to the environment from these systems. SGTS is normally used

  .under post-accident conditions to maintain secondary containment pressure negative with respect to outside atmosphere, since exhaust from SGTS is processed before being discharged to the environment through an elevated release point.

The third and fourth retainment override statements direct the operator to restart Reactor Zone and/or Refuel Zone Ventilation Systems once it has been confirmed that excessive radioactivity release will not occur from system operation. Reactor Zone and Refuel Zone Ventilation Systems are normally used to maintain secondary containment temperature and differential pressure within operational limits. Once it has been confirmed that restart will not result in an excessive release of radioactivity to the environment, it is appropriate to restart these systems and use them to restore and maintain secondary containment temperature and pressure. Authorization and step-by-step guidance is provided in EOI Appendix 8E to defeat high drywell pressure and low RPV water level Group 6 isolation interlocks for these systems, if necessary, since these isolations serve to limit radioactivity release to the environment, and become dispensable once it is assured that excessive release of radioactivity will not occur. EOI Appendix 8F provides guidance to restore Reactor Zone Ventilation following a Group 6 isolation.

REVISION 1 PAGE 13 OF 73 SECTION O-V-E

CS, LEVEUPOWER CONTROL BASES EOI PROGRAM MANUAL

SECTION O-V-K '

I DISCUSSION: STEP C5-9 I This action step directs the operator to bypass any low RPV water level isolation interlock that may cause the MSIVs to close. System specific EOI Appendices provide step-by-step guidance for bypassing isolation interlocks.

Subsequent actions in this procedure may lower RPV water level to or below the low RPV water level isolation setpoint for MSIVs, when conditions are most desirable to maintain use of the main condenser as the heat sink. The MSIV low-low-low RPV water level isolation interlock is bypassed using the guidance in EOI Appendix 8A to ensure that the main condenser remains available as a heat sink ifRPV water level is lowered past the MSIV isolation setpoint. The reactor building ventilation low RPV water level isolation interlock is bypassed using the guidance in EOI Appendix 8E because reactor building ventilation is necessary for maintaining steam tunnel temperatures below their isolation setpoint. If the steam tunnel temperature exceeds the MSIV isolation setpoint, the main condenser is lost as a heat sink. The drywell control air low RPV water level isolation interlock is bypassed using the guidance in EOI Appendix 8e to ensure that adequate pneumatic pressure is available to maintain open the inboard MSIVs. If drywell control air is lost, the inboard MSIVs may inadvertently close. Other MSIV isolation interlocks (i.e., main steam high flow and steam tunnel high temperature) are not bypassed. These isolations are required to provide automatic protection for conditions where preventing the closure of the MSIVs is not appropriate. Maintaining the MSNs open is conditioned as follows:

1. The main condenser must remain available, since the only reason for maintaining the MSIV s open is to utilize the main condenser as the heat sink.

2. There must be no indication of "gross" fuel failure, since maintaining the MSIVs open with grossly failed fuel could result in a significant release of fission products to the envirorunent. "Gross" fuel failure is specified here to distinguish from small cladding leaks.

The judgment is subjective, based on operator assessment of all available indications.

3. There must be no indication of a steam line break. Maintaining the MSIVs open with a break in the downstream piping could result in an uncontrolled loss of reactor coolant

inventory, release fission products to the environment, and cause personnel injury or significant damage to plant equipment.

REVISION 0 PAGE 23 OF 110 SECTION O-V-K

                            .... _._. __ .__ ... _ - - - - - - - - - - - - - - - - - - - - -...

0610 NRC SRO Exam

19. SRO GENERIC 2.1.12 001lC/A/T3/RPSI21212000G2.1.1212.9/4.0/SRO ONLYINEW 113112008 Unit 1 has just completed a shutdown for a forced maintenance outage, with the following plant conditions:

Recirc Pump '1 A' has been removed from service in preparation for Shutdown Cooling to be placed in service.
RPV pressure is 40 psig and ALL other parameters are normal.

The Unit Operator reports that the Recirc Pump '1A' Discharge Valve will not close. Which ONE of the following describes the required Technical Specification and operational actions? Declare _--,(....;..1.....

                            ) __ inoperable per LCO 3.5.1. Additionally, direct the Unit Operator to close Recirc Pump '1A' Suction Valve and place _---->(=2J....)__ in Shutdown Cooling.

(1) A. RHR Loop I; RHR Loop I o U B ..... RHR Loop II; RHR Loop II (U4 hftL c C. RHR Loop I; RHR Loop II D. RHR Loop II; RHR Loop I KIA Statement: Conduct of Operations 2.1.12 Ability to apply technical specifications for a system. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to correctly determine the required actions per Technical Specifications.

References:

Unit :rTech Specs Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 51

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: Based on current plant conditions, Unit-1 is in Mode 3. In addition, the candidate must recall from memory that the bases for Tech Spec 3.5.1 requires the discharge valve on the associated Recirc Loop to be operable for the RHR Loop to be operable for LPCI injection. Given that condition, recognize that RHR Loop II injects into Recirc Loop "A", therefore, RHR Loop II is inoperable for LPCI injection. Furthermore, recall from memory that aligning RHR for Shutdown Cooling will also make that RHR Loop inoperable for LPCI injection. Based on that determination, aligning RHR Loop II for Shutdown Cooling is the only acceptable alternative to preserve an operable RHR Loop for LPCI injection capability until Mode 4 is achieved. B - correct: A - incorrect: On Unit-1, RHR Loop II injects into Recic Loop "A" for LPCI injection, therefore RHR Loop II is INOP, not RHR Loop I. However, if the first part of answer "A" were correct, aligning RHR Loop I for SO Cooling would be appropriate .. C - incorrect: On Unit-1, RHR Loop II injects into Recic Loop "A" for LPCI injection, therefore RHR Loop II is INOP, not RHR Loop I. However, aligning RHR Loop II for Shutdown Cooling would be appropriate. D - incorrect: RHR Loop II is inoperable per TS 3.5.1, however, placing RHR Loop I in SO Cooling would make BOTH RHR Loops inoperable for LPCI injection. Sunday, February 17, 2008 3:28:27 PM 52

ECCS - Operating B 3.5.1 BASES BACKGROUND at 0.2 seconds when offsite power is available and B, C, and D (continued) pumps approximately 7, 14, and 21 seconds afterwards and if offsite power is not available all pumps 7 seconds after diesel generator power is available). When the RPV pressure drops sufficiently, CS System flow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV. LPCI is an independent operating mode of the RHR System. There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop. The two LPCI pumps and associated motor operated valves in each LPCI subsystem are powered from separate 4 kV shutdown boards. Both pumps in a LPCI subsystem inject water into the reactor vessel through a common inboard

injection valve and de..:: i II 1111: I.

                        ,e          subsystem's common inboard injection valve and recirculation pump discharge valve are powered from one of the two 4 kV shutdown boards associated with that subsystem.

ECCS Preferred Pump Logic In the event of a spurious accident Signal in Unit 2 combined with a real accident in Unit 1 (or a spurious accident signal in Unit 1 with a real accident in Unit 2) the CS and LPCI preferred pump logic will dedicate the Division I CS and LPCI pumps to Unit 1 (1A and 1C) and the Division II pumps to Unit 2 (2B and 2D). Therefore, a spurious accident signal from Unit 2 (which is considered a single failure) results in two RHR pumps and one CS loop (two CS pumps) OPERABLE for Unit 1 (refer to Bases Sections B 3.3.5.1 and B 3.8.1). This is acceptable in (continued) BFN-UNIT 1 B 3.5-3 Revision 0, 33, 47 March 22, 2007

0610 NRC SRO Exam

20. SRO GENERIC 2.2.22 OOllMEM/TECH SPECS/HPCIIIG2.2.22/4.lISROINEW 113112008 Unit 1 was operating at 100% power when a catastrophic failure of the EHC system occurs. Post trip reviews of the transient reveal a pressure spike as follows:

(

St~a_m Dome 13gQJ2§!g
Bottom head::-~r330

                                  '-=--.. . . . . ,-----~,-,.~

psig~

                                                                ,""~~- -.-

Determine if the RPV Safety Limit was violated and the basis for the conclusion. The RPV Safety Limit was _ _-->.-{1;....<.)_ _ , because the Safety Limit is provided for the (2) (1 ) A. violated; Steam Dome B:-t NOT violated; Steam Dome C. violated; entire Pressure boundary D. NOT violated; entire Pressure boundary KIA Statement: Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine if a safety limit was violated and the basis for the limit.

References:

Unit-1 Tech Specs Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of information and apply this to the complicance of technical specification. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:27 PM 53

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must: Recall from memory that the RPV Pressure safety limit is 1325 psig as measured in the steam dome of the reactor vessel. B - correct: A - incorrect: The safety limit of 1325 psig was not violated in the steam dome. C - incorrect: The safety limit of 1325 psig was not violated because the height of water and the corresponding pressure associated with it is already considered in the determination of the limit of 1325 psig. D - incorrect: Although the limit was not exceeded, the height of water and the corresponding pressure associated with it is already considered in the determination of the limit of 1325 psig. Sunday, February 17, 2008 3:28:27 PM 54

RCS Pressure SL B 2.1.2 BASES (continued) APPLICABLE The RCS safety/relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure - High Function have settings established to ensure that the RCS pressure SL will not be exceeded. The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1965 Edition, including Addenda through the summer of 1965 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1, 1967 Edition (Ref. 6), and the additional requirements of GE design and procurement specifications (Ref. 7) which were implemented in lieu of the outdated B31 Nuclear Code Cases - N2, N7, N9, and N10, for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes. SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping. (continued) BFN-UNIT 1 B 2.0-9 Revision 0

0610 NRC SRO Exam

21. SRO GENERIC 2.2.24 00l/C/A/T3/23//2/GEN2.2.24/2.5/3.7/SROIBANK 10127/07 Given the following plant conditions:

Unit 2 is 100% RTP.
Unit 1 is in Mode 5, with initial fuel loading in progress.
Unit 3 is in Mode 5, Control Rod Drive replacement in progress after first series of fuel moves.

                  *  'B1' RHRSW pump is tagged for impeller replacement.
                  *  '2C' RHR Heat Exchanger tagged for eddy current testing.

The Outside Auxiliary Unit Operator (AUO) reports that BOTH Sump Pumps in the RHRSW Pump have failed to start. Initial troubleshooting reveals a shorted motor on BOTH Sump Pumps.

Which ONE of the following describes the minimum required actions imposed by Tech

   .' Specs?

Restore Suppression Pool Cooling, Suppression Chamber Sprays, and Drywell sprays in (1) . Restore RHRSW system and Ultimate Heat Sink in (2) REFERENCE PROVIDED

(1) (2) A. 7 days. 8 hours. B ..... 7 days. 30 days. QIL-C. 30 days. 8 hours. D. 30 days. 30 days. Sunday, February 17, 2008 3:28:28 PM 55

0610 NRC SRO Exam KIA Statement: Equipment Control 2.2.24 Ability to analyze the affect of maintenance activities on LCO status. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine the LCO status resulting from maintenance activities.

References:

U2 TSR Section 3.6 and 3.7 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam REFERENCE PROVIDED: U2 Tech Spec Section 3.6 and 3.7 Plausibility Analysis: In order to answer this question correctly the candidate must know the following:

       - The '0' RHRSW Pump Room sumps result in '01', '02' and '03' RHRSW Pumps being inoperable per Tech Spec Bases.
       - RHRSW flow through the RHR HX is required for SP Cooling, Spray and OW Spray to be operable per Tech Spec Bases. This leads to 2 RHRSW subsystems inoperable for Unit 2.
       - Which RHRSW pumps provide flow to each RHRHX. (system knowledge)
       - Oetermine the LCO for RHRSW pumps, RHRSW subsystem, SP Cooling, SP Spray, and OW Spray.

B - correct: A - incorrect: The 8-hour LCO for RHRSW system applies if 2 units are fueled. (six pumps required) Initial conditions state that ONLY one unit is fueled. C - incorrect: A 30-day LCO is required if ONLY one RHRSW subsystem is inoperable. In addition, the 8-hour LCO for RHRSW system applies if 2 units are fueled. (six pumps required) Initial conditions state that ONLY one unit is fueled. D - incorrect: A 30-day LCO is required if ONLY one RHRSW subsystem is inoperable. Sunday, February 17, 2008 3:28:28 PM 56

. - (REFERENCES

   . PROVIDED TO

( CANDIDATE

RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO 3.6.2.3 Four RHR suppression pool cooling subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore the RHR 30 days pool cooling subsystem suppression pool cooling inoperable. subsystem to OPERABLE status. B. Two RHR suppression B.1 Restore one RHR 7 days pool cooling subsystems suppression pool cooling inoperable. SUbsystem to OPERABLE status. C. Three or more RHR C.1 Restore required RHR 8 hours suppression pool cooling suppression pool cooling subsystems inoperable. subsystems to OPERABLE status. (continued) 3.6-31 Amendment No. ~272 June 8,2001

RHR Suppression Pool Cooling 3.6.2.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. D.2 Be in MODE 4. 36 hours BFN-UNIT 2 3.6-32 Amendment No. ~272 June 8, 2001

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling 31 days subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.3.2 Verify each RHR pump develops a flow rate In accordance

              ;::>: 9000 gpm through the associated heat       with the Inservice exchanger while operating in the suppression     Testing Program pool cooling mode.

BFN-UNIT 2 3.6-33 Amendment No. 253

RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray LCO 3.6.2.4 Four RHR suppression pool spray subsystems shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RH R suppression A.1 Restore the RHR 30 days pool spray subsystem suppression pool spray inoperable. subsystem to OPERABLE status. B. Two RH R suppression B.1 Restore one RHR 7 days pool spray sUbsystems suppression pool spray inoperable. subsystem to OPERABLE status. C. Three or more RHR C.1 Restore required RHR 8 hours suppression pool spray suppression pool spray subsystems inoperable. subsystems to OPERABLE status. D. Required Action and 0.1 Be in MODE 3. 12 hours associated Completion Time not met. AND 0.2 Be in MODE 4. 36 hours BFN-UNIT 2 3.6-34 Amendment No. 253

RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray 31 days subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.4.2 Verify each suppression pool spray nozzle is 5 years unobstructed. BFN-UNIT 2 3.6-35 Amendment No. 253

RH R Drywell Spray 3.6.2.5 3.6 CONTAINMENT SYSTEMS 3.6.2.5 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.5 Four RHR drywell spray subsystems shall be OPERABLE. APPUCABI L1TY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR drywell spray A.1 Restore the RHR drywell 30 days subsystem inoperable .. spray subsystem to OPERABLE status. B. Two RHR drywell spray B.1 Restore one RHR drywell 7 days sUbsystems inoperable. spray subsystem to OPERABLE status. C. Three or more RHR C.1 Restore required RHR 8 hours drywell spray subsystems drywell spray subsystems inoperable. to OPERABLE status. D. Required Action and 0.1 Be in MODE 3. 12 hours associated Completion Time not met. AND 0.2 Be in MODE 4. 36 hours BFN-UNIT 2 3.6-36 Amendment No. 253

RHR Drywell Spray 3.6.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.5.1 Verify each RHR drywell spray subsystem 31 days manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position. SR 3.6.2.5.2 Verify each drywell spray nozzle is 5 years unobstructed. BFN-UNIT 2 3.6-37 Amendment No. 253

RHRSW System and UHS 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS) LCO 3.7.1 .----------------------------------------NOTE---------------------------------------- The number of required RHRSW pumps may be reduced by one for each fueled unit that has been in MODE 4 or 5 for ~ 24 hours. Four RHRSW subsystems and UHS shall be OPERABLE with the number of OPERABLE pumps as listed below:

1. 1 unit fueled - four OPERABLE RHRSWpumps.

2. 2 units fueled - six OPERABLE RHRSW pumps.

3. 3 units fueled - eight OPERABLE RHRSW pumps.

APPLICABILITY: MODES1, 2, and 3. BFN-UNIT 2 3.7-1 Amendment No. 254 September 08, 1998

RHRSW System and UHS 3.7.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required RHRSW A.1 -------------NOTES----------- pump inoperable. 1. Only applicable for the 2 units fueled condition.

2. Only four RHRSW pumps powered from a separate 4 kV shutdown board are required to be OPERABLE if the other fueled unit has been in MODE 4 or 5 for.::: 24 hours.

Verify five RHRSW Immediately pumps powered from separate 4 kV shutdown boards are OPERABLE. OR A.2 Restore required RHRSW 30 days pump to OPERABLE status. (continued) BFN-UNIT 2 3.7-2 Amendment No. 254 September 08, 1998

RHRSW System and UHS 3.7.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One RHRSW subsystem B.1 -------------NOT E------------- inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling

                             - Hot Shutdown," for RHR shutdown cooling made inoperable by the RHRSW system.
                             ---------------------------------

Restore RHRSW 30 days subsystem to OPERABLE status. C. Two required RHRSW C.1 Restore one inoperable 7 days pumps inoperable. RHRSW pump to OPERABLE status. D. Two RHRSW subsystems 0.1 ------------- NOTE ------------- inoperable .. Enter applicable Conditions and Required Actions of LCO 3.4.7, for RHR shutdown cooling made inoperable by the RHRSW System.

                            ----------------------------------

Restore one RHRSW 7 days subsystem to OPERABLE status. (continued) BFN-UNIT 2 3.7-3 Amendment No. 254 September 08, 1998

RHRSW System and UHS 3.7.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Three or more required E.1 Restore one RHRSW 8 hours RHRSW.pumps pump to OPERABLE inoperable. status. F. Three or more RHRSW F.1 -------------N OTE ------------- subsystems inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.7 for RHR shutdown cooling made inoperable by the RHRSW System.

                               ----------------------------------

Restore one RHRSW 8 hours subsystem to OPERABLE status. G. Required Action and G.1 Be in MODE 3. 12 hours associated Completion Time not met. AND G.2 Be in MODE 4. 36 hours OR UHS inoperable BFN-UNIT 2 3.7-4 Amendment No. 254 September 08, 1998

MAR-17-2008 11:11 FROM BFN OPS TRAINING TO 9140456:24854153129 P.01 FAX COVER Send To: Name: Rick Baldwin ()ate: 3/17/2008 Company: USNRC Address: Sam Nunn Federal Building Phone: 404-562-4642 Fax Number: 404-562-4854 Verification Number. Number of pages (including cover): _4~ ____________~_ _ _ _ __

Subject:

Question for review From: Tennessee Valley Authority Name: Robert Spadoni Organization: Operations Training Address: Browns Ferry Nuclear Plant Phone: 256-729~27a3 Fax Number: VerifICation Number: comments: Sent Bv: Time Sent: _~_ _ Date Sent _ _ _ __ Important! If you do not receive all pages, call us back immediately. TVA 15536 (5-1999) [6-1999)

MAR-17-2008 11:11 FROM BFN OPS TRAINING TO 9140456:24854153129 QUESTIONS REPORT for 0610 SRO EXAM VALIDATED

1. SRO OENBRIC 2.3.3 002lMEMJT3111GENEluC 2.3.31ISRO ONLYINEW 3/17/08 A Hi*Storm has just been placed on the Independent Spent Fuel Storage Installation (ISFSI) pad and the Shift Manager is provided with the following data:

The Multi-Purpose Canister (MPC) Overpack ID number is 8FN-0-CASK-079-0100/8.
The MPC Model is 6SFF.
Helium Backfill pressure is 31.2 psig.
Helium Leak Rate is less than 2.0 E-6 atm cc/sec.
Helium purity is 9S.6%
Average surface dose rates on the top of the MPC are 2 mrem/hour.
Average surface dose rates on the side of the MPC are 12 mrem/hour.
Inlet and outlet ducts are clear and reading 14 mrem/hour.

tn a(~oordance with the Technical Specifications for the Hi-Storm 100 Cask System, Whi(~h ONE of the following describes the status of Overpack BFN-O-CASK-079-0100/S. The MPC storage conditions are _~..\-(1!..1.) _ _. According to Techni(",sl Specifications for tIle Hi-Storm 100 Cask System, (2) is required. REF ERENCE PROVIDED (') ('2.) A. ~cceptable verifying surface dose rates within 24 hours after beginning storage operations B. acceptable verifying all overpack inlet and outlet air ducts are free of blockage every 24 hours C~ II nacceptable initiating actions to return the MPC to an analyzed condition within 14 days D. II nacceptable administratively verifying fuel loading within 24 hours Monday. March 17, 2008 12:19:59 PM 1

MAR-17-2008 11:12 FROM BFN OPS TRAINING TO 9140456:24854153129 P.03 SFSC Heat Removal system 3.1.2

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY E;R 3.1.2.1 Verify all OVERPACK inlet and outlet air ducts are 24 hool'S free of blockage. For OVERPACKS with installed temperature 24 hours monitoring ~uipment, verify that the difference between the average OVERPACK air outlet temperature and ISFSI ambient temperature is

                                     $.126 QF.
             -..--------------------------------------------~~~----------

,,-...,.. Certificate of Compliance No. 1014 A)pendixA 3.1.2-2

MAR-17-2008 11:12 FROM BFN OPS TRAINING TO 9140456;24854153129 P.04 OVERPACK Average Surf.;ace Dose Rates 3.2.3

SL: RVEILLANCE REQUIREMI:NTS SURVEILLANCE II FREQUENCY SR 3.2.3.1 ve.my average surface dose rates of the OVERPACK On(;e, within 24 loaded with an MPC containing fuel assemblies are hours after within limits. beginning STORAGE A minimum of 12 dose rate measurements shall be OP!;;RATIONS taken on the side of the OVERPACK in three sets of four measurements. One measurement set shall be taken approximately at the cask mid-height plane, 90 degrees apart around the circumference of the cask. The second and third measurement sets shall be taken approximately 60 inches above and below the mid-height plane, respectively, also 90 degrees apart around the circumference of the cask. The average of the 12 dose rate measurements shaH be compared to the limit specified in LCO 3.2.3.a. A minimum of five (5) dose rate measurements shall be taken on the top of the OVE:RPACK. One dose rate measurement shall be taken at approximately the center of the lid and four measurements shall be taken at locations on the top concrete shield, approximately half w3:f between the center and the edge of the top shield, 90 degrees apart around the circumference of the lid. The average of the 5 dose rate measurements shall be compared to the limit specffied in LeO 3.2.3,b, A dose rate measurement shall be taken adjacent to each inlet and outlet vent duct. The average of the 8 inlet and outlet duct dose rates shan be compared to the limit specified in LCO 3.2.3.c.

   --,----~--~--------------~----------------------~--...~-------

Certificate of Complience No.1 014 Apt:endixA 3.2.3-2 TOTAL P.04

0610 NRC SRO Exam

22. SRO GENERIC 2.3.3 001IMEM/T3111GENERIC 2.3.31ISRO ONLYIBANK 11127/07 RMS During performance of 0-SR-DCS3.1.2.1 ," Spent Fuel Storage Inspection", you receive a report that a pile of leaves and other debris have been found at the base of Overpack BFN-0-CASK-079-0100/8; and it is blocking the air intake.

Which ONE of the following describes the required action(s)? J>~"r4""<:"*" ~ ... "'~~-- ~---'""'~_"""'-_ A. Notify Maintenance anc(Modifications Manage~~riftb remove the debris and clear the blockage. Note thatll1e blockage waSfouna~'ffi1f/NOT cleared, in the Post Test Remarks.

                               ,~j".~~'1:-'.-.n~-~-'-=~>.,"",~'~-'''-"--.~'.~__,""",

B. Coordinate wit~ E§!.~iJ!!i~~_s._~<~n~g~~t0 remove the debris and clear the blockage. Note that the blocKage was found and cleared in the Post Test Remarks. C. Notify the Maintenanc&~agerm remove the debris and clear the blockage. Note that the blockage was found:O~NOT cleared, in the Post Test Remarks. D!' Coordinate wittL~aaiatio~'P~oteCiionlp remove the debris and clear the blockage. Note that the bloC'i<agevmsrocrffcr-and cleared in the Post Test Remarks. KIA Statement: Radiation Control 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems). KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the required actions following discovery of a loss of radioactive material control.

References:

O-SR-DCS 3.1.2.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. 0610 NRC SRO Exam Sunday, February 17, 2008 3:28:28 PM 57

0610 NRC SRO Exam REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly, the candidate must:

1. Determine the appropriate action to take regarding the blockage.

2. Determine the documentation required based on Item 1.

D - correct: A - incorrect: Maintenance and Modifications Management is ONLY notified if the debris CANNOT be cleared. B - incorrect: The Facilities Department is NOT responsible for HI-STORM debris removal. The Facilities Department is responsible for general cleanliness inside the protected area EXCEPT the Independent Spent Fuel Storage Installation (ISFSI) Pad area. C - incorrect: The Maintenance Shift Manager is ONLY contacted if the blockage CANNOT be removed. ( Sunday, February 17, 2008 3:28:28 PM 58

BFN Spent Fuel Storage Inspection 0-SR-DCS3.1.2.1 Unit 0 Rev. 0004 Page 7 of 10 Date 7.2 Single HI-STORM Inspection Prior to Placement on the ISFSI PAD: (continued) [4] IF the HI-STORM(s) located on the ISFI Pad will not be inspected during the performance of this surveillance, THEN MARK the HI-STORM(s) listed on Attachment 2 as N/A. (Otherwise N/A this step.) 7.3 Inspection of HI-STORM(s) Located on the ISFSI Pad [1] PERFORM Attachment 2, HI-STORM Inspection Log. [2] IF any Inlet or Outlet Vents are found to have blockage, THEN PERFORM the following: (Otherwise N/A this section.) [2.1] NOTIFY the Unit Supervisor which HI-STORM ventilation ducts have blockage. [2.2] IF the blockage can be readily removed, THEN PERFORM the following: (Otherwise N/A) A. NOTIFY and COORDINATE with Radiation Protection for debris removal. B. REMOVE the blockage and debris from associated HI-STORM(s). C. RECORD the HI-STORM UNID and specific vent(s) that were blocked and cleared in the Narrative Log for all HI-STORM(s) with blockage.

BFN Spent Fuel Storage Inspection 0-SR-DCS3.1.2.1 Unit 0 Rev. 0004 Page 8 of 10 Date 7.3 Inspection of HI-STORM(s) Located on the ISFSI Pad (continued) [2.3] IF the blockage cannot be readily removed, THEN PERFORM the following: (Otherwise N/A) A. NOTIFY the Shift Maintenance Manager to PERFORM applicable section of MSI-O-079-0CS036 ISFSI Abnormal Conditions Procedure. B. IF acceptance criteria are not met within 4 hours, THEN NOTIFY Maintenance and Modifications Management to start preparation for MPC up load to HI-TRAC. (Otherwise N/A) US 7.4 Completion and Notification [1 ] VERIFY the Inlet and Outlet Vents on all HI-STORM(s) inspected on Attachment 2 are free of blockage. _ _ (AC) I [2] COMPLETE Attachment 1, Surveillance Procedure Review Form, up to Unit Supervisor review. [3] NOTIFY the Unit One, Two, and Three Unit Operators (UOs) this Surveillance Procedure is complete. [4] NOTIFY the Unit Supervisor this Surveillance Procedure is complete and PROVIDE status of any required Corrective Action per SPP-S.1 or unsatisfactory performances. 8.0 ATTACHMENTS Attachment 1: Surveillance Procedu're Review Form Attachment 2: HI-STORM Inspection Log

0610 NRC SRO Exam

23. SRO GENERIC 2.3.9 001lC/A/T3/CONTAINMENT//G2.3.912.5/3.4IR!MOD 113112008 Given the following plant conditions:

Unit 2 is commencing a scheduled reactor shutdown due to a leak in the Drywell.
The Operations Manager has directed that the Drywell and Torus be de-inerted so that an entry team can inspect the Drywell .

Cat1v<" ((.~tPk? s:W~N~)

Containment en,try is scheduled in 22 hours.
The unit is currently at 25% power.

Which ONE of the following describes the earliest time and preferred method for purging the containment to allow for Drywell entry? Purging of containment _ _ _ _ _->(....:.1./-)_ _ _ __ The preferred method of purging the atmosphere is using the _ _ _ _ _->.(=2)1--_ _ _ __ A. (1) CANNOT begin until reactor power is less than or equal to 15%. Standby Gas Treatment System. B. (1) CANNOT begin until reactor power is less than or equal to 15%. Primary Containment Purge Fans and normal Reactor Building Ventillation. C. (1) can begin IMMEDIATELY, -Q.ll--tfie-BryweU-;"BHt,,,t.he,,+9FbJs-GANNo.'T-98-f*.I-r:g.ed GQ-!:l-Gbl-r-Fer:1tl y . Standby Gas Treatment System. D~ (i) can begin IMMEDIATEL,(4JHfte--BryweI"I";-but;-ttte-"F0Flis.,gANN9=F=B&f>l:I'Fged (;0FlGI.:H:reFltl y . (2) Primary Containment Purge Fans and normal Reactor Building Ventillation.

I Sunday, February 17, 2008 3:28:28 PM 59

0610 NRC SRO Exam KIA Statement: Radiation Control 2.3.9 Knowledge of the process for performing a containment purge. KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to correctly determine the process for performing a containment purge.

References:

2-01-64, Rev. 106, section 8.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam ( Sunday, February 17, 2008 3:28:28 PM 60

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: The primary decisions that must be made to correctly answer this question are:

1. Can the purge be started immediately or must it commence after power is reduced

          . below 15%?

2. Which system is preferred to establish containment purge?

Although the TS bases for 3.6.3.2 is clear in describing the ability to commence purging 24 hours prior to going less than 15% RTP, the wording in Applicability (Q) of TS 3.6.3.2 has been known to cause confusion. This makes Distractors A and B plaUsible. The ability to simultaneously purge the Drywell and Torus is well within the capability of the system, however this is prohibited due to an analysis of the potential for containment over-pressurization if a LOCA occurred during the evolution. This makes Distractors A and B plausible because the purge lineup is available. D - correct: The shutdown is scheduled, therefore you can begin de-inerting 24 hours prior to reducing power < 15%. Additionally, Standby Gas Treatment is NOT the preferred method for purging containment. A - incorrect: is incorrect since you don't have to wait until 15% power to start de-inerting. B - incorrect: You don't have to wait until 15% power to start de-inerting however, the Purge Fan and RB Ventilation is the preferred method to purge the containment. C - incorrect: is incorrect since Standby Gas Treatment is NOT the preferred method for purging containment. Sunday, February 17, 2008 3:28:28 PM 61

Primary Containment Oxygen Concentration 3.6.3.2 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxygen Concentration LCO 3.6.3.2 The primary containment oxygen concentration shall be < 4.0 volume percent. APPLICABILITY: MODE 1 during the time period:

a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to

b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 Restore oxygen 24 hours oxygen concentration not concentration to within within limit. limit. B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion POWER to ~ 15% RTP. Time not met. BFN-UNIT 2 3.6-42 Amendment No. 253

BFN Containment Inerting System 2-01-76 Unit 2 Rev. 0062 Page 9 of 82 3.0 PRECAUTIONS AND LIMITATIONS (continued) L. Liquid nitrogen freezes skin on contact. Injury can also result from handling bare piping filled with liquid nitrogen. Caution should be used to prevent exposure from leaks and spills of liquid and insulated gloves should be worn to avoid direct contact with cold piping. M. BFN FSAR stipulates that 2-FCV-84-19 will be maintained closed except during surveillance testing or when directed by EOls. N. The following valves are interlocked closed with Mode Switch in RUN unless Division I and II RUN MODE BYPASS Switches, 2-HS-64-24 and 2-HS-64-25, are in BYPASS position: Div I (BYPASS SW 2-HS-64-24) Div II (BYPASS SW 2-HS-64-25) 2-64-18 2-64-17 2-64-19 2-64-30 2-64-29 2-64-33 2-64-32 2-76-24 O. TOE 0-97-064-0823 evaluated the impact of inerting or purging Suppression Chamber and Drywell concurrently (Both FCV 64-19 and FCV 64-18 open at the same time). This evaluation concluded there is a slight potential to over pressurize primary containment in the event of a large break LOCA with both FCV 64-19 and FCV 64-18 open at the same time with Reactor NOT in Cold Shutdown. This situation could create a large bypass flow path between the Drywell and the Suppression Chamber. Therefore, Suppression Chamber and the Drywell are NOT allowed be inerted or purged at the same time when Reactor is NOT in Cold Shutdown. P. Unless authorized by Shift Manager, applicable CAD TANK, level indicator's in the main control rooms (Unit 1 or 3) is required to be indicating 100% prior to filling/topping off Nitrogen tanks A or B (Refer to 2(3)-01-84 for filling CAD TANKS) Q. Drywell O2 CONCENTRATION indicators on Panels 2:..9-54 and 2-9-55 are no longer calibrated in the high range. PIP-97-189 should be used to correct O2 CONCENTRATION value as a reference until a grab sample is taken. Grab sample values of O 2 concentration are expected to vary from high-range O2 analyzer values.

0610 NRG SRO Exam

24. SRO GENERIC 2.4.21 OOllCIA/T3/RHRlB15/GENERIC 2.4.2111SRO ONLYIMODIF 11127/07 RMS Given the following plant conditions:

During normal full power operation of Unit 3, a loss of 250VDG power causes the RHR System I (Div I) Logic Power Failure Alarm to occur.
I&G investigates and reports a blown fuse in the logic circuit.
Prior to any corrective action being taken, a Loss of Goolant Accident results in the following plant conditions:

RPV Level 8elow TAF (Top of Active Fuel) RPV Pressure 100 psig Drywell Pressure 21 psig Which ONE of the following describes the response of Loop I RHR Pumps and what subsequent actions must be taken to address these plant conditions? A." ALL four (4) RHR pumps auto start in the LPGI mode and should remain there until RPV level is above TAF.

8. RHR Pumps '3A' and '3G' should be manually started in Drywell Spray mode.

RHR Pumps '38' and '3D' auto start in LPGI mode and should remain there until RPV level is above TAF. G. RHR Pump '3G' ONLY will auto start in LPGI mode. RHR Pump '3A 'can be manually started in LPGI mode and should remain there until RPV level is above TAF. RHR Pumps '38' and '3D' should be manually started in LPGI mode. D. RHR Pump '3A' ONLY will auto start in LPGI mode. RHR Pump '3G' can be manually started in LPGI mode and should remain there until RPV level is above T AF. RHR Pumps '38' and '3D' should be manually started in Drywell Spray mode. Sunday, February 17, 20083:28:28 PM 62

0610 NRC SRO Exam KIA Statement: Emergency Procedures IPlan 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including:1 Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment conditions 5. Radioactivity release control. KIA Justification: This question satisfies the KIA statement by requiring the candidate t.o use specific plant conditions to determine the response of RHR system components during an emergency based on a logic failure.

References:

OPL 171.044, 3-ARP-9-3D (5), 1-ARP-9-3D (5), 2-ARP-9-3D (5) Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam MODIFIED FROM OPL 171.044 #49 Sunday, February 17, 2008 3:28:28 PM 63

0610 NRC SRO Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following:

1. Recognize how a loss of Div 1 logic power affects the RHR pump initiation logic.

2. Recognize the difference between Unit-3 logic and Unit 1/2 logic.

3. Determine the correct application of EOI requirements with RPV level below TAF.

A* correct: B

incorrect: '3A' and '3C' RHR pumps auto start in LPCI mode. It is NOT appropriate to divert LPCI injection for containment control until RPV level has been restored above TAF. This is plausible because the RHR Div I pumps on Unit 1/2 will NOT start automatically.

C* incorrect: '3A' RHR pump will auto start in LPCI mode. '3C' and '3D' RHR pumps were NOT affected by the logic power failure and would start in LPCI mode. This is plausible because the CAS logic for Unit 1/2 are divided into "preferred" and "non-preferred" pumps as well as "divisionalized initiation logic." D

incorrect: '3C' RHR pump will start in LPCI mode. It is NOT appropriate to divert LPCI injection for containment control until RPV level has been restored above TAF.

This is plausible because the CAS logic for Unit 1/2 are divided into "preferred" and "non-preferred" pumps as well as "divisionalized initiation logic." Sunday, February 17, 2008 3:28:28 PM 64

OPL 171.044 Revision 15 Page 49 of 159 INSTRUCTOR NOTES

4. Pumps Triplinterlocks

a. Electrical Faults Obj. V.D.8.

b. Loss of suction path Obj. V. C. 7.

c. Each RHR pump has a NORMAUEMERGENCY Obj. V.C.7.

switch at the breaker. (1) With the switch in the EMERGENCY position, the pump can only be started from the breaker (i.e. automatic, local, and control room starts are removed) and then only if the associated pump drain valve is closed. (2) With switch in the EMERGENCY position, the pump will trip only from operation of breaker control switch, Shutdown board load shed, and electrical faults.

d. With an LPCI initiation signal present, the local Obj. V. C. 7.

station cannot be used to stop the pump. (1) The "white" light above the control switch in the control room indicate one of the following: (a) The breaker is tripped with the control switch in the Normal-After-Start or Start position (b) After an automatic start, the white light Red target on would not come on if the pump tripped control switch unless the control switch was place to the Start position. (c) The pump is running at "overload amp

                  . condition".

e. Since Unit 1 and Unit 2 share the U 1/2 diesel as an Obj. V.C.8.

onsite power supply, the possibility existed that the TP-45, 46, 47, diesels could be lost due to an overload conditions if and 48 Unit 1 and Unit 2 equipment were to start on the TP-61 , 62, 63, 64 board at the same time. To prevent this, each Unit was assigned "Preferred and Non-preferred" pumps and logic controls the pumps that are automatically started based upon which unit(s) is/are experiencing an accident.

OPL171.044 Revision 15 Page 50 of 159 INSTRUCTOR NOTES Note: Presently Unit 1 Accident signal will not affect Unit 2 due to DCN H2?35A that lifted wires from relays. Unit 2 will still affect Unit 1. However, the following represents modifications to the inter-tie logic as it will be upon Unit 1 recovery. Obj. V.B.13. Obj. V.C.3 (1 ) Unit 1 Preferred RHR pumps are 1A and 1C Obj. V.C.? Obj. V.D.6 (2) Unit 2 Preferred RHR pumps are 28 and 20 Obj. V.E.II (3) Unit 2 initiation logic is as follows: Div 1 RHR logic initiates Div 1 pumps (A and C), and Div 2 logic initiates Div 2 pumps (B and D)

f. Accident Signal (1 ) LOCA signals are divided into two separate Obj. V.B.13.

signals, one referred to as a Pre Accident Obj. V.C.3 Signal (PAS) and the other referred to as a Obj. V.C.? Common Accident Signal (CAS). Obj. V.D.6 Obj, V.E.II

PAS

                                       -122" Rx water level (Level 1)         Note:

OR It should be clear that the only 2.45 psig DW pressure difference between the two

CAS signals is the

                                       -122" Rx water level (Level 1)          inclusion of Rx OR                            pressure in the CAS signal. The 2.45 psig DW pressure AND <450          PAS signal is an psig Rx pressure                        anticipatory signal that allows the DG's to start on (2)    If a unit receives an accident signal, then all    rising DW its respective RHR and Core Spray pumps            pressure and be will sequence on based upon power source           ready should a to the SD Boards.                                  CAS be received.

(3) All RHR and Core Spray pumps on the non-affected unit will trip (if running) and will be blocked from manual starting for 60 seconds.

OPL171.044 Revision 15 Page 51 of 159 INSTRUCTOR NOTES (4) After 60 seconds all RHR pumps on the non- Operator diligence affected unit may be manually started. required to prevent (5) The non-preferred pumps on the non-overloading SD affected unit are also prevented from boards/DG's automatically starting until the affected unit's accident signal is clear. (6) The preferred pumps on the non-affected unit are locked out from automatically starting until the affected unit accident signal is clear OR the non-affected unit receives an accident signal.

g. 4KV Shutdown Board Load Shed Obj. V.C.B.

(1 ) A stripping of motor loads on the 4KV boards occurs when the board experiences an undervoltage condition. This is referred to as a 4KV Load Shed. This shed prepares the board for the DG ensuring the DG will tie on to the bus unloaded and without faults. (2) The Load Shed occurs when an undervoltage is experienced on the board i.e. or if the Diesel were tied to the board (only source) and one of the units experienced an accident signal which trips the Diesel output breaker. (3) Then, when the Diesel output breaker interlocks are satisfied, the DG output breaker would close and, if an initiation signal is present (CAS) the RHR, CS, and RHRSW pumps would sequence on (4) Following an initiation of a Common Accident Signal (which trips the diesel breaker), if a subsequent accident signal is received from another unit, a second diesel breaker trip on a "unit priority" basis is provided to ensure that the Shutdown boards are stripped prior to starting the RHR pumps and other ECCS loads (5) When an accident signal trip of the diesel Occurs due to breakers is initiated from one unit (CASA or actuation of the CASB), subsequent CAS trips of all eight diesel breaker diesel breakers are blocked. TSCRN relay

REV 0025 UNIT 3 Panel 9-3 3-ARP-9-3D 3-XA-55-3D Page 6 '-~---~-----"-: i . I SENSOR/TRIP POINT: I I RHR SYS I LOGIC

                     !    10A-K14 Relay          Loss of 250V DC power POWER FAILURE I

I c-l SENSOR LOCATION: Panel 3-9-32, Aux Instr Rm, El 593' PROBABLE CAUSE: 1. Failed fuse.

2. Loss of 250V DC power supply at 250V DC RMOV Bd 38.

AUTOMATIC ACTION: None OPERATOR ACTION: 1. DISPATCH personnel to 250V DC RMOV Bd 3B, Breaker lE2, to verify position.

2. DISPATCH personnel to Panel 3-9-32 to check 10-AMP fuses 10A-F1A and 10A-F2A.

3. REFER TO Tech Spec 3.3.5.1, ~, 3.5.2, 3.6.2.3, 3.6.2.4, 3.6.2.5, and TRM 3.3.3.4.

NOTE: IF alarm is valid, THEN the following will occur:

3A and 3C RHR Pumps will not receive an auto start signal from Div I Logic.
3A and 3C RHR Pump will receive an auto start signal from Div II Logic.
SYS'I Inboard Injection Valve will not receive an auto open signal from DIV I Logic.
SYS I Inboard Injection Valve will not manually open from the control room due to loss of 450 psig logic from DIV I.
The SYS I Inboard Injection Valve will receive an auto open signal from DIV II Logic.

REFERENCES:

3-45N620-2; 3-45E712-2; FSAR 8.6.4.2; Tech Spec 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation; 3.5.1, ECCS - Operating; 3.5.2, ECCS - Shutdown; 3.6.2.3, Residual Heat Removal (RHR) Suppression Pool Cooling, 3.6.2.4, Residual Heat Removal (RHR) Suppression Pool Spray; 3.6.2.5, Residual Heat Removal (RHR) Drywell Spray, TRM 3.3.3.4 ECCS and RCIC Trip System Bus Power Monitors.

BFN Panel 9-3 1-ARP-9-3D Unit 1 XA-55-3D Rev. 0021 ( Page 8 of 43 Sensor/Trip Point: RHR SYS I LOGIC 10A-K1A Relay Loss of 250V DC power POWER FAILURE (Page 1 of 1) Sensor 1-PNLA-009-0032 Location: Aux Instr. Rm, EI 593' Probable A. Cleared fuse. Cause: B. Loss of 250V DC power supply. Automatic None Action: Operator A. DISPATCH personnel to 250V DC Rx MOV Bd 1B, breaker 1E2, to Action: verify position. (Rx Bldg, EI 593', R-1 Q-LlNE) D B. DISPATCH personnel to 1-PNLA-009-0032 to check fuses 10A-F1A and 10A-F2A (10 amp). D C. REFER TO Tech Spec Sections 3.3.5.1, 3.5.1, 3.5.2, 3.6.2.3, 3.6.2.4, 3.6.2.5, TRM Section 3.3.3.4. D NOTE IF alarm is valid, THEN the following will occur:

1A RHR Pump will NOT auto start.
1C RHR Pump will NOT auto start.
SYS I Inboard Injection Valve will NOT receive an auto open signal.
SYS I Inboard Injection Valve will NOT manually open from the control room due to loss of 450 psig logic from DIV I.

References:

1-45E602-2 1-45E712-2 1-730E920-4 and -12

BFN Panel 2-9-3 2-ARP-9-3D Unit 2 2-XA-55-3D Rev. 0025 Page 8 of42 Sensor/Trip Point: RHR SYS I 2-RL Y-074-1 OA-K1A Loss of 250V DC power LOGIC POWER FAILURE (Page 1 of 1) Sensor Panel 2-9-32, Aux Instr Rm, EI 593' Location: Probable A. Cleared fuse. Cause: B. Loss of 250V DC power supply, at 250V DC RMOV Board 2B. Automatic None Action: Operator A. DISPATCH personnel to Panel 2-9-32 to check 1O-amp fuses Action: 10A-F1A and 10A-F2A. o B. DISPATCH personnel to 250V DC RMOV Bd 2B Breaker 1E2, to verify position. 0 C. REFERTO Tech Specs 3.3.5.1, 3.5.1,3.5.2, 3.6.2.3, 3.6.2.4, 3.6.2.5, TRM 3.3.3.4. 0 NOTE

1) IF alarm is valid, THEN the following will occur:

2A RHR Pump will NOT auto start.
2C RHR Pump will NOT auto start.
SYS I Inboard Injection Valve will NOT receive an auto open signal.
SYS I Inboard Injection Valve will NOT manually open from the control room due to loss of 450 psig logic from DIV I.

References:

2-45E620-2 2-45E712-2 GE 730E937-5 2-45E765-4 Tech Specs 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation 3.5.1, ECCS - Operating 3.5.2, ECCS - Shutdown 3.6.2.3, Residual Heat Removal (RHR) Suppression Pool Cooling 3.6.2.4, Residual Heat Removal (RHR) Suppression Pool Spray 3.6.2.5, Residual Heat Removal (RHR) Drywell Spray TRM 3.3.3.4, ECCS and RCIC Trip System Bus Power Monitors

0610 NRC SRO Exam

25. SRO GENERIC 2.4.30 002/CIAlREPIIIGENERIC 2.4.301ISRO ONLYINEW 11127/07 RMS ALL three units are operating at full power when the ODS notifies the SM that a major blackout has occurred in the Northeastern part of the United States. Efforts thus far have failed to stabilize the grid. The ODS requests that BFN carry maximum outgoing VARs to hold system voltage constant.

Units 2 and 3 were successful in attaining VARs at "200 Outgoing." Unit 1 tripped on a generator fault. Determine which ONE of the following describes the reporting requirements for these events and appropriate actions to ensure offsite power is available. A (1) report is required to the NRC, and notification to the ODS _....,.(=2}.<.....-_ required to ensure the availability of offsite power for BFN.

      -REFERENCE PROVIDED (1)          (2)

A. 1-hour; is B ..... 4-hour; is C)/~ C. 1- hour; is NOT D. 4- hour; is NOT Sunday, February 17, 2008 3:28:28 PM 65

0610 NRC SRO Exam KIA Statement: Emergency Procedures /Plan 2.4.30 Knowledge of which events related to system operations/status should be reporte~lItside agencies KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the emergency classification level and the associated reporting requirements.

References:

Reportability Matrix, SPP 3.4 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam REFERENCE PROVIDED: Reportability Matrix, SPP 3.4 Plausibility Analysis: I~ order to answer this question correctly, the candidate must determine the following: Current conditions require a 4-hour report to the NRC. In addition, notifying the ODS of current conditions is necessary to maintain offsite power availability for BFN. B - correct: A - incorrect: This situation does not require a 1-hour report to the NRC as long as offsite power is still available. However, notification of the ODS is the appropriate action. C - incorrect: This situation does not require a 1-hour report to the NRC as long as offsite power is still available. In addition, notification of the ODS IS the appropriate action. D - incorrect: This situation does requires a 4-hour report to the NRC. However, notification of the ODS IS the appropriate action. You have completed the test! Sunday, February 17, 2008 3:28:28 PM 66

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0018 Processes Page 18 of 64 Appendix A (Page 2 of 11) 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix D. Operations and the Plant Manager (or Duty Plant Manager) are responsible for making these internal management notifications.

2) NRC NUREG-1022, Supplements and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §50.72 and §50.73.

3.1 Immediate Notification - NRC TVA is required by §50.72 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notified. (Refer to Form SPP-3.5-1). Operations is responsible for making the reportability determinations for §50.72 and §50.73 reports. Operations is responsible for making the immediate notification to NRC in accordance with §50.72. Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile. NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC. A. The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8- hour phone calls. Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification Criteria. B. The following criteria require 1-hour notification:

1. (Technical Specifications) - Safety Limits as defined by the Technical Specifications which have been violated.

2. §50.72 (a)(1 )(i) - The declaration of any of the Emergency classes specified in the licensee's approved Emergency Plan.

NOTE If it is discovered that a condition existed which met the Emergency Plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or misclassified) event, shall be made. However, actual declaration of the emergency class is not necessary in these circumstances.

NPG Standard Regulatory Reporting Requirements SPP-3.S Programs and Rev. 0018 Processes Page 19 of 64 Appendix A (Page30f11) 3.1 Immediate Notification - NRC (continued)

3. §50.72(b).(1)) - Any deviation from the plant's Technical Specifications authorized pursuant to §50.54(x).

C. The following criteria require 4-hour notification:

1. §50.72(b)(2)(i) - The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.

2. §50.72(b)(2)(iv)(A) - Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

3. §50. 72(b )(2)(iv)(8) - Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

4. §50.72(b)(2)(xi) - Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.

D. The following criteria require 8-hour notification: NOTE The non-emergency events specified below are only reportable if they occurred within three years of the date of discovery.

1. §50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

2. §50.72(b)(3)(ii)(8) - Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

3. §50.72(b)(3)(iv)(A) - Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(8) [see list below], except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

a. Reactor protection system (RPS) including: Reactor scram and reactor trip.

NPG Standard Regulatory Reporting Requirements SPP-3.S Programs and Rev. 0018 Processes Page 41 of 64 Appendix D (Page 1 of 2) Site Event Notification Matrix Notification Requirements Event/Condition Duty Plant Plant Manager Ops. Duty Spec. Site VP Corporate Manager (ODS) Duty Officer* ReactorlTurbogenerator trip, unscheduled unit power reduction, or YeS# Yes# Reactor trip and Yes Yes nonscheduled unit shutdown; and when unit is restored to full unscheduled service. shutdown only Unplanned entry into a Limiting Condition for Operation with time Yes Yes No Yes Only for duration of 72 hours or less. duration of 24 hours or less Classification of the Radiological Emergency Plan (REP) at any Yes Yes Yes Yes Yes entry level. Personnel injuries that are potential lost-time injuries, serious Yes Yes Yes Yes Yes recordable injuries, and injuries requiring hQspital admittance or transport to an off-site medical facility. (Refer to Appendix J) Death of any person as a result of injuries received on site or due Yes Yes Yes Yes Yes to medical problems occurring while onsite. Release of oil or hazardous materials to the discharge canal, Yes Yes No Yes Yes ponds or river and violations of the NPDES permit Any event which may be newsworthy to the public. (1) Yes Yes Yes Yes Yes Any Security Contingency Events or any phone call to NRC Yes Yes Yes Yes Yes Operations Center regarding security issues. NRC 1 hour, 4 hour, or 8 hour phone calls. Yes Yes Yes for reactor Yes Yes for 1 hour trips, shutdowns, and 4 hour transport of calls. contaminated or potentially contaminated victim to hospital and for loss of Prompt Notification System.

                                               --  --- ---

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0018 Processes Page 42 of 64 ! Appendix D (Page 2 of 2) Site Event Notification Matrix Notification Requirements Event/Condition Duty Plant Plant Manager Ops. Duty Spec. Site VP Corporate Manager (ODS) Duty Officer* Any unusual radiation exposure to personnel. Yes Yes No Yes Yes Accidental, unplanned or uncontrolled off-site radioactive release. Yes Yes No Yes Yes Any reasonable threat to generation. Yes Yes No Yes Yes Outage critical path extensions exceeding 6 hours. Yes Yes No Yes Yes Any reactivity event or unplanned reactivity change. Yes Yes No Yes Yes

                                                                                                                                                     - - _ ... _--------

NOTE: (1 ) Consider items specified in Appendix Estep 2.2 of this procedure.

Plant Manager should ensure the Plant Managers at the other NPG sites are notified so the Shift Managers and Work Week Managers at the unaffected sites can review their schedules for potential generation-risk activities that may need to be deferred.

If the Corporate Duty Officer cannot be contacted within 15 minutes, the Chief Nuclear Officer should be notified.

Corporate Duty Officer (COO) The COO position is always available and is not part of the REP. This position is not intended to conflict or duplicate any responsibilities of the Central Emergency Control Center (CECC) Director under the REP or the Site Vice President.

(REFERENCES PROVIDED TO ( CANDIDATE

TITLE SPP-3.5 IDl Regulatory Reporting Requirements Rev. 0018 Page 1 of 64 Quality Related DYes 0No NPG Standard PORC Required DYes 0No Programs and Processes Effective Date 08-17-2007 Responsible Peer Team: Licensing Concurred by: Fredrick C. Mashburn for B. A. Wetzel 8/3/07

                                   *Primary Sponsor                              Date Concurred by:                      Robert H. Bryan, Jr.                          8/3/07 Peer Team Mentor                              Date N/A                                   N/A Approved by:

General Manager, NA Date William R. Campbell 8/15/07 Approved by: *Senior Vice President, Nuclear Operations Date

                 *Site-specific changes are approved by Site Sponsor and Site Vice President (see PCF)
                                                                             -}}

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