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Facility Post Examination Documents for the Kewaunee Retake Examination - August 2006
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FACILITY POST-EXAMINATION DOCUMENTS FOR THE KEWAUNEE RETAKE EXAMINATION-AUGUST 2006

Dominion Energy Kewaunee, Inc.

N490 Highway 42, Kewaunee, WI 54216-951 1 AUG 18 2006 Regional Administrator, Region I l l U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 21 0 Lisle IL 60532-4352 minion Serial No.

06-71 1 KPS/LIC:GR RO Docket No.

50-305 License No. DPR-43 Attention: Mr. Dell McNeil DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION OPERATOR LICENSE EXAMINATION In accordance with the requirements of NUREG-1 021, ES-501, the required post-examination material and the written examination facility comments are enclosed.

Forms ES-201-3, Examination Security Agreement, will be forwarded as soon as the required signatures are obtained. At the request of the Chief Examiner, Mr. Dell McNeil, the facility will not be providing Form ES-403-1.

If you have questions or require additional information, please feel free to contact Mr.

Frank Winks at 920-388-8303.

Very truly yours, Leslie N. Hartz II/

Site Vice President, Kewaunee Power Station Enclosure Commitments made by this letter: NONE cc:

Without enclosure NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 06-71 1 ENCLOSURE 1 OPERATOR LICENSE EXAMINATION WRITTEN EXAMINATION FACILITY COMMENTS GRADED WRITTEN EXAMINATIONS AND CLEAN COPY OF ANSWER SHEETS MASTER SRO EXAMINATION AND ANSWER KEY LIST OF QUESTIONS ASKED AND ANSWERS GIVEN DURING EXAM ADMINISTRATION STUDENT EXAM ITEM FEEDBACK AND RESPONSE WRITTEN EXAMINATION SEATING CHART WRITTEN EXAMINATION PERFORMANCE ANALYSIS INFORMATION KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Summary of Grading for Kewaunee Power Station 2006 NRC Retake Written Examination RO Section SRO Section 53 of 75 = 70.66%

63 of 75 = 84.00%

21 of 25 = 84.00%

15 of 25 = 60.00%

Initial wades prior to Feedback Total 74 of 100 = 74.00%

78 of 100 = 78.00%

RO Section SRO Section 55 of 75 = 73.33%

64 of 75 = 85.33%

21 of 24 = 87.50%

16 of 24 = 66.66%

Grades followinq Feedback includina accepted comments Total 76 of 99 = 76.76%

80 Of 99 = 80.80%

Summary of examination chanqes with accepted comments

1.

2.

3.

4.

5.

RO question 25. Initial correct answer was A. Accept either A or D as correct.

Both candidates selected D.

RO question 55. Initial correct answer was D. Accept either A or D as correct.

Candidate Beck selected A and candidate Bunkelman selected B.

SRO question 78. Initial correct answer was B. There is no correct answer. Delete question from examination.

SRO question 91. Initial correct answer was B. Accept either A or B as correct.

Candidate Beck selected B and candidate Bunkelman selected A.

SRO question 95. Initial correct answer was C. Accept either C or D as correct.

Candidate Beck selected D and candidate Bunkelman selected C.

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 8 Answer 6. Cold leg injection flow is decreased due to the filling operation. is also correct. Cold leg flow would decrease while the Accumulator fills. When opening a parallel flow path, flow in the first flow path must go down.

The supplied Engineering Calculation/Evaluation also concludes that, Pump run out should not be a concern for parallel pump operation if a Large Break LOCA were to occur during accumulator fill.

2. Resolution and Response The question is correct as written.

The question will be changed prior to entry into the bank to clarify the given condition for the design basis LOCA. A condition will be added that following initiation of the LOCA SI Pump B fails to start and CANNOT be started manually.

The calculation is correct with two SI Pumps running, however, the question premise provides the following condition, A Design Basis LOCA occurs. The question first gives that SI Pump A is running. USAR Section 14.3.2 describes the Major RCS Pipe Ruptures (LOCA), and subsection Performance Criteria for ECCS describes the ECCS systems response. On page 14.3-8 the following information is provided to define the design basis LOCA.

For the Best-Estimate large break LOCA analysis, one ECCS train, including one high head safety injection (HHSI) pump and one RHR (low-head) pump, starts and delivers flow through the injection lines. One branch of the HHSI injection line spills to the containment backpressure; the other branch connects to the intact loop cold leg accumulator line. The RHR injection line connects directly into the upper plenum. Both emergency diesel generators (EDGs) are assumed to start in the modeling of the containment fan coolers and spray pumps. Modeling full containment heat removal systems operation is required by Branch Technical Position CSB 6-1 (Reference 14) and is conservative for the large break LOCA.

The Technical Specification basis also supports the question as written. TS 3.3.b.5 provides protection from the possibility of one SI pump reaching runout condition during SI accumulator fill concurrent with a large break LOCA. CAPO28956 documents the reason the step 4.3.5.f and h, for entering and exiting a l-hour LCO to fill an SI Accumulator was added to procedure A-SI-33, Abnormal Safety Injection Accumulator Level and Pressure. This CAP references LER 97-004, which was submitted to report the existence of an unanalyzed condition during SI Accumulator filling operations. Specifically, if a LOCA occurred while filling an SI Accumulator, applying single-failure criteria to the Safety Injection pumps could result in only one SI pump delivering flow to both the RCS and the accumulator being filled, and a single SI pump providing flow to both the RCS and an accumulator could be subject to runout, which would result in no operable SI pumps during the accident.

Page 1

2006 Kewaunee NRC Retake Written Examination Question Feedback Concerning the reduction in SI flow, the amount of water being injected to the RCS would be reduced by the amount being diverted to the accumulator. However under the design basis accident, RCS pressure would be at Containment atmospheric pressure. With a maximum value of 46 psig for this pressure, the water flowing to the Accumulator would still be directed to the Cold Leg via the Accumulator outlet. If the flow indication (F1925) were assumed to the operator to be observed indication of Cold Leg Injection flow, then it would rise since the Accumulator fill line is downstream of the flow indicator.

3. Supportina Information Calculation/Evaluation C11026, Rev. Original, Section 6.0 Conclusions and Recommendations.

Drawings OPERXK-100-10, Rev. BG; OPERXK-100-28, Rev. AK and OPERXK-100-29 Rev. Z.

USAR, Rev 19, Section 14.3.2, pages 14.3-7 and 14.3-8 Technical Specification Basis for TS 3.3.b.5 LER 97-004-01 Page 2

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 19 There is no correct answer. E-0 step 5 does not direct you to check the SI ACTIVE Status Panel at this point. Step 17 has you check these items, and has you manually align those items.

2. Resolution and Response The question is correct as written.

The use of alternate indications for status verification is supported by Site and Operations standards.

Dominion Nuclear Operations Standard, DNOS - 0302, Control Board Monitoring, provides the following Expectation, Operators monitor control board indications closely to detect problem situations early. The Standard states, in part, The OATC (operator at the controls) shall use all direct and alternative indications for verification of status.

GNP-03.30.02, Conduct of Operations, Rev. E, also supports the above by reference to the expectations and standards for control board monitoring contained in DNOS - 0302.

Step 5, sub-step c states, FW-12A and B, Feedwater To Steam Generator A(B) Isolation Valves - CLOSED. The Contingency Action sub-step c states, Manually close the valves. BKG E-0 describes actions as, Determine if the valves are closed and close valves as necessary.

Also the instrumentation given is Position indications for these valves.

3. Supportinq Information DNOS - 302, Control Board Monitoring, Rev. 0 GNP-03.30.02, Conduct of Operations, Rev. E, Page 3

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 25 Accept D as correct answer.

Answer A will cause an SI thus emergency boration will not be required. That leaves answer D as the correct answer.

2. Resolution and Response Accept either answer A or D as the correct answer.

Rewrite question to include conditions that has SI Block active in selection A, or remove question from bank.

Not enough information in either the stem or selection A to preclude the assumption that a Safety Injection is NOT actuated due to the condition. Based upon the given information, Selection A with given indications could result in Safety Injection actuation on low Przr pressure (less than 1830 psig) or low steamline pressure (less than 514 psig). This would provide the necessary boration via SI Pump injection into the RCS, rather than by direction of E-CVC-35. The progression through the IPEOPs will be from E-O, Reactor Trip and Safety Injection, to E-2, Faulted Steam Generator Isolation, if the SG pressure is still dropping at step 23 of E-0; or to ES-1.1, SI Termination, at step 26 of E-0 if the btowdown of the SG is complete or stopped by the time step 23 of E-0 is reached.

In E-2, Faulted Steam Generator Isolation, steps wilt be taken to isolate the affected SG and transition would be made to E-1, Loss of Reactor Or Secondary Coolant, at step 9 of E-2. In E-l, Loss of Reactor Or Secondary Coolant, transition to ES-1.1, SI Termination, would be made at step 12, if conditions supported, or by direction of E-1 QRF (Quick Reference Foldout), item 1, SI Termination Criteria. If conditions did not support SI termination, then the transition would be made to ES-1.2, Post LOCA Cooldown And Depressurization, at step 18.b of E-I.

In ES-1.2, Post LOCA Cooldown And Depressurization, charging pumps are started and aligned to the RWST, and then at step 6 RCS cooldown to Cold Shutdown is initiated. In NOTE 1 prior to this step reads, NOTE: During the RCS cooldown, RCS Boron Concentration should be monitored to verify Cold Shutdown Boron Concentration per RD-6.7.

ES-1.1, SI Termination, also directs starting of charging pumps, but directs establishing normal charging and letdown. ES-1.I step 12 has the operator set makeup for automatic control with the Boric Acid Controller set to 11.O. The operator is directed in ES-1.I step 25 to establish stable plant conditions, which includes maintaining RCS temperature at existing value. This would place the plant in INTERMEDIATE SHUTDOWN. ES-1.1 then directs the operator to go to the appropriate Plant Procedure, which would be N-0-05, Plant Cooldown from Hot Shutdown to Cold Shutdown Condition. N-0-05, step 4.1.I and 4.1.2 direct the operator to determine the Cold Shutdown boron concentration and borate the RCS to this value.

Page 4

2006 Kewaunee NRC Retake Written Examination Question Feedback However, in both circumstances, the boration is performed using the normal boration controls for the Reactor Makeup Control System. Therefore, emergency boration is not used.

Selection D is definitive in its conditions and requires an emergency boration to the Hot Shutdown boron concentration. This emergency boration is the largest amount required (648 gallons) from the remaining selections.

Selection A remains an acceptable answer if it is assumed an SI does not occur. This is plausible if based on conditions as allowed in procedures (N-RC-36C, steps 4.2.7.8.b and 4.4.3 when RCS pressure is below 2000 psig and Safety Injection is blocked (Pressurizer SI Blocked). In this case SI would not occur and the appropriate action would be emergency boration. This emergency boration would require 241 6 gallons to reach the Cold Shutdown boron concentration.

3. Supporting Information IPEOPs E-0, ES-1.1 and ES-1.2 as indicated above.

Table TS 3.5-1, Amendment 172 N-RC-36, Pressurizer Pressure Control Page 5

TABLE TS 3.5-1 ENGINEERED SAFETY FEATU RES IN IT1 ATlON INSTRUMENT SETTING LIMITS Lead time constant II 2 540°F ll

~~

~~~

(') Initiates containment isolation, feedwater line isolation, shield building ventilation, auxiliary building special vent, and starting of all containment fans. In addition, the signal overrides any bypass on the accumulator valves.

Confirm main steam isolation valves closure within 5 seconds when tested. d/p = differential pressure Page 1 of 2 Amendment 172 02/27/2004

I WLS;CONSM PWLIC SERVICE CORPORATION I

KEWAUNEENUCLEARPOWERPLANT OPERATING PROCEDURE NO.

N-RC-36C r

I TITLE Pressurizer Pressure Control 1

DATE JUN 17 2005 PAGE 8 of 12 I

NOTE:

Automatic actuation o f Przr PORV PR-2A. Przr Spray Valves, P r z r Heaters, and annunciator PRESSURIZER CONTROL PRESS ABNORMAL (47043-C). i s c o n t r o l l e d by M a s t e r Controller output.

Misadjustment of c o n t r o l l e r setpoint d i a l, manual c o n t r o l l e r operation. manual back-up heater operation. o r high c o n t r o l l e r reset, w i l l a f f e c t actuation points o f above components.

_r

c.

POSITION Przr Spray Control Master c o n t r o l l e r t o automatic:

1.

ADJUST Master c o n t r o l l e r setpoint d i a l u n t i l the deviation meter n u l l s.

( t o p meter)

2.

VERIFY Master C o n t r o l l e r setpoint matches pressurizer pressure channel selected f o r control.

DTE:

Steps 4.2.7.c.3 and 4.2.7.c.4 VERIFY c o n t r o l l e r c i r c u i t r y i s working properly.

3.

POSITION AUTO-BAL-MAN switch t o MAN-BAL.

4.

VERIFY deviation meter i s centered.

5.

POSITION AUTO-BAL-MAN switch t o AUTO.

6.

VERIFY pressurizer spray valves and heater control working properly t o maintain Przr pressure.

8.

ADJUST Master C o n t r o l l e r setpoint d i a l t o slowly INCREASE Pressurizer pressure t o 2235 psig.

a.

Maintain RCS pressure w i t h i n l i m i t a t i o n s o f RD-11.1.

b.

To ensure s a f e t y i n j e c t i o n a c t i v a t i o n c i r c u i t s automatically unblock, WHEN Pressurizer pressure i s greater than 2000 psig.

VERIFY following permissive status l i g h t s, OFF:

0 44905 1101. Pressurizer Perm Block SI 0 44905 1102, Pressurizer S I Blocked

TITLE Pressurizer Pressure Control WLSCOlYSM PWLIC SERMCE CORPORATION KEWAUNEENUCLEARPOWERPLANT JUN 17 2005 I PAGE 10 of 12 I

OPERATING PROCEDURE 4.3 CONTINUED NOTE:

Pressurizer Backup heaters should be energized f o r any o f t h e f o l l o w i n g reasons:

0 To equalize RCS t o pressurizer b o r i c a c i d difference of more than 50 ppm 0 To reduce excessive c y c l i n g o f pressurizer backup heaters during RCS temperature changes 0 To minimize pressure transients during p l a n t evolutions

2.

WHEN required f o r p l a n t conditions. THEN POSITION Backup Przr

3.

WHEN NO longer required f o r p l a n t conditions, THEN POSITION Backup Heater Groups t o ON.

Przr Heater Groups t o AUTO.

4.4 Pressure Control During Cooldown:

NOTE:

Difference i n temperature between Przr and RCS s h a l l NOTE:

Hold Steam Generator pressure a t 600 p s i g u n t i l S I can

1.

POSITION a t l e a s t two backup Pressurizer heater groups t o ON.

2.

(CAS) MAINTAIN RCS pressure w i t h i n l i m i t a t i o n s o f RD-11.1 by NOT exceed 320°F.

be blocked a t 2000 psig i n RCS.

adjusting PRZR Spray Master C o n t r o l l e r setpoint.

WHEN RCS pressure i s 1900-1950 psig. THEN PERFORM t h e f o l l o w i n g t o

3. ~-

block safety i n j e c t i o n :

a.

VERIFY permissive status

b.

POSITION Safety I n j e c t i o n Block S I. ON.

switches t o BLK.

c.

V E R I F Y permissive status Blocked. ON.

i g h t 44905 1101. Pressurizer Perm T r a i n A and T r a i n B Block/Unblock i g h t. 44905 1102. Pressurizer S I CONTINUED

ES-1.2 WISCONSIN PUBLIC SERVICE CORPORATION I NO.

9 OPERATOR ACTIONS POST LOCA COOLDOWN AND KEWAUNJCENUCLEARPOWERPLANT I DEPRESSURIZATION CONTINGENCY ACTIONS EMERGENCY OPERATING PROCEDURES I G - J U L 06 2005 I PAGE 3

of 18 2

Establish Charging Flow:

a. Charging Pumps - AT LEAST ONE

a. Perform the following:

RUNNING

1)

CC flow t o RXCP(s1 Thermal B a r r i e r i s l o s t,

THEN l o c a l l y close CVC-204A(B) t o i s o l a t e seal i n j e c t i o n t o affected RXCP(s) before s t a r t i n g Charging Pumps.

2) S t a r t one Charging Pump.

b. Align Charging Pump suction t o RW ST I

I

1) Open CVC-301. RWST Supply To

2) Close CVC-1. VCT Supply To Charging Pumps Charging Pumps

c. S t a r t a second Charging Pump AND establish maximum Charging f l o w

WISCONSIN PUBLIC SERVICE CORPORATION POST LOCA COOLDOWN AND KEWAUNEENUCLEARPOWERPLANT I DEPRESSURIZATION ES-I. 2 NO.

EbERGENCY OPERATING PROCEDURES DATE JUL 06 2005 PAGE 5

of 18 I

I

a. Locally t h r o t t l e AFW-3A. AFW 5

Maintain AFY Pump Discharge Pressures Greater Than 1000 PSIG:

a. T h r o t t l e AFW-ZA/CV-31315. AFWP

b. T h r o t t l e AFW-ZB/CV-31316. AFWP Pump A Discharge.

A Flow Control B Flow Control I

Pump B Discharge.

AFW Pump Discharge I

pressure could NOT be

b. Locally t h r o t t l e AFW-3B. AFW I

I I

c. Locally t h r o t t l e AFW-2C. T/D

c.

TDAFW Pump discharge maintained greater than 1000 psig. THEN l o c a l l y perform t h e following:

1) P o s i t i o n TDAFW Pump Low Disch Pr&ss T r i p Bypass switch t o BYPASS.

OPERATOR ACTIONS

2) T h r o t t l e AFW-2C as necessary during RCS cooldown t o maintain TDAFW flow less than 260 gpm.

CONTINGENCY ACTIONS NOTE:

During the RCS cooldown. RCS Boron Concentration should be monitored t o v e r i f y Cold Shutdown Boron Concentration per RD-6.7.

RCS cooldown should be performed as f a s t as possible, b u t less than 100" F/hr.

NOTE:

6 I n i t i a t e RCS Cooldown To Cold Shutdown :

a. Maintain cooldown r a t e i n RCS c o l d legs - LESS THAN 10O0F/HR

b. Use RHR System i f i n service

c. Dump steam from i n t a c t SG(s)

c. Dump steam using i n t a c t SG(s) using the Steam Dump System i n STM PRESS mode PORV.

WISCONSM PUBLIC SERVICE CORPORATION ES-1.1 NO.

KEWAUNEE NUCLEARPOWERPLANT EMERGENCYOPERATINGpRocEDuREs 12 13 14 15 16 TITLE SI TERMINATION DATE JUN 21 2005 P-7 of 20 Check VCT Makeup Control System:

F i

a. Makeup Boric Acid Controller -

SET TO 11.0 OPERATOR ACTIONS

b. Makeup s e t f o r automatic control

c. VCT l e v e l - BETWEEN 17% AND 28%

Check Charging Pump Suction -

ALIGNED TO VCT Transfer Steam Dump - TO PRESSURE CONTROL HODE Check RCS Hot Leg Temperature -

STABLE Using Pressurizer Heaters And Normal Pressurizer Spray As Necessary, Maintain Pressurizer Pressure - STABLE I

CONTINGENCY ACTIONS I

a. Set Makeup Boric Acid c o n t r o l l e r t o 11.0.

b. Set Makeup Mode Selector t o AUTO.

c. Re-establish VCT level.

A l i g n suction t o VCT.

IF Condenser Steam Dump NOT available, THEM use Atmospheric Steam Dumps o r Steam Generator PORYs.

Control steam dump and t o t a l feed f l o w as necessary t o s t a b i l i z e RCS temperature.

I F normal spray NOT available AND Letdown i s i n service, THEN use a u x i l i a r y spray.

I F NOT. THEN use one PRZR PORV.

WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNJIE NUCLEARPOWERPLANT EMERGENCY OPERATING PROCEDURES ES-1.1 NO.

TI-S I TERMINATION DATE JUN 21 2005 PJGE 13 of 20 24 25 26 OPERATOR ACTIONS I

D Shut Down Unnecessary Plant Equipment:

a. Refer t o ATTACHMENT B

b. Refer t o N-0-03. PLANT OPERATION GREATER THAN 35% POWER

c. Refer t o N-0-04. 35% POWER TO HOT SHUTDOWN CONDITION Haintain Stable Plant Conditions:

Pressurizer pressure - AT EXISTING VALUE 0 Pressurizer l e v e l - GREATER THAN 19% [42% FOR ADVERSE CONTAINMENT]

RCS temperature - AT EXISTING VALUE 0 I n t a c t Steam Generator narrow range l e v e l s - BETWEEN 4% C15X FOR ADVERSE CONTAINMENT3 and 50%

Verify SI Flow Not Required:

a. RCS subcooling based on Core E x i t TCs - GREATER THAN 30°F 165°F FOR ADVERSE CONTAINMENT3

b. PRZR l e v e l - GREATER THAN 5%

[30% FOR ADVERSE CONTAINMENT]

I CONTINGENCY ACTIONS I

a. Manually s t a r t S I Pumps as necessary.

GO TO E-1. LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

b. Control Charging f l o w t o maintain PRZR l e v e l.

PRZR l e v e l can NOT be maintained, THEN manually s t a r t S I Pumps as necessary.

GO TO E-1. LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

Plant Cooldown from Hot Shutdown t KEWAUlWJt POWER STATION I 'ITLG Cold Shutdown Condition I

OPERATING PROCEDURE I

DATE MAY 06 2006 P

7 of 31 I I 4 - 0 I

4.1 PreDare To Cooldown:

1.

DETERMINE required Cold Shutdown boron concentration using one o f the following:

CPCR0234511 I

NIT1 ALS APPLIES/NA

a. IF a l l control rods are f u l l y inserted, THEN REFER t o RD 6.9, 1% Cold Shutdown A R I Boron Concentration.

APPLIES/NA

b.

any control rod i s not f u l l y inserted, THEN CONSULT w i t h Reactor Engineering t o determine requi red Cold Shutdown boron concentration.

2.

( L ) INITIATE boration o f RCS per N-CVC-35A as follows:

I a.

BORATE t o required Cold Shutdown boron concentration determined i n Step 4.1.1.

BORATED/

I N I T I ATED NOTE:

During RCS cooldown. a t l e a s t two backup groups of Pressurizer Heaters should remain energized t o minimize f a t i g u e stress t o Pressurizer and Pressurizer Surge Line.

3.

(CAS) ESTABLISH, RCS Pressure Control During ESTABLISHED Cooldown. per N-RC-36C.

ESTABLISHED

4.

(CAS) ESTABLISH. Automatic Pressurizer Level Control f o r Shutdown, per N-CVC-35B.

5.

( & I any valves have been backseated per PERFORMED/ N A N-0-01-CLE. THEN PERFORM N-0-01-CLE as necessary t o take applicable valves o f f backseat.

long a s possible during p l a n t cooldown.

6.

(CAS) MAINTAIN Steam Generator Blowdown f l o w as MAINTAINED.

CONTINUED

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 52 Multiple correct answers.

All of the given monitors will indicate the release of activity from the B steam generator.

2. Resolution and Response The question is correct as written.

In A, R-36 is a high-range monitor for the Aux Building Stack; In 6 and D, R-34 is a high-range monitor for the steam line from SG B.

Both R-34 and R-36 have a sensitivity of 1 Whr with an indicated range of 1 Whr to 10,000 Whr.

With a leak of 10 gpm, the expected radiation indication would range about several mWhr at the main steamline. This is about one one-hundredth the sensitivity of the detectors and would not be expected to be displayed for the monitors.

~

3T~upportinq Information Drawing E-2021, Integrated Logic Diagram Radiation Monitoring.

USAR, Rev. 19, Table 11.2-7 E-0-14, Steam Generator Tube Leak.

Page 6

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 55 Answer A is also correct. No timeframe was given for the question. CVC-11 fails closes.

2. Resolution and Response Accept either *A or D as correct.

The question does not provide the time frame for selecting the answer. CVC-11 does have an air accumulator in its supply line and will remain in its current position for a period of time.

However, if air is not restored the accumulator will become depleted and CVC-11 will fail closed.

The line for CVC-11 is a 2 line and the bypass line containing CVC-13 and CVC-14 is a % line.

The purpose of CVC-14 is to provide a relief path due to differential pressure in the line upstream and downstream of the regenerative heat exchanger when CVC-11 is closed. The reduction in flow when CVC-11 closes would be apparent to the operator. Running this condition on the simulator shows that the bypass line only passes approximately 0.6 gpm for F1128 (node point measurement on bypass line). This is below the accuracy of FI-128 meter, so for the operator flow goes to zero. The seal injection pathway flow provides the other flow path for charging.

3. SuDportinq Information System Description 01, Rev. 3, 3.7.4 System Description 35, Rev. 2, 3.3.6 Drawings OPERKK-100-35, OPERXK-100-36 and E-2025 Page 7

Title Station Air & Instrument Air System (AS) I Date 02/11/04 3.7.5 Major SA & IA Manual Valves Page 14 of 21

I Title Chemical and Volume Control System (CVC) I Date 10/02/02 I Page 29 of 73 I CVC-7 is controlled from either the DSP or the Control Room. DSP Switch ES 87128 (two position, LOCALAXEMOTE) is a switch that determines the control location.

+ When ES 87128 is in LOCAL, CVC-7 is controlled using Control Station CS 87209 on the DSP.

+ When ES 87128 is in REMOTE, CVC-7 is controlled using Control Station CS 4320301 in the Control Room.

Charging flow through CVC-7 is sensed by Flow Element FE-128 and indicated locally and in the Control Room. FE-128 is located upstream of CVC-7, but downstream of where seal water injection taps off. Therefore the Control Room Operator must compensate when comparing charging and letdown flow since some of the seal water injection will return to the RCS via the RXCP labyrinth seals.

3.3.6 Charging Stop Valve CVC-ll/CV-31229 The Charging Stop Valve CVC-ll/CV-31229 isolates charging from the RCS.

CVC-11 is an air operated valve and fails closed on a loss of air. CVC-11 is located downstream of the Regenerative Heat Exchanger in the Regenerative Heat Exchanger Room inside Reactor Containment.

CVC-11 is open for normal operation. A bypass line around CVC-11 contains a Check Valve CVC-14. When CVC-11 is closed, CVC-14 will relieve excessive CVC-11 is controlled from the dedicated shutdown panel or the Control Room.

Dedicated Shutdown Panel Control Switch ES 87128 is a two position, LOCALlREMOTE switch used to select the controlling location.

+ When LOCAL is selected, control is from Dedicated Shutdown Panel Switch ES 87129. ES 87129 has two positions, OPEN/CLOSE.

+ When REMOTE is selected, control is from Control Room Switch ES 46238.

ES 46238 has two positions, CLOSE/OPEN.

I

E -. I I

i

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a

LA

r

.x ZF 8

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 75 No correct answer. In the given conditions, SI has occurred. This starts all 4 fan coil units. No indication that any have been secured. Step 2 of FR-Z.3 has you verify fan coil units are running. The answer is yes, thus you do not Start containment fan coil units... They are running.

2. Resolution and Response The question is correct as written.

The premise does not provide the status of the Containment Fancoil units; however, there is nothing that precludes the units from being stopped or not running at this point in the procedure.

The correct action is to verify they are running and if not start the Fancoil Units. If the situation was that the Fancoil Units are always running after SI and continued to run at this point, there would be no need for the Contingency Action statement. The correct action is to start the Fan Coil Units for the purpose provided.

Question could be improved by changing Start... to Verify Containment Fancoil Units running... In this case by definition, verify means check running and if not running start.

3. Supporting Information FR-Z.3 BKG FR-Z.3 Page 8

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 78 The question does not take into consideration the full effects of the Bus 6 lockout outside of Bus

62. On a bus 6 lockout from 100% power, the secondary plant transient would lead to a condition that would require a reactor trip to place the unit in a safe condition. The loss of power to Bus 6 has various effects but most notably at 100% power are the loss of both Heater Drain Pumps and the loss of Service Water Train B.

Both Heater Drain Pumps are lost due to de-energization of MCC-62C and MCC-62E, leading to the loss of power to the Heater Drain Pump magnetic couplings and de-coupling. Per A-CD-03, the loss of two Heater Drain Pumps requires turbine impulse pressure to be reduced to less than or equal to 425 psig or approximately 70% using VPL Lower to maintain adequate suction pressure to the Main Feedwater Pumps. This is not accounted for in the question considering the stem has the plant at 97% following the transient which is not possible based on the previous discussion.

Additional consideration needs to be made based on the loss of Service Water. The Bus 6 lockout causes Service Water Pumps 1 B1 and 1 B2 to be lost due to the loss of power. Header pressure would be expected to drop to less than 72 psig in the B header quickly closing SW-3B if not already closed (in-plant condition). With the Turbine Building SW Selector Switch in the B position, all Service Water flow would be lost to the Turbine Building. Temperatures on various equipment and components such as Feedwater and Condensate Pump bearings and slot gas temperatures would rise very quickly. Temperature limits would be exceeding requiring equipment stopped/tripped within a matter of minutes including a trip of the main turbine. At the onset of the transient the following procedures need to be addressed A-CRD-49, A-EHV-39, A-SW-02, A-CD-03, and various other abnormals associated with high temperatures on equipment in the Turbine Building. During review and performance of the listed procedures trip criteria would be met prior to opening of SW4A in A-SW-02 which is late in the procedural actions.

Therefore, based on the above discussion, the loss of Bus 6 puts the unit in a transient that is more severe than is stated in the question. Based on actual plant performance and response, the trip of the unit is the correct conservative response to make given this situation.

Answer C is correct.

Page 9

2006 Kewaunee NRC Retake Written Examination Question Feedback

2. Resolution and Response There is no correct answer.

The question is to be deleted from the exam.

The loss of the Heater Drain Pumps was not addressed in development, review or validation of the examination, only the affect on IRPls (rod indication). The action is correct to reduce load in accordance with A-CD-03, Condensate System Abnormal Operation. The plant would not be stabilized at 97% power; therefore, the premise of the question is invalid. This makes the question incorrect. As stated above, A-CD-03 requires with no Heater Drain Pump running that turbine load be reduced to less than or equal to 425 psig impulse pressure, or approximately 72% power.

The loss of Service Water is not considered since the question does not specify the Turbine Building SW Header is selected to either train. Assumption that the Turbine Building SW Header is aligned to Train B is not prudent since the header is just as likely to be aligned to Train A.

This should have been addressed during the examination as a question on alignment of SW.

C cannot be a correct answer. Even if the reactor is tripped due to the given conditions, the emergency boration of 360 gallons for each control rod is not required. This makes the selection also incorrect.

3. Supporting Information A-CD-03, Rev. Q, step 4.4.1 Drawings E-233 and E-1530 Page 10

NO. A-CD-03 I WIScONSIN PUBLIC SERVICE CORPORATION OPERATING PROCEiDUIE Condensate System Abnormal KEWAUNEENUCLEARPOWERPLANT DATE JUN 01 2005 P-5 o f 9

~

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INITIATING EVENT Condensate Pump T r i p

. O SUBSEQUENT ACTIONS IMPULSE PRESSURE s275 p s i g 4.1 4.2 4.3 4.4 Feedwater Pump T r i p Two Heater Drain Pumps T r i p Condensate Pump A Hotwell Level Trip:

1.

VERIFY Feedwater Pumps A and B. OFF.

2.

VERIFY AFW Pumps A and B running.

3.

VERIFY Turbine tripped.

5285 p s i g 5425 p s i g

4.

power i s above P-10, THEN VERIFY reactor,rip an' Condensate Pump B Hotwell Level Trip:

None Condensate Pump Overcurrent Trip:

None Feedwater Htr Bypass Alert:

GO TO E-0.

1.

Based on i n i t i a t i n g event, REDUCE t u r b i n e impulse pressure, using Manual VPL Lower. t o value recommended i n t a b l e below:

~

~~

~

One Heater Drain Pump T r i p 5520 p s i g 1

2.

Feedwater Pump suction pressure can NOT be maintained greater than 260 psig. THEN REDUCE Turbine load u n t i l suction greater than 260 psig.

CONTINUED

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TITLES HEATER DRAIN COUPLING ( SHEET 1 OF 2)

FOR MlR. QPR.

RHW.

MOR.^.

(XK 136-4)

'I NOTE VENDOR DWG'S Mh! NOT RL~LEGT KNPP RELAY DESIGNATION

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 80 Not enough information. If the recirculation fluid temperature and pressure exceed design conditions, then answer D is correct.

2. Resolution and Response The question is correct as written.

The premise that the feedback addresses is based on the precaution and limitation from the system normal operating procedure, N-RHR-34, Residual Heat Removal System Operation.

This P&L is applicable to normal operation conditions only. The conditions given in the question is post-accident. Thus the limitations for the systems are the design limits. Consideration must be then given to RHR, SI and Containment Spray systems that may be supplied on recirculation flow. From the USAR the following design values are:

Desiqn Temperature Desiqn Pressure RHR Pump 4OOOF 600 psig (also for valves & pipes)

RHR Hx SI Pump 3OOOF 2485 psig ICs Pump 3OOOF 500 psig 400° (tube) / 35OOF (she11)600 psig (tube) / 150 psig (Shell Based on the accident analysis the maximum expected Containment Sump temperature is approximately 280OF. This correlates to a maximum Containment pressure of 46.0 psig. These valves fall within the acceptable design limits for the associated systems that may receive the recirculated fluid under accident conditions. As shown by the graph, the actual temperature in the Containment Sump at the time ES-1.3 is being performed (approximately 20 minutes following the initiation of the accident) will be closer to 25OOF (maximum). This is confirmed in USAR Table 6.2-1 2 that gives 25OOF as the maximum operating conditions for RHR components during post LOCA recirculation.

3. Supportinq Information BKG ES-1.3, step 5 N-RHR-34, Precaution & Limitation 2.8 USAR Rev. 16, Table 6.2-6, Table 6.2-7, Table 6.2-1 2, Table 6.4-2 USAR Rev. 16 Figure 14.3.4-12 USAR Rev. 16 9.3.3, page 9.3-13 Page 11

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 85 Answer A is correct based on procedure use and adherence rules. E-EDC-38A Symptoms Section 2.1 lists 47101-A, DC VOLTAGE LOW, as an entry condition for procedure. Procedure E-EDC-38A may be implemented due to a symptom being met, therefore answer A is correct.

2. Resolution and Response The question is correct as written.

The single symptom, BRA-102 DC VOLTAGE LOW (47101-A), is not indicative of a loss of power to BRA-104, which is the purpose of E-EDC-38A. It would not be appropriate to perform E-EDC-38A for these conditions. The Alarm Response Procedure in and of itself directs, GO TO A-EDC-38.

Other than the first Subsequent Action of E-EDC-38A, 4.1, Contact Plant Electricians to investigate cause for loss of power, the remaining steps of E-EDC-38A address actions for systems lost or affected by the loss of DC. These do not address the conditions provided in the premise. A-EDC-38, which also has the same symptom, does directly address actions to address the specific conditions by investigating and checking local indications and components.

The UG-0 provided section on Procedure Entry refers to dual-column format procedures only, as indicated by the hierarchy.

1. Performance of all dual-column format procedures shall start at section 4.0.

Detailed Procedure, Step 1.

a. Section 1.O. Introduction, and Section 2.0. Symptoms, may be used as necessary to determine if the procedure is applicable.

E-EDC-38A and A-EDC-38 are single-format procedures. The condition may apply to these procedures, but the key statement is to determine if the procedure is applicable.

3. Supportinq Information E-EDC-38A, Purpose, Symptoms and subsequent actions (page 2)

A-EDC-38, Purpose, Symptoms and subsequent actions (page 2)

GNP-03.01.03, Rev. U, 6.1.8.3 UG-0 section 6.3.1 Page 12

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feed back Question Number 86 Following establishment of Bleed and Feed, the PORV relieving to the PRT causes the rupture disc to rupture. It is anticipated that Containment may reach Adverse Conditions of 4 psig. Step 27 of FR-H.l has criteria for a hot, dry SG at 5% [20% for Adverse]. In either case this is met and FW flow should be limited to 60 to 100 gpm until level is greater than 5% [20% Adverse].

Levels per the question are 2% and 3%. Answer A is correct.

2. Resolution and Response The question is correct as written.

Hot Dry SG conditions do exist. However, with the RCS temperatures rising, the urgency of establishing a heat sink predominates, and full feedwater flow should be established to one SG.

As indicated in the BKG FR-H.1,...it is advisable to reestablish feedwater to only one steam generator regardless of the size of the plant or number of loops. Thus, if a failure occurs due to excessive thermal stresses, the failure is isolated to one steam generator.

This last statement makes A incorrect since it establishes flow to both SGs.

3. Supporting Information BKG FR-H. 1, Rev., 2.4 Page 13

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 91 Answer A is correct. The premise of the question states that SP-36-082 is completed with a leak rate of 1.1 lgpm of total RCS leakage. Also given are the combination of the RXCP annunciator and RCDT level change both of which are indicative of a Number 2 Seal issue on the B RXCP. Using the provided Operator Aid the leakage from the Number 2 seal can be calculated to be approximately 0.75gpm. (change of the last hour minus normal leakage based on RCDT level change or 0.8 gpm - 0.04gpm = -0.75 gpm) This leakage is now considered identified leakage. It is known where it is coming from, that it is contained, and that it is within the limitations of the RCDT. Being that the total leakage identified by SP-36-082 was 1.1 lgpm and 0.75gpm has been determined to be identified leaves the remaining 0.36gpm as unidentified. Technical Specifications allows not in excess of 1 gpm unidentified leakage and not in excess of 10 gpm identified leakage. Neither of these thresholds have been met. Therefore no actions are required since unidentified leakage remains less than lgpm per Technical Specifications.

2. Resolution and Response Accept either A or B as correct.

Rewrite question to specify that only the mass balance leakrate calculation (of SP-36-082) has been completed per section 6.1, and has the following value of 1.1 1 gpm.

The premise does state, SP-36-082, Reactor Coolant Leak Rate Check, has just been completed. The intent of this was to signify that the Mass balance Leakrate Calculation had been completed. In this event, the investigation is required. This was apparent during development, review and validation, since no comments were received that disputed the selected correct action.

However, the statement in the premise can be interpreted to mean the entire procedure is complete, which includes section 6.3, the Investigation and Evaluation. If this is performed and the RCDT level change is attributed to the RXCP #2 seal problem, then no other action is required. Data Sheet 4, INVESTIGATION AND EVALUATION, would have been completed to describe the source of the leak, the effect on plant operation and the determination the plant operation may safely continue. This also agrees with the actions of A-RC-36C, Attachment b, Response to Abnormal #2 Seal Leakoff, with #2 seal leakoff flow greater then 0.5 gpm but less than 1.1 gpm.

3. Supportins Information SP-36-082 Rev. AJ A-RC-36C, Attachment B Page 14

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DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE NUCLEAR

@ YES SAFETY RELATED 0 NO I

I NO. SP-36-082 I REV A3 (FREQ D)

PORC REVIEW YES SRO APPROVAL OF YES REQUIRED TEMPORARY CHANGES 0 NO REQU I RED 0 NO I

Reactor Cool ant System Leak Rate I

Check 1 DATE MAR 02 2006 I PAG 1

of 14 I I

Jefiey Simon APPROVED BY I

Michael Sevey REVIEWED BY I

1.0 PLANT INITIAL CONDITIONS 1.1 Mass Balance Leakrate Calculation shall be performed a t l e a s t weekly whenever the Reactor i s a t power o r i n a Hot Shutdown condition.

requirement can NOT be waived. CPCR0166551 This 1.2 Mass Balance Leakrate Calculation performance i s desired on a d a i l y basis whenever the Reactor i s a t power o r i n a Hot Shutdown condition.

The S h i f t Manager may waive t h i s requirement i f p l a n t conditions do NOT allow d a i l y performance. CPCR0166553 Reactor power and xenon are stable, such t h a t letdown d i v e r s i o n t o the Holdup Tanks w i l l _NOT occur during the t e s t.

1.3 2.0 PRECAUTIONS 2.1 I f one o r more o f the f o l l o w i n g events occur. the t e s t i s v o i d and s h a l l be repeated:

.I Emergency boration

.I Diversion o f letdown t o the Holdup Tanks

.I Makeup from any source which does not go through the Boric Acid o r Diversion o f Excess Letdown t o the RCDT I

Makeup Water t o t a l i z e r s A Reactor Coolant System sample i s taken v i a the C / R primary sampling valves o r the HRSR manual valves 2.2 An increase i n containment humidity i s i n d i c a t i v e o f an external leak from the Reactor Coolant System.

However, since t h i s i s l e s s s e n s i t i v e ( 2 gpm t o 10 gpm) an increase i n humidity due t o a leak i n t h e Reactor Coolant System should also show a s i g n i f i c a n t increase i n t h e other monitors.

2.3 I f e i t h e r the Containment Sump A Level Detection System Sump Pumps become inoperable, r e f e r t o Section 6.0.

Step 6.4.

NO. SP-36-082 DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION Reactor Coolant System Leak Rate I 'ITLE Check MAR 02 2006 I PAC33 2

of 14 I

SURVEILLANCE PROCEDURE DATE 3.0 L I M I T I N G CONDITIONS FOR OPERATION 3.1 The f o l l o w i n g L i m i t i n g Conditions are based on Technical Specifications TS 3.1.d.

1.

Any Reactor Coolant System leakage i n d i c a t i o n i n excess of 1 gpm s h a l l be the subject o f an i n v e s t i g a t i o n and evaluation i n i t i a t e d w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months o f the indication.

Any i n d i c a t e d leak s h a l l be considered t o be a real leak u n t i l i t i s determined t h a t no unsafe condition e x i s t s. I f the Reactor Coolant System leakage exceeds 1 gpm AND the source o f leakage i s NOT i d e n t i f i e d w i t h i n 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

then the Reactor s h a l l be placed i n the Hot Shutdown condition u t i l i z i n g normal operating procedures.

I f the source o f leakage exceeds 1 gpm AND i s NOT i d e n t i f i e d w i t h i n 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then the Reactor s h a l l be placed i n the Cold Shutdown condition u t i l i z i n g normal operating procedures.

2. I f the sources o f leakage have been i d e n t i f i e d i t i s evaluated t h a t continued operation i s safe, then operation o f the Reactor w i t h a t o t a l Reactor Coolant System leakage r a t e not exceeding 10 gpm s h a l l be permitted.

I f leakage exceeds 10 gpm, then the Reactor s h a l l be placed i n the Hot Shutdown c o n d i t i o n w i t h i n 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months u t i l i z i n g normal operating procedures.

I f t h e leakage exceeds 10 gpm f o r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the Reactor s h a l l be placed i n t h e Cold Shutdown condition u t i l i z i n g normal operating procedures.

3. Primary t o secondary leakage i s l i m i t e d t o 150 gallons per day through any one steam generator.

With tube 'leakage greater than t h e above l i m i t, reduce the leakage r a t e w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months o r be i n Cold Shutdown w i t h i n the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

4. If any Reactor Coolant leakage e x i s t s through a non-isolable f a u l t i n a Reactor Coolant System component ( e x t e r i o r w a l l o f the Reactor Vessel, piping, valve body, R e l i e f Valve leaks.

Pressurizer, Steam Generator Head. o r Pump Seal l e a k o f f ). then the Reactor s h a l l be shut down AND a cooldown t o the Cold Shutdown condition s h a l l be i n i t i a t e d w i t h i n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months o f detection.

.O GENERAL INSTRUCTIONS 4.1 None

.O EQUIPMENT REQUIRED 5.1 None

Reactor Coolant System Leak Rate I 'ITUE Check KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE c

DATE MAR 02 2006 PAGE 3

of 14 DATE 6.0 PROCEDURE NOTE:

This c a l c u l a t i o n i s performed d a i l y.

I f the PPCS i s i n service, l -

use Step 6.1.1; otherwise use Step 6.1.2.

NOTE:

Containment Sump A leakrate can JOJ be used for Mass Balance determination.

NOTE:

I f necessary, the s t a r t time o f the computer c a l c u l a t i o n may be adjusted t o a value less than 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

However, reducing the t e s t i n t e r v a l w i l l reduce the accuracy o f the calculation.

Minimum t e s t i n t e r v a l f o r the PPCS c a l c u l a t i o n i s 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

6.1 Mass Balance Leakrate Calculation 6.1.1 Cornouter Calculation

a.

On PPCS Main Menu, C L I C K on Applications Menu.

b.

On Applications Menu, C L I C K on On Demand RCS Leakage.

c.

VERIFY values are provided f o r a l l RCS Leakage data points.

d.

VERIFY appropriate value f o r VCT Level Over 56%.

e.

CLICK on Calculate and V E R I F Y p r i n t o u t o f RCS leakage c a l c u l a t i o n results.

f.

RECORD calculated RCS leak r a t e on D a t a Sheet 1 and ATTACH RCS leakage c a l c u l a t i o n p r i n t o u t t o D a t a Sheet 1.

g.

Mass Balance leakrate c a l c u l a t i o n i s negative. THEN PERFORM one o f the following:

3.

GO TO Step 6.1.1 and REPEAT Computer Calculation.

OR

2.

GO TO Step 6.1.2 and PERFORM Manual Calculation.

3.

GO TO Step 6.1.1.h and ACCEPT negative leakrate r e s u l t s.

CONTINUED

DOMINION ENERGY KEWAUNEE KEWAIJNEE POWER STATION SURVEILLANCE PROCEDURE NO.

SP-36-082 Reactor Coolant System Leak Rate Check DA!CE MAR 02 2006 I PAGE 4

of 14 DATE 6.1.1 CONTINUED

h.

RECORD leak r a t e i n the Control Room Log and on S h i f t Manager's status board.

i. fl Mass Balance leakrate calculation indicates t h a t leakage from Reactor Coolant System i s negative leakage i s greater than 0.2 gpm. THEN GO TO Step 6.3.

6.1.2 Manual Calculation

a.

Special Precautions

1.

Reactor Coolant System temperature should be s t a b i l i z e d and held constant f o r approximately one hour before s t a r t i n g the t e s t.

2.

Reactor Makeup System i s i n automatic.

A l l Makeup s h a l l go through the Boric Acid Blender.

3.

Reactor power should be s t a b i l i z e d and held constant (plus o r minus 51) f o r approximately one hour before s t a r t i n g the t e s t and f o r the duration o f the t e s t.

4.

Pressurizer temperature and pressure and Reactor Coolant System temperature f i n a l values s h a l l equal i n i t i a l p r e - t e s t values.

A change i n RCS temperature o f 1" F w i l l v a r y pressurizer l e v e l by approximately 1.51.

b.

RECORD i n i t i a l readings on D a t a Sheet 2.

c.

A f t e r a t l e a s t 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , V E R I F Y Reactor Coolant temperature and Pressurizer temperature and pressure are the same value as recorded i n i t i a l l y.

1.

Reactor Coolant temperature and Pressurizer temperature and pressure a r e the same as recorded i n Step 6.1.2.b.

THEN RECORD on D a t a Sheet 2 and Go Step 6.1.2.d.

CONTINUED

DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE NO. SP-36-082 Reactor Coolant System Leak Rate Check DA&

MAR &! 2006 I P m

5 of 14 DATE 6.1.2.c CONTINUED

2.

Reactor Coolant temperature o r Pressurizer temperature o r pressure have changed, THEN PROCEED as follows:

A.

ADJUST parameters t o value recorded i n Step 6.1.2.b.

6.

ALLOW conditions t o s t a b i l i z e (approx 15 minutes).

C.

RECORD f i n a l readings on D a t a Sheet 2.

d.

CALCULATE Reactor Coolant System leak r a t e using information recorded and the formula on D a t a Sheet 2.

e.

Mass Balance leakrate c a l c u l a t i o n i s negative. I@

PERFORM one o f the following:

1.

GO TO Step 6.1.2 and REPEAT Manual Calculation.

2.

GO TO Step 6.1.1 and PERFORM Computer Calculation.

3.

TO Step 6.1.2.f and ACCEPT negative leakrate results.

f.

RECORD leak r a t e i n the Control Room Log and on S h i f t Manager's status board.

g. E Mass Balance leakrate calculation indicates t h a t leakage from Reactor Coolant System i s negative OR leakage i s greater than 0.2 gpm. THEN GO TO Step 6.3.

DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEUANCE PROCEDURE NO. SP-36-082 Reactor Coolant System Leak Rate Check DATE MAR 02 2006 I PAGE 6 of 14 DATE 6.2 Containment Sump PumD Run (A-MDS-30) 6.2.1 Each time a CONTAINMENT SUMP A LEVEL H I G H (47031-0) a l a r m i s received, CALCULATE the corresponding leakrate w i t h i n containment from sump pump run h i s t o r y as follows:

a.

RECORD date, time. and pump A o r B f o r each Containment Sump A Pump run on D a t a Sheet 3.

b.

The volume o f Containment Sump A between the h i g h l e v e l a l a r m and the automatic pump shutoff i s 339.0 gallons

( H i - H i A l a r m i s 412.5 gallons).

c. CALCULATE leakage t o Containment Sump A using 339.0 gallons and the time between pump actuations ( t o nearest 1/10 of a minute).

d.

RECORD leakage on D a t a Sheet 3.

e.

RECORD pump run time t o detect pump degradation.

f.

Upon completion o f weekly Reactor Coolant System Leak Rate Test (Step 6.1).

ATTACH D a t a Sheet 3 t o the r e s u l t s of the t e s t.

g. E indicated leakage i s greater than 1 gpm. THEN PERFORM a Reactor Coolant System Mass Balance leakrate c a l c u l a t i o n per Step 6.1.

6.2.2 a sump pump has NOT run w i t h i n the past 23 days, THEN VERIFY CONTAINMENT SUMP A LEVEL H I G H (47031-0) a l a r m operable by performing one o r more o f the following: CPCR0158511

a.

DRAIN PRT t o Sump A using a l t e r n a t e method per N-RC-36B.

OR

b.

DRAIN RCDT t o Sump A as follows:

1.

OPEN RC-534/CV-31218, Rx Clnt Drain Tank To Cntmt Sump A.

LEVEL HIGH (47031-0) a l a r m actuates, THEN CLOSE RC-534.

2.

WHEN RCDT lowers t o 12% l e v e l CONTAINMENT SUMP A

DOMINION ENERGY KEWAUNEE mWAIJNEE POWER STATION SURVEILLANCE PROCEDURE NO.

SP-36-082 Reactor Coolant System Leak Rate Check DATE MAR 02 2006 I PAGE 7

of 14 DATE 6.3 I n v e s t i g a t i o n and Evaluation 6.3.1 Reactor Coolant System leakrate i s determined t o be negative, THEN PERFORM t h e following:

CPCR0077821

6. 3. 2 NOTE:

A-RC-36F may be used f o r reference.

a.

INVESTIGATE source o f inleakage.

b.

INITIATE an Action Request (AR) documenting the r e s u l t s of t h i s investigation.

I F Reactor Coolant System l e a k r a t e i s determined t o be greater than 0.2 gpm. THEN an i n v e s t i g a t i o n and evaluation s h a l l be s t a r t e d w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months o f the i n d i c a t i o n.

Document t h i s i n v e s t i g a t i o n using D a t a Sheet 4.

The f o l l o w i n g leak paths s h a l l be investigated:

a.

RCS Leak t o Containment

1.

Containment Sump A Run History. recorded on D a t a

2.

Containment P a r t i c u l a t e Monitor, R-11 o r R - 2 1.

Sheet 3 per Step 6.2.1.

trending.

3.

Containment Gas Monitor, R-12. trending.

4.

Containment Humidity Detector trending.

b.

RCS Leak t o Component Cooling System

1.

Comp Cooling Liquid Monitor. R-17. trending.

2.

Component Cooling Surge Tank l e v e l trending us,ng Operator Aids book.

c.

RCS Leak t o Steam Generators

1.

A i r Ejector Exhaust Mon

2.

S / G Blowdown L i q u i d Mon t o r, R-15. trending.

t o r, R-19. trending.

CONTINUED the

DOMINION ENERGY KEWAUNEE JCEWAUNEE POWER STATION SURVEILLANCE PROCEDURE NO.

SP-36-082 Reactor Coolant System Leak Rate

'ITLE Check

~~

~

~~~~

DATE MAR 02 2006 I PAGE 8 of 14 DATE 6.3.2 CONTINUED

d.

RCS Leak t o Waste Disposal System

1.

PRT l e v e l trending using the Operator Aids book.

2.

RCDT l e v e l trending using the Operator Aids book.

6.3.3 E NONE o f the leak paths l i s t e d above i n d i c a t e leakage from Reactor Coolant System, THEN t h e Technical Specifications are s a t i s f i e d : however, an i n v e s t i g a t i o n should be performed t o i d e n t i f y the source o f t h i s leakage external t o the Reactor Coolant System (e.g.

charging pump seal leakage).

6.3.4 The f o l l o w i n g items are i n d i c a t o r s o f p o t e n t i a l leakage sources external t o the Reactor Coolant System:

a.

Charging Pump Leak-off

b.

Changes i n Tank Levels 0 RCDT 0 CVC HUT 0 Waste HUT 0 Deaerated Drain Tank

c.

Sump Pump Run Times 0 Rx Cavity 0 Cntmt Sump A RHR Pump P i t Sump Pump 0 Waste A r e a Sump d.

Excessive Makeup t o VCT

e.

Valve Stem Leakoff Lines

f.

F i l t e r Vent and Drain Lines

g.

Demineralizer Vent and Drain Lines

h.

Visual Inspection o f Aux Building CONTINUED

Reactor Coolant System Leak Rate Check I

DATE MAR 02 2006 I PAGE 9

of 14 n

DOMlNION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE DATE 6.3.4 CONT I NU ED

i. Visual Inspection o f Containment

j. LO-13 Bellows Failure

k.

HRSR Drains to:

RHR Pump P i t Sump Sump Tank DDT 6.4 Containment Basement (592 0 EL) InsDection 6.4.1 E conditions o f Precaution 2.0.

Step 2.3 e x i s t, THEN PERFORM a weekly visual inspection f o r water i n the general area of Sump A and Sump 6.

6.4.2 COMPLETE an Action Request f o r each inspection noting r e s u l t s o f the inspection.

r0NTTNll01 TS 1 TSF

DOMlNlON ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE I

8.4 RCS Leakrate has been determined per t h i s procedure a t l e a s t once per I

I 7 days.

(Table TS 4.1-3 Item 8)

NO.

SP-36-082 Reactor Coolant System Leak Rate Check

~

DATE MAR 02 2006 I PAGE 10 of 14 OAT E INITIALS 7.0 PROBLEMS YESINO 7.1 Any problems encountered during t e s t ?

I N I T I A T E D I N A 7.2 IF yes, THEN INITIATE an Action Request (AR) per GNP-11.08.01, Action Request Process.

AR#

3.0 ACCEPTANCE C R I T E R I A 8.1 Any Reactor Coolant System leakrate o f l e s s than zero gallons per minute has been investigated as required by t h i s procedure.

8.2 Any Reactor Coolant System leakrate o f greater than 0.2 gallons per minute has been investigated as required by t h i s procedure.

8.3 Reactor Coolant System leakage meets the L i m i t i n g Conditions f o r Operation per Technical Specification 3.1.d.

1.0 REFERENCES

9.1 Technical Specifications 3.1.d 9.2 KAP 00-003466 9.3 PCR007782 9.4 KNPP Pumps And Valves IST Plan 9.5 Comtrac 91-205

NO.

SP-36-082

~

VCT Level (L0112A)

PRZR Press (P8023G)

Reactor Coolant System Leak Rate Check

~

DATE MAR 02 2006 I PAGE 11 of 14 DATA SHEET 1 REACTOR COOLANT LEAKAGE CALCULATION BY COMPUTER Parameter I

D a t a a t S t a r t o f Calculation D a t a a t End o f Calcul a t i on Time I

VCT Temp (T0140A)

VCT Press (P0139A)

~

~~

~

PRZR Level (L8015G) rAVG (TO4446 1

<MW

( G a l )

3A

( G a l )

ont.

E l. 626' Amb.

4ir Temp (15187)

DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE Reactor Coolant System Leakage 9 Pm I F leakage i s negative leakage i s greater than 0.2 gpm problems were encountered during t e s t, THEN DESCRIBE on an Action Request.

I F leakage i s negative NOTE:

Attach Containment Sump Pump D a t a Sheet.

Commen t s :

PERFORMED BY DATE S H I FT MANAGER DAT F SPVR NUCLEAR SHIFT OPERATIONS DATE leakage i s greater than 0.2 gpm. THEN REFER t o Step 6.3.

~

I

ontainment iumidi t y (%) (41517)

Difference i n time:

Hrs-Mi n-

I DOMINION ENERGY KEWAUNEE KJtWAUNEE POWER STATION SURVEILLANCE PROCEDURE NO. SP-36-082 Reactor Coolant System Leak Rate Check DATE MAR 02 2006 PAGE 12 of 14 I

DOMINION ENERGY KEWAUNEE KJtWAUNEE POWER STATION SURVEILLANCE PROCEDURE NO. SP-36-082 Reactor Coolant System Leak Rate Check DATE MAR 02 2006 PAGE 12 of 14 hCT Level "F(1)

Required:

No change P0429A. P0430A P0431A. P0449A (Computer p r i n t o u t 1 T I -425

-0R-T I -424 15187 41517 Pressurizer Pressure (2)

Pressurizer Temperature Cont E l 626' Amb A i r Temp Containment Humidity Time o f t e s t psig( 1) Required:

No change DATA SHEET 2 MANUAL REACTOR COOLANT SYSTEM LEAKAGE CALCULATION I

(Step 6.1.2-b:

I n i t i a l Readin:

gal I n i t i a l B =

ga 1 44560/

RMW Batch YIC-111 Integrator BA Batch I Integrator 44559/

YIC-110 LIT-141 LIT-112 L I -426 L I -427 L I -428 Pressurizer Level Reactor Cool ant Temperature

( T a w ) (2)

T0400A. T0401P T0420A. T0421A

( Computer p r i n t o u t )

psig(1)

Required:

No change OF N / A I

N/A O F N/A I

N/A A =

minutes (3)

Leak Rate (gpm) = 1.3798 + 1.41C - 17.390 - 43.29F =

9 Pm A

leakage i s negative leakage i s greater than 0.2 gpm problems were encountered during t e s t, THEN DESCRIBE on an Action Request.

I F leakage i s negative NOTE:

Attach Containment Sump A Pump D a t a Sheet.

Comments :

PERFORMED BY DATE SHIFT MANAGER DATE SPVR NUCLEAR SHIFT OPERATIONS DATE leakage i s greater than 0.2 gpm. THEN REFER t o Step 6.3.

I

DOMINION ENERGY KEWAUNEE KEWAUNEE POWER STATION SURVEILLANCE PROCEDURE NO.

SP-36-082 Reactor Coolant System Leak Rate Check DATE MAR 02 2006 I PAGE 13 of 14 DATA SHEET 3 CONTAINMENT SUMP A PUMP RUN HISTORY Gallons Time Since Last GPM Pump Pum ed Pump Actuation 11 Run Time CNTMT Pump Date Time H i A Y a r m (minl (min)

HUM I

~~

~

339.0 339.0 339.0 1

I 1

I I

339.0 I

I Hi-Hi = 412.5 G a l I N IT1 ALS I

I F l e v e l detection system o r sump pumps f o r Containment Sump A are inoperable. THEN R F E R t o Step 6.4.

Yas e i t h e r sump pump run w i t h i n the past 23 days?

Yes No -

If n e i t h e r sump pump has run w i t h i n the past 23 days, THEN REFER t o Step 6.2.2.

jump leakage l e s s than o r equal t o 1 gpm?

Yes No -

sump leakage i s greater than 1 gpm AND Technical S p e c i f i c a t i o n 3.1.d i s applicable, rHEN DESCRIBE on an Action Request AND REFER t o Step 6.2.1.9.

IPCR0003501 Jpon completion o f the leak r a t e t e s t, ATTACH t h i s D a t a Sheet t o the r e s u l t s.

arry over l a s t pump run t o the new D a t a Sheet 3.

PVR NUCLEAR SHIFT OPERATIONS DATE

DOMINION ENERGY KEWAUNEE LOCATION INDICATION VALUE To Containment Leakage t o sump 9 Pm R-11 C Pm Humidity (41517)

Cont. E l. 626' Amb.

R-12

Pm A i r Temp (15187)

O F

To Component R-17 CPm Cool i ng Surge Tank Level Change 9 Pm To Steam R-15 C Pm

enera t o r R-19 C Pm ro Waste PRT Level Change gPm 3 i sposal RCDT Level Change gPm I NO. SP-36-082 NORMAL gPm C Pm CPm O

F N/A Z

CPm NA CPm CPm NA NA Reactor Coolant System Leak Rate KEWAUNEE POWER STATION 1"

Check SURVEILLANCE PROCEDURE I DATE MAR 02 2006 I PAGE 14 of 14 DATA SHEET 4 INVESTIGATION AND EVALUATION INITIAL I N D I C A T I O N OF LEAKRATE >0.2 GPM DATE TIME LE AKRAT E 9 Pm INVESTIGATION STARTED:

DATE TIME INVESTIGATION OF LEAKAGE:

DESCRIPTION OF LEAK:

SOURCE :

EFFECT ON PLANT OPERATION :

PLANT OPERATION MAY SAFELY CONTINUE:

YES NO PERFORMED BY DATE APPROVED BY

DATE Director Nuclear Operations & Maintenance, Operations Manager.

Spvr Nuclear S h i f t Operations, or S h i f t Manager

WISCONSIN PUBLIC SERVICE COapORATION KEWAUNEENUCLEARPOWERPLANT EMWGENCY OPEdRATING PROCEDURES A'ITACHMRNT B RBSPONSE TO ABNORMAL #2 SEAL LEAKOFF FLOW

1.

NOTE:

Determine #2 Seal Leakoff Flow by monitoring t h e change i n RCDT l e v e l.

High standpipe l e v e l i n d i c a t e s excessive #2 seal leakage.

l e v e l i n d i c a t e s excessive #3 seal leakage.

Low standpipe 2.

Perform 47015-1(47015-L). RXCP A(B) Standpipe High/Low t o determine i f standpipe l e v e l i s high o r low.

NOTE:

Total Seal Leakoff Flow equals #1 Seal Leakoff Flow plus 82 Seal Leakoff Flow.

A-RC - 36C NO -

TITLE ABNORMAL RXCP OPERATION DATE OCT 06 2005 PAGE 17 of 18

3.

4.

5.

I F Total Seal Leakoff Flow is greater than 8.0 gpm. THEN perform an immediate Z C P shutdown:

a.

b.

Stop affected RXCP(s).

c.

Close PS-lA(1B). Pzr Spray, i n affected loop.

d.

Ip reactor is c r i t i c a l. THEN t r i p r e a c t o r AND GO TO 6-0.

WHEN RXCP has come t o a complete stop. THEN c l o s e CVC-207A(B). #1 Seal Leakoff Isol.

e.

Refer t o TS 3.1.d f o r RCS leak rate l i m i t a t i o n s.

I F #2 Seal Leakoff Flow greater than 1.1 gpm. THEN perform an orderly E C P shutdown :

a.

b.

c.

Close PS-lA(1B). Pzr Spray. i n affected loop.

d.

Refer t o TS 3.1.d f o r RCS leak rate l i m i t a t i o n s.

I F #2 Seal Leakoff Flow greater than 0.5 gpm. THEN perform t h e following:

a.

Monitor pump and seal indications.

b.

Trend parameters on PPCS.

c.

I n i t i a t e normal p l a n t shutdown per N-0-03.

Stop affected RXCP(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Consult with s t a t i o n management t o determine i f affected RXCP(s) should be stopped.

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 95 Accept D as a correct answer.

N-FH-53CLD, Refueling Daily Checklist, has you check R-12 operating. If the Checklist or an item on the Checklist is not verified, you cannoffshould not move fuel. The action is conservative.

2. Resolution and Response Accept either C or D as correct.

Either remove the question from the bank or change selection D to be an incorrect action.

The ODCM requires either R-12 or R-21 to be OPERABLE and addresses action for a purge in progress, but does not specifically address Fuel Handling. Likewise the normal procedure N-RM-45, Radiation Monitoring System, addresses the removal of R-12 from service, and ensuring R-21 is operating and aligned to sample the appropriate location.

Using Conduct of Operations and the Standard DNOS - 0101 for Nuclear Safety and Conservative Decision Making, it is not unreasonable to stop fuel movement while verifying the status of required components. The operations guidance for fuel handling, N-FH-53-CLC, Pre-Refueling Checklist, and N-FH-53-CLD, Refueling Daily Checklist, both require R-12 and R-21 to be operating during fuel movement. There is no specific guidance on actions to take if one of the monitors fails. Therefore, Conservative Decision Making is applied, and fuel movement should be stopped. This is also supported by the General Notes of RF-03.01, Fuel Movement During a Refueling Outage, 2.7 which states, in part, Each member of the refueling team needs to understand that they have the authority to stop refueling activities to resolve an issue.

3. Supportina Information ODCM Table 3.2, 1.b N-RM-45, 4.3.10 N-FH-53-CLC, 2.8.3.a & b.

N-FH-53-CLD, 2.9.3.a & b.

GNP-03.30.02, 6.1 DNOS - 01 01, Expectations & Standards RF-03.01, 2.7

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Page 15

TABLE 3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Page 1 of 2)

Instrument

1.

Noble Gas Activity Monitor

a.

R-13 or R-14

- Waste Gas Holdup System

- Auxiliary Building Ventilation

- Containment Purge 2 line (auto-isolation)

System (auto-isolation)

b.

R-12 or R-21

- Containment purge-36 duct (auto-isolation)

c.

R-15 Condenser Evacuation System Radioiodine & Particdate Samplers

a.

Containment Building Vent (R-21)

b.

Auxiliary Building Vent (R-13 or R-14)

Sampler Flow Rate Measuring Devices

a.

Containment Building Vent Sampler

b.

Auxiliary Building Vent Sampler (R-2 1)

(R-13 or R-14)

Minimum Channels I

1 1

Applicability Action -

4 5

6 6

5 7

7 8

8

Atalltimes 3-18 REV. 9 12l0212005

TABLE 3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Page 2 of 2)

Action 4 -

Action 5 -

4ction 6 -

4ction 7 -

4ction 8 -

TABLE NOTATIONS With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tanks contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.4.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may 3-19 REV. 9 12/02/2005

DOMINTON ENERGY KEWAUNEE I NO. N-RM-45 TITLE Radiation Monitoring System DATE APR 12 2006 P

m 19 of 27 KICWlLDNEE POWERSTATloN OPERATING PROCEDURE 4.3.8 CONTINUED

b.

there i s new f u e l i n t h e New Fuel P i t, THEN REQUEST Radiation Protection set up portable monitor with an audible a l a r m s e t a t 15 mr/hr.

C.

d.

9.

R-

a.

b.

C.

d.

e.

f.

10.

POSITION keyswitch t o OFF.

PERFORM d a i l y source check on portable monitor.

, Conta i nment P a r ti cul ate:

IF Containment PurgdVent i s i n progress. THEN VERIFY R-12 OR R-21 operating and aligned t o sample stack. OR STOP Purge/Vent.

I F Containment Purge/Vent i s NOT i n progress, THEN ALIGN R-21 t o sample Containment per Step 4.2.10.

NOTIFY Radiation Protection.

REFER t o Tech Spec 3.1.d.

POSITION R-11/12 Pump Control switch t o OFFIRESET.

POSITION Keyswitch t o KEYPAD t o defeat automatic functions.

NOTE:

P o s i t i o n i n g Keyswitch t o o f f w i l l actuate automatic functions.

9.

R-12. Containment Gas:

a.

IF automatic functions are required. THEN POSITION Keyswi t c h t o OFF.

fi Containment Purge/Vent i s i n progress, JHEN VERIFY R-21 operating and aligned t o sample stack, STOP Purge/Vent.

IF Containment Purge/Vent i s Ji0-J i n progress, THEN ALIGN R-21 t o sample Containment per Step 4.2.10.

b.

c.

NOTIFY Radiation Protection.

d.

REFER t o Tech Spec 3.1.d and ODCM.

CONTINUED

DOMINION ENERGY KEWAUlWE II;EWAUNEE POWER STATION OPERATING PROCEDURE NUCLEAR YES SAFETY RELATED 0 NO Michael swey REVIEWED BY PORC REVIEW YES SRO APPROVAL OF YES REQU I RED TEMPORARY CHANGES 0 NO REQU I RED 0 NO

~

NO. N-FH-53-CLC -1 k V S

~~

TIT=

Pre-Refueling Checklist DATE JUN 13 2006 I PAGE 1

of 7 James Langex APPROVED BY FIRST SECOND OPER OPER

~-

1.0 PLANT REQUIREMENTS 1.1 N-FH-53-CLA Or one door i n ea N-FH-53-CLB complete. except a t l e a s t h personnel a i r l o c k s h a l l be capable o f being closed i n 30 minutes o r l e s s w i t h administrative controls i n place t o ensure closure.

I n addition, a t l e a s t one door i n each personnel a i r l o c k s h a l l be closed when the reactor vessel head o r upper i n t e r n a l s are l i f t e d.

(TS 3.8) 1.2 RCS boron concentration s h a l l be v e r i f i e d greater than o r equal t o COLR Specified Refueling Boron Concentration p r i o r t o i n i t i a l movement and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during f u e l movement.

Reactor Cavity Boron Concentrat RHR System Boron Concentration:

COLR Refueling Boron Concentrat on :

P Pm PPm on :

PPm 1.3 SFP boron concentration shall be v e r i f i e d greater than or equal t o COLR Specified Refueling Boron Concentration p r i o r t o i n i t i a l movement and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during f u e l movement.

SFP Boron Concentration:

COLR Refueling Boron Concentrat 1.4 Reactor has been v e r i f i e d subcr 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months . [CAP122551 PPm on :

PPm t i c a l for greater than 1.5 RCS and Spent Fuel Pool Temperature s h a l l be v e r i f i e d greater than o r equal t o 50°F p r i o r t o s t a r t o f f u e l movement.

CCA0192021 COMPLETED 2 REQUIRED 2 REQUIRED V E R I F I E D V E R I F I E D

DOMINION ENERGY KEWAUNEE OPERATING PROCEDURE N - FH CLC I No-DATE JUN 13 2006 PAGE 2

of7 KEWAUNEE POWJER STATION I TITLE Pre-Refuel i ng Check1 i s t DATE FIRST SECOND OPER OPER 2.0 SYSTEM EQUIPMENT STATUS SET UP 2.1 Reactor Engineering core status system set up i n Control Room.

2.2 Residual Heat Removal System:

1.

A t l e a s t one t r a i n o f RHR System operable.

OPERABLE

2.

RHR pump suction temperature less than or equal t o

~ 1 4 0 "

F 140" F.

2.3 A u x i l i a r y B u i l d i n g Special V e n t i l a t i o n system operable.

OPERABLE NOTE:

Control Room Post-Accident Recirculation System may be considered if one t r a i n i s operating operable i f both t r a i n s are operable i n r e c i r c mode, provided i t s emergency D/G i s a l s o operable t o support f u e l handling accident analysis assumptions.

2.4 Control Room Post-Accident Recirculation System OPE RAB LE Operable.

2.5 Boric Acid I n j e c t i o n Flow Path:

1.

System p i p i n g and valves are operable f o r a t l e a s t one flow path f o r b o r i c acid i n j e c t i o n t o RCS.

0 P E RAB LE 2.6 Reactor Cavity Level i s greater than o r equal t o 64%.

264%

2.7 Plant Process Computer:

1.

Point L9053A o r L9054A. Refueling Water Level A ( B ) WR:

N I A

a.

VERIFY L9053A or L9054A on SCAN.

SCAN N I A

b.

VERIFY a l a r m l i m i t s f o r selected p o i n t L9053A

>64%

or L9054A established per PPCS RHR Mid-Loop Limits screen are greater than 64%.

2.

Point N8031G. SR N-31 Counts (N8032G. SR N-32 Counts 1 :

a.

VERIFY N8031G (N8032G) on SCAN.

SCAN N / A CONTINUED

~

DOMINION ENERGY KEWAUNEE KEWAUNEE POWEP STATION OPERATING PROCEDURE

~~~ I NO. N-FH-53-CLC I TITLE Pre-Refuel i n g Check1 i s t I DAm-JUN 13 2006 I PAGE 3

of 7 DATE FIRST SECOND OPER OPER 2.7.2 CONTINUED

b. IF reactor vessel contains f u e l. JtJJ CPSINA N I A CONTACT Reactor Engineering t o provide an estimate o f expected source range counts.

c.

V E R I F Y Low A l a r m L i m i t i s 0.4 Baseline count VERI F I ED N I A rate.

d.

VERIFY High A l a r m L i m i t i s 1.7 Baseline count VERIFIED N I A rate.

e.

Single Point Change required.

CONTACTEDINA N I A THEN CONTACT Nuclear Computer Group.

3.

computer i s available for trending. THEN APPLIESINA N I A TREND the following points (PPCS Operations -

Protected 1. Group 5) every 10 minutes:

Point I D S Points F0626A GOO1 1G GOO126 T0627A T0630A R81 OOG N8031G N8032G LO1 12A Lo Head Trn A RHR To RCS R - 1 1 Containment Part R - 1 2 Containment Gas RHR Hx Out Loop Hdr Temp RHR Pump Suct Hdr Temp RCS Boron Conc Calc SR N-31 Counts SR N-32 Counts Volume Control Tank Lvl

DOMITUON ENERGY KEWAUNEE I NO. N-FH-53-CLC I

KEWAUNEE POwEa STATION TITLE Pre-Refueling Checklist OPERATING PROCEDURE I DAm JUN 13 2006 I PAGE 4

of 7 I DATE 2.8 Radiation Monitoring:

1.

R-2/81041. Containment Area Monitor OPERATING.

2.

R-5/81042. Fuel Handling Area Monitor OPERATING.

3.

PERFORM the following:

a.

VERIFY R-12/81049. Containment Gas.

OPERATING.

b.

VERIFY R-21/81058. Containment Vent, OPERATING.

c.

VERIFY Containment Vent I s o l a t i o n w i l l actuate on h i g h r a d i a t i o n signal as follows:

CPCR0177871

1. IF Reactor B u i l d i n g Vent i s NOT operating.

THEN OPEN t h e f o l l o w i n g and PLACE i n AUTO.

A.

RBV-l/CV-31125. Cntmt Purge/Vent Supply Valve A B.

RBV-2/CV-31126. Cntmt Purge/Vent Supply Valve B C.

RBV-3/CV-31124, Cntrnt Purge/Vent Exhaust Valve A D.

RBV-4/CV-31123, Cntmt Purge/Vent Exhaust Valve B

2.

PERFORM functional t e s t of R-12 per N - RM - 45.

3.

VERFIY REV-2 CLOSED.

4.

V E R F I Y RBV-3 CLOSED.

5.

PERFORM functional t e s t of R-21 per N-RM-45.

6.

V E R F I Y RBV-1 CLOSED.

7.

VERFIY RBV-4 CLOSED.

FIRST SECOND OPER OPER OPERATING OPERATING OPE RAT I NG OPE RAT I NG APPLIES/NA 0 PEN / AUTO OPEN/AUTO OPEN/AUTO OPEN/AUTO P E RFO RMED CLOSED CLOSED PERFORMED CLOSED CLOSED CONTINUED

DOMINION ENERGY KEWAUNEE i

I KEWAUNEE POWEB STATION OPERATING PROCEDURE DATE NO.

N-FH-53-CLC TITLE Pre-Refuel i n g Check1 i s t DA!l!E JUN 13 2006 I PAL;E 5 o f 7 2.8.3.c CONTINUED

8. IF Reactor Building Vent was operating.

THEFl SHUT DOWN Reactor Building Vent per N-RBV-18B.

9. IF operation o f Reactor Building Vent i s desired. THEN START Reactor Building Vent per N-RBV-18B.

4.

R-23/81061. Control Room Vent Monitor

a.

R-23. OPERATING.

b.

VERIFY Control Room Post-Accident Recirxulation System w i l l actuate on high r a d i a t i o n signal, GO TO N-RM-45.

2.9 Nuclear Instrumentation:

1.

Two Source Range Detectors are monitoring Neutron f l u x and each provide visual i n d i c a t i o n i n Control Room.

2.

One Source Range Detector shall have an audible count r a t e i n containment.

3. IF any fuel i s i n the reactor vessel, THEN Source Range Detectors' High Flux a t Shutdown a l a r m set a t approximately 3 times t h e i r e x i s t i n g count rate.

4. IF reactor vessel contains f u e l, THEN Source Range Detectors' High Flux a t Shutdown a l a r m set a t value established during core off-load.

5.

High Flux a t Shutdown a l a r m and Containment a l a r m test:

a.

POSITION Level T r i p switch t o BYPASS.

b.

ROTATE Level Adjust potentiometer f u l l y counter-clockwise.

CONTINUED FIRST OPER SHUT DOWN/NA STARTED/NA OPERATING V E R I F I ED CONFIRMED CONFIRMEO ALARM SET/NA ALARM SET/NA BYPASS ROTATED SECOND OPER

DOMINION ENERGY KEWAUNEE I NO. N-FH-53-CLC TITLE Pre-Refueling Checklist I

KEWAUNEE POwEa STATION OPERATING PROCEDURE I MTE JUN 13 2006 I PAGE 6

of 7 2.9.5 CONTINUED

c.

POSITION Operat ADJUST.

DATE on Selector switch t o LEVEL

d.

ROTATE Level Adjust potentiometer clockwise u n t i l High Flux a t Shutdown alarm actuates.

e.

VERIFY alarm i n containment actuated.

f.

WHEN High Flux a t Shutdown a l a r m t e s t complete. THEN PERFORM the following:

1.

ROTATE Level Adjust Potentiometer f u l l y counterclockwise.

2.

POSITION Operation Selector Switch t o NORMAL.

3.

POSITION Level T r i p switch t o NORMAL.

2.10 VERIFY Containment Evacuation A l a r m on MCC A operable.

2.11 Communications established between Control Room, Containment. and Spent Fuel Pool.

2.12 Senior Reactor ODerator i n r e f u e l i n g area t o observe f u e l movement.

2.13 Radiation monitor

1.

Radiation mon on Manipulator Crane OPERATING.

t o r s e t t o a l a r m a t 10 R/hr.

2.14 Fuel Transfer System gate valve open.

2.15 A u x i l i a r y Building Crane i n t e r l o c k operable.

2.16 Spent Fuel Pool:

1.

SFP temperature less than 125°F.

2.

SFP cooling system operating.

FIRST SECOND OPER OPER LEVEL ADJUST ROTATED ACTUATED ROTATED NORMAL NORMAL OPERA8 LE ESTABLISHED CONFIRMED OP ERAT I NG SET 10 R/HR OPEN 0 P ERA B LE

<125" F OPERATING

DOMINION ENERGY KEWAUNEE KEWAUNEE FOWEE STATION OPERATING PROCEDURE DATE 2.17 Spent Fuel Pool Sweep System:

1.

Exhaust Fans ( 2 ) operating.

2.

Charcoal F i l t e r s (2) i n service (Bypass CLOSED).

3.

SP-17-126. SFP Sweep System F i l t e r Testing, has been s a t i s f a c t o r i l y completed.

2.18 A t l e a s t two Containment Fan Coil U n i t fans OPERATING (SW NOT required).

3.0 MONITORING AND ALARM REQUIREMENTS 3.1 Containment I s o l a t i o n Active Status Panel operable.

3.2 Sequenti a1 Events Recorder operable.

4.0 REMOTELY OPERATED AND AUTOMATIC VALVES 4.1 NONE 5.0 LOCAL VALVE POSITIONS 5.1 NONE NO. N-FH-53-CLC TITLE Pre-Refueling Checklist MTE JUN 13 2006 PAGE 7

of7 FIRST SECOND OPER OPER 0 PE RAT I NG I N SERVICE COMPLETED OP ERAT I NG OPERABLE OPERABLE Nf A Performed By I

I n i t Date Performed By f

I n i t Date Performed By I

I n i t Date S h i f t Manager Date SPVR Nuclear S h i f t Ops Date Name ( P r i n t )

S i gna t u r e Name ( P r i n t )

Signature Name ( P r i n t )

S i gna t u r e (PrintISign)

(Print/Sign)

WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNFX NUcLeABpowERPLhlyT

~

~~

NUCLEAR YES SAFETY RELATED OPERATING PROCEDURE PORC REVIEW REQUIRED Mark Larger RgvIgwED BY YES 0 NO SRO APPROVAL OF YES TEMPORARY CHANGES REQUI RED 0 NO NO. N-FH-53-CLD I -

P

~

TITLE Refueling Daily Checklist DATE OCT 11 2005 I P m

1 of 10 Je5ey Simon APPROVED BY I

DATE 1.0 PLANT REQUIREMENTS 1.1 N-FH-53-CLC has been completed.

1.2 Reactor Cavity and Residual Heat Removal System boron concentration s h a l l be v e r i f i e d greater than or equal t o 2500 ppm every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during f u e l movement.

1.3 Spent Fuel Pool boron concentration shall be v e r i f i e d greater than o r equal t o 2500 ppm every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during f u e l movement.

1.4 RCS and Spent Fuel Pool Temperatures s h a l l be v e r i f i e d g r e a t e r than or equal t o 50°F during fue movement.

CCA0192021 2.0 SYSTEM EQUIPMENT STATUS 2.1 RECORD Tagout Control Sheet number for l o c a l valves.

2.2 Residual Heat Removal System:

1.

A t l e a s t one t r a i n o f RHR System operable.

2.

RHR pump suction temperature less than or equal t o 240°F.

2.3 A u x i l i a r y Building Special V e n t i l a t i o n System operabl e.

2.4 Control Room Post-Accident Recirculation System Operable.

Control Room Post-Accident Recirculation System may be considered operable i f both t r a i n s a r e operable provided i t s emergency D/G i s also operable t o support f u e l handling accident analysis assumptions.

i f one t r a i n i s operating i n r e c i r c mode,

-~

FIRST SECONC OPER OPER

-~

COMPLETED V E R I F I E D VERIFIED VERI FI ED RECORDED 0 PERAB LE 4 4 0 ° F OP E RAB LE OPERABLE

WISCONSIN PUBLXC SERVICE CORpoRATIoly KEWAUNEENUCLEARPOWERPW OPERATING PROCEDURE NO. N-FH-53-CLD TITLE Refueling D a i l y Checklist

~~

DATE OCT 11 2005 I PAGF, 2

of 10 DATE 2.5 Boric Acid I n j e c t i o n Flow Path:

1.

System p i p i n g and valves are operable f o r a t l e a s t one f l o w path f o r b o r i c acid i n j e c t i o n t o RCS.

2.6 Reactor C a v i t y leakage t o Containment Sump C:

1.

RECORD t o t a l amount o f leakage Total gal

2.

RECORD l a t e s t leak r a t e 9 Pm 2.7 Reactor C a v i t y Level greater than o r equal t o 64%.

2.8 Plant Process Computer:

1. IF computer i s available f o r trending. THEN TREND the following points (PPCS Operations -

Protected 1. Group 5) every 10 minutes:

Poi nt I Ds Points F0626A Lo Head Trn A RHR To RCS G O O l l G R-11 Containment P a r t GOO12G R-12 Containment Gas T0627A RHR Hx Out Loop Hdr Temp T0630A R8100G RCS Boron Conc Calc RHR Pump Suct Hdr Temp N8031G SR N-31 Counts N8032G SR N-32 Counts L0112A Volume Control Tank Lvl FIRST SECOND OPER OPER

-~

OPERA6 LE RECORDED APPLIES/NA N / A

WlsCoNSIN PUBLIC SERVICE CORPORATION I

(

E w

A

~

E

~

P

~

I TITLE Refueling Daily Checklist NO. N-FH-53-CLD I

OPERATING PROCEDURE DATE DA!CE OCT 11 2005 PAGE 3

of 10 2.9 Radiation Monitoring:

1.

R-2/81041 Containment Area Monitor OPERATING.

2.

R-5/81042 Fuel Handling Area Monitor OPERATING.

PERFORM the following:

a.

VERIFY R-12/81049. Containment Gas, OPERATING.

4 OPERATING.

b.

VERIFY R-21/81058. Containment Vent.

c.

VERIFY Containment Vent I s o l a t i o n w i l l actuate on high r a d i a t i o n signal as follows:

C PCROl7787 1

1.

2.

3.

4.

5.

6.

7.

I F Reactor Building Vent i s )JOT operating, THEN OPEN the following and PLACE i n AUTO.

A.

RBV-l/CV-31125. Cntmt Purge/Vent Supply V a l v e A Supply V a l v e B Exhaust Valve A

6.

RBV-Z/CV-31126. Cntmt Purge/Vent C.

RBV-3/CV-31124. Cntmt Purge/Vent

0.

RBV-4/CV-31123. Cntmt Purge/Vent Exhaust Valve B PERFORM functional t e s t o f R-12 per N-RM-45.

V E R F I Y RBV-2 CLOSED.

VERFIY RBV-3 CLOSED.

PERFORM functional t e s t o f R - 2 1 per N - RM-45.

VERFIY RBV-1 CLOSED.

VERFIY RBV-4 CLOSED.

CONTINUED FIRST SECOND OPER OPER 0 PE RAT I NG OPERATING OPE RAT I NG 0 P E RAT I N G APPLIES/NA 0 PEN /AUTO OPEN / AUTO OPEN/AUTO OPEN /AUTO PERF 0 RMED CLOSED CLOSED PERFORMED CLOSED CLOSED

WISCONSIN PUBLIC SERVICE CORPORATION

~ W A U N J t E N U ~ P O W E R P W N T NO. N-FH-53-CLD TITLE Refueling D a i l y Checklist DATE OPERATING PROCEDURE 2.9.3.c CONTINUED DATE OCT 11 2005 PAGE 4

of 10

8. TF Reactor Building Vent was operating.

THEN SHUT DOWN Reactor B u i l d i n g Vent per N-RBV-188.

9.

operation o f Reactor B u i l d i n g Vent i s desired, THEN START Reactor B u i l d i n g Vent per N - RBV-188.

4.

R-23/81061 Control Room Vent Monitor OPERATING.

2.10 Nuclear Instrumentation:

1.

Two Source Range Detectors are monitoring Neutron f l u x and each provide visual i n d i c a t i o n i n Control Room.

2.

One Source Range Detector s h a l l have an audible count r a t e i n containment.

3.

any f u e l i s i n the reactor vessel, rHEN Source Range Detectors' High Flux a t Shutdown a l a r m set a t approximately 3 times t h e i r e x i s t i n g count rate.

4.

reactor vessel contains NO f u e l,

THEN Source Range Detectors' High Flux a t Shutdown a l a r m set a t value established during core o f f - 1 oad.

5.

High Flux a t Shutdown a l a r m and Containment a l a r m t e s t :

a.

POSITION Level T r i p switch t o BYPASS.

b.

ROTATE Level Adjust potentiometer f u l l y counter-clockwise.

c.

POSITION Operation Selector switch t o LEVEL ADJUST.

d.

ROTATE Level Adjust potentiometer clockwise u n t i l High Flux a t Shutdown a l a r m actuates.

FIRST SECONC OPER OPER SHUT DOWN/NA STARTED/NA OPERATING CONFIRMED CONFIRMED ALARM SET/NA ALARM SET/NA BYPASS ROTATED LEVEL ADJUST ROTAT ED CONTINUED

WISCONSJN PUBLIC SERVICE CORPORATION KEWAUNEE NDCLEARPIlWERPWNT OPERATING PROCEDURE 2.10.5 CONTINUED

-~

NO. N-FH-53-CLD TITLE Refueling D a i l y Checklist DATE OCT 11 2005 I PAGE 5 of 10 OAT

e.

VERIFY a l a r m i n containment actuated.

f.

WHEN High Flux a t Shutdown a l a r m t e s t complete, THEN PERFORM the fo11 owi ng:

1.

ROTATE Level Adjust Potentiometer f u l l y counterclockwise.

2.

POSITION Operation Selector Switch t o NORMAL.

3.

POSITION Level T r i p switch t o NORMAL.

2.11 VERIFY Containment Evacuation A l a r m on MCC A operable.

2.12 Communications established between Control Room, Containment, and Spent Fuel Pool.

2.13 Senior Reactor Operator i n r e f u e l i n g a r e a t o observe f u e l movement.

2.14 Radiation monitor on Manipulator Crane OPERATING.

2.15 Steam Generator Nozzle Dam leakage acceptable.

2.16 Shield Building Annulus Gates:

1.

V E R I F Y gate a t Personnel Airlock CLOSED and LOCKED.

2.

VERIFY gate a t Emergency A i r l o c k CLOSED and LOCKED.

2.17 A u x i l i a r y B u i l d i n g Crane i n t e r l o c k operable.

FIRST SECOND OPER OPER

~-

ACTU AT ED ROTAT ED NORMAL NORMAL OPERABLE ESTABLISHED CONFIRMED OPERATING ACCEPTABLE CLOSE0 and LOCKED CLOSED and LOCKED OPERABLE

NO. N-FH-53-CLD WISCONSIN PUBLIC SERVICE CORPORATION.

OPERATING PROCEDURE

~ W A ~ E N I J C L E A R ~ O W E R P L A ~ J T I TITLE Refuel i n g D a i l y Check1 i s t DATE OCT 11 2005 PAGE 6

of 10 DATE 2.18 Containment Vessel F l e x i b l e Seals:

1.

Containment Vessel Pressurization t e s t :

a.

Penetration (42N):

Loop Seal water l e v e l i n NORMAL range.

OR 0 Fiber Optic Cable Penetration Seal OR Contai nment Vessel B1 i nd F1 ange INSTALLED.

B l i n d Flange Test Conn CAPPED.

b.

Refueling Cable Penetration (43N):

0 Loop Seal water l e v e l i n NORMAL range.

OR 0 Containment Vessel B l i n d Flange INSTAL AND 0 B l i n d Flange Test Conn CAPPED.

2.19 Spent Fuel Pool Sweep System:

1.

Exhaust Fans ( 2 ) OPERATING ED.

2.

Charcoal F i l t e r s (2) i n service (Bypass CLOSED) 2.20 A t l e a s t two Containment Fan Coil U n i t Fans operating.

(SW NOT required) 3.0 MONITORING AND ALARM REQUIREMENTS 3.1 Containment I s o l a t i o n Active Status Panel operable.

3.2 Sequential Events Recorder operable.

FIRST SECOND OPER OPER NORMAL INSTALLED INSTALLED CAPPED NORMAL INSTALLED CAPPED OP ERAT I NG I N SERVICE 0 PE RAT I NG OPERABLE 0 PE RAB LE

1 NO. N-FH-53-CLD WEECONSIN PUBLIC SERVICE CORPORATION KEWAUNEENUCLEARPOWERPLANT TITLE Refueling Daily Checklist I

OPERATING PROCEDURE I DATE OCT 11 2005 I PAGE 7

of 10 DATE 4.0 REMOTELY OPERATED AND AUTOMATIC VALVES NOTE:

Each l i n e t h a t has automatic Containment I s o l a t i o n valves s h a l l have an operable automatic i s o l a t i o n valve OR a closed i s o l a t i o n valve.

For inoperable valves denote the associated Tagout number i n the "Second Oper" i n i t i a l s column.

4.1 Mechanical Console C NG-107/CV-31253. Nitrogen Supply t o S I Accumulators CC-601A/MV-32084. Component Cooling t o RXCP A CC-612A/MV-32086.

RXCP A Component Cooling Return I s o l CC-601B/MV-32085. Component Cooling t o RXCP B CC-612B/MV-32087. RXCP 6 Component Cooling Return I s o l CC-653lMV-32082. Excess Letdown Hx Comp Cooling Return 4.2 Mechanical Console.E LD-6/CV-31234. Letdown Line I s o l a t i o n CVC-212/MV-32115.

RXCP S e a l Water Return I s o l a t i o n CVC-211/MV-32124. RXCP Seal Water Return I s o l a t i o n FIRST SECONC OPER OPFR 0 PERAB LE CLOSED OR cc SYS INTACT TO RXCP A CLOSED OR cc SYS INTACT TO RXCP A CLOSED OR cc SYS INTACT TO RXCP B CLOSED OR cc SYS INTACT TO RXCP B OPERABLE OPERABLE 0 P E RAB LE OPERABLE

WISCONSIN PUBLIC SERVICE CORPORATION KEWAUNEENOCLEAaPOWERPLANT OPERATING PROCEDURE DATE 4.3 V e r t i c a l Panel B AS-l/CV-31383. Containment A i r Sample I s o l a t i o n A AS-Z/CV-31384. Containment A i r Sample I s o l a t i o n 6 AS-32/CV-31385. Containment A i r Sample I s o l a t i o n C MD(R)-134/CV-31136. Cntmt Sump Pumps Discharge Header 1 sol MD(R)-135/CV-31137. Cntmt Sump Pumps Discharge Header 1 sol 4.4 V e r t i c a l Panel A RC-402/CV-31263, Pressurizer Steam Sampling I s o l a t i o n RC-412/CV-31264. Pressurizer L i q u i d Sampling I s o l a t i o n RC-422/SV-33092.

Rx Coolant Hot Leg Sampling I s o l a t i o n RC-403/CV-31267. Pressurizer Steam Sampling I s o l a t i o n RC-413/CV-31268, Pressurizer L i q u i d Sampling I s o l a t i o n RC-423/SV-33327. Rx Coolant Hot Leg Sampling I s o l a t i o n MU-1010-1/CV-31261, Przr R e l i e f Tank Make Up Water I

sol MG(R)-512/CV-31259.

Przr R e l i e f Tank Gas Sampling I s o l MG(R)-513/CV-31260. Przr R e l i e f Tank Gas Sampling I s o l NG-302/CV-31298. Przr R e l i e f Tank Nitrogen Supply I s o l RC-507/CV-31134. Rx C l n t Drain Pump Disch Header I s o l RC-508/CV-31135. Rx Clnt Drain Pump Disch Header I s o l MG(R)-509/CV-31132. RCDT Vent To Waste Gas Header MG(R)-510/CV-31133. RCDT Vent To Waste Gas Header MG(R)-503/CV-31216.

RCDT To Gas Anzr Header I s o l a t i o n CONT I NU ED NO. N-FH-53-CLD TITLE Refueling Daily Checklist DATE OCT 11 2005 PAGE 8 of 10 FIRST SECOND OPER OPER

~-

0 PE RAB L E OPERABLE OPERABLE OPERABLE OPERABLE OP E RAE LE OPE RAE LE OPERABLE OPERABLE OPERABLE OPERABLE 0 PERAB LE OPERABLE OPERABLE OPERABLE 0 P E RAE! LE 0 P E RAB LE OPERABLE 0 P E RAB LE OPERABLE

WISCONSIN PUBLIC SERVICE CORPORATION TITLE Refueling Daily Checklist I

KEWAUNEE NUCLEARPOWERPLANT NO. N-FH-53-CLD L

OPERATING PROCEDURE OAT E I

DATE OCT 11 2005 PAGE 9

of 10 4.4 CONTINUED MG(R)-504/CV-31217.

RCDT To Gas Anzr Header I s o l a t i o n MD(R)-323A/MV-32390. Deaerated Drains Tank Cntmt Oisch Isol A MD(R)-323B/MV-32391, Deaerated Drains Tank Cntmt Disch Isol B WG-31O/SV-33655. Deaerated Drains Tank Vent Outside Cntmt CVC-54/SV-33651. VCT Vent t o Cntmt VB-lOA/CV-31337.

Power Operated Cntmt Vacuum Breaker A VB-lOB/CV-31338. Power Operated Cntmt Vacuum Breaker B LOCA-201B/CV-31727. Post LOCA Hydrogen Recombiner 8 To Cntmt LOCA-lOOB/CV-31725. Post LOCA Hydrogen t o Recombiner B SA-7003B/MV-32148. Hydrogen D i l u t i o n To Containment LOCA-2B/MV-32146. Post LOCA Hydrogen Cntmt Vent Isol B RBV-l/CV-31125. Cntmt Purge/Vent Supply Valve A RBV-4/CV-31123. Cntmt Purge/Vent Exhaust Valve A RBV-2/CV-31126. Cntmt Purge/Vent Supply Valve B RBV-3/CV-31124. Cntmt Purge/Vent Exhaust Valve B FIRST SECONC OPER OPER OPERABLE OPERABLE OPERABLE OPERABLE OPE RAE LE 0 PERAB LE OPERABLE 0 P E RAE LE OP E RAB LE OPERABLE 0 P E RAE LE OPERABLE OPERABLE 0 PE RAB L E OPERABLE

1 NO. N-FH-53-CLD WISCONSIN PUBLIC SERVICE COIWORATION KEWAUNEE NUCLEARPOWERPLANT TITLE Refueling D a i l y Checklist OPERATING PROCEDURE I DATE OCT 11 2005 I PAGE 10 of 10 DATE 4.5 Mechanical Console A BT-31A/CV-31334. S/G Sample Isol Vlvs BT-31B/CV-31270. S/G Sample I s o l Vlvs BT-32A/CV-31335. S/G Sample I s o l Vlvs BT-32B/CV-31271. S/G Sample Isol Vlvs BT-3A/MV-32078 S / G A Blowdown I s o l a t i o n Valve A2 BT-3B/MV-32080 S/G B Blowdown I s o l a t i o n Valve 82 I

I 5.0 LOCAL VALVE POSITIONS 5.1 None FIRST SECOND OPER OPER OPERABLE OPERABLE 0 PE RAB LE OPERABLE OPERABLE OPERABLE PERFORMED BY DATE PERFORMED BY DATE PERFORMED BY DATE S H I FT MANAGER DATE SPVR NUCLEAR SHIFT OPERATIONS DATE

6.0 Procedure 6.1 Nuclear Safety And Conservative Decision-Making 6.1.1 The expectations for Nuclear Safety And Conservative Decision-Making are contained in DNOS-0101, Nuclear Safety And Conservative Decision-Making.

6.2 Human Performance - Self-Checking (STAR) 6.2.1 The expectations for Self-checking are contained in DNAP-1907, Human Performance (HU) Program.

6.3 Human Performance - Peer-Checking 6.3.1 The expectations for Peer-Checking are contained in DNAP-1907, Human Performance (HU) Program.

6.3.2 Additional requirements for peer-checking are:

6.3.2.1 6.3.2.2 6.3.2.3 6.3.2.4 6.3.2.5 The SM or US shall assess the complexity or the importance of task performance to determine peer checker qualifications. The peer checker shall have sufficient knowledge to evaluate the task in progress and is knowledgeable of the peer checking process.

Peer-checking requires the performance of the action to be observable and methodical in the actions being taken, with discrete pauses to allow observers to correct any errors.

Peer-checking never relieves an Operator from the responsibility to self-check, nor the responsibility for achieving the desired outcome of the action being taken.

Peer-checking is encouraged but not required for all Control Board manipulations.

When the pace of activities increases to the point where obtaining a peer check constitutes a distraction in itself, operators and supervisors shall remain aware of the need to prevent operational events by increasing the attention paid to other error prevention techniques, such as use of STAR.

Peer-checking is required for all reactivity manipulations, except during seIected activities where it is not practical (e.g., rapid power reductions, or transients, etc.).

The peer check requirement may be waived only with concurrence of shift supervision.

INFORMATION USE

DNOS - 0101 Revision 0 DOMINION NUCLEAR OPERATIONS STANDARD NUCLEAR SAFETY AND CONSERVATIVE DECISION MAKING Expectation :

9 Nuclear and industrial safety are the overriding station concerns.

> The reactor and its supporting systems are maintained within the bounds of analyzed equipment alignments and approved procedures.

3 Risks and challenges associated with plant operations are anticipated and a healthy respect is maintained for the stored energy within the reactor core.

> Operators faced with unexpected or uncertain conditions will place the plant in a safe condition and will not hesitate, if necessary, to reduce power or trip the reactor.

Standards:

3 9

9 3

9 9

9 P

Operators shall recognize when degraded conditions exist that could challenge plant safety or reliability.

Information shall be gathered and analyzed from relevant sources and appropriate personnel in order to clearly define and provide options for resolution of operational concerns.

Short-and long-term risks, consequences, and the aggregate impact associated with decision options shall be critically and objectively considered.

Implementation plans to resolve operational concerns shall be developed that include contingencies and compensatory measures to maintain or enhance safety or probabilistic risk margins.

Decision-makers and their roles and responsibilities shall be clearly identified.

Command and control responsibilities shall be carried out in accordance with sitespecific procedures.

The bases for decisions shall be communicated throughout the organization.

The effectiveness of decisions shall be periodically evaluated.

Human performance tools and group input shall be utilized to avoid inappropriate actions and unexpected responses when reaching operating decisions.

When faced with time-critical decisions, operators:

+

Do not allow production or cost to override safety.

+

Do not challenge the safe operating envelope.

+

Question and validate available information.

+

Utilize available alternate indications to validate information.

+' Assume the available indications are valid until proven otherwise.

Use all available resources, including people offsite, if necessary.

+

Develop contingency actions, if time allows.

+

Do not proceed in the face of uncertainty.

Approved By:

e on File Date: On File Page 1 of 1

WISCONSIN PUBLIC SERVICE CORP.

Kewaunee Nuclear Power Plant No.

RF-03.O 1 Rev.

J Refueling Procedure Reviewed By Tim Wiltman Nuclear 0 Yes Safety Related 0 No Date OCT 14 2004 Page 1 of 20 PORC p~ yes Review Required 0 No

- - INFREQUENTLY PERFORMED TEST ***

SRO Approval Of 0

Yes Temporary Changes Required 0 No 1.0 Purpose 1.1 This procedure provides instruction for fuel movement during a refueling outage.

2.0 General Notes 2.1 2.2 2.3 2.4 2.5 2.6 2.7 Contact the Reactor Engineering Group (RXE) for resolution of any problems, concerns, or questions.

Signoffs in this procedure shall be M E unless identified otherwise or unless designated by the RXE.

Working copies of this procedure may be used to facilitate fuel movement and signoffs may initially be made in the working copies. However, signoffs are to be transferred to the master copy as soon as practical.

During performance of this procedure, the master copy of all forms in this procedure (FAHDR, FAMS, ICRR) shall be retained in the control room. The containment and spent fuel pool copies shall be maintained for informational purposes only.

Operations (OPS) (SRO), Chemistry, and Quality Control (QC) work groups are needed to support performance of this procedure.

During refueling operations, the big picture must be kept in focus at all times. Care must be taken to NOT fixate on only one solution to a probledissue encountered during fuel movement. The entire refueling team should be included in all problem solving and the problem observed from as many directions and angles as possible prior to deciding on and implementing a solution. The Management team needs to be kept informed of the same.

During refueling operations, schedule pressure needs to be conservatively managed. The time required to refuel the reactor safely, successfully, and error free is the correct duration. Each member of the refueling team needs to understand that they have the authority to stop refueling activities to resolve an issue. Only the Refueling SRO shall restart refueling activities.

plga459.doc-Amy KudicWDori Ziegler-Tim Wiltman CONTINUOUS USE

2006 Kewaunee NRC Retake Written Examination Question Feedback

1. Examinee Feedback Question Number 98 Answer B is correct answer. When DG is started it is done per A-DGM-1 OB, and Bus 6 is energized per step 4.6. Step 4.6.4 states Sequentially start safeguards equipment as required.

This guidance can be used to start the SI Pump B. No conflicting guidance is given in ECA-0.0.

2. Resolution and Response The question is correct as written.

Step 7 of ECA-0.0 directs the operator to place specific equipment in PULLOUT, including the SI Pumps. A NOTE prior to step 9 (Dispatch personnel to locally restore Emergency AC Power) reads. Pre-planning of power restoration efforts based on the event and available sources is required. When power is restored to the Bus the operator is directed to continue actions of ECA-0.0 at step 37. Three CAUTIONS exist prior to step 38. The two applicable CAUTIONS read, The loads placed on the energized emergency AC Bus should not exceed the capacity of the power source, and, If an SI signal exists or if an SI signal is actuated during this procedure, it should be reset to permit manual loading of equipment on an emergency AC bus. At step 40, the operator transitions to the appropriate recovery procedure (ECA-0.2 in this case).

The NOTE at the beginning of the procedure states (Likewise in ECA-O.O), CSF Status Trees should be monitored for information only. Function Restoration Procedures should not be implemented prior to completion of Step 10. At step 5 in ECA-0.0, the operator is directed to manually load safeguards equipment on AC Emergency Bus, including SI Pumps.

BKG ECA-0.0 specifically states in 1. Introduction, If plant conditions have deteriorated significantly, the operator may have insufficient or conflicting indications as to plant status and a concurrent event may be contributing to the deterioration of RCS conditions. Under these RCS conditions, the operator is instructed to implement IPEOP ECA-0.2 and initiate plant recovery utilizing Safety Injection (SI) operational systems. IPEOP ECA-0.2 functions to start safe-guards equipment as appropriate and then directs the operator to go to IPEOP E-1...)

Also (section 3)

The loss of all ac power procedures are unique within the IPEOP set. With the exception of these procedures, all IPEOPs are written on the premise that at least one ac emergency bus is energized and associated equipment can be powered from the energized ac emergency bus. Consequently, the guidance provided in other procedures in the IPEOP set is not applicable following a loss of all ac power. Therefore, ECA-0.0 has priority over all other procedures in the IPEOP set.

Page 16

2006 Kewaunee NRC Retake Written Examination Question Feedback (3.1.4)

Following restoration of ac power, the operator is instructed to stabilize steam generator pressures, if secondary depressurization is in progress, and to evaluate the status of the energized ac bus. These actions verify that certain select equipment has automatically loaded on the ac emergency bus and provides the operator with information that will aid him in loading subsequent equipment on the energized ac emergency bus in recovery procedures.

Step 6 CAUTION 1 (for the transition to step 37) basis reads To minimize the deterioration of plant conditions, recovery actions should be started as soon as ac power is restored. Procedure ECA-0.0 is written such that recovery actions step can be entered from any step that follows this CAUTION.

Procedures ECA-0.0, ECA-0.1 and ECA-0.2 are written to establish the appropriate systems operation and alignments before transitioning the operator to other IPEOPs.

UG-0, Users Guide For Emergency and Abnormal Procedures, sets the priority for implementing procedures. Section 6.2 identifies the general order of priority: 1) FRPs; 2) ORPs;

3) EOPs and 4) AOPs. As mentioned above ECA-0 series is special in that it takes priority over FRPs. In this case the direction of ECA-0 series should take the highest priority in actions to be performed.

The action in B will start the SI Pump without having a Component Cooling Water Pump running for support.

3. Supportinq Information A-DGM-1 OB, step 4.6 BKG ECA-0.0 and ECA-0.2 UG-0, 6.2.1 & 6.2.4 Page 17

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1.0 REFERENCES

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