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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 4 2 , Kewaunee, WI 54216-951 1 minion AUG 18 2006 Regional Administrator, Region I l l Serial No.06-711 U. S. Nuclear Regulatory Commission KPS/LIC:GR RO 2443 Warrenville Road Docket No. 50-305 Suite 210 License No. DPR-43 Lisle IL 60532-4352 Attention: Mr. Dell McNeil DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION OPERATOR LICENSE EXAMINATION In accordance with the requirements of NUREG-1021, 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.06-711 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 Initial wades prior to Feedback RO Section SRO Section Total 53 of 75 = 70.66% 21 of 25 = 84.00% 74 of 100 = 74.00%

63 of 75 = 84.00% 15 of 25 = 60.00% 78 of 100 = 78.00%

Grades followinq Feedback includina accepted comments RO Section SRO Section Total 55 of 75 = 73.33% 21 of 24 = 87.50% 76 of 99 = 76.76%

64 of 75 = 85.33% 16 of 24 = 66.66% 80 Of 99 = 80.80%

Summary of examination chanqes with accepted comments

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

Both candidates selected D.

2. RO question 55. Initial correct answer was D. Accept either Aor D as correct.

Candidate Beck selected Aand candidate Bunkelman selected B.

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

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

Candidate Beck selected B and candidate Bunkelman selected A.

5. 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.5provides 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 Awith 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 .Istep 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 .Istep 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 2416 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 FEATURES INIT1ATlON INSTRUMENT SETTING LIMITS Lead time constant l 2 540°F I

~~ ~~~

(') 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 Amendment 172 Page 1 of 2 02/27/2004

WLS;CONSM PWLIC SERVICE CORPORATION NO. N-RC-36C I r I KEWAUNEENUCLEARPOWERPLANT TITLE P r e s s u r i z e r Pressure Control 1

OPERATING PROCEDURE DATE JUN 17 2005 PAGE 8 of 12 I

NOTE: Automatic a c t u a t i o n o f P r z r PORV PR-2A. P r z r 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 C o n t r o l l e r output. Misadjustment of c o n t r o l l e r s e t p o i n t d i a l , manual c o n t r o l l e r o p e r a t i o n . manual back-up h e a t e r o p e r a t i o n . 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 p o i n t s o f above components.

c. POSITION P r z r Spray C o n t r o l 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 s e t p o i n t d i a l u n t i l t h e d e v i a t i o n meter n u l l s . ( t o p meter)

2. V E R I F Y Master C o n t r o l l e r s e t p o i n t matches p r e s s u r i z e r pressure channel s e l e c t e d f o r c o n t r o l .

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 p r o p e r l y .

3. P O S I T I O N AUTO-BAL-MAN s w i t c h t o MAN-BAL.

4. V E R I F Y d e v i a t i o n meter i s centered.

I 5. POSITION AUTO-BAL-MAN s w i t c h t o AUTO.

6. V E R I F Y p r e s s u r i z e r spray valves and h e a t e r c o n t r o l working p r o p e r l y t o m a i n t a i n P r z r pressure.

8. ADJUST Master C o n t r o l l e r s e t p o i n t d i a l t o s l o w l y INCREASE P r e s s u r i z e r pressure t o 2235 p s i g .

a. M a i n t a i n 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 a u t o m a t i c a l l y unblock, WHEN P r e s s u r i z e r pressure i s g r e a t e r t h a n 2000 p s i g .

V E R I F Y f o l l o w i n g p e r m i s s i v e s t a t u s l i g h t s , OFF:

0 44905 1101. P r e s s u r i z e r Perm Block SI 0 44905 1102, P r e s s u r i z e r S I Blocked

_r

WLSCOlYSM PWLIC SERMCE CORPORATION KEWAUNEENUCLEARPOWERPLANT TITLE P r e s s u r i z e r Pressure C o n t r o l OPERATING PROCEDURE I JUN 17 2005 I PAGE 10 of 12 4.3 CONTINUED NOTE: P r e s s u r i z e r Backup heaters should be e n e r g i z e d f o r any o f t h e f o l l o w i n g reasons:

0 To e q u a l i z e RCS t o p r e s s u r i z e r b o r i c a c i d d i f f e r e n c e of more than 50 ppm 0 To reduce excessive c y c l i n g o f p r e s s u r i z e r backup h e a t e r s d u r i n g RCS temperature changes 0 To m i n i m i z e pressure t r a n s i e n t s d u r i n g p l a n t evolutions

2. WHEN r e q u i r e d f o r p l a n t c o n d i t i o n s . THEN POSITION Backup P r z r Heater Groups t o ON.

3. WHEN NO l o n g e r r e q u i r e d f o r p l a n t c o n d i t i o n s , THEN POSITION Backup P r z r Heater Groups t o AUTO.

4.4 Pressure C o n t r o l During Cooldown:

NOTE: D i f f e r e n c e i n temperature between P r z r and RCS s h a l l NOT exceed 320°F.

NOTE: Hold Steam Generator pressure a t 600 p s i g u n t i l S I can be b l o c k e d a t 2000 p s i g i n RCS.

1. P O S I T I O N a t l e a s t two backup P r e s s u r i z e r h e a t e r groups t o ON.

2. ( C A S ) 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 a d j u s t i n g PRZR Spray Master C o n t r o l l e r s e t p o i n t .

3. WHEN RCS p r e s s u r e i s 1900-1950 p s i g . THEN PERFORM t h e f o l l o w i n g t o

~-

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

a. VERIFY permissive status i g h t 44905 1101. P r e s s u r i z e r Perm Block S I . ON.

b. P O S I T I O N S a f e t y I n j e c t i o n T r a i n A and T r a i n B Block/Unblock switches t o BLK.

c. V E R I F Y permissive status i g h t . 44905 1102. P r e s s u r i z e r S I Blocked. ON.

CONTINUED

WISCONSIN PUBLIC SERVICE CORPORATION I NO. ES-1.2 KEWAUNJCENUCLEARPOWERPLANT I POST LOCA COOLDOWN AND DEPRESSURIZATION EMERGENCY OPERATING PROCEDURES I G - J U L 06 2005 I PAGE 3 of 18 9

OPERATOR ACTIONS CONTINGENCY ACTIONS 2 Establish Charging Flow:

a. Charging Pumps - AT LEAST ONE a. Perform t h e f o l l o w i n g :

RUNNING

1) CC f l o w t o R X C P ( s 1 Thermal B a r r i e r i s l o s t ,

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

2) S t a r t one Charging Pump.

b. A l i g n Charging Pump s u c t i o n t o RW ST

1) Open CVC-301. RWST Supply To Charging Pumps I

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

c. S t a r t a second Charging Pump AND e s t a b l i s h maximum Charging flow

WISCONSIN PUBLIC SERVICE CORPORATION NO. ES-I. 2 KEWAUNEENUCLEARPOWERPLANT I POST LOCA COOLDOWN AND DEPRESSURIZATION EbERGENCY OPERATING PROCEDURES DATE JUL 06 2005 PAGE 5 of 18 OPERATOR ACTIONS CONTINGENCY ACTIONS 5 M a i n t a i n AFY Pump Discharge Pressures Greater Than 1000 PSIG:

a. T h r o t t l e AFW-ZA/CV-31315. AFWP A Flow C o n t r o l I a. L o c a l l y t h r o t t l e AFW-3A. AFW Pump A Discharge. I I

b. T h r o t t l e AFW-ZB/CV-31316. AFWP B Flow C o n t r o l I b. L o c a l l y t h r o t t l e AFW-3B. AFW Pump B Discharge. I

c. L o c a l l y t h r o t t l e AFW-2C. T/D c. TDAFW Pump d i s c h a r g e I

I AFW Pump Discharge p r e s s u r e c o u l d NOT be maintained greater than 1000 p s i g . THEN l o c a l l y perform the following:

1) P o s i t i o n TDAFW Pump Low D i s c h P r & s s T r i p Bypass s w i t c h t o BYPASS.

2) T h r o t t l e AFW-2C as necessary d u r i n g RCS cooldown t o m a i n t a i n TDAFW f l o w l e s s t h a n 260 gpm.

NOTE: During t h e RCS cooldown. RCS Boron C o n c e n t r a t i o n s h o u l d be monitored t o v e r i f y Cold Shutdown Boron Concentration p e r RD-6.7.

NOTE: RCS cooldown should be performed as f a s t as p o s s i b l e , b u t l e s s t h a n 100" F/hr.

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

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

b. Use RHR System i f i n s e r v i c e

c. Dump steam f r o m i n t a c t S G ( s ) c. Dump steam u s i n g i n t a c t S G ( s )

u s i n g t h e Steam Dump System i n PORV.

STM PRESS mode

WISCONSM PUBLIC SERVICE CORPORATION NO. ES-1.1 KEWAUNEE NUCLEARPOWERPLANT TITLE SI TERMINATION EMERGENCYOPERATINGpRocEDuREs DATE JUN 21 2005 P- 7 of 20 Fi OPERATOR ACTIONS I CONTINGENCY ACTIONS I 12 Check VCT Makeup Control System:

a. Makeup B o r i c A c i d C o n t r o l l e r - a. Set Makeup B o r i c Acid SET TO 11.0 c o n t r o l l e r t o 11.0.

b. Makeup s e t f o r automatic c o n t r o l b. Set Makeup Mode S e l e c t o r t o AUTO.

c. VCT l e v e l - BETWEEN 17% AND 28% c. R e - e s t a b l i s h VCT l e v e l .

13 Check Charging Pump Suction - A l i g n s u c t i o n t o VCT.

ALIGNED TO VCT 14 Transfer Steam Dump - TO PRESSURE -IF Condenser Steam Dump NOT CONTROL HODE a v a i l a b l e , THEM use Atmospheric Steam Dumps o r Steam Generator PORYs .

15 Check RCS Hot Leg Temperature - C o n t r o l steam dump and t o t a l feed STABLE f l o w as necessary t o s t a b i l i z e RCS temperature.

16 Using Pressurizer Heaters And I F normal

- spray NOT a v a i l a b l e AND Normal Pressurizer Spray As Letdown i s i n s e r v i c e , THEN use Necessary, Maintain Pressurizer a u x i l i a r y spray. I F NOT. THEN use Pressure - STABLE one PRZR PORV.

WISCONSIN PUBLIC SERVICE CORPORATION NO. ES-1.1 KEWAUNJIE NUCLEARPOWERPLANT TI- S I TERMINATION EMERGENCY OPERATING PROCEDURES DATE JUN 2 1 2005 PJGE 13 of 20 I

OPERATOR ACTIONS D

I CONTINGENCY ACTIONS I 24 Shut Down Unnecessary Plant Equipment:

a. R e f e r 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 25 H a i n t a i n Stable Plant Conditions:

P r e s s u r i z e r p r e s s u r e - AT EXISTING VALUE 0 P r e s s u r i z e r 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%

26 V e r i f y SI Flow N o t Required:

a. RCS s u b c o o l i n g based on Core a. Manually s t a r t S I Pumps as E x i t TCs - GREATER THAN 30°F necessary. GO TO E - 1 . LOSS OF 165°F FOR ADVERSE CONTAINMENT3 REACTOR OR SECONDARY COOLANT, Step 1.

b . PRZR l e v e l - GREATER THAN 5% b. C o n t r o l Charging f l o w t o

[30% FOR ADVERSE CONTAINMENT] m a i n t a i n 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.

I KEWAUlWJt POWER STATION I P l a n t Cooldown from Hot Shutdown t

'ITLGCold Shutdown C o n d i t i o n I OPERATING PROCEDURE DATE MAY 06 2006 P ! 7 of 31 I INIT1ALS II 4-0 4.1 PreDare To Cooldown:

1. DETERMINE r e q u i r e d Cold Shutdown boron c o n c e n t r a t i o n u s i n g one o f t h e f o l l o w i n g :

CPCR0234511

a. IF a l l c o n t r o l rods a r e f u l l y i n s e r t e d , THEN APPLIES/NA REFER t o RD 6.9, 1%Cold Shutdown A R I Boron Concentration.

b. any c o n t r o l r o d i s n o t f u l l y i n s e r t e d , APPLIES/NA THEN CONSULT w i t h Reactor Engineering t o determine r e q u i r e d Cold Shutdown boron concentration.

I 2. ( L ) INITIATE b o r a t i o n o f RCS per N-CVC-35A as follows:

a. BORATE t o r e q u i r e d Cold Shutdown boron c o n c e n t r a t i o n determined i n Step 4.1.1.

BORATED/

I N I T IATED NOTE: During RCS cooldown. a t l e a s t two backup groups of P r e s s u r i z e r Heaters should remain energized t o minimize f a t i g u e s t r e s s t o P r e s s u r i z e r and P r e s s u r i z e r Surge Line.

3. (CAS) ESTABLISH, RCS Pressure C o n t r o l During ESTABLISHED Cooldown. p e r N-RC-36C.

4. (CAS) ESTABLISH. Automatic P r e s s u r i z e r Level ESTABLISHED C o n t r o l f o r Shutdown, p e r N-CVC-35B.

5. (&I any valves have been backseated p e r PERFORMED/ NA N-0-01-CLE. THEN PERFORM N-0-01-CLE as necessary t o t a k e a p p l i c a b l e valves o f f backseat.

6. ( C A S ) MAINTAIN Steam Generator Blowdown f l o w as MAINTAINED .

l o n g a s p o s s i b l e d u r i n g p l a n t cooldown.

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~upportinqInformation 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 Page 14 of 21 3.7.5 Major SA & IA Manual Valves

I Title Chemical and Volume Control System (CVC) Date I 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 Check Valve CVC-14. When CVC-11 is closed, CVC-14 will relieve excessive a

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.

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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 1B1 and 1B2 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

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

. O SUBSEQUENT ACTIONS 4.1 Condensate Pump A H o t w e l l Level T r i p :

1. VERIFY Feedwater Pumps A and B. OFF.

2. VERIFY AFW Pumps A and B running.

3. VERIFY T u r b i n e t r i p p e d .

4. power i s above P-10, THEN VERIFY r e a c t o r , r i p an' - GO - TO E-0.

4.2 Condensate Pump B H o t w e l l Level T r i p :

None 4.3 Condensate Pump Overcurrent T r i p :

None 4.4 Feedwater Htr Bypass A l e r t :

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

~ ~~ ~

INITIATING EVENT IMPULSE PRESSURE Condensate Pump T r i p s275 p s i g Feedwater Pump T r i p 5285 p s i g Two Heater D r a i n Pumps T r i p 5425 p s i g

~ ~~ ~ ~

One Heater D r a i n Pump T r i p 5520 p s i g 1

2. Feedwater Pump s u c t i o n pressure can NOT be m a i n t a i n e d g r e a t e r t h a n 260 p s i g . THEN REDUCE T u r b i n e l o a d u n t i l s u c t i o n g r e a t e r than 260 p s i g .

CONTINUED

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RHW. MOR.^. HEATER DRAIN COUPLING ( SHEET 1 OF 2)

(XK 136-4)

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 400° (tube) / 35OOF (she11)600 psig (tube) / 150 psig (Shell SI Pump 3OOOF 2485 psig ICs Pump 3OOOF 500 psig 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 6.2-12 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 6.2-12, 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 Feedback 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 Aincorrect 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 1.1lgpm 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 1.1lgpm and 0.75gpm has been determined to be identified leaves the remaining 0.36gpm as unidentified. Technical Specifications allows not in excess of 1gpm 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 1.11 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 I NO. SP-36-082 I REV A3 (FREQ D)

I KEWAUNEE POWER STATION I Reactor Cool a n t System Leak Rate Check I

SURVEILLANCE PROCEDURE 1 DATE MAR 02 2006 I PAG 1 of 14 I I REVIEWED BY Michael Sevey I APPROVED BY Jefiey Simon I

NUCLEAR @ YES PORC REVIEW YES SRO APPROVAL OF YES SAFETY RELATED REQUIRED TEMPORARY CHANGES 0 NO 0 NO REQU I RED 0 NO 1.0 PLANT I N I T I A L CONDITIONS 1.1 Mass Balance Leakrate C a l c u l a t i o n s h a l l be performed a t l e a s t weekly whenever t h e Reactor i s a t power o r i n a Hot Shutdown c o n d i t i o n . This requirement can NOT be waived. CPCR0166551 1.2 Mass Balance Leakrate C a l c u l a t i o n performance i s d e s i r e d on a d a i l y b a s i s whenever t h e Reactor i s a t power o r i n a Hot Shutdown c o n d i t i o n .

The S h i f t Manager may waive t h i s requirement i f p l a n t c o n d i t i o n s do

-NOT a l l o w d a i l y performance. CPCR0166553 1.3 Reactor power and xenon a r e s t a b l e , such t h a t letdown d i v e r s i o n t o t h e Holdup Tanks w i l l _NOT occur d u r i n g t h e t e s t .

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

I Emergency b o r a t i o n I D i v e r s i o n o f letdown t o t h e Holdup Tanks

.I Makeup from any source which does n o t go through t h e B o r i c A c i d o r Makeup Water t o t a l i z e r s D i v e r s i o n o f Excess Letdown t o t h e RCDT A Reactor Coolant System sample i s taken v i a t h e C / R p r i m a r y sampling v a l v e s o r t h e HRSR manual valves 2.2 An i n c r e a s e i n containment h u m i d i t y i s i n d i c a t i v e o f an e x t e r n a l l e a k f r o m t h e Reactor Coolant System. However, s i n c e 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 i n c r e a s e i n h u m i d i t y due t o a l e a k i n t h e Reactor Coolant System should a l s o show a s i g n i f i c a n t i n c r e a s e i n t h e other monitors.

2.3 I f e i t h e r t h e Containment Sump A Level D e t e c t i o n System Sump Pumps become i n o p e r a b l e , r e f e r t o S e c t i o n 6.0. Step 6.4.

DOMINION ENERGY KEWAUNEE NO. SP-36-082 KEWAUNEE POWER STATION I Reactor Coolant System Leak Rate

'ITLE Check SURVEILLANCE PROCEDURE I MAR 02 2006 I PAC33 2 of 14 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 a r e based on T e c h n i c a l S p e c i f i c a t i o n s 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 t h e s u b j e c t o f an i n v e s t i g a t i o n and e v a l u a t i o n 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 t h e i n d i c a t i o n . Any i n d i c a t e d l e a k s h a l l be considered t o be a r e a l l e a k u n t i l i t i s determined t h a t no unsafe c o n d i t i o n e x i s t s . I f t h e Reactor Coolant System leakage exceeds 1 gpm AND t h e 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 .

t h e n t h e Reactor s h a l l be p l a c e d i n t h e Hot Shutdown c o n d i t i o n u t i l i z i n g normal o p e r a t i n g procedures.

I f t h e 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 t h e Reactor s h a l l be p l a c e d i n t h e Cold Shutdown c o n d i t i o n u t i l i z i n g normal o p e r a t i n g procedures.

2. I f t h e 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 c o n t i n u e d o p e r a t i o n i s s a f e , then o p e r a t i o n o f t h e Reactor w i t h a t o t a l Reactor Coolant System leakage r a t e n o t exceeding 10 gpm s h a l l be p e r m i t t e d . I f leakage exceeds 10 gpm, t h e n t h e Reactor s h a l l be placed i n t h e Hot Shutdown c o n d i t i o n w i t h i n 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months u t i l i z i n g normal o p e r a t i n g 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 t h e Reactor s h a l l be placed i n t h e Cold Shutdown c o n d i t i o n u t i l i z i n g normal o p e r a t i n g procedures.

3. Primary t o secondary leakage i s l i m i t e d t o 150 g a l l o n s per day t h r o u g h any one steam generator. With t u b e 'leakage g r e a t e r than t h e above l i m i t , reduce t h e 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 t h e 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 t h r o u g h a n o n - i s o l a b l e 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 t h e Reactor Vessel, p i p i n g , v a l v e body, R e l i e f Valve l e a k s .

P r e s s u r i z e r , Steam Generator Head. o r Pump Seal l e a k o f f ) . t h e n t h e Reactor s h a l l be shut down AND a cooldown t o t h e Cold Shutdown c o n d i t i o n 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 d e t e c t i o n .

.O GENERAL I N S T R U C T I O N S 4.1 None

.O EQUIPMENT REQUIRED 5.1 None

KEWAUNEE POWER STATION I c

'ITUE Check Reactor Coolant System Leak Rate SURVEILLANCEPROCEDURE DATE MAR 02 2006 PAGE 3 of 14 DATE 6.0 PROCEDURE l - NOTE: T h i s c a l c u l a t i o n i s performed d a i l y .

use Step 6.1.1; o t h e r w i s e use Step 6.1.2.

I f t h e PPCS i s i n s e r v i c e ,

NOTE: Containment Sump A l e a k r a t e can O JJ be used f o r Mass Balance determination.

NOTE: I f necessary, t h e s t a r t t i m e o f t h e computer c a l c u l a t i o n may be a d j u s t e d t o a value l e s s than 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . However, r e d u c i n g t h e t e s t i n t e r v a l w i l l reduce t h e accuracy o f t h e c a l c u l a t i o n .

Minimum t e s t i n t e r v a l f o r t h e 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 C a l c u l a t i o n 6.1.1 Cornouter C a l c u l a t i o n

a. On PPCS Main Menu, C L I C K on A p p l i c a t i o n s Menu.

b. On A p p l i c a t i o n s Menu, C L I C K on On Demand RCS Leakage.

c. VERIFY values are p r o v i d e d f o r a l l RCS Leakage data p o i n t s .

d. VERIFY a p p r o p r i a t e v a l u e f o r VCT Level Over 56%.

e. CLICK on C a l c u l a t e 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 c a l c u l a t e d RCS l e a k 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 l e a k r a t e c a l c u l a t i o n i s n e g a t i v e . THEN PERFORM one o f t h e f o l l o w i n g :

3. GO TO Step 6.1.1 and REPEAT Computer C a l c u l a t i o n .

OR

2. GO TO Step 6.1.2 and PERFORM Manual C a l c u l a t i o n .

3. GO TO Step 6.1.1.h and ACCEPT n e g a t i v e l e a k r a t e results.

CONTINUED

DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate KEWAIJNEE POWER STATION Check SURVEILLANCE PROCEDURE DA!CE MAR 02 2006 I PAGE 4 of 1 4 DATE 6.1.1 CONTINUED

h. RECORD l e a k r a t e i n t h e C o n t r o l Room Log and on S h i f t Manager's s t a t u s board.

i. fl Mass Balance l e a k r a t e c a l c u l a t i o n i n d i c a t e s t h a t leakage from Reactor Coolant S y s t e m i s n e g a t i v e leakage i s g r e a t e r than 0.2 gpm. THEN GO TO Step 6.3.

6.1.2 Manual C a l c u l a t i o n

a. Special Precautions

1. Reactor Coolant System temperature should be s t a b i l i z e d and h e l d c o n s t a n t f o r approximately one hour b e f o r e s t a r t i n g the test.

2. Reactor Makeup System i s i n automatic. A l l Makeup s h a l l go through t h e B o r i c Acid Blender.

3. Reactor power should be s t a b i l i z e d and h e l d c o n s t a n t

( p l u s o r minus 51) f o r approximately one hour b e f o r e s t a r t i n g t h e t e s t and f o r t h e d u r a t i o n o f t h e t e s t .

4. P r e s s u r i z e r 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 p r e s s u r i z e r 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 P r e s s u r i z e r temperature and pressure a r e t h e same value a s recorded i n i t i a l l y .

1. Reactor Coolant temperature and P r e s s u r i z e r temperature and p r e s s u r e a r e t h e 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 NO. SP-36-082 Reactor Coolant System Leak Rate KEWAUNEE POWER STATION Check SURVEILLANCE PROCEDURE DA& MAR &! 2006 I P m 5 of 14 DATE 6.1.2.c CONTINUED

2. Reactor Coolant temperature o r P r e s s u r i z e r temperature o r pressure have changed, THEN PROCEED a s follows:

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

6. ALLOW c o n d i t i o n s 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 l e a k r a t e u s i n g i n f o r m a t i o n recorded and t h e formula on D a t a Sheet 2.

e. Mass Balance l e a k r a t e c a l c u l a t i o n i s n e g a t i v e . I@

PERFORM one o f t h e f o l l o w i n g :

1. GO TO Step 6.1.2 and REPEAT Manual C a l c u l a t i o n .

2. GO TO Step 6.1.1 and PERFORM Computer C a l c u l a t i o n .

3. TO Step 6.1.2.f and ACCEPT n e g a t i v e l e a k r a t e results.

f. RECORD l e a k r a t e i n t h e Control Room Log and on S h i f t Manager's s t a t u s board.

g. E Mass Balance l e a k r a t e c a l c u l a t i o n i n d i c a t e s t h a t leakage from Reactor Coolant System i s n e g a t i v e OR leakage i s g r e a t e r than 0.2 gpm. THEN GO TO Step 6.3.

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

a. RECORD date, t i m e . and pump A o r B f o r each Containment Sump A Pump r u n on D a t a Sheet 3.

b. The volume o f Containment Sump A between t h e h i g h l e v e l a l a r m and t h e automatic pump s h u t o f f i s 339.0 g a l l o n s

( H i - H i A l a r m i s 412.5 g a l l o n s ) .

c. CALCULATE leakage t o Containment Sump A u s i n g 339.0 g a l l o n s and t h e t i m e between pump a c t u a t i o n s ( t o n e a r e s t 1/10 of a minute).

d. RECORD leakage on D a t a Sheet 3.

e. RECORD pump r u n t i m e t o d e t e c t pump d e g r a d a t i o n .

f. Upon c o m p l e t i o n o f weekly Reactor Coolant System Leak Rate Test (Step 6.1). ATTACH D a t a Sheet 3 t o t h e r e s u l t s of t h e test.

g. E i n d i c a t e d leakage i s g r e a t e r t h a n 1 gpm. THEN PERFORM a Reactor Coolant System Mass Balance l e a k r a t e c a l c u l a t i o n per Step 6.1.

6.2.2 a sump pump has NOT r u n w i t h i n t h e p a s t 23 days, THEN VERIFY CONTAINMENT SUMP A LEVEL H I G H (47031-0) a l a r m o p e r a b l e by p e r f o r m i n g one o r more o f t h e f o l l o w i n g : CPCR0158511

a. D R A I N PRT t o Sump A u s i n g a l t e r n a t e method p e r N-RC-36B.

OR

b. D R A I N RCDT t o Sump A a s f o l l o w s :

1. OPEN RC-534/CV-31218, Rx C l n t D r a i n Tank To Cntmt Sump A.

2. WHEN RCDT lowers t o 12% l e v e l CONTAINMENT SUMP A LEVEL H I G H (47031-0) a l a r m a c t u a t e s , THEN CLOSE RC-534.

DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate mWAIJNEE POWER STATION Check SURVEILLANCE PROCEDURE 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 E v a l u a t i o n 6.3.1 Reactor Coolant System l e a k r a t e i s determined t o be n e g a t i v e , THEN PERFORM t h e f o l l o w i n g : CPCR0077821 NOTE: A-RC-36F may be used f o r r e f e r e n c e .

a. INVESTIGATE source o f inleakage.

b. INITIATE an A c t i o n Request ( A R ) documenting t h e r e s u l t s of t h i s investigation.

6. 3 . 2 -IF Reactor Coolant System l e a k r a t e i s determined t o be g r e a t e r t h a n 0.2 gpm. THEN an i n v e s t i g a t i o n and e v a l u a t i o n 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 t h e 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 u s i n g D a t a Sheet 4 . The f o l l o w i n g l e a k paths s h a l l be i n v e s t i g a t e d :

a. RCS Leak t o Containment

1. Containment Sump A Run H i s t o r y . recorded on D a t a Sheet 3 per Step 6.2.1.

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

trending.

3. Containment Gas M o n i t o r , R - 1 2 . t r e n d i n g .

4. Containment Humidity D e t e c t o r t r e n d i n g .

b. RCS Leak t o Component C o o l i n g System

1. Comp C o o l i n g L i q u i d M o n i t o r . R-17. t r e n d i n g .

2. Component Cooling Surge Tank l e v e l t r e n d i n g us,ng t h e Operator Aids book.

c. RCS Leak t o Steam Generators

1. A i r E j e c t o r Exhaust Mon t o r , R-15. t r e n d i n g .

2. S / G Blowdown L i q u i d Mon t o r , R-19. t r e n d i n g .

CONTINUED

DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate JCEWAUNEE POWER STATION 'ITLE Check I PAGE

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SURVEILLANCE PROCEDURE DATE MAR 02 2006 8 of 14 DATE 6.3.2 CONTINUED

d. RCS Leak t o Waste D i s p o s a l System

1. PRT l e v e l t r e n d i n g u s i n g t h e Operator A i d s book.

2. RCDT l e v e l t r e n d i n g u s i n g t h e Operator Aids book.

6.3.3 E NONE o f t h e l e a k 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 S p e c i f i c a t i o n s a r e 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 t h e source o f t h i s leakage e x t e r n a l t o t h e Reactor Coolant System (e.g. c h a r g i n g pump seal leakage).

6.3.4 The f o l l o w i n g items a r e i n d i c a t o r s o f p o t e n t i a l leakage sources e x t e r n a l t o t h e Reactor C o o l a n t System:

a. Charging Pump L e a k - o f f

b. Changes i n Tank Levels 0 RCDT 0 C V C HUT 0 Waste HUT 0 Deaerated D r a i n Tank

c. Sump Pump Run Times 0 Rx C a v i t y 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 L e a k o f f L i n e s

f. F i l t e r Vent and Drain L i n e s

g. D e m i n e r a l i z e r Vent and D r a i n Lines

h. V i s u a l I n s p e c t i o n o f Aux B u i l d i n g CONTINUED

n DOMlNION ENERGY KEWAUNEE Reactor Coolant System Leak Rate KEWAUNEE POWER STATION Check SURVEILLANCE PROCEDURE DATE MAR 02 2006 I PAGE 9 of 14 DATE 6.3.4 CONT I NU ED

i. V i s u a l I n s p e c t i o n o f Containment

j. LO-13 Bellows F a i l u r e

k. HRSR D r a i n s t o :

RHR Pump P i t Sump DDT Sump Tank 6.4 Containment Basement (592 0 EL) I n s D e c t i o n 6.4.1 E c o n d i t i o n s o f P r e c a u t i o n 2.0. Step 2.3 e x i s t , THEN PERFORM a weekly v i s u a l i n s p e c t i o n f o r water i n t h e general area of Sump A and Sump 6.

6.4.2 COMPLETE an A c t i o n Request f o r each i n s p e c t i o n n o t i n g r e s u l t s o f the inspection.

I r0NTTNll01TS 1TSF

DOMlNlON ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate KEWAUNEE POWER STATION Check I

~ ~ _ _ _~ _ ~

SURVEILLANCE PROCEDURE DATE MAR 02 2006 PAGE 10 of 14 OAT E INITIALS 7.0 PROBLEMS 7.1 Any problems encountered d u r i n g t e s t ? YESINO 7.2 IF yes, THEN INITIATE an A c t i o n Request ( A R ) p e r INITIATEDINA GNP-11.08.01, A c t i o n Request Process.

AR#

3.0 ACCEPTANCE C R I T E R I A 8.1 Any Reactor Coolant System l e a k r a t e o f l e s s t h a n zero g a l l o n s p e r minute has been i n v e s t i g a t e d as r e q u i r e d by t h i s procedure.

8.2 Any Reactor Coolant System l e a k r a t e o f g r e a t e r t h a n 0.2 g a l l o n s per minute has been i n v e s t i g a t e d as r e q u i r e d by t h i s procedure.

8.3 Reactor Coolant System leakage meets t h e L i m i t i n g C o n d i t i o n s f o r Operation p e r Technical S p e c i f i c a t i o n 3.1.d.

I I

8.4 RCS L e a k r a t e has been determined p e r t h i s procedure a t l e a s t once per I 7 days. ( T a b l e TS 4.1-3 I t e m 8 )

1.0 REFERENCES

9.1 Technical S p e c i f i c a t i o n s 3.1.d 9.2 KAP 00-003466 9.3 PCR007782 9.4 KNPP Pumps And Valves IST Plan 9.5 Comtrac 91-205

DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate KEWAUNEE POWER STATION Check SURVEILLANCE PROCEDURE DATE MAR 02 2006 I PAGE 11 of 14 DATA SHEET 1 REACTOR COOLANT LEAKAGE CALCULATION BY COMPUTER Parameter I Data a t S t a r t of Calculation D a t a a t End o f C a l c u l a t ion Difference Time I i n time:

Hrs-Mi n-VCT Temp (T0140A)

VCT Press (P0139A)

~ ~

VCT Level (L0112A)

P R Z R Press (P8023G)

~ ~~ ~ ~ ~ ~

P R Z R Level (L8015G) r A V G (TO4446 1

<MW (Gal) 3A (Gal)

ont. E l . 626' Amb.

4ir Temp (15187)

~

ontainment iumidi t y ( % )(41517) I Reactor Coolant System Leakage 9Pm

-IF leakage i s n e g a t i v e leakage i s g r e a t e r t h a n 0.2 gpm problems were encountered d u r i n g t e s t , THEN DESCRIBE on an A c t i o n Request.

-I F leakage i s n e g a t i v e leakage i s g r e a t e r t h a n 0.2 gpm. THEN REFER t o Step 6.3.

NOTE: A t t a c h Containment Sump Pump D a t a Sheet.

Comment s :

PERFORMED BY DATE S H I FT MANAGER DATF SPVR NUCLEAR S H I F T OPERATIONS DATE

I DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate KJtWAUNEE POWER STATION Check SURVEILLANCE PROCEDURE DATE MAR 02 2006 PAGE 12 of 14 DATA SHEET 2 MANUAL REACTOR COOLANT SYSTEM LEAKAGE CALCULATION I (Step 6.1.2-b:

I n i t i a l Readin:

Initial 44560/ RMW Batch YIC-111 Integrator gal B = ga 1 44559/

YIC-110 IBA Batch Integrator LIT-141 L I T - 112 hCT Level L I -426 L I -427 Pressurizer L I -428 Level T0400A. T0401P Reactor T0420A. T0421A Cool a n t "F(1) Required: No change

( Computer Temperature printout) (Taw) (2)

P0429A. P0430A P0431A. P0449A P r e s s u r i z e r psig(1) p s i g ( 1) Required: No change (Computer Pressure ( 2 )

printout1 T I -425

-0R- Pressurizer OF Required: No change T I -424 Temperature 15187 Cont E l 626' Amb A i r Temp O F N/A I N/A 41517 Containment Humidity N/A I N/A Time o f t e s t A = minutes ( 3 )

I Leak Rate (gpm) = 1.3798 + 1.41C - 17.390 - 43.29F = 9Pm A

leakage i s n e g a t i v e leakage i s g r e a t e r than 0.2 gpm problems were encountered d u r i n g t e s t , THEN DESCRIBE on an A c t i o n Request.

-I F leakage i s n e g a t i v e leakage i s g r e a t e r than 0.2 gpm. THEN REFER t o Step 6.3.

NOTE: A t t a c h Containment Sump A Pump D a t a Sheet.

Comments :

PERFORMED BY DATE S H I F T MANAGER DATE SPVR NUCLEAR SHIFT OPERATIONS DATE

DOMINION ENERGY KEWAUNEE NO. SP-36-082 Reactor Coolant System Leak Rate KEWAUNEE POWER STATION Check SURVEILLANCE PROCEDURE DATE MAR 02 2006 I PAGE 13 of 14 DATA SHEET 3 CONTAINMENT SUMP A PUMP RUN H I S T O R Y Gallons Time Since L a s t GPM Pump Pum ed Pump A c t u a t i o n 11 Run Time CNTMT I N I T 1ALS I Pump Date Time Y H i A arm (minl

~~ ~~

(min)

~

HUM I 339.0 339.0 339.0 1 I 1 I I I I I I 339.0 I I Hi-Hi = 412.5 G a l I F l e v e l d e t e c t i o n system o r sump pumps f o r Containment Sump A are i n o p e r a b l e . THEN R F E R t o Step 6.4.

Yas e i t h e r sump pump r u n w i t h i n t h e p a s t 23 days? Yes No -

If n e i t h e r sump pump has r u n w i t h i n t h e p a s t 23 days, THEN REFER t o Step 6.2.2.

jump leakage l e s s t h a n o r equal t o 1 gpm? Yes No -

sump leakage i s g r e a t e r than 1 gpm AND Technical S p e c i f i c a t i o n 3.1.d i s a p p l i c a b l e ,

rHEN DESCRIBE on an A c t i o n Request AND REFER t o Step 6.2.1.9. IPCR0003501 Jpon c o m p l e t i o n o f t h e l e a k r a t e t e s t , ATTACH t h i s D a t a Sheet t o t h e r e s u l t s .

a r r y over l a s t pump r u n t o t h e new D a t a Sheet 3.

PVR NUCLEAR SHIFT OPERATIONS DATE

DOMINION ENERGY KEWAUNEE I NO. SP-36-082 KEWAUNEE POWER STATION 1" Reactor Coolant System Leak Rate Check SURVEILLANCE PROCEDURE I DATE MAR 02 2006 I PAGE 14 of 1 4 DATA SHEET 4 INVESTIGATION AND EVALUATION INITIAL I N D I C A T I O N OF LEAKRATE >0.2 GPM DATE TIME LEAKRAT E 9 Pm I N V E S T I G A T I O N STARTED:

DATE TIME INVESTIGATION OF LEAKAGE:

LOCATION INDICATION VALUE NORMAL To Containment Leakage t o sump 9 Pm gPm R-11 C Pm C Pm R-12 ;Pm CPm Humidity (41517) N/A Z Cont. E l . 626' Amb.

A i r Temp (15187) O F O F To Component R- 17 CPm CPm Cool ing Surge Tank Level Change 9 Pm NA To Steam R - 15 C Pm CPm

enera t o r R- 19 C Pm CPm ro Waste PRT Level Change gPm NA 3 i sposal RCDT Level Change gPm NA D E S C R I P T I O N OF LEAK

SOURCE :

EFFECT ON PLANT OPERATION :

PLANT OPERATION MAY SAFELY CONTINUE: YES NO PERFORMED BY DATE APPROVED BY

DATE

D i r e c t o r Nuclear Operations & Maintenance, Operations Manager.

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

WISCONSIN PUBLIC SERVICE COapORATION NO - A - RC - 36C KEWAUNEENUCLEARPOWERPLANT TITLE ABNORMAL RXCP OPERATION EMWGENCY OPEdRATING PROCEDURES DATE OCT 06 2005 PAGE 17 of 18 A'ITACHMRNT B RBSPONSE TO ABNORMAL #2 SEAL LEAKOFF FLOW

1. Determine #2 S e a l Leakoff Flow by m o n i t o r i n g t h e change i n RCDT l e v e l .

NOTE: High s t a n d p i p e l e v e l i n d i c a t e s e x c e s s i v e #2 s e a l l e a k a g e . Low s t a n d p i p e l e v e l i n d i c a t e s e x c e s s i v e #3 s e a l l e a k a g e .

2. Perform 47015-1(47015-L). RXCP A(B) S t a n d p i p e High/Low t o d e t e r m i n e i f s t a n d p i p e l e v e l i s h i g h o r low.

NOTE: T o t a l S e a l Leakoff Flow e q u a l s #1 S e a l Leakoff Flow p l u s 82 S e a l Leakoff Flow.

3. I F T o t a l S e a l Leakoff Flow is g r e a t e r t h a n 8.0 gpm. THEN perform a n immediate Z C P shutdown:

a. Ip r e a c t o r 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 .

b. S t o p a f f e c t e d RXCP(s).

c. C l o s e PS-lA(1B). P z r Spray, i n a f f e c t e d l o o p .

d. WHEN RXCP h a s come t o a complete s t o p . THEN c l o s e CVC-207A(B). #1 S e a l Leakoff Isol.

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

4. I F #2 S e a l Leakoff Flow g r e a t e r t h a n 1.1 gpm. THEN perform an o r d e r l y E C P shutdown :

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

b. S t o p a f f e c t e d RXCP(s) w i t h i n 8 h o u r s .

c. C l o s e PS-lA(1B). P z r Spray. i n a f f e c t e d l o o p .

d. Refer t o TS 3 . 1 . d f o r RCS l e a k r a t e l i m i t a t i o n s .

5. -

I F #2 S e a l Leakoff Flow g r e a t e r t h a n 0.5 gpm. THEN perform t h e f o l l o w i n g :

a. Monitor pump and s e a l i n d i c a t i o n s .

b. Trend p a r a m e t e r s on PPCS.

c. C o n s u l t w i t h s t a t i o n management t o d e t e r m i n e i f a f f e c t e d RXCP(s) should b e 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 - 0101, Expectations & Standards RF-03.01, 2.7

~

Page 15

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

Minimum Instrument Channels Applicability Action

1. Noble Gas Activity Monitor

a. R-13 or R-14

- Waste Gas Holdup System (auto-isolation)

- Auxiliary Building Ventilation System I

- Containment Purge 2 line 4 (auto-isolation) 5

b. R-12 or R-21 6

- Containment purge-36duct 1 *

(auto-isolation) 6

c. R-15 1 *

- Condenser Evacuation System 5

. Radioiodine & Particdate Samplers

a. Containment Building Vent (R-21)

b. Auxiliary Building Vent (R-13 or 1

7 R-14) 1

7

, Sampler Flow Rate Measuring Devices

a. Containment Building Vent Sampler (R-2 1)

b. Auxiliary Building Vent Sampler 1

8 (R-13 or R-14) 1

8

Atalltimes 3-18 REV. 9 12l0212005

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

TABLE NOTATIONS Action 4 - 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.

Action 5 - 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 .

4ction 6 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

4ction 7 - 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.

4ction 8 - 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 KICWlLDNEE POWERSTATloN TITLE R a d i a t i o n M o n i t o r i n g System OPERATING PROCEDURE DATE APR 12 2006 P m 19 of 27 4.3.8 CONTINUED

b. t h e r e i s new f u e l i n t h e New Fuel P i t , THEN REQUEST R a d i a t i o n P r o t e c t i o n s e t up p o r t a b l e m o n i t o r w i t h an a u d i b l e a l a r m s e t a t 15 mr/hr.

C. POSITION k e y s w i t c h t o OFF.

d. PERFORM d a i l y source check on p o r t a b l e m o n i t o r .

9. R- , Conta inment P a r t i c u l a t e :

a. -

IF Containment P u r g d V e n t i s i n progress. THEN VERIFY R-12 OR R - 2 1 o p e r a t i n g and a l i g n e d t o sample stack. OR STOP Purge/Vent.

b. -

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

C. NOTIFY R a d i a t i o n P r o t e c t i o n .

d. REFER t o Tech Spec 3.1.d.

e. POSITION R-11/12 Pump C o n t r o l s w i t c h t o OFFIRESET.

f. POSITION Keyswitch t o KEYPAD t o d e f e a t a u t o m a t i c f u n c t i o n s .

NOTE: P o s i t i o n i n g Keyswitch t o o f f w i l l a c t u a t e automatic functions.

9. IF a u t o m a t i c f u n c t i o n s a r e r e q u i r e d . THEN POSITION Keyswi t c h t o OFF.

10. R-12. Containment Gas:

a. fi Containment Purge/Vent i s i n progress, JHEN VERIFY R-21 o p e r a t i n g and a l i g n e d t o sample stack, STOP Purge/Vent.

b. IF Containment Purge/Vent i s Ji0-J i n p r o g r e s s , THEN A L I G N R-21 t o sample Containment p e r Step 4.2.10.

c. NOTIFY R a d i a t i o n P r o t e c t i o n .

d. REFER t o Tech Spec 3.1.d and ODCM.

CONTINUED

DOMINION ENERGY KEWAUlWE NO.

N-FH-53-CLC

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-1 k V S

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II;EWAUNEE POWER STATION TIT= Pre-Refueling Checklist OPERATING PROCEDURE DATE JUN 13 2006 I PAGE 1 of 7 Michael swey James Langex REVIEWED BY APPROVED BY NUCLEAR YES PORC REVIEW YES SRO APPROVAL OF YES SAFETY RELATED REQUIRED TEMPORARY CHANGES 0 NO 0 NO REQUI RED 0 NO FIRST SECOND

~-

OPER OPER 1.0 PLANT REQUIREMENTS 1.1 N-FH-53-CLA Or N-FH-53-CLB complete. except a t l e a s t COMPLETED one door i n ea h personnel a i r l o c k s h a l l be capable o f b e i n g closed i n 30 minutes o r l e s s w i t h a d m i n i s t r a t i v e c o n t r o l s i n p l a c e 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 c l o s e d when t h e r e a c t o r vessel head o r upper i n t e r n a l s a r e l i f t e d . (TS 3.8) 1.2 RCS boron c o n c e n t r a t i o n s h a l l be v e r i f i e d g r e a t e r t h a n 2 REQUIRED o r equal t o COLR S p e c i f i e d R e f u e l i n g Boron C o n c e n t r a t i o n 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 d u r i n g f u e l movement.

Reactor C a v i t y Boron Concentrat on : P Pm RHR System Boron Concentration: PPm COLR R e f u e l i n g Boron Concentrat on : PPm 1.3 SFP boron c o n c e n t r a t i o n s h a l l be v e r i f i e d g r e a t e r than 2 REQUIRED o r equal t o COLR S p e c i f i e d R e f u e l i n g Boron C o n c e n t r a t i o n 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 d u r i n g f u e l movement.

SFP Boron Concentration: PPm COLR R e f u e l i n g Boron Concentrat on : PPm 1.4 Reactor has been v e r i f i e d subcr t i c a l f o r g r e a t e r than VERIFIED 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months . [CAP122551 1.5 RCS and Spent Fuel Pool Temperature s h a l l be v e r i f i e d VERIFIED g r e a t e r 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

DOMINION ENERGY KEWAUNEE I No- N - FH CLC KEWAUNEE POWJER STATION I TITLE Pre-Refuel ing Check1 is t OPERATING PROCEDURE DATE JUN 13 2006 PAGE 2 of7 DATE FIRST SECOND OPER OPER 2.0 SYSTEM EQUIPMENT STATUS 2.1 Reactor E n g i n e e r i n g c o r e s t a t u s system s e t up i n SET UP C o n t r o l 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 s u c t i o n temperature l e s s t h a n o r 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 S p e c i a l V e n t i l a t i o n system OPERABLE operable.

NOTE: C o n t r o l Room P o s t - A c c i d e n t R e c i r c u l a t i o n System may be considered operable i f b o t h t r a i n s a r e operable if one t r a i n i s o p e r a t i n g i n r e c i r c mode, p r o v i d e d i t s emergency D / G i s a l s o o p e r a b l e t o support f u e l h a n d l i n g a c c i d e n t a n a l y s i s assumptions.

2.4 C o n t r o l Room P o s t - A c c i d e n t R e c i r c u l a t i o n System OPE RAB LE Operable.

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

1. System p i p i n g and valves a r e operable f o r a t l e a s t 0 P E RAB LE one f l o w p a t h f o r b o r i c a c i d i n j e c t i o n t o RCS.

2.6 Reactor C a v i t y Level i s g r e a t e r than o r equal t o 64%. 264%

2.7 P l a n t Process Computer:

1. P o i n t L9053A o r L9054A. R e f u e l i n g Water Level A ( B ) WR:

a. V E R I F Y L9053A o r L9054A on SCAN. SCAN NIA

b. VERIFY a l a r m l i m i t s f o r s e l e c t e d p o i n t L9053A >64% NIA o r L9054A e s t a b l i s h e d per PPCS RHR Mid-Loop L i m i t s screen a r e g r e a t e r than 64%.

2. P o i n t N8031G. SR N-31 Counts (N8032G. SR N-32 Counts 1:

a. VERIFY N8031G (N8032G) on SCAN. SCAN N/A CONTINUED

DOMINION ENERGY KEWAUNEE

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I NO.

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N-FH-53-CLC KEWAUNEE POWEPSTATION I TITLE Pre-Refuel i n g Check1 is t OPERATING PROCEDURE I DAm- JUN 13 2006 I PAGE 3 of 7 DATE FIRST SECOND OPER OPER 2.7.2 CONTINUED

b. IF r e a c t o r vessel c o n t a i n s fuel. J tJJ CPSINA NIA CONTACT Reactor Engineering t o p r o v i d e an e s t i m a t e 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 V E R I F I ED NIA rate.

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

e. S i n g l e P o i n t Change required. CONTACTEDINA NIA THEN CONTACT Nuclear Computer Group.

3. computer i s a v a i l a b l e f o r trending. THEN APPLIESINA NIA TREND t h e f o l l o w i n g p o i n t s (PPCS Operations -

P r o t e c t e d 1. Group 5 ) every 10 minutes:

Point I D S Points F0626A Lo Head T r n A RHR To RCS GOO1 1 G R - 1 1 Containment P a r t GOO126 R - 1 2 Containment Gas T0627A RHR Hx Out Loop Hdr Temp T0630A RHR Pump Suct Hdr Temp R81OOG RCS Boron Conc Calc N8031G SR N-31 Counts N8032G SR N-32 Counts LO11 2 A Volume C o n t r o l Tank L v l

DOMITUON ENERGY KEWAUNEE I NO. N-FH-53-CLC KEWAUNEE POwEa STATION OPERATING PROCEDURE TITLE P r e - R e f u e l i n g C h e c k l i s t I DAm JUN 13 2006 I PAGE 4 of 7 II DATE FIRST SECOND OPER OPER 2.8 Radiation Monitoring:

1. R-2/81041. Containment Area M o n i t o r OPERATING. OPERATING

2. R-5/81042. Fuel Handling Area M o n i t o r OPERATING. OPERATING

3. PERFORM t h e f o l l o w i n g :

a. V E R I F Y R-12/81049. Containment Gas. OPE RAT I NG OPERATING.

b. V E R I F Y R-21/81058. Containment Vent, OPE RAT ING OPERATING.

c. V E R I F Y Containment Vent I s o l a t i o n w i l l a c t u a t e on h i g h r a d i a t i o n s i g n a l a s f o l l o w s :

CPCR0177871

1. IF Reactor B u i l d i n g Vent i s NOT o p e r a t i n g . APPLIES/NA 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 0 PEN / AUTO Supply Valve A B. RBV-2/CV-31126. Cntmt Purge/Vent OPEN/AUTO Supply Valve B C. RBV-3/CV-31124, Cntrnt Purge/Vent OPEN/AUTO Exhaust Valve A D. RBV-4/CV-31123, Cntmt Purge/Vent OPEN/AUTO Exhaust Valve B

2. PERFORM f u n c t i o n a l t e s t of R-12 per P E RFO RMED N - RM - 45.

3. VERFIY REV-2 CLOSED. CLOSED

4. V E R F I Y RBV-3 CLOSED. CLOSED

5. PERFORM f u n c t i o n a l t e s t of R-21 p e r PERFORMED N- RM- 45.

6. V E R F I Y RBV-1 CLOSED. CLOSED

7. V E R F I Y RBV-4 CLOSED. CLOSED CONTINUED

!i DOMINION ENERGY KEWAUNEE NO. N-FH-53-CLC I KEWAUNEE POWEB STATION TITLE Pre-Refuel i n g Check1 i s t OPERATING PROCEDURE DA!lE JUN 13 2006 I PAL;E 5 of7 DATE FIRST SECOND OPER OPER 2.8.3.c CONTINUED

8. IF Reactor B u i l d i n g Vent w a s o p e r a t i n g . SHUT DOWN/NA THEFl SHUT DOWN Reactor B u i l d i n g Vent per N- RBV- 18B.

9. IF o p e r a t i o n o f Reactor B u i l d i n g Vent i s STARTED/NA d e s i r e d . THEN START Reactor B u i l d i n g Vent p e r N-RBV-18B.

4. R-23/81061. C o n t r o l Room Vent M o n i t o r

a. R-23. OPERATING. OPERATING

b. VERIFY C o n t r o l Room Post-Accident V E R I F I ED R e c i r x u l a t i o n System w i l l a c t u a t e on h i g h r a d i a t i o n s i g n a l , GO TO N-RM-45.

2.9 Nuclear I n s t r u m e n t a t i o n :

1. Two Source Range Detectors a r e m o n i t o r i n g Neutron CONFIRMED f l u x and each p r o v i d e v i s u a l i n d i c a t i o n i n C o n t r o l Room.

2. One Source Range Detector s h a l l have an a u d i b l e CONFIRMEO count r a t e i n containment.

3. IF any f u e l i s i n t h e r e a c t o r vessel, THEN Source ALARM SET/NA Range D e t e c t o r s ' High Flux a t Shutdown a l a r m s e t a t approximately 3 times t h e i r e x i s t i n g count rate.

4. IF r e a c t o r vessel c o n t a i n s f u e l , THEN Source ALARM SET/NA Range D e t e c t o r s ' High F l u x a t Shutdown a l a r m s e t a t value e s t a b l i s h e d d u r i n g c o r e o f f - l o a d .

5. High F l u x a t Shutdown a l a r m and Containment a l a r m test:

a. P O S I T I O N Level T r i p s w i t c h t o BYPASS. BYPASS

b. ROTATE Level A d j u s t potentiometer f u l l y ROTATED counter-clockwise.

CONTINUED

DOMINION ENERGY KEWAUNEE I NO. N-FH-53-CLC KEWAUNEE POwEa STATION I TITLE P r e - R e f u e l i n g C h e c k l i s t OPERATING PROCEDURE I MTE JUN 13 2006 I PAGE 6 of 7 DATE FIRST SECOND OPER OPER 2.9.5 CONTINUED

c. POSITION Operat on S e l e c t o r s w i t c h t o LEVEL LEVEL ADJUST ADJUST.

d. ROTATE Level A d j u s t potentiometer c l o c k w i s e ROTATED u n t i l High F l u x a t Shutdown alarm actuates.

e. VERIFY alarm i n containment actuated. ACTUATED

f. WHEN High F l u x a t Shutdown a l a r m t e s t complete. THEN PERFORM t h e f o l l o w i n g :

1. ROTATE Level A d j u s t Potentiometer f u l l y ROTATED counterclockwise.

2. POSITION Operation S e l e c t o r Switch t o NORMAL NORMAL.

3. P O S I T I O N Level T r i p s w i t c h t o NORMAL. NORMAL 2.10 V E R I F Y Containment Evacuation A l a r m on MCC A operable. OPERA8LE 2.11 Communications e s t a b l i s h e d between C o n t r o l Room, ESTABLISHED 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 CONFIRMED f u e l movement.

2.13 R a d i a t i o n m o n i t o r on M a n i p u l a t o r Crane OPERATING. OP ERAT ING

1. R a d i a t i o n mon t o r s e t t o a l a r m a t 10 R/hr. SET 10 R / H R 2.14 Fuel T r a n s f e r System g a t e v a l v e open. OPEN 2.15 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. 0 P ERA B LE 2.16 Spent Fuel Pool:

1. SFP temperature l e s s t h a n 125°F. <125" F

2. SFP c o o l i n g system o p e r a t i n g . OPERATING

DOMINION ENERGY KEWAUNEE NO. N-FH-53-CLC KEWAUNEE FOWEE STATION TITLE Pre-Refueling Checklist OPERATING PROCEDURE MTE JUN 13 2006 PAGE 7 of7 DATE F I R S T SECOND OPER OPER 2.17 Spent Fuel Pool Sweep System:

1. Exhaust Fans ( 2 ) o p e r a t i n g . 0 PE RAT I NG

2. Charcoal F i l t e r s ( 2 ) i n s e r v i c e (Bypass CLOSED). I N SERVICE

3. SP-17-126. SFP Sweep System F i l t e r T e s t i n g , has COMPLETED 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 C o i l U n i t f a n s OPERATING OP ERAT I NG (SW NOT r e q u i r e d ) .

3.0 M O N I T O R I N G AND ALARM REQUIREMENTS 3.1 Containment I s o l a t i o n A c t i v e S t a t u s Panel operable. OPERABLE 3.2 Sequenti a1 Events Recorder operable. OPERABLE Nf A 4.0 REMOTELY OPERATED AND AUTOMATIC VALVES 4.1 NONE 5.0 LOCAL VALVE P O S I T I O N S 5.1 NONE Performed By I I n it Date Name ( P r i n t ) S i gna t u r e Performed By f I n it Date Name ( P r i n t ) Signature Performed By I I n it Date Name ( P r i n t ) S i gna t u r e S h i f t Manager Date (PrintISign)

SPVR Nuclear S h i f t Ops Date (Print/Sign)

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

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KEWAUNFX NUcLeABpowERPLhlyT TITLE R e f u e l i n g D a i l y C h e c k l i s t OPERATING PROCEDURE DATE OCT 11 2005 I P m 1 of 10 Mark Larger Je5ey Simon RgvIgwED BY APPROVED BY

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I NUCLEAR YES PORC REVIEW YES SRO APPROVAL OF YES SAFETY RELATED REQUIRED TEMPORARY CHANGES 0 NO REQUI RED 0 NO

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DATE FIRST SECONC

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OPER OPER 1.0 PLANT REQUIREMENTS 1.1 N-FH-53-CLC has been completed. COMPLETED 1.2 Reactor C a v i t y and Residual Heat Removal System boron VERIFIED c o n c e n t r a t i o n s h a l l be v e r i f i e d g r e a t e r than o r equal t o 2500 ppm every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months d u r i n g f u e l movement.

1.3 Spent Fuel Pool boron c o n c e n t r a t i o n s h a l l be v e r i f i e d VERIFIED g r e a t e r than o r equal t o 2500 ppm every 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months d u r i n g f u e l movement.

1.4 RCS and Spent Fuel Pool Temperatures s h a l l be V E R I FI ED v e r i f i e d g r e a t e r than o r equal t o 50°F d u r i n g fue movement. CCA0192021 2.0 SYSTEM EQUIPMENT STATUS 2.1 RECORD Tagout C o n t r o l Sheet number f o r l o c a l valves. RECORDED 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. 0 PERAB LE

2. RHR pump s u c t i o n temperature l e s s than or equal 440° F t o 240°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 OP E RAB LE operabl e.

2.4 C o n t r o l Room Post-Accident R e c i r c u l a t i o n System OPERABLE Operable. C o n t r o l Room Post-Accident R e c i r c u l a t i o n System may be considered operable i f b o t h t r a i n s a r e operable i f one t r a i n i s o p e r a t i n g 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 h a n d l i n g a c c i d e n t a n a l y s i s assumptions.

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

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OPERATING PROCEDURE DATE OCT 11 2005 2 of 10 DATE F I R S T SECOND

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OPER OPER 2.5 B o r i c A c i d I n j e c t i o n F l o w Path:

1. System p i p i n g and valves a r e operable f o r a t OPERA6LE l e a s t one f l o w p a t h f o r b o r i c a c i d 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 RECORDED Total gal

2. RECORD l a t e s t l e a k r a t e 9 Pm 2.7 Reactor C a v i t y Level g r e a t e r t h a n o r equal t o 64%.

2.8 P l a n t Process Computer:

1. IF computer i s a v a i l a b l e f o r t r e n d i n g . THEN APPLIES/NA N/A TREND t h e f o l l o w i n g p o i n t s (PPCS Operations -

P r o t e c t e d 1. Group 5) every 10 minutes:

Poi nt I Ds Points F0626A Lo Head T r n A RHR To RCS GOOllG R-11 Containment P a r t GOO12G R-12 Containment Gas T0627A RHR Hx Out Loop Hdr Temp T0630A RHR Pump Suct Hdr Temp R8100G RCS Boron Conc Calc N8031G SR N-31 Counts N8032G SR N-32 Counts L0112A Volume C o n t r o l Tank L v l

WlsCoNSIN PUBLIC SERVICE CORPORATION NO. N-FH-53-CLD I

I I ( E w A ~ E ~ TITLE

~ ~ R e f~u e l i nPg D a~i l y C h e c k l i s t OPERATING PROCEDURE DA!CE OCT 11 2005 PAGE 3 of 10 DATE FIRST SECOND OPER OPER 2.9 Radiation Monitoring:

1. R-2/81041 Containment Area Monitor OPERATING. 0 PE RAT ING

2. R-5/81042 Fuel Handling Area Monitor OPERATING. OPERATING PERFORM t h e f o l l o w i n g :

4 a.

b.

c.

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

VERIFY R-21/81058. Containment Vent.

OPERATING.

VERIFY Containment Vent I s o l a t i o n w i l l a c t u a t e on h i g h r a d i a t i o n s i g n a l as f o l l o w s :

C PCROl7787 1 OPE RAT ING 0 P E RAT I NG

1. -

I F Reactor B u i l d i n g Vent i s )JOT APPLIES/NA o p e r a t i n g , 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 0 PEN /AUTO Supply V a l v e A

6. RBV-Z/CV-31126. Cntmt Purge/Vent OPEN / AUTO Supply V a l v e B C. RBV-3/CV-31124. Cntmt Purge/Vent OPEN/AUTO Exhaust Valve A

0. RBV-4/CV-31123. Cntmt Purge/Vent OPEN /AUTO Exhaust Valve B

2. PERFORM f u n c t i o n a l t e s t o f R-12 p e r PERF 0 RMED N- RM-45.

3. V E R F I Y RBV-2 CLOSED. CLOSED

4. V E R F I Y RBV-3 CLOSED. CLOSED

5. PERFORM f u n c t i o n a l t e s t o f R - 2 1 p e r PERFORMED N - RM- 45.

6. V E R F I Y R B V - 1 CLOSED. CLOSED

7. V E R F I Y RBV-4 CLOSED. CLOSED CONTINUED

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

~WAUNJtENU~POWERPWNT TITLE R e f u e l i n g D a i l y C h e c k l i s t OPERATING PROCEDURE DATE OCT 11 2005 PAGE 4 of 10 DATE FIRST SECONC OPER OPER 2.9.3.c CONTINUED

8. TF Reactor B u i l d i n g Vent w a s o p e r a t i n g . SHUT DOWN/NA THEN SHUT DOWN Reactor B u i l d i n g Vent p e r N-RBV-188.

9. o p e r a t i o n o f Reactor B u i l d i n g Vent i s STARTED/NA d e s i r e d , THEN START Reactor B u i l d i n g Vent per N - RBV- 188.

4. R-23/81061 Control Room Vent M o n i t o r OPERATING. OPERATING 2.10 Nuclear I n s t r u m e n t a t i o n :

1. Two Source Range Detectors a r e m o n i t o r i n g CONFIRMED Neutron f l u x and each p r o v i d e v i s u a l i n d i c a t i o n i n C o n t r o l Room.

2. One Source Range D e t e c t o r s h a l l have an a u d i b l e CONFIRMED count r a t e i n containment.

3. any f u e l i s i n t h e r e a c t o r vessel, ALARM SET/NA rHEN Source Range D e t e c t o r s ' High F l u x a t Shutdown a l a r m s e t a t approximately 3 times t h e i r e x i s t i n g count r a t e .

4. r e a c t o r vessel c o n t a i n s NO f u e l , ALARM SET/NA THEN Source Range D e t e c t o r s ' High F l u x a t Shutdown a l a r m set a t value established during core o f f - 1oad.

5. High F l u x a t Shutdown a l a r m and Containment a l a r m test:

a. P O S I T I O N Level T r i p s w i t c h t o BYPASS. BYPASS

b. ROTATE Level A d j u s t p o t e n t i o m e t e r f u l l y ROTATED counter-clockwise.

c. P O S I T I O N Operation S e l e c t o r s w i t c h t o LEVEL LEVEL ADJUST ADJUST.

d. ROTATE Level A d j u s t p o t e n t i o m e t e r c l o c k w i s e ROTAT ED u n t i l High F l u x a t Shutdown a l a r m actuates.

CONTINUED

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NO. N-FH-53-CLD WISCONSJN PUBLIC SERVICE CORPORATION KEWAUNEE NDCLEARPIlWERPWNT TITLE R e f u e l i n g D a i l y C h e c k l i s t OPERATING PROCEDURE DATE OCT 11 2005 I PAGE 5 of 10 OAT FIRST SECOND

~-

OPER OPER 2.10.5 CONTINUED

e. VERIFY a l a r m i n containment actuated. ACTU AT ED

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

1. ROTATE Level A d j u s t Potentiometer f u l l y ROTAT ED counterclockwise.

2. P O S I T I O N Operation S e l e c t o r Switch t o NORMAL NORMAL.

3. P O S I T I O N Level T r i p s w i t c h t o NORMAL. NORMAL 2.11 V E R I F Y Containment Evacuation A l a r m on MCC A operable. OPERABLE 2.12 Communications e s t a b l i s h e d between C o n t r o l Room, ESTABLISHED 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 CONFIRMED f u e l movement.

2.14 R a d i a t i o n m o n i t o r on Manipulator Crane OPERATING. OPERATING 2.15 Steam Generator Nozzle Dam leakage acceptable. ACCEPTABLE 2.16 S h i e l d B u i l d i n g Annulus Gates:

1. V E R I F Y gate a t Personnel A i r l o c k CLOSED and CLOSE0 and LOCKED. LOCKED

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

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

~WA~ENIJCLEAR~OWERPLA~JT I TITLE Refuel i n g D a i l y Check1 is t OPERATING PROCEDURE DATE OCT 11 2005 PAGE 6 of 10 DATE FIRST SECOND OPER O P E R 2.18 Containment Vessel F l e x i b l e Seals:

1. Containment Vessel P r e s s u r i z a t i o n t e s t :

a. Penetration (42N):

Loop Seal water l e v e l i n NORMAL range. NORMAL OR 0 F i b e r O p t i c Cable P e n e t r a t i o n Seal INSTALLED OR Contai nment Vessel B1 ind F1ange INSTALLED. INSTALLED B l i n d Flange Test Conn CAPPED. CAPPED

b. R e f u e l i n g Cable P e n e t r a t i o n (43N):

0 Loop Seal water l e v e l i n NORMAL range. NORMAL OR 0 Containment Vessel B l i n d Flange INSTAL ED. INSTALLED AND 0 B l i n d Flange Test Conn CAPPED. CAPPED 2.19 Spent Fuel Pool Sweep System:

1. Exhaust Fans ( 2 ) OPERATING OP ERAT ING

2. Charcoal F i l t e r s ( 2 ) i n s e r v i c e (Bypass CLOSED) I N SERVICE 2.20 A t l e a s t two Containment Fan C o i l U n i t Fans o p e r a t i n g . 0 PE RAT ING (SW NOT r e q u i r e d )

3.0 MONITORING AND ALARM REQUIREMENTS 3.1 Containment I s o l a t i o n A c t i v e Status Panel operable. OPERABLE 3.2 Sequential Events Recorder operable. 0 PE RAB LE

I WEECONSIN PUBLIC SERVICE CORPORATION KEWAUNEENUCLEARPOWERPLANT 1 NO. N-FH-53-CLD TITLE R e f u e l i n g D a i l y C h e c k l i s t OPERATING PROCEDURE I DATE OCT 11 2005 I PAGE 7 of 10 DATE FIRST SECONC OPER OPFR 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 v a l v e s s h a l l have an operable automatic i s o l a t i o n v a l v e OR a c l o s e d i s o l a t i o n v a l v e . For i n o p e r a b l e valves denote t h e a s s o c i a t e d Tagout number i n t h e "Second Oper" i n i t i a l s column.

4.1 Mechanical Console C NG-107/CV-31253. N i t r o g e n Supply t o S I Accumulators 0PERAB LE CC-601A/MV-32084. Component Cooling t o RXCP A CLOSED OR cc SYS INTACT TO RXCP A CC-612A/MV-32086. RXCP A Component C o o l i n g Return I s o l CLOSED OR cc SYS INTACT TO RXCP A CC-601B/MV-32085. Component C o o l i n g t o RXCP B CLOSED OR cc SYS I N T A C T TO RXCP B CC-612B/MV-32087. RXCP 6 Component C o o l i n g Return I s o l CLOSED OR cc SYS I N T A C T TO RXCP B CC-653lMV-32082. Excess Letdown Hx Comp C o o l i n g Return OPERABLE 4.2 Mechanical Console .E LD-6/CV-31234. Letdown L i n e I s o l a t i o n OPERABLE CVC-212/MV-32115. RXCP S e a l Water Return I s o l a t i o n 0 P E RAB LE CVC-211/MV-32124. RXCP Seal Water Return I s o l a t i o n OPERABLE

WISCONSIN PUBLIC SERVICE CORPORATION NO. N-FH-53-CLD KEWAUNEENOCLEAaPOWERPLANT TITLE R e f u e l i n g Daily C h e c k l i s t OPERATING PROCEDURE DATE OCT 11 2005 PAGE 8 of 10 DATE F I R S T SECOND

~-

OPER OPER 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 0 PE RAB L E AS-Z/CV-31384. Containment A i r Sample I s o l a t i o n 6 OPERABLE AS-32/CV-31385. Containment A i r Sample I s o l a t i o n C OPERABLE MD(R)-134/CV-31136. Cntmt Sump Pumps Discharge Header OPERABLE 1s o l MD(R)-135/CV-31137. Cntmt Sump Pumps Discharge Header OPERABLE 1s o l 4.4 V e r t i c a l Panel A RC-402/CV-31263, P r e s s u r i z e r Steam Sampling I s o l a t i o n OP E RAE LE RC-412/CV-31264. P r e s s u r i z e r L i q u i d Sampling I s o l a t i o n OPE RAE LE RC-422/SV-33092. Rx Coolant Hot Leg Sampling I s o l a t i o n OPERABLE RC-403/CV-31267. P r e s s u r i z e r Steam Sampling I s o l a t i o n OPERABLE RC-413/CV-31268, P r e s s u r i z e r L i q u i d Sampling I s o l a t i o n OPERABLE RC-423/SV-33327. Rx Coolant Hot Leg Sampling I s o l a t i o n OPERABLE MU-1010-1/CV-31261, P r z r R e l i e f Tank Make Up Water 0 PERAB LE Is o l MG(R)-512/CV-31259. P r z r R e l i e f Tank Gas Sampling I s o l OPERABLE MG(R)-513/CV-31260. P r z r R e l i e f Tank Gas Sampling I s o l OPERABLE NG-302/CV-31298. P r z r R e l i e f Tank N i t r o g e n Supply I s o l OPERABLE RC-507/CV-31134. Rx C l n t D r a i n Pump Disch Header I s o l 0 P E RAE! LE RC-508/CV-31135. Rx C l n t D r a i n Pump Disch Header I s o l 0 P E RAB LE MG(R)-509/CV-31132. RCDT Vent To Waste Gas Header OPERABLE MG(R)-510/CV-31133. RCDT Vent To Waste Gas Header 0 P E RAB LE MG(R)-503/CV-31216. RCDT To Gas Anzr Header I s o l a t i o n OPERABLE CONT I NU ED

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

KEWAUNEE NUCLEARPOWERPLANT II TITLE R e f u e l i n g D a i l y C h e c k l i s t OPERATING PROCEDURE DATE OCT 11 2005 PAGE 9 of 10 OAT E F I R S T SECONC OPER OPER 4.4 CONTINUED MG(R)-504/CV-31217. RCDT To Gas Anzr Header I s o l a t i o n OPERABLE MD(R)-323A/MV-32390. Deaerated D r a i n s Tank Cntmt Oisch OPERABLE Isol A MD(R)-323B/MV-32391, Deaerated D r a i n s Tank Cntmt Disch OPERABLE Isol B WG-31O/SV-33655. Deaerated Drains Tank Vent Outside OPERABLE Cntmt CVC-54/SV-33651. VCT Vent t o Cntmt OPE RAE LE VB-lOA/CV-31337. Power Operated Cntmt Vacuum Breaker A 0 PERAB LE VB-lOB/CV-31338. Power Operated Cntmt Vacuum Breaker B OPERABLE LOCA-201B/CV-31727. Post LOCA Hydrogen Recombiner 8 To 0 P ERAE LE Cntmt LOCA-lOOB/CV-31725. Post LOCA Hydrogen t o Recombiner B OP E RAB LE SA-7003B/MV-32148. Hydrogen D i l u t i o n To Containment OPERABLE LOCA-2B/MV-32146. Post LOCA Hydrogen Cntmt Vent I s o l B 0 P E RAE LE RBV-l/CV-31125. Cntmt Purge/Vent Supply Valve A OPERABLE RBV-4/CV-31123. Cntmt Purge/Vent Exhaust Valve A OPERABLE RBV-2/CV-31126. Cntmt Purge/Vent Supply Valve B 0 PE RAB L E RBV-3/CV-31124. Cntmt Purge/Vent Exhaust Valve B OPERABLE

WISCONSIN PUBLIC SERVICE COIWORATION KEWAUNEE NUCLEARPOWERPLANT 1 NO.

TITLE N-FH-53-CLD Refueling Daily Checklist OPERATING PROCEDURE I DATE OCT 11 2005 I PAGE 10 of 10 DATE FIRST SECOND OPER OPER 4.5 Mechanical Console A BT-31A/CV-31334. S / G Sample I s o l V l v s OPERABLE BT-31B/CV-31270. S/G Sample I s o l V l v s OPERABLE BT-32A/CV-31335. S / G Sample I s o l Vlvs 0 PE RAB LE BT-32B/CV-31271. S/G Sample Isol V l v s OPERABLE I BT-3A/MV-32078 S / G A Blowdown I s o l a t i o n Valve A2 OPERABLE I BT-3B/MV-32080 S/G B Blowdown I s o l a t i o n Valve 82 OPERABLE 5.0 LOCAL VALVE P O S I T I O N S 5.1 None 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 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.

6.3.2.2 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.

6.3.2.3 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.

6.3.2.4 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.

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

9 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.

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

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

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

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

9 The bases for decisions shall be communicated throughout the organization.

9 The effectiveness of decisions shall be periodically evaluated.

P 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. No. RF-03.O 1 Rev. J Kewaunee Nuclear Power Plant Refueling Procedure Date OCT 14 2004 Page 1 of 20 Reviewed By Tim Wiltman Nuclear 0 Yes PORC p~yes SRO Approval Of 0 Yes Safety Review Temporary Related 0 No Required 0 No Changes Required 0 No

- - INFREQUENTLY PERFORMED TEST ***

1.0 Purpose 1.1 This procedure provides instruction for fuel movement during a refueling outage.

2.0 General Notes 2.1 Contact the Reactor Engineering Group (RXE) for resolution of any problems, concerns, or questions.

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

2.3 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.

2.4 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.

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

2.6 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.

2.7 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-1OB, 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 Bwill start the SI Pump without having a Component Cooling Water Pump running for support.

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

ML070240208

Text

1.0 REFERENCES

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