ML052240188

Facility Post-Examination Comments for the Palisades Initial Examination - May 2005

ML052240188
Person / Time
Site:	Palisades
Issue date:	05/23/2005
From:	Domonique Malone Nuclear Management Co
To:	NRC/RGN-III/DRS/OLB
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Text

FACILITY POST-EXAMINATION COMMENTS FOR THE PALISADES INITIAL EXAMINATION - MAY 2005

Palisades Nuclear Plant Operated by Nuclear Management Company, LLC June 7, 2005 NUREG 1021, ES-402 Regional Administrator U.S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Initial License Examination Comments In accordance with NUREG-1021, ES-402, Nuclear Management Company LLC (NMC) is submitting comments on the initial license examination administered at the Palisades Nuclear Plant during May 2005. contains the requested information Daniel J. Malone Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosure (1)

CC Project Manager, Palisades, USNRC (w/o enclosure)

Resident Inspector, Palisades, USNRC (w/o enclosure)

Bruce Palagi, Region 111, USNRC 27780 Blue Star Memorial Highway Covert, Michigan 49043-9530 Telephone: 269.764.2000 ntlE/1/

ENCLOSURE 1 PALISADES INITIAL LICENSE EXAMINATION COMMENTS 42 Pages Follow

Question # 13 During a Station Blackout what indication(s) are available to determine when Battery No. 1 (Dol) is approaching a fully discharged condition?

A. ONLY Voltage indication for Battery No. 1 can be used.

B. EITHER Voltage or Amperage indications for Battery No. 1 can be used C. ONLY Amperage indications for Battery No. 1 can be used.

D. EITHER Voltage, Amperage, CR annunciator, or Frequency indications for Battery No. 1 can be used.

NRC Answer Key: B Facilitv Comment:

Distractor B: "Either Voltage OR Amperage..."could be interpreted to imply that either voltage ALONE, or amperage ALONE could be used, but NOT both. While amperage does respond and may be helpful in diagnosing a battery near a fully discharged condition it cannot be used alone. High or low amperage can be indicative of battery loading. Without a relative voltage reading, amperage indication alone is not adequate for diagnosing a battery approaching a fully discharged condition.

EOP-3.0 Station Blackout, requires that if bus voltage drops to 105 volts that the shunt trip push button be pressed for that bus. This ensures the battery can perform its safety function prior to being overdutied. The requirement does not mention bus amperage. Therefore, Distractor A is also acceptable.

Facility Recommendation: Accept both A and B as correct.

Additional Facilitv References (Attached):

1. EOP-3.0, Station Blackout Recovery, Step 20 excerpt, page 22 of 52

2. EOP-3.0, Station Blackout Recovery, Step 20 Basis, page 49, 50 of 177

TITLE: STATION BLACKOUT RECOVERY lNSTRUCTlONS CAUTlON The following step should only be performed as a last resort since it results in separating the respective DC Bus from the Station Battery.

Buses D l l A and D21A will still be supplied from the Station Batteries.

20. E there is an obvious DC Bus problem which can NOT be immediately corrected, PERFORM ALL of the following:

a. E the condition is indicated on 125V DC Bus D10 pB bus voltage drops to 105 volts, PUSH Shunt Trip pushbutton "D-11 Incoming Power Trip' on Panel D11A.

b. E the condition is indicated on 125V DC Bus D20 pB bus voltage drops to 105 volts, PUSH Shunt Trip pushbutton "D-21 Incoming Power Trip" on Panel D21A.

c. GO TO EOP-9.0, "Functional Recovery Procedure" Bhlp REFER TO ONP-25.2, "Alternate Safe Shutdown Procedure."

0 = Continuously applicable step &= Hold Point

I PALISADES NUCLEAR PLANT EOP-3.C EMERGENCY OPERATING Revision YYSL.*..LI*l PROCEDURE BASIS 49 Of 171 I TITLE: STATION BLACKOUT RECOVERY BASIS CFN-153 SBO Step:

None The intent of this step is to respond to an obvious DC Bus problem that can not be immediately corrected. By stripping the battery from D10 or D20 the potential for having the DC buses for operation of the D/G and its associated loads is greatly improved.

Depending on the affected bus, the operator is directed to separate DC Bus D10 or D20, from it's respective Station Battery by pushing the associated Shunt Trip pushbutton on Panel D11Aor D21A.

In order to maintain a full compliment of safety grade instrumentation when the Reactor is not in a refueling condition, three of the four Preferred AC Buses must remain energized. In order to maintain three of the four Preferred AC Buses energized, at least three of the four vital DC bus sections must be energized (requiring both DC Buses D10 and D20 to be energized) or two inverters must be powered by one vital DC bus and a third inverter powered by the Bypass Regulator.

Stripping one or both of the DC Buses D10 or D20 will result in less than three of the four Preferred AC Buses being energized. This condition will warrant exiting this procedure and going to EOP-9.0, "Functional Recovery Procedure." ONP-25.2, "Alternate Safe Shutdown Procedure," is referenced to assist the operator with equipment control since DC control power from the deenergized DC Bus is not available.

Associatedes. C a u t i o n s . :

Caution alerts the operator that these actions should only be performed as a last resort since it results in separating DC Bus D10 or D20 from the Station Battery. Buses D11A and D21A will still be supplied from the Station Batteries and provide power to relays required to open the battery supply breakers to Buses D10 and D20 respectively.

De-energizing D10 or D20 does not remove battery power to Bus D11A or D21A.

viatiQ&mmEG:

A step to respond to a DC Bus problem that can not be immediately corrected is an addition to CEN-152 requirements.

'roc No PALISADES NUCLEAR PLANT

?evision EMERGENCY OPERATING 50 Of 177 PROCEDURE BASIS "lael ... .".

I TITLE: STATION BLACKOUT RECOVERY BASIS or Dew-:

Indication of an electrical ground or low voltage condition on either DC Bus must be promptly addressed in order to prevent the problem from affecting the battery's capability to perform to its analyzed availability rating. If the shunt trip breaker is open separating the DC Bus from its associated battery, then the safety function status check condition that three of the four Preferred AC Buses be energized will not be met. This condition creates the necessity of exiting to EOP-9.0, "Functional Recovery Procedure."

Question #23 The plant is operating at 100% Rx power when a failure of Cooling Tower Pump P-39A has caused condenser vacuum to degrade. Loss of Condenser Vacuum procedure ONP-14 has been entered. A rapid power reduction (per ONP-26) was ordered by the SRO. Following the power reduction, and reactor trip, condenser pressure stabilized at 15" Hg. During the rapid downpower, what was the fastest allowable rate of power reduction, and assuming condenser pressure remains constant what would PCS temperature be after the reactor trips?

A. 60%/Hr and 532°F B. 300%/Hr and 532°F C. 60%/Hr and 535°F D. 300%/Hr and 535°F NRC Answer Key: B Facilitv Comment; The question stem asks, "what would PCS temperature be." The briefing provided to the candidates just prior to the exam, in accordance with Appendix E of NUREG 1021, Rev. 9, instructed them to answer all questions based on actual plant operation, procedures, and references, and that if they believe the answer would be different based on simulator operation or training references, they should answer based on the actualplant.

By design, the turbine bypass valve (TBV) does control main steam header pressure at 900 psia (531.95 degrees F at saturation). However, pressure losses between the main steam header and the steam generators, along with efficiency losses in the steam generators, result in a stable Tave of slightly less than 535 degrees F.

This question and answer B reflect system design, but not actual plant response. Please see attached copies of both actual plant data and simulator response that show that actual PCS temperature (Tave) stabilizes at approximately 535 degrees F with the turbine bypass valve available.

Facility Recommendation: Change correct answer t o D.

Additional Facilitv References IAttachedL

1. PPC trend page from reactor trip report dated 07/21/98 with TBV available

2. PPC trend page from simulator reactor trip with TBV available.

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psig 531.87 FSISi 533.63 psia 885.86

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I 1/21/1998 14:44:2O 011211199a 14:50:20 G712111998 14:56:19 011211199a i4:41:59 A v e r a g e Temp H o t & ' C o l d L e g s deg F 534.38 TAVG PCS i'SP1 Tav deg F 534.57 PCS LG9P2 Tav deg F 535.06

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Question #66 A plant shutdown is required for refueling. When can the Operating Crew declare that they have reached Mode 6?

A. When the Reactor Head is removed with SDM > 1%

B. When the Reactor Head is removed with SDM N/A C. When the first Reactor Vessel Closure Bolt less than fully tensioned with SDM > 1%

D. When the first Reactor Vessel Closure Bolt less than fully tensioned with SDM N/A NRC Answer Key: D Facilitv Comment:

The question does not ask for the definition of Mode 6. The stem presents a decision point and asks, When can the Operating Crew declare that they have reached Mode 6? As soon as the first reactor vessel closure bolt is less than fully tensioned, the conditions of the stem are met.

Since both answers C and D contain this condition (less than fully tensioned), and since SDM is N/A for Mode 6, answers C and D are both correct.

Answers A and B are not correct, since the crew would have declared Mode 6 entry long before the conditions of A and B are true.

Facility Recommendation: Accept both C and D as correct.

Additional Facilitv References (Attachedh

1. Technical Specification Table 1.I-1, MODES

Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES

~~

% RATED AVERAGE PRIMARY REACTIVITY THERMAL COOLANT MODE TITLE CONDITION POWER@) TEMPERATURE (kd)

Power Operation 2 0.99 >5 NA Startup 2 0.99 55 NA Hot Standby < 0.99 NA 5 300 Hot ShutdownIb1 c 0.99 NA 300 > T,, > 200 Cold Shutdown@) c 0.99 NA 5 200 Refueling") NA NA NA (a) Excluding decay heat.

(b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned.

Palisades Nuclear Plant 1.1-7 Amendment No. 189

Question #73 The following plant conditions exist:

All Waste Gas Decay Tanks are full except the tank currently in service A Containment Purge is in Progress DIG 1-2 is currently running for surveillance testing Minimum crew manning is onsite due to a Holiday Waste Gas Decay Tank T-68B needs to be released but Radiation Monitor RE-I 113 is NOT OPERABLE. What conditions must exist for the WGDT to be released?

A. Radiation Monitor RE-I 113 must be returned to OPERABLE status The Containment Purge must be secured B. Two independent verifications of the release rate calculation are performed Two qualified Aux. Operators independentlyverify the WGDT discharge line-up Plant Stack Radiation Monitor is continuously monitored throughout the release C. Two independent tank samples are collected Two independent verifications of the release rate calculation are performed Two qualified Aux. Operators independently verify the WGDT discharge line-up The Containment Purge must be secured D. Two independent tank samples are analyzed Two independent verifications of the release rate calculation are performed Two qualified Aux. Operators independently verify the WGDT discharge line-up Plant Stack Radiation Monitor is continuously monitored throughout the release NRC Answer Key: C Facilitv Comment:

The stem of this question only lists some of the conditions needed to be in place to release a gas batch. It does not list ALL required conditions (e.g., main exhaust fan must be in service).

Answer C is correct, since it is reasonable to assume that a "collected" sample would also be "analyzed."

The last requirement in answer D, "Plant Stack Radiation Monitor is continuously monitored throughout the release," was originally intended to be incorrect, with the other three items being correct. However, the attached references show that the plant stack radiation monitor is continuously used as a monitoring instrument. Therefore, D is also correct.

Facility Recommendation: Accept both C and D as correct.

Additional Facilitv References (Attached):

1. SOP-IaA, Radioactive Waste System - Gaseous, p. 12, 13

2. Design Basis Document (DBD) 3.2.1.10, p. 67

3. Health Physics Procedure HP 6.6, Evaluation and Release of Waste Gas Decay Tank, p. 12

4. Health Physics Procedure HP 6.51, Radioactive Effluent Monitoring instrumentation and Equipment Requirements, Attachment 5, p. 2

5. Initial License Training Lesson Plan RMS, Radiation Monitoring System, p. 65 through 69.

PALISADESNUCLEARPLANT Proc No SOP-IeA SYSTEM OPERATING PROCEDURE Revision 36 Page 12 of 32 TITLE: RADIOACTIVE WASTE SYSTEM GASEOUS -

NOTE: The reason that MV-WG5OO is closed in the following step is that DT-1111 andlor CK-WG404 have historically been problematic. Leakage past this check valve or drain trap can result in fission gas in the Auxiliary Building or the Waste Gas Decay Tank Discharge Header and ultimately make its way to the Stack.

e. WHEN WGST pressure stops decreasing, THEN CLOSE the following valves:

1. MV-CRW782, DT-1111 Bypass Valve. (Waste Gas Surge Tank Room)

2. MV-WG500, WGST Drain Valve. (Waste Gas Surge Tank Room)

f. PLACE in AUTO C-50A and C-506 C-54.

7.5 TO RELEASE GAS FROM WASTE GAS DECAY TANKS

a. CHECK Containment purge is NOT in progress. REFER TO HP 6.14, Attachment 1, "Containment Purge Data."

b. OBTAIN Authorized Batch Release Order from Shift Manager.

c. CHECK Radiation Monitor RE-1113 operable per the following applicable procedure:

a Technical Specification Surveillance Procedure DWO-1, "Operator's DailyMleekly Items Modes 1, 2, 3, and 4" a Technical Specification Surveillance Procedure DWO-2, "Operator's DailyNVeekly Items Modes 5 and 6"

d. E RE-I 113 is not operable, THEN the WGDT may be released providing all of the following conditions are met:

1. At least two independent samples of the tank contents are-prior to release.

2. At least two independent verifications of the release rate calculations have been performed.

3. At least two qualified Auxiliary Operators independently verify the WGDT discharge line-up correct.

PALISADES NUCLEAR PLANT Proc No SOP-10A SYSTEM OPERATING PROCEDURE Revision 36 Page 13 of 32 TITLE: RADIOACTIVE WASTE SYSTEM GASEOUS -

e. ENSURE that at least one V-6A or V-6B Main Exhaust Fan is operating and record on HP Form 6.6-3.

f. IF the Main Exhaust Fans trip during WGDT release, THEN immediately STOP batch release. This may be done from the Control Room by lowering the trip setpoint on RIA-I 113 until CV-1123, WGDT Discharge, trips closed.

NOTE: Pressure settings for operation of HIC-1123 are as follows:

Position Full Closed 118 Open I Pressure I 0-3psig 1 4-9 psig I

Full Open I 10-15 psig

1. PLACE to CLOSED position HIC-1123, Waste Gas Discharge To Stack.

2. ENSURE CLOSED the air supply valves for the following WGDT Discharge Valves:

MV-CA363, INST AIR SUP CV-l119A, T-68A DISCHARGE MV-CA364, T-68B OUTLET CV-1120A AIS MV-CA365, DECAY TANK T-68C VLV CV-1121A AIR SUP MV-CA367, T-IOIA OUTLET CV-1160 AIS MV-CA368, T-IOIB OUTLET CV-1161A AIS MV-CA369, T-I 01C OUTLET CV-1162A AIS

3. ENSURE CLOSED the following:

MV-WG719, T-68NBIC Outlet Isolation MV-WG718, T-101AIBIC Outlet Isolation MV-WG71EA, T-IOIAIBIC Outlet Isolation

g. AIR PURGE RE-I 113 to reduce background count rate AND RECORD on HP Form 6.6-3.

h. SOURCE CHECK RE-I 113 prior to batch release AND RECORD cpm on HP Form 6.6-3.

I. RECORD RIA-I 113 background on Form HP 6.6-3.

j. DETERMINE Hi Alarm setpoint. (This is the sum of the observed background and the established release limit provided on Form HP 6.6-3.)

PALISADES NUCLEAR PLANT DBD-1.07 DESIGN BASIS DOCUMENT Revision 4 Page 67 of 184 TITLE: AUXILIARY BUILDING HVAC SYSTEMS

~ ~

3.2.1.10 Main Exhaust Fan System The air exhausted from the Radwaste Area HVAC System, the Fuel Handling Area HVAC System and the Containment Air Room Purge System all combine in the Main Exhaust Plenum. The plenum is located between the 590 and 607 elevations of the Auxiliary Building. Access to the plenum is via the 607 elevation in the Main Steam Line Penetration Room (2nd Level CCW Room). Two Main Exhaust Fans V-6A (EMB-1215) and V-6B (EMB-1111) draw a suction from the plenum and discharge to the Plant Exhaust Stack via dampers PO-I816 (V-6A) and PO-I818 (V-6B). The fans are currently only run one at a time and are used to keep the main exhaust plenum at a negative pressure compared to atmosphere. This ensures that the other HVAC system exhaust fans (V-8s, V-l4s, V-68s, V-70s and Containment) have a place to exhaust their air without reversing flow on one of the other fans. The negative pressure main exhaust plenum also provides the location for the Vent Gas Collection Header to exhaust (the Vent Gas Collection Header, though, is outside the scope of this DBD) with the V-6s providing the negative pressure motive force for gasses to be exhausted via the header. The stack provides a mixing area for gasses to be discharged from the plant (these gases and their release points are outside the scope of this DBD). The gasses that are released include the discharge line from the Waste Gas Decay Tanks, the exhaust from the Steam Jet Air Eje and the Flash Tank and Blowdown Tanks reliefhen-e from the stack are monitored by the Radioactive System (which is also outside the scope of this DBD). RGEM the gas in the stack, monitors it for radioactivity and then the stack via the air ejector exhaust line.

- u Safety Classification The Main Exhaust Fan System is not safety-related.

PALISADES NUCLEAR PLANT Proc No HP 6.6 HEALTH PHYSICS PROCEDURE Revision 17 Page 12 of 14 TITLE: EVALUATION AND RELEASE OF WASTE GAS DECAY TANK 5.10.2 Verify that the calculated release rate is greater than 150 CFM (as established in EA-CR-940839-OI), and authorize the release by signing Form HP 6.6-3. In the event that calculated release rate for any waste gas decay tank is less than 150 CFM, further holdup (decay) of the tank is required prior to release.

5.11 BATCH RELEASE 5.11.1 Deliver batch authorization to Shift Manager for release. The Shift Manager shall be responsible for completion of sections labeled Auxiliary Operator, Control Room Operator, Shift Manager, and Approval to Release on Form HP 6.6-3.

5.12 RETS EFFLUENT LOGBOOK 5.12.1 The RETS Analyst shall enter the following data in the RETS Effluent Logbook:

a. Release start date and time

b. Release stop date and time

c. CQ,/ECi = Total "jiCi/EC" (cc) + release time (sec)

d. Average annual release fraction for the period of release Fraction = CQi/ECi + 4.7E+11 cc/sec 6.0 ACCEPTANCE CRITERIA 6.1 Release shall be reviewed for limits being met and for completion of required data.

Work Order or Action Request shall be filed if appropriate.

6.2 Following isolation and/or prior to release, WGDT contents shall be sampled and analyzed for compliance with Palisades ODCM Appendix A, Sections III.A, 111.8, and 1II.E and Table B-1 Item A.

6.3 Waste gas holdup time meets requirements of Palisades ODCM Appendix A, Section 1II.E.

6.4 During release of gaseous wastes to the Plant vent stack, the following conditions shall be met:

6.4.1 At least one main exhaust fan (V-6A or V-6B) should be in operation.

6.4.2 If RIA-1 113 is not operational, ensure that the provisions of Palisades ODCM Appendix A, Table A-I are met.

TABLE 6.51-5 Proc No HP 6.51 Attachment

.-. - .. 5 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION Revision 10 AND EQUIPMENT REQUIREMENTS Page 2 of 7 INSTRUMENT OPERABLE APPLICABILITY ACTION

8. STACK GAS EFFLUENT SYSTEM

a. Noble Gas Activity Monitor (RIA-2326) (1) At All Times 7A

b. lodine/Particulate/Sampler/Monitor (RIA-2325) (11 At All Times 7

c. Sampler Flow Rate Monitor (FE-2346) (11 At All Times 6

d. Hi Range Noble Gas (RIA-2327) (1) Above 210°F 8 (Modes 1,2, 3,4)

9. STEAM GENERATOR BLOWDOWN VENT SYSTEM

a. Noble Gas Activity Monitor (RIA-2320) (11 Above210"F 5 (Modes 1,2, 3,4)

10. MAIN STEAM SAFETY AND DUMP VALVE DISCHARGE LINE

a. Gross Gamma Activity Monitor (11 Above 325°F 8 (RIA-2323 and RIA-2324) per Main (Modes 1,2, 3,4)

Steam Line

11. ENGINEERED SAFEGUARDS ROOM VENT SYSTEM

a. Noble Gas Activity Monitor (1) Above 210°F 12 (RIA-1810 and RIA-1811) per Room (Modes 1,2, 3,4)

12. WASTE GAS DECAY TANKS

a. T-68A, T-68B, T-68C, (1) Min 10 T-IOIA, T-lOIB, and T-1O1C 15-60 day decay

13. CONTAINMENT HIGH RANGE GAMMA MONITORS

a. RIA-2321, RIA-2322 (2) Above 210°F 11 (Modes 1,2,3,4)

LESSON NAME: Radiation Monitoring System RMS Rev. 1 IVA-81 RIA4712 Auto Actions

4) Automatic Actuations a) On a HIGH alarm, RIA-5712 automatically trips Auxiliary Building ventilation fan V-69.

b) The exhaust fan selected for standby, V-70A or V-708, may or may not trip.

c) For fans that trip, the associated damper closes.

0. RIA-2325, 2326, and 2327, Radioactive Gas Effluent Monitoring System (RGEMS) 1VA-82 RGEMS Purpose I

1) Purpose a) RIA-2325, RIA-2326, and RIA-2327 are the process radiation detectors in the Radioactive Gas Effluent Monitoring System (RGEMS).

(1) Analog channels RIA-2325 and RIA-2326 are normal range particulate/iodine and noble gas detectors, respectively.

(2) Digital channel RIA-2327 is a high-range noble gas monitor.

b) The RGEMS monitors actual stack effluent for radioactive contamination greater than expected.

(1) It serves as a backup to all effluent streams that are discharged through the stack, such as the ventilation exhaust monitors, the gaseous waste discharge monitor, and the condenser off-gas monitor IVA-83 RGEMS FIOW Path

2) Flow Path Page 65 of 109

LESSON NAME: Radiation Monitoring System RMS Rev. 1 Flgure 15. RGEMS Flow Path a) An isokinetic nozzle draws sample gas from the stack flow volume.

(1) Two 100% capacity diaphragm-operated variable-speed positive displacement pumps control the flow rate into the isokinetic nozzle.

(a) Only one pump is in service at a time.

(2) By controlling pump speed, a flow controller matches the flow speed of the gas sample entering the isokinetic nozzle with the flow speed of gas as it travels up the stack. This ensures that the RGEMS monitors a very representative sample of actual stack flow.

b) Normally, the sample gas flows into a lead-shielded assembly that contains a particulate and iodine filter and radiation detector RE-2325.

(1) The detector monitors the buildup of particulates and iodine from the gas sample that collect on the filter.

Page 66 of 109

LESSON NAME: Radiation Monitoring System RMS Rev. 1 c) Filtered sample gas leaves the particulateliodine assembly and then enters the suction of the positive displacement pumps, P-2301A and P-2301B.

d) From the pump discharge, the gas sample enters another lead-shielded assembly that contains radiation detector RE-2326.

(1) RE-2326 monitors the gas sample for radioactive noble gases.

e) After leaving the noble gas assembly, the sample gas is returned to the stack base enclosure, where it mixes with all the air entering the stack for discharge.

IVA-84 RGEMS Alert Flow Path f) On an ALERT condition (from RIA-2326 only),

the sample gas is automatically routed to a grab-sample bomb for 15 seconds.

(1) The diversion to the sample bomb is upstream of the RE-2325 iodinelparticulate filter chamber.

IVA-85 RGEMS High Flow Path g) On a HIGH radiation condition (from RIA-2326 only), the normal monitoring loop is bypassed.

(1) The sample gas is routed through two parallel paths. About 0.2 cfm is routed to Manual Accident Filter VF-2302.

(a) This flow path is known as the "minor loop." '

(2) The remainder of the gas sample (about 1.8 cfm) is routed through a filter assembly from which it enters a 20-litre lead-shielded chamber that is monitored by RE-2327.

(a) This flow path is known as the "major W t y of sample flow loop." goes through -loop.

(3) Both of these parallel flow paths recombine at the suction of the sample pumps.

(4) As with the normal flow path, the pump discharge flow is routed to the stack base enclosure.

Page 67 of 109

LESSON NAME: Radiation Monitoring System RMS Rev. 1 h) The sample lines in the RGEMS are heat-traced to prevent moisture condensation.

3) Alarms and Setpoints a) RIA-2325 ALERT alarm occurs at 1.4E5 CPM, and actuates Control Room annunciator EK-0219, "STACK EFF RAD C-169 ALERT" (Reflash).

b) RIA-2326 ALERT alarm occurs at 1.6E4 CPM, and actuates Control Room annunciator EK-0219, "STACK EFF RAD C-169 ALERT (Reflash).

c) RIA-2325 HIGH alarm occurs at 1.5E5 CPM, and actuates Control Room annunciator EK-0207, "STACK EFF RAD C-169 HIGH" (Reflash).

d) RIA-2326 HIGH alarm occurs at 1.3E6 CPM, and actuates Control Room annunciator EK-0207, "STACK EFF RAD C-169 HIGH" (Reflash).

e) A low signal output from RIA-2325, RIA-2326, or RIA-2327 actuates Control Room annunciator EK-0231, "STACK EFF/HT C-169/C-172 FAWTROUBLE" (Reflash)

This alarm is also actuated by the following conditions:

(1) Heat trace failure, sensed by low temperature (< 105 degrees F) at temperature switch TS-2302.

(2) Sampling system failure as sensed by programmable controller JIC-2301A or JIC-2301B.

(3) Heat tracing circuit low current at C-172 (4) Heat tracing failure at C-169 (5) A loss of power to RIA-2325, RIA-2326, or RIA-2327 (a) If the loss of power occurs on RIA-2326, then the channel ALERT and HIGH alarm automatic actions will also occur.

Page 68 of 109

LESSON NAME: Radiation Monitoring System RMS Rev. 1 IVA-86 RGEMS Auto Actions

4) Automatic Actuations a) On an ALERT alarm from RIA-2326 (only),

sample flow is routed through the sample bomb for 15 seconds. This is enough time to flush existing air from the sample bomb and store a representative sample of stack effluent for retrieval and manual analysis.

b) On a HIGH alarm from RIA-2326 (only), the normal sample flow path is bypassed and sample flow is diverted to the accident filters and high-range noble gas monitor (RE-2327).

p. RIA-2323 and RIA-2324, Main Steam Monitors

[ VA-87 RIA-2323 6 RIA-2324 PufpOSe

1) Purpose a) Analog channels RIA-2323 and RIA-2324 monitor the main steam flow leaving Steam Generators B and A, respectively.

b) The monitors are designed to detect radioactive contaminationin the main steam flow that would indicate a primary-to-secondary leak into a steam generator.

[VA-88 RIA-2323 6 RIA-2324 Flow Path

2) Flow Path a) Each detector resides in a lead-shielded collimator adjacent to its respective main steam pipe, near the main steam safety valves.

3) Alarms and Setpoints a) RIA-2323 WARN alarm occurs at 215 CPM and activates Control Room annunciator EK-0217, MAIN STEAM E-50B RIA-2323 ALERT.

b) RIA-2324 WARN alarm occurs at 215 CPM and activates Control Room annunciator EK-0218.

MAIN STEAM E-50A RIA-2324 ALERT.

c) RIA-2323 HIGH alarm occurs at 1260 CPM and activates Control Room annunciator EK-0205, MAIN STEAM E-50B RIA-2323 HIGH.

Page 69 of 109

Question #82 Given the following:

Power level is stable at 100%.

0 Pressurizer level is being controlled by Pressurizer Level Controller LIC-0101A.

0 The output of level controller LIC-0101A has just failed at 100% output signal.

No other failures occur.

Assuming no ODerators actions, what will charging flow be after the level controller output failure, and what is the expected plant response?

A. 0 gpm; and the Reactor will trip on Thermal MarginlLow Pressure.

B. 33 gpm; and Pressurizer level cycles in an approximately 11% band.

C. 44 gpm; and Pressurizer level stabilizes at 57%.

D. 133 gpm; and the Reactor will then trip on High Pressurizer Pressure.

NRC Answer Key: B Facility Comment:

This question has no correct answer. The correct answer was selected originally based on an understanding of the backup pressurizer level control system design, specifically, that it controls in an approximately 11 percent band. However, with the pressurizer level control malfunction standing, the pressurizer level will actually oscillate over a 2 percent range, the range between where the backup program takes control (-6%) and where it gets a signal to reset (- - 4%).

Facility Recommendation: Delete question from exam since no correct answer is provided.

Additional Facilitv References (AttachedL

1. System Operating Procedure SOP-2A, Attachment 2, Pressurizer Level Control Table

Proc No SOP-2A Attachment 2 Revision 57 PRESSURIZER LEVEL CONTROL TABLE Page Iof 1 PRESSURIZER LEVEL CONTROL PROGRAM Output Signal Rising_ _

Trip P4SB Open CV-2006 Trip P-SSC 1 *I 0 en CV-2004 C L Iili Ill

~~

. Output Signal Falling 8ICkUO PZR Level Control Program Deviation From Calculated Setpoint

- t 6% High I Low Alarm On

- t 5% P-55 B&C, -#2 8 3 Orifice, Backup Heaters On

- t4% (Reset)Hi I Low Alarm Off

- t 2% (m) Backup Heaters Off, Permit P-556 & C Start, Permit #2 & 3 orifice closure

- - 4 % (m) Hi I Low Alarm Off, Permit P-556 8 C Stop, Permit #2 & 3 orifice open

--6 % P-556 8 C, -#2 8 3 Orifice, High I Low Alarm On

Question #95 All plant equipment functioned as designed following a Large Break LOCA. When and why are the Charging Pump suctions aligned to the SIRWT in EOP-4.0, Loss of Coolant Accident Recovery?

A. Approximately 30 to 45 minutes; to reduce the effects of boric acid precipitation in the core

8. Approximately 30 to 45 minutes; to prevent Charging Pump cavitation due to loss of suction C. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; to ensure adequate SIRWT inventory is injected into the PCSlContainment D. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; to ensure adequate shutdown margin is established NRC Answer Key: A Facilitv Comment:

The concern for boric acid precipitation in the core is addressed by securing emergency boration. Refer to EOP-4.0 Basis, Step 19, and EOP Supplement 40 Basis.

Re-aligning charging pump suction from the concentrated boric acid storage tanks to either the volume control tank (VCT) or the safety injection refueling water tank (SIRWT) is done for the purpose of flushing the lines associated with boric acid injection. However, it does assist in reducing the effects of boric acid precipitation in the core. Therefore, answer A is correct.

During a LBLOCA, once shutdown margin (SDM) requirements are met, emergency boration would be secured. However, prior to this, charging pump suction would be re-aligned to a lower boron source (SIRWT) for the purpose of flushing the injection lines, as noted in the previous paragraph. Refer to EOP Supplement 40, Charging Pump Suction Alignment. At this point, boron is still being injected, only from a lower concentration source. Once the flush is complete, boration would be secured by shutting off charging pumps. This action is the one that addresses the concern for excess boron in the PCS (boron precipitation).

The stem is worded to ask when and why the suction source of the charging pumps would be re-aligned to the SIRWT. It is reasonable that answer D is also correct; Le., the action given in the stem (swapping suction to SIRWT) is done only after emergency boration is secured, and emergency boration is only secured once adequate SDM is established. This would occur within one hour of the condition stipulated in the stem of the question. Refer to EOP-4.0 Basis, Step 43.

Facility Recommendation: Accept both A and D as correct.

Additional Facilitv References (Attached):

1. EOP Supplement 40, Charging Pump Suction Alignment

2. EOP Supplement 40, Charging Pump Suction Alignment Basis

3. EOP-4.0, Loss of Coolant Accident Recovery Basis for Step 19

4. EOP4.0, Loss of Coolant Accident Recovery Basis for Step 43

Proc No EOP Supplement PALISADES NUCLEAR PLANT Supplement EMERGENCY OPERATING Revision MCLEIR PUHT PROCEDURE Page 1of2

~

TITLE: Charging Pump Suction Alignment 1.O ALIGN CHARGING PUMP SUCTION TO THF VCT

- 1. Open Charging Pumps Suction VCT Outlet Valve, MO-2087.

- 2. Stop the Boric Acid Pumps.

.. P-56A P-56B

- 3. Close the following valves:

. Boric Acid Pump Feed Valve, MO-2140

. Gravity Feed Valves 0 MO-2169 0 MO-2170

- 4. Ensure closed Charging Pumps Suction From SIRWT, MO-2160.

- 5. Operate each Charging Pump for at least five minutes.

- 6. WHEN each pump has been operated at least five minutes, THEN operate Charging Pumps as necessary to maintain safety functions.

2.0 ALIGN C W G PUMP SUCTION TO THE S I R W

- 1. Open Charging Pumps Suction From SIRWT Valve, MO-2160.

- 2. Stop the Boric Acid Pumps.

.. P-56A P-568

Proc No EOP Supplement PALISADES NUCLEAR PLANT Supplement 40 EMERGENCY OPERATING Revision 5 PROCEDURE I -

Paae 2 of 2 TITLE: Charging Pump Suction Alignment

- 3. Close the following valves:

Boric Acid Pump Feed Valve, MO-2140

. Gravity Feed Valves 0 MO-2169 0 MO-2170

- 4. Ensure closed Charging Pumps Suction VCT Outlet Valve, MO-2087.

- 5. Operate each Charging Pump for at least five minutes.

- 6. WHEN each pump has been operated at least five minutes, THEN operate Charging Pumps as necessary to maintain safety functions.

Completed By:

Datemime: /

Reviewed By: (SS)

PALISADES NUCLEAR PLANT tevision EMERGENCY OPERATING 68 of 77 PROCEDURE BASIS TITLE: EOP SUPPLEMENTS BASIS EOP Supplement 40 Technical Basis:

The intent of this supplement is to provide the steps necessary to align operating Charging Pumps from the Boric Acid supply to either the VCT or the SIRWT.

Emergency boration is performed during the initial stages of plant events to ensure an adequate Reactor shutdown margin exists. This supplement provides the steps for changing the Charging Pump suction to a lower boron concentration source to prevent excess boron in the PCS and possible boron precipitation. Each Charging Pump is operated at least 5 minutes to flush the high concentration of boric acid out of the associated piping.

This is a plant specific addition to CEN-152 guidance.

In numerous locations throughout CEN-152, the guidelines state to control Charging and Letdown. If Charging Pump suction was changed due to emergency boration or SlAS then the Charging Pumps will have suction from the Boric Acid Storage Tank.

When adequate boration has occurred the Charging Pumps must have the suction source swapped to a source with PCS boron concentration. This supplement controls the Charging and Letdown system by providing the appropriate suction to the Charging Pumps and a location for Letdown flow.

EMERGENCY OPERATING TITLE: LOSS OF COOLANT ACCIDENT RECOVERY BASIS STEP 19 NQE: IE emergency boration is in progress, ItlEhl cooldown may commence/continuewhile the required shutdown margin value is calculated.

0 19. VERIFY PCS boron concentration 19. IE PCS boron concentration is less greater than or equal to required than required boron concentration, boron concentration as verified by ItlEhl PERFORM BOTH of the sample or hand calculation. Refer following:

to EOP Supplement 35. 3 k m , d , m 04 ha,w - v bOc&J;,, a. ENSURE emergency boration is

a. IE Emergency boration is in Rpq~'~.noJ, in progress.

progress Bhlp PCS boron concentration b. WHFN required boron is greater than or equal to concentration is reached, required boron concentration, ItiEhlSECURE emergency ItlEhlSECURE emergency boration. Refer to EOP boration. Refer to EOP Supplement 40.

Supplement 40.

flease see pf6JIded CFN-152 I OCA Step: EOP s u 4 6 , jnd. bas'js .

PP None The intent of this step is to maintain the [reactor shutdown] throughout the cooldown.

To accomplish this, the operator verifies boron concentration is greater than or equal to the required boron concentration by sample results or hand calculation using EOP Supplement

35. The operator may have to emergency borate to obtain the required PCS boron concentration.

If a PT Curve maximum cooldown rate is maintained, Pressurizer level may not be able to be maintained constant during the initial stages of the cooldown. Therefore, Pressurizer level may lower. This outsurge from the Pressurizer will tend to dilute the PCS boron concentration. The possible dilution effects should be considered in determining cold shutdown boron concentration.

EMERGENCY OPERATING TITLE: LOSS OF COOLANT ACCIDENT RECOVERY BASIS Cooldown is allowed prior to establishing the required boron concentration as long as emergency boration is in progress. Emergency boration will maintain the required shutdown margin when the cooldown rate is within the allowed technical specification cooldown rate. If plant conditions do not dictate immediately cooling down, then the boron concentration should be established prior to cooling down to, ,T less than 525°F. This is a judgement call that must be made by the on-shift SROs based on present plant conditions.

and W m  :

This note provides information that the cooldown may commence or continue if emergency boration is in progress. Calculations (EA-MAB-97-003 Rev 0) show that if emergency boration is in progress at 30 gpm using a concentrated boric acid tank and the plant is cooled down within technical specification rates, the reactor shutdown margin will remain within the required technical specification value.

In the event that shutdown boron concentration is not met, cooldown is allowed as long as emergency boration is in progress. Analysis show that boration at 33 gpm indicated (30 gpm actual) will maintain the reactor shutdown by the required shutdown margin during the cooldown if the cooldown rate is within the allowed technical specification cooldown rate.

n for Devi-:

Engineering Analysis EA-MAE-97-003 Rev 0, "Maintaining Shutdown Margin While Emergency Borating And Simultaneously Cooling the PCS At The Technical Specification Limit" allows a cooldown with Technical Specification limits to take place as long as Emergency Boration is in progress. This provides additional options to the operators without jeopardizing shutdown margin.

)roc No PALISADES NUCLEAR PLANT levision EMERGENCY OPERATING 122 of 311 PROCEDURE BASIS TITLE: LOSS OF COOLANT ACCIDENT RECOVERY BASIS STEP 43

43. required shutdown boron concentration has been established (approximately 30 to 45 minutes using all charging pumps),

ItlEhlALIGN Charging Pump suction to SIRWT. Refer to EOP Supplement 40.

CFN-153 I O m  :

• 33. [IF the charging pumps are taking suction from a concentrated borated water source, THEN suction to the RWT or other suitable source within [l hour] after the start of the Loss of Coolant Accident].

Technical B&:

The intent of this step is to realign the Charging Pumps from the Concentrated Boric Acid Tanks (CBATs) to the Safety InjectionlRefueling Water Tank (SIRWT) to reduce the effects of boric acid precipitation in the core which may occur due to boil off during a large break LOCA.

For large breaks, the Reactor Vessel refills only to the elevation of the break. Borated water is injected into the Reactor Vessel via the Charging and Safety Injection Pumps while steam is boiled away. This may result in boric acid being concentrated in the Reactor Vessel. Switching the suction of the Charging Pump to the SIRWT helps limit the excessive buildup of boric acid in the Reactor Vessel while still allowing for sufficient long-term reactivity control.

Switching of Charging pumps suction to the SIRWT should be done when required cold shutdown boron concentration has been established (approximately 30 to 45 minutes using all charging pumps) {time to swap charging pumps suction}.

Associated Notes. C a w . and W a r n :

None

EMERGENCY OPERATING TITLE: LOSS OF COOLANT ACCIDENT RECOVERY BASIS from FPG:

In addition to the time constraints, added boric acid quantity constraints.

boron precipitation problem. The term required shutdown boron m required by technical specifications. The upper limit of the of one boric acid tank was chosen to provide adequate boron concentration for refueling concentranon. I n e baSisf01.sa S~L-

.. . nical specification for the quantity of boric acid in one concentrated boric tank with a level of 118 inches of 6 1/4 weight percent boric acid has sufficient boron to establish [adequate shutdown margin].

Question #96 Following a refueling outage, during core reloading in what manner is the core reloaded and why?

A. The core reloading is started at the center of the core and loaded towards the periphery to ensure both source range detectors are monitoring the core 6 . The core reloading is started near an operable source range detector and loaded to the center of the core so that core uncoupling does not occur C. The core reloading is started at the center of the core and loaded towards the periphery to ensure a potential critical configuration is not shielded from the source range detectors D. The core reloading is started near an operable source range detector and loaded to the center of the core so that the initial fuel assemblies are supported by the core barrel NRC Answer Key: B Facilitv Comment:

Answer D is also correct. When reloading the core, or any fuel bundle, procedures require that the bundle be supported on at least one side by either another fuel assembly, or by the core shroud. This is done by starting loading on the peripheral of the core, and working inward, as noted in the provided in Reference 1. Reference 2 describes that the core shroud is an integral part of the core barrel.

Facility Recommendation: Accept both B and D as correct.

Additional Facilitv References (Attached):

1, EM-04-29, Guidelines for Preparing Fuel Movement Plans, Step 6.1.19

2. Lesson Plan RXVI. "Reactor Vessel and Internals," pages 14 - 19

PALISADES NUCLEAR PLANT Proc No EM-04-29 ENGINEERING MANUAL PROCEDURE Revision 2 Page 4 of 12 TITLE: GUIDELINES FOR PREPARING FUEL MOVEMENT PLANS 6.0 PROCEDURE I USER ALERT REFERENCE USE PROCEDURE Refer to the procedure periodically to confirm that all procedure segments of an activity will I

be or are being performed. Where required, sign appropriate sign-off blanks to certify that all segments are complete.

6.1 GENERAL REQUIREMENTS NOTE: This procedure is intended to provide guidance for fuel movements. Reactor Engineering personnel may override guidance in this procedure if the proper justification exists.

6.1.1 Fuel movement steps shall be developed by a qualified person and independently reviewed by a qualified person.

6.1.2 Attachment 1 contains a fuel movement example sheet. The functional equivalent can be used for developing and performing fuel movements.

6.1.3 Avoid placing freshly discharged assemblies in locations within two cells of the Spent Fuel Pool walls to minimize gamma heating. (PCR023425 implementing OE13221, "Gamma Heating of Spent Fuel Pool Walls and Floor"- " ) I 6.1.4 Fuel assemblies should decay for one year before placing them into Region II, to reduce gamma irradiation of the boraflex absorbers in the racks.

6.1.5 An evaluation shall be performed before placing fuel assemblies into Region II.

This evaluation shall include all Technical Specifications requirements for placing I fuel into Region II. This evaluation is typically documented in an Engineering Analysis.

6.1.6 Refer to Technical Specifications DSGN 4.3 for enrichment limits for fuel storage in the New Fuel Storage racks. Technical Specifications essentially limit storage to every other cell of the outer rows 'X' and ' Z for new fuel storage. The center row 'Y' I

should not be used due to maximum planar average enrichment limits.

PALISADES NUCLEAR PLANT Proc No EM-04-29 ENGINEERING MANUAL PROCEDURE Revision 2 Page 5 of 12 TITLE: GUIDELINES FOR PREPARING FUEL MOVEMENT PLANS 6.1.7 Fuel assemblies shall not be moved from the Reactor Core to the Spent Fuel Pool sooner than that permitted by the current Spent Fuel Pool heat load analysis, as in EA-CCW-87-01, Revision 2, . . , "Spent Fuel Pool Heat Load and I Required SFPHX Component Cooling Water Flow During a DBA." Both minimum decay time and maximum assembly addition rate to the pool shall be observed.

6.1.8 The Spent Fuel Handling Machine (SFHM) mast shall be at an orientation of 30, 120, or 210 degrees when accessing the transfer boxes.

6.1.9 The SFHM camera cannot be turned toward the SFP wall when accessing the cells closest to the wall.

6.1.10 Reference the System Operating Procedure SOP-28, "Fuel Handling System," for unusable cells (ie, stuck bundles) and various other requirements such as use of the Region II racks in the SFP. SOP-28 should be reviewed before designing most fuel shuffle plans.

6.1.11 It is preferable to have the SFHM mast at 90 degrees when accessing the New Fuel Elevator (NFE); however, 0" also works well and 180" will work when the proper attention to the drop light is given. A mast orientation of 270" shall not be used when entering the elevator due to interference with the SFHM camera and the wall.

6.1.12 When making fuel movement plan amendments during the shuffle, pay particular attention to the orientation and location of bundles in the SFP that will go back into the core. This can present a potential error for misloading the core.

6.1.13 Ensure plenty of clearance between the refueling machine camera and the core shroud when planning the refueling machine orientation. A one row buffer is necessary, but two rows are preferred.

6.1.14 K-I core location requirements: ONLY NE and SW fuel assembly orientations should be used.

6.1.15 Ensure the mast will not contact the SFP walls with the aux hoist installed.

6.1 . I 6 All fuel assemblies to be placed in the North Tilt Pit shall have decayed for greater than one year.

6.1.17 While handling control rods in the core, the Refueling Machine Mast orientation must be 45", 135", 225", or 315". Control rods are handled at the same orientations used for fuel assemblies in all other locations.

PALISADES NUCLEAR PLANT Proc No EM-04-29 ENGINEERING MANUAL PROCEDURE Revision 2 Page 6 of 12 TITLE: GUIDELINES FOR PREPARING FUEL MOVEMENT PLANS 6.1.18 If possible, while placing freshly irradiated assemblies in Region I racks, scatter load the assemblies to reduce sharing common walls between the assemblies. This will help reduce the gamma dose rates on the rack walls, decreasing the required wall vent rate and reducing the risk for an assembly becoming stuck in a cell due to a blocked vent hole.

6.1 . I 9 Fuel assemblies in the core must be supported on at least one side by either another fuel assembly or the core shroud.

6.1.20 If practical and efficient, orient the refueling mast (reactor side) so that the camera is not placed over an empty cell. The operators have a better visual reference when the camera is oriented over an occupied core location.

6.2 CORE OFFLOAD 6.2.1 Temporary placement of any fuel bundle to any core periphery position (adjacent to core shroud) is allowed, provided the nearest fuel bundle is a minimum of two core locations away.

6.2.2 The core shall be off loaded in a manner such that core uncoupling does not occur.

This can be accomplished by working from the center of the core toward operable excore detectors. It is imperative that a potential critical configuration is not shielded from the excore detectors.

6.3 CORE RELOAD REQUIREMENTS 6.3.1 Core reload fuel bundle placement in the core shall be limited to the following:

a. Final core location

b. Symmetric Core location

c. Temporary placement of a fuel bundle to any core periphery position (adjacent to core shroud), provided nearest fuel bundle is a minimum of two core locations away.

d. Final core location of lesser burnt fuel 6.3.2 The core shall be loaded in a manner such that core uncoupling does not occur.

This can be accomplished by working from operable excore detectors toward the center of the core. It is imperative that a potential critical configuration is not shielded from the excore detectors.

LESSON NAME: Reactor Vessel and lnternals RXVl Rev. 1 IVA-land11 I

4. Core Support Assembly ajor support member of the reactor internals is the core support assembly. This assembled structure consists of the core support barrel, the core support plate and support columns, the core shrouds, the core support barrel to pressure vessel snubbers and the core support barrel to guide structure guide pins. ,

b. The core support assembly is supported at its upper flange from a machined ledge in the reactor vessel flange. The lower end is restrained in its lateral movement by six core support barrel to pressure vessel snubbers. Within the core support barrel are axial shroud plates and former plates that are attached to the core support barrel wall and the core support plate and form the enclosure periphery of the assembled core.

c. The core support plate is positioned within the barrel at the lower end and is supported both by a ledge in the core support barrel and by 52 columns. The core support plate provides support and orientation for the fuel bundles. Also within the core support barrel just below the nozzles are four guide pins, which align and prevent excessive motion of the lower end of the upper guide structure relative to the core support barrel during operation.

d. The core support plate, 1-1/2 inches thick, is a perforated member with flow distribution and pin locating holes for each fuel bundle. The plate is supported by a ledge and by columns. The ledge on the CSB supports the periphery of the plate, and the plate is pinned, bolted and lock welded to the ledge for maintaining accurate location of the plate.

e. A series of columns are placed between the plate and the beams across the bottom of the core support barrel. The columns provide stiffness and transmit the core load to the bottom of the core support barrel.

Page 14 of 38

LESSON NAME: Reactor Vessel and Internal8 Wl Rev. 1

5. (Core Support Barrel (CS-

a. Cylindrical in shape with an inside diameter of 150 inches and a wall thickness of 1 inch.

b. Suspended by 4-inch thick flange from the reactor vessel core support ledge.

c. The core support barrel carries the entire weight of the core and other internals.

d. A 1.5-inch thick core support plate containing flow distribution holes and fuel assembly alignment pinholes is supported by a ledge inside the core rt barrel and by 52 core support columns.

re shroud is attached to the core s u. ~. w r t

,( plate and limits core bypass flow. 1 IVA- 13-15

6. Core Support Barrel Snubbers

a. The core support barrel snubbers are attached to the reactor vessel internal surface near the bottom of the core support barrel 1 ) Six core support barrel snubbers provide a close fit between the core support barrel and the lower vessel wall. (Tongue-and-groove arrangement. Tongue is located on the vessel).

b. Snubbers prevent flow-induced vibration and seismically-causedmovement of the CSB

c. Cap screws, secured by locking pins. are used to attach two shims to each CSB stabilizing lug.

d. The CSB Snubbers are sometimes referenced as CSB Stabilizing Lugs.

Page 15 of 38

LESSON NAME: Reactor Vessel and lnternals RXVl Rev. 1

7. Core Support Lugs (Core Stops)

a. Nine core support lugs are attached to the vessel lower wall. The Core Support Lugs are designed tc catch the core barrel in the event that the upper core supports failed. The Core Support Lugs contain a small yield pad that is intended to deform and absorb the energy of the falling core.

b. The Core Support Lugs are sometimes referred to as the Core Stops.

[vi% 16

8. Core Shroud

a. The core shroud follows the perimeter of the core and limits the amounts of coolant bypass flow.

The shroud consists of rectangular plates 5/8 inch thick, 145 inches long and of varying widths. The bottom edges of these plates are fastened to the core support plate by use of anchor blocks.

b. The critical gap between the outside of the peripheral fuel bundles and the shroud plates is maintained by seven tiers of centering plates attached to the shroud plates and centered during initial assembly by adjusting bushings located in the core support barrel. The overall core shroud assembly, including the rectangular plates, the centering plates, and the anchor blocks, is a boltec and lock-welded assembly.

c. In locations where mechanical connections are used, bolts and pins are designed with respect to shear, binding and bearing stresses. The core shroud assembly is designed with some inherent flexibility to minimize internal stresses at fastener locations while maintaining necessary clearances.

Because pressure is equalized across inner and outer shroud faces at both the upper and lower ends of the shroud, differential pressure across tht shroud during transients will remain relatively low.

d. All bolts, block and pins used in assembly of the core shroud are lock-welded and pre-designed to be captured in the event of a fracture.

e. Holes in core support plate allow some bypass flow upward between core shroud and CSB -

minimize thermal stress, eliminate stagnant pockets of coolant.

Page 16 of 38

LESSON NAME:

~~

Reactor Vessel and lnternals RXVl Rev. 1 IVA- 17.18 1

9. Flow Skirt

a. The lnconel flow skirt is a perforated (2-1/2 inch diameter holes) right circular cylinder, reinforced at the top and bottom with stiffening rings. The flow skirt is used to reduce inequalities in core inlet flow distributions and to prevent formation of large vortices in the lower plenum.

b. The flow skirt is a cylinder broken into 900 2.5-inch holes that break the large annular wall jet into approximately 900 2.5-inch diameter jets directed toward the center of the reactor vessel.

c. The skirt provides a nearly equalized pressure distribution across the bottom of the core support barrel. The skirt is hung by welded attachments from the core stop lugs near the bottom of the pressure vessel and is not attached to the core support barrel.

Weight is around 27

10. Upper Guide Structure tons

a. This assembly consists of a flanged grid structure, 45 control rod shrouds, a fuel bundle alignment plate and a ring shim. The upper guide structure aligns and supports the upper end of the fuel bundles, maintains the control rod channel spacing, prevents fuel bundles from being lifted out of position during a severe accident condition and protects the control rods from the effect of coolant cross flow in the upper plenum. It also supports the incore instrumentation guide tubing.

The upper guide structure is handled as one unit during installation and removal.

b. The upper end of the assembly is a flanged grid structure consisting of a grid array of 18-inch-deep long beams in one direction with 9-inch-deep short beams at 90 degrees to the deeper beams. The grid is encircled by an 18-inch-deep cylinder with a 3-inch-deep flange welded to the cylinder.

Page 17 of 38

LESSON NAME: Reactor Vessel and lnternals RXVl Rev. 1

c. The periphery of the flange contains four accurately machined and located alignment keyways, equally spaced at 90-degree intervals, which engage the core barrel alignment keys. The reactor vessel closure head flange is slotted to engage the upper ends of the alignment keys in the core barrel. This system of keys and slots provides an accurate means of aligning the core with the closure head. The grid aligns and supports the upper end of the control rod shrouds

d. The fuel bundle alignment plate is designed to align the upper ends of the fuel bundles and to support and align the lower ends of the control rod shrouds. Precision machined and located pins attached to the fuel bundle alignment plate align the fuel bundles. The fuel bundle alignment plate also has four equally spaced slots on its outer edge which engage with stellite hard-faced pins protruding out from the core support barrel to prevent lateral motion of the upper guide structure assembly during operation. Since the weight of a fuel bundle under all normal operating conditions is greater than the flow lifting force, it is not necessary for the upper guide structure assembly to hold down the core. However, the assembly does capture the core and would limit upward movement in the event of an accident condition.

11. Control Rod Shrouds

a. The control rod shrouds are of cruciform configuration and extend from about 1 inch above the fuel bundles to about 2 inches above the top of the pressure vessel flange. They enclose the control rods in their fully withdrawn position above the core, thereby protecting them from adverse effects of flow forces.

b. The shrouds consist of 4 formed plates, 0.187 inch thick by approximately 138 inches long, which are welded to 4 end bars to form a cruciform-shaped structure. The shrouds are fitted with support pads at the upper end machined for a bolted and lock-welded attachment to the flanged grid structure.

Page 18 of 38

LESSON NAME: Reactor Vessel and lnternals RXVl Rev. 1

c. The lower ends of the shrouds are also fitted with support pads machined for a bolted and lock-welded attachment to the fuel bundle alignment plate. The cruciform design provides a stiff section, resulting in low stresses and deflections. In the area of maximum cross flow, the shroud is supported between the flanged grid structure and the fuel bundle alignment plate as a beam with fixed ends.

12. Surveillance Capsules

a. Surveillance capsules are installed inside the reactor vessel to measure the long term effects of temperature and radiation on reactor vessel materials.

1) RTNDTand PTS concerns (why we utilize surveillance capsules) a) The exposure of the reactor vessel walls to fast neutron flux during operation weakens the steel because the crystal structure is disturbed (neutron embrittlement). This weakening causes the reference temperature for nil ductility transition (RTNDT) to shift to a higher value. In general terms, the material becomes less ductile (more susceptible to brittle fracture).

b. Neutron Fluence Surveillance capsules are mounted in the reactor vessel at eight (8) locations. Six locations are on the inside wall of the vessel and two locations are on the outside wall of the core support barrel shell.

c. Surveillance capsules contain the following items:

1) Reactor vessel metal samples.

a) Base metal, weld metal and the heat affected zone metal (base metal affected by the heat of the welding process) samples are included in the capsules.

Page 19 of 38