ML021350211
| ML021350211 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/18/2002 |
| From: | Diane Jackson Public Service Enterprise Group |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| References | |
| -nr, 05000354/2002-301 05000354/2002-301 | |
| Download: ML021350211 (170) | |
Text
PSEG Nuclear LLC Hope Creek Initial License Operator Training NRC Written Examination Administered 3/18/2002 Post Exam Changes:
Exam RO/SSRO Facility Recommendations Record # Exam Question #
4 4 RO Only Delete question. No correct answer. Answer choice "C" does not limit flow to only "system" flow.
HC.OP-SO.BC-0002 step 5.2.31 and 5.2.33 allow throttling flow with F003A(B) which are NOT affected by the stated bus loss. Also, as stated in Caution 5.2.31 and 5.2.33, the F003A(B) may be used to throttle flow through the heat exchanger. System flow is normally established with the FO 15A(B). In the question, both FO 15 A and B valves lose power from the 10A404 bus loss.
13 11/12 Recommend accept 2 correct answers B and D. The Technical Specification Bases 3/4 3.9 supports answer B, which states the reason for the high RPV Level Main Turbine trip is to prevent Main Turbine damage. The bases is brief and does not contain detailed discussion of the high level trip. Lesson Plan 0301-000.OOH-000002-15 page 13 and 14 supports answer choice D which states HPCI, RCIC, and Reactor Feed Pumps are tripped to prevent RPV overfill and flooding the main steam lines, then states the significant safety concerns if overfill occurred including "Stressing of the reactor main steam line nozzles, steam line snubbers, pipe supports and hangers as a result of: - The weight of water in the main steam lines; and the dynamic transient loads caused by water flow in the main steam lines." This is further supported by NRC INFO NOTICE 88-77, which is also referenced in the lesson plan and addresses RPV overfill and flooding the main steam lines. This is further supported by the bases in Improved Standard Technical Specifications.
31 21/28 Recommend Answer Key change to D. Upon trip of the CRD Pump, HCU Accumulator charging water check valves (V1 15) begin leaking through the valve seats, causing the accumulator pistons to move. The N2 gas pressure lowers when the accumulator piston moves. (See P&ID M-47 Sheet 1) The surveillance requirements of HC.OP-IS.BF-0103 demonstrate the leakage rate is low enough to prevent accumulator trouble alarms for greater than 2 minutes. The accumulator alarms when nitrogen gas side pressure lowers to 940 psig. The original assumption was that there was no leakage past the check valves, however, as stated previously there is some leakage past the check valves that would affect gas pressure and cause it to lower.
Page 1 of 4
PSEG Nuclear LLC Hope Creek Initial License Operator Training NRC Written Examination Administered 3/18/2002 Exam RO/SRO Facility Recommendations Record # Exam Question #
46 34 RO Only Delete question. RO candidates were not provided the HC Event Classification Guide to correlate the ALERT level EOP 103/4 entry condition to a radiation level. ECG Section 6.1 provides that correlation.
Without the ECG, the question becomes a Level of Difficulty 5 memory question relying on memory of wording contained in Lesson Plan 0302-000.00H-000127-12. The text for Learning Objective 2 states:
"The entry condition for Radioactivity Release Control corresponds to an action level defined in the site Emergency Plan."
ECG Bases document states that the ECG Initiating Condition is entered when radioactive release rates reach levels corresponding to 200 times 10CFR20, Appendix B Limits. These levels are high enough that they will not occur during normal operation, but still low enough that the immediate health and safety of the general public is not threatened by the release."
48 36/43 Recommend Answer Key change to D. The Hydrogen alarms are set to alarm at 2.0% Hydrogen concentration on the H2/02 Analyzers. Conditions provided in the stem indicate the reactor core would be degraded and producing hydrogen. EOP 102 Step PC/H3 directs the Hydrogen Recombiners to be placed in service if H2 concentration reaches 0.5%. Placing the H2 Recombiners in service IAW step PC/H3 would be the required action. (Answer D)
Hydrogen alarms are clear indicating Hydrogen Concentration is less than 2.0%, therefore, EXIT EOP 102 and enter SAG is not required. (Answer B) 68 55 RO Only Delete question. Level of Difficulty 5. The question requires memorization of the prerequisite 2.6.2 of HC.OP-SO.SB-0001. The procedure should have been referenced in the question stem.
73 59/60 Delete question. No correct answer.
LPRM 32-33-C is assigned to LPRM Group A IAW HC.RE-ST.SE-0003 Attachment 1. All answer choices affect an APRM. The answer would have been correct if the LPRM chosen belonged to APRM C or D. Additionally, the candidates were not given a reference to determine which APRM the LPRM was assigned. The readings of the LPRM and APRM before the failure occurred would also be necessary to determine if the average went up or down after the failure.
Page 2 of 4
PSEG Nuclear LLC Hope Creek Initial License Operator Training NRC Written Examination Administered 3/18/2002 Exam RO/SRO Facility Recommendations Record # Exam Question #
106 87/83 Delete question. No correct answer. Immediate Operator Action of HC.OP-AB.ZZ-0129 3.1 states "If smoke OR toxic gases are detected in the control room air supply, isolate the Control Room ventilation and place CREF in the RECIRC MODE". Keyed answer D contains part of that answer. Pressing the Control room EMER FILTER UNIT A and B RECIRC MODE pushbuttons alone will not start CREF or place CREF in the RECIRC MODE. CREF must be running for the Recirc Dampers to open.
116 92 RO Only Delete question. No Correct answer. IAW NC.NA-AP.ZZ-0049, Definitions 7.2. "Formal declaration of Suspension of Core Alterations or Fuel Handling is performed by the Refueling SRO or a condition required by Technical Specifications." None of the answer choices contained conditions required by Technical Specifications that would require Suspension of Core Alterations.
Additionally, Lesson Plan 302-000-00H-000 113-10 Obj 66 states... "Determine the conditions under which handling of fuel must be suspended, IAW NC.NA-AP.ZZ-0049. (SRO ONLY)". Therefore the question is not appropriate for the RO candidates.
120 94/93 Delete question for candidates: Breslin, Hernandez, Klass, and Panagotopulos due to question contained a typographical error that allowed no correct answer choice. These 4 students had completed and turned in their exams prior to the discovery of the error. The error was in the stem, Entry 2, Neutron dose should have read 24 instead of 54 mrem. Calculating the answer based on the error resulted in a remaining dose of 1491 mrem. (Reference NC.NA-AP.ZZ-0024)
Previous history TEDE = DDE + CEDE; TEDE = 210 + 45; TEDE = 255 mrem Todays dose: Gamma dose + Neutron dose Entry 1 + Gamma dose + Neutron dose Entry 2 = DDE Entry 1 = (52 + 24) + Entry 2 = (124 + 54)
Todays dose = 76 + 178 = 254 mrem Remaining dose = Admin limit (2000 mrem TEDE) - Previous history (255 mrem) - todays dose (254 mrem)
Remaining dose = 1491 mrem The closest answers were A: 1488 mrem and B: 1521 mrem, both of which are incorrect for the question asked.
Once the error was identified, it was made known to the 4 remaining candidates; Hanna, Kopsick, McKeown, and Baker. The correction was then written on the board. The question will remain valid for these 4 candidates.
Page 3 of 4
PSEG Nuclear LLC Hope Creek Initial License Operator Training NRC Written Examination Administered 3/18/2002 Exam RO/SRO Facility Recommendations Record # Exam Question #
126 98/97 Recommend accept 2 correct answers A and C. SH.OP-AP.ZZ-0102 step 5.5.2 supports answer C.
However, there are numerous 100 series Abnormal procedures that are operational transients. Abnormal procedures HC.OP-AB.ZZ-0138 Main Turbine Trip and HC.OP-AB.ZZ-0 110 Loss of an RPS Channel are examples of operational transients. Answer A stated "100 series are operational transient procedures" which is technically correct.
There is currently a major effort underway to correct identified deficiencies to enhance these procedures and governing documents.
129 100/100 Delete question. No correct answer. Procedure HC.OP-IO.ZZ-0008 was revised to remove reference to using HCU Accumulator pressures as verification means. The procedure now uses SPDS, CRIDS or RMCS Activity Control Cards to verify rods full in if forced to leave the Control Room.
Page 4 of 4
SENT BY:TRAINING CENTER : 4 2 : 15:45 : PSE&O NUCLEAR- 6103375320;# 2 Hope Creek ILOT 2002-01 NRC Exam JPM RO-A.4 The NRC requested validation of the safety significance and critical nature of the step to circle LPCI flow rate in administrative JPM RO-A.4. The JPM required the RO candidates to perform the licensed operator review of an operational status board for a given set of plant parameters following a small break LOCA event. The control room integrated display (CRIDs) computer was not available requiring the candidates to use alternate (backup) hardwire indication to verify and validate the plant parameters provided on the status board. The 3PM, as written, comjbjned verification and validation of the plant status information and circling recorded LPCI flow rates into a single "critical step."
Upon post examination review, it was determined that for this JPM's initial conditions, circling the LPCI flow rates had little safety significance. From a technical perspective, the reasoning is as follows:
> Given the initial JPM conditions and plant status parameters recorded, it is physically impossible for LPCI to be injecting into the RPV at the flow rates indicated. Reactor pressure was at 325#, which is nearly shutoff head for the RHR pumps, and based on the pump curves (attached) the maximum injection flow would be about 500 gpm, not the 10000 gpm indicated for both LPCI pumps. In addition, several other parameters such as the reactor still being at 325# 20 minutes after a small break LOCA indicate that the leak is very small and it is clear that RPV level in being maintained in the normal band with feedwater flow (feedpumps or secondary condensate pumps). Thus, circling LPCI flow to indicate LPCI is operating in a mode other than injection, although required by procedure, is of low safety significance and its omission is of low consequence (i.e. it does not provide clarifying information).
- Additionally, if the candidate had circled the LPCI flow rates indicating that LPCI was operating in a mode other than RPV injection, the procedure does not require that the alternate LPCI mode be recorded on the status board so the receiver of the completed status board would still need to ask for clarifying information (the alternate LPCI modes are: suppression pool cooling, suppression chamber sprays, or drywell sprays). A procedure change request has been generated (attached) to enhance this procedure to require not only circling LPCI flows but to annotate the mode of operation as well.
APR-01-2002 15:39 6093393997 91% P. 02
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SENT BY:TRAIN[NG CENTER ; 4 2 ; 15:48 ;. PSE&G NUCLEAR--. 6103375320;# 6 B Pg I(i Notification Overview Run Date: 04/01/2002 Page: 1 of 1 Notification 20095417 Notification 20095417 Notification type N1 Description Enhancement to 11C ECG Art 8 Nuc. Maint. Request Reporter NUNFC X-1267 11:52:18 Notification date 04/01/2002 Start date 04/01/2002 End date Start time 11 :52: 18 End lime 00: 00:00 Priority Sig. Level X Main WorkCrr. O-EPO0 Funct. location ECG-ATT. 08 SECONDARY COMMUNICATOR LOG Equipment Assembly Order PM planner grp 099 Nuclear Z efault 04/01/2002 12:02:10 NICOLA CONICELLA (NUNFC)
Enhance Hope Creek ECG Attachment 8, secondary communicator log, page 6 of 8, Operational Status Board - Hope Creek, to require annotating the alternate mode that A or B RHR pumps are operating in if not injecting into the RPV in LPCI mode (i.e. suppression pool cooling, suppression chamber sprays, drywell sprays, or shutdown cooling). The status board currently requires that LPC1 flow is circled if not injecting but does not require that the specific mode be annotated. This enhancement was a result of post-NRC license exam review for Hope Creek ILOT class 2002-01. Assign to C. Banner of the emergency preparedness group.
End of repon APR-01-2002 15:42 6093393997 91% P. 06
Given the following:
- The reactor is in Operational Condition 4
"-"A"RHR Pump is in Shutdown Cooling at rated flow 10A404 4.16KV 1E Bus trips on bus differential overcurrent 1-Which one of the following describes the effect the bus loss will have on Shutdown Cooling?
Fa The Shutdown Cooling common suction- ine -isolates and CANN-OT be reset The AP228 Jockey pump trips causing Shutdown Cooling Loop "A" to lose keepfill IBoth "A" and "B" Shutdown Cooling Loops Iose abiIity to adjust flow
'~B" Reactor Recirc Pump discharge valve-automatically opens bypassing core flow ffj *i el Comprehension Hope Creek I1' a*41 03/12/2002 T Emregeny and Abnormal Plant Evolutions tiEE
[ l .295003G409 295003 Partial or Complete Loss of A.C. Power 4,
- 2. ]Emergency Procedures and Plan 2.4.9...... Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation 3.3 3.9
'strategies.
E i"A"RHR SDC Return valve F015A is powered from "D"Channel 1E 480VAC. Loss of D Bus fails this Ivalve as is. Adjusting flow via RHR HX outlet valve and /or bypass valve is not proceduralized. AP228 provides keepfill to HPCI only. B RRP disch valve is controlled by NON 1E power.
.HC.OP-SO.BC-0002 HC.OP-SO.SB-0001 qi ObjPctives 000028E008 (R) Given a system which physically connects to or is required to support the operation of the RHR System or components therein, explain the function of the supporting system, lAW the RHR System Lesson Plan.
Thursday, March 21, 2002 3:37:01PM Page 4 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q)
CAUTION 5.2.31 -6ý l,Po4, 18" Y,,. ar',-c-(,1( v,.-
S. ;* 17ý ..-
'A*-i - -. .
A. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is used to throttle RHR flow through the RHR Heat Exchanger, then BC-HV-FO48A(B) A(B) RHR HX SHELL SIDE BYP MOV must be fully open. [CD-503B]
B. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is CLOSED fully, then RHR INLET TO HX A(B) TE N004A(B) (CRIDS point A2380(A2382) or TR R605 point 1(2)), will not be accurate.
When BC-HV-FO03A(B) Is fully closed then utilize RHR DISCH FROM HX A(B) TE N027A(B) (CRIDS point A2381(A2383) or TR-R605 point 3(4)), to MONITOR system temperature.
C. BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV should be opened slowly to avoid a "cold slug" discharge Into the recirculation piping.
5.2.31 PERFORM the following as necessary to maintain the Shutdown Cooling return to RPV temperature relatively constant, as monitored on TR-R605 point 4 OR CRIDS A2383, while maintaining the required RHR Shutdown Cooling flow, simultaneously:
A. IF temperature is increasing, THEN, PERFORM the following:
- 1. Slowly THROTTLE OPEN on the BC-HV-FO03A(B)
- 2. IF the BC-HV-FO03A(B) is fully opened, fTHEN THROTTLE CLOSED BC-HV-FO48A(B)
A(B) RHR HX SHELL SIDE BYP MOV Continued Next Page Hope Creek Page 42 of 111 Rev. 5
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q) 5.2.31 (Continued)
B. IF temperature is decreasing, THEN, PERFORM the following:
- 2. IF the BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV is fully open, THEN, THROTTLE CLOSED on the BC-HV-FO03A(B)
C. PERFORM the following to maintain shutdown cooling flow:
THROTTLE BC-HV-FO 15B RHR LOOP B RETURN TO RECIRC as necessary to maintain required flow on FI-R603B or FR-R608(B) - CRIDS A3139.
D. PERFORM the following to maintain head spray flow:
THROTTLE BC-HV-F023B RHR LOOP B HEAD SPRAY OUTBD ISLN MOV as necessary to maintain 300 gpm on FI-R607. ICD-935E]
5.2.32 MAINTAIN the RPV Metal Temperatures and RHR SDC System water temperature difference < 240 0 F.
REFERENCE Crids point A2383 OR Recorder TR-R605 - point 4 for RHR SDC system, AND CRIDS points A3569 & A3578 for Vessel flange.
Hope Creek Page 43 of 111 Rev. 5 C;
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Ony HC.OP-SO.BC-0002(Q)
NOTE 5.2.33 Cooldown until the final desired reactor coolant temperature is reached (90F -110F is recommended, although other temperature(s) within TS limits may be used to support integrated plant operations).
CAUTION 5.2.33 A. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is used to throttle RHR flow through the RHR Heat Exchanger, then BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV must be fully open.
B. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is CLOSED fully, then RHR INLET TO HX A(B) TE N004A(B) (CRIDS point A2380(A2382) QR TR-R605 point 1(2)), will not be accurate when BC-HV-FO03A(B) is fully closed, then utilize RHR DISCH FROM HX A(B)
TE N027A(B)(CRIDS point A2381(A2383) OR TR-R605 point 3(4)), to MONITOR system temperature. [CD-503B]
C. BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV should be opened slowly to avoid a "cold slug" discharge into the recirculation piping.
5.2.33 PERFORM the following as necessary to initiate the cooldown AND Control the cooldown rate, while maintaining the required RHR Shutdown Cooling flow, simultaneously: [CD-133B]
A. PERFORM the following to increase the cooldown rate:
- 1. Slowly THROTTLE OPEN on the BC-HV-FO03A(B)
- 2. IF the BC-HV-FO03A(B) is fully open, THEN. THROTTLE CLOSED on the BC-HV-F048A(B)
A(B) RHR HX SHELL SIDE BYP MOV.
Continued Next Page Hope Creek Page 44 of 111 Rev. 5
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q) 5.2.33 (Continued)
B. PERFORM the following to decrease the cooldown rate:
- 2. IF the BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV is fully open, THEN, THROTTLE CLOSED on the BC-HV-FO03A(B)
C. PERFORM the following to maintain Shutdown Cooling flow:
THROTTLE the BC-HV-FO 15A(B) RHR LOOP A(B) RET TO RECIRC, as necessary, to maintain the required RHR Shutdown Cooling flow.
5.2.34 RECORD Reactor Vessel temperatures and pressures IAW Integrated Operating Procedure HC.OP-1O.ZZ-0004(Q);
Shutdown from Rated Power to Cold Shutdown AND HC.OP-DL.ZZ-0026(Q), Attachment 3s; Surveillance Log.
Hope Creek Page 45 of 111 Rev. 5 C-)
Given the following:
- The reactor is in Operational Condition 4
"-"A"RHR Pump is in Shutdown Cooling at rated flow
- 10A404 4.16KV 1E Bus trips on bus differential overcurrent Which one of the following describes the effect the bus loss will have on Shutdown Cooling?
FaI The Shutdown Cooling common suction line isolates and CANNOT be reset II The AP228 Jockey pump trips causing Shutdown Cooling Loop "A" to lose keepfill Both "A" and "B" Shutdown Cooling Loops lose ability to adjust flow "B" Reactor Recirc Pump discharge valve automatically opens bypassing core flow AvR= Cognitive Level Comprehension Facilty ýHope Creek [l"ate:l 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions L 2 SR Gru 1 295003G409 295003 Partial or Complete Loss of A.C. Power lRecord Number 41 2.4 Emergency Procedures and Plan 2.4.9 Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation 3.3 3.9 Istrategies.
"Exlntonof "A"RHR SDC Return valve F015A is powered from "D"Channel 1E 480VAC. Loss of D Bus fails this valve as is. Adjusting flow via RHR HX outlet valve and /or bypass valve is not proceduralized. AP228 provides keepfill to HPCI only. B RRP disch valve is controlled by NON 1E power.
Reference Title HC.OP-SO.BC-0002
ýHC.OP-SO.SM-'001 Learning Objectives 000028E008 (R) Given a system which physically connects to or is required to support the operation of the RHR System or components therein, explain the function of the supporting system, lAW the RHR System Lesson Plan.
Material Required for ExamiNetio.
[~esin Source] New [5uestion Modification Method:ý uesionS ource Comments:
Thursday, March 21, 2002 3:37:01 PM Page 4 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q)
CAUTION 5.2.31 -- ,416" tk S. i-f* lypa J
A. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is used to throttle RHR flow through the RHR Heat Exchanger, then BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV must be fully open. [CD-503B1 B. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV is CLOSED fully, then RHR INLET TO HX A(B) TE N004A(B) (CRIDS point A2380(A2382) or TR R605 point 1(2)), will not be accurate.
When BC-HV-FO03A(B) Is fully closed then utilize RHR DISCH FROM HX A(B) TE N027A(B) (CRIDS point A2381(A2383) or TR-R605 point 3(4)), to MONITOR system temperature.
C. BC-HV-F003A(B) RHR HX A(B) OUTLET VLV should be opened slowly to avoid a "cold slug" discharge Into the recirculation piping.
5.2.31 PERFORM the following as necessary to maintain the Shutdown Cooling return to RPV temperature relatively constant, as monitored on TR-R605 point 4 OR CRIDS A2383, while maintaining the required RHR Shutdown Cooling flow, simultaneously:
A. IF temperature is increasing, THENI PERFORM the following:
- 1. Slowly THROTTLE OPEN on the BC-HV-F003A(B)
- 2. IF the BC-HV-F003A(B) is fully opened, THEN THROTTLE CLOSED BC-HV-F048A(B)
A(B) RHR HX SHELL SIDE BYP MOV Continued Next Page Hope Creek Page 42 of 111 Rev. 5
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q) 5.2.31 (Continued)
B. IF temperature is decreasing, THEN, PERFORM the following:
- 2. IF the BC-HV-FO4SA(B) A(B) RHR lIX SHELL SIDE BYP MOV is fully open, THEN, THROTTLE CLOSED on the BC-HV-F003A(B)
C. PERFORM the following to maintain shutdown cooling flow:
THROTTLE BC-HV-F015B RHR LOOP B RETURN TO RECIRC as necessary to maintain required flow on FI-R603B or FR-R608(B) - CRIDS A3139.
D. PERFORM the following to maintain head spray flow:
THROTTLE BC-HV-F023B RHR LOOP B HEAD SPRAY OUTBD ISLN MOV as necessary to maintain 300 gpm on FI-R607. ICD-935E]
5.2.32 MAINTAIN the RPV Metal Temperatures and RHR SDC System water temperature difference < 240 0F.
REFERENCE Crids point A2383 OR Recorder TR-R605 - point 4 for RHR SDC system, AND CRIDS points A3569 & A3578 for Vessel flange.
Hope Creek Page 43 of 111 Rev. 5
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q)
NOTE 6.2.33 Cooldown until the final desired reactor coolant temperature is reached (90°F -1 10°F is recommended, although other temperature(s) within TS limits may be used to support integrated plant operations).
CAUTION 5.2.33 A. If BC-HV-F003A(B) RHR HX A(B) OUTLET VLV is used to throttle RHR flow through the RHR Heat Exchanger, then BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV must be fully open.
B. If BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV Is CLOSED fully, then RHR INLET TO HX A(B) TE N004A(B) (CRIDS point A2380(A2382) OR TR-R605 point 1(2)), will not be accurate when BC-HV-FO03A(B) is fully closed, then utilize RHR DISCH FROM HX A(B)
TE N027A(B) (CRIDS point A2381(A2383) OR TR-R605 point 3(4)), to MONITOR system temperature. [CD-503B]
C. BC-HV-FO03A(B) RHR HX A(B) OUTLET VLV should be opened slowly to avoid a "cold slug" discharge Into the recirculation piping.
5.2.33 PERFORM the following as necessary to initiate the cooldown AND Control the cooldown rate, while maintaining the required RHR Shutdown Cooling flow, simultaneously: [CD-133B]
A. PERFORM the following to increase the cooldown rate:
- 1. Slowly THROTTLE OPEN on the BC-HV-FO03A(B)
- 2. IF the BC-HV-F003A(B) is fully open, ThEN, THROTTLE CLOSED on the BC-HV-F048A(B)
A(B) RHR HX SHELL SIDE BYP MOV.
Continued Next Page Hope Creek Page 44 of Ill Rev. 5
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020321 PSEG Internal Use Only HC.OP-SO.BC-0002(Q) 5.2.33 (Continued)
B. PERFORM the following to decrease the cooldown rate:
I. THROTTLE OPEN on the BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV.
- 2. IF the BC-HV-F048A(B) A(B) RHR HX SHELL SIDE BYP MOV is fully open, THEN, THROTTLE CLOSED on the BC-HV-FO03A(B)
C. PERFORM the following to maintain Shutdown Cooling flow:
THROTTLE the BC-HV-FO 15A(B) RHR LOOP A(B) RET TO RECIRC, as necessary, to maintain the required RHR Shutdown Cooling flow.
5.2.34 RECORD Reactor Vessel temperatures and pressures IAW Integrated Operating Procedure HC.OP-IO.ZZ-0004(Q);
Shutdown from Rated Power to Cold Shutdown AND HC.OP-DL.ZZ-0026(Q), Attachment 3s; Surveillance Log.
Hope Creek Page 45 of 111 Rev. 5
,A malfunction of the Digital Feedwater Level Controller has resulted in an INCREASING reactor
.water level. The Reactor Feedwater Pumps are automatically tripped on a high reactor water level s.n...a.l.t.0._.p.
.signal ....
to....... ..v.e. i.................................................................................................................................................................................................................................................................................
prevent:
Ei feed pump damage due to increasing pump discharge flow rate and head.
I lmain turbine damage due to water impingement on turbine blades.
El 'reactor vessel damage due to completely filling and overpressurizing the vessel.
E imain steam line piping and hanger damage due to filling the main steam lines.
ie- o............................
.* p~ r e . .. ........................................... -----
bontv B 1[emory Lee Hope Creek 031/20 Tir LEmergency and Abnormal Plant Evolutions 2] [ 2p 295008K304
.295008 -High Reactor Water Level 1
iAK3. Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL:
AM&.64 ]Ractor feed pump trip: Plant-Specific 3L.5 E a io Feedpumps are tripped to prevent reactor overfill and damage to the main turbine.
ITC Bases 3/4.3.9 000002E008 (R) Given a list of reactor vessel pressure and/or level setpoints determine the automatic action that occurs lAW the Lesson
...... *p.......
a ................ ....... .........................................................
~~~~~~........
INPO Exam Bank } n i MEditorially Modified FaI.-S~ol SorceComent: IQID--#6*574 Dresden- 03/1 111i9906 Thursday, March 21, 2002 4:11:06 PM -.-...
P -1 .......1......
o.f. 9.
LESSON NAME: 0301 -000.OOH-000002-115 NUCLEAR BOILER INSTRUMENTATION - 01/11100 Figure 6 Obj. 6d
- 3) Steam carryunder - mixing small amounts of steam from the reactor vessel steam dome area and steam separators with water in the downcomer area.
a) Caused by steam entering the bulk water region below the steam dryer skirt due to low water level b) This results in lower downcomer water density which:
- Decreases indicated water level 0 Decreases recirculation pump NPSH 0 Decreases jet pump NPSH 0 Reduces core inlet subcooling c) Steam carryunder increases with as reactor water level decreases Obj. 6c
- 4) Moisture carryover - mixing of small amounts of moisture with the steam exiting the reactor vessel steam dome to the main steam lines a) Caused by reduced moisture separator/steam dryer efficiency due to high water level b) Excessive carryover will damage the main turbine and feedwater pump turbine blading.
c) Moisture carryover increases with increasing water level.
- c. LEVEL 8 (+54")
- 1) Above LEVEL 8, gross moisture carryover can occur, therefore protection of downstream components is necessary.
- 2) A LEVEL 8 trip initiates the following actions:
a) Main turbine trip S:\VDrive\0PS\HC\Master LPs\00000215.DOC Page 13 of 49 yL~
LESSON NAME: 0301 -000.001-1000002-15 NUCLEAR BOILER INSTRUMENTATION - 01/11100 b) Reactor Feed Pump turbine (RFPT) trip.
c) RCIC and HPCI turbine trips.
- 3) The main turbine is tripped to protect it from blade damage caused by water impingement. The HPCI, RCIC and RFP turbines are tripped to prevent overfilling the vessel and flooding the main steam lines.
NRC IN 88-77 Obj. Ila, b, c
- 4) The NRC regards a reactor vessel overfill event as a significant safety concern, identifying the following four safety issues:
a) Hydrodynamic effects of water or two-phase fluid being discharged through the SRVs. This process could damage the SRVs.
b) Stressing of the reactor vessel main steam line nozzles, steam line snubbers, pipe supports and hangers as a result of:
The thermal transient caused by colder water flowing into the hot main steam line and reactor vessel; The weight of water in the main steam lines; and The dynamic transient loads caused by water flow in the main steam lines.
c) Potential for MSIVs not to close if the main steam lines are filled with water d) Placing the plant in a condition that has not been analyzed in the Final Safety Analysis Report (FSAR)
- d. LEVEL 7(+39")
- 1) At LEVEL 7, moisture carryover in the main steam lines is expected to increase significantly.
S:\VDrive\OPS\HC\vaster LPs\00000215.DOC Page 14 of 49
.BASES 3/4.3.8 DELETE 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of a high reactor vessel water level (Level 8) to mitigate potential damage to the main turbine.
HOPE CREEK B 3/4 3-7 Amendment No. 8 1
Given the following:
- The plant is at 37% power Both CRD pumps are tripped on low suction pressure
- The Reactor Building Operator is swapping CRD suction filters
- CRD ACCUM TROUBLE Overhead Annuciator C6-D4 is clear (Assume NO other operator actions)
Which one of the following describes the effect on gas pressure in the HCU Accumulators 2
ýminutes following the pump trip?
SStays the same because reactor pressure holds the charging water check valve closed
.Stays the same because accumulator pressure holds the charging water check valve closed I1Lowers because the reactor scrams I!Lowers because the accumulator piston moves when charging water header pressure is lost
?b_] LB .i]
[e L mreeso j-pecek- 03/12/002
- [......... ..
Emergn-cy and Abn~ormal Plaant Eoluions ] *.2]1 [ __2,, 1295022K203 AK2. Knowledge of the interrelations between LOSS OF CRD PUMPS and the following:
AK2.03- Accumulator pressures. _ _ _ _ _ 3.4 [3.4 3
SofCharging water check valve 115 maintains water volume on a loss of charging pressure from the CRD pumps. N2 gas pressure will remain the same as long as the check valve holds. If the check valve does not hold, the piston will stroke and N2 pressure will drop causing low accumulator pressure alarm.
iHC.OP-IS.BF-0103 Purpose
~~~~~ -P ....
000006E017 ()Given the appropriate procedure or access to the procedure, summarize the accumulator trouble alarms and thei epit aasssociated with each CRD HCU and how these problems may impact CRDH System Operation, lAW the Lesson Pln.
I114APPY,ý'IARI&Neffft-New i-Friday, March 22, 2002 3:21:17 PM Page 33 3 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200O20322,, ~ iii . L~~._,Xv r1%-.%Jjr-1a.jDr -Ul U.3kz)
CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE - REFUEL INSERVICE TEST 1.0 PURPOSE The purpose of this procedure is to demonstrate during plant refuel, the operability of the Control Rod Drive Accumulator Charging Water Check Valve, 1-BF-V 115, as required by Technical Specification 4.0.5. This is performed by verifying each individual accumulator check valve maintains the associated accumulator pressure above the alarm setpoint for greater than or equal to 2 minutes, starting at normal operating pressure, with no Control Rod Drive Pump operating. [CD-270A]
2.0 PREREQUISITES I
2.1 Charging Water Check Valve Exercise Test 2.1.1 Permission to perform this test has been obtained from the OS/CRS as indicated by the completion of Attachment 1, Section 1.0.
2.1.2 All personnel involved in the performance of this procedure. should complete Attachment 1. Section 3.0, prior to performing any part of this procedure.
2.1.3 No other testing OR maintenance is in progress that would adversely effect the performance of this test.
NOTE 2.1.4 All Control Rod Drive Scram Accumulators need not be OPERABLE to perform this test provided INOPERABLE accumulators are tracked lAW HC.OP-AP.ZZ-0 108, Removal and Return of Equipment to Service, and this surveillance is listed in Part B of Attachment 1, Action Statement Log Sheet as required to restore the equipment to operability.
2.1.4 The Control Rod Drive system is in service AND all OPERABLE Hydraulic Control Units are charged to normal operating pressure IAW HC.OP-SO.BF-000I (Q), CRD Hydraulic System Operation.
2.1.5 All insertable control rods are inserted except for rods removed lAW Technical Specifications 3.9.10.1 and/or 3.9.10.2.
2.1.5 The plant is in Condition 4 or 5.
Hope Creek Page 2 of 20 Rev. I
4f c ,-*-r Which one of the following gaseous radioactive release limits corresponds to the EOP-104 entry
.condition?
Si500 mRem to the Thyroid CEDE 1 5000 mRem to the Thyroid CEDE H i2 times 10CFR 20 Appendix B limits IJ E1200 times 10CFR 20 Appendix B limits Ais ee~[ ]e9 ~ *Mmory.Hp re 03/1 2/2002
~j Emergency"'andAb ... .......
n.ormal
...... .. Pl'ant"
.ii
.. .. 1[...........
E'volutions e o*.... . U.....P*:
- i; .... *o P ..... e...
Ce.............
- '. i .... 95038A203 ...
EA2. Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:
EA_2.0-3 [Radiation levels 3.5([4. 3 JUSTIFICATION:
CORRECT - lAW ECG Section 6 and Lesson plan 0302-000.OOH-0001 27, the alert value is 200 times
.the 10CFR20 Appendix B value i ECG Section 6.0
!LP 0302-000.OOH-000127 000127E0(2 I Given a set of plant conditions, analyze and determine if entry conditions into HC.OP-EO.ZZ-0103/4 exists.
Qeston Mdfcton Metod:
6r" C
Friday, March 22, 2002 9:09:45 AM -Page P 48 of 139
L.P. NO.: 0302-000.00H-000127-12 0303 0304 0305 PREVIOUS L.P. NO.: 0302-000.00H-000127-11 0302-000.00H-000128-09 NUCLEAR TRAINING CENTER LESSON PLAN PROGRAM TITLE: HOPE CREEK LICENSED OPERATOR TRAINING COURSE SECTION/MODULE: OPERATING PROCEDURES TOPIC/SUB MODULE: EMERGENCY OPERATING PROCEDURES LESSON: HC.OP-EO.ZZ-0103/4 REACTOR BUILDING & RAD RELEASE CONTROL DURATION: 2 HOUR (INITIAL) 1 HOUR (REQUAL)
PREREQUISITES: HOPE CREEK SYSTEMS LP NO: 0302-000.00H-000121 JTA NO. OR QUALIFICATION STATEMENT NO.:
AUTHOR: F. W. Berg DATE: 06/09/99 I ... REVIEW/APPROVAL SIGNATURES I SUBMITTED BY: Pete Doran DATE: 06/22/99 BARGAINING UNIT REPRESENTATIVE: DATE:
Badge#
PRINCIPAL TRAINING SUPERVISOR: DATE:
Badge #
LINE SUPERVISOR: DATE:
COPY RECEIVED Badge #
WORD PROCESSING INITIALS:
LESSON NAME: 0302-000.01H-0001 27-12 HC.OP-EO.ZZ-01 0314 REACTOR BUILDING & RAD RELEASE CONTROL - 06/22199 INSTRUCTOR
REFERENCES:
- 1. BWROG Emergency Procedure Guidelines/Severe Accident Guidelines
- 2. BWROG Emergency Procedure Guidelines/Severe Accident Guidelines, Appendix A, B, C and D
- 3. HCGS Plant Specific Technical Guidelines
- 4. HCGS Technical Specifications
- 6. HCGS HC.OP-EO.ZZ-0103/4 Reactor Building & Rad Release Control
- 7. HCGS HC.OP-EO.ZZ-0103 Reactor Building & Rad Release Control Flowchart TRAINING MATERIAL REQUIRED:
Lesson Plan HC.OP-EO.ZZ-0103/4 Flowcharts (Current Revision)
HC.OP-EO.ZZ-0103/4 Bases document Transparencies (If Applicable)
Procedures referenced will be available in the classroom for student reference STUDENT HANDOUTS:
Lesson Plan 0302-000.00H-000127 HC.OP-EO.ZZ-0103/4 Flowcharts (Current Revision)
HC.OP-EO.ZZ-0103/4 Bases document Procedures referenced will be available in the classroom for student reference SPECIAL CLASSROOM REQUIREMENTS:
No special consideration for the classroom size or arrangement are required.
S:WDrive\OPS\HC\Master LPs\00012712.DOC 2 of 2
LESSON NAME: 0302-000.OOH-0001 27-12 HC.OP-EO.ZZ-010314 REACTOR BUILDING & RAD RELEASE CONTROL - 06122/99 LEARNING OBJECTIVES:
TERMINAL LEARNING OBJECTIVES:
Provided a scenario or plant/system status or previous plant condition, the trainee will be able to:
1.0 Perform actions required to implement HC.OP-EO.ZZ-0103/4, for Secondary Containment Control. (1-6) 2.0 Recognize the entry condition(s) of HC.OP-EO.ZZ-0103/4, Reactor Building & Rad Release Control and utilize the flowchart to control reactor building parameters. (1-6) 3.0 Derive an expected actuation or continuation of the event based on basic knowledge presented during previous training. The trainee's response shall be consistent with and contain the essential elements identified on an answer key. (1-6)
S:VDrive\OPS\HCwvaster LPs\00012712.DOC 3 of 3
LESSON NAME: 0302-000.OOH-0001 27-12 I-C.OP-EO.ZZ-01 0314 REACTOR BUILDING & RAD RELEASE CONTROL - 06/22199 ENABLING LEARNING OBJECTIVES:
Note: All ELOs are required to be covered during initial training. Those with an "R" prefix are the minimum required to be covered during requal training.
- 1. State the three purposes of the Reactor Building & Rad Release Control procedure, HC.OP-EO.ZZ-0103/4.
- 2. Given a set of plant conditions, analyze and determine if entry conditions into HC.OP EO.ZZ-01 03/4 exists.
R3. Define the term "Maximum Safe Operating Temperature".
R4. Define the term "Maximum Safe Operating Radiation Level".
R5. Define the term "Maximum Safe Floor Level".
R6. Given any step in the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by the step.
S:\VDrive\OPS\HCWlaster LPs\00012712.DOC 4 of 4
LESSON NAME: 0302-000.OH-0001 27-12 KC.OP-EO.ZZ-010314 REACTOR BUILDING & RAD RELEASE CONTROL - 06122/99
- b. The purpose of this step is to direct the operator to restore Reactor Building parameters to normal levels, which will assure the Reactor building integrity and to limit any potential releases to the environment.
- c. The operator remains at this step until the exit criteria of the RB-1 override statement are met.
IV. RAD RELEASE CONTROL A. The procedure is entered when radioactive release rates reach levels corresponding to 200 times 10CFR20, Appendix B Limits. These levels are high enough that they will not occur during normal operation, but still low enough that the immediate health and safety of the general public is not threatened by the release.
B. HC.OP- EO.ZZ-0103/4 isolates primary system discharges and controls RPV pressure through sequentially executed steps as required to minimize the offsite release of radioactivity. These steps provide the interface between individual events specifically addressed by the site Emergency Plan and the symptomatic control of RPV, primary containment, and reactor building parameters.
V. RAD RELEASE CONTROL PROCEDURE Obj. 2 A. Conditions for Entry The entry conditions to this procedure is:
GASEOUS RADIOACTIVE RELEASE ABOVE ALERT.
B. Procedural Steps
- 1. RR-1 Retention Override Steps IF while executing the following steps: THEN:
SAG entry is required EXIT this procedure AND ENTER SAG
- a. If primary containment flooding is required or hydrogen above 2%
is generated, all EOPs are exited and the SAGs are entered. The SAGs then remain in effect until an emergency no longer exists.
- b. Review HC.OP-EO.ZZ-01 03/4 bases for these steps.
Obj. 6
- 2. RR-2 RESTORE Rad Release Rate below ALERT level
- a. Entry from RR-1, or any of the following:
S:\VDrive\OPS\HC\Master LPs\00012712.DOC 18 of 18
LESSON NAME: 0302-000.001-0001 27-12 HC.OP-EO.ZZ-01 0314 REACTOR BUILDING & RAD RELEASE CONTROL - 06122/99
"* Yes response on step RR-6
"* No response on RR-10
"* Completion of actions from steps RR-9 and RR-1 1
- b. The entry condition for Radioactivity Release Control corresponds to an action level defined in the site Emergency Plan.
Table 3
- c. Refer to section 6.1 of HCGS ECG
- d. Review HC.OP-EO.ZZ-0103/4 bases for this step.
Obj. 6
- 3. RR-3 Can Release Rate be Maintained below ALERT Level
- a. Entry from step RR-2
- b. A yes response directs operator to exit procedure.
- c. A no response directs operator to additional actions in step RR-4
- d. Review HC.OP-EO.ZZ-0103/4 bases for this step.
Obj. 6
- 4. RR-4 Retention Override Steps IF while executing the following steps: THEN:
Turbine Bldg Ventilation System is RESTART Turbine Bldg Ventilation System shutdown
- a. Entered on a no response from RR-3.
- b. Review HC.OP-EO.ZZ-0103/4 bases for this step.
Obj. 6
- 5. RR-5 Except systems required to:
"* Assure adequate core cooling
"* Shutdown the reactor ISOLATE all primary systems discharging into areas outside Primary Containment and Rx Bldg
- a. Entry from RR-4
- b. Review HC.OP-EO.ZZ-01 03/4 bases for this step.
S:VDrive\OPS\HC\Master LPs\0001 2712.DOC 19 of 19
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1> -
n 111 tFI TAd
-o eli z2ozooz 3sfl MIl aOJi 03HOViiV 38 ismlV S39NVHD lOdS-3Hi-NO 3AIiOV liV
.Given the following:
- A large break LOCA has occurred inside the Drywell
- Multiple equipment failures occurred
- Drywell pressure is 15 psig
- Steam cooling was required until water level was restored above TAF with Fire Water
- The Containment H2/02 Analyzers were placed in service
- The High Hydrogen alarms are clear Which one of the following actions is required lAW EOP-1 02?
EIIIVent the Drywell because Hydrogen concentration is above 2%
K] Exit EOP-102 and enter SAG because Hydrogen concentration is above 2%
']Vent the Suppression Chamber because Hydrogen concentration is below 2%
K Place the Hydrogen Recombiners in service because Hydrogen concentration is below 2%
~s#~ eve~[B ] C' u ves Apiain ]V~ Hope creek-mDae 03112/2002.
i50 00 C-n-an11 eni ý yd rg n6nen I~gr ..................
-ion-7 1F.... E........
-- 48-r Emergency and Abnormal Plant Evolutions J e1500000K303 i 5000 1High Containment Hydrogen Concentration RcrNubr48:
IEK3. Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT
'HYDROGEN CONCENTRATIONS:
EK3.03 Operation of hydrogen and oxygen recombiners 3.0 3.51 Exp a io High H2 alarms come in at 2% Hydrogen. Since the H2 concentration is above 2%, EOP-102 step
!PC/H1 directs exit from EOP-102 and enter SAG rtEOP-102 step PC/Hi S.................................................... .................... ................. ............................
LerigObjcie 00126AE004 Recall the reasons why the following are used for determining the entry condition and Ior subsequent actions lAW the Primary containment Control - Drywell Lesson Plan.
- a. Drywell Pressure
- b. Average Drywell Temperature
- c. H2 and 02 concentrations in the drywell t.. .........
[material Required forl*Exa a IEOP. Flowcharts without entry conditions ew q.. .. -
Friday, March 22, 2002 9:22:22 AM Page 50 of 139
IJ IF while executing the following step: THEN:
Primary containment isolation PLACE H 2 /0 2 analyzers back in service, occurs if necessary.
H 2 /0 2 analyzer system is OR SAMPLE dIwl AND supp chamber for H2 becomes unavailable after warmup AND 02 H2 concentration exceeds 2 % EXIT this procedure and ENTER SAG PC/H-1 the MONITOR H2 AND 02 concentrations in Supp Chamber AND the Drwl PC/H-2 I
IF H2 concentration in the Drwl reaches .5%,
PLACE the H2 recombiners in service.
PC/H-3
Which one of the following describes when the Reactor Mode Switch Shutdown position scram may may be bypassed? _P~sse e~b ? ............................................. ...................................
E[ When moving the mode switch from REFUEL to SHUTDOWN I1When moving the mode switch from SHUTDOWN to REFUEL W When testing the "One Rod Out Interlock" a control rod blade is being uncoupled SiWhen FAn a x 'e [R ]7 Cgi~tiv MmrHope Creek 03/12/2002:
Ter: [Plant Systems 1]12120-09G1123 21200 Reactor Protection System [Rcr ubr68i 2.1 'Conduct of Operations 2.1.23 [Ability to perform specific system and integrated plant procedures during different modes of plant 3.9 [40]
loperation.
ExpTn o he Reactor -M-lode Switch Shutdown positi0on scram may be bypassed -to move tie-MS-from-refuel-to l nswer Shutdown when all control rods are fully inserted or the reactor is defueled.
Refrenc6Jitle 1HC.OP-SO.SB-0001 Prereq 2.6.2 _
9000221E.004 . (R) From memory, identify the parameters which initiate a Reactor Scram, list the scram initiation setpoints for each identified parameter, and determine when the parameter is bypassed, lAW the Lesson Plan.
Q New io M f tioM :
Srde* 'mentsr:
Friday, March 22, 2002 9:10:43 AM P Page 73 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE
{0020322 PSEG Internal Use Only Page 1 of 1 PSEG NUCLEAR L.L.C.
HOPE CREEK GENERATING STATION HC.OP-SO.SB-0001(Q) - Rey. 16 REACTOR PROTECTION SYSTEM OPERATION USE CATEGORY: II FIELD COPY EXISTS REVISION
SUMMARY
- 1. Order 80013429 (Pro-Trac #2306) This procedure has been revised to include the words "Field Copy Exists" at the top. This change has been incorporated to provide a second verification for the clerks during implementation of procedures.
- 2. IAW Generic procedure change identified under CR 990219121, (Closed To File), Limitations 3.2.2 was removed from this procedure. The direction contained in the Limitations regarding partial procedure performance and sequential performance of steps is adequately contained within higher tier administrative procedures including NC.NA-AP.ZZ-0001 (Q) and NC.NA-WG.ZZ-0001(Q).
- 3. All procedure references to WCCS have been deleted. Editorial change, conforms to existing procedures.
versus Process Computer may need OD-3 EDIT run for a rescan.
- 5. Order 80018874 Revised step 5.1.7 wording to restore to as written in revision 10 when it was distinct from similar direction in step 5.1.2.D. Revision I I attempted to make the steps similarly written for consistency but inadvertently introduced a typographical error in describing steps 5.1.7 check as an "unloaded' check, when in actuality it had been correctly identified as a "loaded' check in revision 10 and prior.
- 6. Based on writer review all occurrences of the action term "CHECK" have been replaced with "VERIFY".
This entire revision can be considered editorial in nature.
IMPLEMENTATION REQUIREMENTS None Effective date /1/1 APPROVED: , _ ___ ___ __f C)
Manager - Hope Creek Operations Date
"ALL AC-TIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-SO.SB-0001(Q)
REACTOR PROTECTION SYSTEM OPERATION TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPO SE ............................................................................................................. 2 2.0 PREREQ U ISITES .................................................................................................... 2 3.0 PRECAUTIONS AND LIMITATIONS .................................................................. 4 4.0 EQUIPMENT REQUIRED ...................................................................................... 7 5.0 PRO CEDU RE ........................................................................................................... 8 5.1 System Operability Observation .............................................................. 8 5.2 RPS Scram .............................................................................................. 16 5.3 Resetting RPS Trips ............................................................................... 18 5.4 Transfer of RPS Bus A(B) Power - from RPS MG Set A(B) to RPS Alternate Transformer A(B) ............................................................ 23 5.5 Transfer of RPS Bus A(B) Power - from RPS Alternate Transformer A(B) to RPS MG Set A(B) ................................................ 27 5.6 Bypassing Reactor Mode Switch Scram ............................................... 33 6.0 RECO RD S .............................................................................................................. 35 7.0 REFEREN CES ........................................................................................................ 35 ATTACHMENTS Attachment 1 Independent Verification - Reactor Protection System Operation ........... 37 Attachment 2 RPS TRIP LOGIC CHANNEL Al Simplified Drawing ....................... 45 Attachment 3 LOSS OF RPS BUS ............................................................................... 46 Attachment 4 RPS Power Distribution Simplified Drawing ........................................ 48 Attachment 5 Bypassing Reactor Mode Switch Scram ............................................... 49 Hope Creek Page 1 of 49 Rev. 16
"ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE
-00'2022 PSEG Internal Use Only HC.OP-SO.SB-0001(Q)
REACTOR PROTECTION SYSTEM OPERATION 1.0 PURPOSE This procedure outlines the steps necessary for the operation of the Reactor Protection System (RPS).
2.0 PREREQUISITES 2.1 System Operability Observation 2.1.1 All personnel performing any steps in this procedure should complete Attachment 1, Section 2.0, prior to performing any part of this procedure. __
2.1.2 Reactor Protection System Electrical Lineup is aligned for the applicable plant condition.
2.2 RPS Scram Automatic OR Manual Scram initiated.
2.3 Resetting RPS Trips 2.3.1 All personnel performing any steps in this procedure should complete Attachment 1, Section 2.0, prior to performing any part of this procedure.
2.3.2 Automatic OR Manual Half OR Full Scram initiated and the initiating trip signal(s) have cleared.
2.3.3 Reset ARJ/RRCS prior to resetting the Scram, IF it had initiated.
2.4 Transfer of RPS Bus A(B) Power - from RPS MG Set A(B) to RPS Alternate Transformer A(B) 2.4.1 All personnel performing any steps in this procedure should complete Attachment 1, Section 2.0, prior to performing any part of this procedure.
2.4.2 Reactor Protection System Electrical Lineup for the Alternate Feed is aligned for the applicable plant condition.
Hope Creek Page 2 of 49 Rev. 16
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-SO.SB-OOO1(Q) 2.4.3 The opposite RPS Bus B(A) is NOT being supplied by its RPS Alternate Feed.
2.4.4 IF the Main Turbine is shutdown (Turbine Stop and/or Control Valves closed),
THEN the EOC-RPT bypass switches are in the BYPASS position.
2.5 Transfer of RPS Bus A(B) Power - from RPS Alternate Transformer AM) to RPS MG Set AM) 2.5.1 All personnel performing any steps in this procedure should complete Attachment 1, Section 2.0, prior to performing any part of this procedure.
2.5.2 Reactor Protection System Electrical Lineup for the RPS MG Set A(B),
to be placed in service, is aligned for the applicable plant condition.
2.5.3 IF the Main Turbine is shutdown (Turbine Stop and/or Control Valves closed),
THEN the EOC-RPT bypass switches are in the BYPASS position.
2.6 Bypassing Reactor Mode Switch Scram 2.6.1 Permission to perform this test has been obtained from the OS/CRS.
2.6.2 All Control Rods are inserted OR Reactor is defueled.
2.6.3 No Core Alterations are in progress.
2.6.4 Banana Jacks have been installed on the terminals to be jumpered in order to facilitate the required jumpering.
Hope Creek Page 3 of 49 Rev. 16
PSEG Internal Use Only HC.OP-SO.SB-0001(Q)
NOTE 5.6 This section temporarily bypasses the Reactor Mode Switch in SHUTDOWN scram signal, until the signal is automatically bypassed after the 10 second time delay.
This is not a normal evolution and should only be used to bypass this scram signal when placing the Reactor Mode Switch from REFUEL to SHUTDOWN. [PR 981028191]
The Mode Switch shall be considered inoperable while the jumpers are installed.
The Mode Switch should be considered operable when removal of the jumpers has been independently verified and documented on Attachment 5.
5.6 Bypassing Reactor Mode Switch Scram 5.6.1 ENSURE all prerequisites of Section 2.6 are satisfied AND INITIAL Attachment 5 to document completion and verification.
5.6.2 DIRECT I&C to perform the following:
A. CONNECT a jumper between terminal Z 9 AND terminal ZZ 23 in BAY A of 1O-C609 AND I&C and Verifier INITIAL Attachment 5.
B. CONNECT a jumper between terminal A 9 AND terminal BB 23 in BAY F of 10-C609 AND I&C and Verifier INITIAL Attachment 5.
C. CONNECT a jumper between terminal Z 9 AND terminal ZZ 23 in BAY A of 1O-C611 AND I&C and verifier INITIAL Attachment 5.
D. CONNECT a jumper between terminal A 9 AND terminal BB 23 in BAY F of 1O-C611 AND I&C and verifier INITIAL Attachment 5.
Hope Creek Page 34 of 49 Rev. 16
TRAINING MATERIAL REQUIRED:
Lesson Plan Transparencies Technical Specifications STUDENT HANDOUTS:
Reactor Protection System Lesson Plan 0301-000.00H-000022 Instructor References (as required)
SPECIAL CLASSROOM REQUIREMENTS:
No special considerations for classroom size or arrangements are necessary Page 5 of 4343 C:\WlNDOWS\Profiles\cesktop\CALGCVXH.docC:\Wd'NDOWS\Profilesocmd\Desktop\CALGCVXH.docS :Wr,.,*.OpSý.Q.artor 11- -n-200=
01 9.d-"'cA=: 12-PLars-on." plaR..doc
LEARNING OBJECTIVES:
Provided a scenario of plant/system status or previous plant conditions associated with the operation of the Reactor Protection system, in accordance with the applicable trainee handout and system operating procedures, the trainee will:
1.0 State the purpose/design criteria of the system. (1,2,3,4,8,13,14) 2.0 Identify system configuration including major system components, flowpaths, and protective devices. (2,3,6,8,14) 3.0 Identify electrical power sources (4.16 KV, 480 VAC, 250 VDC, 1E, etc.), actuation signals, interlocks and alarms associated with system operation. (2,4,6,7) 4.0 Describe Technical Specifications requirements associated with the system. (13,18) 5.0 Critique plant problems and industry events associated with the system. (12) 6.0 Use plant procedures associated with system operation. (6,10,11,16) 7.0 Evaluate system interrelationships. (2,4,6,11) 8.0 Describe system configuration following emergency system actuation. (4,5,9,11,14)
ENABLING LEARNING OBJECTIVES:
NOTE: All ELOs are required to be covered during initial training. Those with an "R" prefix are the minimum required to be covered during continuing/requal training.
- 1. From memory, state the purpose of the Reactor Protection System (RPS), lAW the Lesson Plan.
- 2. Given the Hope Creek approved Electrical Load List and applicable electrical drawings:
- a. Identify the normal and alternate sources of power to RPS Bus A and RPS Bus B, lAW the Lesson Plan.
Page 6 of 4343 C:\WINDOWS\Profiles\\md CDesktopNCALGCVXH.docC:\WINDOWS\Profiles "md\Desktop\CALGCVXH.doCS:"-v1 r......o'QP
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R4. From memory, identify the parameters which initiate a Reactor Scram, list the scram initiation setpoints for each identified parameter, and determine when the parameter is bypassed, lAW the Lesson Plan.
R5. From memory, define the following terms:
- a. Scram, JAW the Lesson Plan.
- b. Half-scram, lAW Lesson Plan.
R6. Given the appropriate system operating procedure explain the effects on the reactor protection system when the power source is transferred from normal to alternate, and vice versa, lAW the RPS System Operating Procedure.
- 7. From memory, state the purpose of the time delay after a scram, lAW the Lesson Plan.
- 8. Given labeled diagrams/drawings of the RPS trip logics, explain the purpose of the RPS shorting links, lAW the Lesson Plan.
R9. Given plant conditions, evaluate the response of RPS to an electrical failure, lAW the Lesson Plan.
- 10. From memory, explain how a scram can be manually initiated, lAW the RPS System Operating Procedure.
- 11. From memory, state the immediate operator action(s) for a Loss of RPS channel lAW HC.OP-AB.ZZ-0110.
R12. Given plant problems/industry events associated with the Reactor Protection System.
- a. Discuss the root cause of the plant problem/industry events, lAW the Lesson Plan.
- b. Discuss the HCGS design and/or procedural guidelines that mitigate/reduce the likelihood of the problem/industry event, lAW the Lesson Plan.
- c. Discuss the "Lessons Learned" from the plant problem/industry event, lAW the Lesson Plan.
Page 7 of 4343 c:\wINDOwS\Profies~jmd\Desktop\cALGCVxH~docc:\wINDOwS\Profies~md\DesktoP\cALGcvxH~docc:\wINDOws\Profiesljmd\
Desktp\CALGvxHdocC:\WINDO1S\Prolesmd\DesktoP\CALGCVXH dOGS-ý.ýlrv ` tr . _9.doc_.A.R:PS I scriop plapAor
Table 1 REACTOR SCRAMS Parameter Scram Setpoints Bypassed
- 1. IRM Neutron Flux Upscale 120/125 of full scale Rx. Mode Sw in RUN
- 2. APRM Neutron Flux Upscale 15% of Rated Thermal Rx. Mode Sw in RUN
- 3. Flow Biased Simulated Th. Pwr. .66(W-AW)+51% Clamped-@
Upscale 113.5%
- 4. Fixed Neutron Flux Upscale 118% Rx Mode Sw Out of RUN
- 5. Reactor Vessel Steam Dome 1037 psig Pressure - High
- 6. Reactor Vessel Water Lvl - Low Level < 12.5" (Level 3) 3
- 7. Main Steam Line Isolation Valve < 92% open Rx Mode Sw Out of RUN
- 8. Primary Containment Pressure High 1.68 psig
- 9. Scram Discharge Volume water level Elev. 110' 10.5" Rx Mode Sw in S/D or REFUEL &
- High Disch. Vol. Scram Bypassed
- 10. Turbine Stop Valve - Closure < 95% open <30% Rx Power (135.7 psig Turb 1 st Stage Pressure)
- 11. Turbine Control Valve Fast Closure 530 psig (ETS Oil Pressure) <30% Rx Power Trip Oil Pressure Low (135.7 psig Turb 1 St Stage Pressure)
- 12. Rx. Mode Switch in Shutdown N/A 6 (+1-4) seconds after the Mode Switch is in SHUTDOWN
- 13. Manual Scram N/A 5
- 14. SRM High Count Rate 2x10 cps Shorting Links Installed
- 15. IRM Inoperative
- High Voltage P.S. Low, Rx Mode Switch in RUN
- Function Switch Not In Operate,
- Module Unplugged
- 16. APRM Inoperative & Function Switch Not in Operate
- <14 LPRM Inputs
- Module unplugged Page 44 of 4,43 c:\wsNDOwS\ProesCmd\Desktop\cALGcvxHCdocc:\w'NDPSwSrrocles1md\Desktop\cALGcvxHcdocc:\wINDOws\Proesjmd\
Desktop\CALG CVXH.dorS............ .............. .. .................... . 1 ..... .......
Given the following:
- Local Power Range Monitor (LPRM) detector 32-33-C has just failed downscale
- Subsequently, Control Rod 30-31 is selected Which one of the following describes the effect of the failure on the associated APRM and RBM channels?
The LPRM input:
be automatically bypassed and removed from the APRM only. SwillThe APRM and RBM readings will be lower than actual power.
Swill be automatically bypassed and removed from both the APRM and RBM. The APRM and RBM readings will remain the same.
Hwill be automatically bypassed and removed from the APRM only. The APRM reading will remain the same and the RBM reading will be lower than actual power.
EIwill be automatically bypassed and removed from the RBM only. The APRM and the RBM
,readings will be lower than actual power.
Ahsniiv dLev~ [ pel --
Comprehension Hope Creek 03/1212002.
Tier: IPlant Systems jR Gop ~ ~ 215005K305 215005 1Averagle Power Rangle Monitor/Local Power Range Monitor System 7 3 jKnowledge of the effect that a loss or malfunction of the APRM/LPRM will have on following:
K3.06- ~Ractoir poer indication 3.8 j3.8j E* oThe
] LPRM must be manually bypassed to remove from the APRM averaging circuit. The LPRM is f automatically bypassed in the RBM Count Circuit ifthe detector is reading <4%. Since the LPRM is still feeding the APRM avg, the indicated avg will be lower. Since the control rod is selected after the LPRM
,fails downscale so the gain change circuit will null to the now lower APRM reference signal.
HC0.OP-SO.S F-0002 000017E008 Given the applicable drawing, determine how the Rod Block Monitor (RBM) System interrelates with the following systems:
- a. Local Power Range Monitoring (LPRM) System
- b. Average Power Range Monitoring (APRM) System
- c. Recirculation Flow Units
- d. 120 VAC Instrument Power System
- e. 120 VAC Un-interruptible Power Supply System
- f. Reactor Manual Control System (RMCS) lAW the Rod Block Monitor (RBM) System Lesson Plan INPO Exam Bank Questo Modific tMSignithcantly Modified QID# 12556 Limerick 11/10/1995 Friday, March 22, 2002 11:34:45 AM Page 78 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 HC.RE-ST.SE-0003(Q)
ATTACHMENT 1 Calculation/Data Sheets Page 1 of 8 APRM CHANNEL A LPRM EV Io Calibration Ic Location GAF* (V) (uA) Verified By (uA)32-49B 16-33B 48-33B 32-17B 16-49D 48-49D 32-33D 16-17D 48-17D 24-57A 08-41A 40-41A 24-25A 56-25A 40-09A
- _ _ _1_....
_--1 _ _ - _ Wm t-__;-__,___ ire,40-57C 24-41C 56-41 C 08-25C 40-25C 24-09C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 64 of 79 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE
.20020522 HC.RE-ST.SE-0003(Q)
ATTACHMENT I Calculation/Data Sheets Page 2 of 8 APRM CHANNEL B LPRM GAF* EV lI Calibration I, Location (V) (uA) Verified By (uA)08-49B 40-49B 24-33B 56-33B 08-17B 40-17B 24-49D 08-33D 40-33D 24-17D 56-17D 32-57A 16-41A 48-41A 32-25A 16-09A 48-09A 16-57C 32-41C 16-25C 48-25C 32-09C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 65 of 79 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE
-20020322 HC.RE-ST.SE-0003(Q)
ATTACHMENT I CalculationlData Sheets Page 3 of 8 APRM CHANNEL C LPRM GAF* EV Io Calibration Ic Location (V) (uA) Verified By (uA)24-57B 08-41B 40-41B 24-25B 56-25B 40-09B 40-57D 24-41 D 56-41D 08-25D 40-25D 24-09D 16-49A 48-49A 32-33A 16-17A 48-17A 32-49C 16-33C 48-33C 32-17C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 66 of 79 Rev. 11
/
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2002C03.22 HC.RE-ST.SE-0003(Q)
ATTACHMENT I Calculation/Data Sheets Page 4 of 8 APRM CHANNEL D LPRM GAF* EV 1o Calibration 1c Location (V) (uA) Verified By (uA)32-57B 16-41 B 48-41 B 32-25B 16-09B 48-09B 16-57D 32-41 D 16-25D 48-25D 32-09D 24-49AO 08-33A 40-33A 24-17A 56-17A 08-49C 40-49C 24-33C 56-33C 08-17C 40-17C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 67 of 79 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020"322 HC.RE-ST.SE-0003(Q)
ATTACHMENT 1 CalculationlData Sheets Page 5 of 8 APRM CHANNEL E LPRM GAF* EV 1. Calibration I, Location (V) (uA) Verified By (uA)16-49B 48-49B 32-33B 16-17B 48-17B 32-49D 16-33D 48-33D 32-17D 40-57A 24-41A 56-41A 08-25A 40-25A 24-09A 24-57C 08-41C 40-41C 24-25C 56-25C 40-09C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 68 of 79 Rev. 11
,~.
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE
-2002C7322 HC.RE-ST.SE-0003(Q)
ATTACHMENT 1 Calculation/Data Sheets Page 6 of 8 APRM CHANNEL F LPRM GAF* EV 10 Calibration Ic Location (V) (uA) Verified By (uA)24-49B 08-33B 40-33B 24-17B 56-17B 08-49D 40-49D 24-33D 56-33D 08-17D 40-17D 16-57A 32-41A 16-25A 48-25A 32-09A 32-57C 16-41C 48-41 C 32-25C 16-09C 48-09C Pre-calibration APRM power level: V Post-calibration APRM power level: V
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 69 of 79 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200203"22 HC.RE-ST.SE-0003(Q)
ATTACHMENT 1 Calculation/Data Sheets Page 7 of 8 LPRM GROUP A LPRM GAF* EV I Calibration 1c Location (V) (uA) Verified By (uA)40-57B 24-41B 56-41B 08-25B 40-25B 24-09B 24-57D 08-41D 40-41 D 24-25D 56-25D 40-09D 32-49A 16-33A 48-33A 32-17A 16-49C 48-49C 32-33C .16-17C 48-17C
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 70 of 79 Rev. 11
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 HC.RE-ST.SE-0003(Q)
ATTACHMENT 1 CalculationlData Sheets Page 8 of 8 LPRM GROUP B LPRM GAF* EV 1 Calibration Ic Location (V) (uA) Verified By (uA)16-57B 32-41 B 16-25B 48-25B 32-09B 32-57D 16-41D 48-41D 32-25D 16-09D 48-09D 08-49A 40-49A 24-33A 56-33A 08-17A 40-17A 24-49C 08-33C 40-33C 24-17C 56-17C
- Calibrated LPRM Reading instead of the GAF Hope Creek Page 71 of 79 Rev. 11
Given the following:
- The plant is operating at 100% power
"- "A" Control Room HVAC train and Chilled Water system is running
- A light haze with an acrid odor is noticed in the Main Control Room
- No alarms are received that could explain the origin of the haze and odor
-HC.OP-AB.ZZ-01 29, High Radiation, Smoke or Toxic Gases in the Control Room Air Supply is entered
ýBased on plant conditions, which one of the following is an immediate action lAW HC.OP-AB.ZZ i0129?
Fa] Verify that the Control Room Supply Ventilation has automatically isolated I Verify that the "A" Control Room Emergency Filter Unit automatically started
'Press the CONTROL ROOM EMER FILTER UNIT A and B OA pushbuttons WPress the CONTROL ROOM EMER FILTER UNIT A and B RECIRC MODE pushbuttons Lvl B it~iv Lev'elMmoya Cognd~ai Hope Creek Dt. W/1 2/2002' 03a fTier: 'Plant Systems [OGru 2.ý 12. 290003K501 K5.j Knowledge of the operational implications of the following concepts as they apply to CONTROL ROOM iHVAC K5.01 1lAirborne contamination (e.g., radiological, toxic gas, smoke) control 3.2!3.51 Explanation of Press the CONTROL ROOM EMER FILTER UNIT A and B RECIRO MODE push-buttonls. For a tox-ic g-a-s!
in the Control Room Supply, isolate Control Room Ventilation and place CREF in the Recirc Mode.
INCORRECT - Press the CONTROL ROOM EMER FILTER UNIT A and B OA pushbuttons. CREF must be in the Recirc Mode for a toxic gas event.
INCORRECT - Verify that the Control Room Supply Ventilation has automatically isolated. Toxic gas will not automatically isolate Control Room Ventilation. Only high rad.
INCORRECT -Verify that the "A' Control Room Emergency Filter Unit automatically started. Does not automatically start on toxic gas, only high rads.
,HC.OP-AB.ZZ-0129
......... Supply, Abnormal Operating Procedure.
F l Exam .......
B . . . .. Significantly Modifie Q e t o S o rc .. .. .. .... . ... . . ..... . .. ...... .. . . .. ........... . .... . ................... .. ... ... ................. .. .. . . .. .. ..................... . . . . ............... ............ . . . . . ..... .......... ....
d . . .......... ..
VISION BANK QID# Q61261 Friday, March 22, 2002 2:10:12 PM Page 114 of 139
HC.OP-AB.ZZ-0129(Q)
Approved: ,
Operations anage '- HCO Date CATEGORY II HIGH RADIATION, SMOKE OR TOXIC GASES IN THE CONTROL ROOM AIR SUPPLY 1.0 SYMPTOMS 1.1 Alarms
- a. RADIATION MONITORING ALARM/TRBLA
- b. CONTROL RM AIRBORNE ACTIVITY HI
- c. FIRE PROT PANEL 10C671 1.2 Increased activity reading for the control room air supply radiation monitors.
1.3 Smoke in the Control Room air supply 1.4 Pungent odor in the Control Room air supply 2.0 AUTOMATIC ACTIONS 2.1 Control room ventilation isolates and Control Room Emergency Filter Unit (CREF) starts (2 x 10-5 micro curies/cc) in OA Mode.
3.0 IMMEDIATE OPERATOR ACTIONS 3.1 If smoke OR toxic gases are detected in the control room air supply, isolate the Control Room ventilation and place the CREF in the RECIRC MODE.
3.2 Ensure that all appropriate automatic actions are complete.
4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 Ensure that all appropriate immediate operator actions are complete.
4.2 IF the control room atmosphere becomes smoke filled OR is suspected of being contaminated by toxic gases OR airborne radioactivity, don protective clothing AND respiratory equipment as necessary.
4.3 IF high radiation is detected in the air supply intake request the Radiation Protection Department to survey the control room AND limit access as necessary.
Hope Creek Page I of 3 Rev. 3
HC.OP-AB.ZZ-0129(Q) 4.4 Notify Site Protection IF smoke OR toxic gases are detected in the control room air supply.
4.5 Notify plant personnel of the high radiation, smoke, OR toxic gas in the control room air supply.
CAUTION 4.6 Protective clothing and respiratory equipment maybe required for the personnel involved.
4.6 Determine the source of the high radiation, toxic gas, OR smoke AND initiate corrective action.
CAUTION 4.7 Control Room personnel should not remove protective clothing OR respiratory equipment until Radiation Protection Department has lifted the requirements.
4.7 Implement the HCGS Emergency Plan IF the appropriate entry criteria is satisfied.
4.8 After the source of the contaminated atmosphere has been located AND isolated, ventilate the control room.
4.9 IF it is determined that it is necessary to evacuate the control room, implement procedure HC.OP-AB.ZZ-O130(Q).
5.0 DISCUSSION 5.1 The Control Room Ventilation System will automatically isolate on high radiation in the outside air supply or a LOCA signal.
5.2 The ISOLATE pushbutton for the control room supply (CRS) unit will close one inlet isolation damper to each CRS unit and the Control Area Exhaust (CAE) fan, trip the C.\ 1i fan, and start the associated CREF unit.
Continued Next Page Hope Creek Page 2 of 3
HC.OP-AB.ZZ-0129(Q) 5.3 (Continued) 5.3 The CREF has two modes of operation:
- a. OA MODE (pressurizing) is an automatic mode following detection of high airborne radioactivity in the control room normal air intake, in which 1000 cfm of outside air is mixed with 3000 cfm of control room return air, thus pressurizing the control room.
- b. RECIRC MODE (recirculation or isolation) is when the outside air intake isolation damper for the CREF unit is closed. In this mode 4000 cfm of return air is circulated through the CREF unit without the introduction of outside air.
However, this mode is not used following a radiological accident.
5.4 The loss of airflow in the CREF system automatically trips and isolates the operating train and alarms in the main control room. Manual operation is required to start the standby train. Loss of airflow AND/OR high pressure differential across the filter train are alarmed in the main control room.
5.5 The protective clothing described in Step 4.2 is non-radiological protective clothing. Use of Radiological protective clothing is restricted to the RCA only.
5.6 The existence of this procedure fulfills the requirements of the following Closing Documents:
CD-181X NHO HSAR F06-0126-00 CD-176X NHO HSAR F06-0119-00 CD-175X NHO HSAR F06-0118-00 Hone Creek Page 3 of 3 Rev. 3
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020327 PSEG Internal Use Only HC.OP-SO.GK-0001(Q) 5.3 Manual Isolation NOTE 5.3 A. The system can be manually isolated 50 sec after a Process Inhibit Signal is generated by a LOP or LOCA.
B. Isolation of Control Area Supply Unit should be for the running unit.
5.3.1 ENSURE that all prerequisites have been satisfied IAW Section 2.3.
5.3.2 PRESS HD-9598A(B) CONTROL AREA ISOLATION DAMPERS A(B) ISOLATE PB.
5.3.3 OBSERVE that CONTROL RM EMER FILTER UNIT A(B)V400 START is ON.
NOTE 5.3.4 The Control Room Emergency Filter Unit is only intended to be operated in the recirculation mode IF smoke or toxic gases are detected in the control room air supply.
This mode is not automatically initiated following a radiological accident 5.3.4 IF required to operate Control Room Emergency Filter Unit in recirculation mode, PERFORM the following:
A. PRESS CONTROL RM EMER FILTER UNIT A(B) RECIRC MODE.
B. OBSERVE that OA DAMPER HD-9593A(B) CLOSED is ON.
Hope Creek Page 15 of 22 Rev. 9
Given the following:
- Tech Spec compliance has been verified lAW "Refueling Operations". [HC.OP-IO.ZZ-0009]
- Multiple Control Rod Drive Mechanisms are being removed lAW Technical Specification 3.9.10.2
- Spiral Fuel offload is in progress per directions of Reactor Engineers and Fuel Handling Control
.Core Alteration forms. [HC.RE-FR.ZZ-0001]
- 14 Fuel Assemblies are remaining in the Vessel Which one of the following conditions would require a formal declaration of Suspension of Core Alterations as described in plant procedures?
.Spent Fuel Storage Area Radiation Monitor in alarm while transporting LPRMS through the
.Cattle Shute IEAll SRMs indicate between 2.1 & 2.6 cps El Mode Switch position change from Shutdown to Refuel for Rod Speed adjustments per system operating procedure
!Refueling Bridge Platform surveillance identifies Frame Mounted hoist up travel stops are out of Technical Specification tolerance aa -]~~~~~~~...
Answer13 ...e ...--
x... .... .. -]
R ognitive Lev ....... [pliaio
ca jo ............ Hope ii i e* iii iiii i ~
Creek ~............... 603/12/20021 i 2 2 -2 Tier Generic
[G e e ri -K Knowledge and Abilities n wie lg.........................................................
a nc~ *ii..e ].. ii [..[........0..G.2.
i ............ 0....
.. 0 6 22
,GENERICor ume 116 2.2 Equipment Control 2.2.27 [knowledg of the refueling process. 2.61 3.5J Expanaio Justification lneHC.OP-IO.ZZ-0009, directs use of NC.NA-AP.ZZ-0049, for direction on formal suspension of fuel
.handling activities, adverse radiological conditions are one of the criteria.
Additionally, Refuel Radiation Area Alarms is an entry condition for HC.OP-AB.ZZ-01 01 "Irradiated Fuel Damage" which directs suspension of all refueling operations.
- Other L.~~~
choices
.. ~
~~
. .~
~. ~~~~~~~ are all within the Allowable Technical Specification ........... ............. ... ... .....for
. ... ....... .................................. ... . . ... ........ ........... ...........boundaries Core Alterations.
0NC.NA-AP.ZZ-004g 001 121E004 (R) Apply Precautions, Limitations and Notes while executing the. REFUELING OPERATIONS Integrated Operating Procedure S.
[Q. = nSu l Facility Exam Bank i Mo c Metod: Editorially Modified Q SoucI CVision Bank QID# 058930 Friday, March 22, 2002 11:53:56 AM I Page 125 of 139
LESSON NAME: 0302-000.00H-0001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06/00 the following terms, lAW NC.NA-AP.ZZ-0037:
- a. SPCC [Spill Prevention, Control, and Countermeasure Plan]
- b. DPCC [Discharge Prevention, Containment, and Countermeasure Plan]
- 55. Provided access to control room references, initiate chemical item classification permit, lAW NC.NA-AP.ZZ-0038.
- 56. Given access to control room references, evaluate a CICP and determine limitations on use/disposal prior to start of job, lAW NC.NA AP.ZZ-0038.
- 57. State the purpose of Station Aids, lAW NC.NA-AP.ZZ-0044.
- 58. Determine if a proposed Station Aid may be posted, lAW NC.NA AP.ZZ-0044.
- 59. Define the following terms, lAW NC.NA-AP.ZZ-0049:
- a. Core Alterations
- b. Suspension of Fuel Handing (or Core Alterations
- 61. State the responsibilities of the Refueling Bridge Operator, lAW NC.NA AP.ZZ-0049.
- 62. State the responsibilities of the Control Room Refuel Monitor, lAW NC. NA-AP.ZZ-0049.
- 63. Provided access to control room references, determine the required actions if a fuel assembly is found in an incorrect core location, lAW NC.NA-AP.ZZ-0049.
- 64. State the minimum fuel handling crew requirement for fuel handling involving core alterations, lAW NC.NA-AP.ZZ-0049.
- 65. State the minimum fuel handling crew requirement for non-core alteration fuel handling, lAW NC.NA-AP.ZZ-0049.
- 66. Determine the conditions under which handling of fuel must be suspended, lAW NC.NA-AP*.ZZ0049. (SRO ONLY)
- 67. Determine the types of maintenance activities that will typically require a Post Maintenance Test and/or Operability Retest, lAW NC.NA-AP.ZZ 0050 and NC.NA-TS.ZZ-0050.
- 68. Provided access to control room references, determine the conditions that would necessitate changing Post Maintenance Testing or S:VDrive\OPS\HC\Master LPs\O001131 O.doc 13 of 39
-_L ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 5.2.4 (Continued)
G. All material nonconformances identified during the performance of Fuel Handling shall be documented in accordance with NAP-0.
H. (Hope Creek) All LPRM and Control Blade replacements shall be performed in accordance with procedures in the sequence prescribed by Reactor Engineering.
- 1. (Hope Creek) During irradiated fuel movement, the time spent with fuel in transit between the spent fuel pool and reactor pressure vessel shall be minimized to avoid the potential for high radiation doses in the upper regions of the drywell. [CD-612X]
J. (Salem) During irradiated fuel movement, the time spent with fuel in the transfer tube shall be minimized to avoid the potential for high radiation doses in the transfer tube area.
K. When sufficient manpower exists, Fuel Handling should be conducted using more than one Fuel Handling crew per operating shift. This allows the periodic rotation of Fuel Handling crew personnel and decreases the individual radiation dose received.
5.2.5 The suspension of Fuel Handling may be directed by the OS/CRS or Refueling SRO as appropriate. Prior to the resumption of Fuel Handling, applicable Technical Specification Surveillances shall be verified current. The following conditions require the suspension of Fuel Handling activities:
"* Refueling floor radiological conditions that require the termination of refuel work activities.
" Any non-compliance with the applicable unit Technical Specifications governing the performance of Fuel Handling involving Core Alterations.
" Any neutron monitoring channel indicates unexpected increasing count rate.
" Any fuel damage occurs including the dropping, bumping, scraping or general mishandling of fuel or other suspended loads during handling.
" (Hope Creek) For any reactor core fuel loading, if any control rod or associated control rod drive mechanism has been removed from the reactor core or pressure vessel.
Nuclear Common Page 11 of 13 Rev. I
- j/,1 L.P. NO.: 0302-000.00H-000113-10 0303 0304 0305 PREVIOUS L.P. NO.: 0302-000.00H-000113-09 NUCLEAR TRAINING CENTER LESSON PLAN PROGRAM TITLE: HOPE CREEK OPERATOR TRAINING COURSE SECTION/MODULE: OPERATING PROCEDURES TOPIC/SUB MODULE: ADMINISTRATIVE PROCEDURES LESSON: NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES DURATION: VARIES DEPENDING ON PROCEDURE(S)
PREREQUISITES:
JTA NO. OR QUALIFICATION STATEMENT NO.:
AUTHOR: Peter Doran DATE: 01/06/00 I REVIEW/APPROVAL SIGNATURES I SUBMITTED BY: Peter Doran DATE: 01/06/00 BARGAINING UNIT REPRESENTATIVE: DATE:
PRINCIPAL TRAINING SUPERVISOR: DATE:
LINE SUPERVISOR: DATE:
COPY RECEIVED WORD PROCESSING INITIALS:
LESSON NAME: 0302-000.001-0001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01106100 INSTRUCTOR
REFERENCES:
Number Title
A. PROCEDURE
S NC.NA-AP.ZZ-0001 Nuclear Procedure System NC.NA-AP.ZZ-0002 Nuclear Department Organization NC.NA-AP.ZZ-0005 Station Operating Practices NC.NA-AP.ZZ-0009 Work Control Process NC.NA-AP.ZZ-0013 Control of Temporary Modifications NC. NA-AP.ZZ-0015 Safety Tagging Program NC.NA-AP.ZZ-0020 Control of Nonconforming Components and Structures NC.NA-AP.ZZ-0023 Scaffolding Program NC.NA-AP.ZZ-0024 Radiation Protection Program NC.NA-AP.ZZ-0025 Operational Fire Protection NC.NA-AP.ZZ-0037 Environmental Control NC.NA-AP.ZZ-0038 Chemical Control Program NC.NA-AP.ZZ-0044 Station Aids and Labels NC.NA-AP.ZZ-0049 Conduct of Fuel Handling NC.NA-AP.ZZ-0050 Station Testing Program NC.NA-AP.ZZ-0059 10CFR50.59 Reviews and Safety Evaluations NC.NA-AP.ZZ-0069 Work Control Coordination NC.NA-AP.ZZ-0070 Inservice Testing and MOV Testing Programs NC. NA-AP.ZZ-0071 Fuel Integrity Program NC.NA-AP.ZZ-0083 Transient Loads NC.NA-AP.ZZ-0084 Conduct of Infrequently Performed Tests or Evolutions HC.SA-AP.ZZ-0002 Station Organization HC.SA-AP.ZZ-0021 Station Cleanliness Program HC.RA-AP.ZZ-0051 Leakage Reduction Program SH. O P-AP.ZZ-0107 Shift and Relief Turnover HC.OP-AP.ZZ-0005 Department Operating Practices HC.OP-AP.ZZ-0012 Technical Specification Surveillances HC.OP-AP.ZZ-0101 Post Reactor Scram/ECCS Actuation Review and Approval Requirements HC.OP-AP.ZZ-0102 Use of Operations Department Procedures HC.OP-AP.ZZ-0108 Operability Assessment and Equipment Control Program HC.OP-AP.ZZ-0109 Equipment Operational Control HC.OP-AP.ZZ-0110 Use and Development of Operating Logs PSEG Nuclear Confined Space Entry Safety Manual (section 9.0)
HC.OP-EO.ZZ-103 Reactor Building Control SH.OP-DD.ZZ-0004 Operations Standards HC.OP-DD.ZZ-0067 Personnel Qualification and Training S:\VDriveXOPS\HC\Master LPs\0001 131 O.doc 2 of 39
LESSON NAME: 0302-000.001-000113-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06100 Number Title NC.WM-AP.ZZ-0000 Notification Process NC.WM-AP.ZZ-0001 Work Management Process NC.WM-AP.ZZ-0003 Regular Maintenance Process NC.NA-TS.ZZ-0050 Maintenance Testing Matrix B. DOCUMENTS DOCUMENT DESCRIPTION SER 29-82 (PTS-306) INPO Significant Event Report 29-82;
Subject:
Trip of (CD-515A) Both Reactors at Site following Condensate Pump Trip to be discussed with NC.NA-AP.ZZ-0013 AR-M20-89-0109 Potential/Impact of Unauthorized Mods; to be discussed with NC.NA-AP.ZZ-0013 DCP 4HC-0214 Deletes requirements to use post trip log for review and disables automatic log printout SOER 85-5 (PTS INPO Significant Operating Experience Report: Internal 1437) (CD-265E) Flooding of Power Plant Buildings INPO IS 912 SOER 85-5 Recommendation Clarification IE Notice 85-27 Notification to the NRC Operations Center and Report Events in Licensee Event Reports NLR-191426 ESF System Actuation Reportability NLR-191148 Engineered Safety Features System Actuations ASME Section XI, Division 1, Articles IWP and IWV CD-358F Inservice testing of DG fuel oil transfer pumps 10CFR50.54(x) and (y) Code of Federal Regulations 10CFR50.59 Code of Federal Regulations IN 87-21 Shutdown order issued because Licensed Operators asleep while on duty IN 88-20 Unauthorized individual manipulating controls and performing control room activities IN 92-30 Falsification of Records GEK-90300 Reactor Assembly and Servicing Equipment, Vol. II Part 2
FSAR Auxiliary Systems, Vol. 13, Chapter 9.1 T/S HCGS Technical Specifications S:\VDrive\OPS\HC\Master LPs\0001 131 O.doc 3 of 39
LESSON NAME: 0302-000.001-000113-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01106100 TRAINING MATERIAL REQUIRED:
- 1. Lesson Plan
- 2. NC.NA-AP.ZZ-0001, Nuclear Procedure System
- 3. NC.NA-AP.ZZ-0002, Nuclear Department Organization
- 4. NC.NA-AP.ZZ-0005, Station Operating Practices
- 5. NC.NA-AP.ZZ-0005, Station Operating Practices
- 6. NC.NA-AP.ZZ-0013, Control of Temporary Modifications
- 7. NC.NA-AP.ZZ-0015, Safety Tagging Program
- 8. NC.NA-AP.ZZ-0020, Control of Nonconforming Components and Structures
- 9. NC.NA-AP.ZZ-0023, Scaffolding Program
- 10. NC.NA-AP.ZZ-0024, Radiation Protection Program
- 11. NC.NA-AP.ZZ-0025, Operational Fire Protection Program
- 12. NC.NA-AP.ZZ-0037, Environmental Control
- 13. NC.NA-AP.ZZ-0038, Chemical Control Program
- 14. NC.NA-AP.ZZ-0044, Station Aids and Labels
- 15. NC.NA-AP.ZZ-0049, Conduct of Fuel Handling
- 16. NC.NA-AP.ZZ-0050, Station Testing Program
- 17. NC.NA-AP.ZZ-0059, 10CFR50.59 Reviews and Safety Evaluation
- 18. NC.NA-AP.ZZ-0069, Work Control Coordination
- 19. NC.NA-AP.ZZ-0070, Inservice Testing and MOV Testing Programs
- 20. NC.NA-AP.ZZ-0071, Fuel Integrity Program
- 21. NC.NA-AP.ZZ-0083, Transient Loads
- 22. HC.SA-AP.ZZ-0002, Station Organization
- 23. HC.SA-AP.ZZ-0021, Station Cleanliness Program
- 24. SH.OP-AP.ZZ-0107, Shift and Relief Turnover
- 25. HC.OP-AP.ZZ-0005, Department Operating Practices
- 26. HC.OP-AP.ZZ-0012 Technical Specification Surveillances
- 27. HC.OP-AP.ZZ-0101, Post Reactor Scram/ECCS Actuation Review and Approval Requirements
- 28. HC.OP-AP.ZZ-0102, Use of Operations Department Procedures
- 29. HC.OP-AP.ZZ-0108, Operability Assessment and Equipment Control Program
- 30. HC.OP-AP.ZZ-0109, Equipment Operational Control
- 31. HC.RA-AP.ZZ-0051, Leakage Reduction Program
- 32. SH.OP-DD.ZZ-0004, Operations Standards
- 33. HC.OP-DD.ZZ-0067, Personnel Qualification and Training
LESSON NAME: 0302-000.001-0001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 011106100 STUDENT HANDOUTS:
- 1. Lesson Plan I Terminal and Enabling Learning Objectives LATEST REVISION OF THE FOLLOWING PROCEDURES
- 2. NC.NA-AP.ZZ-0001, Nuclear Procedure System
- 3. NC.NA-AP.ZZ-0002, Nuclear Department Organization
- 4. NC.NA-AP.ZZ-0005, Station Operating Practices
- 5. NC.NA-AP.ZZ-0005, Station Operating Practices
- 6. NC.NA-AP.ZZ-0013, Control of Temporary Modifications
- 7. NC.NA-AP.ZZ-0015, Safety Tagging Program
- 8. NC.NA-AP.ZZ-0020, Control of Nonconforming Components and Structures
- 9. NC.NA-AP.ZZ-0023, Scaffolding Program
- 10. NC.NA-AP.ZZ-0024, Radiation Protection Program
- 11. NC.NA-AP.ZZ-0025, Operational Fire Protection Program
- 12. NC.NA-AP.ZZ-0037, Environmental Control
- 13. NC.NA-AP.ZZ-0038, Chemical Control Program
- 14. NC.NA-AP.ZZ-0044, Station Aids and Labels
- 15. NC.NA-AP.ZZ-0049, Conduct of Fuel Handling
- 16. NC.NA-AP.ZZ-0050, Station Testing Program
- 17. NC.NA-AP.ZZ-0059, 10CFR50.59 Reviews and Safety Evaluation
- 18. NC.NA-AP.ZZ-0069, Work Control Coordination
- 19. NC.NA-AP.ZZ-0070, Inservice Testing and MOV Testing Programs
- 20. NC.NA-AP.ZZ-0071, Fuel Integrity Program
- 21. NC.NA-AP.ZZ-0083, Transient Loads
- 22. HC.SA-AP.ZZ-0002, Station Organization
- 23. HC.SA-AP.ZZ-0021, Station Cleanliness Program
- 24. SH.OP-AP.ZZ-0107, Shift and Relief Turnover
- 25. HC.OP-AP.ZZ-0005, Department Operating Practices
- 26. HC.OP-AP.ZZ-0012 Technical Specification Surveillances
- 27. HC.OP-AP.ZZ-01 01, Post Reactor Scram/ECCS Actuation Review and Approval Requirements
- 28. HC.OP-AP.ZZ-0102, Use of Operations Department Procedures
- 29. HC.OP-AP.ZZ-0108, Operability Assessment and Equipment Control Program
- 30. HC.OP-AP.ZZ-0109, Equipment Operational Control
- 31. HC.RA-AP.ZZ-0051, Leakage Reduction Program
- 32. SH.OP-DD.ZZ-0004, Operations Standards
- 33. HC.OP-DD.ZZ-0067, Personnel Qualification and Training
- 34. PSEG Nuclear Safety Manual CLASSROOM REQUIREMENTS:
No special considerations for the classroom size or arrangements are required.
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LESSON NAME: 0302-000.001-1-01 013-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06100 INSTRUCTIONAL OBJECTIVES:
1.0 Provided access to control room references and/or a scenario of plant/system status or previous plant conditions associated with the operation of Hope Creek Generating Station, the trainee will interpret the applicable Nuclear Common Administrative Procedure to ensure plant operations are being conducted, and are in compliance with, the requirements of the following procedures:
A. NC NA-AP ZZ-0001, Nuclear Procedure System B. NC NA-APZZ-0002, Nuclear Department Organization C. NC.NA-AP-ZZ-0005, Station Operating Practices D. NC. NA-AP-ZZ-0009, Work Control Process E. NC. NA-AP-ZZ-0013, Control of Temporary Modifications F. NC.NA-AP-ZZ-0015, Safety Tagging Program G. NC.NA-AP-ZZ-0020, Control of Nonconforming Components and Structures H. NC.NA-AP-ZZ-0023, Scaffolding Program I. NC.NA-AP-ZZ-0024, Radiation Protection Program J. NC.NA-AP-ZZ-0025, Operational Fire Protection Program K. NC.NA-AP-ZZ-0037, Environmental Control L. NC.NA-AP-ZZ-0038, Chemical Control Program M. NC. NA-AP-ZZ-0044, Station Aids and Labels N. NC.NA-AP-ZZ-0049, Conduct of Fuel Handling
- 0. NC.NA-AP-ZZ-0050, Station Testing Program P. NC.NA-AP-ZZ-0059, 10CFR50.59 Reviews and Safety Evaluation Q. NC.NA-AP-ZZ-0069, Work Control Coordination R. NC.NA-AP-ZZ-0070, Inservice Testing and MOV Testing Programs S. NC.NA-AP-ZZ-0071, Fuel Integrity Program T. NC.NA-AP-ZZ-0083, Transient Loads U. NC.NA-AP.ZZ-0084, Conduct of Infrequently Performed Tests V. NC.WM-AP.ZZ-0000, Notification Process W. NC.WM-AP.ZZ-0001, Work Management Process X. NC.WM-AP.ZZ-0003, Regular Maintenance Process Y. NC.NA-TS.ZZ-0050, Maintenance Testing Matrix S:\VDrive\OPS\HC\Master LPs\0001 1310.doc 6 of 39
LESSON NAME: 0302-000.001H-0001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01106/00 2.0 Provided access to control room references and/or a scenario of plant/system status or previous plant conditions associated with the operation of Hope Creek Generating Station, the trainee will interpret the applicable HCGS Operations Administrative Procedure to ensure plant operations are being conducted, and are in compliance with, the requirements of the following procedures:
A. SH.OP-DD.ZZ-0004, Operations Standards B. HC.OP-AP.ZZ-0012 Technical Specification Surveillances C. HC.OP-DD.ZZ-0067, Personnel Qualification and Training D. HC.OP-AP.ZZ-0005, Department Operating Practices E. SH.OP-AP.ZZ-01 07, Shift Turnover Responsibilities 3.0 Provided access to control room references and/or a scenario of plant/system status or previous plant conditions associated with the operation of Hope Creek Generating Station the trainee will interpret the applicable HCGS Station Administrative Procedure to ensure plant operations are being conducted and are in compliance with the requirements of the following procedures:
A. HC.RA-AP.ZZ-0051,Leakage Reduction Program 4.0 Provided access to control room references and/or a scenario of plant/system status or previous plant conditions associated with the operation of Hope Creek Generating Station, the trainee will interpret PSEG Nuclear Safety Manual, section 9.0 for Confined Space Entry, to ensure plant operations are being conducted, and are in compliance with, the requirements of the procedure.
S:WDdve\OPS\HC\Master LPs\0001 131 0.doc 7 of 39
LESSON NAME: 0302-000.00H-0001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06100 ENABLING LEARNING OBJECTIVES:
- 1. State who may delegate authority to an individual, when that function is not delineated in the individual's position description, lAW NC.NA AP.ZZ-0002.
- 2. Describe the process of telephone approval and documents requiring signature approval lAW NC.NA-AP.ZZ-0002.
- 3. Explain, from memory, whom and when the OS may call-out and the approvals required to be obtained prior to making such call-outs, lAW NC.NA-AP.ZZ-0002
- 4. Determine the requirements for a licensed operator to be present "at the controls" at all times during operation of the facility, lAW NC.NA AP.ZZ-0005, and HC.OP-AP.ZZ-0005.
- 5. Given a copy of the control room layout, identify the area denoted "at the controls," lAW NC.NA-AP.ZZ-0005.
- 6. Determine who is permitted (including conditions) to manipulate controls which directly or indirectly affect reactivity or power level, lAW NC.NA-AP.ZZ-0005, and HC.OP-AP.ZZ-0005.
- 7. Given plant conditions and/or access to control room references, determine the following lAW NC.NA-AP.ZZ-0005, and HC.OP-AP.ZZ 0005:
- a. The level of licensing required for the OS, CRS, and RO/PO.
- b. Minimum shift manning requirements for all plant conditions.
- c. Normal shift staffing levels.
- d. When a person can serve a dual role as CRS/STA or OS/STA
- 8. From memory, choose the correct operator response to instrument indications, lAW NC.NA-AP.ZZ-0005, and HC.OP-AP.ZZ-0005.
- 9. Provided access to control room references, apply the overtime guidelines lAW NC.NA-AP.ZZ-0005.
- 10. Provided access to control room references, determine the requirements for maintaining an operator license active, lAW HC.OP DD.ZZ-0067.
- 11. Explain, from memory, the circumstances and approval required for Licensed Operators to deviate from Technical Specifications or license conditions, lAW NC.NA-AP.ZZ-0005 and 1OCF50.54(x).
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LESSON NAME: 0302-000.001-000113-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01106/00
- 12. Describe the control room restrictions pertaining to the following lAW NC.NA-AP.ZZ-0005:
- a. Control room access during normal and transient operation.
- b. The conduct of plant related technical or administrative business or personal business.
- c. Access for non-shift personnel
- 13. State the conditions which require Operations Manager notification, lAW NC.NA-AP.ZZ-0005 and SH.OP-AP.ZZ-0004
- 14. Determine the requirements for Independent Verification lAW NC.NA AP.ZZ-0005.
- 15. Describe how to perform an Independent Verification lAW NC.NA AP.ZZ-0005
- 16. Provided access to control room references, determine the Reactor Shutdown considerations lAW SH.OP-DD.ZZ-0004.
- 17. Given a set of conditions, determine when a Motor Operated Valve must be declared Inoperable due to manual operation, lAW NC.NA AP.ZZ-0005.
- 18. Given a set of conditions, determine when a manually operated valve shall be locked, and the correct method to perform this function, lAW NC.NA-AP.ZZ-0005
- 19. State the purpose of the Notification Process, lAW NC.WM-AP.ZZ 0000,
- 20. Describe the responsibilities of All NBU Personnel, lAW NC.WM-AP.ZZ 0000.
- 21. Given access to control room references, describe how to validate a corrective maintenance action request, lAW NC.WM-AP.ZZ-0003.
- 22. lAW NC.WM-AP.ZZ-0000 describe the actions for OS/CRS review of a Notification to determine the following: (SRO ONLY)
- a. If operability is required
- b. Determine Operability
- c. Reportability requirements
- 23. State the responsibility of the OS/CRS for the work control process lAW NC. NA-AP.ZZ-0009.
- 24. State the two purposes of an EMIS tag, lAW NC.WM-AP.ZZ-0000 S:VDrive\OPS\HC\Mlaster LPs\0001 131 O.doc 9 of 39
LESSON NAME: 0302-000.00H1-000 113-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01106/00
- 25. Determine the proper method of completing and hanging an EMIS tag, lAW NC.WM-AP.ZZ-0000.
- 26. Provided access to control room references, determine when it is necessary to obtain OS/CRS permission prior to starting work, lAW NC.WM-AP.ZZ-0001.
- 27. Determine the proper method of initiating a Work Package for Unscheduled LCO entry due to TS Equipment Declared INOPERABLE lAW NC.WM-AP.ZZ-0001.
- 28. Determine the responsibilities of the Job Supervisor, lAW NC.NA AP.ZZ-0009.
- 29. Given plant conditions, evaluate plant status to determine if work may be conducted as scheduled, lAW NC.WM-AP.ZZ-0001 for the following:. (SRO ONLY)
- a. Minor Maintenance
- b. Work Package
- 30. Determine the proper method of closing and completing a Work Package, lAW NC.WM-AP.ZZ-0001.
- 31. Given plant problems/industry events associated with station maintenance:
- a. Discuss the root cause of the plant problem/industry event, lAW SOER 85-5.
- b. Discuss HCGS procedural guidelines that mitigate/reduce the likelihood of the problem/industry event at HCGS, lAW:
- OP-EO.ZZ-103, Reactor Building Control EOP
- NC.NA-AP.ZZ-0015, Safety Tagging Program.
- NC.WM-AP.ZZ-0001, Work Management Process
- NC.NA-AP.ZZ-0050, Station Testing Program
- 32. Explain the appropriate actions to be taken prior to authorizing performance of a Surveillance and/or Inservice Test, lAW NC.WM AP.ZZ-0003 and HC.OP-AP.ZZ-0012. (SRO ONLY)
- 33. Explain the appropriate action to be taken when Surveillance and/or Inservice test results are determined to be:
- a. acceptable
- b. unacceptable lAW NC.WM-AP.ZZ-0003 and HC.OP-AP.ZZ-0012 S:VDrive\OPS\HCwlaster LPs\O001 131 O.doc 10 of 39
LESSON NAME: 0302-000.001-000113-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06100
- 34. State the appropriate actions to be taken when Temporary Modification Tags are missing, lAW NC.NA-AP.ZZ-0013.
- 35. State the meaning of the following terms, lAW NC.NA-AP.ZZ-0013:
- b. Lifted Lead
- c. Electrical Jumper
- d. Independent Verification
- 36. Given the procedure for Control of Temporary Modifications and a specific group of proposed actions, determine whether or not a Temporary Modification is required, lAW NC.NA-AP.ZZ-0013. (SRO ONLY)
- 37. Provided with a Troubleshooting Plan for which a Temporary Modification Package (TMP) is not required and access to control room references, determine whether the Troubleshooting Plan should be approved, lAW NC.NA-AP.ZZ-0013 and SH.OP-AP.ZZ-0008. (SRO ONLY)
- 38. Provided with a Temporary Modification Package (TMP) which affects floor drain plug status and access to control room references, determine whether the Temporary Modification Package should be approved, lAW NC.NA-AP.ZZ-0013. (SRO ONLY)
- 39. Identify the responsibilities of the OS/CRS for Control of Temporary Modifications for the following, lAW NC.NA-AP.ZZ-0013: (SRO ONLY)
- a. Installation of a T-Mod
- b. Reviewing installed T-mods
- c. Removing T-Mods
- d. Preparing a T-Mod for an alarm bypass
- e. Expedited Temp Mod packages
- 40. Explain the responsibilities of the "User" of Measuring and Test Equipment, lAW NC.NA-AP.ZZ-0022.
- 41. Describe what the worker is acknowledging when signing a RWP prior to use, lAW NC.NA-AP.ZZ-0024, Radiation Protection Program.
- 42. State the definition of the following terms, lAW NC.NA-AP.ZZ-0024:
- a. Contaminated Area
- d. Radiation Area
- e. Restricted Area S:\VDrive\OPS\HC\Master LPs\0001 131 O.doc 11 of 39
LESSON NAME: 0302-000.01-10001 13-10 NUCLEAR COMMON, HCGS OPERATIONS, AND HCGS STATION ADMINISTRATIVE PROCEDURES - 01/06/00
- f. Very High Radiation Area
- g. Airborne Radioactivity Area
- h. Declared Pregnant Woman (DPW)
- i. Total Effective Dose Equivalent (TEDE)
- 43. State the responsibilities of the following personnel for issuance of keys to locked High Radiation Areas, lAW NC.NA-AP.ZZ-0024:
- a. Key Holder
- b. OS
- c. Radiation Protection Supervisor
- 44. Given a set of exposure conditions, identify the personnel responsible for approval of the following dose extension, lAW NC.NA-AP.ZZ-0024:
- a. Yearly Dose Extension
- b. Declared Pregnant Women Dose Extension
- c. Lifetime Dose Extension
- 45. State the actions of an individual discovering a fire, lAW NC.NA-AP.ZZ 0025.
- 46. Explain the controls utilized over the following, lAW NC.NA-AP.ZZ 0025:
- a. Combustible Material
- b. Flammable Liquids and Gases
- c. Ignition Sources
- 47. Explain the controls utilized over impairments to fire protection systems, lAW NC.NA-AP.ZZ-0025.
- 49. Explain the use of the Transient Combustible Load Allowances Table, lAW NC.NA-AP.ZZ-0025.
- 50. Explain the responsibilities of the Job Supervisor for entry into cleanliness Zone II, lAW NC.NA-AP.ZZ-0031. (SRO ONLY)
- 51. State the requirements of Zone II cleanliness controls based on work scope as delineated in NC.NA-AP.ZZ-0031.
- 52. State when a Pre-startup Walkdown is required as delineated in NC.NA-AP.ZZ-0031.
- 53. Determine the Group that the OS will initially notify to respond to a major spill, lAW NC.NA-AP.ZZ-0037. (SRO ONLY)
- 54. Provided access to control room references, determine the meaning of S:\VDrive\OPS\HC\Master LPs\O001 131 O.doc 12 of 39
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only Page 1 of I PSEG NUCLEAR L.L.C.
HOPE CREEK GENERATING STATION HC.OP-IO.ZZ-0009(Q) - Rev. 24 REFUELING OPERATIONS USE CATEGORY: I A. Biennial Review performed Yes No N/A "
B. Change Package(s) and Affected Document Number(s) incorporated into this revision.
C. OTSC(s) incorporated into this revision:
- OTSC No(s) or None /
REVISION
SUMMARY
- 1. Based on request made under Order 80033269 (T/S Amendment 137) all references to Refueling Operations Technical Specifications 3/4.9.4 Decay Time, 3/4.9.5, Communications, and 3/4.9.6, Refueling Platform, were replaced with new Updated Final Safety Analysis Report references. As part of T/S Amendment 137, the above mentioned T/Ss and their associated bases are being moved to section 9.1.4.2.12 of the UFSAR. These changes are reflected in procedure steps 1.0, 2.1.1, 3.3, 3.4, Note 5.1, steps 5.1.10, 5.1.12, 5.1.13, Caution 5.3.1.E, steps 5.3.1 .E.8, 5.3.l.E.10, and associated signoffs and references. These changes can be considered editorial based on allowances in NC.DM-AP.ZZ-0001(Q) for incorporating a change that has already been reviewed and approved in accordance with another approved process.
- 2. Based on Writers Review, the following editorial changes were made:
- Converted Note 5.1.4, item A, into step 5.1.4 due to the action contained therein to record time and date of shorting link removal.
- Updated reference to SH.OP-AP.ZZ-0108(Q), in step 5.1.14.B.
- Added RECORDS section 6.0.
- Removed reference to HC.OP-MD.KE-0001(Q), Refueling Platform 7-Day Operational Checks, at step 5.1.12.A.1. These checks are now captured within HC.OP-ST.KE-0001, Refuel Interlock Operability Functional Test, which is verified current under 5.1.12.B.
- Reformatted signoff under step 5.1.12.A.2.
IMPLEMENTATION REQUIREMENTS Effective date _____/____
Implementation of T/S Amendment 137.
APPROVED: hZt C3/5-' I Manager - Hope Creek Operations Date
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q)
REFUELING OPERATIONS TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ............................................................................................................ 2 2.0 PREREQUISITES .................................................................................................... 2 3.0 PRECAUTIONS AND LIM ITATIONS ................................................................... 3 4.0 EQUIPM ENT REQUIRED ..................................................................................... 4 5.0 PROCEDURE ........................................................................................................... 5 5.1 Administrative Controls for Starting CORE ALTERATIONS ........................................................................... 5 5.2 CORE ALTERATIONS and In-vessel Tests, Inspections and Maintenance Activities ................................................... 14 5.3 Administrative Controls for Suspension and Resumption of CORE ALTERATIONS .................................................. 19 6.0 RECORDS 24 I
7.0 REFERENCES
24 ATTACHMENTS Attachment I CORE ALTERATIONS Technical Specification / UFSAR I Requirem ents Checklist ............................................................................ 27 Attachment 2 Technical Specifications Required for CORE ALTERATIONS Review List .................................................................. 31 Attachment 3 CORE ALTERATIONS and In-vessel Testing, Inspection and Maintenance Completion Review .................................... 32 Attachment 4 Resuming Core Alterations Technical Specification / UFSAR Requirem ents Checklist ............................................................................ 33 Attachment 5 Placing the Plant in Alternate Decay Heat Removal Mode of Operation .................................................................... 36 Attachment 6 Vessel Level Instrumentation Temperature Com pensation Curves ............................................................................. 38 Hope Creek Page I of 40 Rev. 24
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q)
REFUELING OPERATIONS START TIME DATE BY TERMINATION TIME DATE BY COMPLETION TIME DATE BY 1.0 PURPOSE This procedure provides a mechanism for ensuring that the TECH SPEC / UFSAR requirements are satisfied prior to conducting CORE ALTERATIONS. In addition this procedure provides a means of identifying and coordinating in-vessel activities with other plant activities while the plant is in OPERATIONAL CONDITION 5 - Refueling. [CD-443X]
2.0 PREREQUISITES 2.1 Plant Initial Conditions 2.1.1 The plant is in OPERATIONAL CONDITION - 5.
All TECH SPEC / UFSAR requirements for this operating condition have been satisfied AND are being maintained.
2.1.2 The Reactor Vessel preparations for refueling have been completed IAW HC.OP-IO.ZZ-0005(Q), Cold Shutdown to Refueling, AND indicated on Attachment 1.
2.1.3 The Reactor Cavity (AND Dryer/Separator Pool if applicable) have been flooded to a level equal to the normal level of the Spent Fuel Pool.
The Fuel Pool Gates have been removed.
2.1.4 One of the following is in-service, if required, to remove decay heat AND provide Reactor Core circulation:
"* Residual Heat Removal System in Shutdown Cooling IAW HC.OP-SO.BC-0002(Q), Decay Heat Removal Operation.
"* Alternate Decay Heat Removal operation IAW Attachment 5 OR
"* Alternate Fuel Pool Cooling Assist Mode of RHR operation 1AW HC.OP-SO.BC-0002(Q). (For Full Core Offload and the transition to and from a Full Core Offload, This mode of operation should be available prior to Offloading 1/31 of the core).
Hope Creek Page 2 of 40 Rev. 24
ALL -ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 2.2 Zone II housekeeping cleanliness requirements have been established for the Spent Fuel Pool, Reactor Cavity, AND Dryer/Separator Pools IAW NC.NA-AP.ZZ-003 1(Q), Artificial Island InspectionfHousekeeping Program.
2.3 A list of CORE ALTERATIONS AND in-vessel tests, inspections, and maintenance activities has been obtained that are unique to the current refueling outage from the Outage Planning Department.
3.0 PRECAUTIONS AND LIMITATIONS 3.1 In the event that plant conditions require a delay during some part of this procedure, the Operations Superintendent/Control Room Supervisor (OS/CRS) shall retain this procedure UNTIL it is continued OR terminated.
3.2 IF this procedure is terminated PRIOR to completion, THEN the OS/CRS shall note the reason, time, AND date of termination on this procedure..
3.3 The TECH SPEC / UFSAR requirements described in Section 5.1 shall be satisfied PRIOR to the start of any activity resulting in a CORE ALTERATION.
3.4 IF the CORE ALTERATION requirements of TECH SPEC / UFSAR CAN NOT be maintained, THEN the following are required:
3.4.1 The OS/CRS shall direct the Refuel Floor Supervisor (RFS) to suspend those activities resulting in a CORE ALTERATION.
3.4.2 The RFS shall direct personnel performing CORE ALTERATIONS to place hoisted fuel OR core components in a stible configuration AND suspend subsequent CtRE ALTERATIONS.
3.5 Section 5.2 of this procedure describes tests, inspections, AND CORE ALTERATIONS. The procedure stqdescribed in this section are NOT required to be performed in order. The exact -auence of test, inspections, AND CORE ALTERATIONS is determined by the*khedule of events prepared by the Outage Planning Department.
Hope Creek Page 3 of 40 Rev. 24
ALL -ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 3.6 To prevent affecting the Reactor Shutdown Margin DO NOT allow Reactor temperatures to go below 70 0F(when fuel is in the vessel).
Cooldown below 68'F could result in an invalidation of Shutdown Margin calculations which are based in part on the Reactor being in the shutdown condition; cold i.e. 68 0 F AND Xenon free.
3.7 CORE ALTERATIONS AND in-vessel activities have the potential for affecting the Reactor shutdown margin, exposing personnel to high levels of radiation, contamination, AND other safety hazards. The OS/CRS AND Control Room personnel shall:
3.7.1 Be aware of all in-progress tests, inspections, AND CORE ALTERATIONS.
3.7.2 Direct, control, AND coordinate the alignment AND operation of plant systems with the activities in-progress on the refuel floor.
3.8 A designated "spotter" is required for all bridge activities which require any grapple to be loaded.
3.9 NO individual should perform bridge activities for greater than six consecutive hours.
3.10 Testing of IST Valves need NOT begin in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after Cold Shutdown when entering a Refueling Outage, provided all valves required to be tested during Cold Shutdown/Refueling will be tested before Plant Startup.
All Cold Shutdown, Refueling Outage, AND Containment Deine valves shall be tested before startup from Refueling Outqes, unless testing has been completed within the previdfis 92 days.
- IF an outage lasts beyond 92 days, THEN all Cold Shutdown'testing shall be completed AND all Cold Shutdown testing shall continue such that all applicable components havo.been tested within the last 92 days of the shutdown.
4.0 EQUIPMENT REQUIRED None Hope Creek P-age 44.f 40 Rev. 24
ALL ACiIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2U020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 5.3 Administrative Controls for Suspension and Resumption of CORE ALTERATIONS NOTE 5.3A Suspension of CORE ALTERATIONS shall be a formal declaration by the Refuel Floor SRO, OR a suspension of CORE ALTERATIONS as required by Technical Specifications.
Section 5.2.5 of NC.NA-AP.ZZ-0049(Q), Conduct of Fuel Handling, discusses the suspension of fuel handling.
5.3.1 IF CORE ALTERATIONS were suspended, THEN PERFORM the following PRIOR to resumption of CORE ALTERATIONS:
A. RECORD the time, date, AND reason for suspending CORE ALTERATIONS in Remarks section of Attachment 1.
NOTE 5.3.1.8 Surveillance requirements shall be re-done as required prior to resuming CORE ALTERATIONS.
B. VERIFY the surveillance requirement completion on Refueling Daily Log of HC.OP-DL.ZZ-0026(Q), Surveillance Log, are current.
C. IF CORE ALTERATIONS are suspended due to required maintenance on Refuel Position Interlocks, THEN PERFORM the applicable Section(s) of HC.OP-ST.KE-O001(Q), Refuel Interlock Operability Test, as a retest, to ensure ititerlock operability PRIOR to resumingOORE ALTERATIONS AND RECORD thetImne, date, AND ENTER initiil-s-n Attachment 4 to indicate when this surveillance requirenoft is satisfied. IT/S 4.9.1.31 D. WHEN ready for resption of CORE ALTERATIONS, THEN REQUEST CRS to complete items I & 2 of Attachment 4.
Continued Next Page Hope Creek Page 19*1f 40 Rev. 24
ALL ACYIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 5.3.1 (Continued)
CAUTION 5.3.1.E A. If the original TECH SPEC I UFSAR surveillance requirements for starting CORE ALTERATIONS are maintained current during the time when actual CORE ALTERATIONS are NOT taking place, then the requirement to perform the surveillance PRIOR to CORE ALTERATIONS is redundant and Is NOT necessary for resumption of CORE ALTERATIONS.
B. If the TECH SPEC I UFSAR surveillance requirements have NOT been I maintained during the period of time when CORE ALTERATIONS are NOT actually occurring, then the "Prior to CORE ALTERATIONS" surveillance requirement Is necessary to comply with the associated TECH SPEC I UFSAR requirement.
E. PERFORM the following as applicable AND RECORD time, date, AND ENTER initials on Attachment 4 as applicable to indicate when surveillance requirement is satisfied: [TIS 4.9.1.1.a.21
- 1. IF required (If applicable within last 7 days before resumption of CORE ALTERATIONS),
THEN DEMONSTRATE the Source Range Monitors are OPERABLE by requesting I&C to complete the following as required: [T/S 4.9.2.b.21 HC.IC-FT.SE-000 I(Q), NUC Instrument System Source Range Monitor HC.IC-ITSE-0002(Q), NUC Instrument System Source R4pge Monitor
- HC.IC.zFTSE-0003(Q), NUC Instrument System Source Rahge Monitor HC.IC-FT.SE-0004(Q), NUC Instrument System Source Range Monitor HC.IC-*TUSE-0025(Q), NIS SRM Fuel Loading Non Coincident Trips Continued Next Page Hope Creek Page 20 of 40 Rev. 24
ALLACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20-020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 5.3.1 .E (Continued)
- 2. ENSURE that at least 22 feet 2 inches of water is being maintained over the Reactor Vessel flange (within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) [T/S 4.9.8]
- 3. VERIFY that the Reactor Mode Switch is locked (key removed) in the SHUTDOWN OR REFUEL position (within the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, unless Mode Switch was unlocked, then within the last 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
[T/S 4.9.1.1.a.2, 4.9.1.1.bJ
- 4. VERIFY Reactor Protection System shorting links are removed OR adequate Shutdown Margin has been demonstrated IAW T/S 3.1.1 (within the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) [T/S 4.9.2.dl
- 5. VERIFY the following PRIOR to the start of removal of a single Control Rod, OR the associated Control Rod Drive Mechanism from the Core OR Reactor Pressure Vessel (within the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />):
[T/S 4.9.10.1]
- a. The Reactor Mode Switch is OPERABLE LAW Attachment 2 of HC.OP-DL.ZZ-0026(Q),
Surveillance Log. [CD-404E]
Surveillance Log. [CD-404E]
- c. The SHUTDOWN MARGIN requirements IAW Attachment 2 of HC.OP-DL.ZZ-0026(Q),
Surveillance Log. ICD404E]
Continued Next Page Hope Creek Page 21 of 40 Rev. 24
ALL, ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 26020322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 5.3.1.E.5 (Continued)
- d. All other control rods in a five-by-five array centered on the Control Rod being removed are inserted AND electrically OR hydraulically disarmed OR the four Fuel Assemblies surrounding the Control Rod OR Control Rod Drive Mechanism to be removed from the Core AND/OR Reactor Vessel are removed from the core cell.
- e. All other control rods are fully inserted.
- f. All fuel loading operations are suspended.
g Restricted Core Alteration Forms provided by procedure HC.RE-FR.ZZ-000l(Q) shall be available to enhance administrative control during Control Rod withdrawal during refueling.
- h. The Reactor Mode Switch is Locked (key removed) in the SHUTDOWN OR REFUEL position.
- 6. VERIFY the following prior to the start of removal of any number of Control Rods, OR the associated Control Rod Drive Mechanisms from the Core OR Reactor Pressure Vessel (within the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />): IT/S 4.9.10.2.1l
- a. The Reactor Mode Switch is OPERABLE JAW Attachment 2 of HC.OP-DL.ZZ-0026(Q),
Surveillance Log. [CD-404E]
Surveillance Log. [CD-404EJ
- c. The SHUTDOWN MARGIN requirements IAW Attachment 2 of HC.OP-DL.ZZ-0026(Q),
Surveillance Log. [CD-404EJ Continued Next Page Hope Creek Page 22 of 40 Rev. 24
ALL AC'TIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2C0"20322 PSEG Internal Use Only HC.OP-IO.ZZ-0009(Q) 5.3.1.E.6 (Continued)
- d. All other control rods are either inserted OR have the surrounding four Fuel Assemblies removed from the core cell. [T/S 4.9.10.2.1.d]
- e. The four Fuel Assemblies surrounding each Control Rod and/or Control Rod Drive Mechanism to be removed from the core and/or Reactor Vessel are removed from the core cell. jT/S 4.9.10.2.1.e]
f All fuel loading operations are suspended.
g Restricted Core Alteration Forms provided by procedure HC.RE-FR.ZZ-000 1(Q) shall be available to enhance administrative control during Control Rod withdrawal during refueling.
- h. The Reactor Mode Switch is Locked (key removed) in the SHUTDOWN OR REFUEL position.
(within the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) IT/S 4.9.2.cl
- 8. VERIFY the Reactor has been subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as indicated by the date AND time when all Control Rods were fully inserted as recorded in Attachment 1.
(within the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) [UFSAR 9.1.4.2.12.11
- 9. VERIFY all Control Rods are fully inserted, OR the withdrawal of one Control Rod under the control of the Reactor Mode Switch REFUEL position one-rod-out interlock may be withdrawn.
(within last 2 -hours for starting, within last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for continuation) IT/S 4.9.3.a & bl Continued Next Page Hope Creek Page 23 of 40 Rev. 24
HC.OP-AB.ZZ q101(Q)
APPROVED: ...- ýý. -,o:_ A . a?//
e/I Operdtions Manage* Date CATEGORY II IRRADIATED FUEL DAMAGE 1.0 SYMPTOMS 17 1.1 Alarms A. REFUELING FL AIRBORNE ACTIVITY HI B. R B AIRBORNE ACTIVITY HI C. RADIATION MONITORING ALARM/TRBL D. NEW FUEL CRITICALITY RAD HI 2.0 AUTOMATIC ACTIONS 2.1 Reactor Building Ventilation System Isolation on the following:
Reactor Building Exh Hi Rad (1.0 X 10-3 uci/cc)
Refuel Floor Exh Hi Rad (2.0 X 10-3 uci/cc) 2.2 Filtration, Recirculation and Ventilation System (FRVS) automatic start.
3.0 IMMEDIATE OPERATOR ACTIONS 3.1 SUSPEND all refueling operations.
3.2 ENSURE all appropriate automatic actions are complete.
Hope Creek Page I of 2 Rev. 3
HC.OP-AB.ZZ-0101(Q) 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 ENSURE all appropriate immediate operator actions are complete.
4.2 EVACUATE all unnecessary personnel from the Reactor Building.
4.3 ENSURE that secondary containment is in effect.
4.4 DIRECT the Radiation Protection Department to take air samples AND control access to the reactor building and refuel floor, if necessary.
4.5 DETERMINE the FRVS release rate AND ACTIVATE the appropriate emergency plan.
5.0 DISCUSSION 5.1 A damaged fuel assembly attached to the fuel handling grapple should be set down in the fuel pool storage area OR isolated in the defective fuel storage container if a high area radiation condition does not exist.
5.2 A failure of fuel cladding during refueling operations will release gaseous fission products to the reactor building. The severity of this accident will depend upon the exposure history of the fuel bundles.
5.3 Operation of the spent fuel pool cooling system, with irradiated fuel damage, can result in increased radiation levels in the spent fuel pool cooling piping.
Hope Creek Page 2 of 2 Rev. 3
ALL ACFihVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 Page 1 of I PSE&G NUCLEAR BUSINESS UNIT NC.NA-AP.ZZ-0049(Q) - REV. 1 CONDUCT OF FUEL HANDLING SPONSOR ORGANIZATION: Hope Creek Reactor Engineering REVISION
SUMMARY
- Biennial Review performed Yes X No This is an editorial revision. Minor editorial, reference, cross-reference and organizational changes that reflect present processes are not shown with revision bars.
Editorial changes shown with revision bars are:
- 1. Sections 3.13, 5.1.1, 5.1.4.B, 5.2.2.B.4 and 5.2.4.D, added the following "or electronic equivalent". This was added to clarify that an electronic equivalent, e.g., software program, may be used in lieu of the fuel tag board. (CR 951216070 CRCA 2)
- 2. Section 3.3, deleted former third bullet, unnecessary reference to Radiation Protection Technician responsibilities.
- 3. Deleted former Section 3.6.4, these duplicated those in Section 3.5.
- 4. Section 3.6, deleted responsibility specific to Technical Specifications adherence. This is not required to be stated, this is a requirement of the license.
- 5. Section 3.18, deleted responsibility for revising procedures. This does not have to be stated.
- 6. Section 8.0, added cross-references: NC.NA-AP.ZZ-0000(Q), Action Request Process; NC.NA-AP.ZZ-0003(Q), Document Management Program.
- 7. Section 8.0, deleted cross-references not applicable to the procedure: NC.NA-AP.ZZ 0006(Q), Incident Report/Reportable Event Program and Quality/Safety Concerns Reporting System; NC.NA-AP.ZZ-001 1(Q), Records Management Program; NC.NA AP.ZZ-0020(Q), Nonconformance Program and NC.NA-AP.ZZ-0032(Q), Preparation, Review and Approval of Procedures.
IMPLEMENTATION REQUIREMENTS: Effective Date: 94 s /9 7 APPROVED: __________________
Manager - Hope Creek System Engineering Date APPROVED: v.
Gpera an r pe/Creek Operations [5at i(
APPROVED: _
General Manage - Salem Operations I D te
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q)
CONDUCT OF FUEL HANDLING TABLE OF CONTENTS Section Title 1.0 PURPOSE ................................................................................................ 2 2 .0 S CO P E .................................................................................................. ..2 3.0 RESPONSIBILITIES .............................................................................. 2 4.0 PROCESS DESCRIPTION ..................................................................... 6 5.0 PROCEDURE ......................................................................................... 6 5.1 Fuel Handling Not Involving Core Alterations ............................... 6 5.2 Fuel Handling Involving Core Alterations ..................................... 8 5.3 General Housekeeping Considerations ....................................... 12 6 .0 R E C O R D S ................................................................................................ 13 7.0 DEFINITIONS ...................................................................................... 13
8.0 REFERENCES
..................................................................................... 13 Nuclear Common Page 1 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 1.0 PURPOSE Establish the responsibilities and requirements necessary to ensure Fuel Handling is conducted in a safe and efficient manner. [CD-644A]
2.0 SCOPE 2.1 This procedure is applicable to Fuel Handling activities performed on the refuel floor and the Spent Fuel Handling Building (Salem) except as noted in Section 2.2.
2.2 New fuel receipt is outside the scope of this procedure. New fuel receipt is defined as the time fuel arrives at the plant security gate until its initial placement into the spent fuel pool.
3.0 RESPONSIBILITIES 3.1 Operations Manager is responsible for:
"a Ensuring that Refueling SROs are qualified in accordance with NC.NA-AP.ZZ-0014(Q),
Training, Qualification, and Certification (NAP-14). [CD-168A]
"* Hope Creek, ensuring that Fuel Handling Operators are qualified in accordance with NAP-14. [CD-168A]
"* Hope Creek, ensuring that approved procedures are available for Refueling Bridge Operation.
"* Salem, ensuring that Fuel Handling Operators, not involving core alterations, are qualified in accordance with NAP-14.
" Ensuring approved procedures are available for coordination of refueling operations involving Core Alterations and that these procedures contain prerequisites, documentation of operation's surveillance and log readings for Core Alterations.
3.2 General Manager - Nuclear Maintenance is responsible for ensuring maintenance is performed on all equipment related to Fuel Handling activities.
3,3 Superintendent. Radiation Protection is responsible for:
- Hope Creek, controlling access to the upper elevations of the drywell during irradiated Fuel Handling in the reactor vessel. [CD-612X]
0 Salem, controlling access to the transfer tube area during irradiated Fuel Handling in the reactor vessel.
Nuclear Common Page 2 of 13 Rev. I
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 3,4 Manager - System Engineering is responsible for:
"* Ensuring that Reactor Engineers assigned to the Fuel Handling crews are qualified in accordance with NAP-14.
"* At Salem, ensuring that Fuel Handling Operators, involved in core alterations, are qualified in accordance with NAP-14. [CD-168A]
" Ensuring procedures are available for:
=> Fuel Transfer Documents
=> Fuel Channeling and Dechanneling (Hope Creek)
= Local Power Range Monitors (LPRM) removal and installation (Hope Creek)
=> Control Blade removal and installation (Hope Creek)
=> Manipulator Crane Operation (Salem)
=> Insert change outs (Salem)
= Fuel Transfer System (Salem) 3.5 Supervisor - Reactor Engineering is responsible for:
"* Preparing fuel transfer documents.
"* Preparing necessary documentation for LPRM and Control Blade removal and installation. (Hope Creek)
"* Monitoring all Fuel Handling and associated activities
"* Verifying proper location and orientation of all fuel assemblies, as required.
"* Maintaining accurate records of the location history for nuclear fuel and core components.
" Providing and maintaining the underwater cameras, video monitors, video recorders, cables and video control equipment used for Fuel Handling activities.
" Prior to the refueling outage, conducting "pre-refueling training and briefing" meetings with personnel involved with the Fuel Handling, Core Alterations, reactor pressure vessel work and in-vessel work. Meeting agenda should include: [CD-827D]
= General outline of the activities to be performed.
SAny unique activities to be performed.
SGeneral practices of the refuel floor.
= Actions during an emergency situation.
z Potential emergency situation and a review of past industry incidents.
Nuclear Common Page 3 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 3.6 Refuel Senior Reactor Operator (SRO) is responsible for:
" Supervising and controlling activities related to Fuel Handling involving Core Alterations. [CD-217B, CD-897E, CD-168A]
" Assisting the Fuel Handling Coordinator in resolving problems that halted Fuel Handling.
" Directing Refueling Bridge (Hope Creek) or Manipulator Crane (Salem) operations during Fuel Handling involving Core Alterations in the sequence defined by Reactor Engineering.
"* Ensuring compliance with directions from Radiation Protection personnel to minimize the spread of radioactive contamination and reduce personnel radiation exposure. [CD-897E]
"* Initiating immediate actions during abnormal events, such as irradiated fuel damage, loss of fuel pool inventory or refuel floor high radiation conditions.
"* Suspending Fuel Handling, as he deems necessary or as required by procedure. (Refer to Section 5.2.5)
" Controlling access to the Refuel Bridge (Hope Creek) or Manipulator Crane (Salem) to ensure safety and minimize potential personnel exposure. [CD-528A]
" Determining whether activities should be suspended if problems or erratic functioning of refueling tools or cranes exist. [CD-422A]
3.7 Refueling Bridge Operator (Hope Creek), Manipulator Crane Operator (Salem),
and Fuel Handling Operator (Salem)
"* Performing nuclear fuel moves in accordance with the sequence for fuel transfer provided by Reactor Engineering.
"* Informing the Refuel SRO of any problems or erratic functioning of the refueling tools or cranes. [CD-422A]
3.8 Control Room Refuel Monitor is responsible for:
"* Communicating with the refuel floor personnel during Fuel Handling involving Core Alterations and related activities.
"* Monitoring the Nuclear Instrumentation during fuel movement and informing the refuel floor personnel of any unexpected increasing count rate.
" Updating the fuel tag boards or electronic equivalent in the Control Room.
Nuclear Common Page 4 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 3.9 Fuel Handling Coordinator is responsible for:
"* Ensuring that work performed on the refuel floor meets the requirements of NC. NA-AP.ZZ-0031 (Q), Inspection/Housekeeping Program (NAP-31).
"* Reviewing refuel floor schedule to ensure required tools, equipment, materials, etc. are in place to support upcoming activities.
"* Attending daily outage meetings and report on refuel floor activities.
" Ensuring that Work Orders associated with Fuel Handling activities are performed in accordance with NC.NA-AP.ZZ-0009(Q), Work Control Process (NAP-9).
" When equipment or tools require maintenance during Fuel Handling activities, ensuring that the repair is being performed expeditiously and documented, as required.
3.10 Radiation Protection Technician is responsible for:
"* Sampling the refueling floor and the 130" el. of the Fuel Handling building (Salem) air for the detection of airborne radioactive contamination.
" Directing personnel to prevent over-exposure in the event that radiological conditions change on the refuel floor or the Fuel Handling building.
" Advising Fuel Handling personnel of any changes in radiological conditions affecting the task in progress.
"* Monitoring Fuel Handling floor activities to ensure that personnel overexposure, contamination or ingestion of radioactive materials does not occur.
"* Performing radiation surveys of all tools and equipment to be removed from the reactor cavity, transfer pool (Salem) or spent fuel pool. [CD-217B, CD-897E]
3.11 Spotter is responsible for aiding the operator in the movement or placement of fuel in the reactor vessel or the spent fuel pool.
3.12 Upender Operator in containment is responsible for performing upending operations in containment in accordance with Reactor Engineering procedures. (Salem) 3.13 Upender Operator in the Fuel Handling building is responsible for (Salem):
"* Performing upending operations in the Fuel Handling building in accordance with Reactor Engineering procedures.
"* Aiding the Fuel Handling operator in the movement or placement of fuel in the spent fuel pool.
Nuclear Common Page 5 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 3.14 Core Physics Monitor is responsible for performing Inverse Count Rate Ratio plots during core reload and core shuffle. (Salem only) 3.15 Gate Valve Operator is responsible for closing the Transfer Canal Isolation Valve when directed by the Refuel SRO. (Salem only) 3.16 Tool Control Monitor is responsible for assuring personnel/material accountability in the reactor cavity and associated areas.
3.17 Department Managers are responsible for ensuring that the overtime guidelines specified in NC.NA-AP.ZZ-0005(Q), Station Operating Practices, are utilized for Fuel Handling Crew personnel, under their direction.
3.18 Manager - Quality Assessment is responsible for specifying QA hold or notification points in Fuel Handling procedures.
4.0 PROCESS DESCRIPTION None.
5.0 PROCEDURE 5.1 Fuel Handling Not Involving Core Alterations 5.1.1 Prior to commencing Fuel Handling, not involving core alterations, Reactor Engineering shall:
"* Initiate an Action Request in accordance with NC.NA-AP.ZZ-OOO0(Q), Action Request Process (NAP-0) for the required work activities.
"* Ensure that the system operability and appropriate Technical Specification requirements for irradiated Fuel Handling have been satisfied.
" Ensure that appropriate documents are prepared depicting the transfer of fuel.
A set of approved documents is required on the refuel floor (Hope Creek) or Fuel Handling building (Salem).
"* Verify that the Control Room fuel tag boards or electronic equivalent reflect the location of nuclear fuel at the applicable unit.
"* Obtain permission from the Operations Superintendent/Control Room Supervisor (OS/CRS) to commence Fuel Handling Activities.
"* Notify the Radiation Protection Department prior to the commencement of Fuel Handling.
"* Brief the Fuel Handling crew (non-core alterations).
Nuclear Common Page 6 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 5.1.2 Fuel Handling Crew for non-core alterations Fuel Handling activities.
A. The minimum crew for non-core alterations Fuel Handling activities is:
"* Fuel Crane Operator
"* Radiation Protection Technician
"* Reactor Engineer
"* Spotter B. The Reactor Engineer may fulfill the duties of the spotter.
C. All members of this crew shall be on the refueling floor (Hope Creek) or in the Fuel Handling building (Salem) during Fuel Handling activities.
5.1.3 The transfer of nuclear fuel shall be performed as described below.
A. Fuel movements shall be performed in accordance with approved procedures.
B. The sequence of fuel movement shall be performed in accordance with the fuel transfer documents.
C. Following the completion of fuel moves, the Fuel Crane Operator and the spotter shall sign the completed fuel move on the fuel transfer documents.
D. Hope Creek, any required fuel dechanneling or channeling operations is performed in the fuel preparation machines by the Operations Department.
E. The removal of tools or equipment stored underwater in the spent fuel pool, cask storage pit (Hope Creek) or transfer pool (Salem) shall not to be allowed without the prior knowledge and consent of the Radiation Protection Technician. The Radiation Protection Technician shall be present for these activities. [CD-217B, CD-897E]
F. All material nonconformances identified during the performance of Fuel Handling shall be documented in accordance with NAP-O.
5.1.4 Upon completion of Fuel Handling activities, the Reactor Engineer shall:
A. Notify the OS/CRS and the Radiation Protection Department.
B. Ensure that the Control Room fuel tag boards or electronic equivalent are updated to reflect the location of nuclear fuel at the applicable unit.
Nuclear Common Page 7 of 13 Rev. 1
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2D0020322 NC.NA-AP.ZZ-0049(Q) 5.2 Fuel Handlinq Involving Core Alterations 5.2.1 Refueling operations involving Core Alterations are coordinated through department procedures which provide the following:
A. Prerequisite signatures for the initiation of Core Alterations including system operability requirements. (Operations)
B. Documentation governing the resumption of Core Alterations. (Operations)
C. Refueling logs or other means necessary to document periodic Refueling surveillance items. (Departments which are responsible for Fuel Handling related Surveillances.)
5.2.2 Prior to commencing Fuel Handling, involving core alterations, the following shall be completed:
A. The Refueling SRO shall request permission from the OS/CRS and notify the Radiation Protection Department.
B. Reactor Engineering shall:
- 1. Prepare the appropriate documents depicting the transfer of fuel.
- 2. Prepare the necessary documentation to direct the performance of the following tasks:
"* LPRM removal and installation (Hope Creek)
"* Control blade removal and installation (Hope Creek)
"* Fuel Channeling and De-channeling (Hope Creek)
"* Insert change outs (Salem)
- 3. Ensure that copies of the documents for steps 1 and 2 above are available at the following locations:
"* Control Room
"* Refuel floor
"* Fuel Handling Building (Salem)
Nuclear Common Page 8 of 13 Rev. 1
ALL AC'TIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 5.2.2 (Continued)
- 4. Verify that the Control Room fuel tag boards or electronic equivalent are reflect the location of nuclear fuel on the refuel floor and the Fuel Handling building (Salem).
- 5. Conduct pre-refueling training and briefing.
C. The Fuel Handling Coordinator shall ensure that Work Orders are prepared for the required work activities.
D. The Refueling SRO shall direct Radiation Protection Technicians to establish restricted access to the upper regions of the drywell (Hope Creek).
5.2.3 Fuel Handling Crew for Fuel Handling activities involving core alterations.
A. The following is the minimum crew for Fuel Handling activities involving core alterations. Those designated with an asterisk (*), shall be on the refueling floor during Fuel Handling activities. [CD-168A]
"* *Refueling SRO
"* *Refueling Bridge Operator (Hope creek)
"* *Manipulator Crane Operator (Salem)
"* *InvesselSpotter
"* *Radiation Protection Technician
"* Reactor Engineer
"* Upender Operator in containment (Salem only)
"* Fuel Handling Building Crane Operator (Salem only)
"* Radiation Protection technician in the Fuel Handling Building (Salem only)
"* Upender Operator in Fuel Handling Building (Salem only)
- Control Room Refuel Monitor
- Core Physics Monitor (Salem only during core reload or shuffle))
- Fuel Handling Coordinator
- Gate Valve Operator
- Tool Control Monitor B. The Refueling SRO or the Reactor Engineer may fulfill the duties of the Invessel Spotter. This shall be determined by the Refueling SRO.
C. The Invessel Spotter or the Reactor Engineer may fulfill the duties of the Fuel Handling Coordinator.
Nuclear Common Page 9 of 13 Rev. 1
ALL AC*TIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2-0020322 NC.NA-AP.ZZ-0049(Q) 5.2.4 All Fuel Handling involving core alterations shall be performed as described below.
A. Communications shall be established and maintained between the Control Room Refuel Monitor located in the Control Room and a member of the Fuel Handling Crew on the refuel floor. [HC & S TIS 3.9.51
"* Salem, communications is also required with the Fuel Handling Building.
"* Hope Creek, it may be necessary to maintain communications with the undervessel area.
B. All fuel movement shall be performed by approved Fuel Handling procedures.
C. The sequence of fuel movement shall be conducted in accordance with the approved fuel transfer documents.
D. Following the completion of individual fuel moves, the following shall be performed:
" The Refueling Bridge Operator (Hope Creek), Manipulator Crane Operator (Salem) and the Refueling SRO (or designee) complete all required signatures on the associated fuel transfer document. The spotter may sign for the operator.
" The Control Room Refuel Monitor located in the Control Room signs the associated Control Room document (independent verification of Control Room documentation is not applicable).
"* (Salem) The spotter, in the Fuel Handling Building, signs the associated fuel transfer document.
"* Control Room fuel tag boards or electronic equivalent are updated by the Control Room Refuel Monitor.
E. (Hope Creek) Any required fuel dechanneling or channeling operations are performed in the fuel preparation machines by the Operations department per appropriate Reactor Engineering procedures.
F. The removal of tools or equipment stored underwater in the reactor, spent fuel pool, cask storage pit or transfer canal shall not be performed without the prior knowledge and consent of the Radiation Protection Technician. The Radiation Protection Technician shall be present for these activities.
[CD-217B, CD-897E]
Nuclear Common Page 10 of 13 Rev. 1
ALL A4CTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 5.2.6 If a fuel assembly is found in an incorrect core location, the following actions shall be taken:
A. Ifthe error is found during a fuel move, return the latched fuel assembly to its original location.
B. Contact the Reactor Engineer to prepare modified fuel transfer documents to put the core in an analyzed configuration, with respect to shutdown margin.
Then move the fuel assembly to the analyzed configuration.
C. Terminate Fuel Handling and Core Alterations.
D. Initiate an Action Request to evaluate the consequences of the Fuel loading error in accordance with NAP-0.
E. If a personnel error in the fuel movement occurred, the event shall be reviewed by the Operations Manager. This should involve a review of the refueling qualifications of the Refuel Platform Operator (Hope Creek),
Manipulator Crane Operator (Salem) and Refueling SRO (and any Spotter and/or Designee) on the refuel platform at the time.
F. Recovery Fuel Transfer Documents are prepared and provided by the Reactor Engineer.
G. Fuel Handling shall not recommence without the approval of the Operations Manager.
5.2.7 Upon the completion of Fuel Handling involving Core Alterations, Reactor Engineering shall verify the reactor core fuel locations and orientations.
(Hope Creek) This should be performed prior to any control rod motion (in that particular cell) once reactor fuel loading operations commence.
5.3 General Housekeeping Considerations 5.3.1 The Fuel Handling Coordinator is responsible for housekeeping and cleanliness on the refuel floor during and following the completion of Refueling activities until operational housekeeping is restored.
5.3.2 Provisions for personnel/material accountability established in accordance with NAP-31 are maintained on the refuel floor and Fuel Handling Building during the performance of Fuel Handling.
5.3.3 The temporary suspension of tools and equipment from the refueling platform or manipulator crane and temporary placement of such items within the reactor cavity should only be allowed if such items are required for work presently in process.
Nuclear Common Page 12 of 13 Rev. 1
ALL A4TIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 NC.NA-AP.ZZ-0049(Q) 5.3.4 Additional groups involved in Fuel Handling such as ISi personnel or fuel vendor representatives, shall ensure that refuel floor housekeeping and cleanliness is maintained during the performance of work on the refuel floor.
5.3.5 Special care shall be maintained to avoid the contamination of completed Fuel Transfer Documents and documentation of other activities located on the refuel floor.
6.0 RECORDS 6.1 The master transfer documents shall be retained in accordance with NC.NA-AP.ZZ-0003(Q), Document Management Program.
6.2 Completed Work Orders shall be retained in accordance with NAP-9.
6.3 Qualification records shall be retained in accordance with NAP-14.
6.4 Personnel/Material Accountability Control Logs shall be retained in accordance with NAP-31.
7.0 DEFINITIONS 7.1 Core Alterations - This definition is specific to Salem or Hope Creek. See the applicable Technical Specification.
7.2 Suspension of Fuel Handling (or Core Alterations) - A formal declaration by the Refuel SRO or a condition required by Technical Specifications.
8.0 REFERENCES
8.1 UFSAR, Hope Creek Generating station, Chapter 9.
8.2 UFSAR, Salem Generating station, Chapter 9.
8.3 Hope Creek and Salem Technical Specifications.
8.4 Cross References 8.4.1 NC.NA-AP.ZZ-0000(Q), Action Request Process 8.4.2 NC.NA-AP.ZZ-0003(Q), Document Management Program 8.4.3 NC.NA-AP.ZZ-0005(Q), Station Operating Practices 8.4.4 NC.NA-AP.ZZ-0009(Q), Work Control Process 8.4.5 NC. NA-AP.ZZ-0014(Q), Training, Qualification and Certification 8.4.6 NC.NA-AP.ZZ-0031(Q), Inspection/Housekeeping Program 8.5 Closing Documents CD-168A (NRC Circular 80-21) CD-217B (INPO O&MR 111)
CD-442A (INPO SER 59-81) CD-827D (INPO SOER 84-01R06)
CD-528A (INPO SER 43-82) CD-897E (INPO SER 12-87)
CD-644A (INPO O&MR 65) CD-612X (Hope Creek UFSAR F01-0091-01)
Nuclear Common Page 13 of 13 Rev. 1
An operator has the following exposure history this year until today:
Deep Dose Equivalent (DDE) 210 mrem Committed Effective Dose Equivalent (CEDE) 45 mrem Shallow Dose Equivalent (SDE) 33 mrem Today, the operator was required to make two entries into the Drywell at 5 percent reactor power:
,Entry 1: Gamma dose: 52 mrem; Neutron dose: 24 mrem
,Entry 2: Gamma dose: 124 mrem; Neutron dose: 54 mrem How much radiation exposure is available to the operator without extension if he has to make additional entries?
ýHis available Non-Emergency margin for the year is...
.1488 mrem S1521 mrem W 1599 mrem A 1712 mremr Answitir bee Ho0pe c-reek FaciBli[omreenioty 03 /12/2 002 Tier i d
,Generic Knowledge and Abilities 1
........ c o r.Nu m b r.12
}U[1294001 G301
.Re 0 i2.3 Radiological Controls 2.3.1 [Kolde of 10 CFR 20 and related facility radiation control requirements. 1. 30 Ex* CORRECT ANSWER. Gamma and neutron dose are summed for DDE. DDE and CEDE are summed
.together to obtain TEDE. The Dose limit without extension is 2000 mrem/year TEDE Reerne Title
- NC.NA-AP.ZZ-0024 000113E059 a. Identify the personnel responsible for approval of the following dose extension:
Yearly Dose Extension Declared Pregnant Women Dose Extension Lifetime Dose Extension t 0 INPO Exam Bank ]uI i onIMeo 'Significantly Modified
..SouPrEA*
.... .M- BANK QUESTION IDw#3324. Braidwood 1 09/14/1998 Saturday, March 23, 2002 8:34:49 AM Page 129 of 139
An operator has the following exposure history this year until today:
Deep Dose Equivalent (DDE) 210 mrem Committed Effective Dose Equivalent (CEDE) 45 mrem Shallow Dose Equivalent (SDE) 33 mrem Today, the operator was required to make two entries into the Drywell at 5 percent reactor power:
Entry 1: Gamma dose: 52 mrem; Neutron dose: .4 mrem
.Entry 2: Gamma dose: 124 mrem; Neutron dose: /54"rrem 2., r*"",
How much radiation exposure is available to the operator without extension if he has to make additional entries?
- His!Ha.v available ....
_E ...
~ Non-Emergency
.s._a !..... _!
.... e .....
g_e_n.
I.... m ~ ....
margin f.for the y
... yeara r is ... ...........
is........ . ..... ..... ........ . ......... ....... .. .. ...... ...
R91488 mrem I]:1521 mrem S:1599 mrem 1712 mrem S b1
- B iComprehesioni*1pe creek 03/12/2002 eKnowle dge andAbl e ... 294001G301A
ýGENERIC ............. ........... .... ~i
~
11 Radiological Controls 1 nowledge of 10 cFR 20 and related facility radiation control requirements.
[K2.3 2.6. 3:0 CORRECT ANSWER. Gamma and neutron dose are summed for DDE. DDE and CEDE are summed together to obtain TEDE. The Dose limit without extension is 2000 mrem/year TEDE SNC.NA-AP.ZZ-0024 0001113E059 a. Identify the personnel responsible for approval of the following dose extension:
Yearly Dose Extension Declared Pregnant Women Dose Extension Lifetime Dose Extension 0,_14i-41 IQ ýM INPO Exam Bank flMelotiii cV ld:iSignificantly Modified m.ents:;.. I INPO EXAM BANK QUESTION ID #3324. Braidwood 1 09/14/1998 Saturday, March 23, 2002 8:34:49 AM Page 129 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200ý0323 NC.NA-AP.ZZ-0024(Q) 7.21 Public Dose - Dose received by a member of the public from station operations, or from another source under the control of PSEG.
It dose or dose received from background radiation, does not include occupational as a patient from medical practices, or as a medical research volunteer.
7.22 Radiation Area - Any area accessible to personnel with radiation dose rates that exceed 5 mrem/hour DDE at 30 cm from any 7.23 source.
Radiation Work Permit - A document used to provide workers with instructions for radiological controls associated with work in the RCA.
7.24 Radiation Worker - A worker who has completed Radiation Worker Training.
7.25 Radioactive Sources - Radioactive material used or stored for calibrating or testing station installed or portable instrumentation.
7.26 Radioactive Waste - Licensed radioactive material that has been determined to be no longer useful and that requires disposal.
7.27 Radiological Effluent Technical Specifications (RETSIODCM) - Specifications contained in the Technical Specifications or the Offsite Dose Calculations Manual.
7.28 Radiologically Controlled Area(RCA) - An area within the Restricted Area that has the potential for significant personnel radiation control over personnel access. The term "Radiological exposure and has positive Control Area" is synonymous.
7.29 Restricted Area - Any area where access is controlled by the licensee to protect individuals from undue risks from exposure to radiation and radioactive materials.
The Restricted Area is normally the area within the PSEG security fence (the Protected Area).
7.30 Self Monitor - An individual trained in radiation protection procedures who is not a member of the Radiation Protection staff, but protection instrumentation for personal radiationwho is qualified to use radiation protection in High Radiation Areas for observation, sampling, or tours.
7.31' Station ALARA Committee (SAC) - The committee coordination of the ALARA program within the responsible for overall station. The committee meets at least semi-yearly to review the status of the ALARA Program and is composed of individuals from the major functional departments 7.32 of the station.
(Radiological)S tIop Work Order - A directive halting all work within a specified area, or associated with a specified activity, or performed by a specified group. The directive prevents unplanned radiation exposure or loss of control of radioactive material. The order can be lifted only by Radiation Protection Manager.
7.33 .° !gvalkt1 TnE - The sum of the external whole ose tU) body nd the internal whole 1ody dose (MDE).
7.34 Unrestricted Area - The area outside of the Restricted Area (beyond the "owner controlled area").
Nuclear Common Page 29 of 40 Rev. 11 K,1
N>
or->
0>
NC.NA-AP.ZZ-0024(Q) 1iC) z 0
I ATTACHMENT I ADMINISTRATIVE DOSE CONTROL LEVELS AND (I)
Page I of 2 EXTENSION REQUIREMENTS 0
Control Level Whole Body Dose Control Levels - TEDE Description Action atControl Level M
. u..rrent year dose control level Increase Approval C/)
Dose control level may be increased to Radiation Protection Supervisor S~~3000 3000 mrem/year TEDE mremea z
C Extended current year dose control level. Dose control level may be increased to Radiation Protection Manager co 4000 mrem/year TEDE Final current year dose control level (may not Incremental increase up to 4750 mrem. Vice President - Operations M beexceededin non-emergency situations).
WholeBody Dose Control Levelfor theDeclared Pregnant Woman (DPW)
Dose Control Level DescriptionAction 50 mrem/month atContro Level Increase Mosthly dose control level fo DWs.
or less TEDE Total dose for the gestation period Radiation Protection Manager should not exceed 450 mrem TEDE.
Control level should not be increased beyond 50 mrem in any month unless absolutely necessar.
Internal DoseMonitorin T ld for the Declar 50 mrem/year CEDE dose monitoring threshold for DPWs: n nt man (DPW)
Mnnitrin fhk 1,1 .Interal R~adiation Protection Manager I!
... ,s~,,=;
? l. aIl.ll nolll t we SConfirmatory monitoring may be provided.- exceeded unless monitoring prog.ram is IConfirmatory monitoring may be provided. xceedd I U3LdUIIIItU unless monitoring uy ,aoiauon I.Jau 31 lOu I BUt program Protection.U is Kaalatlon Protection Manager I
Nuclear Common Page 32 of 40 Rev. 11
Which one of the following describes organizational grouping of Abnormal Operating Procedures (ABs) lAW SH.OP-AP.ZZ-01 02 "Use of Procedures".
L 100 series are operational transient procedures S200 series address component failures H 300 series apply at all times S000 series address fire and medical emergencies Answ#tv cL]evel ne ri.. KnGew.g na Ao ~e ..........
[ ] rO [Memory
........................................ i ;;i op -e c--r H-c k Ie-e
}i e --0.311 -21 2 40 ! 40 2002ý Tier: ,Generic Knowledge and Abilities ] 0[i294001 G40 5 126:
- 2.4 Emergency Procedures and Plan
ý2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and 2.9 3.6 iemergency evolutions.
Explanatioaiigf Jus..t i..fi..c-a ti -on-Answer -lAW SH.OP-AP.ZZ-0102, section 5.5.2 SH.OP-AP.ZZ-0102, section 5.5.2 00011 3E005 a. Summarize the guidelines for the use of the following types of procedures:
Abnormal Operating Procedures Emergency Operating Procedures Alarm Response procedures
, ~~ ~~~~~~~
....... ..... ~~~~~~
ImAterial Rq~i~re*~ ~fr 1,9 I ~~~~~~~~~
iaExam iiy x ona mga k........................... ~~~~~~ .....................
IQuestion SourEI Facility; Exam Bank I lQuestion Modification Method: Direct From Source I
SOrc Comns VISION BANK QID# Q57004 Friday, March 22, 2002 4:53:06 PM P Page 136 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200*2 0322 SH.OP-AP.ZZ-0102(Q) 5.4.3 SO and AOP Interrelationship
"* In general, the AOP is the preferred method of dealing with abnormal or transient conditions, and will normally take precedence over the implementation of normal operating procedures during these situations.
"* The interrelationship is not always clearly defined between the two procedure types and operator skill and knowledge is required to determine this balance. Since every possible situation cannot be addressed in the abnormal operating procedure, the control room supervisor is allowed to exercise good judgment in the concurrent use of normal operating procedures during the implementation of the AOPs. Concurrent use of the normal operating procedure should not impede the progress through the AOPs, or counteract the alignment of the plant in mitigating the transient within the AOP.
5.5 Hope Creek Abnormal Operating Procedures (AB) 5.5.1 Purpose - describes the actions to be taken when important parameters or systems are in jeopardy yet, in most cases, the Reactor Protection System has not yet activated.
5.5.2 Types - the Abnormal Operating Procedures have been divided into two groups:
operational transients and abnormal conditions.
0 100 Series Abnormal Condition addresses system or component failures which pose significant problems to the operator. In addition, these procedures deal with non system related problems which may adversely effect operation (i.e., plant fire, flood, etc.). they are brief and rely on the operator's training and knowledge.
0 200 Series Operational Transient procedures are symptom oriented and deal with plant conditions which input directly into the RPS. The procedures are written to the point of entry into the EOPs. They are not written to cover actual system isolation and scram but rather to prevent these occurrences. They include any immediate steps for mitigating the consequences of the transient and subsequent steps for determining the origin of the event.
The discussion section provides additional information to aid in assessing the transient. In general, the procedure directs the operator to look at key items and activities and relies on the operator's training and systems knowledge for specific component manipulations.
300 Series Reactor Power Oscillations applies at all times. The 100 and 200 series may provide entry conditions for the 300 series.
Salem/Hope Creek Page 31 of 42 Rev. 3
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2OMt20322 SH.OP-AP.ZZ-0102(Q) 5.5.3 Use - guidelines for use of both types of ABs are as follows:
"* All Immediate Actions shall be committed to memory.
"* All expected Automatic Actions shall be verified to have occurred.
"* If a scram condition or other EOP entry condition is met the appropriate EOP shall be entered.
"* The Subsequent Actions shall be performed with the procedure in-hand or at the direction of a person with the procedure in-hand.
The AB, when in use, should be utilized as part ofthe Control Room Log(s) and should be marked in a manner so as to allow for re-creation of the event.
This should be accomplished by writing on the procedure information pertaining to major steps.
5.6 Alarm Response Procedures (AR) 5.6.1 Purpose - directs operator response to an alarm on an overhead panel, control room console, Plant Computer, or local panel in the plant 5.6.2 Types
"* Overhead Alarm Response Procedures - the overhead AR directs the operator in responding to control room overhead alarms. Since many of the overhead ARs represent multiple inputs with a variety of digital or console alarms feeding the alarm, the overhead ARs are prepared as packages containing the overhead summary sheet along with the associated digital and console alarm. However, due to the volume of the packages, the procedures are split in the control room by panel. They are placed within easy reach and provide the operator with quick information on what may have caused the alarm as well as the major concerns associated with the alarm. The digital, console, and non-indicating alarm response procedures provide the operator with more specific information on causes and actions with the alarm.
"* Local Alarm Response Procedures - the local ARs direct operator response to an alarm at a local panel.
Salem/Hope Creek Page 32 of 42 Rev. 3
Which one of the following describes how a scram is verified in accordance with HC.OP-IO.ZZ 0008 Shutdown from Outside the Control Room?
I~IHCU nitrogen pressure verified to be less than 800 psig at each HCU
.]Reactor vessel pressure verified less than 920 psig IiRPS power distribution circuit breakers verified to be open W Scram air header pressure verified to be less than 100 psig ATnsw-e7r, a1 el~ B 1 pgniv Lyevel1 [Memory Pacitlf 'Hope Creek 03/12/2002
.iJ i*~~~~. _. _
Tier IGe ne ric Knowl edge and AbiIi ties ] ~ P IS ~~[1294.001 G434
,.GENERIC-: . . . . . . .e. .129................... 129 2.4 !Emergency Procedures and Plan 2.4.34i Knowledge of RO tasks performed outside the main control room during emergency operations 3.8_3.61 including system geography and system implications.
Expl io The scram is verified outside the control room via HCU Accumulator pressures < 800 psig at each HCU HC. OP-hO.ZZ-0008 00112HE004 (R) Apply Precautions, Limitations and Notes while executing the SHUTDOWN FROM OUTSIDE THE CONTROL ROOM Integrated Operating Procedure.
UeSti Sour Facility Exam Bank [Qbeston Modification Method: Direct From Source vision Exam Bank QID:#Q540018 Friday, March 22, 2002 4:53:56 PM Page 139 of 139
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2P..02032Z 5.0 PROCEDURE NOTE 5.0 A. Each step should be initialed upon completion of the step.
B. Attachment 2 should be referred to for RSP redundant instrumentation/
controls.
C. Attachment 3 should be referred to for placing 'A' Loop RHR in Suppression Pool Cooling.
D. Attachments 6 and 7 should be referred to for plant communications information. When dispatching an operator to a remote shutdown control station, the operator should be provided with a sound-powered phone OR radio to assist with communication.
5.1 Establish Control from Outside the Control Room 5.1.1 ENSURE that all prerequisites have been satisfied IAW Section 2.0 of this procedure.
NOTE 5.1.2 A. IF the Reactor was NOT scrammed AND the MSIVs are still open, then the Feedwater System AND the Main Turbine Bypass Valves may be regulating Rx level AND Rx pressure at this time.
B. Opening the circuit breakers listed in Step 5.1.2 will deenergize the RPS busses, scramming the plant, AND deenergize the NSSSS busses, closing the MSIVs.
C. 10C410(10C41 1) RPS PWR Dist. Panels A(B) are located in Control/DG Bldg.
El. 54'.
5.1.2 IF the Reactor was NOT scrammed prior to Control Room evacuation, THEN OPEN the following circuit breakers: [CD-987X]
A. CB2A, CB3A, CB5A, CB7A AND CB8A (RPS PWR DIST PNL A 10C4 10).
B. CB2B, CB3B, CB5B, CB7B AND CB8B (RPS PWR DIST PNL B 10C41 1).
Hope Creek Page 8 of 70 Rev. 14
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020322 5.1.3 IF the Rx scram was NOT verified prior to evacuating the Control Room, THEN VERIFY Rods Full In. (SPP$XCiDý (TSC)
OR RMCS Activity Control Card( other).)
5.1.4 NOTIFY Chemistry to verify that the Hydrogen/Oxygen System has tripped IAW HC.CH-SO.AX-000 1(Q).
5.1.5 Upon arriving at the RSP, MONITOR the RSP System indications AND CHECK specifically for the following:
A. REACTOR VESSEL PRESSURE PR-7853D (905 - 1045 psig)
CAUTION 5.1.5.8 IF the rate of rise of RPV level indicates HPCI is injecting AND the Control Room is unmanned, THEN HPCI will have to be tripped using Attachment 8 when no longer required OR prior to exceeding the high level trip (Level 8). The high level trip may NOT function in the event a fire occurs in the relay room. [CD-012Z3 B. REACTOR VESSEL LEVEL LR-7854 (12.5 - 54")
C. RCIC System status (standby OR auto-initiated)
D. PSV-F013F,H,M SRV status (standby OR cycling open/closed)
E. SUPPRESSION CHAMBER WATER TR-3647J (AND M) (average less than 95'F)
F. DIESEL GENERATOR 1A(B,C,D)G400 TRIP/CLOSED Status (closed IF a loss of offsite power has occurred).
5.1.6 IF a loss of offsite power has occurred, THEN SEND an operator to the Diesel Generator Remote Control Panel (Aux. Bldg El. 130') to monitor Diesel Generator operation, AND IMPLEMENT HC.OP-AB.ZZ-0135(Q), Loss of Offsite Power, concurrent with this procedure.
Hope Creek Page 9 of 70 Rev,. 14
I-b PSE&G NUCLEAR-I 6103375320;# 1
-a -
To: From:
NAME NAME PHONE COMPANY PSEG Nuclear, LLC DEPARTMENT Nuclear Training Department W'Nu4cear Training "244Chesuiut Street FAX Salem, NJ 08079 Fax: (856).139-3997
( 0 W-33 7- *0 PHONE El Nuclear Services Team 244 Chestnut StrCCt Salem, NJ 08079 Fax: (856)339-.2382 TOTAL PAGES (INCLUDING COVER)
O PSEG NuclearLLC APR-02-2002 13:55 S093393997 P. 01
SENT BY:TRAINING CENTER ; 4 2 ; 14:02 PSE&G NUCLEAR 6103375320;# 2 PSEG HOPE GREEK OPS TEL:856-339-3014 P.002 APR.-02'02(TUE) 13:51 Work Order Shop Papers 08/04/2001 SAFETY: The Only C.HOI.C.E.
Conrmitnrrnt &Ip fiversighl InvDlvemenT Page 1 of 1 APR-02-2002 13:57 6093393997 95% P. 02
SENT BY:TRAIN1NG CENTER ; 4 2 14:04 ; PSE&G NUCLEAR-I 6103375320;# 3 AR.,-U2oU2`l'UE) 13:53 PSEG HOPE CREEK OPS TEL: 856-33Y-3014 P.003 I
Operation List Summary 0A 08/04/2001 C)M 50020180 ur' *UD worK Ouscriptio n Start Date Work No Durtn Op. Center 0010 0-H PERF OP-ISBF-0103 C 1011012001 a 2 4 RD ACC CHG VLVS 550D Page 1 of 1 APR-02-2002 13:5B 6093393997 95% P. 03
SENT BY:TRAINING CENTER ;4 2 14:04 ; PSE&G NUCLEAR 6103375320;# 4 APR,-02O2(TUE) 13:53 PSEG HOPE CPREEK OPS TEL:856-339-3014 P.004 Operation Key Info A10n4m 08104/2001 MRq 500201 80 Order: 50020180 PERF OP-IS.BF-103 CRD ACC CHG VLVS 5500 Operation; 0010 PERF OP-IS.BF-0103 CRD ACC CHG VLVS 550D Permit No's Work oenter: O-H NNUC Status: REL PRC SETC Number of People: 2 Scheduled Dates; Start; 10/10/2001 Finish: 10/11/2001 Planned Hours; 8 Actual Dat es: Start: Finish: Actual Hours; 0 Personnel Number: _ Completion Confirmation Number:
1006454 Confirmation Text:
Signature:
Descrptilon of Work:
PERF OP-IS.BF-0103 CRD ACC CHG VLVS 550D PERFORM CRD ACCUMULATOR CHARGING WATER CHECK VLVS 5500 IN SERVICE TEST-REFUELING OP-IS.BF-103({).
MTE CALiBRATED STOPWATCH 444 *4****
- DOCUMENTS CLASS MANUAL/DRAWING NUMBER SEQUENCE OP-IS.BF-0103 Q(
°'"*'
- CODE JOB PACKAGE ASSIGNED PLANNER POWELL 3643 ACTIVITY COMPONENT ID NONE Page I of 1 APR-02-2002 13:58 6093393997 P. 04
SENT BY:TRAINING CENTER ; 4 2 ; 14:05 ; PSE&(i NUCLEAR-' 6103375320;# 5 APR -02 02(TUE) 13;53 PSEG HOPE CREEK OPS TEL:856-339-3014 P. 005 PSEG Internal Use Only Page 1of 3 HOPE CREEK GENERATING STATIOP HC.OP.IS.BF-0103(Q) - Rev. I CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER'CHE*K*VALVE.
REFUEL - INSERVICE TEST USE CATEGORY; .
- 1. This procedure has been converted from Professional Write to Microsoft Word.
- 2. The conversion of this procedure required portions of the Attachments to be re-typed.
- 3. Organizational title changes were made in this revision to bring the procedure in line with guidelines, as containcd in NC.NA-AP.ZZ-0002(Q), Nuclear Business Unit Organization, Attachment 1 and are considered editorial based on an allowance in NC.NA-AP.ZZ-0001 (Q),
Attachment 7 for "changing personnel titles to reflect organizational changes (without ehanging authority or responsibilities)." Due to the extensive changes, revision bars were omitted,
- 4. The following changes were made in this revision to bring the procedure in line with the rules governing procedure format, content, and writer/reviewer guidelines, as contained in NC.NA.
WG.ZZ-0001 (Q), Procedure Writers Guide and can be considered Editorial in nature. Due to the extensive changes, revision bars were omitted.
- Added "RECORDS" Section 6. 0 Capitalization and Bolding of action verbs
- Changed procedure Font from "Arial" to -Timnes New Roman"
- Revised Cautions and Note boxes format from margin to margin
- Moved "Commitmenm Document" numbers from left margins to end of applicable steps
- Added Placekeeping/Step completion signoffs throughout procedure
- 5. This procedure hal been revised to add Document Secur.ity classification statement "PSEG Internal Use Only" to procedure header.
- 6. LIST OF EFFECTIVE PAGES has been deleted, this is a aeneric change.
- 7. Removed Note 2.0 from Prerequisites Section that staTed "Prerequisites within a subsection may be completed in any order". NC.NA-AP.ZZ-0001 (Q), Nuclear Department Procedure System, states that Prerequisites need not be completed in order unless specifically stated, so, this Note is no longer required in any Implementing procedure. This change can be considered editorial in nature.
- 8. CAUMION 2.1.4 has been changed to NOTE 2.1.4. The information contented does not satisfy the criteria for a caution.
(Coniinued)
INELEMP NTATION REOUIREMENTS Effective Date-t APPROVED: ______ ________
Operations Manager Date APR-02-2002 13:59 609339399? P. 05
SENT BY:TRAINING CENTER ; 4 2 ; 14:05 PSE&G NUCLEAR 6103375320;# 6 Af,.-U2 U2IIUb) 16:ý4 FNLU HUFL X(E Ut'S '11! ýb-3)9-3U 14 P. 006 PSEG In ter-qal Use Only HC.OP-S.B'-O 103(Q)
!g2I 5.1.7 Addctional Attachment 3 forms may be used for performing the following steps.
5.1.7 VALIDATE the Control Rod Drive Accumulator low pressurc zdarm by performing the following steps:
A, LIST the identification number of each HCU recorded ont Artachmenrt 2 from Step 5.1.5.C on Attachment 3.
B. ELIMINATE the alarms caused by high water lAW HC.OP-SO.BF-0002(Q), Individua* CRD HCU Operation, Subsection 5.6, Draining Accumulator (nitrogea side) of Water.
C. IF the catse fot the Accumulator zlarm can not be attributed to high water, ENTER VALID on Attachment 3, otherwiise ENTER INVALID.
D. INITIAL the space provided on Attachmznt 3.
E. RETURN this procedure to the NCO for the :ormp[etion of this test. 4-5.1.8 RECORD SAT on Attachment 2 for any Control Rod Drive Accuxmulator Alams invalidated in Step 5.1.7, otherwise RECORD UNSAT.
5.1.9 RECORD SAT oa Attachmcnt 2 for all Control Rod Drive Accumulators that did not alarm during the performance of Step 5.1.5.
S.1.10 INITIAL the space provided on Attachment 2. +/-
5.1.11 LOG rest end time in the Contol Room log(s),
5.1.12 SUBMIT this procedure to the OS/CRS for review AND complction of Attachment 1.
4 Hope Creek Page 5 of20 Rev. 1 APR-02-2002 14:00 6093393997 95% P. 06
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ATTACHMENT I Page I of 3 OS/CRS DATA AND SIGNATURE, SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHCIK VALVE REFUEL - INSERVICE TEST 1.0 PIU2EST INFORMATION 1.1 Reo>son toritheTest 1.11.1 Regular Survoillance INITIALS 1.1.2 Retest INITIALS 1.1.3 IF not performing the complete test, LIST subsections to be performed AQ Accumulators that this procedure is testing.
SUBSECTION(S) 1.2 Peagt Condition5 1.2.1 Operational Condition 1.2.2 Reactor Power Level 1-2-3 OMWe 13 -ermigsiou to Perform the Tes-1.3.1 A review of NC.NA-AP.ZZ-0005(Q); Station Operating Practices for a list of systems requiring an independent verification has been completed. The OS/CRS has placed an NIA in the applicable space(s) on Attachment 2 which DO NOT requirn ýiCpcndent vcgfication.
DATE-TIME 1.3.2 Penmission granted to perform this teSL 0S/CRS DATE-TIME Hope Creek Page 7 of 20 Rev. I APR-02-2002 14:00 6093393997 P. 07
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ATTACHMENT 1 Page 2 of 3 OS/CRS DATA AND SIGNATURE SHEET CONTROL ROD DRIVE ACCUN(II'LATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 2.0 POSTITEST INFOIRMATTON 2.1 The data acquired duting the performance of this tos'c Fas been reviewed for completeness and compliance with Technical Specification 4.0.5 and the test is considered.
2.1,1 .ATTSFAqIT (All acceptance criteta is marked SAT)
(S/CRS DATE-TIvM 2.1.2 UNSATISFACTORY (Any test evaluations are marked UNSAT). I TAKE action IAW NC.NA-AP.ZZ-0070(Q).
OStCRS DAT'E-TIME 2.1.3 Test results which are related to Technical Specification 4.0.5 have been evaluated for acceptability. IF required. an Action Request has been generated to incorporate new baseline data, for Inservi e Test components contained in ihis procedure. ML3H PR 01-T 1ST r~G a fAU f5dN ER DATIE-T1&M 2.1.4 Wotik Order No.
2.1.5 Remarks Hope Creek Page s or20 Rev. I APR-02-2002 14:00 6093393997 P. 08
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ATTACHMENT I Page 3 of 3 OSICRS DATA AND SIGNATIUR SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 3.0 P?1,, (-.LRE PERFORMER S] ANI) VER.FI ER(S) 3.1 1 have read and understand the steps of this procedure that I am required to perform.
(All Depairmencs)
PRINT NAMj SIGNATURE INATIALS Al / die-GIC -*7 _ .t //J l I--'"* o I/ i,'c"-t" 6 1 aus7 la I/ ///o
___ _ _ _/__ _ __ _!_ _ / __ _ _
___ _ _ _/_ I_ _ _ /__ _ _ _
___ _ _ _I _ _ _ _ /__ _ _ _ __ _ _
___ _ __ /_ __ _ _ /__ _ _/_ _ _
_ _/ I _ *.l, I
_ _ _ _I__ __ _ / _ _ / _ _ _ _
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A..ACHMVIENT 2
?.1ze 1 of 10 COYFT-AC--. R~OOM DATA SHEET CONTROL ROD DRIVE ACCU'-F.--,J0R CHARGINU WATER CHIECK VALVE K -J:. CENSER VICE TEST Con~J,&d Drive Exercise "HCU STEP 5. STEP 5.1.8 A~ND 5.1.9 PR NXTNMBER SAT OR UNSAT (UNSAT FOR VALID, W~SOR S.k SAT WFD INVALID) 02-19 9A/lA.Y 02-237 Is.
02-35 ZýA-_ __
- 02-39 ___________
N' 06-19 A 4' 062.3 A/ I - A
- 06-27 /V T 06-2S
- 06-39 S4 A4
- 06-43 )A -
- 06-47 Ž .. _______%'
- 10-11 *
- 10-15
- 10-19 A
- Acceptance Criterion - the A.-. -:Kokmust be marked SAT.
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ATTACHMENT 2 Page 2 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER C-ECK VALVE REFUEL - INSERVICE TEST I Control Rod Drive Exercise STEP 5.1.5 STEP 5.1.8 AND 5.1.9 PERF N1UMBER ACCUMULATOR SAT OR UNSAT ALA.IR (UNSAT FOR VALID, YES OR N/A SAT FOR INVALID) 10-27 .. 7. F __ *
. 0-35
- M0-39 _IV____ ______
10-43)
. 0-47 _,__"_,_A Zi i *
,_ ]+J ... _',t ***s
- 14-07 N_ __ __ __
14-11 14
- 14-35 14-27 AJ) ".
- 14-39 AlP 4'
"14-39 /vl,+ "+
- 14-47 __
- 14-53 re .
- Acceptaace Critcrion - the SATIUNNSAT block must be m~arked SAT.
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ATTACHN1ENT 2 Page 3 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARIGING WATER CHECK VALVE REFUEL - INSERVICE TEST Conrrmi Rod Drive Everejij HCU STEP 511.5 STEP 5.1.8 AND 5.1.9 MMI*MBER ACCUMULATOR SAT OR UNSAT ALARM (UNSAT FOR VALID, YES OR N/A SAT FJ)R INVALID) 03
- A.-07 v!1,*]
- .'.I * ,
I ~A , ý.,- ÷,SAT . *
- 8-23 _4,.T
- s8-27 _______, _ ______
- -31 B... SA f.* .1' IV 6F 18-55
- 22-07 _ _ _ 4
- _+/- *
":2-19 p- *
- Accepiance Critenion - the SAT/UJNSAT block must be marked SAT.
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ATTACHMENT 2 Page 4 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST Control Rod Drive Exercise HCU STEP 5.1.5 STEP 5.1.8 AND 3.1.9 PERF NUMBER ACCUMULATOR SAT OR UNSAT ALARM (UNSAT FOR VALID,
__ _ YES OR N/A SAT FOR INVALIDI ,
22.23 ... & . ...- _) *g 22-27 A/1 .i-i g 22-31 ____"
22-35 ~i
- 22-39 PId 0 _ _"___A __
22,43 ._
22-47 ,~ ._________________
- 22-51AA ____
4 4
22-59 Ak __ _ _ __ _ _ _ _ _ _
26.03 V(
26-07 _A 4
26-11 ___(A____
26-15 Al ('
Is 26-19 A~_ _ _ _ _ _ _ _ _
26-23 M A, _ __ _ _ __ A 26-27 ______
26.35 26-39 /VIA . *
- Acceplance Criterion - the SAT/UNSAT block must be marked SAT.
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.PSEG Intern1l Us OnLy RC.OP-IS.BF-01O3(Q)
ATTACHMENT 2 Page 5 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 1.1 Control Rod Drive Exercise H0: STEP 5.1.5 STEP 5.1., ANND 5.1.9 PER"F NUMBER ACCUMULATOR SAT OR UNSAT YEAOARIN(JNSAT FOR V.ALID, YES OR NIA
... SAT FO8 INV.A-IuD) 26-4.i: A)_"__ _" __ _ _
- 4.
26-*9 "v.
30-03 j V . ,_.
30 , .
30-11, Al, r ,
3 0 - 1; (L ., - ..
- 30-19 n "*'-"l' 30.23 di 4.
30-3o. 1 Al "--T" 4' 30-36 t JA ' ..
- 4'
_. 30-39 _", :4.....
K 30-51 AY A__________________
"AcceptanctCriterion - the SAT/UNSAT block must be marked SAT.
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ATTACHMENT 2 Page 6 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 1.1 Control Rod Drive Exerrise HCU STEP 5.1. STEP 5.1.8 AND 5.1.9 FERF (NUMBER ACCUMULATOR SAT OR LUNSAT ALARM (-NSAT FOR VALID, YES OR N/A SAT FOR INVALID) 34-03 AM_ _ _ _ _ _
34-07 34-11 4
34-15 P14_____
- 34-19 v,_ . ..
34-27 rulk .,
- 34-31 /VIA 34-43 AlJA '__
34-47
- l.-,* -*
4.
34-51*,*' ' IC 34-59.. i S T, I' 38-03 38-15 /
38-19 /V#
I LK_
" Acceptance Criterion - the SAT/UNSAT block must be marked SAT.
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ATTACHMENT 2 Page 7 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST
- 1. 1 Qontrol Rod Drive Exercise
- Acceptance Critc-ion - the SAT/UNSAT block must be rarked SAT.
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ATTACHMENT 2 Page 8 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 1.1 Control Rod Drive Exercise HCU STEP 5.1.5 STEP 5. 1.8 AND 5.1,9 ACCUMULATOR PERF I NUMBER SAT OR UNSAT ALARM (TJNSAT FOR VALID, YES OR N/A SAT FOR INVALrn*
42-43 SAT . - --
- 42-47 A...,
- 42-51 ,O A"
- 42-59 '
- 46-7.. ...
- 46-11 46-19 46-15 AJ lik. .,",*.: !- ,
- 46.23 -- k
- 46-2 1 46-35 46-39l 46-55 All* "+' 'T*
- 46-43A
- 46-47 A~ _ _ _ _ _ _ _ _ _
46-51 50-11 IVA/A___
- 15 1v4- *
- Acceptance Criterion - the SATIIUNSAT block must be marked SAT.
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ATTACHMENT 2 Page 9 of I0 CONTROL ROOM DATA SHEET CO-TROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 1,1 Cotrrol Rod Drivc Exercise HCU STEP 5.1.5 STEP 5.1.8 -AND 5.1.9 ACCUMULATOR SAT OR UNSAT PERF
.'UMBER ALARM (UNSAT FOR VALID, YES OR N/A SAT FO')l/INVALI*3 so-:N A" -ll
__ J-- _ _ _ _ __
- __,
I*
50,5: 1 ---- AAA -- .
- ...tr:* ,
tV 1.4
- o--t,4_______ *__
54-F.
_-_ ~4 A-' 77--4
- m 14.. ,, *
' Accep tc Criterion - ihe SATAJNSAT block must be marked SAT.
I.'.
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}-IC.OP-IS.3F-01 O:3(Q)
ATTACHMENT 2 Page 10 of 10 CONTROL ROOM DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST 1.1 Control Rod Drive Exerise Hcu STEP s.L1s STEP 5.1.8 AND 5.1-9 NUMBER ACCUMULATOR SAT OR UNSAT ALARM (UNSAT FOR VALID, YES OR N/A SAT FOR INVAL1D) 58-31 S.8-39 58-43 A//A-AA
- Acceptance Criterion - the SAT/UNSAT block must be marked SAT.
.1L2 *
- Accpiance Criterion - the SAT/UNSAT block must be marked SAT or actual time must be > 2 minutes.
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ATTACHMENT 3 Page 1 of 1 INPLANT DATA SHEET CONTROL ROD DRIVE ACCUMULATOR CHARGING WATER CHECK VALVE REFUEL - INSERVICE TEST Hope Creek Page 20 of 20 Rev. 1 APR-02-2002 14:05 6093393997 P. 20
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 1 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION Applicability:
Salem 1 Salem 3 (Gas Turbine) NBU Common Salem 2 XXX Hope Creek Common to Salem 1 &2 Common to Hope Creek &Salem COMPLETION AND APPROVAL MARK CIRELLY 09109/00 PREPRERYIN) ATE NAME (PRINT) QUAL EXPIRS JOHN THOMPSON 04/02/99 PE R REV ER (SIGN NAME (PRINT) OVuA EXPIRES LEN RAJKOWSKI 0311 o/99X OV (IGN)DAIE NAME (PRIN)
- /1 - DRTe RVIW, Safety Evaluation No. N. ?-0ooy SORC Chairman: Mtg. No. LJ~-~ Date _ _ _ _ _
(Hope Creek)
Sta. GM Approval: _____,/_/ __________ Date i (i-lope Cr k) )
SORG Chairman: / Mtg. No. Date (Salem) J, "
Sta. GM Approval: Date (Salem)
Safety Evaluation and associated documentation sent to Nuclear Review Board (NRB)
MIC N38: SORC
[UFSAR 17.2.1.1.2.1] Presenter: Date:
Nuclear Common Page I of 16 Rev. 7 APR-01-2002 15:42 6093393997 P. 09
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FORM-3 10CFRS0.59 SAFETY EVALUATION Page 2 of 16 Revision 0 I.D. Numbers!Reference/Revision: HC.OP-SO.SB-0001 (0) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION 1.0 10CFR50.54 PRE-SCREENING YES NO XXX a. Could the proposed change affect the Quality Assurance Program Description Included In the UFSAR?
IfYES, STOP. Contact Quatity Assessment for assistance.
_XXX b. CouTd the proposed change affect the Security Plan?
If YES, STOP. Contact NuclearSecurity for assistance.
XXX c. Could the proposed change affect the Emergency Plan?
IfYES. STOP. Contact Emergency Preparedness for assistance.
2.0 1DCFR50.59 APPLICABILITY REVIEW - 10CFR50.59 applies because:
2.1 The proposal changes the facility as described in the SAR.
YES XXX NO Explain: Sections 7.2.1.1.11 and 7.2.2.3.7 of the UFSAR describe the RPS scram signal generated when the Reactor Mode Switch is placed the "Shutdown" position. Insofar as the proposed procedure revision provides instructions to bypass this feature, it constitutes a change to the facility as described in the SAR.
2.2 The proposal changes procedures as described in the SAR.
YES NO XXX Explain: Although the subject procedure is listed in the SAR, it is not described. Therefore, the proposed revision does not change procedures as described in the SAR.
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FORM-3 IOCFR5O.59 SAFETY EVALUATION Page 3 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION 2.3 The proposal involves a test or experiment not described In the SAR.
YES NO XXX Explain: The activities associated with the proposed procedure revision do not constitute a test or experiment as defined in NC.NA-AP.ZZ-0008 (Q) and NC.NA-AS.ZZ-0059 (Q). Therefore, it does not involve a test or experiment not described in the SAR.
3.0 LICENSING BASIS DOCUMENTATION 3.1 UFSAR REVISION DETERMINATION - Does the proposal require a UFSAR change?
YES NO XXX UFSAR Change Notice No. XXX The proposed procedure revision provides instructions on bypassing the RPS scram signal generated when the Mode Switch is placed in the 'Shutdown" position. This will only be done when the reactor is shutdown with all rods inserted. The scram signal will only be bypassed for approximately 30 seconds each time the mode switch is required to be moved to the shutdown position and then restored. Since this will not be a normal mode of operation, and the system is restored to the configuration described in the SAR each time the procedure Is performed, a change to the SAR is not required.
3.2 TECHNICAL SPECIFICATION REVISION DETERMINATION - Does the proposal require a Technical Specification change?
YES NO XXX If a change is required, STOP. Contact Nuclear Licensing for assistance in preparation of a License Change Request.
Identify the pertinent Technical Specification sections that were reviewed to make the determination:
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FORM-3 10CFR50o.59 SAFETY EVALUATION Page 4 of 16 Revision 0 I.D. NumberslReference/Revision: HC.OP-SO.,SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION 2.2 LIMITING SAFETY SYSTEM SETTiNGS- REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 3-3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 3.3.6 CONTROL ROD BLOCK INSTRUMENTATION 3.9.1 REFUELING OPERATIONS - REACTOR MODE SWITCH
4.0 DESCRIPTION
4.1 Describe the modification or activity being evaluated 'and its expected effects.
JUSTIFICATION FOR PROPOSED CHANGE:
When the Reactor Mode switch is placed in the shutdown position, a RPS scram signal is generated and then is automatically bypassed after a six (+/- four) second time delay.
This feature is required to be operable in Operational Conditions (Op Con) 1 through 5 in accordance with Specification 3.3.1. With the feature not operable in Op Con 5, the required action is to suspend core alterations and insert all insertable control rods within one hour During the course of a refueling outage, it is necessary to periodically move the mode switch between the refueling and shutdown positions for testing and surveillance purposes and to conduct core alterations. However, the resultant scram signals cause CRD scram valves to reposition and scram accumulators to discharge. Empirical experience has shown this to be a major contributor to control rod withdrawal difficulties during startup due to nitrogen intrusion into the CRD HCUs. Additionally, the repeated scram signals place mechanical and hydraulic stresses on internal control rod drive mechanism components that accelerate the degradation of the drives.
DESCRIPTION OF PROPOSED CHANGE:
UFSAR Section 7.2.1.1.10 describes a manual reactor scram as being initiated by use of the four RPS manual pushbuttons. Additionally, the statement is made thai "Manual reactor scram is diverse to all automatic reactor trip signals." Section 7.2.1.1.11 then describes the reactor mode switch manual scram signal. Therefore, the scram signal Nuclear Common Page 4 of 16 Rev. 7 APR-01-2002 15:43 6093393997 P. 10
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FORM-3 IOCFR50.59 SAFETY EVALUATION Page 5 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(0) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION generated by the mode switch is redundant to the manual scram pushbuttons, which in turn are redundant to all automatic scram signals.
In order to preclude control rod withdrawal difficulties during startups, the proposed procedure revision is being incorporated to bypass the RPS scram signal generated when the mode switch is shifted from the refuel to the shutdown position. The prerequisites of the procedure require permission to be obtained from the OS/CRS, all control rods to be inserted (or the core to be off-loaded) and no core alterations to be in progress. These prerequisites will ensure that Technical Specification operability is addressed, and that the required actions have been completed prior to bypassing the mode switch scram signal. A note will alert the operator that bypassing the mode switch scram signal will render the switch inoperable in accordance with Technical Specifications. Additionally, the prerequisites require the verification that banar)a jacks have been installed on the affected terminals in order to facilitate the required jumpering. The installation of the banana jacks will be performed under existing approved procedures. This ensures compliance with the guidance of Regulatory Guide 1.118 and it's referenced documents pertaining to lifted leads and jumpers as discussed in UFSAR Sections 1.8.1.118 and 7.1.2.4, Item #17.
Once the prerequisites are complete, the mode switch position scram signal, and only that signal, is bypassed by installing four jumpers. The mode switch is then placed in the shutdown position. After approximately thirty seconds, the jumpers are removed and the mode switch is returned to operable status. The value of approximately 30 seconds is used to provide sufficient time for the six (+J- four) second timer to time out and then automatically bypass the scram signal. However, a delay in the removal of the jumpers would neither affect nor invalidate this evaluation or its conclusions since there are no time-dependent factors introduced. At this point, the jumpers are no longer required and can be removed. Independent verification of the prerequisites is performed as procedural steps. Installation and removal of the jumpers are documented and second verified in Attachment 5 of the procedure.
DESCRIPTION OF THE RPS LOGIC AFFECTED BY PROPOSED CHANGE:
The part of the RPS logic that is affected by the proposed revision is shown on GE Elementary Drawing PN1 -C71-1020-0006, sheet 12. This is the reactor manual scram trip logic for each of the four RPS subchannels (Al, A2, B1 and 82). The normally energized logic feeds the K15A-D relays. If the K15 relays become de-energized, they in turn open the K15 contacts in the automatic scram logic shown on sheet 13, thus de energizing the K14 contacts and causing a reactor scram. The K15 relays can be de Nuclear Common Page 5 of 16 Rev. 7 APR-01-2002 15:44 6093393997 95% P. 11
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 6 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION energized by the Nuclear Instrumentation (NI) non-coincident trip contacts, the reactor mode switch being placed in the shutdown position or the manual RPS scram pushbuttons. The mode switch part of the logic consists of four contacts in parallel, K32, K34, K35 and K36. Each contact is associated with a mode switch position, (run, startup, refuel and shutdown respectively) and is closed when the switch is in that position. The K36 contact, which is closed when the mode switch is in the shutdown position, also has a time delayed contact in series with it, the KIB contact. These contacts close six (+/- four) seconds after the mode switch is placed in shutdown.
Consequently, when the mode switch is placed in shutdown, the K36 contacts close; however, logic power is interrupted for six (+/- four) seconds until the K36 contacts close. This is how the scram signal is generated and then automaticaliy bypassed.
The jumpers installed by the proposed revision are placed across terminals Z-9 and ZZ 23 in the Al RPS logic. A-9 and BB-23 In the A2 logic, Z-9 and ZZ-23 in the B1 logic and A-9 and BB-23 in the 82 logic. As can be seen on GE Elementary Drawing PN1 C71-1020-0006, sheet 12, jumpers at these locations will provide a logic path around contacts K32, K34, K35 and K36 thus preventing a scram signal when the mode switch is placed in shutdown. The rest of the manual scram logic is unaffected by the jumpers since opening of any of the other contacts will de-energize the respective Ki 5 contacts and result in a scram signal. Additionally, it can be seen on sheets 2 and 12 that there is no other mode switch functions affected by the jumpers. The signals associated with the run, startup and refuel positions are unaffected as seen on sheet 2 and the non scram signals associated with the shutdown position (to the Nuclear Steam Supply System, Rod Block interlocks, MSIV isolation bypass and Scram Discharge Volume High Level bypass logics) remain totally unaffected. Additionally, the NI and manual RPS push button logics remain unaffected.
4.2 Identify the parameters and systems affected by the change.
The proposed procedure revision affects the RPS scram signal generated when the mode switch is placed in the shutdown position. No other signals or actuations are affected. Refer to preceding discussion in Section 4.1 for technical discussion and references.
4-3 Identify the credible failure modes associated with the change.
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FORM-3 10CFR5O.59 SAFETY EVALUATION Page 7 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SOSB-O0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION The credible failure modes associated with the proposed procedure revision include incorrect installation or removal of the subject jumpers and failure of the RPS manual scram pushbuttons necessitating use of the mode switch to initiate a manual scram.
4.4 Provide references to location of information used for the Safety Evaluation.
UFSAR Sections 1.8, 7.2, 7.5, 7.7, 15.7. 15.8, 15.9 Technical Specifications 2.2, 3.3.1, 3.9.1, 3.3.6, GEK-90348B - Reactor Protection System (PN1-A41-8010-0044 (2))
ANSIIANS-58.9-1981 -Single Failure Criteria for Light Water Reactor Safety-related Fluid Systems Regulatory Guide 1.47 - BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS Regulatory Guide 1.118 - PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS - REVISION2 NC.NA-AP.ZZ-0008 (Q)
NC.NA-AS.ZZ-0059 (Q) 4-5 Other Discussion, If applicable.
UFSAR Section 1.8.1.47 states that the Hope Creek Generating Station complies with Regulatory Guide (Reg Guide) 1.47, "Bypassed And Inoperable Status Indication For Nuclear Power Plant Safety Systems". The Reg Guide delineates requirements to provide automatic or manually actuated control room indication when the protective action of some part of a protection system has been bypassed or deliberately rendered inoperative. Although the proposed procedure will bypass the reactor mode switch position scram signal, it will only be performed when all control rods have been inserted or the core has been off-loaded. Therefore, the protective action that is affected by the bypass, control rod insertion, will already have occurred. Per reg. Guide 1.47 automatic indication of bypass or inoperable conditions apply when all of the following conditions exist::
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FORM-3 IDCFR50.59 SAFETY EVALUATION Page 8 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION
- 1. The bypass or inoperable condition affects a system that is designed to perform automatically a function that is important to the safety of the public. (Placing the mode switch to shutdown is a manual function, so this condition does not apply)
- 2. The bypass will be utilized by plant personnel or the inoperable condition can reasonably be expected to occur more frequently than once per year. (Although this procedure will only be performed during an outage, it will occur several times during the outage and therefore applies)
- 3. The bypass or inoperable condition is expected to occur when the affected system is normally required to be operable. (As controlled by the procedure, the bypass will occur when the mode switch is inoperable and appropriate technical specification action statement wilI be in affect, so this condition does not apply)
Based on the preceding discussion, It is concluded that although the requirements of the regulatory guide do not apply, the proposed revision is nonetheless in compliance with it.
5,0 USQ DETERMINATION - Is an Unreviewed Safety Question (USQ) involved?
5.1 Which anticipated operational transients or postulated design basis accidents previously evaluated in the SAR are considered applicable to the proposal?
Insofar as the mode switch shutdown position scram is a redundant means of initiating a manual reactor scram, the Anticipated Transient Without Scram event is applicable to the proposed change.
5.2 May the proposal:
- a. Increase the probability of an accident previously evaluated in the SAR?
YES NO XXX DISCUSSION: The proposed procedure revision will facilitate bypassing of the RPS scram signal generated when the mode switch is placed in the shutdown position. This is accomplished by installation of four jumpers.
There are no other signals or actuations affected by installation of the specified jumpers. All other scram signals will remain unaffected and the rod block generated when the mode switch is in the shutdown position will remain in effect. In accordance with prerequisites included in the proposed change.
the evolution will not be performed unless permission has been obtained from Nuclear Common Page 8 of 16 Rev. 7 APR-01-2002 15:45 6093393997 P. 14
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 9 of 16 Revision 0 I.D. Numbers/ReferencelRevision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION the OSJCRS, all control rods have been inserted (or the core is off-loaded) and no core alterations are in progress. Independent verification of these prerequisites is performed as procedural steps. Per the proposed revision, once the mode switch has been placed in shutdown and approximately thirty seconds have elapsed, the jumpers are removed and the system configuration is restored to normal. Since the prerequisites remain in effect until the jumpers are removed, a delay in the removal of the jumpers would have no adverse impact on plant safety. Installation and removal of the jumpers are documented and second verified in accordance with an attachment included in the revision. Based on the preceding, it is concluded that the prerequisites will ensure the proposed procedure will only be performed when the safety function of RPS, control rod insertion, has already been met (or the core is off-loaded) and the probability of a failure of RPS leading to an ATWS event is completely precluded.
Based on the preceding, it is concluded that the proposed change will not affect any accident precursors and therefore will not increase the probability of any accidents previously evaluated in the SAR.
- b. Increase the consequences of an accident previously evaluated in the SAR?
YES NO XXX DISCUSSION: UFSAR Section 7.2.1.1.10 describes a manual reactor scram as being initiated by use of the four RPS manual pushbuttons.
Additionally, the statement is made that "Manual reactor scram is diverse to all automatic reactor trip signals." Section 7.2.1.1.11 then describes the reactor mode switch manual scram signal. Therefore, the scram signal generated by the mode switch is redundant to the manual scram pushbuttons, which in turn are redundant to all automatic scram signals.
In accordance with prerequisites included in the proposed change, the evolution will not be performed unless permission has been obtained from the OS/CRS, all control rods have been inserted (or the core is off-loaded) and no core alterations are in progress. Since the reactor is already shutdown, there are no credible accidents in which the manually initiated scram signal could be a potential mitigating factor. Therefore, the consequences of a Nuclear Common Page 9 of 16 Rev. 7 APR-01-2002 15:46 6093393997 955 P. 15
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 10 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001((Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION failure to scram (ATWS), including potential radioactive release, are unaffected. It is therefore concluded that the proposed change will not increase the consequences of any accident previously evaluated in the SAR.
5.3 What malfunctions of equipment important to safety that were previously evaluated in the SAR are considered applicable to the proposal?
The scram signal generated by the mode switch is redundant to the manual scram pushbuttons, which in turn are redundant to all automatic scram signals.
Therefore, failure of the RPS system to initiate an automatic reactor scram is the most pertinent malfunction.
5.4 May the proposal:
- a. Increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?
YES NO XXX DISCUSSION: The RPS logic is de-energized to trip. When the mode switch is placed in the shutdown position, a contact opens which then interrupts power to the logic relay and results in a scram signal. When the scram signal is bypassed in accordance with the proposed revision, jumpers will be installed around the contacts associated with the mode switch position. This will maintain a path for current flow to the rest of the logic train when the switch position contacts open, thus defeating the scram signal.
However, the jumpers will have no affect on the rest of the logic since the opening of any of the other contacts associated with automatic scram setpoints and the RPS manual push buttons will still de-energize the logic and result in a reactor scram signal. Therefore, although the mode switch position scram signal is one feature that can be used to mitigate an ATWS event, it will not affect the probability of the system failing to automatically scram the reactor when any of the sensed parameters exceed their specified setpoints during an anticipated transient. The preceding discussion notwithstanding, the proposed procedure wIll not be performed unless the reactor is already fully shutdown. Requiring either all control rods to be Nuclear Common Page 10 of 16 Rev. 7 APR-01-2002 15:46 609339399? P. 16
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Title:
REACTOR PROTECTION SYSTEM OPERATION inserted or the core to be off-loaded ensures this. Therefore, the conditions leading to an ATWS event are entirely precluded.
UFSAR Section 15.8 provides an assessment of the Anticipated Transient Without Scram (ATVVS) event. The discussion focuses on operator response delineated by the Emergency Operating Procedures and the use of "diverse, highly redundant, and very reliable scram systems". The description mentions the following design features: normal scram systems, Alternate Rod Insertion System, ATWS recirculation pump trip, manual rod insertion, Standby Liquid Control System. feedwater runback, and the scram discharge volume. It does not describe or list the reactor mode switch or the associated scram signal.
As a mitigation feature (not described in the SAR), the mode switch is placed in shutdown in accordance with the EOPs in order to insert a redundant scram signal and also to bypass the MSIV isolation interlock if reactor pressure decreases to the specified setpoint. This retains the main condenser as a heat sink after a reactor trip. The proposbd change will have no impact on this feature.
Potential errors that could occur during installation and removal of the subject jumpers were assessed as part of this safety evaluation. During installation of the jumpers, the jumpers could be misplaced. The misplaced jumper could:
- 1. bypass a different function than intended
- 2. short out a power bus
- 3. cross-connect circuits such that the logic operates in a different manner.
In any of these scenarios, there is no safety impact. The procedure is only being used when the plant is shut down and after all control rods are verified to be full in or the core is off-loaded. Therefore, the conditions leading to an ATWS event are entirely precluded and the safety function of RPS has already been completed. If a jumper is misplaced, a half scram will occur and be detected when the mode switch is placed in shutdown. RPS is fail-safe and a power bus short would result in a blown fuse and half scram in that channel. A single jumper mistake would be limited to one sub-channel (Al, A2, B1, and B2) due to the physical separation of channels in the RPS panel bays, Nuclear Common Page 11 of 16 Rev. 7 APR-O1-2002 15:47 6093393997 95% P. I?
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FORM-3 I0CFR50.69 SAFETY EVALUATION Page 12 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION Installation and removal of the subject jumpers are documented and second verified in accordance with an attachment included in the revision.
Additionally, the prerequisites require the verification that banana jacks have been installed on the affected terminals in order to facilitate the required jumpering. The installation of the banana jacks will be performed under separate station procedures and will provide additional assurance that the jumpers are correctly installed, do not come loose while installed and are properly removed. This ensures compliance with the guidance of Regulatory Guide 1.118 and it's referenced documents pertaining to lifted leads and jumpers as discussed in UFSAR Sections 1.8.1.118 and 7.1.2.4, Item #17.
Where temporary alterations are required (e.g., Jumpers and/or Lifted Leads), Hope Creek is committed to follow the guidance in Office of Inspection and Enforcement (IE) Information notice 84-37, "Use of Lifted Leads and Jumpers During Maintenance and Surveillance Testing', which recommends a combination of administrative controls and functional test to verify the restoration of proper system configuration following surveillance test. This allows for additional procedure checks, recordings, and independent verifications as outlined in this referenced procedure. Therefore, it would require multiple human errors for the jumpers to be incorrectly removed. Additionally, it would require multiple equipment failures (i.e. the systems described in UFSAR Section 15.8) in order for this malfunction to impact the consequences of an ATWS. Based on the preceding discussion it is concluded that the proposed change will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.
- b. Increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR?
YES NO XXX DISCUSSION: As discussed in the preceding discussion, the sequence of events which would be required in order to necessitate use of the mode switch to initiate a manual scram are precluded by the proposed procedure revision. Therefore, the consequences of an ATWS event are unaffected.
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 13 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15 Title; REACTOR PROTECTION SYSTEM OPERATION The proposed procedure will only be used when all control rods are inserted or the core is off-loaded. Thus, the safety function of RPS, control rod insertion, have already been fulfilled and the conditions leading to an ATWS event are precluded, It is therefore concluded that the proposed revision will not increase the consequences of, or radioactive release associated with, a malfunction of equipment important to safety previously evaluated in the SAR.
5.5 May the proposal:
- a. Create the possibility of an accident of a different type from any previously evaluated in the SAR?
YES NO XXX DISCUSSION: The proposed procedure revision will only be performed when the reactor is shutdown with all rods inserted (or the core off-loaded) and no core alterations in progress, There will be no impact to any mechanical or electrical distribution systems and no potential to cause undesired initiations or failures.
The proposed change will bypass the scram signal associated with the reactor mode switch shutdown position. This will only be done when all rod s are inserted or the core Is off-loaded. Once the prerequisites are met, there are no compensatory actions required while the subject scram signal is bypassed. Therefore, there are no manual actions taking the place of automatic actions.
It is therefore concluded that the proposed change will not create the possibility of an accident of a different type from any previously evaluated in the SAR.
- b. Create the possibility of a malfunction of a different type from any previously evaluated in the SAR?
YES NO XXX Nuclear Common Page 13 of 18 Rev. 7 APR-01-2002 15:47 609339399? P. 19
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FORM-3 IOCFR5O.59 SAFETY EVALUATION Page 14 of 16 Revision 0 I.D. Numbers/Reference/Revision: HC.OP-SO.SB-001I(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION DISCUSSION: The jumpers used to bypass the mode switch position scram signal will have no impact on any other scram signal or mode switch interlock (e.g. Rod block). As discussed previously, incorrect installation of the jumpers or restoration of the mode switch scram feature would require multiple human errors and is therefore not a credible event. Further, even if this were to occur, the RPS manual pushbuttons would still be available to initiate a scram as described in UFSAR Section 7.2.1.1.10. It is therefore concluded that the proposed change will not create the possibility of a malfunction of a different type from any previously evaluated in the SAR.
5.6 Does the proposal reduce the margin of safety as defined in the basis for any Technical Specifications?
YES NO XXX Discuss the bases for the determinations and identify the pertinent Technical Specification sections that were reviewed to make the determination (use continuation sheets if required).
Specification 2.2, LIMITING SAFETY SYSTEM SETTINGS- REACTOR PROTECTION SYSTEM SETPOINTS, lists the RPS scram signals and their respective setpoints. The reactor mode switch shutdown position is listed as Item #11 on Table 2.2.1-1 with no trip setpoint or allowable value specified. The basis for this specification states that the reactor mode switch shutdown position provides additional manual reactor trip capability. With a RPS setpoint set less conservative than specified (or with the mode switch position scram unavailable),
Specification 2.2 requires 1he affected channel to be declared inoperable and the actions specified in Specification 3.3.1 to be taken. This action will apply when the mode switch scram signal is bypassed.
Specification 3.3.1provides the operability requirements for RPS instrumentation.
The basis for this specification states that RPS automatically initiates a reactor scram to: 1) preserve the integrity of the fuel cladding; 2) preserve the integrity of the reactor coolant system; 3) Minimize the energy which must be absorbed following a LOCA; and 4) prevent inadvertent criticality. Table 3.3.1-1 requires the reactor mode switch position scram signal to be operable in Operational Conditions 1 through 5. The actions required to be taken ifthe mode switch Nuclear Common Page 14 of 16 Rev. 7 APR-01-2002 15:48 6093393997 P. 20
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FORM-3 10CFR50.59 SAFETY EVALUATION Page 15 of 16 Revision 0 I1D. Numbers/Reference/Revision: HC.OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION position scram is inoperable are to suspend all core alterations and insert all insertable control rods within one hour. These actions will be required to be taken in accordance with the prerequisites of the proposed revision prior to bypassing the mode switch shutdown position scram. Therefore, the action requirements of Technical Specifications will be satisfied and the function of RPS in shutdown conditions, to prevent inadvertent criticality, will be fulfilled.
Specification 3.9.1 requires the mode switch to be operable and locked in the refuel or shutdown position in operational condition 5. The basis for this specification state that locking the mode switch in refuel or shutdown ensures that the restrictions on control rod withdrawal and refueling platform movement during refueling operations are properly activated. For the shutdown position this applies to the control rod block generated when the mode switch is in the shutdown position. Although the mode switch will be declared inoperable when the shutdown position scram is bypassed, the rod block will be unaffected and will therefore fulfill the intent of this specification. Refer to Section 4.1 for technical discussion and references.
The rod block that is generated when the mode switch is In the shutdown position is required to be operable in Operational Conditions 3 and 4 in accordance with Specification 3.3.6, CONTROL ROD BLOCK INSTRUMENTATION. This function will be unaffected by implementation proposed procedure revision. Refer to Section 4.1 for technical of the discussion and references.
The proposed procedure revision requires the installation of four jumpers; each of which will be installed in a separate bay in two separate panels.
The jumpers will be installed into banana jacks, which will be readily accessible.
Removal of the jumpers will be documented and independently verified. Based on the simplicity of the task and the ease of accessibility for removal and verification of removal, operability of the mode switch function will be re-established by documentation of jumper removal and independent verification.
No other re tests will be required.
Based on the preceding discussion, it is concluded that the proposed change will not reduce the margin of safety as defined in the basis for any Technical Specifications.
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FORM-3 10CFR50.69 SAFETY EVALUATION Page 16 of 16 Revision 0 I.D. Numbers/ReferencelRevision: HC OP-SO.SB-0001(Q) REVISION 15
Title:
REACTOR PROTECTION SYSTEM OPERATION 6.0 10CFR50.59(b)(2) REPORT - Provide a brief description of the change and a summary of the Safety Evaluation.
The proposed procedure revision provides instructions on bypassing the RPS scram signal generated when the Mode Switch is placed in the "Shutdown" position. This will only be done when the reactor is shutdown with all rods inserted. The scram signal will only be bypassed for approximately 30 seconds each time the mode switch is required to be moved to the shutdown position and then restored. The safety evaluation concludes that the proposed change does not constitute an unreviewed safety question and no Technical Specification changes are required.
7.0 CONCLUSION
If ALL answers in Section 5 are "NO," the proposal does NOT involve a USQ.
If ANY answer in Section 5 is "YES," the proposal DOES involve a USQ.
Is a USQ involved?
YES NO XXX If a USQ is involved, refer to NC.NA-AP.ZZ-0035 (Q) and obtain assistance from Licensing for additional processing.
LCR Number: NA Nuclear Common Page 16 of 16 Rov. 7 APR-01-2002 15:49 609339399? 96% P. 22
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- - - - S -
- S To: From:
NAME NAME PHONE COMPANY PSEG Nuclear, LLC DEPARTMENT Nuclear Training Department Gt3 37 - 5ý3- 0 Q-'Nuclear Training 244 Chcstnut Street FAX Salon, NJ 08079 Fax: (856)9?3-9-3997 6,o - 337- So05 PHONE 0 Nuclear Services Team 244 Chestnut Street Salem, NJ 08079 Fax: (856)439-2382 TOTAL PAGES (INCLUDING COVER)
O PSEG Nuclear LLC U PLEASE COMMENT 0 REPLY ASAP U~pLESF,COMMENT 0 URGEVNT! 0 FOR YOUR REP7EW APR-02-2002 09:57 6093393997 P. 01
SENT BY:TRAINING CENTER ; 4 2 ; 10:04 ; PSE&G NUCLEAR-I 6103375320;# 2/12 ALL ACTVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HOPE CREEK GENERATING STATION Page 1 of 2 HC.OP-IO.ZZ-0008(Q) - Rev. 10 SHUTDOWN FROM OUTSIDE CONTROL ROOM USE CATEGORY: 1 REVISION
SUMMARY
REV. 10
- 1. Attachment 1 - step B.1.5 has been changed from "TCIC" to RCIC.
- 2. Attachment 5 has been revised to delete that portion containing the curve drawing. This change satisfies the requirements of revision request OP-95-1 139.
- 3. The following changes have been incorporates to satisfy the requirements of revision request OP-9S-1 164:
- Step 5.9.12 has been revised to delete the reference to T.S Fig 3.4.6.1-1 on Attachment 1.
a Step 5.9.12 has been revised to reference Tech Spec 3.4.6.1
- 4. Attachment 5 - step 5.4 has been revised to reflect the deletion of the graphics.
Reviewer's comment.
- 5. The following changes have been incorporated as per procedure reviewer's comments:
- Step 3.1.9.C has been revised to change HV-FO07A to HV-F007B. This change conforms with the valve description and the body of this procedure. This procedure only addresses the HV-F007B.
- Step 3.1.9.D has been revised to delete the reference to HV-FO07A. This procedure provides no guidance for lowering Reactor water level utilizing the HV-FO07A.
IMPLEMEMNTAION REQUIREMENTS This procedure revision is only effective for use afe a oeoer copy is issued to the Control Room.
The previous revision remajný in effect until then, regardless of the approval date or the DCS update date. ,\,v APPROVED: _____ //9 -*(
Operations Manager Date APR-02-2002 09:58 6093393997 P. 02
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SUMMARY
REV. 10
- 5. The following changes have been incorporated as per procedure reviewer's comments:
Step 5.11.2.G has been restructured to included substeps.
" Caution 5.1.12.1. D has been revised to indicate "failure to reset NSSSS prior to transferring to the MCR may result in a Shutdown Cooling Isolation' versus "NSSS."
"* Step 5.10.2 has been revised to indicate Reactor coolant temperature is < 120°F prior to route RHR to Radwaste. This change conforms with step 5.10.2.A, 3.1.9.F, 5.11.2.B and Caution 5.11.2.
- Step 5.9.11 has been revised to add the following:
"(Maintain an administrative temperature range of 90"F -1 10'F. Other temperature(s) within TS limits may be used to support specific plant operations, as necessary).' This change conforms with HC.OP-IO.ZZ-0004(Q).
- 6. Attachment 4 - step 4.1.3 has been revised to correct valve number. The valve referenced to as 1 BC-V26 changed to I BC-V262.
- 7. The following changes have been incorporated as per reviewer's comments:
- Step 3.1.9.A has been revised to indicate only the HV-F0041B,
- Step 3.1.9.B has been revised to Indicate only the HV-F024B.
APR-02-2002 09:58 6093393997 P. 03
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6.0 REFERENCES
............................................................................................ 24 ATTACHMENTS Attachment 1 RSP Transfer Switch Automatic Action ...................................... 25 Hope Creek Page 1 of 61 Rev. 10 APR-02-2002 09:59 6093393997 P. 04
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36 Attachment 3 A RHR Loop Suppression Pool Cooling .......................................
39 Attachment 4 B RHR Loop Shutdown Cooling Operation ..................................
44 Attachment 5 Reactor Coolant System Temperature/Pressure Data .................
48 Attachment 6 Communications Systems for Safe Shutdown Areas ....................
Remote Shutdown Communication System .................................. 54 Attachment 7 HPCI Shutdown From "A" Diesel Generator Control Room .......... 57 Attachment 8 Attachment 9 Vessel Level Instrumentation Temperature Compensation Curves .................................................................. 58 Hope Creek Page 2 of 61 Rev. 10 APR-02-2002 09:59 6093393997 P. 05
SENT BY:TRAINING CENTER ; 4 2 ; 10:05 ; PSE&G NUCLEAR-' 6103375320;# 6/12 ALL ACTIVE ON-THE-SPOI CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 SHUTDOWN FROM OUTSIDE CONTROL ROOM START TIME DATE BY TERMINATION TIME DATE BY COMPLETION TIME DATE BY 1.0 PURPOSE This procedure provides guidelines for the shutdown of the plant from outside the Control Room, and reestablishing control in the Control Room.
2.0 PREREQUISITES 2.1 HC. OP-AB.ZZ-O1 30(Q), Control Room Evacuation, is complete, if possible.
2.2 Establish communications between each local panel or breaker and the Remote Shutdown Panel as the steps are performed 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Administrative 3.1.1 This procedure is to be used as a guideline for the shutdown of the plant from outside the Control Room. It is not required that each section/step be performed in precise sequence as long as the sections/steps are performed in a timely manner in keeping with the intent of this procedure.
3.1.2 In the event plant conditions require a delay during some part of this procedure, the Senior Nuclear Shift SupervisorJNuclear Shift Supervisor (SNSS/NSS) shall retain this procedure until it is continued or terminated.
3.1.3 Ifthis procedure is terminated prior to completion, the SNSS/NSS shall note the reason, time, and date of termination on this procedure.
3.1.4 For any unit scram, refer to the Event Classification Guide for the appropriate classification and notifications.
3.1.5 Ensure compliance with the Reactor Coolant System temperature and pressure requirements of Technical Specification 3.4.6.1.
Hope Creek Page 3 of 61 Rev. 10 APR-02-2002 10:00 6093393997 95Z% P. 06
SENT BY:TRAINING CENTER ; 4 2 ; 10:06 ; PSE&G NUCLEAR-I 6103375320;# 7/12 At L ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 . '- ,,- . v 3.1.6 Ensure compliance with the Reactor Vessel # Head Flange temperature limits of Technical Specification 3.4.6.1 ,d.
3.1-7 Observe the Suppression Chamber temperature requirements of Technical Specification 3.6.2,1.
CD-9048 3.1.8 When the RSP Transfer Switches are placed in EMER, all trips and CD-695A auto starts associated with the following equipment are bypassed:
A. SACS PUMPS B and D B. SSWS PUMPS B and D C. RHRPUMPB D. RCIC system (The backup mechanical overspeed trip of 125% rated speed is still provided (RCIC). This turbine trip will close the trip and throttle valve (HV-4282). This overspeed trip must be locally reset to allow relatching of the Turbine Trip Throttle valve. The limitorque must be manually run to the full closed position to relatch the valve. After locally resetting, valve control is restored to the Control Room.
3.1.9 The following precautions and limitations are related to the RHR System:
CD-847E A. HV-FO04B RHR PMP SUPP POOL SUCT MOV will drain the CD-696A Reactor Vessel to the Suppression Pool if opened in shutdown cooling, CD-847E 1. HV-F024B RHR LOOP TEST RET MOV will drain the Reactor CD-695A Vessel to the Suppression Pool if opened in shutdown cooling.
CD-847E C. HV-F0079 B RHR PMP MIN FLOW MOV will drain the Reactor CD-695A Vessel to the Suppression Pool if opened in shutdown cooling.
CD-847E D. Opening HV-FO07B to lower the reactor water level is only to be CD-695A done when absolutely necessary. Opening this valve provides the potential for an uncontrolled drainage path from the Reactor to the Suppression Pool.
E. Opening HV-F009 SHUTDOWN COOLING lNBD ISLN MOV may cause a decrease in Reactor water Level.
Hope Creek Page 4 of 61 Rev. 10 APR-02-2002 10:00 6093393997 P. 07
SENT BY:TRAINING CENTER ; 4 2 ; 10:06 ; PSE&G NUCLEAR-- 6103375320;# 8/12 FIELD USE
,LL ACTIVE ON-THE-SPOT CHANGES MUST BE AITACHED FOR 20020402 3.1.9. F. Do not allow discharge of water > 120°F to Liquid Radwaste System.( TI-4401, Disch to LRW-DISCH TEMP)
RCIC 3.1.10 The following precautions Ind limitations are related to the System:
A. To prevent RCIC Turbine exhaust piping and check valve vibration to problems, RCIC Turbine speed should be rapidly increased
> 2150 RPM.
should B. To prevent possible bearing damage, RCIC Turbine speed be limited to > 2150 RPM.
HPCI is CD-012Z 3.1.11 Ithe rate of rise of the Reactor Pressure Vessel level indicates to be injecting and the Control Room is unmanned, HPCI will have no longer required or prior to tripped using Attachment 8 when
- 8) may exceeding the high level trip (Level 8). The high level trip (Level not function in the event a fire occurs in the relay room.
CD-847E 3.1.12 When the RSP transfer switch is placed in EMER, the LOW REACTOR LEVELJHIGH REACTOR PRESSURE isolation logic associated with shutdown cooling is bypassed.
A. High Reactor pressure will prevent opening of the shutdown cooling valves but will not isolate the valve if pressure rises above setpoint.
3.1.13 Operation of SRVs at low Reactor pressures (below 700 PSIG) may result in failure of an SRV to reclose when required.
level 3.1.14 During plant Cooldown/Depressurization, monitor similar Rx water instrumentation for significant deviation, indicating possible reference have line De-gasing. Also, terminate all maintenance activities which the potential of draining the Rx vessel.
CD-473G 3.1.15 The HPCI Exhaust Diaphragms, if ruptured, relieve exhaust steam directly to the Torus Chamber area creating a severe personnel safety hazard. ENSURE all personnel are clear of the Torus Chamber Area prior to starting HPCI and during system operation, except as part of SA-AP.ZZ-0051 (Q) walkdowns.
3.1.16 Contact Radiation Protection prior to performing venting and/or draining in this procedure. The individual(s) performing the venting and/or draining shall obtain instructions and approval from the RP Shift Technician or RP Supervisor.
Page 5 of 61 Rev. 10 Hope Creek APR-02-2002 10:00 6093393997 P. 08
SENT BY:TRAINING CENTER ; 4 2 ; 10:07 ; PSE&G NUCLEAR 6103375320;# 9/12 ALL ACTIVE ON-THL-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 n% ,vr-%u .".--wUU jwI 3.2 Q 3.2.1 The precautions and limitations in the appropriate SOP's will be applicable when the SOP's are used in this procedure.
to 3.2.2 AP21I(BP21i) A(B) FUEL POOL COOLING PUMP(s) may trip due Cooling Demineralizers should be placed in channel transfers, Fuel Pool hold as required.
4.0 EQUIPMENT REQUIRED 4.1 Sound powered phones 4.2 Radios 4.3 Keys for Security Doors and MCC Keylock Switches (located in key cabinet, Remote Shutdown Panel Room)
Page 6 of 61 Rev. 10 Hope Creek APR-02-2002 10:01 609339399? 95% P. 09
SENT BY:TRAINING CENTER 4 2 ; 10:07 ; PSE&G NUCLEAR-, 6103375320;# 10/12 MUST BE ATTACHED FOR I ItLU U*t ALL ACTIVE ON -THE-SPOT CHANGES 20020402 5.0 PROCEDURE 4qI s.0 A. Initial each step upon completion of the step.
B. Refer to Attachment 2 for RSP redundant instrumentatlon/controls.
C. Refer 1o Attachment 3 for placing 'A' Loop RHR in Suppression Pool Cooling.
D. Refer to Attachments 6 and 7 for plant communications a remote Information. When dispatching an operator to the operator with a sound shutdown control station, provide powered phone or radio to assist with communication.
Initials 5.1 Eslablish Control from Outside the Control Room 5.1.1 Ensure that all prerequisites have been satisfied lAW Section 2.0 of this procedure.
NOTE 5.1.2 are still A. If the Reactor was not scrammed and the MSIVs the Main Turbine Bypass open, the Feedwater System and pressure at this Valves may be regulating Rx level and Rx time.
will B. Opening the circuit breakers listed in step 5.1.2 the plant, and deenergize the RPS busses, scramming deenergize the NSSSS busses, closing the MSIVs.
are located In C. 10C410(10C411) RPS PWR Dist. Panels A(B) the Control/DG Bldg. El. 54'.
Rev. 10 Hope Crook Page 7 of 61 APR-02-2002 10:01 6093393997 95Z P. 10
SENT BY:TRAINING CENTER ; 4 2 ; 10:07 ; PSE&G NUCLEAR 6103375320;#11/12 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HC.0 P-IU.ZZ-O008(Q)
Initials CD-987X 5.1.2 If the Reactor could not be scrammed prior to Control Room evacuation OPEN the following circuit breakers A CB2A, CB3A, CB5A, CB7A and CBSA (RPS PWR DIST PNL A 10C41 0).
B. CB2B, CB3B, CB5B, CB76 and CBB8 (RPS PWR DIST PNL B 10C41 1).
5.1.3 Ifthe Rx scram was not verified prior to evacuating the Control Room, send an operator to each HCU to check local nitrogen side pressure indicators for a low (< 800 psig) pressure indication verifying the Reactor Scram.
5.1.4 Notify Chemistry to verify that the HydrogeniOxygen System has tripped lAW HC.C H-SO.AX-O001 (0) 5.1.5 Upon arriving at the RSP, monitor the RSP system indications and check specifically for the following:
A. REACTOR VESSEL PRESSURE PR-7853D (between 905 psig and 1045 psig)
CAUTION 5.1.5.5 CD-01 2Z Ifthe rate of rise of the Reactor Pressure Vessel level indicates HPCI is Injecting and the Control Room is unmanned, HPCI will have to be tripped using Attachment a when no longer required or prior to exceeding the high level trip (Level 8). The high level trip (Level 8) may not function in the event a fire occurs in the Relay Room.
5.1.5. B. REACTOR VESSEL LEVEL LR-7854 (between 12.5" a 54")
C. RCIC System status (standby or auto-initiated)
D. PSV-FO13F,H,M SRV status (standby gr cycling open/closed)
E. SUPPRESSION CHAMBER WATER TR-3647J (and M) (average less then 950F)
Hope Creek Page 8 of 61 Rev. 10 APR-02-2002 10:02 6093393997 P. 11
SENT BY:TRAINING CENTER ; 4 2 ; 10:08 ; PSE&G NUCLEAR-' 6103375320:#12/12 ALL. ALUIVE ON--TIHE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200204.02 MGU.Ur-sV.4z-4,UUG; w Mt;.uv-Iu.* In-tual s1 5.1.5. F. DIESEL GENERATOR 1A(BC,D)G400 TRIP/CLOSED Status (closed ff a loss of offsite power has occurred).
5.1.6 If a loss of offsite power has occurred, send an operator to the Diesel Generator Remote Control Panel (Aux. Bldg El. 130') to monitor Diesel Generator operation and implement HC.OP-AB.ZZ.0135(Q), Loss of Offsite Power, concurrent with this procedure.
CD-462Y 5.1.7 P the following RSP switches to EMER:
A. CH "A" TRANSFER B. CH "i" TRANSFER C. CH "C" TRANSFER D. CH 'D" TRANSFER E. CHANNEL "NON-lE" TRANSFER F. PMP BP202 XFR - B RHR PUMP G. PMP BP502 XFR - B SERV WIR PUMP H. PMP OPS02 XFR - D SERV WTR PUMP I. PMP BP210 XFR - B SACS PUMP J- PMP DP210 XFR - D SACS PUMP K MOTOR BK400 TRANS L. MOTOR BK403 TRANS 5.1.8 Ensure appropriate automatic actions have occurred on the transfer lAW Attachment 1.
Hope Creek Page 9 of 61 Rev. 10 APR-02-2002 10:02 6093393997 P. 12
OFNT RY:TRAININQ rFNTFR :4 9 : 10):09 ; PSE&G NUCLEAR- 6103375320;# 1/14 Nuclear Taining ep
_ - - - I- -- - - -
m m m - m m m II m - m m m To: From:
NAME NAME PHONE COMPANY PSEG Nuclear, LLC DEPARTMENT Nuclear Training Department
[2/Nuclear Training FAX 244 Chestnut Street Salem, NJ 08079 Fax: (856)339-3997 PHONE El Nuclear Services Team 244 Chestrut Street Salem, NJ 08079 Fax; (856) 339-2382 TOTAL PAGES (INCLUDING COVER) o PSEG Nuclear LLC APR-02-2002 10:03 6093393997 95% P. 13
SENT BY:TRAINING CENTER ; 4 2 ; 10:09 ; PSE&G NUCLEAR-' 610U3757832 0;# 2/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HOPE CREEK GENERATING STATION Page I of 4 HC.OP-IO.ZZ-008(Q) - Rev. I I SHUTDOWN FROM OUTSIDE CONTROL ROOM USE CATEGORY: I REVISION
SUMMARY
REV. II
- 1. Added initials blocks to Sections 2 and 3 and Attachments 3, 4, 5 and 8.
- 2. Changed substeps in 4.1. 10 of Attachment 4 to bullcted steps.
- 3. Added page numbers specific to thati attachment to each attachment (also to TOC)
- 4. Changed Step 8.2 in Attachment 8 to Note 8.1.
- 5. Modified Step 4.1 .7.E to notify Radwaste of impending receipt of water at 200'F.
- 6. Edited Attachment 6 to improve ease of reading; rev bars not used.
- 5. Deleted BC-IIV-F0i 1B, F026B, F052B and IIV-4429 from step B.2.1 of Attachment I since these valves arc permanently out of service, with power removed.
- 6. Corrected Valve number from fHV-F0221" to (IV-F022 in B.2.2.
- 7. Incorporated revision request OP 97-1600 IPR 970630279, PR 9707022471 to correct the valve nuinbcr from IEG-HV-2496B TO IEG-HV-2491A in Attachmeti 3, Step 3.3.4.
- 8. Incorporated revision request OP 96-1601 to add clarity to the operation of Attachment 3, Steps 3.5.1. 3.5.2, 3.5.3, 3.5.6, and 3.5.7. Specific direction is given to operate the key switch at the respective breaker vice the breaker itself to conduct valve manipulations.
Also addresses portion of OP 95-0352 [PR 960723097]
(Continued)
IMPLEMENTATION REOUIREMENI$
Effective Date )
APrROVED: ... . -7 rations Manager Date APR-02-2002 10:03 6093393997 P. 14
SENT BY:TRAINING CENTER 4 2 ; 10:09 ; PSE&G NUCLEAR- 6t03375.32L);# 3/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST O3F ATTACHED FOR FIELD USE 20020402 HOPE CREEK GENERATING STATION Page 2 of 4 HC.OP-IO.ZZ-0008(Q) - Rev. II SHUTDOWN FROM OUTSIDE CONTROL ROOM USE CATEGORY: I RIEVISION
SUMMARY
- CONTINUATION SHEET REYýBl
- 9. Incorporated revision request OP 96-1614 and OP 97-0083 [PR 9611222661 by adding the following:
Removed Prerequisites 2.2 which stated Establish communications between each local panel or brcakcr go the Remote Shutdown Panel as the steps are performed. The content of this prerequisite is captured in Note 5.0-1).
Changed the discharge temperature limit for liquids going to liquid radwaste in Precaution 3.1.9 and Steps 5.10.2, 5.10.2.A and 5.11.2.B and Caution 5.11.2 frvlu 120 degrees to 200 degrees. This coincides with 1IC.OP-SO.BC-000 1(Q).
Also addresses portion of OP 95-0352, IPR 9607230971 Changed the maximum SACS temperature from 100 degrees to 95 degrees in step 5.9.3.
This coincides with HC.OP-SO.EG-0001(Q), Step 3.2.8.
- 10. Incorporated revision request OP 96-0057 to add the noun names of the valves operaied in Attachment 1, Section D and to change Section D.2.7 from HV-2494D to HV-2494B and from HV-2355D to HV-2355B for proper valve identification.
- 11. Incorporated reviewer comments JPR 970514283, PR 9707022471 to include the following changes:
Added Note 5.1.8 stating.. If running, chillers BK400 and BK403 will trip when their respective transfer switches are taken to emergency.. to note expected system response.
Added Step 5.12.I.D stating.. 1'IfWIC is running from the RSP, ensure the RCIC flow controller in the M('R is matched with actual RCIC flow., to ensure continuity of monitored readings prior to shifting control from the RSP to the MCR.
Changed breaker number 52-40104 to 52-40304 in Attachment 3, Step 3-4.2 to indicate the correct breaker associated with "C" SACS pump.
"* Added a provision to reset PCIS in step 5.12. !.C.
"* Deleted Attachment 8, Step g.1.2 open circuit breaker 26, HPCI INVERTER VERTICAL.
BOAIR 10C620 bccause it is a spare.
(Continued)
APR-02-2002 10:03 6093393997 P. 15
SENT BY:TRAINING CENTER ; 4 2 ; 10:10 ; PSE&G NUCLEAR-A 610337532C :# 4/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHFD FOR FIELD USE 20020402 HOPE CREEK GENERATING STATION Page 3 of 4 HC.OP-IO.ZZ-0008(Q) - Rev. II SHUTDOWN FROM OUTSIDE CONTROL ROOM USE CATEGORY: I REVISION
SUMMARY
- CONTINUATION SHEET REV. 11
- 12. Incorporated revision request OP 95-0352 [PR 9607230971 to change the following:
"* Deleted Step 5.8.2.D so that the procedure would leave HV-FO07B in the open position which is its nomnal position.
"* See revison summary item 8.
- See revision summary item 9, second bullet.
- Revised direction on Attachment 3 at step 3.5.6, and 3.5.7 (previously Note and step 3.5.5) by: converting the note into a step; clarify "hook up' statement in noLe to mean hooking up of sound powered phone equipment, and added a reference to Attachment 7; clarified instrument lo bc used in lower relay room; and converted tihe statcd indication to achieve, to one commensurate with the range of the instrument (previously the step directed operator to achieve 83.3% indication, yet the instrument range is from 0 to 30%. The new value listed is 83.3% of 30 (25%).
- 13. Based on Department reviewer comments, the following changes were made:
- Removed "Control Room", and added "RSP" to Note 3.1 .8.D
- Split precaution 3.1.13 into a note and a step.
,, Delctcd Precaution 3.1 .9,C. This precaution was adequately addressed in 3.1.9.D.
- Added Letters (A, B, C, D) to Note 5.0 to differentiate statements.
- Removed implied interlock from 5.2. .G lIr an auto swap of RCJC suctions on a high suppression chamber level.
a Corrected nomenclature of instrument BI! l-N652A to match field labeling, and as refered to in Attachments 2, 3, and 7.
- Corrected title of Attachment 6 as refered to in the table of contents.
- Added "r.ocalr - step 5.3.I.F.
- Added steps 5.5. I.C and 5.5.1.E to verify repositioned I X Inlet Valves.
This action is similar to that tfund in SOP for placing system in service.
- Changed "CLOSE" to "REPOSITION" and added "as necessary" in step 5.8.2.E.
- Added steps 5.4.1 .D and 5.4. .J to open the HV-2 197 Backwash Valvcs. This action is similar to that found in SOP for placing system in Fwrvice.
"* Added direction to ensure the HV-FO03B 13 RHR I IX OUTLET MOV is open in step 5.8.1 .G.
"* Added new direction at step 5.11.3 for utilizing the A R.HR for Torus Level reduction if B R.HR is in SI)C (Continued)
APR-02-2002 10:04 6093393997 P. 16
SENT BY:TRAINING CENTER ; 4 2 ; 10:10 ; PSE&G NUCLEAR- 6103375320:# 5/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HOPE CREEK GENERATING STATION Page 4 of 4 HC.OP-IO.ZZ-0008(Q) - Rev. 11 SHUTDOWN FROM OUTSIDE CONTROL ROOM USE CATEGORY: I REVISION
SUMMARY
- CONTINUATION SHEET REV, 11
- 13. Based on Departmcnt rcviewer comments, the following changes were made:
"* Changed "HPCI" to "RCIC" in precaution 3.1.16.
"* Modified step 4. 1.8 of Attachment 4 to match direction for operation of 11V-17015 found in RHR SOP.
- 14. Incorporates revision request OP-97-0443 (BP 970811169) by adding Precaution 3.1.12 and step 5.1.7 to warn of the potential effect a fire in the relay room would have on BJ-lIV-F008.
- 15. Incorporates revision rcquest B3P 970908166 by adding precaution 3.1.18 and NOTE 5.8 to warn of a possible water hammer condition if the HV-F024A(B) and/or HV-F027A(B) arc open and the associated RHR pump is stopped.
APR-02-2002 10:04 6093393997 P.$17
6103375320:# 6/14 SENT BY:TRAINING CENTER ; 4 2 ; 10:10 ; PSE&G PSE&G NUCLEAR-NUCLEAR-I 6103375320;# 6/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HC.O]P-IO.Zz.01108(Q)
SHUTDOWN FROM OUJTSIDE CONTROL ROOM Inaitab TABLE OF CONTENTS 5EEDDA PURPOSE.... PAGE 1.0 PUPS .... H... .. ........ ,°.,,, ................................................... o...
2.0 PREREQUISITES .............................................. . ...........
3,0 PRECAUTIONS AND LIMITATIONS ............................... 3 4.0 EQUIPMENT REQUIRED ..................... 6 5.0 PROCEDURE.................................................7 5.1 Establish Control from Outside the Control Room ................. 7 5.2 Place RCIC in Sevice ........................................................................... 10 5.3 Remove RCIC From Service .................................................................. 13 5.4 Place SSW Loop Rin Service ................................. 14 5.5 Place SACS Loop B in Service ............................... ............. 15 5.6 Place Control Area Chilled Water in Service ........................................ 16 5.7 Place the IE Panel Room Chilled Water System in Service .................. 16 5.8 RHR Suppresion Pool Cooling Mode ........................... 17 5.9 Plant C ooldow n ........................................................................................ 18 5.10 Lower Reactor Water Level ....................... 21 5.11 Lower Suppression Pool W ater Level .................................................... 22 5.12 Control Room Re .entry .................................................................... 24 6.0 RECORDS................ ............................. ............................ 25 I
7.0 REFE RE N C ES ....................................................................................................... 25 Hope Creek Page I of 64 Rev. 11 APR-02-2002 10:04 6093393997 P. 18
SENT BY:TRAINING CENTER ; 4 2 ; 10:11 ; PSE&G NUCLE? R-4 6103375320;# 7/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HC.OP-IO.Z1Z0008(Q)
Initial SHUTDOWN FROM OUTSIDE CONTROL ROOM TABIE OF CONTENTS SE CTION 11L A'ITACHMENTS Attachment 1 RSP Transfir Switch Automatic Action .................................................. 26 Attachment 2 RSP Redundant lnstruzm tation/Controls ............................................... 32 Attachment 3 A RHR Loop Suppression Pool Cooling ............................ 38 Attachment 4 B R-HR Loop Shutdown Cooling Operation ............................................ 42 Attachment 5 Reactor Coolant System Temperaturc/Pres-ure Data ............................... 47 Attachment 6 Communications And Emergency Lighting Systems For Safe Shutdown Areas ......................................................................................................... 51 I Attachment 7 Remote Shutdown Communication Systcm ............................................. 57 Attachment 8 HPCI Shutdown From "A"Diesel Generator Control Room ................... 60 Attachmnen( 9 Vessel Level Instrumentation Temperature Com pensation Curves ......................................................................... 61 Hope Creek Page 2 of 64 Rev. 11 APR-02-2002 10:05 6093393997 95% P. 19
SENT BY:TRAININO CENTER ; 4 2 ; 10:11 ; PSE&G NUCLEAR- 6103375320;# 8/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HC.OP-IO.ZZ.0008(Q)
In1itials SHUTDOWN FROM OUTSIDE CONTROL ROOM START TIME DATE -BY-.....
TERMINATION TIME DATE BY COMPLETION TIME DATE BY 1.0 PlMECIS This procedure provides guidelines for the shutdown of thc plant from outside the Control Room.,
AND for re-establishing control in the Control Room.
2.0 PI QREOISITES 2.1 I IC.OP-AB.ZZ-0 130(Q), Control Room Evacuation, completc if possible.
3.0 PRECAUTrIONS AND LIMITATIONS 3.1 Administrativ 3.1.1 This procedure is to be used as a guidelinc for shutdown of the plant from outside the Control Room. It is NO'[ required that each section/step be performed in precise sequence a.s long as the sections/steps are performed in a timely manner, in keeping with the intent of this procedure.
3.1.2 In the event plant conditions require a delay during performance of this procedure, the Senior Nuclcar Shift Supervisor/Nuclear Shift Supervisor (SNSSINSS) should retain this procedure Until it is continued OR terminated.
3.1.3 IF it is terminated prior to completion, the SNSS!NSS should note the reason, time, AND date of termination on this procedure.
3.1.4 For any unit scram, the Event Classification Guide should be referred to for the appropriate classification AND notifications..
3.1.5 Reactor Coolant System temperature AND pressure requirements of f/S 3.4.6.1 shall be complied with.
Hope Creek Page 3 of 64 Rev. 11 APR-02-2002 10:05 6093393997 P. 20
- 4 2
- 10:12 ; PSE&G NUCLEAR- 6103375320;* 9/14 SENT BY:TRAINING CENTER ALL ACTIVE ON-THE-SPOT CHANGES MUST 817 ATTACHED FOR FIELD USE 20020402 HG.OP-IO.ZZ-0008(Q) 3.1.6 The Reactor Vessel AND Ilead Flange temperature limits of "T/S3.4.6. 1.d shall be complied with.
3.1.7 The Suppression Chamber tcmpcrature requirements of T/S 3.6.2.1 shall be complied with.
3.1.8 When RSP Transfer Switches are placed in EMER, all trips and auto starts associated with the following equipment are bypassed:
ICD-904B, CD-695Al A. SACS PUMPS B and D B, SSWS PUMPS B and D C. RHIR PUMP B NMQTJ 3.1.8.D The RCIC backup mechanical overspeed trip of 125% rated speed is still provided. This turbine trip will qlos& the trip and throttle valve (HV-4282). This overspeed trip must be locally reset to allow relatching of the Turbine Trip Throttle valve. The limitorque must be manually run to the full closed position to relatch the valve. After locally resetting, valve control is restored to the RSP.
D. RCIC system 3.1.9 The following are related to the RHR System:
A. 1L opened during Shutdown Cooling operations,
.*t HV-FO04B RHR ]IMP SUPP POOL SJCT MOV will drain the Reactor Vessel to the Suppression Pool
[CD-847E, CD-695AI B, If opened during Shutdown Cooling operations, THEN HV-F024B RIIR LOOP TEST RET MOV will drain the Reactor Vessel to the Suppression Pool.
ICD-847E, CD-695AJ Continued Next Page Hope Creek Page 4 of 64 Rev. I I APR-02-2002 10:06 6093393997 P. 21
SENT BY:TRAINING CENTER ; 4 2 ; 10:12 ; PSE&G NUCLEAR-* 6103375320;#10/14 HC.OP-IO.7zZ-OOO(Q) 3.1.9 (Continued)
C. Opening HV-F007B to lower the reactor water level is to be done QhLY when absolutely necessary. Opening this valve provides the potential for an uncontrolled drainage path from the Reactor to the Suppression Pool.
(CD-847E, CD-695Aj D. Opening 1IV-F009 SI-UTDOWN COOLING iNBD ISLN MOV may cause a dccrease in Reactor water Level.
E. Water >200'F should b.(1 be discharged to the Liquid Radwaste System (11-4401, Disch to LRW-DISCH TEMP).
3.1.10 The following precautions AM limitations are related to the RCIC System:
A. To prevent RCIC Turbine exhaust piping AND check valve vibration problems, RCIC Turbine speed should be rapidly increased to ? 2150 rpm.
B. To prevent possible bearing damage, RCTC Turbine speed should be limited to 2150 rpm.
3.1.11 IF the rate of rise of RPV level indicates UIPCI is injecting AND the Control Room is unmanned, :IMBN-N IfPCI will have to be tripped using Attachment B when no longer required OR prior to exceeding high level trip (Level 8). High level trip may NOT function in the event a fire occurs in the relay room.
[CD-0l2Z1 3.1.12 A fire in the relay room can cause BJ-HV-FOO8, HPCI TEST BYPASS TO CSf to spuriously open. If HPCJ suction is aligned to the Suppression pool, this valve must be closed from it's MCC - 1OD251103.
NOTE 3.1.13 High Reactor pressure will prevent opening of the shutdown cooling valves but will LQI isolate the valve IL pressure rises above setpoint.
3.1.13 1)YH*t4 the RSP transfer switch is placed in EMER, IJ.[iE the LOW REACTOR LEVEL/HIGI I REACTOR I'R.ESSI IRE isolation logic associated with shutdown cooling is bypassed.
[CD-847EJ Hope Creck Page 5 of 64 Rev. I I APR-02-2002 10:06 6093393997 95% P. 22
SENT BY:TRAINING CENTER ; 4 2 ; 10:12 ; PSE&G NUCLEAR-- 6103375320 ;#11/14 AI.[. ACTIVE ON-THE-SPOT CHANGFS MUST BE ATTACHED FOR FIELD USE 200204102 -COP-IOZZ-0008(Q) 3.1.14 Operation of SRVs at low Reactor pressures (below 700 PSIG) may result in failure of SRV(s) to reclose when required.
3.1.15 During plant Cooldown/Deprcussurization, similar Rx water level instrumentation should be monitored for significant deviation.
indicating possible reference line degassing.
In addition, all maintenance activities having the potential for draining the Rx vessel should be terminated.
3.1-16 IF ruptured, the RCIC Exhaust Diaphragms relieve exhaust steam directly to the Torus Chamber area, creating a severe personnel safety hazard.
ENSURE that all personnel arc clear of the Torus Chw-nber Area PER1O to starting RCIC AND during system operation, except as part ofSA-AP.ZZ-0051(Q) walkdowns. ICD-473GJ 3,1.17 Radiation Protection should be contacted prior to performing venting AND/OIL draining. The individual(s) performing the venting ADi1LQ draining should obtain instructions ANID approval from the RP Shift Technician or RP Supervisor, 3.1.18 IF at any time a situation develops whereby HV-F024A(B) and/or HV-F027A(B) are open with the associated RHR pump not in operation, a potential system drain down will occur. A subsequent start of the RHR pump following this situation will cause water hammer.
Therefore, IF the valves are open AND the pump either trips or is not running IHL ENSURE both valves are closed, AND PERFORM a system fill & vent PRIOQ to starting the RHR pump. [970908166) 3.2 01cr 3.2.1 The precautions h limitations in the appropriate SOPs will be applicable when the SOPs are used in this proecdurc.
3.2.2 AP21 1(BP21 1) A(B) FUEL POOL COOLING PUMP(s) may trip due to channel transfers. Fuel Pool Cooling Demincralizers should be placed in hold, as required.
4.0 EQUIPMENT REQITI:RI)
- Sound powered phones
- Radios
- Keys for Security Doors AND MCC Keylock Switches (located in key cabinet, Rcrnotc Shutdown Panel Room)
Hope Creck Page 6 of 64 Rev. 11 RPR-02-2002 10:0? 6093393997 P. 23
SENT BY:TRAINING CENTER ; 4 2 ; 10:13 ; PSE&G NUCLEAR- 6103375320;#12/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USr 20020402 HC.OP-I O.ZZ-0008(Q)
Initials 5.0 PRQCEDE NOTE 5.0 A. Each step should be initialed upon completion of the step.
B. Attachment 2 should be referred to for RSP redundant instrumentation/
controls.
C. Attachment 3 should be referred to for placing 'A' Loop RHR in Suppression Pool Cooling.
D. Attachments 6 and 7 should be referred to for plant communications information. When dispatching an operator to a remote shutdown control station, the operator should be provided with a sound-powered phone QE radio to assist with communication.
5.1 Establish Control frm Olutsidc the Control Room 5.1.1 ENSURE that all prerequisites have been satisfied lAW Section 2.0 of this procedure.
N.4_IE 5.1.2 A. If the Reactor was NM scrammed ANDŽ the MSIVs are still open, then the Feedwater System AND the Main Turbine Bypass Valves may be regulating Rx level AND Rx pressure at this time.
B. Opening the circuit breakers listed in step 5.1.2 will deenergize the RPS busses, scramming the plant, AND deenergize the NSSSS busses, closing the MSIVs.
C. 10C410(10C411) RPS PWR Dist. Panels A(B) are located in Control/DG Bldg.
El. 54'.
5.1.2 It the Reactor was NOT1 scrammed prior to Control Room evacuation, flUF,1 OPEN the fbllowing circuit breakers: ICD-987X1 A. CB2A, CB3A, CB5A, CB7A AND CB8A (KPS PWR DIST PNL A 10C410).
B. CB2B. CB3B, (I'SB. CB7B AND CS38 (RPS PWR DIST PNL. B 10C41 1).
Hope Creek Page 7 of 64 Rev. 11 APR-02-2002 10:07 609339399? 95% P. 24
SENT BY:TRAINING CENTER ; 4 2 ; 10:13 ; PSE&G NUCLEAR- 6103373320;#13/14 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20020402 HC.OP-IO.ZZ-0008(Q) 5.1.3 IE the Rx scram was N-I verified prior to evacuating the Control Room, THEN VERIFY Rods Full In. (SPDS/CRIDS (TSC) QR R MCS Activity Control Cards DR Other).
5.1.4 NO'riFY Chemistry to verify that the Hydrogen/Oxygcn System has tripped lAW HC.CH4-SO.AX-000I(Q).
5.1.5 Upon arriving at the RSP, MONITOR the RSP system indications AND CHECK speci fically for the following:
A. REACTOR VESSEL PRESSURE PR-7853D (905 - 1045 psig)
GAyJulQ 5.1 .5.B IEthe rate of rise of RPV level Indicates HPCI is injecting ANU the Control Room Is unmanned, tEN HPCI will have to be tripped using Attachment 8 when no longer required DR prior to exceeding the high level trip (Level 8). The high level trip may NOT function in tho event a fire occurs in the relay room. [CD-012Z]
B. REACTOR VESSEiL LEVEL LR-7854 (12.5 - 54 ")
C. RCIC System status (standby (R auto-initiated)
D. PSV-YOI3F,1T.M SRV status (standby .2 cycling opcn/closcd)
E. SUPPRESSION CHAMBER WATER TR-3647J (AND M) (average less than 95'F)
F. DIESEL GENERATOR IA(B,C,D)G400 TRIP/CLOSED Status (closed IE a loss of offsile pow"r has occurred).
5.1.6 F-a loss of offsite power has occurred, TfHiEN SEND an operator to the Diesel Generator Remote Control Panel (Aux. Bldg El. 130') to monitor Diesel Generator operation. AND IMPLEMENT F-C.OP-AB.ZZ-0135(Q), Loss of Offsite Power, concunrnt with this procedure.
Hope Creek Page 8 of 64 Rev. II APR-02-2002 10:07 6093393997 P. 25
NC.NA-AP.zz-oooi(w FORM-1
-D Page of 0
PROCEDURE COMMENT FORM TO: POCRT1EmE WRITRR Dept: BQ cc: ________________Conraent Due Date; ____
FROM: ___________Dept:
0)
DOCUMENT NO., REVISION, AND TITLE:
HC.OP-IO.ZZ-0008(Q), Rev. 11 Page _of_
SHUTDOWN FROM OUTSIDE CONTROL ROOM Pageil'ara 1Comments or Recxomnendations Number Comment Disposition (A neqat!ive disposition requires justifi cation)
-. 3 A73 T P 7- Ss
'qL'zc T A(,-CU-C/ A, U.I, aAddd o. e CAlJ 61f4-4 rt~(4 9J'(
tClS F4AK, uf- .7( el P AlqC FlLt Z Ey(,I... CAz"J9., ;qA l.0 REVISWEJR:
DATE: ~i DrsPOSITIoNED BY:
DATE; j1 ~i Nucrear Common Page 42 of 50 0 44
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