ML15009A252

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Post Exam Comments and Technical References (Folder 1)
ML15009A252
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Site: Salem  PSEG icon.png
Issue date: 12/23/2014
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Salem December 2014 NRC Exam RO Question #61 Given the following conditions:

- Unit 2 is operating at 100% power with an identified small fuel pin leak.

- A 5 gpm tube leak occurs on 22 SG.

Of the following, which is the only radiation monitor that will NOT show a change from this tube leak?

a. 2R19B, 22 SG Blowdown.
b. 2R46A, 22 Main Steam Line.
c. 2R15, Condenser Air Ejector.
d. 2R41D, Plant Vent Release Rate.

Correct Answer: b Proposed Change: Delete question based on no correct answer Justification: The R46 monitors are designed to provide continuous monitoring of high-level, post-accident releases of radioactive noble gases via the safety-relief valves, atmospheric dump valves, and auxiliary feed pump turbine and are capable of functioning both during and following an accident. The monitors are designed to meet the requirements ofNUREG-0737 II.F.1 and the intent ofRG 1.97. A range of lE-1 to 1E3 micro Ci/cc is required. The monitor's alarm function (7 mR!hr) is used in the EOPs to identify a SGTR event/EOP entry point, and identify which SG has ruptured.

However, when discussing the post-accident ability of the monitor, the most recent DCP to update the R46 monitors installed in 2005/2006, DCP 80057587, says on page 22 of 109.... "However, these monitors will exhibit a response to N-16, but typically only at levels corresponding to leak rates in the gallon per minute range. Because of the N-16 response, these monitors can provide a clear indication of a SGTR if the rupture occurs while the unit is at power." (my emphasis)

Since the 2R46A will exhibit a response to the SGTL described in the stem (5 gpm at 100% power), as well as the remaining three distracters, we believe there is no correct answer to this question, and request it be deleted.

CHANGE NO:

---'8::...::0:...::.0..::..57::...::5:..:::8...:....7 ____________ REVISION NO.:

1 gamma rays (such as Xe-133 with an 80 KeV photon) is minimal. These monitors would only respond if there was a significant RCS (Reactor Coolant System) source term. As a result, these monitors cannot be used for low level leak rate d~tection and are limited to post-accident assessment of significant releases. However, these monitors will exhibit a response to N-16, but typically only at levels corresponding to leak rates in the gallon per minute range. Because of the N-16 response, these monitors can provide a clear indication of a SGTR if the rupture occurs while the unit is at power. However, once the reactor trips the readings will typically return to normal levels.

Because they are not sensitive to small primary-to-secondary leak changes, Main Steam Line Monitors installed to meet RG 1.97 requirements will provide no useful primary-to-secondary leak trend information.

The off-line sampling arrangement currently used is not discussed but would have a considerably lower N-16 response due to sample transport time and the short N-16 half-life of -7 seconds. The sensitivity of the existing arrangement to low energy gammas is somewhat higher due to less pipe thickness.

2.10 Existing SAP Classification The existing R46 Monitors are classified in SAP as EQ Mild; Safety Related; QA Required Yes; Seismic Classification 1; Safety Significance 002; Component Classification Data Q07 (Safety Sy M/1/Cont Pen Cmp ); Safety Class/Quality Grp Code 1 E; Safety Class (Nuclear) 1 E; Tech Spec Accpt Criteria lnst 2.11 Steam Line Penetration Room Environmental Information Per S-C-ZZ-SDC-1419, Salem Generating Station Environmental Design Criteria, the mechanical steam line penetration rooms have the following design-basis-accident environmental conditions: 5.4 PSIG, 467 DEG F, 100% RH, 187 mR/Hr, 5.63 RAD, and 356RTID.

2.12 Design Bases Summary The R46 monitors are required to provide continuous monitoring of high-level, post-accident releases of radioactive noble gases via the safety-relief valves, atmospheric dump valves, and auxiliary feed pump turbine and are to be capable of functioning both during and following an accident. The monitors are designed to meet the requirements of NUREG-073711.F.1 and the intent of RG 1.97. A range of 10-1 to 103 f.lCi/cc is required although equivalency to Xe-133 not explicitly required by RG 1.97. RG-1.97 acknowledges that monitors placed adjacent to main steam lines would not be able to detect low energy gamma emissions and a lower energy response of 500 Kev is acceptable, as long as the concentration of lower energy-emitting isotopes can be Page 22 of 109 NC. CC-AP.ZZ-0080(Q). Rev 15

Salem December 2014 NRC Exam SRO Question #6 Given the following conditions:

- Unit 1 is operating at 100% power.

- A breaker fault occurs on the 2-6 500 KV breaker.

- The 2-6 500 KV breaker does NOT trip, but should have.

15 seconds after the breaker failure, Unit 1 has NOT tripped.

Which of the following identifies how the Unit 1 CRS should proceed?

a. Direct the RO to manually trip the reactor and go to EOP-TRIP-1, Reactor Trip or Safety Injection. Concurrently with EOP implementation, initiate S 1.0P-AB.LOOP-0001, Loss of Offsite Power, and perform Attaclunent 2, Loss of Group Buses, Part A, Loss of 1E and 1H 4KV Group Buses.
b. Direct the RO to manually trip the reactor and go to EOP-TRlP-1. Concurrently with EOP implementation, initiate Sl.OP-AB.LOOP-0001, and perform, Loss of Group Buses, Part B, Loss of lF and 1 G 4KV Group Buses.
c. EnterS 1.0P-AB.LOOP-0003, Partial Loss of Off-Site Power, then enterS 1.0P-AB.CW-0001 Circulating Water System Malfunction, and perform a power reduction to 83% power or less.
d. Enter Sl.OP-AB.LOOP-0003, then enter Sl.OP-AB.CW-0001 and open the Hood Spray Bypass valves 11-13MC62s.

Correct Answer: d Proposed Change: Accept choice c as correct also.

Justification: Choices c and d both have the correct procedure progression, with choice d having an action directed at step 3.12 of AB.CW with 3 circulators running, with one running on each waterbox. Choice c contains the action to reduce power to :S83% power, which would be performed directly lAW step 3.9 ifboth the 11A & 11B circulators are out of service. Additionally, the Continuous Action Summary, Step 5.0 (page 11 of 32 in AB.CW) states...

"IF AT ANYTIME any ofthe following conditions exist or are approaching:

Condenser L\\T >27°F Condensate suction temperature > 135°F Flashing occurs in the Condenser Hotwell or Condensate Pump suction piping as indicated by erratic Condensate Pump Amp Indication OR erratic SGFP suction pressure indication.

THEN INITIATE a load reduction lAW S1.0P-AB.LOAD-000l(Q),

Rapid Load Reduction, concurrently with this procedure until parameters stabilize within operational limits."

Based on their knowledge of plant response to a loss of circulators, 4 of 8 SRO candidates assumed that condenser L'l T would rise rapidly and exceed the 27°F limit, and based their answer on having to do a rapid load reduction to some power level below 83%. The accompanying simulator charts reflecting a breaker failure on the 2-6 500KV breaker corroborates that rapid rise in condenser temperature, and the necessity for initiating the power reduction. As shown on the Plant Computer screen shot, Rx power was lowered to approximately 81% to reduce Condenser L'lT to 27.1 °F. An additional simulator computer graph shows that a Rx power reduction to 81% power was required to lower condenser L'lT to 27 °F.

S l.oP-AB.CW-0001(Q)

ATTACHMENT 1 (Page 2 of7)

CONTINUOUS ACTION

SUMMARY

4.0 IF AT ANYTIME a load reduction is required to maintain condenser backpressure within the Allowable Operating Region of Attachment 4, Condenser Absolute Pressure Limits, THEN:

NOTE Turbine load reduction ramp rates of ~5%/min are desirable to prevent operation of Steam Dumps, which could degrade the low vacuum/high temperature condition.

The degradation of condenser vacuum and possible loss of Steam Dumps, may lead to the use of the MS 10 atmospheric reliefs for temperature control.

+

INITIATE a Rapid Load Reduction lAW Sl.OP-AB.LOAD-OOOI(Q) concurrently with this procedure until parameters stabilize within the operational limits of Attachment 4.

5.0 IF AT ANYTIME any of the following conditions exist or are approaching:

+

Condenser A'F">21RF *t

+

Condensate suction temperature;;:>: l35°F

+

Flashing occurs in the Condenser Hotwell or Condensate Pump suction piping as indicated by erratic Condensate Pump Amp Indication OR erratic SGFP suction pressure indication.

THEN INITIATE a load reduction lAW Sl.OP-AB.LOAD-OOOI(Q),

Rapid Load Reduction, concurrently with this procedure until parameters stabilize within operational limits.

6.0 IF AT ANYTIME Hotwelllevel is <32 inches AND there are indications of Condensate Pump cavitation, THEN:

6.1 STOP the affected CN Pump.

6.2 INITIATE Sl.OP-AB.CN-OOOl(Q),

Main Feedwater/Condensate System Abnormality, concurrently with this procedure.

Time Time Time Salem 1 Page 11 of 32 Rev. 37

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CIRCULATING WATER SYSTEM MALFUNCTION 1.0 ENTRY CONDITIONS DATE: ___ TIME:

1.1 Any indication of abnormal operation of the Circulating Water System.

1.2 Any indication of a system rupture, in the Circulating Water System, external to the condenser.

1.3 Any indication of a gross condenser tube failure.

1.4 Removal of two or more Circulating Water Pumps from service.

2.0 IMMEDIATE ACTIONS None 3.0 SUBSEQUENT ACTIONS 3.1 INITIATE Attachment 1, Continuous Action Summary.

3.2 Is there a pipe rupture in the Circulating Water System as indicated by any of the following?

+

Report from personnel of excessive leakage from the Circulating Water System.

+

OHA G-43, TURB AREA LVL HI PMP START NO I

YES --> GO TO Step 3.20 v

3.3 Are two or more Circulating Water Pumps 0/S?

YES I

NO-->

GO TO Step 3.26 v

Salem 1 Page 1 of32

{C0361)

Time Time Rev. 37

sl.OP-AB.CW-OOOl(Q)

NOTE Steam Dump operation is blocked to those Condensers which have both Circulating Water Pumps out of service, or CNDSR OUTLET MOVs do NOT indicate open.

3.4 Is at least one Circulating Water Pump operating on each condenser?

NO I

YES-->

GO TO Step 3.11 v

3.5 IF llA and liB Circulating Water Pumps are BOTH out of service, THEN GO TO Step 3.9 NOTE Time When 11 - 14GB4 valves are closed then SGBD sodium is no longer a representative arameter of Steam Generator chemist.

3.6 IF Steam Generator Blowdown is in service to 12 Condenser, Salem 1 AND 12A and 12B Circulating Water Pumps are BOTH out of service, THEN A.

CLOSE ll-14GB4, BLOWDOWN ISOLATION VALVEs.

B.

INITIATE Sl.OP-SO.GBD-0002(Q), Steam Generator Blowdown Operation, to complete Steam Generator Blowdown isolation, as required.

C.

IF 12A and 12B are the ONLY Circulating Water Pumps out of service, THEN CONTINUE to monitor CAS limits and evaluate the need to initiate a load reduction.

D.

PLACE SGBD in service to 13 Condenser as directed by the SM/CRS.

E.

GO TO Step 3.8 Page 2 of32 Rev.37

S l.OP-AB.CW -OOOl(Q)

NOTE When 11 - 14GB4 valves are closed then SGBD sodium is no longer a representative arameter of Steam Generator chemist.

3.7 IF Steam Generator Slowdown is in service to 13 Condenser, AND 13A and 13B Circulating Water Pumps are BOTH out of service, THEN:

A.

CLOSE 11-14GB4,BLOWDOWNISOLATIONVALVEs.

B.

INITIATE S1.0P-SO.GBD-0002(Q), Steam Generator Slowdown Operation, to complete Steam Generator Slowdown isolation, as required.

C.

IF 13A and 13B are the ONLY Circulating Water Pumps out of service, THEN CONTINUE to monitor CAS limits and evaluate the need to initiate a load reduction.

D.

PLACE SGBD in service to 12 Condenser as directed by the SM/CRS.

3.8 Are conditions stable and CAS limits maintained?

NO I

YES-->

GO TO Step 3. 11 v

NOTE Time Turbine load reduction ramp rate of 5%/min or less is desirable to prevent operation of Steam Dumps, which could exacerbate the low vacuum/high temperature condition.

3.9 INITIATE a Rapid Load Reduction to ~83% Reactor Power lAW Sl.OP-AB.LOAD-OOOl(Q), Rapid Load Reduction, to prevent flashing in the Condensate System, while continuing with this procedure.

3.1 0 When Reactor Power is stable at ~ 83% due to two Circulating Water Pumps out of service in any waterbox, EVALUATE plant conditions AND PERFORM one of the following while continuing with this procedure:

+

Maintain power stable to maintain CAS limits.

+

Lower power level to maintain CAS limits.

[C0346]

Time

+

Raise power (if desired) until CAS limits are approached IA W Attachment 6, Raising Power with Two Circulating Water Pumps Out of Service in any Waterbox.

Salem 1 Page 3 of32 Rev. 37

sl.OP-AB.CW-OOOl(Q) 3.11 IF unable to maintain the affected condenser level(s) >33 inches, THEN SEND an Operator to locally throttle the applicable (A or B) condenser hotwell isolation valve(s) to maintain between 33 and 65 inches:

[C0346]

llA Hotwell 11 CN79, llA CONDENSER HOTWELL ISOL 11 B Hotwell 11 CN83, llB CONDENSER HOTWELL ISOL 12A Hotwell 12CN79, 12A CONDENSER HOTWELL ISOL 12B Hotwell 12CN83, 12B CONDENSER HOTWELL ISOL 13A Hotwell 13CN79, 13A CONDENSER HOTWELL ISOL 13B Hotwell 13CN83, 13B CONDENSER HOTWELL ISOL 3.12 SEND Operators to:

CAUTION The use of LP Hood Spray is permissible >15%power as a mitigating strategy for existing conditions. However, the use of LP Hood Spray is not permitted under normal circumstances >15% power because prolonged use at >15% power can result in LP Turbine Blade damage.

[70039711]

Salem 1 OPEN the appropriate 11MC62, 12MC62, or 13MC62, TURB HOOD SPRAY BYP V, on the affected condenser(s).

+

INITIATE monitoring of Condenser Hotwell and Condensate Pump suction piping for indications of flashing.

+

DETERMINE cause of Circulating Water Pump(s) failure.

Page 4 of32

[C0346]

Rev. 37

S l.OP-AB.CW -0001 (Q)

ATTACHMENT 1 (Page 2 of7)

CONTINUOUS ACTION

SUMMARY

4.0 IF AT ANYTIME a load reduction is required to maintain condenser backpressure within the Allowable Operating Region of Attachment 4, Condenser Absolute Pressure Limits, THEN:

NOTE Turbine load reduction ramp rates of ~5%/min are desirable to prevent operation of Steam Dumps, which could degrade the low vacuum/high temperature condition.

The degradation of condenser vacuum and possible loss of Steam Dumps, may lead to the use of the MS10 atmospheric reliefs for temperature control.

+

INITIATE a Rapid Load Reduction lAW Sl.OP-AB.LOAD-OOOI(Q) concun*ently with this procedure until parameters stabilize within the operational limits of Attachment 4.

5.0 IF AT ANYTIME any of the following conditions exist or are approaching:

+

Condenser ~T >27°F

+

Condensate suction temperature ~ 135°F

+

Flashing occurs in the Condenser Hotwell or Condensate Pump suction piping as indicated by erratic Condensate Pump Amp Indication OR erratic SGFP suction pressure indication.

THEN INITIATE a load reduction lAW Sl.OP-AB.LOAD-OOOI(Q),

Rapid Load Reduction, concurrently with this procedure until parameters stabilize within operational limits.

6.0 IF AT ANYTIME Hotwell level is <32 inches AND there are indications of Condensate Pump cavitation, THEN:

6.1 STOP the affected CN Pump.

6.2 INITIATE Sl.OP-AB.CN-OOOI(Q),

Main Feedwater/Condensate System Abnonnality, concurrently with this procedure.

Time Time Time Salem 1 Page 11 of32 Rev. 37

Salem December 2014 NRC Exam SRO Question #8 Given the following conditions:

- Unit 2 is operating at 100% power.

- Operators receive the following alarms:

- OHA B-13 21 SW HDR PRESS LO

- OHA B-14 22 SW HDR PRESS LO SW header pressure indication in the control room reads 98 psig for both headers.

No other ORA's have annunciated.

Which of the following describes both the possible location of a Service Water System leak which would cause these indication, and how the CRS should respond?

Assume each of the leaks is large enough to cause the indications present in the control room.

a. 4 Service Water Bay. Split SW Bays by closing 21 SW17 and 22SW17 IA W S2.0P-AB.SW-0003, Service Water Bay Leak.
b. Nuclear header x-over line between the 21SW23 and the 22SW23. Shut EITHER SW23 using Attachment 6, Service Water Valve Malfunctions, ofS2.0P-AB.SW-0001.
c. 2B EDG Lube Oil Cooler. Isolate BOTH SW supply header isolation valves and BOTH return header isolation valves to 2B EDG lAW S2.0P-AB.SW-0001 Loss of Service Water Header Pressure.
d. 21 CCW HX end bell. Ensure 22 CC HX controller set lower than 21 CCW HX, and isolate 21 Service Water Header using Attachment 4, Service Water Header Isolation, of S2.0P-AB.SW -0001.

Correct Answer: c Proposed Change: Delete question based on no correct answer.

There is no correct answer to question #8. As shown on drawing 205342, Sheet 3, the 22SW39 is the 2B EDG Jacket Water and Lube Oil Coolers SW Supply Valve, which is nonnally closed with the EDG out of service. The 2A, 2B, and 2C EDGs are normally out of service, and no indication was given in stem that 2B EDG was in service. A leak on the EDG Lube Oil Cooler will not reduce SW Header pressure. Drawing 246689 shows that when the EDG is stopped the SW39 is demanded closed and does not go open until the EDG is started. There is an additional choice (b), which also had a leak location on a line which is normally out of service, so it cannot be assumed that the leak location was active.

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Question R0#61 Given the following conditions:

Unit 2 is operating at 100% power with an identified small fuel pin leak.

A 5 gpm tube leak occurs on 22 SG Of the following, which is the only radiation monitor that will NOT show a change from this tube leak?

a.

2R19B, 22 SG Slowdown

b.

2R46A-22 Main Steam Line

c.

2R 15-condenser Air Ejector

d.

2R41 D-Piant Vent Release Rate Answer: b Licensee Recommendation: The licensee proposed to delete this question based upon there is no correct answer. Although the 2R46 monitors were designed to provide continuous monitoring of high-level, post-accident releases of radioactive noble gases via the safety relief valves, the atmospheric dump valves, and the auxiliary feed pump turbine, they will exhibit a response to N-16 at levels corresponding to leak rates in the gallon per minute range according to DCP 80057587. Thus, these monitors can provide indication of a tube lead while the unit is at power.

NRC

Conclusion:

This question was initially acceptable for the examination based upon the (incorrect) premise that the 2R46 monitors would only provide indication during high levels of post-accident releases of radioactive noble gases. However, based upon documentation stating that these monitors can provide indication in response to N-16 during low steam generator tube leak rates, the NRC concluded that Choice b is incorrect. The NRC will delete this question based upon the determination that there is no correct answer.

Question SR0#6 Given the following conditions:

Unit 1 is operating at 1 00% power.

A breaker fault occurs on the 2-6 500 KV breaker.

The 2-6 500 KV breaker does NOT trip, but should have.

15 seconds after the breaker failure, Unit 1 has not tripped.

Which of the following identifies how the Unit 1 CRS should proceed?

a.

Direct the RO to manually trip the reactor and go to EOP-TRIP-1, Reactor Trip or Safety Injection. Concurrently with the EOP implementation, initiate S1.0P-AB.LOOP-0001, Loss of Offsite Power, and perform Attachment 2, Loss of Group Busses, Part A, Loss of 1 E and 1 H 4KV Group Buses.

b.

Direct the RO to manually trip the reactor and go to EOP-TRIP-1. Concurrently with EOP implementation, initiate S1.0P-AB.LOOP-0001, and perform

, Loss of Group Busses, Part B, Loss of 1 F and 1 G 4KV Group Buses.

c.

Enter S1.0P-AB.LOOP-0003, Partial Loss of Off-Site Power, then enter S1.0P-AB.CW-0001 Circulating Water System Malfunction, and perform a power reduction to 83% power or less.

d.

Enter S1.0P-AB.LOOP-0003, then enter S1.0P-AB.CW-0001 and open the Hood Spray Bypass valves 11-13MC62s.

Answer: d Licensee Recommendation: The licensee proposed that Choice c be accepted as an additional correct response. According to the licensee, several applicants assumed that for the given conditions that condenser delta-T would increase rapidly, therefore the Continuous Action Summary of S1.0P-AB.CW-0001 would come into effect because condenser delta-T would exceed 27 oF. This would require the crew to initiate a load reduction in accordance with S1.0P-AB.LOAD-0001, thus making Choice c to be a correct response. The applicants' assumption was substantiated by a simulator plant computer screen shot showing the rapid increase in condenser delta-T when the conditions provided in the question were run on the simulator and, within a short period of time, condenser delta-T reached 30°F.

NRC

Conclusion:

This question was initially acceptable for the examination based upon the (incorrect) premise that the Continuous Action Summary would not come into effect.

However, based upon the simulator modeling of condenser delta-T for the given conditions, the NRC concluded that Choice c is a correct response. The NRC will accept Choices c and d as correct responses to this question.

Question SR0#8 Given the following conditions:

Unit 2 is operating at 100% power.

Operators receive the following alarms:

OHA B-13 21 SW HDR PRESS LO OHA B-14 22 SW HDR PRESS LO SW header pressure indication in the control room reads 98 psig for both headers.

No other OHA's have annunciated.

Which of the following describes both the possible location of a Service Water System leak which would cause these indication, and how the CRS should respond?

Assume each of the leaks is large enough to cause the indications present in the control room.

a.

4 Service Water Bay. Split SW Bays by closing 21SW17 and 22SW171AW S2.0P-AB.SW-0003, Service Water Bay Leak.

b.

Nuclear header x-over line between the 21 SW23 and the 22SW23. Shut EITHER SW23 using Attachment 6, Service Water Valve Malfunctions, of S2.0P-AB.SW-0001.

c.

2B EDG Lube Oil Cooler. Isolate BOTH SW supply header isolation valves and BOTH return header isolation valves to 2B EDG lAW S2.0P-AB.SW-0001, Loss of Service Water Header Pressure.

d.

21 CCW HX end bell. Ensure 22 CC HX controller set lower than 21 CCW HX, and isolate 21 Service Water Header using Attachment 4.

Answer: c Licensee Recommendation: The licensee proposed to delete this question based upon there is no correct answer. For the given plant conditions (no emergency diesel generators running), the service water supply header isolation valves (SW39s) would already be closed. These valves are normally closed and are open if their respective EDG is running. Therefore, if a leak developed on the 2B EDG Lube Oil Cooler (Choice c), it would not reduce SW header pressure because the oil cooler is already isolated.

The proposal to delete this question is based up the SW flow path on Drawing 205342 Sheet 3 and the valve control logic on Drawing 246689.

NRC

Conclusion:

This question was initially acceptable for the examination based upon the (incorrect) premise that the service supply header isolation valves (SW39s) were in the open position. However, based upon the plant conditions provided in the question, and documentation supporting that these valves would have already been closed, The NRC concluded that Choice c is incorrect. The NRC will delete this question based upon the determination that there is no correct answer.

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