ML20268A079
| ML20268A079 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/06/2020 |
| From: | Bryan Bergeon Entergy Nuclear Operations |
| To: | NRC/RGN-III/DRS/OLB |
| Bergeon B | |
| Shared Package | |
| ML19213A180 | List: |
| References | |
| Download: ML20268A079 (28) | |
Text
Exam Question Comment Form Source of Comment: Facility Licensee Applicant (Docket #: ______)
List the Question, answer, and reference below:
7 ID: 853534 Points: 1.00 The plant has just been shutdown from a 100 day run at full power.
Component Cooling Water Temperature is 95°F E-60A, Shutdown Cooling (SDC) Heat Exchanger is unavailable Given these conditions, complete the following statements:
The SDC system ___(1)___ meet its design basis.
With SDC in service and a loss of air to CV-3212, CV-3213, CV-3223, and CV-3224, SDC HX Isolation valves occurs, PCS flow through the SDC HXs ___(2)___.
(1)
(2)
A.
can stops B.
can continues C.
can NOT stops D.
can NOT continues Answer:
D
Exam Question Comment Form Answer Explanation Answer: D D. can NOT, continues; is correct Explanation:
Per the reference (PL-SDC Shutdown Cooling Lesson Plan, rev 7) page 7:
Design Criteria SDC HXs will cool the PCS to 130°F 27.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown with 90°F CCW temperature and a fully irradiated core. (FSAR requires BOTH HXs.)
page 14:
SDC Heat Exchangers E-60A, E-60B Remove decay heat during cooldown and cold shutdown With both Hxs in service, they will lower and hold PCS at 130F, with CCW at 90F, 27.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown with an irradiated core. Thus part one is correct because one of the SDC HXs is NOT available, and CCW temperature is >90°F.
- 9. CV-3212, 3213, 3223, 3224 SDC Hx Isolations
- a. Keyswitch operated from either C-03 or C-33.
Normally Electrically Locked Open (both solenoids de-energized).
Switches at C-33 and Control Room are in parallel, either switch placed in CLOSE will energize solenoids and valve will close. If any of the two switches is in CLOSE valve cannot be opened from other location. Since the valves were open, and fail as is, part two is correct.
Distractors:
The distractors with 'can' in part one are plausible for the examinee who is unclear about the design bases of the system.
The distractors with 'Stops' in part two are plausible for the examinee that doesn't know how a loss of air affects the positions of these valves or doesn't know the positions of the valves when the HXs are in service.
KA: 025 Loss of RHRS AK2. Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:
(CFR 41.7 / 45.7)
AK2.01 RHR heat exchangers............................................. 2.9 2.9 Test item meets KA. At Palisades the SDC system is the RHR system. Question poses a loss of the RHR system / SDC HX (via an unavailable SDC HX, then also a loss of air to isolation valves), and requires knowledge of the interrelation (partially based on the design basis of the SDC cooling system) between PCS and SDC to determine the impact on the PCS.
Exam Question Comment Form HIGH COG - need to predict / draw conclusion based on knowledge how the line up of the valves is affected when in service, and of the fail positions of system valves, the effect on flows through and around the heat exchangers, and how the PCS flow responds.
Objectives:
Explain the purpose including design criteria of the Shutdown Cooling System in accordance with the FSAR. (G2.1.27)
(SDC_CK01.0)
For the following Shutdown Cooling System major components: E-60A/B, Shutdown Cooling Heat Exchangers Describe the operational design of each component in accordance with the FSAR. (G2.2.28)(SDC_CK02.0)
Describe the effects of a loss or malfunction of the Shutdown Cooling System has on the Primary Coolant System in accordance with the FSAR. (005 K3.01) (SDC_CK11.0b)
From memory, describe the effects a loss or malfunction of the Component Cooling Water system has on Shutdown Cooling operations in accordance with the FSAR. (005 A2.01)
Comment: The station does not agree with the candidate's assertion for the reason below:
Part 1 of this question asks if the system meets its design basis, which is described in FSAR 6.1 and PL-SDC as:
The system is designed to cool the primary system from 300°F to refueling temperature with the low-pressure injection pumps and 90°F component cooling water. The maximum pressure of the primary coolant during this cooldown is 270 psia.
The stem specifies that CCW inlet temperature is 95°F, which is above the design basis temperature of the SDC system.
This means distractor B cannot be correct. The second part of this question is not being challenged.
Recommendation:
That the grading of this question remains as approved.
Exam Question Comment Form Facility Licensee position (for applicant comment only):
Challenge to Question 7. There should be two correct answers. Both (B) and (D) are correct based on an ambiguous time reference and/or no reference to Primary Coolant System (PCS) temperature.
When reviewing the question, the first sentence states, The plant has just been shutdown from a 100-day run at full power. The initial conditions, specifically the word just, is ambiguous and allows interpretation of the frame of reference. For instance, if you have a 100-day frame of reference for a refueling outage, it is reasonable to discuss two days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) into the outage that the plant was just shut down.
Additionally, PCS temperature is not provided in question 7. At Palisades, shutdown cooling (SDC) is only placed in service during modes 4, 5 and 6. These three modes are less than 300 degrees Fahrenheit PCS Tcold.
Based on the above, it is required to make assumptions on current plant parameters i.e. Tcold and time after shutdown.
Either parameter is vital to answer the first part of the question correctly. Without knowing the time after shutdown (specifically greater or less than 27.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) or Tcold (greater or less than 130 degrees F) either answer (B) or (D) could be correct based on the assumption of the student.
Reference that supports comment: Lesson plan PL-SDC; FSAR 6.1 Ensure NRC has a copy of the reference. If not, include a copy of the reference to the NRC.
Copy of reference included with written exam package submittal.
Copy of reference previously submitted to Chief Examiner.
Exam Question Comment Form Source of Comment: Facility Licensee Applicant (Docket #: _______)
List the Question, answer, and reference below:
63 ID: 848215 Points: 1.00 The plant is in MODE 1 at 13% power, preparing for synchronization, Main Turbine is at 1800 RPM The Primary, Backup, and Coastdown Relays (386P, B, and C) have NOT been reset
- Then, Actual main turbine speed reaches 1900 RPM Reactor Power rises to 16% by Nuclear Instrumentation At the turbine controls in the Main Control Room, the operator observes the closure of...
A.
ONLY the Main Stop valves B.
ONLY the Governor and Intercept valves C.
ONLY the Governor and Reheat Stop valves D.
ALL the Main Stop, Governor, Intercept, and Reheat Stop valves Answer:
D Answer Explanation Answer: D.
D. ALL the Main Stop, Governor, Intercept and Reheat Stop valves is Correct.
Explanation:
During an actual speeding up of the main turbine the OPC circuits protect from overspeed at 103% of 1800 RPM (1854 RPM) by closing the GVs and IVs, until the overspeed condition clears. It is NOT a Turbine Trip.
On a Turbine Trip, the ASO header depressurizes and ALL the valves close by Springs (MS, GV, IV, RS).
When power is greater than 15% and the generator output breakers are both open (with 386 P, B, and C devices are NOT reset), a reactor trip and turbine trip occur. (Loss of Load) ALL the valves close.
The Overspeed Trip Setpoint is 110% of 1800 RPM (1980 RPM) per EK-0102, turbine Overspeed Trip, and the lesson plan PL-EHC r08, pg. 39.
Distractors:
A, B, C are plausible for the examinee that doesn't know which valves close under the given conditions or the part that Nuclear Power > 15% plays in enabling the Loss of Load Trip of the Turbine and reactor.
Exam Question Comment Form KA: 045 Main Turbine Generator, A3 Ability to monitor automatic operation of the MT/G system, including:
(CFR: 41/7 / 45.5) A3.07 Turbine stop/governor valve closure on turbine trip.................... 3.5 3.6 Test item meets KA. Question asks what do you expect to see / monitor indications when a turbine overspeed and trip occur.
HIGH COG - multiple mental step to determine if low load trip will apply. Comprehension of inputs (prep for synchronization)
Objective: From memory, describe the design features and interlocks that provide the following Electro-Hydraulic Control System functions:
- a. Trip/Auto Start of EHC Pumps P-19A/B
- b. Turbine Runback (045 K4.12)
- c. Turbine Trip/Overspeed Trip (045 K4.11, 045 K4.13)
- d. DEH Operator Auto Control (045 A3.05, 045 K4.07) in accordance with ARP-1, E-17 Sh. 9 and SOP-8.
Comment: The drawing depicts the contacts in the shelf state (de-energized) position, however this circuit is an energized to actuate and trip the turbine vice a safety circuit that would normally de-energize to cause the trip. SOP-8, Main Turbine and Generating System, section 7.1, K-1 Turbine Generator, step 7.1.1.z directs that relays 386P, Gen Direct Trip Lockout Relay (primary), 386B, Gen Direct Trip Lockout Relay (backup), and 386C, Generator Indirect Trip Lockout Relay, be reset. Without resetting these relays, DEH will continue to dump and will not be able to actuate the Main Stop, Governor, Intercept, and Reheat Stop valves. Since the stem indicates that the Main Turbine is operating at 1800 RPM, these relays MUST be reset, therefore the conditions of the stem are technically incorrect.
Recommendation: Throw out the question as it is technically incorrect and therefore has no correct answers.
Facility Licensee position (for applicant comment only): N/A Reference that supports comment: SOP-8; E-17 Ensure NRC has a copy of the reference. If not, include a copy of the reference to the NRC.
Copy of reference included with written exam package submittal.
Copy of reference previously submitted to Chief Examiner.
Exam Question Comment Form Source of Comment: Facility Licensee Applicant (Docket #: ______)
List the Question, answer, and reference below:
74 ID: 852017 Points: 1.00 Answer Explanation
Reference:
2.4.28 Knowledge of procedures relating to a security event (non-safeguards information).
(CFR: 41.10 / 43.5 / 45.13) l IMPORTANCE RO 3.2 SRO 4.1 Test item meets KA. Q/A requires knowledge of IMMEDIATE ACTIONS of a Security based AOP.
MEMORY LEVEL Objective: Given a Security threat, implement the required actions in accordance with AOP-44, "Response to Attack on Palisades."
Comment: The station supports the applicants assertion for the reason below:
While the K/A is applicable to Palisades, Fleet Procedure EN-TQ-113, Initial License Operator Training Program, places topics such as Extreme Damage Mitigation Guidelines and B.5.b in the Post-NRC Exam transition training portion of the training program. Candidate performance on this question was 100% failure as the topic and AOPs have yet to be presented to the class.
Question 74 withheld from public disclosure due to security related content.
Question 74 withheld from public disclosure due to security related content.
Exam Question Comment Form Recommendation: Throw out this question as the subject matter is directed for training in the post-NRC exam portion of the training program.
Facility Licensee position (for applicant comment only):
This topic was not trained on prior to the exam and the procedures were not made available to candidates until the training was conducted (security sensitive procedures not for public disclosure).
Reference that supports comment: EN-TQ-113, Initial License Operator Training Program, Attachment 9.1,Section IX; AOP-43; AOP-44 Ensure NRC has a copy of the reference. If not, include a copy of the reference to the NRC.
Copy of reference included with written exam package submittal.
Copy of reference previously submitted to Chief Examiner.
Exam Question Comment Form Source of Comment: Facility Licensee Applicant (Docket #: ______)
List the Question, answer, and reference below:
84 ID: 851530 Points: 1.00 SRO ONLY:
The Plant is in Mode 5 for a maintenance outage.
An UNPLANNED ENTRY into a Higher Risk Plant Operating States (HRPOS) is required.
Given these conditions, complete the following statements:
Any Hot Work in progress ___(1)___.
LCO 3.0.9, which addresses situations where required barriers are unable to perform their related support functions and provides instructions and conditions for meeting the supported system LCOs ___(2)___ applicable to Fire Barriers.
(1)
(2)
A.
must be stopped is B.
must be stopped is NOT C.
can continue is D.
can continue is NOT Answer:
B
Exam Question Comment Form Answer Explanation Answer: B, must be stopped, is NOT; is correct.
Explanation:
Per the reference, Admin 4.49, Section 7.1 requires an announcement that any hot work must be stopped when an unplanned entry into a HRPOS occurs. Thus, part one is correct. (allowing hot work to continue is an acceptable distractor)
Per the Tech spec Bases for LCO 3.0.9, the provisions provided by this LCO do NOT apply to Fire Barriers. Thus, part two is correct. (implying the opposite is an acceptable distractor)
Distractors:
All are plausible for the examinee that is not familiar with the requirements of Admin 4.49, and / or LCO 3.0.9, but incorrect because they don't have both parts correct.
KA: 067 Plant Fire on Site 2.2.38 Knowledge of conditions and limitations in the facility license.
(CFR: 41.7 / 41.10 / 43.1 / 45.13)
IMPORTANCE RO 3.6 SRO 4.5 Test item meets KA. Tech Specs refer to many other programs that are required to meet various regulations. The Fire Protection Implementing Procedures are one of them. One specific SRO Responsibility stated in FPIP-4 Fire Protection Systems and Fire Protection Equipment, is that the SRO has the responsibility to ensure equipment status and compensatory measures are maintained as required by Palisades Admin Procedure 4.49, 'Non Power Operation Fire Risk Management'. This is the procedure that directs the stoppage of hot work, and LCO 3.09.
MEMORY level SRO ONLY:
A. Conditions and Limitations in the Facility License [10 CFR 55.43(b)(1)] Examples of SRO exam items for this topic include the following:
- reporting requirements when the maximum licensed thermal power output is exceeded
- administration of fire protection program requirements, such as compensatory actions associated with inoperable sprinkler systems and fire doors
- required actions necessary when a facility does not meet the administrative controls listed in Technical Specifications (TS), Section 5 or 6, depending on the facility (e.g., shift staffing requirements)
- National Pollutant Discharge Elimination System requirements, if applicable
- processes for TS and final safety analysis report changes Objective: Given the need to block open a Fire Door or Barrier, implement the requirements of AP 4.02 and FPIP-04.
Comment: The station does not support the applicants assertion for the reason below:
This question requires knowledge of the applicability of LCO 3.0.9 to Fire Doors. The site team reviewed LCO 3.0.9 bases, which states:
This provision does not apply to barriers which support ventilation systems or to fire barriers. The Technical Specifications for ventilation systems provide specific Conditions for inoperable barriers. Fire barriers are addressed by other regulatory requirements and associated plant programs.
Exam Question Comment Form The question stem was determined to be correct as written and requires no further information to answer the question posed.
Recommendation: No change to the grading of this question.
Facility Licensee position (for applicant comment only): The question does not specify a barrier in question. Several barriers in the plant are both fire and flood or HELB doors. These barriers fire functions are not covered by LCO 3.0.9, but 3.0.9 does still apply to them for other functions. Question seems too non-specific to blatantly state 3.0.9 does not apply. Recommend accepting answers a and b.
Reference that supports comment: LCO 3.0.9, Admin 4.49.
Ensure NRC has a copy of the reference. If not, include a copy of the reference to the NRC.
Copy of reference included with written exam package submittal.
Copy of reference previously submitted to Chief Examiner.
Exam Question Comment Form Source of Comment: Facility Licensee Applicant (Docket #: ______)
List the Question, answer, and reference below:
89 ID: 851480 Points: 1.00 SRO ONLY:
The plant is in MODE 1, with the instrumentation issues listed below:
Instrument Issue LI-0702, 'A' SG Narrow Range Level failed low LI-0757B, 'A' SG Wide Range Level failed high FI-0727A Aux Feedwater Flow to 'B' SG failed high LI-0752D, 'B' SG Safety Channel failed low Given these conditions:
The number of INOPERABLE Post Accident Monitoring (PAM) Instrument(s) is ___(1)___.
A.
one B.
two C.
three D.
four Answer:
A
Exam Question Comment Form Answer Explanation Answer: A A. one; is correct.
Explanation:
The Tech Spec Bases for LCO 3.3.7 and FSAR Appendix 7C identify the PAMS that cover SG Level as LI-0757A and B, for the 'A' SG, and LI-0758A and B for the 'B' SG.
Distractors:
All plausible as they are each a real SG Level or flow Instrument, but they each have different purposes and are all affected by the operation of AFW.
LI-0702 is a narrow range level indicator on panel C-01 that is part of the high-level override circuit and not tied to any tech spec.
FI-0727A is a flow indication in the control room on C-11 and is part of the circuit that determines if there is flow from AFW pumps P-8A and /or P-8B for proper sequencing of AFW pump starts.
LI-0752D is a level instrument that is part of the RPS and AFAS instrumentation and tied to tech spec LCO 3.3.1, 3.3.3, and 3.3.8.
The distractors are plausible for the examinee that believes either the Narrow Range Level, or the AFW flow indicator or the Safety Channel are PAMs. (They are NOT).
KA:
061 AFW 2.4.3 Ability to identify post-accident instrumentation.
(CFR: 41.6 / 45.4)
IMPORTANCE RO 3.7 SRO 3.9 Test item meets KA. Answer requires the ability to identify from a list, the PAMs associated with a Steam Generator most likely after AFW initiation. Per the TS bases, the only PAM instrument that mentions AFW is the Condensate Storage Tank Level Indication. (A suitable test item could not be developed around just a single PAM for AFW). AFW feeds the SG and level is monitored via these indicators. They are the only PAMS for this system.
MEMORY SRO ONLY:
SRO ONLY, because the information found to answer the bases part of the question is in the TS bases.
Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic the following:
- application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)
- application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).
- knowledge of TS bases that are required to analyze TS-required actions and terminology
Exam Question Comment Form Objective: Given plant conditions and Technical Specifications 3.3.3, 3.3.4, 3.3.7, 3.3.8, and 3.7.5 determine the following for the Auxiliary Feedwater System in accordance with Technical Specification 3.3.3, 3.3.4, 3.3.7, 3.3.8, and 3.7.5 BASES for the Auxiliary Feedwater System, LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)
- a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times
- b. Time period during which any associated Surveillance Requirements must be performed>>
Comment: The station does not support the candidates assertion for the reason below:
Post-Accident Monitoring Instrumentation (PAM) is discussed in LCO 3.3.7 and its basis. From LCO 3.3.7 basis:
The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. The required instruments are identified in FSAR Appendix 7C (Ref. 1) and address the recommendations of Regulatory Guide 1.97 (Ref. 2), as required by Supplement 1 to NUREG-0737, "TMI Action Items" (Ref. 3).
It is the stations position that Regulatory Guide 1.97 is not the governing document for determination as to whether an instrument is a PAM instrument or not, but that LCO 3.3.7 is the governing document and that Regulatory Guide 1.97 recommendations have already been addressed. Based on this, only one PAM instrument is INOPERABLE as described in LCO 3.3.7.
Recommendation:
The question should be graded as approved.
Exam Question Comment Form Facility Licensee position (for applicant comment only):
Challenge to Question 89. The correct answer should be (B), two inoperable instruments based on the stem of the question.
When answering question 89, both FI-0727A, Aux Feedwater Flow to Steam Gen E-50B, and LI-0757B, A Steam Generator Wide Range Level, Post Accident Monitoring (PAM) Instruments were inoperable. The answer key states the correct answer is (A), One PAM instrument is inoperable. The question stem does not specify that only Technical Specification (TS) 3.3.7 PAM instruments should be considered to answer the question. Additionally, LI-0757B is a TS 3.3.7 indication, but the question analysis states LI-0757B is the only PAM instrument inoperable. The following points support that FI-0727A is considered a PAM instrument while answering this question:
- 1) When reviewing question 89, I recognized that LI-0757B was a TS 3.3.7 instrument and realized it was inoperable. I also recognized that FI-0727A is required post-accident to monitor auxiliary feedwater flow to the Bravo Steam Generator (B S/G). There was no reference to TS 3.3.7 in the question stem and interpreted the question to be asking about all Regulatory Guide 1.97, Instrumentation For Light-Water Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident instruments. I even wrote Reg Guide 1.97 on my exam. (Also, ORM section 3.17.6, See explanation below)
- a. Reg Guide 1.97 states auxiliary feedwater flow indication is considered a Type D, Category 2 instrument for PWRs that is required to be available to the operator to monitor operation of the safety system and inform the operator of the necessity for unplanned actions to mitigate the consequences of an accident.
(See reference 2, highlighted sections page 3 paragraph 2 and page 27)
- b. At Palisades, FI-0727A is listed as a Type D, Category 2 instrument in Appendix 7C, Palisades Plant Regulatory Guide 1.97 Rev 3 Parameter Summary Table, of the Palisades Final Safety Analysis Report (FSAR). (see reference 1, document page 15). FSAR Section 7.4.3.2.1 Auxiliary Feedwater Flow Controls and Isolation Design Basis states that Reliable AFW flow and steam generator level instrumentation are necessary in order to adequately determine and control, from the control room or alternate shutdown stations, the performance of the safety-related portion of the Auxiliary Feedwater System since the operation of this system is considered as an anticipated operational occurrence by 10 CFR 50, Appendix A, GDC 13. (See Reference 5, first paragraph highlighted on page 7.4-12 of 7.4-21)
- c. Additionally, FSAR section 7.4.3.2.3 states, The performance of the safety-related portion of the AFW system can be assessed by the AFW flow indicators, two for each steam generator located in the control room and alternate stations outside the control room and a wide-range water level indicator for each steam generator. (see Reference 5, page 7.4-15 of 7.4-21)
- 2) The question stem states INOPERABLE. The Palisades Operating Requirements Manual (ORM) Table 3.17.6 items 6.1 and 6.2 address inoperability of the auxiliary feedwater flow indicators and specifically uses the word inoperable in the condition statements and in the Specification statement for section 3.17.6 of the ORM. (see reference 3) When I read the question, I knew the ORM specifically required operability of the auxiliary feedwater indications and further convinced me that the question was asking about all Reg Guide 1.97 instruments in the stem.
- 3) Instrument labeling in the control room is controlled by EN-OP-129, Operations Equipment Labeling, and specifically drawing E-49, Nameplate Standards. Note 7 of E-49 requires that all Reg Guide 1.97 instruments will be labeled in blue. Reference 4 depicts FI-0727A with a blue label and provides the note from drawing E-49.
Reference that supports comment: T.S. 3.3.7; Reg Guide 1.97; ORM: FSAR; EN-OP-129 Ensure NRC has a copy of the reference. If not, include a copy of the reference to the NRC.
Copy of reference included with written exam package submittal.
Copy of reference previously submitted to Chief Examiner.