ML17355A651

11 Post-Examination Comments

ML17355A651
Person / Time
Site:	Wolf Creek
Issue date:	12/04/2017
From:	Stucker W Wolf Creek
To:	Olsterholtz C NRC Region 4
References
TR 17-0020
Download: ML17355A651 (14)

Text

December 4, 2017 TR 17-0020 Clyde Olsterholtz NRC Chief Examiner U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Blvd Arlington, TX 76011-4511 Wolf Creek Generating Station, Unit 1 Facility Operating License No. NPF-42 Docket No. STN 50-482

Subject:

Submittal of Graded Written Examination and Analysis

Dear Mr. Olsterholtz,

Enclosed are the written examination materials, supporting the initial license exam given 11/14/2017, at Wolf Creek Generating Station.

This submittal includes the graded Reactor Operator and Senior Reactor Operator Written Exam signed cover sheets, two clean copies of each applicants answer sheets, the list of applicant clarifying questions and answers, seating chart, high miss question analysis and requests to modify the grade for three questions. The completed post-examination security statement forms (ES-201-3) are also enclosed.

In accordance with NUREG-1021, Revision, 11, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning the examination materials, please contact (620) 364-8831, Extension 5018 for Andrew Servaes, or Extension 5058 for Bob Meyer.

Respectfully, William J. Stucker Supt. Operations Support Operations Training Cc: Document Services P. O. Box 411 / Burlington, KS 66839 / Phone:(620) 364-8831 An Equal Opportunity Employer M/F/HC/VET

Wolf Creek 2017 Written Exam Question Miss Analysis

1. ID Answer Clarification R1 R3 R4 R5 R6 U1 I1 I2 I3 I4 I5 Results Bank Changes 1 #1 003 / K2.02 C 11/11 2 #2 003 / A1.09 D B 10/11 3 #3 004 / K5.09 C 11/11 4 #4 005 / K4.07 B 11/11 5 #5 005 / A4.02 A 11/11 6 #6 006 / 2.1.20 A 11/11 7 #7 007 / A3.01 B 11/11 8 #8 008 / A2.02 C 11/11 9 #9 010 / K5.01 B 11/11 10 #10 010 / K6.03 A 11/11 11 #11 EPE 007 / EK3.01 D Yes A A 9/11 Improve 12 #12 APE 008 / 2.4.4 B 11/11 13 #13 EPE 011 / EK3.13 A 11/11 14 #14 APE 015017 / AK1.02 D B B B B B B 5/11 Improve 15 #15 APE 025 / AK2.03 B D D 9/11 16 #16 APE 026 / 2.1.27 D 11/11 17 #17 APE 027 / AK2.03 C 11/11 18 #18 EPE 038 / EA2.01 C Yes 11/11 Improve 19 #19 WE12 / 2.4.21 D Yes 11/11 Improve 20 #20 APE 054 / AK1.02 A 11/11 21 #21 APE 056 / AK3.02 A C 10/11 22 #22 APE 057 / AA1.06 A 11/11 23 #23 APE 058 / AA2.03 B A A A C A C C 4/11 24 #24 APE 062 / AA1.05 D B B 9/11 25 #25 APE 065 / AA2.06 A 11/11 26 #26 WE04 / EK2.2 D A 10/11 27 #27 APE 011/EA1.3 B 11/11 28 #28 APE 077 / AK1.02 C A A D 8/11 29 #29 APE 001 / AA2.05 C 11/11 30 #30 APE 003 / AA1.07 D 11/11 31 #31 APE 028 / AK3.02 A 11/11 32 #32 APE 033 / AK1.01 A Yes 11/11 Improve 33 #33 APE 037 /2.4.6 C A D 9/11 34 #34 APE 068 / AA2.09 A C C 9/11 35 #35 APE 067 / AK1.02 B A 10/11 36 #36 APE WE13 / EA1.1 C C D 9/11 Improve 37 #37 WE16 / EK2.2 A R3 10/11 38 #38 012 / K3.01 C 11/11 39 #39 013 / A1.06 B 11/11 40 #40 013 / A3.01 D 11/11 41 #41 022 / K4.04 C D D D 8/11 42 #42 026 / K1.01 A C 10/11 43 #43 039 / A3.02 C 11/11 Improve 44 #44 059 / K3.03 B 11/11 45 #45 059 / A4.03 C D B 9/11 46 #46 061 / K6.01 C 11/11 47 #47 062 / A4.04 B Yes D A D A A C D A D 2/11 Improve 48 #48 063 / K2.01 A 11/11 49 #49 064 / K4.02 D B 10/11 50 #50 064 / A1.03 C 11/11

Wolf Creek 2017 Written Exam Question Miss Analysis

1. ID Answer Clarification R1 R3 R4 R5 R6 U1 I1 I2 I3 I4 I5 Results Bank Changes 51 #51 073 / 2.1.28 C D A 9/11 52 #52 076 / K1.19 B 11/11 53 #53 078 / K3.01 A 11/11 54 #54 103 / K1.02 B 11/11 55 #55 103 / A2.03 A B 10/11 56 #56 001 / K5.07 B Yes C 10/11 Improve 57 #57 011 / K3.02 B A A A D 7/11 58 #58 015 / 2.2.42 D B B 9/11 59 #59 017 / K4.01 C Yes 11/11 60 #60 033 / A1.02 B A A 9/11 61 #61 045 / A3.11 B C C D C C D 5/11 62 #62 068 / K6.10 B 11/11 63 #63 071 / A4.14 A C B D C 7/11 64 #64 064 / A2.03 D 11/11 65 #65 056 / K1.03 C D D 9/11 66 #66 2.1.1 A C B 9/11 67 #67 2.1.36 A/C A A A B D B A A 3/11 CR 117618 TR 2017-0246-1 68 #68 2.1.45 B 11/11 69 #69 2.2.13 B D C C C C D 5/11 CR 117337 70 #70 2.2.38 D 11/11 71 #71 2.3.4 A Yes B 10/11 72 #72 2.3.15 D B B B 8/11 73 #73 2.4.5 D Yes C C C 8/11 Improve 74 #74 2.4.32 C A B B B A A B D A 2/11 TR 2017-0252-1 75 #75 2.4.34 D 11/11 76 #76 EPE 009 / EA2.11 (SRO) A 6/6 77 #77 APE 022 / AA2.01 (SRO) A 6/6 78 #78 EPE 029 / 2.4.41 (SRO) D 6/6 79 #79 EPE 038 / 2.2.44 (SRO) C Yes B B 4/6 Improve 80 #80 EPE 055 / EA2.02 (SRO) D 6/6 81 #81 WE05 / 2.2.37 (SRO) D A A A 3/6 Improve 82 #82 APE 005 / 2.2.25 (SRO) B 6/6 83 #83 WE15 / 2.4.4 (SRO) B 6/6 84 #84 APE 069 / AA2.01 (SRO) C Yes 0/0 CR 117619 85 #85 WE02 / EA2.1 (SRO) A 6/6 86 #86 004 / 2.1.23 (SRO) C 6/6 87 #87 012 / A2.01 (SRO) A 6/6 88 #88 026 / 2.2.40 (SRO) A Yes 6/6 89 #89 061 / A2.05 (SRO) A 6/6 90 #90 062 / A2.12 (SRO) C B 5/6 91 #91 002 / 2.4.18 (SRO) D 6/6 92 #92 014 / A2.06 (SRO) A 6/6 93 #93 035 / A2.01 (SRO) C 6/6 94 #94 2.1.35 (SRO) D A 5/6 95 #95 2.1.43 (SRO) B A 5/6 96 #96 2.2.11 (SRO) A B B 4/6 97 #97 2.2.18 (SRO) C 6/6 98 #98 2.3.13 (SRO) B D D D D 2/6 Delete 99 #99 2.4.16 (SRO) C A A A 3/6 CR 117620 100 #100 2.4.30 (SRO) C B B D A D D 0/6 TR 2017-0256-1

2017 Wolf Creek ILO High Miss Question Analysis

1. 14 - Six of eleven applicants missed. High miss rate attributed to the applicants misreading the question and selecting the response associated for the affected S/G, even though we emphasized UNAFFECTED for human performance factoring. Additionally, feedback from the applicants included there is no need to specify Initially; as this would be applicable to a S/G Level question. Lesson Plan material covers expected secondary response for the resultant RCP trip. The bank question can be improved to specify The three unaffected S/G pressures should Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 23 - Seven of eleven applicants missed. High miss rate attributed to lack of knowledge of the loads powered by bus PK03 and the expected actions to address the loss of the power supply.

Lesson plan material covers the actions necessary to respond to this bus loss. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 47 - Nine of eleven applicants missed. High miss rate attributed to question construction and/or lack of understanding of the mitigative strategy associated with LOCAL EDG startup.

The applicants indicated they didnt select the intended correct answer because the listed action alone did not reset the anti-pumping relays. There are three parts to resetting the anti-pumping relays listed in Step 4c, 1) place Master Transfer Switch to LOC/MAN 2) hold for at least 4 seconds, and then 3) return the switch to AUTO. Four of the applicants at least understood there is a local action to close the breaker once the anti-pumping relays were reset, while five selected distractors that manipulated main control board switches, which are not expected during performance of a procedure for LOCAL operation of the EDG. The lesson plan material covers the expected actions. The bank question can be improved by changing the correct answer to Reset the anti-pumping relay by operating KJ HS-9, Master Transfer Switch. and distractor D to Reset the anti-pumping relay by operating NE HIS-25, NB01 Emergency Supply Breaker Handswitch. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 61 - Six of eleven applicants missed. High miss attributed to lack of knowledge regarding turbine trips. The applicants indicated they understood the question was asking for which trip was NOT considered Vital, but they chose wrong. The lesson plan material covers the turbine trips, including which are vital. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 67 - Eight of eleven applicants missed. High miss attributed to lack of refueling operation knowledge. Training Request 2017-0246-1 documents the need for improved training on the refueling topic. Exam construction error, documented on CR00117618, resulted in changing the correct answer and re-grading the question. The bank question will be changed to correct the noted construction error. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 69 - Six of eleven applicants missed. High miss attributed to confusion with regard to Tags Plus procedure requirements. The procedure allows two valves in series to be tagged, but this is not standard practice and led four applicants to select this answer choice. Confusion also existed with regards to Breakers containing D-rings; just because they are exempt from tags plus requirements does not mean they meet tags plus requirements. The applicants received Clearance Order Preparer / Tagging Authority training week of 27 Nov 2017, CO1232101.

CR00117337 previously identified the need to perform Clearance Order Preparer / Tagging Authority training earlier in the training program. Applicants were remediated on Tags Plus procedure requirements on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 74 - Nine of eleven applicants missed. High Miss attributed to lack of knowledge regarding power supplies to annunciators and the effect of the loss of these power supplies. Training Request 2017-0252-1 documented enhancements that could be made to LO1732439 OFN PK-029 lesson plan material. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 81 - Four of six SRO applicants missed. High Miss attributed to question construction by placing the applicants in a situation where conditions that direct the crew to perform foldout page actions AND Exit EMG C-0 were BOTH met. The applicants demonstrated EMG C-0 procedure knowledge for the given scenario. They knew they were at Step 11, and would exit the procedure at step 14 after taking NO physical action. Step 12 checks Red Train ESW alignment, which is satisfied by applying Appendix E guidance. The applicants were therefore faced with choice to exit the procedure and perform the required mitigation action to protect the fuel per EMG FR-H1, or delay the protective mitigation strategy to perform a foldout page action what would also be directed in EMG FR-H1. The latter option was selected as the correct choice because it is the NEXT step and answered the question asked. The bank question has been modified for future use by eliminating the restoration of power to bus NB01. Adverse Containment conditions at 20 psig were also not realistic for just a loss of offsite power/station blackout scenario. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 98 - Four of six SRO applicants missed. High Miss attributed to lack of knowledge and question construction regarding the answer choices provided. No lesson plan improvement is required as LO1733258, AP 21-001, Conduct of Operations, covers the tested information. AP 25B-100, RADIATION WORKER GUIDELINES is the actual procedure that contains requirements for RCA fast entry. The single paragraph on page 65 of 89 of AP21-001 CONDUCT OF OPERATIONS is a summary of the guidelines. An answer choice that contains exposure control and personnel protection is too plausible an answer choice when asking what procedure governs radiological controls. Even though we changed the question from During Emergency Procedure Usage, the applicants still determined EPP procedures would be applicable for a fire that last lasts more than 15 minutes. However, even if the Emergency Plan Procedures were entered, the distractor selected does NOT provide guidelines for Fast RCA entry. This question will not be added to the exam bank. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 99 - Three of six SRO applicants missed. High Miss attributed to exam construction error as documented on CR00117620. The question attempted to test the general rules of usage per AP 15C-003, PROCEDURE USERS GUIDE FOR ABNORMAL PLANT CONDITIONS, but was too specific with regard to the regulatory and license basis requirements for the crew to respond to an active fire in a safety related space. AP 15C-003 recognizes there are times when general guidelines are not always applicable and three of the applicants determined we were testing a specific time when the general requirements did NOT apply. By announcing a fire and dispatching the fire brigade, the CRS is concurrently performing both OFN KC-016 and EMG ES-03. This question will not be added to exam bank. Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

1. 100 - All six SRO applicants missed. High Miss attributed to AP 21C-001, WOLF CREEK SUBSTATION procedure not being taught. This discrepancy was identified during validations as a gap in our training program. Section 5.3 of this procedure lists 27 responsibilities of the SM or designee, including notifying the Transmission System Operator (TSO) for a change of automatic voltage regulator status (Paragraph 5.3.26) within 30 minutes. Training Request 2017-0256-1 was created to recommend adding this procedure to Lesson Plan LO1734021, Reportability and Emergency Plan to cover the required attributes of NUREG 1122, K/A 2.4.30.

Applicants were remediated on 30 Nov 2017 as documented on LO109024, NRC Written Exam Review.

Original Question Question 67 According to GEN 00-009, REFUELING, when moving control rods in containment, the crew should verify refueling pool water level is at a MINIMUM of at least 23 feet ___1)___.

While monitoring the Main Control Board Wide Range Loop Level indicators, the channels should indicate within a MINIMUM of___2)___ of each other.

A. 1) Over the top of the reactor vessel flange

2) 10 inches B. 1) Over the top of the reactor vessel flange

2) 3 inch C. 1) Above the top of the fuel assemblies in the reactor vessel

2) 10 inches D. 1) Above the top of the fuel assemblies in the reactor vessel

2) 3 inch Answer: A Explanation: GEN 00-009, Rev 39, in multiple locations states refueling pool water level must be at least 23 feet above the top of the Reactor Vessel flange Precaution and Limitation 4.7.2 specifies NR Loop Level channels should indicate within 3 inch of each other if using NPIS and 10 inches of each other if using Main Control Board indicators.

A is correct See explanation.

B 1) is correct 2) is wrong but plausible in that 3 inches is required range if using NPIS.

C 1) is wrong because refueling pool water level must be a minimum of at least 23 feet above the top of the flange when moving control rods. Plausible because GEN 00-009, paragraph G.6 specifies refueling pool water level must be a minimum of at least 23 feet over the top of the fuel assemblies in the reactor vessel for movement of control rods in the reactor vessel. 2) is correct D 1) is wrong because refueling pool water level must be a minimum of at least 23 feet above the top of the flange when moving control rods. Plausible because GEN 00-009, paragraph G.6 specifies refueling pool water level must be a minimum of at least 23 feet over the top of the fuel assemblies in the reactor vessel for movement of control rods in the reactor vessel. (2) is wrong but plausible in that 3 inches is required range is using NPIS.

Technical

References:

GEN 00-009, REFUELING, Rev 39,

New information that supports question re-grade request:

For the question asked, movement of Control Rods requires at least 23 feet above the fuel assemblies in the reactor vessel and the MCB WR Loop Level Indicators must agree within 10 inches per GEN 00-009, REFUELING, Step G.6.1 and step 4.7.2. This set of conditions corresponds with answer choice C while answer choice A was selected as the proposed correct answer.

Recommendation:

Recommend changing the correct answer to C from A. It is expected that water level in the refueling pool be at least 23 feet above the top of the fuel assemblies during movement of control rods and the Main Control Board Wide Range loop level indicators should indicate within 10 inches of each other per GEN 00-009 requirements therefore answer choice C is correct. Wolf Creek CR00117618 written to document the exam writing error, which resulted in required regrade evaluation of the question by the NRC.

Original Question Question 84 Per Technical Specification Bases for LCO 3.6.3 for Containment Isolation Valves, the single failure criterion imposed during the plant safety analysis for a loss of containment integrity is:

A. RCP seal injection valves B. 36-inch shutdown purge valves C. 18-inch purge isolation valves D. Category 1 containment Isolation valves Answer: C Explanation:

A is wrong because...these are left open per a note in the LCOs.

B is wrong because these valves must be closed with a blind flange before entry to mode 4.

They are so large that they are assumed to be unable to be shut against DBA pressure.

C is correct because this is the single failure criterion of this TS and these valves per the TS bases. They are opened intermittently at power for various reasons so the inboard and outboard isolation valves have separate power supplies, etc.

D is wrong because this is a group of valves listed in the table that have the shortest LCO completion time due to risk and other factors but these are not the single failure criterion for TS 3.6.3.

Technical

References:

TS Bases rev 75, page 3.6.3-3.

New Information that supports removal of question from the exam:

The question was written to test the applicants knowledge of LCO 3.6.3 Bases. However, as noted in the Tech Spec Bases Background, single failure criterion applies to ALL containment Isolation valves.

The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetration flow path not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on a containment isolation signal. These isolation devices are either passive or active (automatic). Manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration flow path so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. One of these barriers may be a closed system. These barriers (typically containment isolation valves) make up the Containment Isolation System.

The applicants were therefore confused with regard to what the question was asking since single failure criterion is imposed on the design consideration for all four answer choices. One clarification request form was submitted for this question during the exam, but no clarification was provided. The specific discussion in LCO 3.6.3 background document about the single failure as it applies to the 18 inch containment mini-purge valves is nothing more than an explanation on why these valves are different from the others, and NOT that they are the ONLY set of containment isolation valves in which single failure criterion was imposed.

Licensing input was requested and obtained with the following results.

10 CFR 50 App. A General Design Criteria 55 and 56 provide the requirements for containment isolation valves:

GDC 55. Reactor coolant pressure boundary penetrating containment-And GDC 56. Primary Containment Isolation-Each line (that meets GDC 55 or 56) shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Answer Choice A (RCP Seal Injection Valves) is correct. These valves meet single failure design criteria, which was modified NRC Letter dated November 3, 2010 transmitted License Amendment 190 and the accompanying Safety Evaluation to add a note allowing the RCP seal injection valves to be considered OPERABLE with the valves open and power removed.

Answer Choice B (36-inch purge valves) is correct. These valves meet single failure design criteria, per Standard Review Plan (SRP) 6.2.4 (Rev 2, July 1981)Section II. Acceptance Criteria, 6 provides a number of acceptable alternatives for containment isolation, including:

f. Sealed closed barriers may be used in place of automatic isolation valves. Sealed closed barriers include blind flanges and sealed closed isolation valves which may be closed manual valves, closed remote-manual valves, and closed automatic valves which remain closed after a loss-of-coolant accident. Sealed closed isolation valves should be under administrative control to assure that they cannot be inadvertently opened. Administrative control includes mechanical devices to seal or lock the valve closed, or to prevent power from being supplied to the valve operator.

Answer Choice D (Category I Containment Isolation valves) is correct as listed in Table B 3.6.3-1 as they meet single design failure criteria per SRP 6.2.4 Section II. Acceptance Criteria, 6

e. Containment isolation provisions for lines in engineered safety feature or engineered safety feature-related systems normally consist of two isolation valves in series. A single isolation valve will be acceptable if it can be shown that the system reliability is greater with only one isolation valve in the line, the system is closed outside containment, and a single active failure can be accommodated with only one isolation valve in the line. The closed system outside containment should be protected from missiles, designed to seismic Category I standards, classified Safety Class 2 (Ref. 9), and should have a design temperature and pressure rating at least equal to that for the containment. The closed system outside containment should be leak tested, unless it can be shown that the system integrity is being maintained during normal plant operations. For this type of isolation valve arrangement the valve is located outside containment, and the piping between the containment and the valve should be closed in a leak tight or controlled leakage housing. If, in lieu of a housing, conservative design of the piping and valve is assumed to preclude a breach of piping integrity, the design should conform to the requirements of SRP Section 3.6.2. Design of the valve and/or the piping compartment should provide the capability to detect leakage from the valve shaft and/or bonnet seals and terminate the leakage.

Recommendation, Recommend removal of this question from the exam. Since Technical Specification Bases for LCO 3.6.3, Containment Isolation Valves specifies ALL containment isolation valves are designed with single failure design criteria considerations, then all four answer choices are correct and the question is invalid. Wolf Creek CR117619 was written to document the exam writing error, which caused the NRC to evaluate the need or remove the question from the exam.

Original Question Question 99 While performing EMG ES-03, SI TERMINATION the crew observes alarms are present on Fire Alarm Control Panel KC008. A few minutes later, an NSO calls the control room and reports there is a fire in MDAFW Pump Room B.

The CRS A. Is required to concurrently implement both EMG ES-03 and OFN KC-016, FIRE RESPONSE B. Is required to complete the actions in EMG ES-03 and then transition to OFN KC-016, FIRE

RESPONSE

C. Is allowed by procedure to implement OFN KC-016, FIRE RESPONSE as long as it does not interfere with performance of EMG ES-03 D. Is allowed by procedure to implement the actions in EMG ES-03 as long as they do not interfere with the actions of OFN KC-016, FIRE RESPONSE Answer: C Explanation: AP 15C-003, Section 6.2.3, says, While performing EMGs, plant conditions may indicate the need to correct problems not directly related to the event mitigation strategy. The operator may perform OFNs and ALRs which address these problems as long as the actions do not interfere with performance of the EMGs. Therefore, the SRO may concurrently perform the Fire OFN as long as it does not interfere with performance of the EMGs.

A is wrong because AP 15C-003 says the SRO may perform OFN and EMG actions concurrently; not that the SRO is required to do so. Plausible because if the fire is not extinguished within 15 minutes, the crew will need to declare an EAL. Therefore, extinguishing the fire as soon as possible will be a priority for the crew; however, it is not required by AP 15C-003 to concurrently perform the OFN with the EMG, even where there is a fire.

B is wrong because AP 15C-003 allows the SRO to perform the OFN concurrently as long as it does not interfere with performance of the EMG. There is no requirement to wait to perform the OFN after performing the EMG actions. Plausible because performance of actions in EMGs are a higher priority than performance of actions directed by an OFN.

C is correct. See explanation.

D is wrong because the EMGs are a higher priority than OFNs. Plausible since concurrent implementation is allowed, and a candidate may think that addressing fire response is a higher priority because a fire has the potential to injure plant personnel, to spread and cause additional damage, which could complicate recovery with the EMGs, and to require entry into the Emergency Plan.

Technical

References:

AP 15C-003, Rev 34, Page 19/60 Lesson Plan LO1733203, Rev 013, Page 16/33

New Information that supports question regrade:

Input was requested and obtained from Licensing, Fire Protection Engineering, and Operations Management with the following results:

10 CFR 50 Appendix R states that (III. K.) administrative controls shall be established to minimize fire hazards in areas containing structures, systems, and components important to safety. These controls shall establish procedures to:

10. Control actions to be taken by the control room operator to determine the need for brigade assistance upon report of a fire or receipt of alarm on control room annunciator panel, for example, announcing location of fire over PA system, sounding fire alarms, and notifying the shift supervisor and the fire brigade leader of the type, size, and location of the fire.

WCNOCs Response to App R III. K. Administrative Controls states (USAR App 9.5E):

Administrative procedures define limitations to minimize fire Hazards in areas containing SSCs important to safety. Administrative procedures are also provided to promote prompt, Appropriate action upon discovery of a fire.

Section 2.C.(5)(a) of the Facility Operating License requires Wolf Creek to maintain in effect all provisions of the approved Fire Protection Program.

The Wolf Creek Fire Protection Program is based on the following defense in depth approach, which is a part of our license basis. (Reference USAR Section, 9.5.1.7.5; AP 10-100, Section 4.3; and Regulatory Guide 1.189, Section B):

• Prevent fires from starting.

• Detect rapidly, control, and extinguish promptly those fires that do occur.

• Protect SSCs important to safety, so that a fire that is not promptly extinguished by the fire suppression activities will not prevent the safe shutdown of the plant.

For the given test question scenario, with the crew performing EMG ES-03, SI TERMINATION, a design basis accident has NOT occurred, which is an assumptions associated with the post-fire safe shutdown (PFSSD) analysis per XX-E-013, POST-FIRE SAFE SHUTDOWN ANALYSIS, Rev 4, Section 3-A-3, and NRC Generic Letter 86-10, Response to Question 7.2.

The analysis documented in E-1F9910 demonstrates that PFSSD can be achieved and maintained following a single fire located in any plant area. With the scenario presented in the test question, the plant is not in a normal configuration, the cause of the safety injection is not indicated, and the fire is in an area containing equipment that is credited for PFSSD. This presents a condition that is not directly analyzed regarding potential impact on PFSSD capability. Therefore, it is important for the Control Room to know and be prepared to complete the OFN KC-016 Attachment E actions that may be required to mitigate fire induced spurious equipment maloperation. Additionally, the fire brigade response, driven by OFN KC-016, is necessary to satisfy the second defense in depth element of the Fire Protection Program so that the final defense element is not challenged.

NRC Inspection Report 05000482/2009007 documents a non-cited violation (NCV) for failure to implement compensatory actions per Station Fire Protection Program requirements in a timely manner.

On 8/19/2009 the plant experienced a momentary loss of offsite power (LOOP). Following the plant trip, multiple trouble alarms were annunciated in the Control Room on main fire alarm panel KC008. The trouble alarms were silenced by the Control Room Supervisor and a building watch was dispatched to walk down the affected areas. However, the fire watch compensatory measures required by AP 10-103, Fire Protection Impairment Control, were not established within one hour as required by the procedure.

Page 26 of the Inspection Report contains the following statement:

The control room supervisor was preoccupied with actions related to the reactor trip and did not perform the required action to initiate a fire protection impairment.

This statement reflects a regulatory position that compliance with the Fire Protection Program is required even when addressing a plant transient. Therefore, it is reasonable to conclude that Control Room and Fire Brigade personnel would need to concurrently address an EMG response and a fire response in a safety related area. Ignoring or delaying fire response would be a violation of the Fire Protection Program, which is required to be maintained by our Operating License.

AP 15C-003, Step 6.2.2 specifies: Under certain plant conditions, procedures of lower priority make take precedence over procedures of higher priority.

ALR KC-888, FIRE PROTECTION PANEL KC-008 ALARM RESPONSE, Step A.1.3.1 States:

Implement OFN KC-016, Fire Response, for fire alarms.

For the given scenario, Operations Management expects required concurrent performance of OFN KC-016 and EMG ES-03 with the following considerations:

• AP 15C-003 allows parallel performance and specifies there are specific conditions where general requirements do not apply universally.

• AP 10-100 REQUIRES response to fires, irrespective of other plant conditions.

• The KC-008 Alarm Panel response REQUIRES entry into OFN KC-016.

• AP 21-004, OPERATOR RESPONSE TIME PROGRAM specifies time critical actions associated with the plant Fire Protection Program Licensing Basis. The CRS cannot choose to ignore an active fire only because performance of an EMG procedure is in progress.

• Scenario placement in EMG ES-03 demonstrates that SI is NOT required and thus the accident in progress is something less than design basis.

• Scenario postulated fire area contains equipment vital to safety, which may well be required to respond to the accident in progress.

• Station Operating Experience includes a NCV for failure to write an impairment during emergency response. Writing an impairment is objectively less important than responding to an active fire in the MDAFW area.

Recommendation Recommend changing the correct answer to A from C. The CRS is REQUIRED to enter and perform BOTH OFN KC-016 AND EMG ES-03 actions concurrently to comply with Operating License,10CFR50, Appendix R, USAR, TECH SPECS and AP10-100 requirements for the given scenario, therefore answer choice A is correct. Wolf Creek CR117620 is written to document the exam writing error, which resulted in required regrade evaluation of the question by the NRC.