Text

Forwards Compilation of Comments on Written Exam Administered to Facility License Candidates on 881107. Comments Were Result of Review of Exams Conducted by Facility Training Staff
	ML20246P405
Person / Time
Site:	Millstone
Issue date:	11/16/1988
From:	Scace S; NORTHEAST NUCLEAR ENERGY CO.
To:	Gallo R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20246P386	List: IR 05000245/1988016; ML20246P380; ML20246P405;
References
	MP-12456, NUDOCS 8903280215
	Download: ML20246P405 (75)
	v • d • e

{{#Wiki_filter:r~ fdccchmed 3 ., u NORTHEAST UTILITIES o.n.,.i ome.,. s io.n si,..i. e.,i,n. conn.ei,cui l .in . woes uc ac w.- P O. BOX 270 m.w. 9..o.vac "'* H ARTFORD. CONNECTICUT 06141-0270 k L J $)$ [,N,cZ, (203) 665-5000 J November 16, 1988 MP-12456 Mr. Robert M. Gallo Branch Chief U. S.. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 i

Dear Mr. Gallo:

Attached is the compilation of comments on the written exam-inations administered to Millstone Unit 1 license candidates on 1 November 7, 1988. These comments were the result of a review of the examinations conducted by members of the Millstone Unit 1 training staff.. Included are both the comments discussed during the exam review meeting of November 11, 1988, plus additional comments resulting from reviews conducted subsequent to this meeting. Attendees at the November 11, 1988 meeting were: Review Participants Observer T. Walker - NRC R. Lueneburg, Supervisor D. Florek - NRC MPl Operator Training M. Spenser, EG&G, Idaho J. Hanek - EG&G, Idaho R. Palmieri - NNECO J. Nowell - NNECO M. Jensen - NUSCO G. Sturgeon - NUSCO M. Schulz - NUSCO C. Tabone - NUSCO J. Rogers - NUSCO D. Meekhoff NUSCO The exam reviews were conducted considering the following: 1. Does the question elicit the correct response? 2. Is the answer key correct? 3. Is there potential for additonal correct response? 4 Is the question appropriate? 8903280215 890321 PDR ADOCK 05000245 V PNU }

T Mr. Robert M. Gallo November 16, 1988 Page 2 References are provided, where necessary, to substantiate the comments. Please contact Mr. Raymond L. Lueneburg, Supervisor, Operator Training, Millstone Unit I with any questions concerning our comments. Sincerely, caul.-- Ste en E. Scace Station Superintendent Millstone SES/RLL/DJM/dk Enclosure c: B. W. Ruth, Manager, Operator Training USNRC Document Control Desk w/o Attach.

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a Facility Comments on the Reactor Operator-q and Senior Reactor Operator Examinat".ons J Administered on November 7, 1988 l 1 i Examiner's Handbook for Developing Operator Licensing Examina-q tions, NUREG/BR 0122, states, "It is important to note that the purpose of licensing is to make reliable and valid distinctions at the minimum level of competency that the agency ] has selected in the-interest of public protection." j For a test to be considered valid, it must be shown to measure what it is intended to measure. Question quality directly affects exam validity. Important topics tested by ambiguous, awkward of unspecific questions ' cannot be a valid measure of a candidate's knowledge. Reliable examinations produce test scores which are an -: ac cu ra te indication of a candidate's ability to protect the health and safety of the public. As exam reliability improves, the informa-tion provided by exam scores becomes more useful. Any test which yields a level of performance lower than actual. competence is i unreliable. When inappropriate knowledges and abilities: are i sampled, or the mechanics of the exam process impact personnel. taking the examination, exam reliability is adversely affected. The attached individual question comments justify Millstone Station's concern over these examinations. The candidates' scores may not be an accurate reflection of their competency or their ability to safely operate the plant to protect the health ~ and safety of the public. l ~

v l The following is a summary of the individual question comments: i ) o Some questions referenced equipment - design features not present on Millstone Unit 1. i o Many questions were ambiguous, unclear, or not specific. ) The candidates were required to spend a significant amount of time attempting to determine the intent of the question. Consequently, many candidates were pressed for time toward the end of the examination. Further, many questions did not elicit the response provided on the answer key. j l l l o Some questions contained statements which were. inaccurate or contradicted Millstone Unit 1 Technical Specifications. I. o The answer key for some questions contained technical inaccuracies. ) o The conditions stated in some questions were altered during the course of the exam, requiring ~ the students who had completed those questions to spend additional time to reassess those questions., and change their answer if

required, one question was discarded and replaced with a new question written on the spot.

o The proctors present for the majority of the exam were not the authors of the exam and they. were not fully knowledgeable of the examination questions. As a result, several candidates were given direction which was misleading. i l L_-_-________.

r v l o Some questions were out of the scope of knowledge Millstone 1 requires of Reactor Operators or Senior Reactor Operators ) and were not supported by K/A references or MP1 lesson objectives'. o Some RO examination questions (Section 4), and SRO exam-ination questions (Section 7),. required recall of the. ] material supplied in the subsequent' action steps of Millstone 1 procedures without access to these procedures. o 9 of the 26 points (34.6%) of Section 8, Administrative. Procedures, Conditions and Limitations, required recall of j Administrative Control Procedures (ACP's) or Technical Specifications without access'to the procedures or app {i-cable technical specifications. Neither the K/A Catalog or MP1 lesson objectives support. requiring recall without access to the ACPs or Technical Specifications. Comments breakdown is as follows: Reactor Operator Examination l Category 1: Comments on 3 of 1; questions - 23% Category 2: Comments on 7 of 13 questions - 54% 3 Category 3: Comments on 9 of 12 questions - 75% j Category 4: Comments on 8 of 11 questions - 73% l Overall: Comments on 27 of 49 questions - 55% Category 5: Comments on 3 of 11 questions - 27% Category 6: Comments on 9 of 11 questions - 82% Category 7: Comments on 7 of 10 questions - 70% Category 8: Comments on 7 of 13 questions - 54% Overall: Comments on 26'of 45 questions - 57% ~3-

e o The majority of these problems would have been identified in a pre-exam review by the. f acility. Most comments ;could have been resolved before the exam was administered to the candidates. Licensing examinations administered at Millstone Unit 2 during July 1988 utilized a pre-exam review. During that review, twenty of the fifty questions were rewritten. Those changes resulted in an improvement in exam quality. REF: NRC Exam Report No. 50-336188-18 (OL). Millstone Station recommends that Region I consider making pre-exam review a standard policy, i f _4_

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q J QUESTION: RO 1.05/SRO 5.05 ~ j COMMENT:. This question initially stated, "The narrow and wide range level instruments are calibrated for the reactor at hot. pressurized conditions with a specified drywell ] temperature." According to MP1 Systems Text 1300B, I volume 1, page 15-16, paragraph 5.4.3, Millstone Unit 1 wide range yarway level instruments are calibrated for cold conditions with no forced recirculation in-the 4 vessel. When-an examinee alerted the proctor to the j misinformation, the words "In a generic BWR" were added I at the beginning of the question. Consequently,. students I were required to evaluate and respond to conditions which were incorrect for the facility on which they were trained. This resulted in unnecessary confusion for the students and required additional time to evaluate non-plant specific conditions. RESOLUTION: None, general comment. QUESTION: RO 1.06/SRO 5.06 COMMENT: This question states "During operation at 100% power, a feedwater train automatically isolates'due to high water level in a heater." This question is confusing and not specific to the facility since, according to MP1 System Text ~1348,-Heater Drains,. Volume 6, page 4, para. 3.2.2, and Text 1346, Extraction Steam, page 12, Millstone 1 has no " Automatic Feedwater Train Isolations". The examinees are therefore unsure whether this question intends to test their knowledge of plant response to a loss of feedwater flow, a loss of feedwater heating, or to test their knowledge that Millstone 1 has no auto-matic feedwater train isolations. When an examinee informed the proctor, all students were informed to answer the question for a " loss of feedwater heating". l This information is inaccurate for Millstone 1 since a l single heater high level will not result in a signifi-cant loss of feedwater heating. Consequently, the candidates encountered unnecessary confusion which l required additional time to evaluate non-plant specific conditions. RESOLUTION: None, general comment. i- _ _ _ _ - _ _ - _ _ _ - _ _ _ _ - _ _ - _. _

m QUESTION: RO 1.08/SRO 5.08 COMMENT: This question requires the calculation of doubling time, given a reactor period. The questions does not specify use of a particular formula for doubling time calcula-tion. However, the answer key awards points for the use of'the P = Poe t/T equation. According to Millstone 1 operating procedure OP 201, Approach to criticality, Rev. 18, page 3, step 4.3, Period = Doubling Time x 1.44. Consequently, a candidate could solve for doubling time by dividing period by 1.44: RESOLUTION: Modify the answer key to avoid points for use of either the P = POe t/T equation or the period = DT x 1.44 equation. QUESTION: RO 2.02a COMMENT: This question states "How will the control rod hydraulic system respond to each of the following conditions or transients?" Substituting condition a.2, " Accumulator I charging flow after a scram" for the words "each of the following conditions or transients", since a.2 is one of the conditions / transients, the questions reads, "How will the control rod hydraulic system respond to accumu-lator charging flow after a scram?" However, for credit, the candidate must explain what happens to charging flow following a scram, NOT how the hydraulic system responds l to charging flow foTTowing a scram. Further, the question does not ask for values or setpoints. Additionally, a.3 implies that each CRD pump has more than 1 suction valve. This information is not specific j to the facility and caused unnecessary confusion for the candidates. RESOLUTION: The questions should be reworded to elicit the response on the key. Also, since no values were required, the answer key for a.2 should be modified to state " flow goes to maximum" for a full credit. Additionally, question a.3 shuld be reworded to state " closing the suction valve of an operating CRD pump" j since each CRD pump has only 1 suction valve. I I

4 QUESTION: RO 2.02b COMMENT: The MP1 lesson objectives and the K/As listed in the answer key reference do not support asking this question. Several K/A references do not even deal, with the CRD system. MP1 systems text 1302, control Rod Drive, Objective 27 requires the operator to state the definition of trip, when operating equipment, Objective 28 requires the candidate to state from memory how CRD flow control valves fail on loss of instrument air, and Objective 31 requires knowledge of how and why a reactor scram affects CRD system flow and charging header pressure. K/A 20100lK109 deals with l the impact of a loss of instrument air in the system. K/A 201001A201 deals with the ability to predict impact I on CRD system of pump trips. K/A 201001K602 deals with the impact of a loss of the CST on the CRD hydraulic system. K/As 293009K106, K108, K109, K110 and K118 deal with core thermal limits. None of the listed objectives or KAs require knowledge of which closed cooling water system cools the.CRD pumps, or which I components on the pump are cooled. RESOLUTION: None, general comment. J QUESTION: RO 2.06a/SRO 6.06a COMMENT: This question required the examinee to state how the discharge piping of the core spray pumps is protected from over pressurization. According to MP1 Systems Text 1336, volume 5, overpressure protection is provided in either of 2 ways, by the system relief valve, as stated in the answer key, or also by the interlock between CS-4A(B) and CS-5A(B) as explained on page 12, paragraph 5.5.1 of the referenced text. RESOLUTION: Modify the answer key to accept either "the relief valve" or "the interlock between core spray admission valves" for full credit.

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QUESTION: RO 2.06b/SRO 6.06b j l COMMENT: The question places the examinee'in an LNP condition with a core spray initiation signal present, then asks which core spray pump will start immediately after 1 the 4160 VAC bus is energized. The question implies MP1 has only one 4160 VAC-bus. Because this wording implies conditions which are not accurate for the l facility, several candidates questioned.the proctor and 1 were directed to list which pump starts as soon as j power is restored. Two interpretations are possible "Which pump starts as soon as power is restored to its 1 I bus," or "which pump starts as soon as power is restored to the plant". Since the emergency diesel generator starts in approximately.5 seconds, 14F will l be energized approximately 30 seconds before the gas j turbine restores power to 14E. Consequently, CS pump A l will start before CS pump B. However,-if the_ candidate interpreted the question as, "Which pump will start as l soon as power is restored to its bus?", then CS pump g s l would be the correct response, since it starts with no t!0s delay af ter 14E is energized. This resulted in unnecessary confusion among the examinees since it'was unclear exactly what the question required. RESOLUTION: Modify the answer key to accept either Core Spray Pump. ) A or Core Spray Pump B, based on wording of the i question and information supplied by the proctor. QUESTION: RO 2.06C/SRO 6.06C COMMENT: This question asks the candidate to give two ways to improve NPSH of the core spray pumps in accident conditions, then goes onto state " Caution 8 of the emergency operating procedures". This statement caused some students to initially think they were required to list or explain caution 8. 1 Caution 8 merely directs the operator to observe the NPSH limits and use the LOCA on Non-LOCA condition curves. The statement provides no additional useful. information to the candidate. Instead, it was distracting, resulting in unnecessary confusion for the examinees. _ _ _ _ _ _

m x s The answer key lists " decreasing core spray.flowLandr 1 decreasing torus water temperature as'means toLenhance. core spray pump NPSH. Referencing EOP 580, containment Control, Section 3, Operator' Actions,.page 3b,,the axis- .for the core spray NPSH curves areLsuppression pool water temperature and. core spray pump fl'ow rate,(GPM).- Therefore,' stating " Stay within the EOP curves, refer-enced in caution 8".is an acceptableLaubstitute for stating " Lowering torus water temperature and reducing core. spray flow"..Since:EOPs are'available during plant operation, operators should not be required to memorize'the axis of'these graphs. RESOLUTION: Modify answer key to accept " Stay within EOP curves referenced in Caution 8".as an acceptable 1 substitute ( ,for full credit. l QUESTION:- RO 2.07/SRO 6.07 COMMENT: This question placed the candidate in'a 90% power situation with all systems normal, then, given 4-separate events, asked the candidate to select which action occurred first: 3

1) RPS trip,

2) Group 1 isolation,

-3) Both, or 4).Neither This question has a logic problem. -How could both-action 1 and 2 occur first, as in action'4? 'How could neither action 1 or 2 occur first, as in-Action.37 When questioned, the proctor deleted the word first from this question. This resulted in two correct actions.for Event A (steam tunnel high temperature at 200"r).- i Action 2 (Group'l isolation) is' correct,.as well as 1 Action 4 (both 1 and 2) since MSIV closure'from the i Group 1 isolation will result'in a. reactor scram. The ) question, as modified, did not' direct ~the candidate to J consider only'first order events. a Additionally, the wording of Event C is'open-ended. The event does not statb how far'in excess of 115% steam flow is. The candidates were unsure:whether the question was tes' ting their knowledge of a the fact that-the technical specification requirement is.120% or whether the actual setpoint'is set conservatively below i 120%. Consequently, if-the student' assumes steam flow at 119% or even 120%, actions other than action 3. (neither) would be correct., 3 _1________________ j

y__ _ _. RESOLUTION: Mofify answer key to accept either action 2 Group 1 isolation, or Action 4. Both 1 and 2.as correct. Reword Event C to be more specific. QUESTION: ' RO 2.08a COMMENT: The question asks the candidate to locate the main steamline rad monitors in relation to the MSIVs. - The answer key states "..... located-in the steam tunnel (.25) just downstream (.25) of-the outboard MSIVs (.25)." Requiring the candidates to. state;" outboard MSIVs" is redundant. The only MSIVs in the steam-tunnel are.the outboard MSIVs, since the inboard MSIVs are in the drywell. RESOLUTION: Modify the answer key to delete the-word " outboard". QUESTION: RO 2.08c COMMENT: The question asks why the main.steamline radiation monitors indicate a significant. background radiation level during power operation. The answer key lists 2 contributors to steamline radiation-as-activated 4 I corrosion product carryover, and water activption. Since water activation primarily produces N and N, 3 gamma decay is the primary cpptributor to steamline radiation levels at power, N should be an' acceptable substitute for " water activation". J RESOLUTION: Modify answer key to accept'either N*' gamma radiation 1 or water activation. QUESTION: RO 2.llb COMMENT: The question states " secondary containment integrity is required if 6 conditions are satisfied. Two of.these - conditions are: 1) The reactor is in cold shutdown. I l 2)' Reactor water temperature is <212 F. List the other 4 conditions that must be satisfied." l, ._N-_---._---.-

v-c 4 Referencing MPl Technical' Specifications, Section 3.7, Containment Systems,'Part;C,oSecondary Containment,-

l statess'" Secondary 1 containment integrity',Has. defined.

i in'section 1, shall be maintained'during all-modes of. l plant. operation except when-all of the following l l conditions are' met: 1)' The reactor is inL the' cold' shutdown' condition and: j Specification 3.3.A'is' met.. 2) The fuel cask or~ irradiated fuel is not being moved in the reactor. building. Several problems arise. The statement in this question-is a directucontradiction of Technical Specifications i since secondary containment 11s always required, and can-only be broken when 2 conditions-are met, and"is not, as stated in this question,. required to be established i only when 6 conditions are: met. Consequently,;this -{ question created a significant amount of confusion'among. 1 the examinees. -It is unclear whether the. question is-l testing the candidates knowledge of when secondary containment is required,.when secondary containment ca'n. be broken, or is a trick question,' testing their ability-q to identify. incorrect statements. Assuming the question 3 intended to' test the candidates' knowledge of what 1 conditions must be met to break secondary containment, only 2 conditions are required: 1) Reactor in cold shutdown and Specification 3.3.A is met,.and i 2) Fuel cask or irradiated fuel.is not:being moved within the Reactor Building. Since " Reactor in cold shutdown" is stated in the' question; all that' remains for.the candidates to list is " Specification 3.3.A is met", and " fuel cask or irradiated fuel is not being moved in the Reactor Building". Further, since operators areL.not required to memorize Technical Specifications, an acceptable substitute for " Specification 3.3.A is met" would be " Shutdown margin is met".. Additionally, the reference quoted by the answer key, l MP1 Operator Training Procedures, EOP. Text,-page 136 l has no relevance to the material being tested in this question. 1 i ( l A

-4 DESOLUTION: Modify answer key to accept: 1) Fuel case or irradiated fuel not being moved in the Reactor Building. 2) (Specification 3.3.a) or (Shutdown Margin) is met. QUESTION: RO 2.12a COMMENT: This question provides the student with a graph of the response of vessel level, reactor pressure, core flow, and average power to a recire pump trip. The examinees are required to explain the response of each parameter. No clear discussion endpoint was defined by the question. Therefore, the examinees should not be expected to carry their discussion beyond the graph enclosed with the exam. The answer key states that reactor pressure decreases l (with power) and stabilizes at the pressure demanded by the main turbine. However, reactor pressure does not stabilize during the period of the transient illustrated,* by the graph, but continues to decrease. l RESOLUTION: Modify the answer key to delete the words "and stablizes at the pressure demanded by the main turbine". Additionally, to eliminate confusion, the question should be reworded from "Briefly describe the reason each of the four curves produced the trace it has during the pump trip transient", since curves do not produce traces during transients. QUESTION: RO 2.12b COMMENT: The question states "Briefly explain why indicated loop flow for an idle recirculation pump is inaccurate and how total core flow is determined." This question is confusing for several reasons. It is unclear whether the question is referring to recirculation pump or jet pump flow, since both are referred to as " loop flow". (MP1 Systems Text 1301A, volume 1, page 30-31, paragraphs 5.4.1 and 5.4.2.) Further, recirculation pump loop flow is not inaccurate when the recirculation pump is idle. Additionally, jet pump loop flow is inaccurate only when the other recirculation pump is operating. The status of the other recire pump is not included in the question. 1

's l The second part of the question is confusing since it is .l unclear whether the student should explain how total I core flow is determined normally or how total core flow i is determined during single loop operation.. RESOLUTION: This question should be reworded to include appropriate ] information. QUESTION: RO 2.13 COMMENT: The question asks the examinee to list the normal and ] alternate power supplies to the 125 VDC switchboard l 101A. ) The answer key lists Battery 18A, Battery Charger #3, l and 125 VDC switchboard 101B.- MP1 Systems Text 1344A, J 125 VDC Distribution System, page 2, uses the following interchangeable terms: Battery 18A or Battery 101A or Battery #1 l- ,'i Battery Charger #3 or Battery Charger C.or j Battery Charger 101C j 1 RESOLUTION: Modify answer key to accept Battery 18A or Battery 101A or Battery #1 and Battery Charger #3 or Battery Charger C or l Battery Charger 101C 1 QUESTION: RO 3.01/SRO 6.01 l 1 COMMENT: Answer key lists a SBGT System initiation as +0 inches Reactor Vessel Water Level. Actual initiation setpoint is +8 inches vessel level. Reference Millstone Unit 1 Operator Training Text, Systems Volume 4, TX 1329, page 11 of 24, paragraph 6.1 2. Additionally, a group II isolation signal is automati-cally initiated by reactor low level at +8" and/or high drywell pressure at 2 psig. Therefore, the 2 psig drywell pressure and +8 inch vessel level starts of SBGT are commonly referred to as a group II isolation start of SBGT. Referencei MP1 Operator Training Text, Systems volume 2, TX 1311A, para. 5.1.10.2, pages 51 and 52. _

RESOLUTION: Modify answer key to state +8" instead of +0" and accept either " group II isolation" or 2# start and +8" stert. QUESTION: RO 3.02/SRO 6.02 COMMENT: The question asks the candidate to explain the response of the turbine generator and the turbine control system during a runback from 90% power if the turbine control valves fail to close. The. answer key requires the examinee to state not only l the response of the generator and pressure control l system during the runback, but also the consequences of the system failing to complete the runback. Those additional responses were not elicited by the stated question. RESOLUTION: Modify answer key to delete 3.02.b.2 and 3.02.b.3. QUESTION: RO 3.03a/SRO 6.03a COMMENT: This question presents the candidate with a situation in which water level is at -40" and drywell pressure is at 3 psig. The examinee is asked to determine if APRs will automatically initiate. i The logic diagram included with the exam illustrates that initiation will not occur-unless water level drops less than -48". However, the answer key indicates the APRs will actuate. During the examination, several examinees questioned the proctor and were informed that level had fallen below -48" and " sealed in". APR logic does not seal in the -48" signal unless an actuation has occurred. RESOLUTION: Due to the inaccuracy of the answer key and the information supplied the examinees during the exam, we request either "yes" or "no" be accepted as correct, or the question be removed from the exam. QUESTION: RO 3.03b/SRO 6.03b I L COMMENT: The question initially presented the candidate with a relay failure in the APR logic which occurred after the time delay completed its cycle. i i I ' L_______-__-____

a l Again referring to the logic diagram included with the exam, analysis reveals that actuation occurs when the timer completes its cycle. Therefore, asking the. 1 student if the system will actuate when a relay failure occurs after actuation has already occurred is confusing. 1 The original question stated the "106" relay. In fact, 3 there are two "106" relays, A & B. A pen and ink i correction was made during the exam to indicate the "106A" relay. ~ RESOLUTION: None, general comment. j QUESTION: RO 3.05 4 i COMMENT: This question is ambiguous and does not elicit the response provided by the answer key. The examinee is asked for the values of the RRMG interlock associated l with the pump discharge' valve and feedwater flow. The l answer key provides setpoints of the parameters moni-tored to impose the interlock and the speed to which the j interlock limits the machine. It.is unclear exactly j what is required by " values of the interlock". 1 Additionally, since the question asks only for the j " bases" of the interlock, it is unclear to the student whether he should state only the condition which the l speed limit prevents, or include the damage which could j result should the interlock not function. j RESOLUTION: Since the question mentions valve position and feedwater flow, delete the reference to RRMG speed in the " values" position of the answer. Since the question is not specific, modify the answer key to state only the condition which the. speed limit prevents, not the damage which could result if the interlock fails. Delete the words "a condition which could result in axial thrust damage to the pump". l l._.____-__a

1 l f QUESTION: RO 3.06a COMMENT: The question requires working knowledge of Technical Specifications tables. Two concerns emerge. The reactor operators were'not provided with the appro-priate technical specifications. Additionally, although several K/As from 215005, APRM/LPRM area are quoted, the most applicable K/A, 1215005G011, an SRO level K/A, is not quoted. .I i This is an SRO level question. RO candidates-should not be required to memorize Tech Spec tables. RESOLUTION: None, general comment. I QUESTION: RO 3.08a COMMENT: The answer key indicates the gas turbine will start on a loss of the ESST transformer during power operation. MP1 Systems Text, Volume 5, TX 1339, para. 2.2, page 2 lists the gas turbine start signals as 2 psig drywell pressure or -48" vessel level, or LNP relay actuation. Loss of ESST is not a gas turbine start signal. RESOLUTION: Modify answer key to. change 3.08.a.3 to NO. j l l QUESTION: RO 3.08b COMMENT: The answer key indicates 4160 VAC buses 14A, C, E, B and D will be automatically supplied by an emergency generator following a LNP. MP1 Systems. Text, Volume 5, TX 1341, 4160 VAC Distribution System, para. 5.9, page 30-31 states that 4160V buses A, C& E will be reenergized through 14G by the gas turbine, and 14F will reenergize from the diesel generator. Buses 14B and 14D are not automatically reenergized by the diesel or the gas turbine. RESOLUTION: Modify the answer key, 3.08b to indicate 14A, 14C, 14E, 14G, and 14F. Delete 14B and 14D. t

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QUESTION: RO 3.09a i 1 COMMENT: The question asks the operator how to determine whether i RBM rod block is due to the intermediate or high level } block. The answer key response is a word-for-word description from TX 1410A, RBM, of how the RBM functions to apply one of 3 trip setpoint levels when a control rod is selected for motion. It does not describe how the operator can determine whether a block is due to an intermediate or high level trip. Rather, procedure IC 410A, Nuclear Instrumentation (RBM Channels), Step 8.1, l RBM Hi Flux or Inop, Step 8.1.5.3 describes operator I response to an RBM Hi Flux rod block. RESOLUTION: Change answer key to state: 1 1) If green "RBM trip set high" light is lit, block I is from highest setpoint. l 2) If " push to setup" light is lit, block is from intermediate setpoint. QUESTION: RO 3.09b COMMENT: The question asks the examinee what indication is available with the RBM meter function switch in each listed position. The answer key, 3.09.b.1 not only states the indication, but provides an explanation of the scale indication as well. The explanation of scale information was not elicited by the question. Additionally, the reference, K/A 218001K50, deals with automatic depressurization and has no bearing on this question. RESOLUTION: Delete the words "the volts to operable inputs ratio is 1 to 1" from the answer key.

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i 1 QUESTION: RO 3.10 COMMENT: The question asks the examinee to explain how the l combination of low ist seal leakage and high second seal leakage flow alarms could be an indication of #2 seal 4 failure-on an operating recirculation pump. 1 i The answer key, however, initially explains what a #1 j seal low flow alarm, by itself, would indicate. This i information is not elicited by the question and should not be required. i Additionally, the objectives referenced by the answer key, TX 1301A, Obj. 15h, i, j, require the operator to know the purpose of individual alarm windows, not combinations of windows. Further, K/A 202001K404 requires knowledge of recirc system features which provided for controlled seal flow, not seal failure. Seal failure is more accurately covered by K/A 202002A2.10. Notably this K/A is marked by a dagger, indicating this' depth of knowledge is better suited to an SRO exam than an RC exam. The remaining K/As, 241000K306, K304, K302, K301 all deal with turbine pressure regulation and have no bearing on this question. RESOLUTION: Alter the answer key to read: "The-combination indicates #2 seal failure large enough to activate the

2 high seal flow alarm and reduce #1 seal flow to the alarm value."

QUESTION: RO 3.11 l COMMENT: The question asks the examinee to state 3 of the 4 local indications of a tripped breaker, assuming the fault was a load side ground over 4 amps. The answer key lists " closing spring status flag chgd" as one indication of a tripped breaker. Referring to i the question reference, MP1 Operator Training Systems, volume 5, 4160 VAC, page 33, 2 lists of indications are provided. These lists present the status of all indica-tions available on the breaker in the closed or tripped conditions. Not all indications available on the breaker will be utilized by the operator to determine the status of the breaker. Closing spring status flag _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

y 4 indication is not useful for determining breaker status since it indicates charged whether'the breaker is open. or closed. RESOLUTION: Delete the words " closing spring status flag 'chgd'" from the answer key. QUESTION: RO 4.02a/SRO 7.02a COMMENT: The examinee is provided a plant condition with both reactor recirculation pumps off, 1 shutdown' cooling pump at maximum flow, with vessel level at 55" and is required to determine if stratification will occur. OP 206, Plant Cooldown to Cold Shutdown, Steps 4.9, 4.10 and 4.11 indicate that stratification is prevented if vessel level is raised above +55". Additionally, MP1, Operator Training Text 1200, Procedures Volume para. 3.6.4.5, pages 69-71 agrees. RESOLUTION: Change the answer key from Yes to No. QUESTION: RO 4.02B/SRO 7.02b COMMENT: OP 206, Plant Cooldown to Cold shutdown, Step 4.10 states " Operation of only the 'B' recirc pump in conjunction with.the shutdown cooling system may result in 'short circuiting' core flow if vessel level is less than +50 inches." MP1 Operator Training Text 1200, Procedures Volume, para. 3.6.4.5, page 20, agrees, stating " Stratification is most probable if only recirculation pump B is operated when the shutdown cooling system is in operation since this can short circuit flow around the core." This question places the candidate in a situation where shutdown cooling is in operation at full flow with only recire pump B operating. However, the examinee-cannot accurately determine if stratification will occur since no information is provided concerning vessel level. If the student assumes vessel level is less than 50", the correct answer would be yes. If the student assumes I vessel level is above 50", the correct answer would be I no. l l l I

RESOLUTION: Either this question should be removed from the exam, or es or no accepted as correct since insufficient n ormatTon was provided to accurately determine if stratification would occur. l QUESTION: RO 4.02d/SRO 7.02d j COMMENT: Since, in this question, vessel level is +58", the references and logic of question RO 4.02a/SRO 7.02a apply. RESOLUTION: Change the answer key from Yes to No. ? 1 j QUESTION: RO 4.03 COMMENT: This question asks the RO candidates why, in EOP 575, Pover/ Level Control, CRD injection is allowed to continue while level is being lowered, how lowering level decreases reactor power and asks for 2 reasons why level is restored to 10-50" when the hot shutdown l weight of boron has been inserted. The reference quoted by the answer key, MP1 Operator 1 Training Text 1500, Procedures volume, contains no objectives requiring RO knowledge of the information tested in this question. Further, of the K/As listed, three apply. K/A 295037K102 (RO 4.l*) requires know-ledge of the effects of reactor water level on reactor power. On an RO level, knowledge of the effects of water level on power translates to stating how changing water level changes reactor power, i.e., lowering level lowers power, NOT knowledge of wh_y lowering vessel level lowers reactor power. K/A 2950 WK209 (4.0) requires knowledge of the interrelations between an ATWS condi-tion and vessel water level. Again, this translates to stating how reactor power is affected by water level, not jh reactor power is affected by water level. K/A w 2950 WK303 (4.l*) requires knowledge of the reason for lowering water level in an ATWS condition. This K/A requires the candidate to know that water level is lowered to lower power to minimize heat addition to the torus., NOT why lowering level lowers reactor power. Neither MP1 objectives nor the K/A catalog supports requiring Reactor Operators to have knowledge of the concepts tested in this question. _ - _ _ _ _ _ _ _ - _.

1 1 RESOLUTION: Delete this question from the Reactor Operator' exam. ) k QUESTION: RO 4.04 l COMMENT: This question asks the candidate to list the two l immediate actions required by ONP 503B (Loss of All j Station AC Power). In fact, there are three procedure i steps in ONP 503B, and.a total of-five. discrete actions. j .Therefore, asking the candidate to list the two J immediate actions is confusing and does not reflect the i actual procedure. Reference ONP 503B, Rev. 1, page. 1. ) 1 l RESOLUTION. Modify the answer key to require only two of'the following: 1) Verify or initiate a reactor scram. ] 2) Close all MSIVs. 3) Place isolation condenser in service. 4) After start of DG/GT, start RBCCW pump. 5) After start of DG/GT, start CRD pump. QUESTION: RO 4.06 COMMENT: This question requires listing 5 of the plant conditions f requiring a manual scram. 'The list provided by the j answer key is incomplete. ] RESOLUTION: Add the following to the list of -atanual scrams: 9. Torus temperature reaches 110*F REF: EOP 580, Step 3.3.4 10. Uncontrolled power oscillations i REF: EOP 526, Step 1.3 l 11. Imminent loss of safety equipment or vital controls or instrumentation REF: ONP 525B, Step 2.2 ONP 525D, Step 2.2 12. Spreading fire in SDC pump area or fire degrading primary containment capability REF: ONP 525E, Step 2.3 l l l 1.

I 13. SS/SCO determination that safe operation of plant is threatened during fire in Gas Turbine Building REF: ONP 525F, STEP 2.3 QUESTION: RO 4.07 COMMENT: This question requires the candidate, given a list of l separate events, select from an index which procedures would be entered. l Event A states " Scram pilot air header pressure - 50 psig" with no explanation why header pressure is at 50 psig. Is there a scram air header pressure regulator malfunction? Perhaps a small leak on the header? Have both air compressors tripped? Is there a large leak in station or instrument air? Further, has the reactor scrammed or does an ATWS condition exist? This event contains insufficient information to make an accurate determination of which procedures to enter. The question is uncisar whether the candidate should limit himself to first effect procedures or consider also second order effects. For example, if the reactor has scrammed due to the low air header pressure, when level drops less than '8", EOPs 570, 571, and 572 are also entered. The answer key to event A requires the student make the assumptions that the reactor has scrammed (ONP 502) and the decrease in scram air header pressure is due to a rapid and total loss of instrument air (ONP 512), and limits itself to first order events. In Event C, no explanation is provided for why reactor power has increased to 1094 psig, contributing to the unrealistic nature of the question. In Event D, both recirc pumps have tripped and power is steady at 45% one minute later. The inclusion of a time delay contributes to the overall confusion present in this question. The candidate is unsure why power operation was allowed to continue for 1 minute with no forced circulation. Or, has the operator attempted a manual scram according to ONP 504, Recire System Failures, and an ATWS condition exists? The answer key lists ONP 504, Recirc System Failures as the applicable first order procedure, but also lists ONP 502, Emergency Plant Shutdown. ONP 502, in this instance, is a second order procedure since ONP 504 directs the manual scram. When this event occurs. ONP 504 is antered initially, not ONP 502. _ _ - _ - _ _ _

m In Event r, condensate pump discharge conductivity high high,.no. explanation is provided for why condensate pump discharge conductivity has increased to 15 umhos, contributing to the overall unrealistic nature of this question. That level of condensate conductivity is difficult to attain without gross failure of main condenser tubes. Additionally, the answer key for Part { F lists ONP 515A, the applicable first order. procedure, but also lists ONP 502, Emere ency Plant Shutdown. ONP 502, in this instance is a sacond-order procedure, since ONP 515A directs the scram. When this event occurs, ONP 515A is entered first, not ONP 502. 1 i i RESOLUTION: Modify the answer key as follows to require only first i order procedures, but allow the candidate to list second l order procedures without losing credit since the question did not specify how far the student should 3 carry his answer. Additionally, delete ONP 512 from ') Event A, since insufficient information about the cause of the low scram air header pressure was provided. 3 a) ONP 502 (ONP 512, EOP 570, 571 572) ) b) ONP 516, ONP 509 l l c) ONP 502, EOP 570, 571, 572 d) ONP 504 (ONP 502, EOP 570,~571, 572) e) None f) ONP 515A (ONP 502, EOP 570, 571, 572) l QUESTION: RO 4.09 COMMENT: This question lists the word-for-word definition of each' l of the four NRC event classifications. This requires i the candidate to supply, from memory, the corresponding j name of each event classification. Neither EPIP 4701 or the EAL tables were supplied. J i l l This question raises several concerns related to both the intent.of the question and the significance of the l material being tested. To adequately answer this question, a student is required to commit to memory the word-for-word definition and corresponding name of each NRC event classification. Yet, those classifications-are provided on each utility's EAL tables. Further, the word definitions are not used by the operators in classifying events. Rather plant-specific symptoms and, conditions are listed in each of the key categories, rendering verbatim knowledge of event definitions useless information. l. - - _ _ -

O t This question encourages rote memorization of information which is not required to safely operate a facility or to protect.the' health and safety of the public. MP1 Operator Training Materials contain no objectives requiring. memorization of NRC event classification definitions. Further, the K/As listed in the reference of the answer key do not support this question. K/A 2950315G02 (3.l* - 4.6*) requires knowledge of which events related to syrtem operation / status should be. reported to outside agency. The knowledge that an event should.be ceported to an outside agency does not equate to verbatim knowledge of the word-for-word definition of the NRC event classifications. Notably, K/A 2950315G02 (3.l* - 4.6*) is marked with a dagger and the RO importance is only 3.1. As. supported by this K/A, event classification at Millstone 1 is the responsibility of the SS/SCO. .Therefore,-requiring reactor operators to have verbatim knowledge of NRC event classifications is beyond the scope of ther responsibilities. 1 Additionally, the question is unclear since it does not state whether the student should supply the NRC term or the State of Connecticut Posture Code term, since both are listed on the EAL table. RESOLUTION: This question is beyond the scope of knowledge required of Reactor Operators. Further, no MP1 objective or NRC j K/A supports testing this information. This question i should be deleted from the examination. J QUESTION: RO 4.10b COMMENT: This question informs the. candidate that he has just completed a tour of the reactor building which was authorized by a blanket RWP. The examinee is required 4 to state where he will log his exposure. The answer key states, "On the weekly incidental exposure RWP form." Also acceptable would be " Incidental exposure sheet". The answer key reference, SHP 4902, page 4, Step 4.4 states, "If (the activity was) not (covered by) an RWP, the readings should be recorded on the Weekly Incidental Exposure RWP form at the end of each day". SHP 4912, Rev. 10, page 3, Step 3.2 states that " Incidental exposure sheets are used to record exensures which are i not recorded by an RWP. 1

x J Further, SHP 4912, Radiation Work Permit Completion and Flow Control, Rev. 10, page.9,LStep 8.1.11 states "The second part of the_RWP may be either a ' daily log' or ' weekly blanket' for access and exposure control (part 2 or 2a, respectively of HP Form 4912-1 enclosed in the-back of SHP 4912). These portions of the RWP-are used. to record personnel exposures." Exposure received during a tour of the reactor building authorized by a blanket RWP will be recorded on the blanket RWP. Since this examination was administered'to Reactor Operator candidates from the Operations Department, and the question did not specify which blanket RWF, the. 'andidates could assume the Operations Blanket RWP as the correct place to log this exposure. RESOLUTION: Change answer key to accept either " blanket RWP" or " Operations (OPS) Blanket RWP". l l L l QUESTION: RO 4.11 COMMENT: This question asks the Reactor. Operator' candidate why venting primary containment is prevented if drywel F temperature exceeds 212* F in an accident condition. The same concerns raised in RO question 4.03 apply'here. The reference quoted by the. answer key, MP1 Operator-Training Text 1500, Procedures volume, contains no objectives requiring knowledge of the information tested in this question. K/A 290528K201 requires knowledge of the interrelations between hi drywell temperature and drywell spray, not the interrelations between drywell temperature and containment venting. Further, K/A 270528K201 is marked with a dagger, indicating a lower knowledya level is required of RO candidates than SRO candii.e-:. Although the-question asks why venting is stopped above 212*F, the answer key explains why drywell spray is not initiated in a steam environment. According to MP1 Operator Training Text 1500B page 136, containment g venting is stopped above 212 F, since above 212 F it i s likely that steam is being admitted to the drywell and venting under those conditions would result in the gradual removal of all non-condensibles. The question does not ask the candidate to explain why spraying a high steam environment could be damecing..

l \\ This material is also tested in SRO 6.11, however the i wording of the question is vastly improved. RESOLUTION: Neither MP1 objectives nor the K/A catalog' support testing Reactor Operator knowledge of this material. Additionally, the on the answer key. question does not elicit the response This question should be deleted from ~ the RO examination. 4 QUESTION: SRO 6.08 l COMMENT: This question asks the candidate to explain the effect of load variations on diesel generator output voltage with the voltage regulator in manual. The question does not state whether the diesel generator is operating as the only generator on the bus, or whether,it is operating in parallel with the reserve station services transformer, or normal station services transformer through 4160 VAC bus 140. The response of the diesel generator output voltage to load changes will differ in l either case. The answer key describes the diesel generator output voltage response if the diesel is the only source on the bus. One candidate requested classification from the proctor and was told to answer the question assuming the diesel generator was in parallel with 14D. Under those condi-tions, starting core spray pump 'A' will have no effect on D/G output voltage, since the additional reactive load will be assumed by the transformer, due to the load i sharing characteristic of the transformer. Since the candidate assumed no change in Part a, and the answers in Part b are determined by Part a, no change becomes a l correct response for both Part 1 and 2 of b. RESOLUTION: Based on the wording of the question, and the information supplied by the proctor, the answer key should be modified to also accept "no change" for Part a and 1 and 2 of Part b. l _ _ - _ _ _ _ - _ _

\\

- - - - A***

QUESTION: SRO 6.09 COMMENT: This questions states, " Explain the reason the Automatic Pressure Relief (APR) system is required to prevent core damage during small break loss of coolant transients." MP1 Operator Training Text TX 1337,. Volume 5, page 1, objective 1 states "The purpose of the APR system is to provide a method to rapidly reduce reactor pressure during a small break loss of coolant accident accompanied by a feedwater coolant injection system failure. MP1 Operator Training Text TX 1334, volume 4, page 1 states, "The purpose of the FWCI system is to provide a source of coolant to adequately cool the core for a small break loss of coolant accident". This question is stated incorrectly, since APR initiation is required during small break LOCAs only when the FWCI system (designed to adequately handle small breaks) fails. Therefore, the examinees were unsure whether their knowledge of the purpose of the FWCI system, the purpose of the APR system, or their ability to identify the statement as incorrect was being tested. Further, the answer key reference, MP1 Operator Training Systems, Volume 5, APR System, objective 1 stated above does not support this question. Answer key reference K/A 218000F301 (4.4-4.4) requires the candidate have a knowledge of the effect.that a loss or malfunction of the APR system will have on restoration of reactor water level after a break that does not depressurize the reactor when required. This K/A requires knowledge of the impact of the failure of the APR system'on level restoration after a break too small to depressurize the vessel and does not support testing the candidates knowledge of the purpose of the APR system. RESOLUTICN: Since most candidates assumed the question was asking for the function of the APR system in a small break LOCA, the answer key should be modified to state: "The APR system opens selected SRVs to rapidly depressurize the reactor to within the injection pressure of core spray and LPCI (Low Pressure ECCS Systems)".

Reference:

MP1 Operator Training Text, volume 5, TX 1337, APR System, page 1, para. 1.0.

s 4 QUESTION: SRO 6.10 q COMMENT: Ths question places the examinee in a situation with an idle recire pump, design basis LOCA in the idle loop,-and a trip failure on the r.unning recire pump. The candidate is then asked 4 separate questions concerning the LPCI loop selection logic. 1 In Part A, the candidates were asked what action load selection would take prior to proceeding with load selection. Since " prior to proceeding with loop selec-tion" was defined as the endpoint for this part,.some I candidates carried their anewer through reactor pressure decreasing to 900 psig, since loop selection will not proceed until pressure drops to 900 psig. The meaning of the word permissive in Part B was unclear. l Candidates were unsure whether the permissive was the permissive to CONTINUE (900 psig) or the permissive to START (2# or -48"). Since some candidates discussed 900 psig in part a, they assumed permissive meant " permissive to start" and listed "2 psig drywell pressure or -48"; vessel level". When questioned,.the proctor lead some candidates to believe Part b required " permissive to j start" signals. Further, " loop selection permissive" could be interpreted as the signal which permits selecting loop A or loop B, which is riser differential pressure. Consequently, the words " loop selection permissive" could. be-logically interpreted 3 different ways. Part d requires the candidate supply the permissive signal for the LPCI loop A injection valve to open. The answer key states " low RPV pressure (.25) 350 psig (.25). Inherent in identifying the signal as " reactor pressure at 350 psig" is the knowledge that 350 psig is low pressure. Therefore, requiring the candidate to state " low reactor pressure" for.25 pts is redundant. RESOLUTION: Modify Part b answer to accept either: 1) RPV pressure at 900 psig (identifying this pressure as " low" is also redundant) gr 2) 2 psig drywell pressure or -48 inches vessel level EE 3) recire loop riser differential pressures Modify Part d answer to delete the word " low".

- - - - .+....**+.. - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ -. _ _ _ _.

9 l l ] QUESTION: SRO 6.11 . COMMENT: The wording of 6.11 is awkward. Question should be reworded from "What condition is indicative if drywell tempe rature is above 212* F7". RESOLUTION: None, general comment. QUESTION: SRO 7.03 COMMENT: This question is identical to RO 4.03.

However, comments on RO 4.03 dealt with requiring RO candidates

{ to have SRO level knowledge of EOP bases. On the SRO exam, however,-one concern arises regarding the answer'to 7.03b as written on the key. Although 7.03b states, "How does lowering RPV water level' decrease reactor-power?", the answer key requires'thad the candidates state " reducing the natural circulation driving. head i due to height differential (.5).results.in increased core voiding and decreased reactor power". Requiring the student to state "and decreased reactor power" is redundant, since that information was supplied with the question. Candidates should not be required to restate l the question in their answer. ) RESOLUTION: Modify the answer key to delete the words "and decreased reactor power". QUESTION: SRO 7.04 { COMMENT: This question stated in part, " Answer the following questions concerning operator actions per ONP 503B, Unit 1 Loss of Off-site and On-site AC Power (Station Blackout)." i This statement confused most candidates since ONP 503B is the Loss of All Station AC Power procedure, while " Unit 1 Loss of Off-site and On-site AC Power (Station Blackout)" is ONP 503C. Since one procedure was mentioned by number, but another by name, each assuming different electrical power availabilities, the candidates were confused as to the actual availability of power or the electric plant alignment. l l ! l l L

s 4 Some students felt the question was testing their - ability to determine that the name and number of. the procedur e did not match. Questions 7.04c and 7.04d are.particularly susceptible, since they state "The procedure " instead of naming a particular procedure, as 7.04a and 7.04b do. Additionally, 7.04b states: " List the reason for the subsequent action stated below during the performance of ONP 503B (Loss of all Station AC power): i verify drywell radiation are normal by selecting " Radioactivity control from SPDS displays or check drywell radiation on CRP 918. Then open drywell nitrogen compressor suction valves and restart the compressor." This question is worded to emphasize the concern for checking radiation levels prior to restarting the nitrogen compressor. Consequently, the question does not elicit the response to explain why the drywell compressor is started, but rather, why drywell radiatkon levels are check prior to starting the compressor. Additionally, the answer key for 7.04b states, in part, " Restores nitrogen, provides operational control-for reactor pressure control equipment (.25) and the drywell cooler dampers." Text 1500A, ONPs, page 40, states "..... provides operational control of the safety relief valves, the MSIVs inside the drywell and the nitrogen J operated drywell cooler dampers". Drywell nitrogen is not supplied to the pressure control system. Question 7.04c requires the candidate to state why the i GT mode switch is placed in "OFF" following a starting l failure. Assuming the question refers to ONP 503C, 1 Station Blackout, the answer key states, " Secure the DC Auxilihry pumps (.25) and conserve the gas turbine battery (.25)". The subsequent action referred to by ) this question, ONP 503C, Step 2.3,

states,

..... place gas turbine mode switch in OFF to secure DC auxiliary j pumps and conserve gas turbine battery". This answer, j therefore, requires a candidate have recall of the subsequent action steps of the ONP. Although MP1 j Operator Training lesson objectives require knowledge of l the reasons for subsequent actions, they do not require 1 memorization of the subsequent action steps, nor do the i K/As listed in the answer key reference. 1. _. _ _ _ _ _ _ _ _ - - _ _ _ _ _

a ~ E l MP1 Operator Training text, Procedures volume, TX 1500A, page 53 states, " Placing the gas turbine mode switch in l off will conserve the gas turbine battery. The intent ) of the step is to conserve the battery, not secure the DC auxiliaries. If the DC auxiliaries could be run without depleting the battery, they would be left to run. RESOLUTION: Modify the answer key for 7.04b to state only, " Ensures i no radioactive release to the reactor building", since the ] question does not elicit the response on the answer key, nor is drywell N supplied to the pressure control system. 2 Modify the answer key for 7.04C to state " conserve gas turbine battery" since operators are not required to recall information supplied in subsequent actions without i l reference to the procedures. While grading 7.04d, consideration should be given to the fact that students were unsure of electrical power avail-ability due to the ambiguity of the opening statement. l QUESTION: SRO 7.05a COMMENT: The candidate is asked to state the negative consequent of opening an SRV prior to level decreasing to -220 while in a steam cooling. { The reference stated on the answer key, MP1 Operator Training, Text TX 1500B, Emergency Operating Procedures, l page 49-50, does not list a negative consequence of opening a relief valve prior to 'J20" vessel level. The purpose of the steam cooling procedure is to minimize core heatup and provide additional time for re-establish-ing a makeup source of water if level is less than -127 and no makeup source exists. To minimize core heatup while providing as much time as possible to establish an alternate makeup source, RPV water level is allowed to I decrease until peak clad temperature reaches 2,000 F. Since fuel temperature cannot be monitored directly, the l fuel uncovery time to reach 2,000 F was determined by l l computer model. The level to which inventory would drop i during this time was calculated to be -220". Establishing this point as an action level relies on an easily moni-tored parameter rather than difficult to measure time intervals. Further, the calculation for -220" assumed the RPV was isolated and no break existed, inventory loss was l 1 _ - __ ____.

M a i 1 I based on. boil-off.alone. This: calculation was.most { conservative'since-any break would result inJa more-rapid level decrease and therefore a1 shorter uncovery time and lower fuel temperature. Therefore, allowing level to drop to -220" providesLthe operator'with an much time.as possible while not allowing clad'temperatureato reach 2200*F. Given'that -220",1 throuah; calculation,.. represents a clad.temperatureLof?2,000 r,Hopening relief valves prior.to -220 is conservative since uncovery time will be: shorter and clad. temperature will therefers be lower.. 1 Opening valves prior to -220 shortens thertime1the l operator has to re-establish'a makeup source ~. 4 l .The answer key states that " Insufficient thermal 1 gradient would be developed resulting.inLinsufficient steam flow to remove decay heat from the core with. ~ L L steam". Although the reference states "Higherofuelu temperatures are required to produce"the thermal-gradient necessary to remove decay hest from.the core with steam", it does not identify.the temperature.at. which the thermal-gradient will be high enough to read,ve heat. The reference also states that 2,000*r clad' e temperature will generate sufficient. heat transfer:while maintaining adequate margin to thermal limits. 1 The-statement justifies' allowing clad temperature.to. reach 2,000 r; it cannot be-turned:around to11mply that insufficient thermal gradient for. steam cooling exists below 2,000 F. The reference goes on to' state, "When level drops to -220 the operator is directed to open l'SRV. The resultant steam flow would.be. sufficient to maintain clad temperature.below 2,200*r." This statement justifies opening only 1 SRV at -220, since the resultant steam flow will be sufficient to l remove decay heat at 2,000* F. It does not imply ~that j 1 SRV will produce insufficient flow for steam cooling l at clad temperatures less than.2,000*F, since steam flow is a function of reactor pressure, NOT thermal gradient. .I Sufficient steam flow exists wi.th 1 SRV until-reactor pressure decreases to 700 psig. RESOLUTION: We therefore request this question be removed from the l exam, since no negative consequences of opening an1 SRV prior to -220" is stated in either the reference or the EOP Appendix B. Additionally, the objective quoted;by the answer key, objective 24, asks the. student to state the purpose of the steam cooling! procedure, not.the-operational impact of not following'the-procedure. Further, the K/As referenced, 295031K101 and=295031G007, require the examinee to relate the operational implica-tion of low water-level to adequate core cooling and to explain system limits and precautiann, respectively, i 1 l

l I { \\ neither-of which support the question.- Therefore, we feel this question does not match the testing objective i or the K/A intent. ] I QUESTION: SRO 7.06 COMMENT: The wording of this question is misleading. EOP 570 and EOP 572 are symptom-based procedures, not transients. The guidance provided by these procedures'is applicable in a wide range of plant events, not just in two 1 instances, as the question implies. This question is ambiguous and non-specific and does not elicit the specificity of the answer key. RESOLUTION: None, general comment. l l 1 QUESTION: SRO 7.07 l COMMENT: As originally written, question 7.07 was based'on a -l procedure ONP 511, whien was cancelled effective 10/28/88. When the practor was alerted, the original l question 7.07 was cancelled and a new 7.07, based on ONP 525A, Degraded Fire in the. Control Room or Cable Vault,- was written on the spc'. Consequently, those candidates who had already complecid the original question 7.07 were required to go bach and answer another 2-part question containing 5 responses. Additionally, the answer key for 7.07.a.1 states the SCO reports to "the IC-3 area". However, ONP 525A, page 1, Personnel Allocation, directs the SCO and control Operators to the reactor building. Although the SCO functions are performed in the reactor building, several are performed well outside the "IC-3 area". This question tests a candidates ability to recall subsequent action steps of ONP 525A which, according to ONP 525A, page 2, need not be accomplished until up to 10 minutes after control room evacuation. TX 1500A, l Procedures volume lists no objectives requiring the i knowledge tested in this question. The K/As listed require the operatore have knowledge of the local control station, and 'the ability to monitor and control reactor pressure, determine or interpret cooldown rate, and locate and operate components, including local l 1 l u -j

's i I 4 controls. None of these K/As support requiring knowledge of these actions without in-hand use of procedures, especially considering the actions can occur up to 10 minutes after control room abandonment. The references'of the deleted question 7.07 were used j since none were supplied with the new question 7.07. ) One candidate was stopped shortly into this procedure on his plant walk-through because the examiner stated he 1 obviously knew the procedure very well. However, the candidate feels he may have failed this question since he could not recall from memory all the information

required, l

RESOLUTION: Modify the answer key for 7.07.a.1 to accept " reactor building" as a correct response. I QUESTION: SRO 7.09 ) i COMMENT: This question requires the candidates have a verbatim l knowledge of a flowpath used in the event of a fire in the Shutdown Cooling Pump room, after the isolation condenser is no longer effective, assuming an electrical failure of both shutdown cooling pumps. Again, this question requires specific verbatim knowledge of subsequent action steps of a procedure. MP1 Operator Training texts contain no lesson objectives requiring verbatim knowledge of subseiuent action steps. Of the K/As listed in the reference on the answer key, K/A 205000A203 (3.2-3.2) applies. It requires "the ability to predict the impacts of AC failure on the shutdown cooling system and based on those predictions, USE PROCEDURES to correct, control or mitigate the consequences of those abnormal conditions or operations." Clearly, this K/A does'not require detailed knowledge of subsequent actions from memory. Part 7.04b requires the candidate list 2 reasons why reactor water level is raised to +100 inches after.the isolation condenser is secured. Subsequent action step 2.7.2 of ONP 525E states, raise reactor water level to +100 inches on wide range GEMAC. This will provide increased head pressure for flow to torus and further assure natural circulation if recirculation t L _ ___________ __

m 4 0 pumps are not running". The answer key for 7.04b is an almost word-for-word quote of the subsequent action step, again requiring memorization of the subsequent actions. Part 7.04c is similar to 7.04b, requiring memorization of the caution on page 3 of ONP 525E. 7.04C is also vaguely worded, since both the LPCI heat exchanger and the SDC heat exchanger are in use, but the question only states, "..... the heat exchanger...". Part 7.04d asks why it will eventually be necessary to run the "A" or "C" ESW pump. Depending on initial conditions assumed, torus cooling may or may not be required. In fact, Step 2.8 of ONP 525E directs initiation of torus cooling only if the torus reaches l 100 F. No specific conditions were stated in the question. l l RESOLUTION: This question should be deleted from the examination, since it requires specific verbatim knowledge of the rubsequent action steps of an off normal procedure. Operators are not required to know subsequent actions,* from memory. QUESTION: SRO 8.02 l ) COMMENT: This question requires the candidate list from memory the 3 requirements a jumper bypass controlling procedure l must meet to exempt the jumper bypass from the require-I ments of ACP 2.06B, Station Bypass and Jumper Control. This question requires memorization of ACP 2.06B, Step 6.1, Exceptions. The answer key reference lists K/A 29400lK102, which requires knowledge of tagging and clearance procedures. No depth of knowledge requirement is indicated. Although not mentioned in the reference, l since none support this specific knowledge, MP1 Operator Training Text, Volume 7, Text TX 1100, lesson objectives 38-43 define the depth of knowledge MP1 requires the Senior Operators regarding ACP 2.(6B. In each case, the operator is provided with the pro:edure, since that procedure is readily available to him in the performance of his job and no actions in that procedure are considered "immediate". l l l 1

s a l RESOLUTION:. This question should be removed from.the exam since neither the K/A catalog or MP1 Operator Training lesson. objectives require memorization of exceptions listed in i the ACPs. i l QUESTION: SRO 8.03b COMMENT: This question requires the operator list, from memory, 4 actions required by ACP-QA-9.02, in the event a-surveillance does not meet the acceptance criteria. ( This question requires. memorization of ACP-QA-9.02, Step j l 6.4.5.3. The answer key reference, K/A 294001A1.03 l requires the ability to locate and use procedures and l l station directives'related to shift staffing and l l activities". This K/A clearly does not support memori-zation of ACP steps. Further,. System Training. Text, Volume 7, TX 1100, Administrative Control Procedures, l contains no objectives supporting memorization of ACPs'. l Rather, the operator is provided with the procedure, e l since that procedure is readily available to him in the performance of his job, and no actions in that procedure are considered "immediate". J k l l RESOLUTION: This question should be removed from the exam, since I neither the K/A catalog or MP1 Operator Training lesson i objectives require verbatim memorization of ACP steps. QUESTION: SRO 8.04b l i COMMENT: This question requires the operator list, from memory, 4 alternative methods of independent verifications which could be used in circumstances when-excessive radiation exposure would result. 1 Again, this question requires memorization of steps in i ACP-QA-2.20. The comments of 8.02 and 8.03 apply. These procedures are readily available to the operator in the performance of his job and no actions in them are considered "immediate". RESOLUTION: This question should be removed from the exam, since neither the K/A catalog or MPl Operator Training lesson objectives require verbatim memorization of.ACP steps. i

v-i 1 l l f QUESTION: SRO 8.05 COMMENT: The question asks the operator, in part, according to Technical Specifications, when is the independent operator verification of rod position performed. The answer key states, " prior to commencing each new rod step". Technical specifications bases, page 3/4, 3-4 states " procedural control'is exercised by-verifying 1 all control rod positions after the. withdrawal of each group, prior to proceeding to the next group." The term rod step reflects the new rod worth minimizer. 4 The terminology in the bases of techical specifications i has not yet been changed. j RESOLUTION: Modify the answer key to accept either " rod step" or " rod group". l I OUESTION: SRO 8.06 COMMENT: This question provides the candidate with a series of 1 events which have occurred, then requires the examinee i select the correct statement, and justify his choice. l Since all 4 statements contain the words " power opera-tion cannot resume", that fact does not enter into this process of choosing. Therefore, in explaining why he chose b, the candidate will not explain why power operation cannot continse, but will justify only why he felt a safety limit was violated. Further, Technical Specification 6.7.1 requires only that the unit be in " hot standby", not " shutdown", nor does Technical Specification Section 6.7.1 prescribe that exceeding a safety limit requires review by the NRC before resumption of power opeation. Additionally, although this event was initiated-by an l inadvertent closure of main steam isolation valves, the K/As stated in the reference all relats to K/A area 295005, main turbine generator trip. K/A 295005K101 and K102 require knowledge of the operational implications of pressure effects on reactor power and core thermal limit considerations as they apply to a main generator trip. K/A 295005K201 requires knowledge of the inter-relations between main generator trip and RPS. ! l 1 l

I K/A 295005K301 requires knowledge of the reason the { reactor scrams when the turbine trips. K/A 295005G005 i requires knowledge of the annunciator alarms and 1 indications and use of response instructions during ] a turbine trip. Clearly, none of these K/As support testing a candidates's knowledge in this area. RESOLUTION: Modify the answer key to state: "The reactor did not scram on the primary source signal", since the question did not ask the candidate to, state the consequences of a safety limit violation. ) Change answer key references to reflect K/A area 295020, , Inadvertent Containment Isolation. ] I

- - - - o***********

QUESTION: SRO 8.07 COMMENT: This question informs the candidate that a rod will by taken out of service according te Tech. Spec. 3.3.A.2 if it cannot be moved with drive pressure. Question 8.07C then asks the candidate, "If the control rod cannot be fully inserted before it is disarmed, what additional Technical Specification requirements must be verified?". The examinees were not provided with the required applicable sections of Tech Specs. MP1 Operator Training texts contain no objectives requiring verbatim knowledge of Tech Specs. K/A 201003K3.03 requires the operator be able to explain the impact that a loss or malfunction of a control rod drive will have on shutdown margin. However, that K/A does not require an operator to recall from memory which sections, in addition to Tech Spec section 3.3.A.2 must be checked in the event a control rod cannot be moved with drive pressure. l RESOLUTION: Remove question 807C from the exam, since the students were not provided with the appropriate sections of Tech

Specs, j

l l l 1 l 1 \\

.e dL l i

- - - - w***

l QUESTION: SRO 8.09 j 1 f COMMENT: This question requires the. candidate, list 4 major concerns the SS/SCO has when implementing'a work order which will place a Category 1 system out of service. Again, this question requires memorization of' ACP-QA-2.02C, Work Orders, Section 5.10. Answer key reference K/A 29400lK102 requires knowledge of tagging and clearance procedurer. ACP-QA-2.02C is. 1 I neither a tagging nor a clearance. procedure. K/A' 294001A103 states " Ability to locate and use procedures and' station directives related to shift staffing and activities". This K/A supports the position'that memorization of ACP sections is not required.' Students ability to use an Administrative Control Procedure should be tested, not.their ability to recall, from' memory, sections of the procedure, since ACPs contain. no actions which can be considered "immediate".. ACP-QA-2.02C, Rev. 19, page 11, Step 5.10 lists 8-responsibilities of the shift. supervisor / supervising l control operator concerning' work orders. " Radiation Controls" is not listed. Further, none of the 8.are identified as major. RESOLUTION: Neither MP1 lesson objectives nor the K/A catalog require verbatim recall of ACP sec*. ions.. Consequently, this question should be deleted from the examination. However, in the event question 8.09 remains, the answer i key should be modified to reflect ACP-QA-2.02C, Rev. 19, page 11, Step 5.10, accepting any 4 of the 8 responsi-bilities listed, since none are designated by the i procedure as " major", and 8.09.4 Radiation Controls' deleted from the key, since it is not listed by ACP-QA-202C. I 1 l _ _ _ _ _ - _ _ _ - _ - - _ _ _ _ _ - _ _ -.

e 8 + I QUESTION: SRO 8.13 COMMENT: This question provided the candidate with 4 separate 1 events, then asks the examinees to " determine the NRC j Deportability Time Requirements". EPIP-form 4701-4 has 5 headings: PIR Category, NRC (10 CFR) Reporting Criteria, Event Descriptions, State _i Posture Codes and Comments, and Notes, Examples and Additional Guidance. Further, under event descriptions, events are grouped in 4 headings, and provided with time limits for reporting the event to the NRC. For example, EPIP Form 4701-4, Rev. 6, page 2 of 14, para. II lists 4 "Non-emerger.cy events - 1 hour reports". Page 4 of 14, f Category III lists "non-emergency events - 4 hours reports", and further explains that, although these are called 4 hour reports, they shall be reported to'the NRC within 1 hour, but identified as 4 hour reports. j Therefore, the wording of this question was ambiguous,. since EPIP Form 4701-4 has,no category labeled "NRC Deportability Time Requirements". The candidate is unsure whether he should classify the events as 1 or 4 hours reports, or as the listed NRC Reporting Criteria, or as the PIR category. When questioned, the proctor directed some candidates to list the NRC (10 CFR) reporting criteria and the time limit listed under event description. 1 However, the answer key lists the PIR category for reach ) of the 4 events. j Part d falls under 2 classifications according to EPIP For.m 4701-4 and 10 CFR 50.72 and 50.73. This event could be classified as a 4-hour report, according to 10 CFR 50.72 (b)(2)(iii), and EPIP Form 4701-4, page 5. However, according to 10 CFR 50.73(a)(2)(v) and EPIP Torm 4701-4, page 11, this event is also a 30-day LER. RESOLUTION: Modify the answer key to accept either the PIR category, NRC (10CFR) reporting criteria or the reporting time requirement listed in the event description for each event as shown below. a. A1, immediate or 1 hr report or 10 CFR 50.72(b)(1)(iiTT l ---_______________a

I'd;; : 1' a b.' A1, immediate orL4 hr report (reported or 10 CFR ,50.72(b)(2)(iiT to NRC.in 4 hours) c. A1, immediate'or 4 hr report (reports or 10 CFR ~~ 50.72(b)(2)(iiTT to NRC in 1 hour); d. Either of 2-categories apply: 10-CFR 50.72(b)(2)(iii) to'NRC in i hour) -OR: 30 day LER or-30 days or 10 CFR'50.73(a)(2)(v)' l l l l I m

n i 1 6 NRC Comment Resolution l l Il The facility questioned'the reliability and validity of the R0 and SR0 i written examinations. The facility expressed a concern that the candidates' scores on the examinations may not be an accurate reflection of their competency or their ability to safely operate the plant to protect the health and safety of the public. q The NRC performed a thorough review of the facility comments and utilized the facility training material; NUREG-1021, ' Operator Licensing Examiner Standards' (Examiner Standards); 10 CFR Part 55; the questions asked by the i candidates during the administration of the examination; and the results of the examinations to address the facility concern. The NRC determined that L the written examinations administered on Nov2mber 7,1988 were a reliable and valid measure of the candidates' competency and ability'to safely operate the plant to protect the health and stfety of the public. Summary Comments l The facility summarized the individual question comments. Each of these summary comments was addressed with respect to the validity of the examinations. The facility commented that some questions referenced' equipment design features not present on Millstone Unit 1. Two questions common to the R0 and SR0 examinations and one question unique to the R0 examination referenced equipment design features not present on Millstone Unit 1. In one case (on both the R0 and SR0 examinations) the design feature was identified in the question as being generic and not specific to Millstone Unit 1. In the second case (R0 and SRO) the non plant specific feature was not a key part of the question and did not affect the answer. The question was clarified for the few candidates that asked about it. The third case (R0 only) was a typographical error. No questions were asked during the administration of the examination concerning this question. The validity of the examinations was not adversely affected by any of these questions. (Questions 1.05/5.05, 1.06/5.06 and 2.02a) The facility commented that many questions were ambiguous, unclear, or not specific. The facility identified four questions (two of which were individual parts of the same question) that were common to the R0 and SRO examinations, six questions (two of which were individual parts of the same question) unique to the RO examination and six questions unique to the SR0 examination that they felt were confusing to the candidates. In the majority of these cases, the facility comment was not supported by candidate questions raised during the-administration of the examination. During the

l = I 2 e a i pre-examination briefing the candidates were told that if they had any questions as to the intent of a question that they should have them clarified by the proctor. All questions were answered by the proctor and when appropriate announced to all the candidates. (Questions 2.06b/6.06b, 2.06c/6.06c, 2.07/6.07, 2.11b, 2.12a, 2.12b, 3.03b/6.03b, 3.05, 4.04, 4.07, 6.08, 6.09, 6.10, 7.04, 7.06 and 8.13) The facility was concerned that the candidates were required to spend a significant amount'of time attempting to determine the intent of several questions, causing them to be pressed for time toward the end of the examinati on. Additional time (approximately 1/2 hour) was allowed for completion of the examination to account for several interruptions during the examination. All the candidates except one completed the examination in less than the allotted time. The one exception stated that he had completed the examination and was only reviewing, when questioned by the proctor. There is no indication that confusion adversely affected the validity of the examination. The facility commented that many questions did not elicit the response provided on the answer key. In six of the nine cases identified by the facility the answer key was changed as recommended by the facility. In one case the question was deleted because the answer key could not be revised I appropriately. The NRC disagreed with the facility comment in two cases so no change was made to the answer key. The comments were resolved with changes to the answer key when required, therefore there was no adverse effect on the validity of the examinations. (Questions 2.02a, 2.12a, 3.02/6.02, 3.05, 3.09b, 3.10, 4.11, 7.06 and 8.06) The facility commented that some questions contained statements which were inaccurate or contradicted Millstone Unit 1 Technical Specifications. The two cases where this occurred were caused by typographical errors. The i errors were not brought to the attention of the proctors during the l administration of the examination, therefore there is no evidence.that they were noticed by the candidates. There is no indication that either of these questions caused any confusion or adversely affected the validity of the examinations. (Questions 2.11b and 7.04) The facility commented that the answer key for some questions contained technical inaccuracies. Because only the answer key was affected and was corrected appropriately, there was no adverse effect on the validity of the examinations. (Questions 3.01/6.01, 3.03a/6.03a, 3.08a, 3.08b, 4.02a/7.02a, 4.02d/7.02d and 7.05a) The facility commented that the conditions in some questions were altered during the administration of the examination, requiring additional time to be-spent by the candidates. The candidates were interrupted three times during the administration of the examination to clarify questions for the entire group. In one of these cases the clarification altered the conditions of the question. The SR0 candidates were instructed to skip a question when it was

d 3 brought to the attention of the proctor that a procedure referenced.in the question had been cancelled. This question was replaced by a similar question written and reviewed by the examiners. The candidates were told that additional time (approximately 1/2 hour) would be allowed for completion of the examination due to the interruptions.. All the candidates completed the examination within the allotted time. There is no indication that interruptions during the administration of the examination adversely affected the validity of the examination. (Questions 1.02e/5.02e, 207/6.07, 3.03a/6.03a, 3.03b/6.03b and 7.07) l The facility commented that several candidates were given misleading direction by the proctors, who were not fully knowledgeable of the examination questions. The proctors present for the majority of the administration of the examination, though not the authors, had reviewed the examinations and were knowledgeable of the questions. When' attempting to clarify questions it is not always possible to supply specific information, therefore it is possible' to inadvertently mislead the candidate. The facility noted five cases in which the candidates could have been misled by clarification supplied by the proctors. In three cases the question was deleted because it was not possible to determine what i assumptions had been made by each candidate..In one case the clarification j provided by. the proctor was taken into account in grading the examinations. In s the final case the alternate wording recommended by the facility was accepted. The facility comments were resolved to ensure that no potentially misleading information provided by the proctors affected the validity of the~ examinations. j (Questions 2.06b/6.06b, 3.03a/6.03a, 6.08, 6.10 and 8.13) i The facility commented that some questions were out of the scope of knowledge Millstone 1 requires of R0s or SR0s and were not supported by K/A references-or MP1 lesson objectives. The facility identified six questions that they felt were beyond the scope of knowledge required for reactor operators, one l-question that was beyond the scope of knowledge required for senior reactor operators, and one question that was beyond the scope-of knowledge required for both R0s and SR0s. In all cases except one the NRC determined that the questions were within the scope of knowledge required for reactor operators and/or senior reactor operators. The questions are supported by.K/A references and, where possible, MP1 lesson objectives. Part of one question was determined to be beyond the scope of knowledge required for R0s and it was deleted from the examination. The facility training material does not differentiate between the knowledge required for reactor operators and that required for senior reactor operators. The questions were determined to be valid with respect to scope of knowledge required for R0s and/or SR0s and do not detract from the validity of the examinations. (Questions 2.02b, 2.06c/6.06c, 3.06a, 3.10, 4.03, 4.09, 4.11 and 6.09) The facility commented that some R0 examination questions (Section 4), and SRO examination questions (Section 7), required recall of the material supplied in the subsequent action steps of procedures. The facility did not identify any questions on the R0 examination that required recall of I I A

a 4 I J subsequent action steps. They identified three questions on the SR0 examination that required recall of subsequent action steps. The NRC determined the questions to be valid in all cases. Three questions (or parts of questions) asked for the purpose for. performing the designated step. The candidates were provided with the portion of the step that listed the action,- i but were not given the portion of the step that discussed the purpose of the -l step.- Operators are required to know from memory the reason for performing 1 subsequent action steps whether or not the purpose of the step is provided in the procedure. Two questions (or parts of questions) asked for general descriptions of methods used in the procedures, not verbatim knowledge of subsequent action steps. The questions do not detract from the validity of the examinations. (Questions 7.04, 7.07 and 7.09). The facility commented that 34.6's of Section 8 of the SR0 examination j required. recall of Administrative Control Procedures (ACPs) or Technical Specifications without access to the procedures or Tech Specs, which is not supported by the K/A catalog or MP1 lesson objectives. The facility j l identified four questions that required recall of ACPs. Three.of these i questions tested the candidates' ability to perform routine administrative i tasks and did not require verbatim recall of administrative procedures. They were determined to be valid questions because the SRO is not required to refer to the ACP when performing routine tasks. These questions'are supported by K/A references. One question was deleted because it was determined that the subject of the question was not a routine task. The facility identified one question that required recall of Technical Specifications without access.to the applicable section of Tech Specs. This question tested for an understanding of related specifications and did not' ask for verbatim recall of a specific section of Tech Specs. This question was determined to be valid and is supported by K/A references. These questions did not adversely affect the validity of the examination. (Questions 8.02, 8.03b, 8.04b, 8.07c and 8.09) The facility stated that the majority-of these-problems would have been identified in a pre exam review by the facility and that most comments could have been resolved before the examination was administered to the candidates. The results of the post-exam review that was performed immediately following the administration of the examination do not support this statement. Only a small percentage of the comments that affected the question portion of the items were identified at the post-exam review. The majority of the comments that identified questions as confusing or inappropriate were generated as'a result of candidate comments following administration of the examinations and were not identified to the NRC _ until two weeks (nine working days) after the administration of the examinations. Since the comments were not-identified l-in the two hour post-exam review, there is no reason to believe that they would have been identified in a pre exam review.

t j 5 Individual Comments Each of the specific facility comments on individual questions have been resolved as follows: j l 1.05/5.05: Disagree with comment. The words "in a generic BWR" were l-added to the' question. prior to administration of the examination and was discussed in the preexamination briefing with the l candidates. l The Examiner Standards, section ES-202,. Scope of Written: Examinations Administered to Reactor Operators - Power Reactors, 1 l allows the use of questions related to reactors in general l (generic BWR) and reactors of the type used at the facility, d . i No questions related to the facility comment were asked by the candidates during the administration of the examination. This question did not result in any unnecessary confusion or prevent-I any candidates from completing the examination in the allotted ) time. ] l No changes were made to the question or answer key. 1.06/5.06: Disagree with comment. The non plant-specific design feature referred to was the cause of the feedwater train isolation. The question clearly stated-that this isolation occurred. J Understanding of the cause of the isolation was not required to answer the question which asked for-the effect the coefficients had on mitigating or increasing the severity.of the transient. The two candidates that asked for clarification were told that' a loss of feedwater heating had occurred. 1 1 The Examiner Standards, section ES-202, Scope of Written i Examinations Administered to Reactor Operators - Power Reactors, i allows the use of questions related to reactors in general j (generic BWR) and reactors'of the type used at the facility. This question did not result in any unnecessary confusion or i prevent any candidates from completing the examination in the allotted time. i No changes were made.'to the question or answer key. l i l 1.08/5.08: Agree with comment. The answer key was changed to accept ~ either P = Poe t/T or Period = DT x 1.44 for full credit based on the additional reference supplied by the facility. l

w i 6 1 q i i 2.02a: Partially agree *with comment. No questions related to the l facility comment were asked by the candidates during the administration of the examination. This question did not result in any unnecessary confusion or prevent any candidates from 1 completing the examination in the allotted time. Question a.2 will be reworded to prior to uploading to the NRC i Exam Question Bank (EQB) as follows: State how accumulator charging flow changes and how it is limited following a scram. In addition a.3 was modified to " suction valve" prior to uploading to the EQB. l Set points or values were not required for full credit. The answer key for a.2 was modified to accept " flow goes to maximum" j for full credit. i 2.02b: Comment noted. The following KA's are applicable to this question: Part a.1 - 201001K109 (3.1-3.2), 201001K603 (3.0-2.9); Part a.2 - 295006K203 (3.7-3.8), 295006A106 -1 (3.5-3.6); Part a.3 - 201001A202 (3.2-3.3); Part b. - 201001K106 (2.8-2.8), 201001K606 (2.8-2.8). These KAs were j added to the reference and all nonapplicable KA's were deleted from the reference prior to uploading to the EQB. l } Lesson objective 39 from the CRD lesson plan requires knowledge of how proceduca t prerequisites support system operation and OP 302 requires TBSCCW as a prerequisite. This objective was added to the reference and lesson objective 27 was deleted from the reference prior to uploading to the EQB. 2.06a/6.06a: Partially agree with comment. The answer key was modified to. i require both "the relief valve (set (a 375 psig) [0.25]" AND "the i interlock between the core spray admission valves [0.25]." Total ) point value remained the same. 1 Page 12 was added to the reference. 2.06b/6.06b: Disagree with comment. Only one candidate questioned the proctor concerning the intent of the question during the administration of the examination, therefore it 'is not reasonable to assume that the question caused any unnecessary confusion or prevented any candidate from completing the examination in the allotted time. w_-_.-_---___._ _._-____.n-__----._-.__a.__-____

7 Because it is not possible to determine how each candidate interpreted this question, it was deleted from the examination. Overall and section point values were reduced by 0.5. 2.06c/6.06c: Disagree with comment. The question specifically asks for two ACTIONS the control room, operator can take during accident conditions to improve the NPSH of the core spray pumps. Changing either of the parameters which are the axis of the E0P NPSH curves is required for full credit. The facility's comment that the E0Ps are available for use by the operators during plant operation is noted. During the I operating portion of the examinations it was noted that the Control Room Operators did not utilize the E0Ps directly, but l were directed in their actions by the Control Room Supervisor. A reactor operator needs to know what parameters to monitor to ensure proper operation of equipment. The facility comment that the statement in parenthesis was confusing and distracting is not supoorted by any questions raised by the candidates to the proctor. The one candidate that asked was supplied with a copy of the caution. The statement in parenthesis was deleted from the question prior to uploading this question to the EQB. 2.07/6.07: Partially agree with comment. No logic problem exists in the question as originally written. High radiation levels in the Main Steam Lines cause two relays to actuate simultaneously initiating a Group 1 Isolation and RPS Trip at the same time. r If neither action occurs then answer 3 is correct. Removing the word "first" from the question changes only part a. of the answer. This change was announced to all the candidates. No questions were asked concerning how far in excess of 115% steam flow was in part c. Part c. of the question was deleted i l from the exam. Category and overall point values were reduced by 0.5. Full credit will be given if the candidates substitute actions 1 I & 2 for action 4 The answer key was changed as follows:

a. 4 (Both I and 2) l

b. 4 (Both 1 and 2) l

c. deleted

d. 1 (RPS Trip)

8 I 1 2.08a: Partially agree with comment. The question was specific in asking the location in relation to the MSIVs. Full credit will . be given if the candidate states that the monitors are located in the steam tunnel, downstream of the MSIVs..The word i " outboard" will not be required as long as it is clear that the monitors are not located between the_ inboard and outboard MSIVs. .) 1 2.08c: Agree with comment. H-16 radiation is acceptable' alternate wording for water activation. 2.11b: Partially agree with comment. This question was asked-in the context in which the information is presented in the training 'l p material. The training material should be corrected to be. I consistent with Technical Specifications. The answer provided' by the facility will be accepted for full credit. The word "not" was inadvertently omitted from the question. No questions were asked by the candidates concerning this, therefore it is not reasonable to assume that any confusion was caused by.this typographical error. '1 E0P text page 136 was. removed from the reference section. 2.12a: Partially agree with comment. The answer key is in. direct agreement with the description of the transient in the training material. No questions were asked by the candidates during administration of the examination concerning this question. Because no endpoint was defined'and it is difficult to determine whether or not pressure stabilized during the transient, -\\ l candidates will be given full credit if they omit the word " stabilizes". The method of pressure control is not required 1 for full credit. The word " curves" was replaced by " plant' parameters" for clarification in the question. i -i l 2.12b: Disagree with comment. The facility comment that the question i as asked was confusing to the candidates is not supported by any l questions raised during the examination. The opening context ofL { the question discusses loss of one recirculation pump from 100% power. Two recirc pumps are required to' achieve 100% power operation, therefore the status of the other recirc pump is not a point of confusion. l l 1

9 ONP 504, Recirculation System Failures, directs the operator to close the affected pump discharge valve followed by a 5 second jog open after 5 minutes. This action will reduce the indicated recirculation pump loop flow to less than the minimum sensitivity for indication in an idle loop. This basic system knowledge should eliminate any confusion on the part of the candidates as to which loop flow the question was specifically addressing, as recirc pump loop flow is not in the indicating range following a pump trip. The second part of the question is not confusing but merely a continuation of the sentence context asking how is total core flow determined with an idle recirculation loop. 3 No changes were made to the question or answer key. 2.13: Disagree with comment. The reference material does not support the interchangeable terms supplied in the facility comment. The descriptions on pages 2 and 7 of the text and figure 1 of the 125 volt DC Distribution System lesson plan, use the following terms for the normal and alternate power supplies to 125 VDC switchboard 101-A: Normal - Battery Charger 101-A Alternate - Battery 101-A or 18-A (Standby) Battery Charger 101-C (125 VDC) Bus 101-B The answer key was changed to indicate the correct terminology in accordance with the facility reference material. Because t?,e p.afix 101 is common to all batteries, battery chargers, and switchboards it will not be required for full credit. 3.01/6.01: Partially agree with comment. The answer key was modified to indicate Reactor Vessel Water Level of +8" as the correct SBGT initiation setpoint. The question was specific in asking for the CONDITIONS which would initiate an automatic start of SBGT. A Group II isolation occurs as a result of the same conditions that initiate SBGT. It is not a condition that causes an initiation of SBGT. Group II isolation will not be accepted in place of high drywell pressure or low vessel level conditions. l l

L ~ 10 3.02/6.02: Cisagree with comment. The question was specific in asking for the overall response of the entire turbine generator system AND. turbine pressure control system. The question specifically. states that the turbine throttle valves fail to close (the-I system fails to runback). The runback cc.idition is not terminated until steady state conditions are reached. All parts of the answer as shown in the answer key are required for full credit. No change was made to the answer key. 3.03a/6.03a: Partially agree with' comment. The' initial assumption that water level is -40" and drywell pressure is 3 psig was stated in the question for the R0 candidates and was announced to the SRO candidates during the exam. Because some candidates may have been confused by the " sealed in" terminology, part a. was deleted from the examination. The section and overall point totals were reduced by 0.5. The -40" was corrected to -48" when the question was uploaded to the EQB. 3.03b/6.03b: Comment noted. Extra time was allowed for completion of the examination to make up for the interruptions during the examination administration. The facility comment that the timing of the relay failure caused confusion is not supported by any candidate questions during the administration of the examination. 3.05: Partially agree with comment. The facility comment that it is unclear exactly what is required by values of the interlock is-not supported by any candidate questions during the administration of the examination. The question refers to recirculation pump speed as part of the interlock. Speed demand limited to 28% will be required for full credit. "A condition which could result in axial. thrust damage to the RR pumps" will not be required for full credit. ~ ~ g. 6 %. - mm 1h> _~ .r ~, e

l '. 11 3.06a: Disagree with comment. The action required by the Technical Specifications was given in the question and did not require the RO to memorize Tech Spec tables. In addition the APRM bypass switches in the Control Room are caution tagged to alert the operators to this specific condition. The KAs referenced by the NRC are applicable to the question. I The KA referenced by the facility is not applicable because the question does not test the candidates' ability to recognize entry conditions for Technical Specifications. l No changes were made to the question or answer key. 3.08.a.3: Agree with comment. The reference discussed by the facility was not supplied to all the requested examiners for examination preparation or with the comment submittal. The correct answer l was verified by a review of the electrical control drawings. i The answer key was changed to indicate N0 as the correct response to part 3.08.a.3. I i 3.08b: Agree with comment. Answer key was changed to indicate 14A, 14C, 14E, 14G, and 14F as the correct responses. 4 l 3.09a: Agree with comment. The facility comment is acceptable i alternative wording to the answer key and is technically ] correct. i The additional reference discussed by the facility was not supplied for the initial examination preparation or with the comments for review. 3.09b: Agree with comment. "The volts to operable input ratio is 1 to 1" will not be required for full credit. KA 215002K501 (2.6-2.8) was added to the reference. The KA reft nced in the facility comment was not listed in the reference. It appeared in a computer generated list that followed the reference, but was not applicable to the question. 3.10: Partially agree with comment. The answer key was modified to accept the alternative wording provided by the facility. The first sentence of the answer key was not required for full credit, but a fully correct answer must indicate that #1 seal I flow is below normal. l L l

O 12 i The facility lesson objectives and section ES-202 of the Examiner Standards support this question. The Examiner Standards state that a reactor operator "should be able to make use of all available instrumentation to provide checks or verification of observed readings". A reactor operator should be able to integrate the information from a combination of alarms in addition to understanding the individual alarms. KA 202001K404 is appropriate for this question, the candidate must understand normal controlled seal flow to determine what is abnormal. The facility referenced KA 202002A210 is not in the current revision of NUREG-1123. It was assumed that the comment referred to KA 202001A210 (3.5-3.9) which was added to the reference section. The KAs referenced in the facility comment as inappropriate were not listed in the reference. They appeared in a computer generated list that followed the reference, but were not applicable to the cuestion. The facility comment concerning daggers in the KA catalog is in l error; NUREG-1123, section 1.6.5 defines the dagger as i I indicating the LEVEL of knowledge or ability is different for SRO and RO candidates, not that is inappropriate for R0 candidates. 1 3.11: Agree with comment. The facility reference material does not have a clear definition of a " Tripped Breaker" in the text to s1pport lesson objective 37. The facility should clarify the text supporting this lesson objective. The question was reworded to " STATE THREE local indications of a TRIPPED 4160 volt breaker" prior to upload to the EQB. Closing st ring status flag "CHGD" was removed from the answer key. l 4.02a/7.02a: Agree with comment. The reference identified in the facility comment was not provided to all the requested locations for examination preparation or-resolution of comments. The question gives an initial condition of one shutdown cooling pump in operation. The referenced text, page 69 of 90, is very explicit when addressing this condition as requiring TWO shutdown cooling pumps operating at maximum flow. Additionally

J 1 13 the shutdown cooling text', page 14 indicates max flow of TWO-shutdown cooling pumps is required to prevent stratification whenever the: recirc pumps are stopped. This information is. ~ confusing when compared to the referenced procedure and should be corrected. The answer key was changed from Yes to No. 'I 4.02b/7.02b: Agree with comment.. The reference identified in the facility comment was not provided to all the requested locations for examination preparation or resolution of comments. Because vessel level is required to accurately determine if stratification will occur this question was deleted.from the exam. Section and overall point values were reduced by 0.5. 4.02d/7.02d: Agree with comment. The reference identified in the facility comment was not provided to all requested locations for examination preparation or resolution of comments. The question states that both shutdown cooling pumps are operating with the A recirc pump.in operation. The referenced. text does not identify a stratification problem.when operating j under these conditions. 1 The answer key was changed to indicate No as the correct. response. 4.03: Disagree with comment. Lesson Objective 26 of the-referenced text, TX-15008, is directly applicable to this question. The objective specifically states that an operator should know how and why vessel level affects reactor power and makes no-differentiation between R0 and SRO knowledge. level. The text provides a-detailed discussion of how and why power is effected by vessel level during an ATWS. I All of the referenced KAs are applicable and of sufficient value to justify this question. The facility interpretations of. the KAs associated with this question are incorrect. An understanding of."how and why" vessel level affects reactor power is required for a complete understanding of the concepts associated with performance of the E0Ps. 4 l

) 14 l Even though there is no differentiation in the training material between R0 and SRO required knowledge, it was determinad that .{ part c. was beyond the knowledge required for a reactor operator. Part c. was deleted from the question. Section and overall point values were reduced by 1.0. No change was made to the answer. key for part b. Exact wording is not required. i 4.04: Partially agree with comment. The facility comment concerning confusion regarding this question is not supported by any questions raised by the candidates. The answer key was modified to include the third _ procedure step from ONP 5038. Candidates will be given full credit for providing any two of the procedure steps stated in ONP 503B. The question was modified prior to uploading to the EQB as follows: List ALL the immediate actions required by ONP 503B, Loss of All Station AC power. ] J 4.06: Agree with comment. The additional answers were not included in the originally referenced material. The referenced material should be corrected. The answer key was modified to include the additional scrams and 3 references provided by the facility. I l 4.07: Partially agree with comment. The facility comment about the overall confusion present-in this question is not supported by l any questions by the candidates. j The facility comment concerning event A is not valid. The question did not require the candidates to analyze the plant conditions prior to the event. Regardless of the cause or automatic actions of standby equipment, an automatic action consisting of a reactor scram would have occurred at 53.5 psig requiring entry of ONP 502. All reactor scrams do not result in vessel level dropping below +8", for example if the scram occurred during startup it is unlikely vessel level would fall to +8", however, if any candidate makes that assumption and lists E0P 570, 571 and 572 credit will not be subtracted. ____________-__-___-L

15 The facility comment concerning event C is not valid..The. L question did not require the candidates to analyze the plant conditions prior to the event. Regardless of the cause of the event,'the procedures listed would be utilized. The answer key was. modified for part d. to list only E0P.504 as the.first order procedure with a point value of (0.5), since the question did.not.specify to include more than first order proceduresi The_ answer key was modified for part f. to list only E0P 515A as the first order-procedure with a point value of (0,5), since the question did not specify to include more than first order 4 procedures. I The parenthesis on the answer key signifies information which is not required for full credit and includes the second order procedures. 4.09: Disagree with comment. The candidates were not required to memorize a word-for-word definition'of each NRC event classification, these were given in the question. 'At a minimum, a reactor operator should have an understanding of the severity of the events.and given the definitions, be able to relate them in order of severity. The KAs in the reference section'were replaced-by the following KAs:.295033K305 (3.6-4.5), 295038K205 (3.7-4.7), 295038K301 (3.6-4.5), 295017K102 (3.8-4.3), 295017K206 (3.4-4.6), 295017K303 (3.3-4.5), 295023A205 (3.2-4.6). These KAs indicate the need for an R0 to understand the relationships between plant conditions and the NRC emergency action level classifications. L Candidates will be given full credit if they provide the correct l Sta.te of Connecticut Posture Code term applicable to the I corresponding NRC term. The question was modified to include NRC in the opening-statement. 1 4.10b: Agree with comment. The alternate wording provided!in the l facility comment is acceptable. l 4.11: Partially agree with comment. This question is in accordance-with section'ES-202 of the Examiner Standards and 10 CFR-55._41. 10 CFR 55.41 does not require that all knowledge requirements be j based on facility learning objectives. The answer to this 4 .1 S - D I

l 16 l question _was taken directly from the facility training material. There is no differentiation between R0 and SR0 level of knowledge in the training material. The facility referenced KAs 290528K201 and 270528K201 are not in l the current revision of NUREG-1123. It was assumed that the l facility was referring to KA 295028K201. This KA is applicable because the restrictions on drywell venting above 212 degrees F. are interrelated with containment spray and the containment vacuum breakers. This relationship is required for a total understanding of the restriction on venting above 212 F. Additional KAs were added to the reference section, 295028K303 (3.6-3.9),223001K501(3.1-3.3),223001A210(3.6-3.8), 223001G010 (3.2-3.6). All of the referenced KAs are applicable and of sufficient value to justify this question. This question was deleted from the examination because it was determined that the question did not elicit the answer provided in the answer key. Section and overall point values were reduced by 2.0. 6.08: Partially agree with comment. The facility comment concerning the answer key is correct assuming the diesel is the only source j on the bus. L I The candidate who was told to assume the diesel was in parallel with bus 14D will be graded in accordance with the assumption stated by the proctor. Any candidate who did not raise a question during the examination but stated his assumptions in his answer will be graded accordingly. All other candidates will be graded in accordance with the answer key. " Assume the diesel generator is the only source of power supplying the bus" was added to the question for clarification prior to uploading this question to the EQB. 6.09: Disagree with comment. The question is supported by the referenced lesson objective, except that the procedure was not provided. An operator should be able to state the purpose of a safety system without referring to the procedure. The wording of the question was specific enough to elicit a discussion of a failure of FWCI to restore level.

37 -1 ) The answer key was'a direct quote'of the training material, but..- for clarity the answer key was modified as follows: In the event. I of a failure of the FWCI system or a break. larger than the capacity of the FWCI system [0.5] the APR system. opens selected SRVs to rapidly depressurize the reactor. [0.5) to within the injection pressure of the Core Spray and LPCI (Low Pressure ECCS Systems) [0,5]. KA 218000G004 (4.0-4.1) was added to the reference section. KA 1 218000K301 is applicable to'this question because understanding of the system purpose is directly related to knowledge.of the impact'of a system failure. 6.10: Partially agree with comment. If a candidate carried the answer through reactor pressure decreasing to 900 psig, no credit will be subtracted. Part b'was deleted from the question and the section and overall point values were reduced by 0.5. The answer key for part d. was modified.to eliminate the ~ designation of. partial credit. Reactor pressure of 350 psig is required for full credit. 6.11: Comment noted. Part a. of the question was deleted prior.to upload to the EQB. 7.03: Agree with comment. No change to the answer key was required. Exact wording is not required for e.hort answer or essay type answers. 7.04: Partially agree with comment. The question did contain an error in the opening statement however, each individual.section of' the question references (if necessary) the correct procedure-by number and name. Part al. is applicable to both procedures. Part a2. obviously refers to ONP 503B because it specifies that the diesel and gas turbine.are'available. Part b. refers to the procedure by the correct number and name. In parts c. and d. knowledge of the electrical lineup is not required to answer the question. i l

m.

i. q.

18 l The facility comment that the question:" confused most L candidates" is not supported by candidate questions during the examination. Only one candidate raised a question concerning i part d. only. The facility comment concerning whether the: ? question was testing the ability of the candidates-to match the correct name and number of the procedure is not supported by any candidate questions raised during therexamination.. The error would have been corrected had it been identified by the candidates during the' administration of the examination. The subsequent action step discussed in part b.is taken directly~- from ONP 5038, Rev. 1, Change No.l. ~The question is a direct-quote of the step.and no emphasis is placed on any part of the-step. The partial credit distribution was changed to give equal' credit to the reason for checking radiation levels and the reason for restarting the compressor. The portion of the answer concerning why the compressor.is-started is taken-from the ONP lesson plan which discusses providing operational control of the safety / relief valves and the MSIVs inside the drywell. The terminology " reactor pressure control equipment" was used as a summation of this equipment'and was not referring to.the turbine pressure control system. It-was assumed that the facility was referring to the. turbine pressure control system in their. comment. l Part c. states a specific action that is taken for_ a particular malfunction and asks the candidates to explain'why this action is taken. The name and number'of the procedure is not required to answer this portion of question. Even though the reason.for the action is stated in the procedure for part c., the operator is still responsible for understanding the purpose of the step. In an emergency situation, the E0Ps-take precedence over the off normal procedures and the ONPs may' not be referred to for a significant period of time. In emergency conditions it is important that the operator have knowledge of the purpose of the steps in the ONPs so that important subsequent action steps such as conserving the gas turbine battery will be taken in a timely manner. The answer key was modified for 7.04c to require only " conserve the Gas Turbine Battery" for full credit. Any assumptions stated by the candidates were taken into considerations when grading part d. i __.___..___..__..m_________

19 7.05a: Partially agree with. comment. While it.is 'true the reference text does not use the exact terminology " negative consequence" it does provide a lengthy discussion of the effectiveness of steam cooling and lengthening the time which would be available-to line up alternate injection systems based on waiting until 4 -220" to open the SRV. The answer key was modified to accept decreased steam cooling effectiveness [0.5] and decreased time available to line up alternate injection systems. [0.5]. I The additional facility reference material, E0P appendix B, was not supplied with the facility comments or with the material supplied to develop the examination. KA 295031K304 (4.0-4.3) was added to the reference. The referenced KAs support this question and are of sufficient value to justify this question. Lesson objective 24 was deleted from the reference section. 7.06: Disagree with comment. The facility' comment that this question is misleading,' ambiguous and non-specific is not supported by any candidate questions-raised during the examination. The question does not infer that there are only two transients covered by.the E0Ps. Each of the referenced procedures require prevention of the blowdown under specific' conditions. A j description of these conditions is not required to. answer the I question. ( q No proposed resolution for a change to the answer. key was provided by the facility. The question was reworded to include the specific steps that require prevention of the blowdown prior to uploading to the EQB. 7.07: Disagree with comment. Although ONP 511 was cancelled prior to administration of the examination, the facility did not inform. the NRC. When it was identified to the proctor that the procedure was no longer in effect, all the candidates were alerted to skip the question until it could be replaced. It was assumed that the candidates had been trained on the procedure change, so to be fair to the candidates and to prevent having to delete the question afterwards, a new quest: ion on the same subject matter was prepared. l

20 The change to this question did not prevent any candidates'from completing the examination in the allotted' time. Although ONP 525A directs.the SCO to'the reactor. building in a i note following step 2.5, step 2.5.1.1 directs him to perform actions in the Isolation Condenser (IC) area which lead to the i function of controlling reactor pressure with the.IC. The question was specific in' stating "briefly describe his FUNCTION at that location". Full credit will be given if the candidate states " reactor building" as long as he' discusses use of the IC l in his response. No change to the answer key was required. This question is in accordance with section ES-402 of the Examiner Standards which states: "The candidate should be able to describe generally the objectives and methods used in the normal, offnormal, and emergency operating procedures and the-methods used to perform the verifications". The candidates were not asked to recall from memory the exact details of the procedure. References were supplied with the answer key provided to the facility however, the KAs were not included because NUREG-1123 was not available during the administration of the examination. The KAs which are applicable to this question are as follows: 295016K202 (4.0-4.1), 295016A108 (4.0-4.0), and 295016A109 (4.0-4.0). None of these KAs require the use of procedures to perform basic functions. 1 The plant walk-through portion of the examination has no bearing on the written examination. The facility comment concerning the question asked during a plant walk-through is irrelevant. 7.09: Disagree with comment. The facility comment concerning. l question 7.04b was assumed to apply to question 7.09b. The facility comment concerning question 7.04c was assumed to apply to question 7.09c. The facility comment concerning question 7.04d was assumed to apply.to question 7.09d. This question is in accordance with section ES-402 of the Examiner Standards which states: "The candidate should be able to describe generally the objectives and methods used in the normal, offnormal, and emergency operating procedures". The candidates were not asked to recall from memory the exact details of the procedure. Part a. asked for a general description of the method used for alternate shutdown cooling. Parts b., c. and d. concerned the objectives of the procedures. Even though the reason for the actions are stated in the procedure for parts b. and c., the operator is still responsible for understanding the objective of the action. \\.

I i .e o 21 KA 205000A203 was deleted from the reference. The remaining KAs q are applicable to this question. The following additional KAs were added to this question: 295021A204'(3.6-3.6),295021A205l J (3.4-3.5), 295021G003 (3.1-3.6), 295021G007 (2.9-3.2), 295021G008 (3.2-3.9). .) The facility comment in reference to the statement;"the' heat exchanger" in part c. being vague is not supported-by any candidate questions raised during the examination. The. question is specific in referring to reactor coolant temperature. The l question does not require knowledgeLof which heat exchanger in ] the flowpath the reactor coolant is flowing through to provide the correct answer. Also the LPCI heat' exchanger is used for torus cooling, not for cooling water directly from the reactor. l With flow to the torus it is expected that torus temperature will " eventually" increase to 100 F. The precedure.does not require running ESW pumps for any other reason. l Question 7.09 was not deleted from the examination. 8.02: Agree with comment. Question 8.02 was deleted from the examination. l I J 8.03b: 01sagree with comment. This question is in accordance with section ES-402 of the Examiner Standards and 10 CFR 55.43. 10 CFR 55.43 does not require that all knowledge requirements be based on facility learning objectives. 1 i This question does not require memorization of the procedure. It tests for an understanding of what is required when a component fails a surveillance.. KA 294001A103 is applicable to this question because the ability to use procedures related to j station activities includes a knowledge of how to perform routine administrative tasks. ACP-QA-3.02, step 6.6.1.4 states that " procedures for actions that are routine and commonly performed are not required to be located at the performance site." For a routine procedure, such as reviewing. surveillance results, an SRO would not be expected to refer to the administrative procedure even though it is readily available. Question 8.03b was not deleted froin the examination. I i

3 L 3 22 8.04b: Disagree with comment. 1*is question is in accordance with-section ES-402 of the F M ner Standards and 10 CFR 55.43. 10 i CFR 55.43 does not resin that all knowledge requirements be based on facility learnNg objectives. l This. question does not require memorization of the procedure. It tests for an understanding of what alternative methods are acceptable for performing verifications. An SR0 should know when the normal method of independent verification is not l-required and what alternative methods are acceptable without referring to the administrative procedure. 1 KA 294001K101 is applicable to this question'because knowledge of the acceptable methods available for performing verifications j is integral to conducting and' verifying valve lineups. Question 8.04b was not deleted from the examination. I 8.05: Agree with comment. The answer key was modified to accept " rod' i group" as alternate wording for " rod step". l l 8.06: Agree with comment. An explanation of why power operation cannot resume is not required for full credit however,-the candidate must justify why a safety limit was violated for full 4

credit, i

The answer key was modified to state that the unit must be placed in Hot Standby. The statement "and review.by the NRC j before resumption of unit operation" was' deleted from the answer ] key. Question point value' remained the same. ] The alternative wording requested by the facility will be j accepted in place of; " Failure of the primary source scram j l signal (steam line isolation valve closure scram) may have allowed a safety limit to be violated". The remaining portions l of the answer are required for full credit. The KAs were changed to the following: 295020K201 (3.6-3.7), 295020A206 (3.4-3.8), 295020G003 (3.1-3.9), and 295020G002 (2.9-4.1). 8.07: Disagree with comment. This question is in accordance with I section ES-402 of the Examiner Standards which states: " Questions concerning the Technical Specifications will require I a thorough knowledge of what items are addressed in the specifications, the bases for the requirements and how to comply with the requirements." i i I I

'~ 23 The answers _for parts a. and b. are found in the Technical Specification Bases. Part c. of the question did not require verbatim knowledge of a' specific requirement. The question tests that the. candidate recognizes that shutdown margin could be affected by a rod that is stuck out. The question is supported by the~ referenced KAs. In' addition KAs 201003K508 (3.1-3.5) and 201003G005 (3.3-3.9) support the question. The additional KAs were added to the reference l~ section. Question 8.07c was not deleted from the examination. 8.09: Disagree with comment. This question is in accordance with section ES-402 of the Examiner Standards and 10 CFR 55.43. 10 CFR 55.43 does not require that all knowledge requirements be based on facility learning objectives. This question does not require memorization of the procedure. It tests for an understanding of what is required when implementing a work order. KA 294001A103 is applicable to-this question because the ability to use procedures related to station activities includes a knowledge of how to perform. routine administrative tasks. ACP-QA-3.02, step 6.6.1.4 states that "proceduresLfor_ actions that are routine and commonly performed are not required to be located at the performance I site.".For a routine procedure,'such as implementing a work order, an SRO would not be expected to refer to the' administrative procedure even though:it'is readily available. KA 294001K102 was deleted from the reference section. L No alternate answers were accepted because the facility reference, Revision 19 of ACP-QA-2.02C was not provided with the facility comments. The word " major" was deleted from the question prior to uploading to the EQB. Part 4 of the answer key was corrected to " assign fire watches if required" in place of " radiation controls". Question 8.09 was not deleted from the examination. a

l l 24 8.13: Partially agree with comment. The facility comment that this question was ambiguous and confusing is.not supported by candidate questions raised during the examination. Two candidates asked if a short answer response was sufficient or if a paragraph explanation was required. They were told that a short answer was sufficient. The term "NRC deportability time requirement" was developed from the document title " DEPORTABILITY". No questions were asked by the candidates concerning the meaning of this term. The PIR j category, the NRC (10 CFR) reporting criteria, or the reporting time requirement was accepted. The events of parts a., b. and d. fall under two classifications j according to EPIP 4701-4 and 10 CFR 50.72 and 50.73. Both j classifications are required for full credit. The answer key was modified as follows: a. A.1, Immediate or 1 hour report or 10 CFR 50.72(b)(1)(iii) [0.25] AND 30 day LER or 10 CFR 50.73a(2)(iii) [0.25] ) b. A.1, Immediate or 4 hour report or 10 CFR 50.72(b)(2)(ii) [0.25] AND 30 day LER or 10 CFR 50.73(a)(2)iv) [0.25] c. A.1, Immediate or 4 hour report or 10 CFR 50.72(b)(2)(iii) [0.50] d. 4 hour report or 10 CFR 50.72(b)(2)(iii) [0.25] l AND 30 Day LER or 10 CFR 50.73(a)(2)(v) [0.25] 1

'l e 0 9 ATTACHMENT 5 -l SIMULATION FACILITY FIDELITY REPORT l Facility Licensee: Northeast Nuc! ear Energy Company Facility Licensee Docket No.: 50-245 Facility Licensee No. : OPR-21 i Operating Tests administered at: Millstone Unit 1 1 Operating Tests diven On: November 8, 9 and 10, 1988 During the conduct of the simulator portion of the operating tests administered November 8, 9 and 10, 1988 the following apparent performance j l and/or human factors discrepancies were observed: 1 There were no apparent performance and/or human factors discrepanc'tes observed during the examinations, however, a question exists as to whether the model for reactor power as a function of water level appropriately accounts for the difference between actual and indicated water level. Wida range level instrumentation which is used for indication when performing l level / power control is calibrated cold and reads approximately five feet lower thar, actual water level at high pressures. This is a concern because actual reactor water level is not being lowered as far as tne instrumentation indicates affecting simulator fidelity with respect to reactor power. i e 1 ___._.________._______m__}}

ML20246P405

Text

Dear Mr. Gallo:

Reference:

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