IR 05000482/1997301

Insp Rept 50-482/97-301 on 970825-29.No Violations Noted. Major Areas Inspected:Operations

ML20151M152
Person / Time
Site:	Wolf Creek
Issue date:	09/17/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20151M149	List: IR 05000482/1997301 ML20151M144
References
50-482-97-301, NUDOCS 9709230059
Download: ML20151M152 (22)

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50 482-License No.: NPF-42 Report No.: .50-482/97-301 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane, NE l

Burlington, Kansas  !

Dates: August 25-29,1997 J

Inspectors: H. Bundy, Chief Examiner, Operations Branch T. McKernon, Examiner, Operations Branch  ;

T. Meadows, Examiner, Operations Branch M. Murphy, Examiner, Operations Branch L. Vick, Examiner, Operator Licensing, Office of Nuclear Reactor Regulation ( Approved By: J. L. Pellet, Chief, Operations Branch l Division of Reactor Safety 1

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ATTACHMENTS:

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Attachment 1: Supplemental Information Attachment 2: Simulation Facility Report Attachment 3: Facility Initial License Written Examination Comments Attachment 4: Final Written Examination and Answer Key I

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-2 EXECUTIVE SUMMARY Wolf Creek Generating Station NRC Inspection Report 50 482/97-301 a

NRC examiners evaluated the competency of five reactor operator and three senior reactor  !

operator applicants for issuance of operating licenses at the Wolf Creek Station facilit The licensee developed the initial license examinations using NUREG-1021, " Operator Licensing Standards for Power Reactors," Interim Revision 8. NRC examiners reviewed, approved, and administered the examinations. The initial written examinations were administered to all eight applicants on August 25,1997, by facility proctors in accordance with instructions provided by the chief examiner. The NRC examiners administered the operating tests on August 26-28, 1997. Three applicants for reactor operator licenses and one applicant for a senior reactor operator license displayed the requisite knowledge and skills to satisfy the requirements of 10 CFR Part 55 and were issued the appropriate license Ooerations

• One of three applicants passed the senior r'eactor operator written examination and three of five applicants passed the reactor operator written examination. Both groups of applicants demonstrated a potential generic knowledge weakness of nuclear instrumentation. The reactor operator applicants demonstrated potential knowledge weakness in the administrative requirements area (Section 04.1).

• All eight applicants passed the operating test. Overall, strong applicant l performance was demonstrated during the dynamic simulator scenarios (Section 04.2).

• The licensee submitted an examination outline which was adequate for examination

development. Several enhancement suggestions provided by the examiners were  ;

l incorporated in the final submittal (Section 05.1.1). l

• The examination materials were acceptable for administration as submitted, but the reactor operator walkthrough tasks were minimally challenging. The licensee made several changes to the reactor operator walkthrough tasks pursuant to Region IV l comments to raise the discriminatory value of the examination to an average level.

! As a result of the post-examination analysis, it was concluded that the written examination contained an excessive number of technical inaccuracies (Section 05.1.2).

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Report Details Summary of Plant Status The plant operated at essentially 100 percent power for the duration of this inspectio ,

I. Operations

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04 Operator Knowledge and Performance 0 Initial Written Examination Inspection Scope On August 25,-1997, the licensee proctored the administration of the written

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examination approved by the chief examiner and NRC Region IV supervision to five

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individuals, who had applied for initialieactor operator licenses, and three individuals, who had applied for initial instant senior reactor operator licenses,. The l licensee graded the written examinations and the staff reviewed the results. The licensee also performed a post-examination question analysis, which was reviewed by the examiners, Observations and Findinos

.The minimum passing score was 80 percent. Scores of applicants for reactor

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operator licenses ranged from 71.7 to 88.9 percent. Three reactor operator license applicants passed with an average score of 88.2 percent. Two reactor operator applicants failed with an average failing score of 74.3 percent. Scores of applicants for senior reactor operator licenses ranged from 75.8 to 90.9 percent. One senior reactor operator applicant passed with a score of 90.9 percent. The two senior reactor operator applicants who failed had an average score of 76.8 percen The above grades reflect the results after examination changes recommended by the licensee as a result of its post examination question analysis were incorporate The Region IV staff reviewed and approved these recommendations based on the technical merits of each recommendation. The licensee initiated a performance improvement request, which will include a root-cause analysis, to investigate "a pass rate that is unacceptable to Wolf Creek Generating Station." It also requested a meeting with Region IV to discuss the results of the performance improvement reques Questions 1 to 75 were the same on both examinations. Questions 76 to 100 were unique to the specific examination. As a result of the post-examination analysis,

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changes were made to the answer keys as follows: For both examinations, Question 6 was deleted, two answers were accepted on Questions 6,9,12,68, nnd 82, and the answer was changed for Question 48. Also, two answers were

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-4 allowed for Questions 88 and 94 on the reactor operator examination and the answer was changed for Question 91 on the senior reactor operator examination, j This resulted in nine changes to the answer key for the reactor operator examination and eight changes to the answer key for the senior reactor operator examinatio Further, the Chief Examiner identified a potential weakness of reactor operator applicants in the administrative area in that Questions 92, 94, 95, and 97, dealing with administrative requirements, were missed by more than half the applicant Also, a potential applicant knowledge weakness relating to the characteristics of nuclear instrumentation was identified in that more than half of all applicants missed Question 41, which related to intermediate range nuclear instrumentation, and half the applicants missed Question 42, which related to source range nuclear instrumentatio l Conclusions One of ihree applicants passed the senior reactor operator written examination and three of five applicants passed the reactor operator written examination. Both groups of applicants demonstrated a potential generic knowledge weakness of nuclear instrumentation. The reactor operator applicants demonstrated potential knowledge weakness in the administrative requirements are j 0 initial Operatina Test Inspection Scope The examination team administered the various portions of the operating examination to the eight applicants on August 26-28, 1997. Each applicant participated in two or three dynamic simulator scenarios. Each also received a walkthrough test which consisted of ten system tasks together with two followup questions for each system. Four of five subjects in four administrative areas were covered by administrative tasks. The remaining administrative subject was covered by two questions, Observations and Findinas All applicants passed all portions of the operating test. With minor exceptions, applicant performance in the dynamic simulator scenarios was stron Communications were good and directions provided to the crews by the senior reactor operator applicants were timely and appropriate. Performance of all applicants in the administrative area was good. Some applicants demonstrated performance problems on specific system tasks during the walkthrough, but overall performance was satisf actory. No broad performance weaknesses were identified.

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- Cnnclusions All eight applicants passed the operating tes. Overall, strong applicant I performance was demonstrated during the dynamic simulator scenario Operator Training and Qualification 0 Initial Licensinn Examination Development The licensee developed the initial licensing examination in accordance with guidance provided in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8, 05. Examination Outline Insoection Scope The licensee submitted the initial examination outline on June 25,1997. The -

examiners reviewed the submittal against the requirements of NUREG .102 Observations and Findinas The initial examination outline was adequate as a gu'de i for development of the examination. However, the reviewing examiner provided several enhancement

. suggestions to achieve the expected level of difficulty for the examinations and provide additional opportunities to evaluate the competency of the applicants. As submitted, the dynamic scenarios provided only the minimum quantitative criteria for evaluating the response of individual applicants to component and instrument f ailures. This could necessitate running additional scenarios because the expected applicant does not always respond to a given event. The examiner also observed that the scenarios were mostly a collection of unrelated events, with little impact on the major transient mitigation strategy and that there was' minimal challenge to the senior reactor operator applicants in prioritizing actions and making decision Specific comments were provided on each scenario for licensee consideration. The written and walkthrough examinations were considered satisfactory with minor  !

enhancement suggestions offered by the examiner. The licensee discussed J

, proposed responses to' the examiner comments with the chief examiner on July 8, 1997, and submitted a revised outline on July 11,1997. The revised outline l

l satisf actorily resolved examiner comments. Extra events with clearer linkage to the major transients had been added to make mitigation strategy more challenging. The outline was condensed in the examination package received on July 25,1997, to l< - reflect fewer examination materials because of the withdrawal of two potential

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- Conclusions The hcensee submitted an examination outline which was adequate for examination development. Several enhancement suggestions provided by the examiners were incorporated in the final submitta . Examination Package Insnection Scope The licensee submitted the completed examination package on July 25,1997. The chief examiner reviewed the submittal against the requirements of NUREG-102 l Observations and Findinas The licensee submitted 125 draft written examination questions of which 75 were designated to be common to both the reactor operator and senior reactor operator examinations. Of the 125 questions, 25 were unique to the reactor operator examination and 25 were unique to the senior reactor examination. The chief l examiner provided comments or questions on a substantial number of questions. In resolving these comments and questions, the licensee modified or replaced live ;

questions, which were common to both examinations, and six questions, which l appeared only on the senior reactor operator examination. Many of these changes were minor editorial or enhancement changes. The chief examiner concurred with the resolution of the comments and the final product. As discussed in Section 04.1, deletion of one question common to both examinations and answer modifications for five questions common to both examinations, three questions appearing only on the reactor operator written examination, and two questions appearing only on the senior reactor operator examination were required as a result of post-examination reviews to make the examinations technically accurate. The quantity of changes is considered excessive in accordance with the 5 percent threshold guidance provided in NUREG-1021, ES-50 The licensee submitted seven dynamic scenarios, including one backup scenario, which was not used during the examination. Also, because of a reduction in the number of applicants, one of the prirnary scenarios was not administered to the applicants. The submitted scenarios were adequate for administration. The licensee subsequently incorporated several enhancement suggestions provided by

the NRC examiners as a result of a table top review and onsite evaluation. Further scenario changes having minimal effect on examination administration were made as follows

in Scenarios Two and Two A Pressurizer Level 460 channel f ailure was

! substituted for Volume Control Tank Level Channel BG LT-149 f ailure because of a

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l problem with the plant procedures for responding to the volume control tank level l channel f ailure, which was discovered during simulator setup; in Scenarios Three l and Three A, Event 3, the f ailing steam generator level channel was changed from 517 to 551 to implement a correction that was identified during preparation week, but inadvertently not entered; and in Scenario 3, Event 3, Nuclear instrument Power Range Channel 44 was inadvertently f ailed instead of 42.

( To support the systems walkthrough section of the operating test, the licensee provided two sets of job perforrnance measures developed to evaluate selected operator tasks. One set was designed for reactor operator applicants and the other set was designed for senior reactor operator applicants. Both sets contained well-

! written task elements, performance standards, and comprehensive evaluator cue The tasks designed for the senior reactor operator applicants were high quality and l acceptable for administration as submitted. However, based on an in-office review, the tasks designed for the reactor operator applicants discriminated at too low a level in contrast to the other set reviewed. Subsequently, as a result of onsite validation, the chief examiner concluded that the reactor operator tasks were marginally adequat or example, although Job Performance Measure 1 contained only one critical step, the task was of high importance and f amiliarity with vital electrical bus l instrumentation and controls together with proper sequencing of preliminary steps was required to successfully complete the task. In response to the comment j concerning task difficulty, the licensee increased task complexity by requiring the l applicant to reenergize a dead bus rather than switching power supplies on a live l bus. Another example involved Job Performance Measure 3, which was essentially i a two critical step task: recognize which valves did not automatically close as l expected on a Phase A isolation signal and manually close the valves. This task l l was of high importance and a significant number of manipulations on several control l l panels were required to successfully complete the task. The licensee enhanced the l difficulty of this task by inserting a valve position indication failure which required

the applicant to use alternate indications to confirm task completion. Also, the licensee upgraded another reactor operator task and replaced one task pursuant to the chief examiner's comments. A second task on the reactor operator walkthrough section was replaced when it was potentially compromised as a result of an j inadvertent encounter, at the task location during NRC validation, with individuals

! not on the security agreement, who were involved in an operator requalification I examination. This was considered to be a scheduling problem on the part of the licensee.

l The final reactor operator walkthrough tasks were considered to be of high quality

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and of average difficulty. Two followup questions associated with each

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walkthrough task were also submitted by the licensee. While the questions were considered acceptable for administration as submitted, the chief examiner provided the licensee with a few enhancement suggestions, which were incorporated by the hcensee. During administration of the test it was discovered that Guestion 14, i

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, I associated with Reactor Operator Simulator Job Performance Measure Three was technically flawed and the wrong answer was given in the answer key. The chief

examiner replaced it with another question for four of the five applicants. The one applicant who correctly responded to the onginal question was given credit and did not have to respond to the replacement questio The licensee submitted two sets of job performance measures and questions to cover the administrative section of the examination. One set was designed for reactor operator applicants and the other set was designed for senior reactor operator applicants. The job performance measures and questions were of high ,

quality with one exception. The submitted job performance measure for the reactor f operator applicant on the emergency plan topic was non discriminatory. The licensee replaced this job performance measure with two discriminatory question The final product was of high qualit c Conclusions The examination materials were acceptable for administration as submitted, but the reactor operator walkthrough tasks were minimally challenging. The licensee made several changes to the reactor operator walkthrough tasks pursuant to Region IV comments to raise the discriminatory value of the examination to an average leve l As a result of the post-examination analysis, it was concluded that the written examination contained an excessive number of technical inaccuracie .2 Simulation Facility Performance Insoection Scope The examiners observed simulator performance with regard to fidelity during the examination validation and administratio Observations and Findinas The simulation facility supported examination administration well. Four minor deficiencies affecting examination administration are discussed in Attachment Three deficiencies, which did not affect examination administration were identified dunng preparation week and are also discussed in Attachment 2. They had a minor l effect on examination validation, but did not require any changes to the scenanos af fected. Other than the issue involving the lamps during loss-of-power conditions, i

none of these deficiencies were repeated during the examination process. The

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licensee made interim simulator adjustments to achieve proper readings for reactor I coolant pump seallbearing water temperatures and allow loading of the emergency

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esel generator. The licensee initiated simulator modification requests to address

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e Conclusions The simulation f acility supported the examination administration wel V. Management Meetings X1 Exit Meeting Summary The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on August 29,1997. The licensee acknowledged the findings presented. Although individual examination scores were not discussed, the licensee indicated that it was not satisfied with the preliminary

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results from the written examination and had formally requested a meeting with l

Region IV to discuss the root causes of the poor results, any operational l implications, and planned corrective action On September 4,1997, Mr. McKernon discussed the excessive number of technical

accuracy issues with the written examination discovered as a result of the post-examination analysis, as well as, potential agenda items for the licensee requested

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meeting with Messrs. Pippin and Guye The licensee did not identify as proprietary any information or materials examined l

during the inspection.

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ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED -

Licensee

. T. Damashek, Supervisor, Licensing ,

R. Flannigan, Manager, Nuclear Engineering, Safety, and Licensing i R. Guyer, Superintendent Operations Training S. Hatch, instructor, Training O. Maynard, President and Chief Executive Officer B.- McKinney, Plant Manager A' Palmer, Lead Senior Instructor, initial Licensed Operator Training

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'J. Pippin, Manager, Training G. Smith, Senior Instructor, Training e C. Warren, Vice President Operations, Chief Operating Officer C. Younie, Manager, Operations

'NRC .

F. Ringwald, Senior Resident inspector

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ATTACHMENT 2 SIMULATION FACILITY REPORT Facihty Licensee: Wolf Creek Nuclear Operating Corporation Facihty Docket: 50 482 Operating Examinations Administered at: Wolf Creek Generating Station Operating Examinations Administered on: April 15-16,1997 These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility, other than to provide information, which may be used in future evaluations. No licensee action is required in response to these observations. Only the first observation was repeatable during the examination Deficiencies identified Durinn Examination Preparation

• The push to test feature lighted lamps on pump and breaker control panels under loss-of-power conditions that should have precluded this feature.

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• A power decrease was in progress and the drain valves on the extraction steam

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lines started cycling for no apparent reaso * Af ter a Charging Pump B trip with all other charging secured, tne following anomalies were observed. Although the seal flow meters on Panel RLOO1 went to zero, other indicators showed approximately 7 gpm per pump flow to the reactor coolant pumps and Charging Pump A boron injection flow increased to 600 gp Also, the seal / bearing water temperature meter was pegged low and the NPIS point indicated 3.5 degrees F while the reactor coolant temperature was 535 degrees Deficiencies identified Durina Examination Administration

• While performing a task involving starting and loading the emergency diesel generator, it was noted that the synchroscope check light did not illuminate when Bus NE02 voltage was adjusted 50-100 volts higher than Bus NB02 voltage in accordance with procedure. The instructor adjusted NE02 volts until the light was illuminated and then set Meter NB El-29 to read 50-100 volts higher than Bus NB02

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voltage. No further problems were noted during subsequent performances of this task.

• While performing Procedure STS SF-001 during a dynamic scenario, when the i

operator returned the rod control bank selector switch to AUTO it appeared that the

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switch did not select AUTO. Rod speed indicated 48 steps per minute instead of The apphcant cycled the switch to manual and then back to AUTO to achieve the desired results.

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• During a dynamic scenario after a trip of Charging Pump B followed by the operator starting the normal charging pump and adjusting flow to restore pressurizar level, the charging flow indication went to zer ;

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ATTACHMENT 3 FACILITY INITIAL LICENSE WRITTEN EXAMINATION COMMENTS

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INITI Al.1.lCENSEI) WRITTEN EXA.Tl CO.\D1ENTS Question 6: Question states immediate Actions of EMG E-0 are being performed - implies very early atter a Reactor Irip. Tavg has dropped 10'F below no load Tavg, and steam pressure is -100 lbs. low, it is not possible to differentiate between a steam line break and feed line break without knowing the trends for S/G level and pressure -- since ' A' S/G level is below the feed ring - break is blowing primarily stea Even on a steam line break the affected S/G level will decrease more rapidly due to flow restriction in the cross-over pipe. Since FWlV's would be closed, break flow would no longer be present if key answer (C)

is correct, but without trends it cannot be determined if break flow is still presen . Insuf6cient information is given to determine the correct respons * Recommend deleting this item from the examinatio Question 9: (Reference BD-EMG C-0) Basis for depressurizing the RCS 260 lbs. in EMG C- Answer (C), key answer, is to reduce leakage through RCP seals. This data appears in BD-EMG C-0 on page 61 under the BASIS discussion for step 2 STATEMENT FROM BACKGROUND: "This step depressurizes intact S/Gs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss."

In the third paragraph of the BASIS on page 61, the following also appears:

STATEMENT FROM BACKGROUND "It is important that the depressurization not reduce S/G pressures in an uncontrolled manner that under shoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS."

Answer (A) is also correct in that the basis for the 260 psig limit is to prevent accumulator nitrogen from injecting into the RC . Recommend accepting Answers A and C, Question 12: (Reference LO N 086 00) Key answer is 'B'. For this answer to be correct, the diesel fire pump would have to start on low pressure along with the electric driven fire pump. After 30 minutes, the electric fire pump would shut off, and the diesel fire pump would supply sufficient flo The electric fire pump starts at a higher pressure than the diesel fire pump and is capable of supplying 3300 gpm (only 2000 is being used). Depending on how fast pressure dropped, the diesel fire pump may or may not have started - if the pressure drop were slow enough, the electric driven pump would start and supply sufficient flow to prevent the diesel pump from starting, if the pressure drop is rapid, the pressure may decrease below the start pressure for the diesel pump before the electric pump comes up to speed. After 30 minutes, the electric pump would shut off, then re-start on low pressure (if the diesel pump is not running)

again before the diesel fire pump would receive a start signa * Recommend accepting Answers A and B Question 16: This question is correct as written. Sufficient information is presented to allow the student to determine the correct response Traming occurred under TIN LO 17 32317 on the effects of natural circulatto * Recommend acceptmg only key Answer D.

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Question 21: This question is correct as written. SufGcient infonnation is presented to allow the student

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to determine the correct response. Training occurred under TIN LO !3 002 0 . itecommend accepting only key Answer . Question 22: his question is correct as written. SufGcient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 17 323 21.

j =: Recommend accepting only key Answer l

! Question 41: His question is correct as written. Sufficient information is presented to. allow the student to L detennine the correct response. Training occuned under TIN LO 13 015 0 e' Recommend accepting only key Answer A.

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Question 42: This question is correct as written. Sufricient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 13 015 0 * Recommend accepting only key Answer D.

i-I Question 48: (Reference LO 13 003 00. Reactor Coolant Pumps) #3 sealis a split seal. 400 cc goes to

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'the Containment Sump (from the upper seal) and 400 cc mixes with #2 seal return and goes to the Reactor Coolant Drain Tank.

f i e Key Answer 'D' - only #2 seal leakoff goes to Reactor Coolant Drain Tank.

! * More Correct Answer: 'B' + leakoff from #2 and #3 seals drain the Reactor Coolant Drain Tank.-

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l' * Recommend changing key to accept only Answer B.

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Question 58: His question is correct as written. Sufficient information is presented to allow the student to l

E determine the correct response. Training occurred under TIN LO 13 026 0 . . Recommend accepting only key Answer A.

l2 Question 59: his question is correct as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 13 028 0 * Recommend accepting only key Answer Question 63: his question is correct as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 15 062 0 e Recommend accepting only key Answer B.

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j Question 68: inero i t/< i ! 16 "r t il'N li/b ile i N10 b-i l and i WN BB-Oli both hase ltetuchne Water ' mace rank swarmer reset prior to ahenmg a tram for shutdown coohng. Only in GEN 00-006 could you end up wah one tram in S D coohng mode and one tram injectmg. OFN BB-031 would has e you mitially secure the S D cooling tram md isolate it from the RCS. then realign later if S/D cooling is required.. Init:al sonditions do not tell what mode the LOCA occurred in. Key answer 'C' would be correct for a Mode 4 LOCA during the sery early stages of OFN BB-031. Answer 'B' is correct for EMG liS-1I or OFN BBdI lineu * Recommend a cepting Answers B and Question 82: (Reference P/R 94-1248 // 7-12-941. securing a condensate pump at power is a non-significant event witin little consequences. WCGS had a condensate pump secured at full power on Sep .1994 (after rerate). The plant continued to operate and S/G levels recovered (the pump was off for a total of 28 minutes). TIN LO 15 059 00 does state that it is doubtful that adequate flow can be maintaine l i

However, actual plant experience is that it can be maintaine ,

e Recommend accepting Answers A and Question 86: This question is correct as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 14 078 0 * Recommend accepting only key Answer Question 88: Key answer 'C' is one possible answer, if operating " valves wide open" { control valves fully open) or on the limiter such that control valves will not open further to maintain Mwe, and an MSR relief valve opened. Answer D is also correct if not operating " valves wide open", control valves will open to try to maintain Mwe. Decreased efficiency due to lower vacuum will cause Mwe to decrease with " essentially constant" steam flow and reactor power".

• Recommend accepting Answers C and j Question 89: This question is correct as w ritten. Sufficient information is presented to allow the student to determine the correct response. Training occetTed under TIN LO 14 089 0 * Recommend accepting only key Answer Question 92: his question is corTect as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 17 332 0 * Recommend accepting only key Answer Question 94: (Reference AP llE-001)

Key answer 'B' is correct. AP 21E-001 allows independent verification to be done by remote indication prior to deenergizing a valse. Answer 'D' is also correct if a blocking device is used, which is allowed by AP 21E-001 (also on page 14). NOTE. Although the procedure does not specifically require the valve to ne de-energized prior to verifying " blocking device." this is a standard operating practic * Recommend accepting Answers B and .- . - _ _ - - __ .-- - -- -- .- . -

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e l{ccommend acceptmg only key Answer Question 9': fhis question is correct as written Sutlicient inf ormation is presented to allow the student to determine the wtrect response frammg occurred under flN 1.0 17 332 01 e llecommend accepting only key Answer 1 Question 107: (Rc/erence. EMG E.0 fo/dout page / This question asks if Adverse Containment values shoald be used. !!MG foidout pages list 2 criteria to determine if Adserse Containment values should be use ) Is containment pressure greater than 5 psig. If pressure has been abos e 5 psig, adverse containment values need not be used once pressure decreases below 5 psi ) is containment radiation greater than 10' R/hr, or integrated dose has exceeded 10' Rad. If radiation levels in containment have es ' i the 10' R/hr level, adverse containment values must be used until an engineering analysis is perfc the TSC to verify integrated exposure has remained acceptabl Answer C is correct since the r.. gated dose level has been exceede Answer D is correct since the m3tantaneous dose rate level has 5een exceeded and no engineering analysis has beer performe e llecommend accepting Answers C and Question 114: This question is correct as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 13 001 0 *

• llecommend accepting only key Answer ,

i Question 116: (Reference M 211-001. Temporary Modification (Rev 2.) The Temporary Modification 1 procedure does not exclude non-safety related equipment. It allows this type of replacement on non-safety related equipment only, but contains no relaxation from the requirement to issue a Temp. Mo * Key answer is not correct 'D' in that a Temp. Mod should be issue . Answer 'B' is correc e llecommend changing the key and accepting only answer Question 119: This question is correct as written. Sufficient information is presented to allow the student to detennine the correct response. Training occurred under TIN LO 13 032 0 e itecommend accepting only key Answer A.

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-M explanation summary M g m ~ 7._ 'E 6 answer Question see attached notes 4 number A A A A A A n 'g

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