ML20199D331

Affidavit of H Peterson in Support of NRC Staff Response to Sp O'Hern Written Presentation.* Affidavit of H Peterson in Support of NRC Staff Proposed Denial of Sp O'Hern Application for Reactor Operator License

ML20199D331
Person / Time
Site:	05532442
Issue date:	01/19/1999
From:	Hironori Peterson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199D323	List: ML20199D318 ML20199D331 ML20199D350
References
99-753-01-SP, 99-753-1-SP, SP, NUDOCS 9901200071
Download: ML20199D331 (52)

Text

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January 19,1999  !

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l UNITED STATES OF AMERICA ,

NUCLEAR REGULATORY COMMISSION ,

l BEFORE THE PRESIDING OFFICER l Before Administrative Judge:

Peter B. Bloch (Dr. Richard F. Cole, Special Assistant)

In the Matter of )

) Docket No. 55-32442-SP SHAUN P. O'HERN ) 1

) ASLBP No. 99-753-01-SP ,

(Denial of Application )

for Reactor Operator License) )

AFFIDAVIT OF HIRONORI PETERSON i IN SUPPORT OF THE NRC STAFF'S RESPONSE l TO SHAUN P. O'HERN'S WRITTEN PRESENTATION l I, Hironori Peterson, having first been duly sworn, do hereby state as follows:

1. My name is Hironori Peterson. I am employed as a Reactor Engineer in i the Operator Licensing Branch, Division of Reactor Safety, NRC Region III, in Lisle, l

Illinois. I reviewed and approved the written examination which was administered by the facility licensee to Mr. O'Hern at the Enrico Fermi Nuclear Station, Unit 2, on April 6, 1998. I was also the NRC chief examiner in charge of the licensing examinations at Fermi. A statement of my professional qualifications is attached hereto.

2. This Affidavit is submitted by the NRC staff (Staff) in response to the l

written presentation dated December 7,1998, submitted by Shaun P. O'Hern j i

l (Presentation), in support of his request for a hearing on the NRC Staff's proposed denial

-9901200071 PDR 990119 l MISC 9901200068 PDR [

Exhibit 1

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j 9 s of his application for a Reactor Operator (RO) license for use at the Enrico Fermi Nuclear Station, Unit 2 (Fermi), operated by the Detroit Edison Company.

j 3. The Presentation includes the sworn written arguments of Mr. O'Hern and eight supporting documents which were not already contained in the Hearing File. ,

4. On November 20, 1998, the Staff transmitted the Hearing File to the Presiding Officer and Mr. O'Hern, along with a numbered index thereto. Items contained l

in the Hearing File are herein referred to by their designated " Item" number, as set forth l in the Hearing File index.

5. Mr. O'Hern does not currently hold an NRC operating license and his i

position at Fermi at the time of license application was as a radioactive waste supervisor.

6. On March 17, 1998, Mr. O'Hern and his employer, Detroit Edison Company, submitted a " Personal Qualification Statement - Licensee" (NRC Form 398) l requesting that Mr. O'Hern be administered an examination for an RO license. (Hearing File Item 1). The NRC Region III Office, Lisle, Illinois, received the application on March 18,1998. On March 23, 1998, Mr. Michael E. Bielby, Reactor Engineer, l'

Region III, reviewed and approved Mr. O'Hern's application as meeting NRC experience and education requirements to be administered an RO licensing examination.

7. I began review of the draft examination prepared by the Fermi staffin the l Region III office on March 6,1998, in accordance with Examination Standard l'

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l " Reactor Operators" are referred to as " Nuclear Supervising Operators" (NSO) at the Fermi Nuclear Station.

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l l (ES)-401, " Preparing Initial Site-Specific Written Examinations." (Hearing File Item 14).

On March 23,1998, two other NRC examiners and I traveled to the Fermi site to continue l the review and validate the draft written examination and the operating test items.2 The written examination review was continued at the Fermi site with the licensee's training department and operations department personnel. During this weck on site, the examiners t

met with the licensee's examination development group and performed the review and l validation of all the examination material. The review of the written examination resulted in several changes and enhancements, which included some new questions proposed by the licensee's development group for NRC approval. The proposed changes to the written examination were mutually agreed upon by both the facility licensee and myself. The changes were made by the facility licensee with additional review and verification performed by facility licensee personnel. The final written and operating examinations, incorporating the changes and enhancements, were submitted by the facility licensee on March 30,1998.

8. Having determined that Mr. O'Hern met the eligibility requirements to take the examination, NRC Region III administered operator licensing examinations to

' Mr. O'Hern and eleven other license applicants at the Fermi facility during the weeks of April 6,1998, and April 13, 1998. On April 6,1998, the written examination was 2

The examination, including the written examination, was prepared by Detroit Edison l

pursuant to a pilot program in which the NRC is evaluating the feasibility of revising 10 C.F.R. Part 55 to require facility licensees to write the operator licensing :xaminations.

Detroit Edison prepared the written examination to the same specifications that the NRC Staff would have applied. The Staff reviewed the written examination test in detail, made revisions as necessary, and approved the final product before it was administered.

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j administered to all twelve candidates and proctored by Detroit Edison training staff. Prior to the written examiration, the NRC examiners, along with facility licensee personnel, read i

l NUREG 1021, Appendix E, " Policies and Guidelines for Taking NRC Examinations,"

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! verbatim, to all of the candidates, including Part B, Paragraph 7, which advises the candidates that "[i]f the intent of a question is unclear, ask questions of the NRC examiner l

or the designated facility instructor only." (Exhibit 5, page 2). All questions asked by the candidates during the exam were documented. (Exhibit 4). No questions regarding I

examinations questions 7,54,59 or 87 were recorded. (Exhibit 4).

9. Mr. O'Hern's licensing examination consisted of a 100-question written i examination and an operating test, which included an individual plant walk-through test i

. (Categories A and B) and a crew-based, integrated plant operations performance test on l

I a dynamic simulator (Category C). The passing grade for the written examination is l

80.0%. (Hearing File Item 17, page 1). Applicants are required to pass both the written examination and the operating test, including Categories A, B and C, in order to receive a license. The licensing examination was prepared in accordance with the instructions of NUREG-1021.

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10. The Staff's expectations regarding RO written examination knowledge are set forth in the following paragraphs.

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11. 10 C.F.R. i 55.41(a) states:

The written examination for an operator will contain a representative selection of questions on the knowledge, skills, and abilities needed to perform licensed operator duties. The knowledge, skills, and abilities will be identified, in part, from learning objectives derived from a systematic analysis oflicensed operator duties performed by each facility licensee and contained in its training program and from information in the Final Safety -

Analysis Report, system description manuals and operating procedures, facility license and license amendments, Licensee Event Reports, and other materials requested from the facility licensee by the Commission.

12. Detroit Edison has implemented a Systems Approach to Training for the

Fermi facility for the licensed operators. This was certified to the NRC in a letter dated May 27,1987. (Exhibit 6).10 C.F.R. f 55.4 defines a Systems Approach to Training as a training program that includes the following five elements

(1) Systematic analysis of the jobs to be performed, (2) Learning objectives derived from the analysis which describe desired performance after training, (3) Training design and implementation based on the learning objectives, (4) Evaluation of trainee mastery of the objectives during training, and (5) Evaluation and revision of the training based on the performance of trained personnel i'1 the job setting.

13. 10 C.F.R. i 55.41 (b) states the written examination for an operator will include a representative sample from a list of fourteen specified items.

14. Pursuant to 10 C.F.R. i 55.41, the Staff has established detailed criteria for the design of the written examination in ES-401, " Preparing Initial Site-Specific Written Examinations," of NUREG-1021. (Hearing File Item 14). ES-401 requires the written examination to contain a representative selection of knowledge, skills, and abilities needed to perform duties at the desired license level.

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15. The Staff expects the RO's to possess adequate knowledge and to be competent in each of the areas. The NRC operator licensing examination process seeks to determine an applicant's level of knowledge, skills and abilities, and then to evaluate whether this level meets that needed for a minimally safe and competent operator. As

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such, the examination is designed to have one correct answer for each question. Responses l

l that display a lack of knowledge, understanding, or familiarity with the subject matter will

be graded as unsatisfactory.

16. The Staff's expectations for following and complying with the facility's procedures are specifically stated in the licenses it issues to ROs pursuant to 10 C.F.R.

l f 55.51. A condition of every operator license issued by the NRC requires the holder to observe the operating procedures and other conditions specified in the facility license authorizing operation of the facility.

17. Detroit Edison provided post-examination comments on ten written examination questions. I reviewed the post-examination comments and determined five of l the comments had merit. I accepted those five comments, in whole or in part, but they did  ;

i i not affect the questions being addressed in this proceeding. Of the remaining five j l

I conunents, one relates to question 87, which is in issue in this proceeding. (Hearing File Item 8).

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18. I graded Mr. O'Hern's written examination answers in accordance with L ES-403 , " Grading Initial Site-Specific Written Examinations," of NUREG-1021. (Hearing File Item 16). The initial grading of Mr. O'Hern's examination showed failure to

. answer 24 of these items correctly. He received a grade of 76% (76 correct out of 100

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total), which was below the required 80.0%. He therefore failed to achieve a satisfactory l

grade for the written examination. This resulted in an overall grade of unsatisfactory for Mr. O'Hern's licensing examination. (Hearing File Item 3). Accordingly, on May 12, l

l 1998, NRC Region III license examiners signed Form ES-303-1 and recommended that l

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! Mr. O'Hern's application for an RO license should be denied, notwithstanding the fact that he had successfully passed the operating test. Melvin N. leach, Region III Chief of the Operator Licensing Branch (OLB), independently reviewed and concurred with this recommendation on May 18,1998.

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19. In a letter dated May 20,1998, Mr. leach informed Mr. O'IIern that the i grading of his written examination indicated that he had failed the written examination and l

the Staff proposed to deny his application for an RO license. (Hearing File Item 4).

Mr. O'Hern was advised that he could request an informal NRC staff review or a hearing within 20 days. If he requested an informal review, he was to indicate which answers he believed were incorrectly graded and provide the basis with supporting documentation for l

his contentions. Upon receipt of that request and supporting information, the Staff would i review his contentions, reconsider its grading and inform him of the results. If he still i failed the examination, Mr. O'Hern could then request a hearing pursuant to 10 C.F.R.

f 2.103(b)(2).

20. On May 29,1998, Mr. O'Hern responded to the NRC Staff's letter of May 20,1998, and requested an informal review of the grading of 10 questions on his l

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written examination' (Hearing File Item 5). The Staff acknowledged Mr. O'Hern's request for informal review in a letter dated June 9,1998. (Hearing File Item 6).

Regarding the questions whic.: are the subject of this hearing, Mr. O'Hern's May 29,1998 submittal contended that each of the four questions had an additional correct answer The

! specific contentions are addressed in this affidavit and in Hearing File Items 5, 7, 9 and 10. j

21. An informal review of Mr. O'Hern's contentions was undertaken by the l Staff in NRC Region III in accordance with the procedures found in ES-502, " Processing l

Requests for Administrative Reviews and Hearings-After Initial License Denial," of i

NUREG-1021. (Hearing File Item 18). The Region III Staff considered the information I supplied by Mr. O'Hern during the informal review phase of his appeal and determined no change to the original grading for any of the contended questions was warranted.

Consequently, on June 16,1998, the Region III Staff recommended continued denial of  ;

Mr. O'Hern's application for an RO license to the Chief of the Operator Licensing and i l

Human Performance Branch (HOHB), Division of Reactor Controls and Human Factors (DRCH), Office of Nuclear Reactor Regulation (NRR). (Hearing File Item 7).

I 22. Following Region III's informal review, again in accordance with ES 502.D of NUREG-1021, a three-person appeal board was selected from other regional offices to consider the ten written examination questions originally placed in contention by Mr. O'Hern. (Hearing File Item 18). The appeal board also reviewed written 3

Mr. O'Hern requested review of questions 2,7,17,25,34,38,45,54,59, and 87.

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9 examination appeals from two other individuals at the Fermi facility. The review of those appeals affected questions on Mr. O'Hern's examination other than those he had appealed.

The appeal board provided its recommendations to Robert M. Gallo, Chief, HOHB, in a 4 1

memorandum dated July 31,1998. (Hearing File Item 9; Exhibit 3'). The appeal board recommended deleting questions 17,38, and 71 (incorrectly answered by Mr. O'Hern);

deleting question 56 (correctly answered by Mr. O'Hern); and giving credit to question 87 for two correct answers (incorrectly answered by Mr. O'Hern). Otherwise, the appeal board agreed with Region III's informal review resu!!s for the remainder of the questions challenged by Mr. O'Hern. The appeal board sustained the denial with a final grade of 79.2% (76 correct out of % questions) and recommended in the cover memorandum that Mr. O'Hern be evaluated as failing the written examination.5

23. The HOHB Staff reviewed the appeal board's findings and recommendations and the results of the Region III informal review. The HOHB Staff agreed with the appeal board's recommendations for question numbers 17,38,71, and 56, but agreed with the Region III recommendation for question 87. In addition, the HOHB Staff recommended deleting question 25 (incorrectly answered by Mr. O'Hern). The

• Exhibit 3 is a complete copy of the appeal board's July 31,1998 recommendation.

- Hearing File Item 9 is a redacted version of the same document.

5 It should be noted that there are administrative errors in Attachment 2 to the appeal board's recommendation. (See Exhibit 3, page 12; Munro Affidavit 17). The errors are transposition and administrative errors and do not change the appeal board's conclusion that Mr. O'Hern failed the written examination. (See Munro Affidavit 17 ). In any event, the appeal board's conclusions are recommendations. The Director of DRCH nukes the final NRC Staff decision in these matters. (See Munro Affidavit 15; Hearing File Item 18, page 5).

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. HOHB Staff detennined that the applicant's grade was 78.9 % (75 correct out of 95 questions). The HOHB Staff concluded that the applicant's performance on the written  !

examination was unsatisfactory and recommended that the denial of Mr. O'Hern's license l application be sustained. On September 7,1998, the Staff transmitted a letter to Mr.

l O'Hern, in which it informed him that it had reviewed the grading of his written i i

examination in light of tie information he supplied, and that the Staff concluded that he  !

still did not pass the written examination. (Hearing File Item 10). Accordingly, the Staff  ;

letermined that the proposed denial of Mr. O'Hern's RO license application should be  :

sustained, and advised him of his right to request a hearing in connection with the proposed denial. The Staff also advised Mr. O'Hern that, if he accepted the proposed  :

denial, he could reapply for a license two months from the date of the letter and that he could request a waiver of the operating test. f l 24. On September 22, 1998, Mr. O'Hern filed a request for hearing in 1

connection with the proposed denial of his RO license application. (Hearing File Item 11).

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! In that document, Mr. O'Hern stated he felt that not all of the questions had been i adequately addressed during the informal review. In his Presentation, filed December 7, 1998, Mr. O'Hern also states: "I do not believe the NRC fully understood the basis for my appeal of questions 7, 54, and 87. . . . I .will also demonstrate the reasons for my contentions that question 59 should be deleted." He also requested question 54 be deleted l

if his technical arguments were not accepted. The master question, answer, and references for these questions are included in the Hearing File as Item 45.

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h' l' 25. Question 7, the Staff's discussion of the question, Mr. O'Hern's l l

contentions, and the Staff's response are set forth in the following paragraphs.

26 RO Ouestion #7. "From full power operation, a transient has occurred.  !

The following annunciators were received: .

3D73, Trip Actuators A1/A2 Tripped 3D74, Trip Actuators B1/B2 Tripped 3D99, APRM [ Average Power Range Monitor] Upscale Neutron / Thermal Trip  !

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Immediately after receipt of these ~ annunciators, the following parameters were reported to the NASS [ Nuclear Assistant Shift Supervisor]:

l Reactor Power 48% and stable l

RPV [ Reactor Pressure Vessel] Level 164 inches, decreasing slowly Reactor Pressure 1085 psig, increasing slowly I With these plant conditions, what is the first action that must be performed, and which 1

indications must be observed to verify proper response?

(a)- Manually operate SRVs [ Safety Relief Valves] to stabilize pressure at less j than 1050 psig; observe Div 1 and 2 post-accident recorders.

(b) Place the SVLCV Bypass Valve Mode Switch in STARTUP, and verify Reactor Pressure Vessel level is not increasing.

-(c) Initiate Alternate Rod Insertion; perform OD-7 option 2.

i (d). Place the Reactor Mode switch in SHUTDOWN; verify blue group scram lights are Off."

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27. The question asks for the first action that must be performed, and which L indication must be observed to verify proper response, followmg a transient from full i L  !

power operation. The correct answer choice per the answer key is (d), "[p] lace the

- Reactor Mode switch in SHUTDOWN; verify blue group scram lights are Off." For the given conditions, with annunciators 3D73, 3D74, and 3D99 actuated, the reactor should l t have automatically shut down (scrammed), and reactor power should be 0%. The question i

indicated the reactor was still at 48 % power, which indicates the reactor had failed to shut

down or scram, due to a fault condition (s) (electrical and/or mechanical). This is referred to as an Anticipated Transient Without Scram (ATWS). When the reactor mode switch is placed in the SHUTDOWN position, a diverse and redundant reactor scram signal is generated by the Reactor Protection System (RPS)* logic. This action specifically addresses the potential for multiple sensor and sensor relay failures in the RPS logic.

l l (Hearing File Item 38, pages 38,62). This redundant scram signal may be successful in l shutting down the reactor depending upon the fault condition causing the ATWS. It also enables the system logic such that certain recovery actions for the ATWS may be I~

• Student Text ST-OP-315-0027-001, " Reactor Protection System," states: "1. RPS prevents the reactor from operating under unsafe or potentially unsafe conditions by j l automatically initiating rapid insertion of the control rods (reactor scram) when trip i setpoints are exceeded. 2. RPS is a dual-trip system consisting of 2 trip systems L designated A and B. The 2 independent trip systems, A and B, are each made up of 2

! scram trip channels which include both automatic and manual functions. These trip channels are designated A1, A2, B1, and B2. If a trip occurs in any trip channel of Trip l' System A and a trip occurs in any trip channel of Trip System B, a reactor scram will result. However, if trips occur in one or both trip channels of the same trip system, a i reactor scram will not result. This condition is called a ' half-scram'." (Hearing File 4

Item 35, page 8).

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l' i performed. For example, if the ATWS condition was due to hydraulic lock of the control 1

rods it will be necessary to reset the scram signals to allow for repeated scrams to insert the hydraulically stuck control rods. By placing the Reactor Mode switch to the SHUTDOWN position, it bypasses certain scram signals, including the Scram Discharge

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Volume (SDV) high water level scram and the Main Steamline Isolation Valve (MSIV)

I closure scram, which then allows the resetting of the scram. (Hearing File Item 35, j page 21; Hearing File Item 38, page 65). Also, positioning the Reactor Mode switch in the SHUTDOWN position is important because it prevents Main Steam Isolation Valve closure on low main steam line pressure, thus maintaining main condenser availability and minimizing the heat load on the containment. (Hearing File Item 38, page 62).

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The blue group scram lights indicate the status of the RPS logic's two trip systems (A and B) and the attendant scram trip channels (A1, A2, B1, and B2). Normally, when l a reactor scram condition occurs, both RPS trip systems actuate and initiate a reactor ,

j scram. This will cause de-energization of the associated scram pilot valve solenoids (the blue group scram lights go off) and insertion of all control rods, assuming that there are no additional failures. Tripping at least one trip channel (Al or A2 AND B1 or B2) in each RPS trip system will also activate annunciators 3D73 and 3D74, respectively. The question indicated that annunciators 3D73 and 3D74 were in alarm, indicating as a i

minimum that one trip channel in both RPS trip systems had actuated. However, the ,

question also gave indications that the reactor did not scram as the reactor was still at 48 %

power. With reactor power at 48% all control rods have not inserted due to a fault condition (s)in the RPS and/or control rod systems. Therefore, all of the blue group scram i

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o lights may not be off. Multiple sensor, sensor relay, and contact failures in the automatic RPS logic including the RPS scram solenoid circuitry could result in an incomplete actuation of the RPS, with both annunciators 3D73 and 3D74 in alarm. For example, if given a scram condition and trip channels Al and B1 actuate, but trip channels A2 and B2 ,

fail to actuate with scram relays K14E and K14F contacts in the RPS scram solenoid circuitry failing closed, then both annunciators 3D73 and 3D74 will alarm. However, groups two and three blue group scram lights will not be extinguished. This example was determined using RPS Schematic Diagrams from Mr. O'Hern's Presentation Item 2 attachments.

28. In accordance with 10 C.F.R. f 55.41 (b)(10), the NRC is required to evaluate applicants through a written examination that tests the applicants' knowledge, skills, and abilities needed to perform licensed operator duties relating to administrative, normal, abnormal, and emergency operating procedures for the facility. The given '

conditions for question 7 pertain to three procedures which require immediate attention by the operators to recognize the abnormal condition and implement mitigating actions to place and/or ensure the reactor is shutdown. The procedures are Abnormal Operating Procedure (AOP) 20.000.21, " Reactor Scram," (Hearing File Item 23), and Emergency Operating Procedure (EOP) 29.100.001 SH 1, "RPV Control," and SH 1 A, "RPV Control ATWS." (Hearing File Items 33 and 34). The conditions provided in the question stem are consistent with two of the four entry conditions for EOP 29.100.01 SH 1: (1) scram condition AND reactor power CANNOT be determined to be < 3%; and (2) RPV water level < 173 inches. (Hearing File Item 33). With reactor power at 48 percent and stable, u

i there can be no confirmation of a reactor scram and all rods full in as required by steps RC-1 and RC-2 of EOP 29.100.01 SH 1. Subsequent step RC-3 then directs performance of step FSRC-1, of EOP 29.100.01 SH 1A. The first action that must be performed, 1

consistent with the governing procedure EOP 29.100.01 SH 1 A and the available question ,

answer choices, is for the operators to " Confirm7 Rx [ reactor] mode switch in S/D

[ shutdown]" according to Step FSQ-1. Since no operator actions were specified as complete according to the question stem, the operators are required to place the Reactor Mode switch in SHUTDOWN. Thus, the correct answer choice is (d). Furthermore, AOP 20.001.21, " Reactor Scram," also specifies, prior to transition to the governing procedure, the first operator immediate action is to, " Place the Reactor Mode Switch in SHUTDOWN." (Hearing File Item 23). In addition, the facility licensee's Operations Department Instruction (ODI)-022, " Reactivity Management," defines the facility licensee management's ' expectations on conservative actions pertaining to reactivity management / manipulations as:

(1) Fermi 2 Operators will maintain strict control and alertness at all times. Conservative actions are required during any unexpected or unexplained situation with regard to reactivity, criticality, power level, or any other anomalous behaviors of the reactor. Fermi 2 management expects that these conservative actions should include rod insertion to lower power, or a reactor scram without hesitation, whenever such unanticipated or unexplained behavior is encountered; [and (2)] All l

7 The " Introduction to Emergency Operating Procedures," Student Text ST-OP-802-3001-001, provides the following definition and usage for "Confinn" listed as an Emergency Operating Procedure Key Word: " Confirm: Use available indications and, as appropriate, physical observation to establish that the specified action has occurred, l, conditions are as stated, etc. Includes an implied requirement to take corrective action if the identified conditions do not exist." (Hearing File Item 39, page 29).

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licensed operators are responsible for shutting down the reactor when reactor protection setpoints have been exceeded and automatic action has not occurred.

(Exhibit 7, page 1). Therefore, consistent with the three aforementioned procedures and facility licensee management's ODI-022 expectation, the first operator action that must be performed is to place the Reactor Mode switch in SHUTDOWN in an attempt to shutdown the reactor. Student Text ST-OP-315-0027-001, " Reactor Protection System," also  :

validates this interpretation for the question's conditions. It states, in section V.E.2.a, that if a condition exists that requires 'a reactor scram and reactor power cannot be determined to be below 3 percent, then the first step is for the Reactor Operator to initiate a reactor scram by taking the Reactor Mode switch to SHUTDOWN if this has not already been done. (Hearing File Item 35).

29. Mr. O'Hern selected answer choice (a), manually stabilize pressure, asserting that it is an additional correct answer. (Hearing File Item 5). In his request for an informal staff review, he stated that if assigned to one of the RO positions, the control room nuclear supervising operator (CRNSO), he would be expected to be given responsibility for level and pressure control. If assigned to the other RO position, panel 603 operator (P603), the Reactor Mode switch would be placed in SHUTDOWN, correct answer choice (d). (Hearing File Item 5). However, although the informal review request included AOP 20.000.21 and acknowledged the validity of the first immediate operator l action step, " Place the reactor mode switch in SHUTDOWN," in his Presentation, Mr. O'Hern now contends that the only correct answer is answer choice (a) and answer choice (d) is incorrect. The Presentation provides two arguments for the revised position.

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i First, as noted above, it states that answer choice (d) is incorrect since the question l

indicates annunciators 3D73 and 3D74 are actuated and that condition would result in all the blue group scram lights being off and therefore cannot be used to verify proper

( response, i.e., the second part of the answer was incorrect or "As previously stated I cannot verify that an action provided the desired response by observing something that has J l already occurred." The Staff disagrees. Mr. O'Hern assumed that all the blue group l scram lights had already de-energized and therefore concluded that he did not have to verify this action again. The conditions of the question clearly indicates an anomalous condition pertaining to an incomplete reactor scram (ATWS). However, the question i provides no indication that all the blue group scram lights were extinguished. As previously explained in paragraph 27, multiple RPS circuitry failures could result in an  ;

incomplete actuation of RPS with both annunciators 3D73 and 3D74 in alarm, but not all l

L the blue group scram lights extinguished. Even if a complete scram did occur, the proper operator action after performing any procedural step is to verify required actions complete and observe the expected plant response with all and any indications available, including l

verifying previously performed or verified indications. The facility licensee's Student Text ST-OP-315-0027-001 lists diverse methods for verification after the Reactor Mode switch l is placed in SHUTDOWN. (Hearing File Item 35, page 29). Examples include the following: (1) verify that annunciator windows 3D77(78), " Manual Trip Actuator A(B) l System Trip," and 3D1113, " Control Rod Withdrawal Blocked," illuminate; and (2) verify that the Manual Scram pushbutton back lights come on and the white and blue Group 1, 2, 3, and 4 lights go off.

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30. The second argument contends that the Staff's review (Hearing File Item i l

l 10) acknowledges that the stabilization of reactor pressure is an allowed concurrent step, l

thereby supporting Mr. O'Hern's answer choice (a) as the correct action. Mr. O'Hern 1

states in Presentation Item 2: " Entering the pressure leg of the EOP's has the ,same l

l priority as entering the power leg." Although entering all legs of the EOP is a concurrent l action, the Staff disagrees that this action alone makes Mr. O'Hern's answer choice I

correct. Student Text ST-OP-802-3003-001 states that 1

[t]he symptomatic approach to emergency response, upon which the EOPs l are based, precludes being able to establish in advance a priority for l executing any of the parallel action paths of RPV Control. Rather, current l values and trerds of parameters and the status of plant systems and l l equipment dictate the relative importance ofindividual RPV Control steps and the relative priority with which they should be accomplished.

l (Hearing File Item 38, page 13). The given plant conditions of question 7 - reactor pressure at 1085 psig slowly increasing and reactor power stable at 48% -- clearly dictate l placing the Reactor Mode switch in SHUTDOWN as the operator's priority action l according to the procedures AOP 20.000.21 (Reactor Scram) and both EOPs 29.100.01 SH 1 (RPV Control) and SH 1 A (RPV Control-ATWS). The action to use SRVs to reduce reactor pressure was not a required action based on plant conditions. This is further l

explained in paragraph 31, below.

31. Mr. O'Hern's selected answer choice, (a), is incorrect because not only l is it not the first action that must be performed as discussed in paragraph 30 above, it is l~

not required for the given conditions. According to procedure EOP 29.100.01 SH 1A, manual operation of the SRVs is required only if any SRV is cycling. (Hearing File Item 1

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34, steps FSP-1 and 2). According to Student Text ST-OP-802-3003-001 SRV cycling is defined a.s multiple, closely sequenced valve actuations with valve opening being initiated i

)

l in response to RPV pressure increasing to/above the lifting setpoint, valve closure being governed by RPV pressure decreasing to/below the reset setpoint. (Hearing File Item 38, page 56). The question states that reactor pressure is 1085 psig and increasing slowly.

This condition is not representative of cycling SRVs. The question's cited pressure of 1085 psig is less than the pressure at which the SRVs would automatically open. While ]

l the Staff agrees that stabilization of reactor pressure is allowed as a concurrent step, the l Staff disagrees that this validates answer choice (a). The NRC's informal review stated i

that the action to stabilize pressure less than 1050 psig is not yet required since reactor l pressure is less than the threshold pressure of step FSP-3 (stabilize RPV pressure < 1093 psig) and the question asked for the action that must be performed. (Hearing File Item 10, page 3).

32. The Staff has reviewed Mr. O'Hern's contentions and has determined the original answer remains the only correct one. Placing the Reactor Mode switch in SHUTDOWN is an immediate action required of the operators for any scram condition, i

according to procedure AOP 20.000.21 referenced by Mr. O'Hern in the request for an informal review. (Hearing File Item 5). Moreover, the governing procedural action for

. the question's scenario, step FSQ 1 of EOP 29.100.01 SH 1 A directs confirmation of this action. (Hearing File Item 34). Student Text ST-OP-315-0027-001, " Reactor Protection System," confirms this interpretation for the question's conditions (condition exists that

requires a reactor scram and reactor power cannot be determined to be below 3 percent).

, l l - (Hearing File Item 35). In section V.E.2.a. of ST-OP-315-0027-001, it states that the first step is for the RO to initiate a reactor scram by taking the Reactor Mode switch to 4 l SHUTDOWN if this has not already been done. The Staff disagrees with the l Presentation's conclusion that "with absolute certainty, that if 3D73 and 3D74 are actuated then the blue scram lights have to be o,ff." The Presentation's conclusion is justified by tracing a scram signal through the RPS logic. H'8 wever, the justification does not consider any RPS logic or electrical component failures and, as a result, the conclusion disregards the potential for multiple sensor, sensor relay, and contact failures in the automatic RPS logic and the scram solenoid circuitry. The question's premise of an ATWS condition is consistent with just such a multiple failure scenario. The multiple sensor relay and contact l failure example provided in paragraph 27 would result in activation of both annunciators 3D73 and 3D74, but not all of the blue group scram lights would be extinguished.

Although the blue group scram lights may or may not have been off, depending on whether the RPS logic operated properly, it does not negate the requirement for the ROs to carry out the immediate operator action of l

AOP 20.000.21 (Reactor Scram), to place the Reactor Mode switch in SHUTDOWN, or to " Confirm Rx [ reactor] mode switch in S/D [ shutdown]" according to EOP 29.100.001

]

SH 1 A. (Hearing File Item 34, step FSQ-1). Failure to confirm this action would violate both the EOP mandate "to take corrective actions if the identified conditions do not exist,"

(Hearing File Item 39, page 29), and facility licensee's management direction to shut down the reactor when reactor protection setpoints have been exceeded and automatic action has l

not occurred. (Exhibit 7, Section 2, page 1). While verification that the blue scram lights I

. . . - - . - - .- _ -- - = . . . - . . --

I

l. l are off may not confirm that all the control rods are inserted, it can assist in confirming proper operation of the Reactor Mode switch with regard to the RPS system. (Hearing File Item 35, page 29).

l 33. As stated in paragraph 16 above, the Staff's expectations for following and l complying with the facility's procedures are specifically stated in the licenses it issues to i

l reactor operators pursuant to 10 C.F.R. f 55.51.

34. . In summary, Mr. O'Hern's selected answer, (a), manually operate SRVs to stabilize reactor pressure, is incorrect because it is not the first action that must be l

performed and it is not required for the given conditions. In his Presentation, Mr. O'Hern argues that placing the Reactor Mode switch in the SHUTDOWN position is, in effect, not l necessary. This is incorrect. The operator action to place the Reactor Mode switch in SHUTDOWN is a necessary action to mitigate the consequences of the event.

Mr. O'Hern's arguments in support of his answer, given his opportunity to review the i

! procedural requirements to the contrary, demonstrate a significant and continuing lack of l understanding of appropriate operator actions for emergency plant conditions.

35. Question 54, the Staff's discussion of the question, Mr. O'Hern's contentions, and the Staff's response are set forth in the following paragraphs.

36. RO Ouestion #54. " Heavy thunderstorms just caused a load-reject from

.100% power. The reactor conditions are:

APRM Power stable at 20%

No indications of control rod position Recirc [ Recirculation] pumps tripped l

l l

f' I

All Main Steam Isolation Valves [MSIVs] are open l

Reactor Level being maintained by feedwater Reactor Pressure being maintained through Turbine Bypass Valves Mode switch in SHUTDOWN ..

The NSOs [ Nuclear Supervising Operators] first actions should be:

(a) Initiate Automatic Depressurization System [ ADS]

(b) Initiate Alternate Rod Insertion [ARI]

(c) Inject Standby Liquid Control [SLC]

(d) Drive Control rods in"

37. The question asks for the first action that must be performed by the NSO i following a load reject from 100 percent power. The correct answer choice per the answer key is (b), Initiate Alternate Rod Insertion [ARI]. For the given conditions, with a load-reject from 100 percent power and the Reactor Mode switch in SHUTDOWN, the reactor should have automatically shut down, (scrammed), and reactor power should be 0 percent.

(Hearing File Item 35, page 15). However, reactor power stable at 20 percent (APRM power) indicates an ATWS condition exists similar to that described in paragraph 27 for question #7. ARI provides an alternate method to initiate control rod insertion and is designed to function as a backup to RPS, in an ATWS situation, by providing a full insertion'of all control rods. (Hearing File Item 38, page 63). ARI is also a relatively

. prompt method of inserting control rods which generically will take approximately 25 to 30 seconds to fully insert the control rods from the time of initiation. The SLC system

provides another backup method to manually shutdown the reactor independent of control l

l l

I

t

, rod insertion. The shutdown is accomplished by injecting Sodium Pentaborate, a strong neutron absorber, to absorb thermal neutrons and thereby terminate the nuclear fission chain reaction. EOP 29.100.01 SH 1A, steps FSQ-11 and 12, call for SLC initiation i

before torus temperature reaches the Boron Injection Initiation Temperature (BIIT).

1 (Hearing File Item 34). However, according.to Student Text ST-OP-802-3003-001, reactor shutdown on control rod insertion alone is preferable to injecting boron for the following reasons: (1) boron injection contaminates the primary system requiring extensive cleanup; (2) if a leak occurs below the elevation of the RPV water level being maintained, l

boron injection may not be successful in shutting down the reactor; and, (3) a reactor shutdown on boron is not necessarily a stable condition, in that, should the boron be j subsequently diluted or displaced, the reactor could return to criticality (Hearing File i Item 38, page 63). Another disadvantage of SLC is that the reactor power reduction will  !

take a number of minutes and is therefore slow when compared to the power reduction achieved from control rod insertion. The question's conditions (Mode Switch in ,

Shutdown, generator load reject, all MSIVs open, and Recirculation pumps tripped) confirm completion of steps FSQ-1 through 6 of EOP 29.100.01 SH 1A. (Hearing File Item 34). The next required operator action according to step FSQ-7 is to " Confirm ARI."

(See footnote 7 for definition of " Confirm"). However, with reactor power at 20 percent and no definitive indications presented in the question to allow the operator to confirm a successful ARI activation according to step FSQ-7 of EOP 29.100.01 SH 1 A (Hearing File Item 34) and Emergency Support Procedure (ESP) 29. ESP.06, " Manual Operation of

t 24 -

Alternate Rod Insertion" (Hearing File Item 27, page 3), the NSOs first required action is to initiate ARI.

38. Mr. O'Hern selected answer choice (c), Inject Standby Liquid Control.

In his request for an informal staff review, Mr. O'Hern requested that answer choice (c) also be accepted as correct since the question does not specify the cause for the trip of the Recirculation Pumps and there are no interlocks that would cause their trip on a load reject. (Hearing File Item 5). Therefore, it is reasonable to conclude that the trip of both Recirculation Pumps was caused by the resulting pressure transient with pressure increasing to greater than the ATWS/ARI setpoint of 1133 psig. Id. The Presentation, item 3, reiterates the above contention "that the same parameters that cause both Reactor Recirc Pumps to trip also cause the ARI valves to reposition," and alleges that the NRC agreed in the September 1998 informal review response with the hypotheses contained .

therein, which confirm the Presentation's conclusion "that initiating ARI would not be my first action because it had aheady occurred." Specifically, Mr. O'Hern states that the NRC agreed that there was no loss of offsite power (LOSP) and that a pressure transient may have caused the trip of the Recirculation Pumps. Furthermore, he references Appendix B of NUREG-1021, " Operator Licensing Examination Standards for Power

~ Reactors" (Hearing File Item 19), as supportive of his position that the tripped status of the Recirculation Pumps in the question stem "had some bearing on the answer being solicited." Finally, he argues that the Staff, by referencing step FSQ-7, is " demanding that I memorize all the steps of the EOPs" which results in a question at a " higher level than what is required to be a licensed reactor operator."

, 39. Mr. O'Hern's assertions that the Staff speed with the two hypotheses discussed in paragraph 38 misconstrues the Staff's response. The Staff's response is presented in the analysis and conclusion section. (IIcaring File Item 10, page 7). The response addressed Mr. O'Hern's assertion that the absence of a concurrent LOSP confirms that the Recirculation Pumps tripped on high pressure, by explaining that "the question does not state the cause for the trip of the Recirculation Pumps, nor does it indicate the pressure transient was of sufficient magnitude to cause an automatic ATWS/ARI actuation." The Staff's acknowledgment that a pressure transient occurred is

, not agreement that it was sufficient to trip the Recirculation Pumps. Although the Staff's response agreed that the trip of both Recirculation Pumps is a resultant automatic action from an ATWS-ARI/ Recirculation Pump Trip (RPT), according to Hearing File Items 25 and 26, the trip, no matter the cause, is not sufficient to confirm ARI according to step FSQ-7 of EOP 29.100.01 SH 1A, and the step's referenced procedure, 29. ESP.06,

" Manual Operation of Alternate Rod Insertion." (Hearing File Item 27). Procedure

29. ESP.M confirms ARI initiation by having the operator verify that the ATWS/ARI valves have changed position to the tripped position. Therefore, the conditions presented in the question stem (reactor power stable at 20 percent, no indications of control rod position, and no confirmation regarding the actuation of the a fWS/ARI valves) do not confirm a proper and successful automatic ARI activation according to step FSQ-7 of EOP 29.100.01 SH 1 A. (Hearing File Item 34).

40. The Presentation references NUREG-1021, Appendix B, Item d, and concludes that this guideline supports the candidate's assumption that the trip of both

. .. - - _ . ~ - . . . ..- - - .- _-.-- -- . . - _ - .- . - - - . . -

l-i Recirculation Pumps "had some bearing on the answer being solicited." (Hearing File Item 19). While the Staff disagrees that guideline C.2.d of Appendix B mandates that all l

information in the question stem in effect must have a bearing on the selection of the correct answer _ choice, the argument is unavailing. The tripped status of both Recirculation Pumps in the question stem is, in fact, relevant in determining the NSO's first action. The tripped pump status was provided to ensure that the candidates would recognize that action consistent with step FSQ-6, " Trip RR pumps," had occuned.

41. Mr. O'Hern also contends that injection of the SLC system is the appropriate answer for the conditions presented. He referenced Operations Training Policy - Operator Expectations Clarification, Tracking Number 97-001, as supportive of this contention. (Hearing File Item 43). The Staff agrees with the Operations Training Position, which directs: "If the Reactor Protection System fails to automatically insert all control rods, and initiation of Alternate Rod Insertion fails to insert all control rods, and greater than one control rod remains withdrawn, then Standby Liquid Control must be injected into the Reactor Pressure Vessel immediately." However, operator action to l initiate ARI should precede SLC injection. The facility licensee's policy for the situation i

presented in the question clearly supports the Staff position that th' e NSO's first action is to initiate ARI prior to injecting SLC.

i

42. The Staff has reviewed Mr. O'Hern's contentions and has determined the original answer remains the only correct one. The applicant's hypothesis that an

{ ATWS/ARI actuation can be confirmed from the information presented in the question stem is invalid. There is no direct relationship or interlock between a load-reject and an i

l l

l \

27 -

I ARI initiation. The ARI system is actuated on low reactor water level, high reactor l l

pressure, or manually. (Hearing File Item 35, page 13). The question's initial conditions l l

eliminate low reactor water level as an actuating condition (reactor water level being l maintained by feedwater) and provide no information allowing the candidates to l conclusively determine that the high pressure ARI actuation setpoint of 1133 psig was reached as a result of the load reject. Furthermore, even if the pressure transient was sufficient to exceed the ARI initiation setpoint, as stated in paragraph 38, the conditions presented in the question stem do not confum a proper and successful initiation according to step FSQ-7 of EOP 29.100.01 SH 1A. (Hearing File Item 34). Mr. O'Hern has  ;

i incorrectly concluded that the trip of both Recirculation Pumps, without conclusively knowing the cause of the trip, is sufficient to properly

• Confirm ARI" activation with no further operator action required. This conclusion is contrary to the specific direction of step FSQ-7, according to the definition and usage information for " Confirm," (See footnote 7), and the 20 percent reactor power condition which clearly indicates ARI has not properly actuated, if at all. Mr. O'Hern's argument that injection of SLC is the next required step is based on his incorrect determination that ARI had automatically initiated.

The limiting factor during an ATWS, which defines the requirement for boron injection, is a challenge to the primary containment. A scram failure coupled with a MSIV isolation results in rapid heatup of the torus due to the steam discharged from the RPV via the SRVs. If torus water :emperature in the primary containment and RPV pressure cannot be maintained below the Heat Capacity Limit, rapid depressurization of the RPV will be required. To avoid depressurizing the RPV with the reactor at power, it is desirable to

,e 28 -

shut down the reactor prior to reaching the Heat Capacity Limit, thus minimizing the quantity of heat discharged to the torus. The BIIT is defined so as to achieve a reactor shutdown without an RPV depressurization. (Hearing File Item 38, pages 62 and 67).

This was not the case for this question since the question clearly states that re, actor pressure is being maintained through the turbine bypass valves (to the Main Condenser),

thereby not discharging to the torus and not challenging the primary containment.

Although SLC injection is required before torus water temperature reaches the BIIT and l may be performed earlier, the question's conditions do not warrant its immediate initiation prior to initiation of ARI. Moreover, as discussed in paragraph 37, reactor shutdown on control rod insertion is preferable for several reasons.

43. Mr. O'Hern contends that the Staff's reference to actions contained in Step FSQ-7 of EOP 29.100.01 SH 1A (Hearing File Item 10, page 7) demands complete memorization of all of the steps of the EOPs which is contrary to the shift team responsibilities discussion in Student Text ST-OP-802-3001-001, " Emergency Operating Procedures,"Section I (F). (Hearing File Item 39, page 10). Mr. O'Hern states that this is clearly above and beyond the scope required for any licensed candidate. In addition, Mr. O'Hern references Student Text lesson Objective 01-04 (Hearing File Item 38, page 5), and states that the appropriate EOP Flowchart should have been provided.

However, Mr. O'Hern arbitrarily selected this 12sson Objective (01-04). The facility licensee's referenced learning objective for the question, lesson Objective 01-21, states:

! "Given an Anticipated Transient Without Scram event analyze plant conditions and j prioritize alternate conttol rod insertion methods." (Hearing File Item 38, page 6).

1 i

,, r + v-- , , - - -

, I i

Although, licensed operators are not required to know the EOP flowchans from memory, they are required to understand their content. At no time according to the RPV Power Control (FSQ) leg of procedure EOP 29.100.01 SH 1 A is it correct to inject'SLC prior to initiat i ng ARI. (Hearing File Item 34). (See reasons for preferring rod insertion over  !

~

boron injection in previous paragraph 37).

l 1

44. Finally, Mr. O'Hern referred to Hearing File item 9 page 12 to support his contention that his answer was correct, stating that the appeal board originally agreed i

with his answer. He was incorrect in this contention. See Munro Affidavit 112-7 . )

l 45 In summary, Mr. O'Hern's selected answer, (c), Inject Standby Liquid -

j Control, is incorrect because it is not the first NSO action that must be performed and it is not required for the given conditions. Mr. O'Hern argues that he verified ARI actuation by noting the trip of the Reactor Recirculation pumps. This is an incorrect method for I

verifying ARI initiation according to the facility licensee's procedures. (Hearing File Items 34 and 27). The operator action to initiate ARI is a necessary action to mitigate the consequences of the event described in the question. Mr. O'Hern's arguments in support l

of his answer, given his opportunity to review the procedural requirements to the contrary, I demonstrate a continuing lack of understanding of appropriate operator actions for emergency plant conditions.

46. Question 59, the Staff's discussion of the question, Mr. O'Hern's

! contentions, and the Staff's response are set forth in the following paragraphs.

l r

-o i

47. RO Ouestion #59. "If the Reactor Mode switch is in START

[STARTUP]/ HOT STANDBY, which one of the following instruments is NOT required to be operable?

a. Reactor Vessel Level 1 for Automatic Depressurization System [ ADS]

b. Reactor Vessel Pressure High for Alternate Rod Insertion [ARI]

c. Reactor Vessel Pressure for High Pressure Scram

d. Reactor Vessel Level 2 Reactor Water Cleanup System Isolation"

48. The question asks which instrument is not required to be operable with the Reactor Mode switch in START [STARTUP]/ HOT STANDBY or Operational Condition -

l 2 - STARTUP. (Hearing File Item 30, page 1-10j. The correct answer choice per the answer key is (b), Reactor Vessel Pressure High for Alternate Rod Insertion. As stated  ;

earlier,10 C.F.R. I 55.41(a) requires that "[t]he written examination for an operator will contain a representative selection of questions on the knowledge, skills, and abilities needed to perform licensed operator duties. The knowledge, skills, and abilities will be identified, in part, from learning objectives derived from a systematic analysis oflicensed operator duties performed by each facility licensee and contained in its training program.

.. The applicant's training for this question is addressed by the licensee's referenced lesson objective,01-10, which states: "Given the conditions or parameters associated with L

i l the Reactor Pressure Vessel Instrumentation, determine if entry into action statements of l

L Technical Specifications would be required." (Hearing File Item 41, page 7). The facility licensee's learning objectives are listed for all licensed operators, including reactor i

[ operator applicants. The facility licensee's Student Text "01" level objectives are for

1

. 1 NSOs, Senior Reactor Operators, and Shift Technical Advisors. The NSOs are the control  !

room licensed ROs. The facility licensee's Operations Department Instruction (ODI) - l 007, " Command and Control," Revision 2, April 29,1997, Section 2, states that the l

NSOs, CRNSO and P603 NSO, are"[d]irectly responsible for the manipulation of control room controls and the supervision of field activities." (Exhibit 8). This question tests I i

whether the applicants can identify which systems are needed and which are not needed I l

when the reactor mode switch is in STARTUP/ HOT STANDBY. With the Reactor Mode i

switch in this position, according to General Operating Procedure (GOP) 22.000.02," Plant 1 i

Startup to 25% Power," reactor operation is limited to less than 10% power (Exhibit 9, pages 35 and 36). The question therefore tests the applicants' knowledge of system operational requirements at high and low reactor power. The Staff believes the question involves general application of system and procedural knowledge regarding system

" operability," and as such comports with 10 C.F.R. i 55.41 (b)(5) and (b)(10), Pursuant j to the Technical Specifications and GOP 22.000.02, the ADS system, the high pressure reactor scram, and the reactor wa'.er clean up isolation are all required with the reactor l

j mode switch in START / HOT STANDBY. General system knowledge of the protective l

l functions for these systems ard protective functions would allow an operator to determine which systems and/or fur.ctions are required for this Reactor Mode switch position and plant Operational Candition, and therefore eliminate those choices (a, c, and d) as incorrect for thir, question. According to the Fermi Updated Safety Analysis Report Section 7.6.1.111, the ARI function responds to an ATWS event and provides a redundant means of inserting control rods and also causes a Recirculation Pump trip. (Exhibit 10).

I 1

l Also, Technical Specification 3.3.4 requires the ATWS-RPT trip function, which includes 1

the reactor vessel pressure high instrumentation for ARI, to be operable in Operational l 1

Condition 1 only. (Hearing File Item 28, page 3/4 3-32).

48. Mr. O'Hern selected answer choice (a), reactor vessel level 1 for the j automatic depressurization system (ADS), as not required for the given plant conditions.

1 In his request for an informal staff review, Mr. O'Hern argued that the ADS l l

instrumentation is not required to be operable below 150 psi reactor steam dome pressure and therefore his answer was correct. (Hearing File Item 5). This question does not require any assumptions regarding reactor pressure. The wording of the question stem j encompasses by definition (Hearing File Item 30, page 1-10), all applicable plant conditions for Technical Specifications Oper tional Condition 2 (Startup/ Hot Standby) and, l therefore, includes any temperature and pressure. This makes answer choice (a) incorrect.  ;

49. In his Presentation, Mr. O' Hem argues that he was being held to a Senior Reactor Operator standard which requires knowledge of Technical Specifications and their bases. He requested that the question be deleted. The Staff has reviewed Mr. O'Hern's contentions in this regard and has determined that the question should not be deleted from the RO examination. The question was selected by the facility licensee based on a valid learning objective (01-10) which does not address Technical Specifications bases. This conclusion was also affirmed by the facility licensee in the submission of NUREG-1021 Form ES-201-2, " Examination Outline Quality Assurance Checklist," which among other checks verifies according to items 1.a and 4.b "that the outline fit (s) the appropriate model l

l per ES-401," and "the 10 C.F.R. f 55.41/43 and 55.45 sampling is appropriate."

l l

. (Exhibit 11). In addition, the question, though modified by the facility licensee, was selected from the licensee's question bank, and was also referenced a 1993 RO written examination question. (Hearing File Item 2, question 59). This is a clear indication by the facility licensee, consistent with the question's referenced learning objective, that the question was at an appropriate level and not in excess of the legal requirements for an RO.

Also, as indicated in paragraph 13, the facility licensee's training program for licensed operators is based on a " systems approach to training," which as defined in 10 C.F.R.'

6 55.4 includes learning objectives derived from the analysis which describe desired

. ... perfonnance after training. Therefore, the question and referenced learning objective are also consistent with the content requirement of 10 C.F.R. 6 55.41 (a).

50. In summary, Mr. O'Hern's selected answer choice (a) is incorrect because l- the question's stated condition (Reactor Mode switch in STARTUP/ HOT STANDBY) is not limited to reactor steam dome pressures less than 150 psig. Mr. O'Hern also argues that he was being held to a Senior Reactor Operator standard, requiring knowledge of l Technical Specifications and their bases, and requested the question be deleted.

Mr O'Hern is incorrect in his assertion. This question was selected by the facility licensee i based on a valid learning objective for ROs. The question does not require detailed memorization of Technical Specifications or understanding of the bases, but rather requires

general application of operational system and procedural knowledge required to recognize l

! system functions for applicable operational conditions. Mr. O'Hern's arguments in f

support of his answer derconstrate that he does not possess the requisite knowledge to perform as a licensed operator.

. . . - . . ~ . . --- - . . . - .- -. .- .- - - -

(..

. -M-

51. Question 87, the Staff's discussion of the question, Mr. O'Hern's contentions, and the Staff's response are set forth in the following paragraphs.

' 52. RO Ouestion 87. "The plant is operating at 96% power with the following indications on the A Recirculation Pump seal:

Seal #1 Pressure 980 psig Seal #2 Pressure 10 psig Annunciator 3D123, RECIRC PMP STAGING SEAL FLOW HIGH/ LOW is alarming

, Flow indication indicates 0.4 gpm l

Which one of the following seal conditions exist?

a. Seal #1 has failed .

i 1

b. Seal #2 has failed l

c. -# 1 Seal Labyrinth is plugged o

d. # 2 Seal Labyrinth is plugged"

53. The question asks the candidate to diagnose the seal conditions for the A Reactor Recirculation pump. The correct answer per the answer key was answer (c), # 1 Seal Labyrinth is plugged. Student Text ST-OP-315-0004-001, " Reactor Recirculation System," Revision 9, states that the indications of the plugging of the #1 seal are #2 seal pressure decreasing with a #1 seal low flow alarm of 0.5 gpm decreasing. (Hearing File Item 42, pages 13,14). The Student Text also states that indications of a #2 seal failure are #2 seal pressure decreasing with a high seal flow alarm of 0.9 gpm increasing. The

l i l

1 l l question stem states that the seal staging flow high/ low annunciator was in alarm and that the seal staging flow was 0.4 gpm.

54. During the week of March 23,1998, the NRC conducted an examination preparation week at the licensee's facility. During this preparation, I noted that question 87, as originally drafted, potentially had two correct answers, a failed #2 seal or a plugged

1. 1 labyrinth. The facility licensee responded that including the existence of tim low flow I

alarm in the stem of the question specifically narrowed the cause of the problem to a l

plugged labyrinth rather than a failed seal. I verified this through the facility licensee's j learning objectives and Student Text, and specifically noted that the low pressure condition 1

on the #2 seal in conjunction with the low flow alarm confirmed that the cause of the problem was a plugged labyrinth. I accepted the facility licensee's submitted question only after the facility licensee added the flow instrument reading of .4 gpm per my request because the annunciator alarm 3D123 may indicate either a high or a low flow condition.

' The specified flow of 0.4 gpm confirmed an actual low flow condition which allowed the candidates to differentiate between the two possible conditions identified by the annunciator alarm.

55. Mr. O'Hern selected answer (b), that seal #2 has failed. In his request for r an informal staff review, Mr. O'Hern stated that answer (b) was also correct in that flow would also fall to less than 0.5 gpm for a failed #2 seal. (Hearing File item 5). He also stated that to differentiate between a failure of the #2 seal and plugging of the #1 labyrinth l

would require seal temperature indication, which was not provided in the stem of the l

question. However, Mr. O'Hern did not indicate any need for additional information

]

regarding the rate of outer seal leakage in order to make a definitive diagnosis of the problem.

56< The Staff confirmed with a facility licensee representative after the i examination that a seal #2 failure will also result in annunciator 3D123 alarming due to low flow, as asserted by Mr. O'Hern. Moreover, the facility licensee provided a post  ;

examination comment that answer choice (b) is also correct based on the statement that 1 annunciator 3D123 may indicate either a seal #2 failure or #1 labyrinth plugging.

(Hearing File Item 8).

57. According to facility licensee Student Text ST-OP-315-0004-001, " Reactor Recirculation System," a #2 seal failure will result in activation of annunciator 3D121,

• RECIRC PUMP A OUTER SEAL LEAKAGE HIGH," in addition to the indications provided in the stem of the question. (Hearing File Item 42 pages 13,14). Since the l

l question does not indicate that annunciator 3D121 is alarming, but that annunciator 3D123 L

is alanning, the application of the system knowledge from the aforementioned Student Text (ST-OP-315-0004-001) leads to the conclusion that the only seal condition that could cause the indications specified in question #87 is plugging of the #1 seal labyrinth.

Consequently, when Region III reviewed Mr. O'Hern's contention, it concluded that the ,

original answer (choice (c)) is the only correct answer to the question as written. (Hearing l File Item 7).

i

58. The informal appeal board accepted Mr. O'Hern's request that answer

, choice (b) be accepted, but it did so for a reason other than that stated by Mr. O'Hern.

3 The appeal board accepted answer (b) because it found that the stem of the question did i

4

( l not contain sufficient information for the applicants to differentiate between answer choices (b) and (c). (Hearing File Item 9, page 11). Sections D.2.g and D.2.h of ES - 502, which specifies the Staff's policies and practices for processing informal staff reviews of initial license application denials, state that the appeal board will submit its m f' dings to the Chief, HOHB, and that the Director, DRCH, will consider the findings and recommendation of the appeal board and make a decision whether to sustain or overturn the applicant's examination failure. (Hearing File Item 18, page 5). It is important to note that the appeal board's findings and recommendations are not final and that the Director's review may result in the applicant's grade increasing (as was the case for question #25) or decreasing, by overturning the Region's and/or the appeal board's recommendations. With regard to question #87, the Director sustained the original grading and answer and stated that it can only be concluded that the seal condition causing the specified indications is plugging of the #1 seal labyrinth. (Hearing File Item 10, page 8). The Director also stated that the l candidate did not provide documentation to support his contention that seal temperature indication would be required for an applicant to differentiate between a failure of seal #2 and a plugging of the #1 labyrinth.

59. In his Presentation, Mr. O'Hern states that there are two answers that will satisfy the conditions given in the stem of the question and that the NRC does not dispute this fact. He contends that the applicant should use (only) the information provided in the l stem of the question and he should not infer information. He then notes that if a non-alarm condition for the outer seal was to be considered it should have been so stated. The presentation also states, "The NRC does not dispute the technical validity of either

~

answer." Mr. O'Hern misconstrued the Staff's position. The Staff's analysis and ,

conclusion for this question clearly states that the NRC does not agree with the applicant's contention and that "it can only be concluded that the seal condition causing the specified indications is plugging of the #1 Seal Labyrinth." (Hearing File Item 10, page 8). With respect to including information in the stem of the question, Section C.I.c of Appendix B, j " Written Examination Guidelines," of NUREG-1021 indicates that all necessary l

information should be provided and that the question should be stated concisely. (Hearing l

File Item 19, page 4). It is the general convention that all abnormal indications necessary l

i i

l to answer the question are provided. But, it would be unreasonable to provide a status for each and every annunciated parameter that is in its normal /non-alarming state. Such extraneous or superfluous information would make the question overly complex. In this t

case, the conditions presented in the stem of question #87 match 60se presented in the l Student Text for a plugged #1 seal labyrinth. (Hearing File Itc d, pge 14). The Staff l agrees that the questions should be answered based on the informa.aon provided in the i

stem. If Mr. O'Hern felt that sufficient information was not available to determine an i

answer, he could have requested clarification from the proctor while the examination was l l

in progress. This policy was orally stated to Mr. O'Hern during the exunination briefing .

l using NUREG 1021, Appendix E, " Policies and Guidelines for Taking NRC l 1

Examinations," as described in previous paragraph 8. Neither Mr. O'Hern nor any of the I other applicants asked for clarifying information regarding the outer seal flow alarm t

! condition, seal temperature indication, or any other aspect of the situation specified in RO I

question #87 (SRO question #86). See Exhibit 4.

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60. In his Presentation regarding question #87, section 1.b (Presentation Itera 5), Mr. O'Hern stated:

i The NRC conveniently uses the opposite logic in their conclusion and analysis of .

Question 59. In that case they argue that the student should not have assumed a  :

reactor pressure when one was not given. No assumptions should ever have to be l . made regarding the information in the stem of the question. T

Again, Mr. O'Hern apparently misunderstood the Staff's September 7,1998 response.

l. l l (Hearing File Item 10, page 8). Indeed, the Staff stated in its analysis and conclusion for question 59 that "The question does not require any assumptions regarding reactor

]

pressure." However, as explained in the analysis, the mode switch position of "Startup/ Hot Standby" clearly places the reactor plant in Operational Condition 2 (Startup)

L which is defined by the facility licensee's Technical Specifications to include "any temperature." (Hearing File Item 30, page 1 - 10). For a boiling water reactor that  :

l includes reactor pressures greater than 150 psig. Therefore, Mr. O'Hern's decision to  !

assume a reactor pressure less than 150 psig, without asking for clarification from the

]

proctor, was contrary to the plant conditions specified in the question stem. As discussed in the previous paragraph the same logic applies to question 87; i.e., the candidate should answer the question using the information provided.

61. In his Presentation, Mr. O'Hern again stated that the plugging of a seal labyrinth would cause a seal high temperature alarm in the control room. Section 2 of the Presentation is quoted as follows:

The NRC, in the final sentence of their analysis and conclusion, states that I did i not provide any justification that the diagnosis of a seal failure requires seal temperature indication. Student Text ST-OP-315-0004-001, " Reactor Recirculation System," discusses the effects of the plugging of a Seal Labyrinth.

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On Page 14 of the Student Text (HF Item 44) it clearly describes the temperature increase that would be seen for a plugged labyrinth. If the Seal temperatures l increased as stated above then 3D100, "RECIRC SYS COOLANT

!. TEMPERATURE HIGH," would be expected to alarm with the conditions given in the stem of the question. The stem of the question failed to address this alarm status.

l The Presentation references page 14 of Student Text ST-OP-315-0004-001. That portion i of the Student Text was neither referenced nor provided in Mr. O'Hern's May 29,1998, j ,

submittal. (Hearing File Item 5). Figure 14, Recirculation Pump Seal Arrangement, was the only information from the Student Text provided at that time.

62. With respect to seal temperature effects, Mr. O'Hern states that if the seal temperatures increased as stated in the Student Text (Hearing File Item 42, page 14) then i

annunciator 3D100 would be expected to alarm. The Staff acknowledges that the l referenced page of'...e Student Text states that #2 seal temperature would decrease if the

1. 2 seal failed and that both seal temperatures would in::rease if the #1 labyrinth were plugged. However, the Student Text does not support Mr. O'Hern's statement that l annunciator 3D100 would be expected to alarm. The May 29,1998 submittal included l annunciator response procedure attachments 3D121 and 3D123 that do indeed advise that seal temperatures be checked, but only in response to an mcreasmg leak rate to the drywell  ;

l equipment drain sump. (Hearing File Items 31 and 32). Per the 3D121 and 3D123

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! annunciator responte procedures, this information is then used to evaluate the l l

Recirculation Pump for possible shutdown and isolation, not to analyze for a plugged j labyrinth. Mr. O'Hern's comments regarding the absence of seal temperature data and the  ;

expected "RECIRC SYS COOLANT TEMPERATURE HIGH" alarm are irrelevant since 1

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the flow alarms and seal pressure changes are sufficient according to 3D123 and ST-OP-315-0004-001 to diagnose a plugged #1 Seal labyrinth with the proper application of Recirculation Pump component design knowledge. Moreover, the flow alarm referenced in the stem of the question will occur before the seal temperatures are substantially affected, because the reactor recirculation pump seals will still have cooling water supplied by the Reactor Building Closed Cooling Water system. (Hearing File Item 42, page 33).

63. Mr. O'Hern's Presentation also refers to a discussion with the facility licensee in which the facility licensee stated that answer choices (b) and (c) are correct.

Mr. O'Hern is correct in that the facility licensee did state this. This was a repeat of the post-examination comment discussed in Paragraph 56 above. The Staff's analysis as described in the preceding paragraphs, including the licensee's own Student Text (ST-OP-315-0004-001), does not support the conclusion that both answer choices are correct.

64. Mr. O'Hern's Presentation references the informal appeal board's analysis for this question (Hearing File Item 9) and states "the Appeal Panel clearly agrees with my position . . . that it was unreasonable to expect me to assume the status of the indication and give me credit for my response, yet I did not receive credit for this question." As stated in paragraph 58, above, the appeal board's analysis and conclusions are recommendations made to the Chief of HOHB. The final determinations is made by the Director, DRCH.

65. In summary, Mr. O'Hern's selected answer choice (b), that seal # 2 has failed, is incorrect. The conditions speci' led in the question stem match those conditions l

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.. in the facility licensee's Student Text for a plugged #1 seal labyrinth. Mr. O'Hern argues that the question stem did not provide sufficient information for him to diagnose the seal condition. The Staff disagrees. The question stem matched the conditions described in the facility licensee's Student Text and did provide sufficient information to perform the ,

l proper diagnosis. Again, Mr. O'Hern's arguments in support of his answer, given his opportunity to review the procedural requirements and Student Text information to the contrary, demonstrate a continuing lack of understanding and ability to properly apply procedures and system knowledge for abnormal plant and system conditions.

66. To date, the following NRC Staff members have evaluated Mr. O'Hern's performance on the written examination and concluded that it was unsatisfactory:

• Myself, Mr. Hironori Peterson , the NRC examiner who approved the written examination which was administered on April 6,1998, and recommended the original written examination failure by signing Form ES-303-1 (Hearing File Item 3) on January 20,1998. Iwas also the NRC l chief examiner for the Fermi initial license examinations.

• Mr. Melvyn I.each, the Chief of the Operator Licensing Branch, charged with making licensing decisions in NRC, Region III, concurred with the examiners' recommendations by signing Form ES-303-1 on May 18,

1998, (Hearing File Item 3) and issuing the original license denial letter i

! (Hearing File Item 4) on May 20,1998. He reevaluated his position in response to Mr. O'Hern's informal request for a regrade (Hearing File i

Item 5), found no basis to revise the original grade for the written i

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examination, and reiterated his conclusion that Mr. O'Hern had failed the written examination in a memorandum dated June 16,1998. (Hearing File Item 7).

!

• Mr. John Pellet, the Chiefof the Operations Branch in NRC, Region IV, was charged with chairing an appeal board that independently reviewed 1

Mr. O'Hern's request for an informal regrade. Although he recommended deleting questions 17, 38, 56, and 71, and reassigning a grade of satisfactory to question 87, he sustained the denial of the license.

(Hearing File Item 9).

• Mr. Thomas Meadows, an examiner from NRC, Region IV, and Mr. Joe D' Antonio, an examiner from NRC, Region I, were the other members l

of the appeal board. They supported Mr. Pellet's review and conclusions t

stated above. (Hearing File Item 9),

e Mr. John Munro, an examiner from HOHB, NRR, charged with l

independently reviewing, assessing and reconciling the applicant's contentions, the Region III informal review results, and the informal review board's results. He identified the appeal board's summary of question changes did not accord with the text of the memorandum. He agreed with the Region III results for question 87 but also found that question 25 should be deleted from the examination. He recommended that the original examiner grading of unsatisfactory for the written i

• examination be sustained. (Munro Affidavit j 6-7).

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• Mr. Robert Gallo, the Chief of the HOHB, NRR, charged with making the final recommendation to the Director of DRCH, NRR regarding the outcome of the informal appeal process, concurred with Mr. Munro's l recommendation and found no basis to change the grading of the written l examination.

i

• Mr. Lee Spessard, Director of DRCH, NRR, charged with making i

l licensing decisions during the examination appeal process, concurred with HOHB's recommendation to sustain the written examination failure in a t

letter to Mr. O'Hern dated September 7,1998. (Hearing File Item 10).

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SUMMARY

OF NRC REVIEW l

l 67. In summary, the NRC Staff has concluded the following based upon its review of all relevant documents, including the Presentation and Mr. O'Hern's affidavit:

• Contrary to the requirements of the NRC Staff and Detroit Edison, l Mr. O'Hern failed to properly identify the correct action to place the Reactor Mode switch in shutdown. The operator action to place the Reactor Mode switch in shutdown is a necessary action to mitigate the l

consequen.es of an ATWS event.

• Contrary to the requirements of the NRC Staff and Detroit Edison, Mr. O'Hern failed to properly identify the correct action to initiate ARI i

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. or an ATWS event. Mr. O'Hern incorrectly argued that verifying the trip of the recirculation pumps verified the initiation of ARI. The operator ,

l action to initiate ARI is a necessary action to mitigate the consequences of f an ATWS event.

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• Contrary to the requirements of the NRC Staff and Detroit Edison, Mr. O'Hern failed to properly identify the correct reactor vessel l

instrumentation requirements. The question was selected by the licensee based on a valid learning objective for ROs and required general application of operational system and procedural knowledge to recognize system functions for applicable operational conditions.

• Contrary to the requirements of the NRC Staff and Detroit Edison, Mr. O'Hern failed to properly diagnose the proper condition for a faulted l

Recirculation Pump seal. The question stem matched the conditions described in the licensee's Student Text and did provide sufficient infonnation to perform the proper diagnosis.

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• '1 NRC StafT's evaluation of Mr. O' Hem's written examination was 1 appropriate and the NRCs grading standards were properly applied.

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68. Based upon the above, the Staff has concluded that Mr. O'Hern's

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$ t submittals & not establish good c::use to change the grading of his written examination.. l j;/""7%O' r)'s final grade for the written examination remains below the minimum I

- ~g " *.755= '

v ,

~~ ' ~p 6 sing grade. Therefore, Mr. O' Hem has failed the written examination and the )

I licensing examination. The NRC Staffs denial of Mr. O'Hern's application for an RO ,

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! license should be ,sustamed 1

l 9; I hereby certify that the foregoing is true and correct to the best of my 1

1 1 , .

j knowle informa __

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Hironori Peterson, Senior License Examiner '

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l Subscribed and swom to before me l I l , this hy of J uary 1999.

/d2 " 9A %A 4

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} Notary Public i My commission expires:<

4 Mdl$$$d'st7h CHRISTOPHER M. MILLER 4

NOTARY PUBLIC STATE OF ILUN0l3

( My Commisalon Empires Feb.20,2002 vvvvyvyyyyyy l l <

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,:: =~M F

E j STATEMENT OF PROFESSIONAL QUALIFICATIONS HIRONORI PETERSON PRESENT POSITION-Senior Operator Licensing Examiner / Reactor Engineer Series 840 GG-14 DUTY STATION: .

NRC Region lll 801 Warrenville Road Lisle,IL 60532 EDUCATION / TRAINING.

B.S. Nuclear Engineering, Oregon State University, Corvallis, Oregon. June 1980.

'* Naval Nuclear Power Training, Orlando, Florida and Idaho Falls, Idaho. June 1980 to July 1981.

Naval Submarine School, New London, Connecticut. July 1981 to October 1981.

Senior Reactor Operator (SRO) Certification, Westinghouse Electric Corporation, Nuclear Training Center, Zion, Illinois. February 1986.

Additionallisting of NRC training upon request.

EXPERIENCE HIGHLIGHTS-NRC' o Senior Resident inspector - Byron Nuclear Power Station, PWR e Certified as Senior Operator Licensing Examiner for the following vendors:

General Electric, Westinghouse, Combustion Engineering, and Babcock and

- Wilcox

• Certified as Resident inspector e Certified as Emergency Preparedness inspector / Analyst Westinghouse e Certified as Westinghouse PWR Senior Reactor Operator (SRO) e Certified as Westinghouse Instructor (Classroom and Sirnulator) e Development of Emergency Preparedness Program and Drill Scenarios e Sta#.up Engir.eer - Technical Specification Surveillance Scheduler and Testing Expeditor Braidwood Startup N:vy e Lieutenant U.S. Navy; Nuclear Submarine Force o Qualified as Engineering Officer of the Watch and Duty Officer e Qualified as Quality Assurance inspector o Communications Security Material System (CMS) Custodian _ I e Honored with two Letters of Commendation for outstanding performance of duty in respect to technical knowledge, organizational ability, and leadership i 1 1:

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EXPERIENCE-1 1

September 1996 Senior Reactor Engineer, Senior Operator Licensing Examiner.

! to Present U. S. Nuclear Regulatory Commission, Region lil, Glen Ellyn, Illinois.

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, Recently, I was reassigned to the Operator Licensing Branch to compensate for the lack

! of experienced examiners to meet the demands of the examination schedule. It has l been approximately two years since reassignment. During which I have participated in L the 1996 Examiner's Conference and the Region lli Examination Workshop. Both of l these activities focused on the new revision to the Examiner Standards, particularly j NUREG 1021, Interim Revision 8. In particular, I was the master of ceremonies (MC) for  ;

j the Examination Workshop which involved presentations and interactions with Regional facility and intemational representatives. Since retuming to the operator licensing l branch, I have identified numerous issues pertaining to both initial operator license and j licensed operator requalification activities. These issues included the identification of

licensee's inadequacy for meeting the experience requirements to take initial operator license examination, licensee's not correctly maintaining active operator licenses, 3

licensee's inadequacy in following emergency operating procedures, licensee's ,

4 examination security problems, and inadequacy of licensee's to develop operator license  !

] examinations in accordance with NUREG 1021, Interim Revision 8. Pertaining to the i j identified issues, I was presented with two Special Act awards. I have also occasionally  ;

( acted as the Branch Chief during his absence and dealt with and resolve issues 1 j pertaining to operator licensing.

M'

y 1993 to Senior Resident inspector, Byron Nuclear Power Station. )

September 1996 U. S. Nuclear Regulatory Commission, Resident inspector Office Byron, Illinois. '

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l As a Senior Resident Inspector, I am the lead NRC inspector responsible for supervising j other inspectors assigned to the site (two resident inspectors, one Illinois Department of Nucles Safety. inspector, and region based inspectors on periodic basis). As the Senior f Resident inspector i represent the NRC to the licensee, state and local officials, and the

i. news media. I have performed in-depth evaluations of incidents and abnormal conditions at the site. Based on the results of the inspections and investigations, I evaluated the overall safety of the reactors at the site, developed proposals for what should be done to correct potentially unsafe conditions arid recommended these proposals through line supervision to the appropriate licensing and standards office in

the NRC. These proposals would take the form of changes in the design of the plant, the l license, or the NRC's rules and regulations. I have identified findings regarding whether

the licensee was in compliance with specific provisions of the license, rules and j regulations; prepared notices of violation to the licensee; recommended escalated enforcement action in form of civil penalties or other orders.  !

' As the Senior Resident inspector at the Byron Nuclear Power Station, I have reviewed safety evaluations, operability decisions, engineering design issues, and requests for enforcement discretion. I was involved in numerous complex technical issues and responded to several abnormal and emergency events. These issues included problems 2

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.h identified in the areas of plant operation, maintenance, radiological protection, radwaste controls, chemistry, fire protection, emergency preparedness, engineering, and quality assurance. Some specific items included enforcement actions on inadequacy of flood seal impairment on auxiliary feedwater tunnel, the absence of an SRO in the control room, and the inadequate design configuration of the containment spray additive system.

Other issues included identification of negative trends conceming personnel errors, poor .

maintenance practice, and engineering problems associated with equipment qualification l and Reg Guide 1.g7. As the lead inspector, I have responded to numerous emergency events. I have evaluated licensee events including reactor trips (control rod drop, turbine trip, and a manual trip due to loss of all non-essential service water), fuel handling events (damaged fuel assembly during core restoration), inadvertent ESF actuations, safety equipment failures (diesel driven auxiliary feedwater pump bearing failure), an unusual event for loss of all station communication, and an unusual event for loss of off-site power.

August 1989 Senior Reactor Engineer, Senior Operator Licensing Examiner. )

to Present U. S. Nuclear Regulatory Commission, Region lil, Glen Ellyn, Illinois.

Certified as a Senior Operator Licensing Examiner for GE BWR and Westinghouse PWR facilities. Developed, prepared and administered, written, oral, and simulator examinations for applicants for operators' and senior operators' licenses for nuclear power plants; evaluated the results and made recommendations for approval or denial of the license; reviewed and processed to final action, as Principal Examiner, applications t for and communications conceming license applications submitted pursuant to 10 CFR .

Part 55. I was responsible as Principal Examiner for all Commonwealth Edison BWR Plants (LaSalle, Dresden, and Quad Cities); reviewed licensee's methods of training, requalifying and evaluating plant staff members; reviewed changes to training programs and participated in continuing development of acceptable criteria; and responsible for overseeing adequacy of licensee training program, examination appeals, SALP, and any enforcement actions relating to Operator Licensing. As a Chief Examiner, responsible for interfacing with the licensee in resolving matters of licensing issues relating to initial ,

and requalification examinations; prepared, administered, and evaluated initial and/or '

requalification examinations for Big Rock Point, Dresden, LaSalle, Quad Cities, Fermi 2, ,

Duane Amold, Clinton, and Monticello. Recommended license renewal or accelerated  !

retraining as needed in observed weak areas. Conducted an Operational Evaluation j (OPEVAL) of licensee UNSAT requalification training program to determine competence and adequacy for continued safe plant operations. I have reviewed and audited regional and consultant prepared and administered examinations to ensure confom1ance with the l NUREG 1021, Examiner's Standards. Participated in special Task Forces, team inspections, or study groups responding to operational occurrences, as assigned. For example: the AEOD Human Factors fact finding team visit involving the Quad Cities uncontrolled criticality event; the augmented inspection of the Washington Public Power Supply System (WPPSS) WNP-2 control room operations due to the unsatisfactory training program; the coordination of the Monticello Augmented incident Investigation concerning the inadvertent scram due to unexpected criticality and the subsequent licensee enforcement actions. I was the team leader for the Big Rock Point simulation facility certification inspection.

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p l September 1987 Resident inspector. U. S. Nuclear Regulatory Commission, Region lli to August 1989 Glen Ellyn, Illinois.

Served as the NRC Resident inspector at the Duane Amold Nuclear Power Station.

Planned and conducted preventive, independent and reactive inspections during all l phases of operations to assure that the plant is operated in accordance with NRC rules '

j and regulations, inspected licensee's activities regarding reactor operations, technical 4

specification interpretations and adherence, surveillance, maintenance, security, health

physics, chemistry, emergency preparedness, quality assurance, fuel handling and i design changes to assure that operations at the facility were conducted so as to protect nuclear material and faciirties, the environment, and the health and safety of the public.

I Performed and/or assisted in numerous Duane Amold activities, in accordance with NRC l Regulations and Policies; for example, Emergency Operating Procedure (EOP) team j inspection, Maintenance Team inspection (MTI), emergency exercise evaluation, NRR i licensing questions, Security inspection (Tampering), and Allei;;ation follow up (Firewatch

, requirement, resume falsification, tampering, and fitness for duty). Responded to l abnormal and emergency events. During the first year of assignment, with the absence i of an assigned Senior Resident inspector (SRI), I acted as the SRI for a period of three

] .

months and interfaced with the licensee on issues pertaining to plant oprations and 1- communicate objectively to the licensee on any safety concems. During iAay of 1988, as j a representative for the NRC, I had the opportunity to tour two Japanese Nuclear j Facilities (Mihama Unit 1, Tsuruga Units 1 and 2) and presented insights on Japanese I Nuclear Power Plants to the Regional Staff and fellow Resident inspectors.

[ October 1986 to Emergency Preparedness Analyst / Inspector. U. S. Nuclear Regulatory i September 1987 Commission, Region ill, Glen Ellyn, Illinois.

, inspected licensee's Emergency Preparedness Program. Performed routine and j emergency exercise inspections as a Team Leader and as a team member for the f following sites; Big Rock Point, Clinton, Fermi 2, Braidwood, Byron, Point Beach, Duane

Amold, Quad Cities, Callaway, LaSalle, Zion, and two fuel facilities at Hemitite, Mo.

i (Combustion Engineering) and Metropolis, IL. (Allied Chemical). As a lead inspector, i prepared inspection reports and assured that all findings were effectively communicated

[ to the licensee and the NRC management in a clear and timely manner. Reviewed and evaluated radiological emergency plan submittal for six assigned facilities (four power plants and two fuel facilities). Served as primary liaison on emergency preparedness matter for the assigned and several unassigned facilities to provide a focal point for responding to site specific inquires and problems. Other duties include; documentation of inspection and exercise evaluation findings, preparation of inspection reports to ,

substantiate the degree of licensee compliance, identify significant safety issues and recommend appropriated enforcement actions. Participated as an evaluator in the 1987 f I

Full Field Exercise (FFE) at the Zion Nuclear Power Plant.

i Fcbruary 1985 Westinghouse Instructor Engineer. Westinghouse Electric Corporation, to October 1986 Zion, Illinois.

Engineer within the Training & Operational Services group. Certified as a Westinghouse 4

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{ Senior Reactor Operator (SRO), SNUPPS. Successfully completed the Westinghouse

Training Program which included nuclear principles, plant system , integrated plant operations training, and SRO certification program. Completed Classroom and

Simulator Instructor Skills and Management course, and certified as a Westinghouse Training Instructor. Participated as simulator instructor for the D. C. Cook operator

! requalification. Developed Computer Aided Instruction (cal) lesson plans for nuclear j operator training. Participated as Core Physics Monitor for Indian Point 2 refueling.

i Completed Post Accident Sampling System (PASS) training, specifically on the Zion ,

l Nuclear Plant High Radiation Sampling System (HRSS). Conducted development work  !

! on Emergency Preparedness Program and Drill Scenarios for Omaha Public Power l j District (OPPD) Fort Calhoun Nuclear Power Facility. Performed Emergency J

. Preparedness training for OPPD corporate staff. I was also one of the forerunners of the 1

Westinghouse / Hydro Nuclear Ten Drill Emergency Preparedness Package. Prior to l j leaving Westinghouse, I was involved in Technical Specification Surveillance test I

scheduling and expediting for the Braidwood Nuclear Power Station startup. )

j Fchruary 1983 Lieutenant, Communications Officer / CMS Custodian. U. S. Navy, j to February 1985 USS Puffer (SSN 652), Pearl Harbor, Hawall.

j Maintained ship's communication readiness, responsible for operation and maintenance i of electronic communications equipment and associated classified material. Maintained l and upgraded top secret classified material as well as technical manuals as the 1 Communications Security Material System (CMS) Custodian. Responsible for operation

, and maintenance of the Nuclear Propulsion Engineering Plant as Engineering Officer of F

the Watch (EOOW). Organized and processed Quality Assurance, ship's emergency pre-overhaul surveillance testing, engineering tagout, preventive maintenance and IMA/ Shipyard maintenance work. Eamed overall grade of excellent on all operational inspections including Communications Readiness Exam. Maintained all elements of the r,MS account resulting in a rating of outstanding during a Naval Security Group CMS inspection. As Communications Division Officer, supervised 6 personnelin operations / maintenance of the ship's communications equipment and readiness.

Received a Letter of Commendation for outstanding performance of duty from the Commander Submarine Force, U.S. Pacific Fleet.

N5vember 1981 Lieutenant (JG), Electrical and interior Communications (l&C) to February 1983 Officer. U. S. Navy, USS Puffer (SSN 652), Pearl Harbor, Hawall. l Responsible for operation and maintenance of the Nuclear Propulsion Engineering Plant  ;

as EOOW. As Electrical Division Officer, supervised 20 personnelin operations / maintenance of the ship's electrical power generation, distribution, and I&C systems. Also, responsible for ship's safety and electrical safety programs.

Reorganized and developed Division Training Program. Engineered the overall upgrading of Divisional Preventive Maintenance System (PMS), after I identified and conducted an investigation of falsification of ship's PMS records. Organized, reviewed and processed Quality Assurance, tagouts, PMS, and IMA/ Shipyard maintenance work.

Conducted troubleshooting diagnostic analysis of electrical and I&C equipment with use of technical manuals, drawings, and troubleshooting guides. Received a Letter of s

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Commendation for outstanding performance of duty in respect to technical knowledge, organizational ability, and leadership from the Commander Submarine Force Squadron One.

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