ML20197J272

Forwards Written Presentation Detailing Deficiencies &/Or Omissions in NRC Proposed Denial of Operating License for O'Hern & Provides Affidavit Affirming Statements.With Eight Oversize Drawings

ML20197J272
Person / Time
Site:	Fermi, 05532442
Issue date:	12/07/1998
From:	Ohern S AFFILIATION NOT ASSIGNED
To:	Bloch P, Cole R Atomic Safety and Licensing Board Panel
References
CON-#498-19792 99-753-01-SP, 99-753-1-SP, SP, NUDOCS 9812150029
Download: ML20197J272 (25)

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00CKETED

/N \T7 USHRC

\ December 7,1998

% DEC 11 PS :08 Peter B. Bloch, Presiding Officer Dr. Rich 9[d(CoM 15gM.F ~

Administrative Judge Administd;e[JtidNe -

Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Board Panel U.S. Nu' lear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 In the Matter of Shaun P. O'Hern (Denial of Reactor Operator License)

ASLBP No. 99-753-01-SP Docket No. 55-32442-SP

Dear Administrative Judges:

Pursuant to the Presiding Ofncer's Order of October 19,1998 (Order) and 10CFR

,o 2.1233, I am submitting the attached written presentation detailing the deficiencies and/or

() omissions in the Nuclear Regulatory Commission's proposed denial of my Operating L,icense.

Affidavit of Shaun P. O'Hern Shaun P. O'Hern, being duly sworn, deposes and says as follows:

I, SH AUN P. O'HERN, candidate for a Reactor Operator License for the Fermi-2 facility, hereby affirm under oath that the statements in the attached response are true and correct to the best of my knowledge and belief.

A Shaun P. O'Hern Subscribed and swom to before me the 7* of December,1998 a121Y '

.2uf otary Public My Commission expires: oo aV JUDITH A. RAY

/-) , Noisy Public, Ybyne Countf. MI

\s' liy Commesico Erpes Apr 4 2@2 9812150029 981207 PDR ADOCK 05000341 Q PDR C MO3 ;

O cc: Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Attention: Rulemakings and Adjudications Staff (plus two copies)

Office of Commission Appellate Adjudication U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Susan L. Uttal, Esq.

Office of the General Counsel Mail Stop - 015 B18 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 O-f ::

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1. Summary of written presentation l

j 2. Written Examination Question #7 with attachments:

6I721-2080-13, Visual Annunciator & Sequence Recorder Alarm Schematics

, Window Box HiIP603D533 l

l 6I721-2155-08, Schematic Diagram Reactor Protection System Trip System "A*

Scram Trip Logic 4

61721-2155-08A, Schematic Diagram RPS Trip System "A2" Scram Trip Logic 61721-2155-09, Schematic Diagram Reac. Prot. Sys. Trip System "B" Scram Trip Logic 61721-2155-09A, Schematic Diagram RPS Trip System "B2" Scram Trip Logic l

6I721-2155-10, Schematic Diagram Reac. Prot. Sys. Scram Solenoids O 29.100.01 SH 1 A Rev. 5, RPV Control - ATWS

3. Written Examination Question #54 with attachment:

29.100.01 SH 1 A Rev. 5, RPV Control - ATWS

4. Written Examination Question # 59

5. Written Examination Question # 87 l

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l (v) Summary of Written Presentation l

l I am contesting the Nuclear Regulatory Commission's (NRC) decision to deny my informal appeal regarding the grading of the written portion of my Reactor Operator's license examination administered on April 6,1998. I do not believe that the NRC fully understood the basis for my appeal of questions 7,54, and 87. I have provided l 1

clarification for the answers which I selected in regards to these three questions. l l

I will also demonstrate the reasons for my contentions that question 59 should be deleted from the examination. The overtuming of the denial of questions 7,54, and 87 plus the deletion of question 59 will result in a composite score of 78/94 = 83%. This final score will result in a satisfactory score for the written examination and issuance of my Reactor C)

V Operator's license.

l Most of the information andjustifications that have already been provided in my original l informal appeal HF Item 5 have been excluded from this presentation to minimize l duplication. I respectfully request that this information be reviewed along with the new information being provided.

l I would also like to bring to the attention of the presiding officer some facts from the Hearing File (HF) which are pertinent to my appeal.

The detailed analysis on HF Item 9 page 12. " Summary of Examination Question Changes" of my informal appeal shows a final score of 80.4% with a final O

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l recommendation to overturn the original denial of my Reactor Operator license It  ;

, should also be noted that two of the questions in concern, #54 & #87, were listed as ,

? l l l agreeing with my original contention. If this was the Appeal Board's final conclusion i l then I should have been issued a Reactor Operator's license. However, this Summary of l

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i Examination Question Changes from HF Item 9 different significantly from HF Item 10 page 10, which is the Region III final conclusion. Region III concluded that my final i

, score should be a 78,9E l \

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RG Ouestion 7 i

in this question the student is asked to identify the correct response to an Anticipated Transient Without Scram,(ATWS). This appeal position is the most complicated position to understand. I will first discuss why the NRC identified correct response, ( d.)

i is incorrect. Then I will explain why my answer, ( a.), is correct. Once these two separate issues are understood it will be easy to conclude that I should be given credit for

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l my response.

Answer (d.) is incorrect for the following reasons: The question stem states that l annunciators 3D73, Trip Actuators A1/A2 Tripped and 3D74 Trip Actuators Bl/B2 Tripped were received. The question also stated that in order to select a correct answer I that I needed to determine the first action that MUST be performed and which indication i

MUST be observed to verify proper response of the action. Answer (d.) stated that I 1

g should verify that the blue group scram lights are off as the correct answer for the V indication that MUST be observed to verify proper response of taking the Mode Switch l to SHUTDOWN. In order to verify that an action provided the desired response there has to be a change in some indication. I will prove that if the 3D73 and 3D74 annunciators actuated as stated in the question, that the blue group scram lights also go off.

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As a result answer (d.) is incorrect because if these lights are already off when the l

l Mode Switch is taken to SHUTDOWN then there would be no observable changes using I this indication to verify that the desired response of the action occurred.

We now need to refer to schematic diagram 61721-2080-13 to determine what will bring in alarms 3D73 and 3D74.

(It is unnecessary to review every K14 relay to show the development for 3D74 and the de-energizing of the group scram lights since they are identical in all cases except for the letter designators on the relays. However, all drawing have been included).

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N At location G8, C71-K14A and/or K14E or C71-K14C and/or K14G contacts must open 1

to bring in Alarm window 3D73, Trip Actuators A1/A2 Tripped.

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l Let's now move on to schematic 61721-2155-08 to see what relay would cause these contacts to open.

l At location D8, de-energizing relays K14A and K14E will cause the associated l contacts (mentioned above) to open. These relays would have been de-energized by the opening of the K12A or K12E contacts from the Neutron Monitoring System Trips (3D99 APRM Upscale Neutron / Thermal Trip actuated as stated in the question stem).

We now need to look at 61721-2155-10.

The pink line represents the normal electrical path used to keep the blue group scram lights on. As we can see, when any of the K14 relays de-energize, then power is lost to

() the blue scram lights.

Thus we can demonstrate, with absolute certainty, that if 3D73 and 3D74 are actuated then the blue scram lights have to be off. As previously stated I cannot verify that an action provided the desired response by observing something that has already occurred.

Knowing this, I determined that the second part of answer (d) was incorrect and therefore answer (d) could not be chosen. I then evaluated the other answers. Both myself and the NRC agree that choices ( b.) and ( c.) are incorrect. This leads to the evaluation of answer ( a.).

In the NRC conclusion and analysis (HF Item 10, page 3), they acknowledge that the stabilization of reactor pressure is allowed as a concurrent step. They are acknowledging the basis behind the implementation of the EOP's. Whenever a flowchart is entered, then all legs of that flowchart are to be entered and carried out concurrently. This is so stated in FSRC-1 at the top of 29.100.01 SH 1 A, RPV Control -

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\ ATWS EOP. Entering the pressure leg of the EOP's has the same priority as entering the power leg. Based on this knowledge, I realized that stabilizing pressure less than 1093 psig was the first real direction block in the pressure leg (see HF ltem 5). I also knew that the SRV's were allowed to be used per Table 4 of the procedure. As for the pressure band of <1050 psig is a reasonable number to be used by the SRO for l controlling pressure in this situation. It was below 1093 psig high enough to prevent a cooldown of the Reactor, which is clearly not allowed per procedure in this situation. It ,

also allows the Low-Low Set SR V . co control pressure freeing the operator to perform l

1 0:her functions (see HF Item 5). Thus, the first part of this answer meets one of my three immediate actions, (control level, pressure, and power). The second half of the question  ;

requires me to observe the pressure response on the Div 1 and Div 2 Post Accident Post Accident recorders. This is a reasonable requirement since these two recorders are designed to trend reactor pressure and reactor level following an accident and they are located in close proximity to the SRV controls.

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I now had an answer that was correct. With the knowledge that the other three answer were incorrect I chose ( a.).

I respectfully request that I be given credit for my response as a correct answer.

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RO Ouestion 54 I do not agree with the NRC Analysis and Conclusion as stated on page 7 of HF Item 10.

The NRC does not agree with my assertion that the information provided in the stem of the question can confirm Altemate Rod Insertion (ARI). The NRC has agreed that there was no loss of offsite power event, (HF Item 10 page 6). HF Item 25 states at the top of the page that the same parameters that cause both Reactor Recire Pumps to trip also cause the ARI valves to reposition. When developing a written examination, NUREG 1021 Appendix B. Item d discusses the purpose of the information placed in the stem of examination questions. Therefore, the candidate must assume that the information provided about the recirc pumps had some bearing on the answer being solicited. Also on page 7 of HF Item 10 i'ae NRC concurs that my hypothesis that a pressure transient may have caused an ATWS/ARI trip of the Recire Pumps. Thus, it is easy to see how I drew the conclusion that initiating ARI would not be my first action because it had already occurred.

We should now refer to HF Item 38, pages 5 and 6. In both the first and last paragraph of HF Item 10 page 7 the NRC references actions per FSQ-7. By referencing this step number in the context listed they are demanding that I memorize all of the steps of the EOP's. This is clearly above and beyond the scope required for any licensed candidate, HF Item 39, page 10 clearly states that only the Entry Conditions for the EOP's are required to be known by the Shift Team Members. HF Item 38, page 5, Objective 01-04 clearly states that we will determine appropriate actions of the EOP's using the l Flowcharts. These Flowcharts were not provided for our examination. This is an example of where I was being examined to a higher level then what is required to be a

! licensed operator.

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n l V However,it is still my contention that the Injection of the Standby Liquid Control (SLC) l system is the appropriate answer for the conditions presented. HF Item 43 provides the necessary direction for this action. It would also appear that the . Appeal Board originally agreed with my position. I am referring to HF Item 9 page 12. There the Appeal Board indicated that they would accept both r.nswer b and answer c. It is unclear to me as to why they reversed this correct decision.

I respectfully request that answer c be accepted as an alternate correct answer.

As an alternative, I would accept deleting this question from the examination based upon the second paragraph above.

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v) RO OUESTION 59 The question asks the examinee to determine which of a group of Tech Spec instruments l

is not required in Mode 2.

In my initial appeal I attempted to get credit for inadequacies in the Stem of the Question.

I selected the Level 1 input to the Automatic Depressurization System,(ADS), as the correct answer. This was based on this signal not being required below 150 psig. (The stem of the question did not give a value for Reactor Pressure). This position was not accepted during the appeal process.

10CFR55 lists the requirements for Operator and Senior Operator exams.

10CFR55.41 lists the requirements for the written examination of a reactor operator.

rs 10CFR55.43 lists the contents for the Senior Operator exam.10CFR55.43(b) in part

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states "The written examination for a senior operator for a facility will incNde a representative sample from among the following seven items and the 14 items specified in 55.41 of this part. ..."

10CFR55.43(b)(2) states the requirements for the Senior Operator to be examined on the facilities Technical Specifications and their bases. This requirement is not in part 55.41 for Reactor Operators.

Based on the above information it appears that the written test was examining the reactor operators to the standards for senior reactor operator as described in the Code of Federal .

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Requirements. This is clearly in excess of the legal requirements for a reactor operator.

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!p) o I respectfully request that this question be deleted from the examination.

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RO OUESTION 87 This question gives a set of conditions associated with a Reactor Recite seal and asks the examinee to determine the failure mode.

There are two answers that will satisfy the conditions given in the stem of the question.

This fact is not disputed by the NRC. In the NRC analysis and conclusion of my informal appeal they acknowledged this fact. They also acknowledged that I should be given credit for my answer,(see the last paragraph below). Their argument for rejecting my appeal consisted of two items:

1. The NRC contends that the absence of alarm 3D121,"RECIRC PMP A OUTER SEAL LEAKAGE HIGH", from the stem of the question should have been used for determining the correct answer. There are a couple of problems with this conclusion.

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• a. The examinee is suppose to use the information provided in the stem of the question. He is not suppose to infer additional information nor should he draw conclusions based upon the lack ofinformation. If the NRC wished to have the examinee use this lack of alarm when formulating his answer then it chould have been so stated in the stem of the question. i.e., they should have stated that this alarm was not alarming.

b. 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 made regarding the information in the stem of the question.

ps N 2. The NRC, in the final sentence of their analysis and conclusion, states that I did not provide anyjustification that the diagnosis of a seal failure requires seal temperature indication. Student Text ST-OP-315-00(M-001," Reactor Recirculation System" ,

discusses the effects of the plugging of a Seal Labyrinth. On page 14 of the Student Text (liF Item 44) it clearly describes the temperature increase that would be seen for a plugged labyrinth. If the Seal temperatures increased as stated above then 3D100, "RECIRC SYS COOLANTS TEMPERATURE HIGH" would be expected to ahtrm with the conditions given in the stem of the question. The stem of the question failed to address this alarm status.

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Based upon the above information either answer ( b.) or ( c.) is correct. The NRC does not dispute the technical validity of either answer. There conclusion was based 1

solely on the information lacking from the stem of the question. As we have seen, their l argument is without justification. l Also, in a discussion between the NRC and the licensees management and engineering representatives on or about August 6,1998 the licensee was asked for their official position and replied that answers (b.) and (c.) were correct for this question.

It is also vital that we examine HF Item 9 page 11. In this Detailed Analysis of my Informal Appeal the Appeal Panel clearly agrees with my position. They reinforced their position in my favor on page 12 of HF ltem 9. I cannot understand how the Appeal Panel can state 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 (HF hem 10, pages 8 and 10).

I respectfully request that I be credited for a correct response.

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