ML20238D291
| ML20238D291 | |
| Person / Time | |
|---|---|
| Site: | 05560755 |
| Issue date: | 10/09/1987 |
| From: | Norris B, David Silk NRC COMMISSION (OCM), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20238D254 | List: |
| References | |
| SP, NUDOCS 8801040187 | |
| Download: ML20238D291 (25) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ADMINISTRATIVE JUDGE In the Matter of
)
)
ALFRED J. MORABITO
)
Docket No.
55-60755
)
(Senior Operator License for
)
Beaver Valley Nuclear Power
)
Station, Unit 1
)
.b<
JOINT AFFIDAVIT OF BARRY S. NORRIS AND DAVID M. SILK Barry 3. Norris and David M. Silk do' depose and say:
1.
I, Barry S.
- Norris, am a
Senior Operations Engineer (Examiner / Inspector)', in the Pressurized Water Reactor Section, Op-erations Branch, Division of Reactor Safety at the United States Nu-clear Regulatory Commission, Region 1.
At the time of the examination of Alfred J. Morabito my position was Reactor Engineer (Examiner) within Section 1C of the Division of Reactor Projects.
My responsibilities relative to that examination were as the certified examiner observing David M.
Silk's administration of the simulator and oral examinations.
I was responsible for the quality assurance review of the grading of the written examination, and i assisted in proctoring that examination.
My professional qualifications are attached.
2.
1, David M. Silk, am an Operations Engineer (Examiner / Inspector) in the Pressurized Water Reactor Section, Operations Branch, Division
. of Reactor Safety at the United States Nuclear Regulatory Commis-sion, Region I.
At the time of the examination of Alfred J. Morabito l
8801040187 871009 l
SECY LIC55 055607 5 l
l
. my position was Reactor Engineer (Examiner) within Section 1C of the Division of Reactor Projects.
My responsibilities relative to that examination were that i prepared, administered, and graded the written examination and administered the oral and simulator examina-tions under the observation of Barry S. Norris.
My professional qualifications are attached.
3.
The NRC examination for operator candidates under 10 C.F.R., Part 55, consists of a written examination, a simulator examination, and an oral examination.
Written examinations are prepared by an NRC examiner in accordance with NUREG-1021, " Operator Licensing Exam-Iner Standards," and NUREG-1122 "Knowledges and Abilities Catalog for Nuclear Power Plant Operators," and based upon facility IIcensee reference material.
Each of the written examination questions admin-istered to car.didates at the Beaver Valley Plant Unit 1 on July 22, 1986 was checked by a second examiner and the Section Chief for technical accuracy, adequate subject coverage, relevance to safe op-eration of a nuclear power plant, content validity, and to ensure regulatory requirements are satisfied.
After the administration of the examination, the facility ilcensee reviewed the examination for I
technical accuracy and recommended changes or additions to the an-
]
l swer key based upon the plant-specific features of the facility.
1 These comments were resolved to the mutual understanding of the NRC and tucility licensee.
The grading of the set of examinations l
was checked by a second examiner and the Section Chief to ensure consistency and accuracy in grading.
To receive a passing grade, l
the candidate must attain an overall score of 80%, and receive at l
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_..-_.___________-.____a
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! I least a 70% in each of the four sections of *.he written examination.
1 Mr. Morabito was tested for a Senior Reactor Operator license as an l
"SRO l'ns' tant", which is a candidate wh'o has not previously been licensed as a
reactor operator at that facility.
Although Mr. Morabito received an overall score in excess of 80%, he received less than a 70% on Section Six of the written examination.
4.
In addition to the written phase, the examination consists of an oral walk-through of the plant with the candidate and a simulator exami-nation.
Mr. Morabito passed the oral walk-through and failed the simulator examination.
5.
Simulator examinations are developed by the NRC to examine usually three candidates simultaneously during three or four different sce-narios, on a plant specific simulator. Each candidate is evaluated by an examiner, observing the candidate to determine whether candidate responses are satisfactory, marginal or unsatisfactory in eight areas of competency.
Based on the candidate's performance, an unsatis-factory evaluation in any one competency area may be considereG an adequate basis for falling the simulator examination.
All unsatisfac-tory evaluations made on the simulator examination require written justification.
Mr. Morabito received an unsatisfactory evaluation in four of the eight competencies.
6.
Although Mr. Morabito provided comments relative to sections of the written examhvions that he passed, the staff has only analyzed Mr. Morabito's answers, relative to Section 6 as this is the only sec-tion in the written examination that he failed.
The question, answer key, candidate response and argument, and staff evaluation follow:
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The Written Examination Section 6:
7.
Question 6.03b: What three design features of the component cooling l
water system minimize the effects of a rupture of the RCP thermal barrier:
(Point Value: 1.5; Grade: 0.5).
Answer Key:
High flow will cause RCP thermal barrier CCR. outlet valves to close; pressure buildup will seat check valve; piping be-tween valves is designed for 2485 psig.
8.
Candidate Answer The answer provided on the day of the examination was:
"1)' Automatic trip close of thermal barrier CCR-Inlet valves on high flow, 2) ability to manually isolate the thermal barriers, 3) auto backup pump start on decreasing CCR pressure, 4) separation of:
thermal barrier CCR supply from pump motor and tube oil cooling supply."
Mr. Morabito argues in the Specification that his responses 2) and 4) are among several correct answers but that the examiner's key listed-only three design features.
9.
Staff Response The candidate's first response ' includes incorrect information about the proper location (inlet versus outlet) of the automatic valves with respect to the thermal barrier heat exchanger.
The candidate was given credit for this response even though it contained incorrect information.
10.
The candidate's second answer that the thermal barrier can be manu-ally isolated does not answer the question.
The capability to manu-ally isolate the component does not, in and of itself, minimize the
. edects of a rupture of the thermal barrier.
A containment entry, l'
which'is a lengthy and involved process, would be required to iso-late the c'omponent and would not provide immediate reduction in the severity of the rupture.
11.
Isolation valves' are normally designed to allow maintenance on a com-ponent and not to minimize the effects of a component failure.
- 12. The candidate's third response is incorrect and shows a significant conceptual error on the part of the candidate.
If a thermal barrier rupture occurs,- the reactor coolant (being at a much higher pres-sure, 2235 psig) would leak into the component cooling water system thus increasing its pressure.
The auto backup pump starts on de-creasing system pressure not on increasing pressure.
- 13. The candidate's fourth response concerns separation of the thermal barrier cooling water supply.
The candidate supplied no documen-tation to support his answer.
A literature search and discussions with technical experts conducted by the NRC staff did not identify any written documentation supporting the candidate's position.
The Beaver Valley Training Department did not recommend any changes to the answer key during the examination review indicating that they expected that a candidate who had completed their training program would respond to the question with the information in the answer key.
14.
However, discussions with engineers at the Westinghouse Reactor Coolant Pump Engineering Group, the Pressurized Water Reactor Sec-tion at the NRC Technical Training Center, and the NRR Plant Sys-tem Branch indicates that the separation of the thermal barrier l
1 J
' component. cooling water supply from the pump motor and the lube oil cooling supply was a consideration in the design.
Based on these discussions' the candidate was given credit for his fourth response.
15.
The credit assigned to the candidate's responses was determined to be 0.75 points based on only two of four correct responses being provided.
This value was determined by taking 2he total number of correct answers divided by the total number of answers supplied.
Thus, although it was initially determined that a candidate giving three design features would result in giving full credit, a candidate who listed 4 or 5 would be assigned a proportional value of the per-centage of the number correct.
This method of determining the as-signment of credit to a question is routinely used by examiners to account for " shot gun" answers where incorrect information and in-discriminate guessing is listed along with responsive information.
i 16.
Question 6.06a:
What is used to control RCS pressure during cold solid plant operations?
(Point Value:
0.4; Grade: 0.)
Answer Key:
Letdown pressure control valve.
17.
Candidate Answer j
The answer provided on the day of the examination was:
"The overpressure protection system. Two PORV's keyed to A loop and C loop WR pressure transmitters." Mr. Morabito argues that the ques-l tion did not ask for " normal control" of RCS pressure but only for i
" control" and also that the word "what" in the question is very am-biguous.
He asserts that his answer is correct.
________-_A
- 18.
Staff Response The candidate's response that the over pressure protection system provides control of the reactor coolant system pressure during cold plant operation is incorrect.
The over pressure protection system provides over pressure protection and does not provide pressure control.
During cold solid plant operations pressure is adjusted and controlled using the letdown pressure control valve.
19.
A safe operator would not be expected to misinterpret the question since he should understand the difference between control systems and protection systems.
The staff does not consider the word "what" to be ambiguous.
The Beaver Valley Training Department did not recommend any changes to the answer key during the exami-nation review, indicating that they expected that a candidate who completed their training program would respond to the question with the information in the answer key.
No credit was given to this answer.
20.
Question 6.06b:
What three plant conditions provide inputs to the interlocks associated with RHR suction valve MOV-RH-701? Setpoints are required.
(Value:
1.5; Grade 1.5 by examiner, 1.3 by regrade. )
Answer Key:
Will not open at RCS pressure greater than 430 psig.
Will auto close at RCS pressure greater than 630 psig. Will not open if pressurizer vapor temperature greater than 475 F.
21.
Candidate Answer i
The answer provided on the day of the examination was:
"Pressur-Izer temperature less than 470 degrees.
A loop wide range pressure less than 430 psig.
A loop wide range pressure more than 630
L L psig. " -.This answer was not appealed for NRR review.
For the pressurizer temperature setpoint the examiner wrote in,the correct setpoint number, 475, but did not delete points.
The regrader de-l leted 0.2 points for the incorrect number.
Mr. Morabito argues this i
1 regrade was petty and vindictive.
l 22.
Staff Response The candidate's temperature set point was incorrect and partial cred-it was deducted during the regrade because an incorrect knowledge of the set point would not allow the candidate to recognize when the
' interlock should be activated, and since the set point is not ad-dressed directly in the procedure, could result in inappropriate valve positioning.
During -an appeal, the objective is to ensure that the candidate was graded fairly and correctly.
Points are routinely deducted during the regrading process where', as here, the staff feels that a candidate initially was awarded points inappropriately.
Examiners unassociated with the initial testing and grading are used during the regrading process to avoid even the appearance of vin-dictive conduct on the part of the challenged examiner.
- 23. Question 6.07a:
Why is the operability of the steam generator code safety valves important during power operation?
(Point Valve: 0.5:
Crade: 0.)
Answer Key: Ensures that secondary system pressure will be limited to within its design pressure during the most severe transient.
24.
Candidate Answer
]
The answer provided on the day of the examination was:
"They provide the first means of protection for Tavg increases above
. i
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program to prevent violating the safety ilmit curve."
This answer was not appealed to NRR.
Mr. Morabito asserts his answer is cor-
)
rect based 'on Exhibit K to his Specification (Duquesne Light Compa-ny Nuclear Division Training Manual),
- p. 44 14 and Attachment 1.
The section referenced states:
" Steam generator safety valves provide protection from exceed-ing [ reactor core safety limit of 2250 psla]."
25.
Staff Response The candidate's response that the steam generator code safety valves are the first means of protection for the T average increases above the program value is incorrect.
The normal first response to a high T average is control rod motion.
In addition, Exhibit K, which is a lesson plan from 'a requalification program conducted in 1984, ex-plains five plant design features which provide T average protection before operation of the steam generator safety valves; additionally Exhibit K states explicitly that T average protection is only provided by the steam generator safety valves at low power levels.
The ref-crence to Exhibit K does not support the answer provided by the candidate because it indicates that the steam generator safety valves provide a secondary or tertiary protection for T average in a limited range of power levels.
26.
The Beaver Valley Training Department did not recommend any changes to the answer key during the examination review, indicating that they ex'pected that a candidate who had completed their training program would respond to the question with the information in the l
answer key, No credit was given to this answer.
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- 27. Question 6.07b:
Give two reasons (NOT CONDITIONS) why the MSIV's are required to close during a steam line rupture.
(Point
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Value: 1.0; Grade: 0 by examiner; question deleted after regrade;
. question reinstated by NRR and 0.5 points given.)
Answer Key:
- 1. Minimize positive reactivity effects of RCS cooldown associated with the blowdown.
- 2. Limit pressure rise within contain-ment during a steam break in containment.
28.
Candidate Answer The answer provided the date of the examination was: "1) to isolate the faulted steam generator; 2) to prevent blowdown of the non-faulted steam generators through the break."
- 29. Mr. Morabito argues that, contrary to the assessment of the NRC staff, his answers are not redundant, and refers to Exhibit M (BVPS Updated Final Safety Analysis Report, pp. 10.3-2 and 10.3-5) as support.
He also describes an accident in which one main steam header ruptures between the main steam isolation valve and the non-return valve to demonstrate that his two answers are different ones.
30.
Staff Response l
The question asks for two reasons the MSIV's are required to close during a main steam line break.
The candidate gives two answers that he states are different.
In fact, in both cases, whether the MSIV shuts as the primary isolator of a leak or as the backup to a non return valve, it is still shutting to isolate the leak, and both answers are essentially the same.
t 4
.. 31 Question 6.09a: '
What two simultaneous conditions will cause the i
quench spray flow cut-back valves (MOV-1 QS-103 A, B) to close?
(Point Value: 0.8; Grade 0.4. )
Answer Key:
Associated quench spray pump running; RWST low low level.
32.
Candidate Answer
'The answer provided the day of the examination was:
"One hour running time elapsed, RWST level at setpoint."
- 33. Mr; Morabito asserts that the NRC answer to this question is incor-rect and that the. p' roof that his answer is correct is shown by Ex-hibit N
(1977 Beaver Valley, Unit 1
" Analysis and System Modification for Recirculation Spray and Low Head Safety injection Pumps Net Position Suction Head").
34.
Staff Response The answer in the answer key is taken directly from the Beaver Val-ley Operating Manual Chapter 13.1.
The Beaver Valley Training De-partment provided no corrections or additions to the answer key.
The candidate in his specification does not address the conditions for which the cut-back valves close and inappropriately references the NRC answer to Question 6.09b.
Exhibit N provides design basis in-formation but does not address the interlocks on the cut-back valves.
The candidate has not provided any additional information which would support changing the grading of. his answer for Ques-tion 6.09a.
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- 35. The candidate's arguments appear to be directed to question 6.09b for which he was given full credit during the review of his examination by Region I.
- 36. As a result of staff evaluation and a regrade of the written portion of Mr. Morabito's exam during the Administrative Appeal the candi-date's point total for Section 6 was increased from 13.5 to 15.4 points.
As a result of the latest regrade the point total for Ques-tion 6.03b was increased by 0.25 points from 0.5 to 0.75 points.
No other changes were made to the point totals of any other question.
The candidate's score is therefore (15.4 + 0.25)/22.6 = 69.2%.
Mr.
Morabito still falls Section 6 and thus fails the written portion of the examination.
37.
The Simulator Examination:
The three scenarios on which Mr. Morabito and two other candidates were tested were:
(1) a boron dilution accident followed sequentially by a vacuum breaker leak, pressurizer reference signal failure, er-ratic governor valve control, and station blackout; (2) the B spray valve failed closed and the Tave (temperature average) input to steam dumps failed low followed sequentially by B loop FRV (feed-water regulating valve) bypass falls open, loop 3 Th (hot leg tem-perature) falls high, turbine governor valves fall closed, PORV block valve failed open for 455D due to a breaker problem and Pzr (pressurizer) PORV 445D failed open; and (3) PRNi (power range nuclear instrument) (44) fails high followed sequentially by S/G (steam generator) tube leak (developing into a rupture), pzr level
.. failed high, FWP-1 A trip from 75% power, and "A" HHSI (high head safety injection) pump falls to auto start on low pressure.51.
38.
The object of the scenarios is to test the competencies of the candi-dates in eight areas during (a) nor aal evolutions (b) instrument failures (c) component malfunctions and (d) major transients in ac-cordance with N UREG-1021, ES302.
The scenarios are developed based on information from the licensee concerning the design of the plant as well as the capabilities of the simulator.
The Staff develops the scenarios to reflect actual or anticipated events.
The scenarios are designed to last about an hour each.
The four competencies Mr. Morabito failed are discussed as follows.
Compliance /Use of Procedures 39.
Examiner's Comment No.1 During the first scenario Mr. Morabito did not consult any procedure when decreasing load to check power range indicator response of two power range indications that were lower than the other two. AOP-10 calls for the plant to be in Mode 3 if two power range channels are malfunctioning.
40.
Answer Mr. Morabito states he did not consider two power range channels inoperable and ordered a small power reduction, for which no proce-dure is required, to determine whether the instruments were faulty.
Having determined the instruments were not faulty, Mr. Morabito halted the power reduction.
. 41.
Staff Response The candidate ordered a 10% load decrease to investigate a suspected instrumentation problem.
The candidate, in effect, performed an unauthorized test of instrumentation accuracy.
If an instrument mal-function was suspected the candidate should have referenced an ab-normal operating procedure or requested a surveillance procedure be conducted to verify the accuracy of the instruments.
A basic tenet of safe operations is to believe all indications unless they can be proven incorrect by prescribed procedures.
42.
Because Mr. Morabito did not use the proper method to verify the instrumentation, the symptoms of the dilution accident were masked by the decrease in load and he was unable to identify the accident in progress.
Therefore, he was not able to identify a condition which was changing the reactivity of the reactor in an unsafe direction.
43.
Examiner's Comment No. 2 During the second scenario while in ES-1.2 step 27 candidate asked "Are RCS hot leg temperatures greater than 395 F?" Candidate did not wait for an operator response and assumed the answer to the question was "yes" by answering "yes" aloud to himself. The opera-tor then indicated the answer to the question was "no."
44.
Answer Mr. Morabito asserts that the action taken was correct and his re-mark prior to the operator's answer is insignificant.
45.
Staff Response The candidate provided no information to refute the observation of the examiner.
A review of the candidate's examination and appeal
comments indicates that during the operating examination he incorrectly assumed reactor coolant hot leg temperature was greater than 395 'F.
The candidate's comments indicate an incorrect analy-sis of the effects of the transient in progress and an improper at-tempt to analyze an event instead of using the symptomatic approach provided by the emergency procedures.
46.
Examiner's Comment No. 3 Mr.
Morabito asserts this comment was deleted during the regrade / review process.
Thus there appears to be no issue regard-Ing this comment.
47.
Staff Response No new information was provided by the candidate.
48.
Examiner's Comment No. 4 During third scenario, after the reactor tripped and Si actuated, I
candidate did not check if LHSi pumps were running as required by immediate action step 11b of E-0, SRO had to remind candidate to i
check if LHSi pumps were running.
49.
Answer Mr. Morabito admits the comment is valid but asserts that it is not sufficient to support a grade of unsatisfactory for this category.
50.
Staff Response The candidate agrees this comment is valid.
Immediate actions are required to be committed to memory and are to be performed without I
reference to the procedure.
Proper performance of the immediate actions of the emergency operating procedures is a mandatory rc -
quirement for a safe operator.
The candidate's inability to perform this step is significant since this was the only evaluation made of the t.
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- candidate's ability to properly perform the required immediate actions of the emergency procedures as a control board operator.
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Control Board Operations (third scenario only) 51.
Examiner's Comment No.1 Following Si actuation as the RCS pressure was decreasing the can-didate misread RCS wide range pressure indication.
Candidate mis-read 1600 psig as 1040 psig and then checked with other operator to confirm RCP trip criteria.
52.
Answer Mr. Morabito asserts that his actions maintained the plant within safety limits and he performed the complex procedure for securing the reactor coolant pumps flawlessly.
53.
Staff Response A review of the candidate's examination and appeal comments indi-cates that he incorrectly tripped reactor coolant pumps due to mis-reading the reactor coolant system pressure instrument.
The fact that he maintained plant safety limits is not relevant because the candidate was being evaluated on his ability to properly read process instrumentation.
Additionally, the fact that he performed an evolu-tion correctly is immaterial considering that the evolution should not have been performed.
The candidate's tripping of the Reactor. Cool-ant Pumps because he misread reactor coolant pressure clearly dem-onstrated his inability to read an important process instrument.
54.
Examiner's Comment No. 2 in Step 4c of E-3, the residual heat release valve was to be checked to ensure !t was closed.
Candidate was looking at the demand
. Indicator for the manual control of residual neat release valve and not at the indication lights for the valve. Candidate was hesitant to respond t'o the check verification and appeared confused until other operator came over and explained the controls and Indications to the candidate.
55.
Answer Mr. Morabito asserts that he was not confused but only hesitant to verify the position of the valve based on the demand indication for the valve alone, and that the conference with the balance of plant operator helped.
He asserts that it is significant that the examiner mistakenly thought there was position indicating lights.
56.
Staff Response A review of the candidate's examination and appeal comments indi-cates that the candidate hesitated and conducted a discussion with another operator prior to completing a verification step in the emer-gency procedures.
The hesitancy and discussion with another indi-vidual indicates a lack of familiarity with control board layout and the policy for completion of the emergency procedure verification step.
- 57. The lack of an Indication light is not relevant because the candidate was being evaluated on his knowledge of the control board indica-tions and his ability to complete a verification step without supervi-sion.
The fact that the candidate did not immediately realize that the demand signal was the only method for verification of valve po-sition supports the unsatisfactory evaluation of his control board operations.
s
. 58.
Examiner's Comment No. 3 in Step 9 of E-3, the containment sump pumps were to.be stopped.
Candidate' stopped one containment-sump pump and the incore instru-ment sump pump.
The other operator came over to show the candi-date where the other containment sump pump switch was located.
59.
Answer Mr. Morabito asserts that even though he turned the wrong switch, he would have discovered ~ his mistake without the other operator.
He also states that his action was precautionary since there is no reason to expect that the containment sump pump will start when the safety injection or containment isolation signals are reset.
He also offers Exhibit U (Step Description Table for E-3) to show that he missed only one of eight actions prescribed.
- 60. Staff Response A review of the candidate's examination and appeal comments indi-cates he did operate an incorrect switch and his mistake was identi-fled by another operator.
Even though the candidate's actions did not degrade plant. conditions, the ~ fact that he incorrectly reposi-tioned a wrong switch without noting his mistake supports the unsat-isfactory evaluation of his control board awareness.
61.
The missing of a single step in a procedure that is being conducted from memory is significant.
The Beaver Valley Operations Depart-ment allows its operators to conduct control board operations without procedures when they are fully familiar with all the required steps.
The candidate, by attempting to conduct the step in the emergency procedure without reference to the procedure or supervisory
i 1
1
- l guidance, and missing a procedural step, was operating outside the I
i guidance set by his facility.
J 62.
Examiner's Comment No. 4 in Step 11 of E-3, CIA was to be reset.
Candidate depressed the CIA Train B button and the CIB Train A button.
CI A did not re-(
{
r set.
Candidate did not verify CIA was reset following his attempt to l
reset CIA.
63.
Answer Mr. Morabito states that he did reset the CIA at the end of the sce-nario, and that there is no indication in the control room of success-fut reset of CIA, and failure to properly reset CIA is not safety significant.
64.
Staff Response A review of the candidate's examination and appeal comments indi-cates that the candidate failed to properly reset CIA when directed by the shift supervisor.
Even though plant conditions were not de-graded, the fact that the candidate was unable to properly accom-plish a procedural step is indicative of his unsatisfactory ability to operate the control board.
Supervisory Ability 65.
Examiner's Comment No.1 in the second scenario, the candidate did not notice that the feed reg bypass valve indicator was indicating that the valve was open during diagnosis of unusual feed reg valve movement.
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- 66. Answer
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Mr. Morabito states that he was the first to notice the open valve; that there-was no alarm to call his attentiorr to a problem; the simu-I lator malfuncibned, but his actions were effectis e since control of B j
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steam generath !evel was maintained.
67.
Staff Response I
A review of the candidate's examinatiem, appeal comments and exam-l Iner's notes indicates that a problem with steam generator feed f5w
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was recognized by the candidate, but the root cause of tPu problem s
was not aggressively pursued and was not identified for twenty minutes.
The fact that there were no alarms to draw the candidate's attention to the incorrectly positioned valve is irrelevant sinc.c. 7 any problems encountered at nuclear, power plants do not initiate an alarm but must be identified from abnormul parameter readings.
(
68.
The malfunctionit.g of the simulator is irrelevant since the candidate was instructed,o respond to all indications as he would in the actual control roca.
A ban: tenet of nuclear power plant operations is to 4
believe cil indications unt!! they can be proven incorrect by other indications.
The candidate's belief that indications were incorrect without extensive verification of all other parameters is inconsistent with the safe operation of a nuclear power plant.
69.
Exami:^.er's Comment No. 2 Unsatisfactory use of procedures and unsatisfactory crew interaction
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supports an unsatisfactory performance in supervisory ability.
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- 70. Answer 7.r
' ihI Mr. Morabito claims the examiner's statement is unsupported and of-1 ;
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fers Exhibits B-1; V-X as evidence 6T his supervisory ability while employed at the Shippingport Atomi Power Station.
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- 71. Staff Responsh -
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I A review 'of the candidate's examination and appeal comments 'ihdi-cates that the car.didate's actions and communications observed dur -
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ing the simulator exaraination displayed a calidusness to procedures ll l
,t and less than precise communications.
For example, the candidate
]
[
' incorrectly enteredI the ' emergency operating procedures after he I
misdiagnosed a loss of all AC power, he conducted procedural steps from-memory which resulted in two errors, and he inappropriately '
]
allowed the use of a hand sigral for communication of radiation l
)
levels.
In one instance the candidate was not aware that his
)
3, reactor operator was having difficulty maintaining rod insertion 3
' limits and controlling Tave during a load reduction whkh was complicated by a dilution accident as evidenced by a low rod inser-tion limit alarm followed later by a loop Tave high alarm.
Thus the candidate offered no supervisory assistance to i bis operator.
Communications / Crew Interactions
- 72. Examiner's Comment No.1
'p s^
' During the first scenario, following the loss :of off>;lte power, the candidate went to ECA-0.0 when he mistakenly observed that he had no emergency busses energized.
Candidate should have relied upon verification of emergency busses from his operator who did properly l
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verify that one emergency bus was energized and informed the l
candidate as much.
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- 73. Answer Mr. Morabito argues that this comment should be deleted because there were no AC busses energized when he assessed the situation.
He claims that after the only available diesel generator was up to speed did it then close onto the bus and load.
He asserts that he observed the symptoms for loss of all AC power and directed perfor-mance of ECA-0.0 in accordance with his observation but that after receiving information from the operator, he properly directed transi-tion to E-0 and maintained safe plant conditions.
74.
Staff Response A review of the candidate's examination and appeal comments indi-cates that during a loss of off site power where a diesel generator started and loaded as designed, the candidate entered the emergency operating procedures incorrectly in that he went to ECA-0.0, Loss of All AC Power, instead of E-0, Reactor Trip / Safety-injection.
This is because he did not request verification by the board opera-tor.
Mr. Morabito failed to recognize the diesel generator does not load for approximately 10 seconds after loss of offsite power.
This oversight was compounded by his failure to obtain in forma-tion from the operator.
Senior operators are responsible for using all available sources of informat!on to ensure proper a c'tlon.
75.
Examiner's Comment No. 2 During the second scenario, Step 6 of E-1 calls for checking secon-dary radiation levels.
An operator checked the monitors and said
] 1 "One indicator is about this much (holding fingers about 1/2 to 1 i
inch apart) higher than normal. "
Candidate proceeded in E-1.
During followup questioning after the scenario the candidate admitted misunderstanding the operator's report of recondary radiation levels and assumed there was no reason to go to E-3.
76.
Answer Mr. Morabit'o claims that he understood the hand signal given; but that he [Mr. Morabito] spoke aloud and used no hand signals. He claims that the NRC staff's prior comment (Exhibit L) that hand sig-nals are unsatisfactory is biased Judgment; and that his action after receiving the hand signal was proper.
He offers Exhibit Y to show that the operator believed Mr. Morabito understood the hand signal at the time.
77.
Staff Response A review of the candidate's examination and appeal comments indi-cates that the candidate relied on hand signals to reach a decision whether secondary radiation was consistent with pre-event levels.
Hand signals are not a reliable means of communications in the control room of a nuclear power plant.
The examiner's comment regarding Mr. Morabito's misunderstanding the hand signal is based on a verbal statement by the candidate.
78.
Examiner's Comment No. 3 During second scenario, while in ES-1.2 step 27, the candidate asked, "Are RCS hot leg temperatures greater than 395oF?"
Candi-date did not wait for an operator resoonse and assumed the answer
l 6
1 l to the question was "yes" and answered "yes" aloud to himself. The I
operator then indicated the answer to the question was "no."
79.
Answer Mr. Morabito states that it is only actions, not words on which he
]
l should be judged; that he should not be criticized for thinking aloud q
since his performance was based on the Information from the opera-tor, not his advance expectation.
80.
Staff Response A review of the candidate's examination and appeal responses indi-cates that the candidate verbalized an incorrect assumption prior to receiving the correct information from the control board operator.
The candidate's comment indicates an incorrect analysis of the effects of the transient in progress and an improper anticipation of plant parameters based on an attempt to analyze an event.
81.
We have reviewed the arguments in the specification at pp. 24-29 concerning the NRR review comments on the simulator examination.
We believe these arguments are repetitions of the arguments ad-dressed above.
Therefore, we have not addressed them separately.
82.
Summary Based on the previous discussion, and for the reasons explained, it is our opinion that Mr. Morabito's performance on both the written and simulator examinations was not adequate to receive a passing grade and that there is no basis to change the falling grade or to issue him an SRO license.
OCT 09 '87 07841 NRC. KING OF PRUSSIA P02
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- 83. The foregoing and attached ' statements of' professional qualifications are true and correctsto the best of our knowledge and belief.
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O Barry S. Norris Senior Operations Engineer EXPERIENT:
U. S. Nuclear Regulatory Cm mission King of Prussia, Pennsylvania 19406 August 1987 - Present Senior Operations Engineer (previously Lead Reactor Engineer).
Coordination of all activities associated with the examination process for ali Region i Babcock & Wilcox and Cmbustion Engineer-ing facilities. and all nonpmer reactor facilities.
Examination of applicants as detailed below.
Certified as an MC Examiner on Cm bustion Engineering technology in accordance with MREG 1021.
May 1985 - August 1987 Reactor Engineer (Examiner)
Examination of applicants for Reactor Operator and Senior Reactor Operator Iicenses in accordance with 10 CFR 55; process includes the preparation, administration and grading of
- written, sinulator and oral examinations.
Certified as an W C Examiner on Westinghouse, Babcock
& Wilcox, and nonpmer technologies in accordance with NLREE 1021.
August 1981 - May 1985 Houston Lighting & Pour,' Co.
South Texas Project, Bay City, Texas 77414 Senior Quality Assurance Specialist-Systems Coordination of the client's Project @
audit and training /certi fication programs, including preparation and conduct of training classes. Assist in the development and review of HLSP's and contractors' @
programs and procedures.
Certi fled Lead Auditor per ANSI N45.2.23/RC 1.146.
Barry S. Norris 2
June' 1980 - August 1982 GlIbert/Canmonweaith, Inc.
Reading, Pennsylvania 19603
~
Ouality Assurance Specialist-Electrical Develop and review procedures and technical spect f! cations for instrumentation / control and fire signaling systems.
Coordination, preparation, and presentation of @/QC courses for both internal training and client organizations _ Foundations of @/QC, Basic F%R/EMR Technology, Fire Protection,
- Auditor, and instructional Techniques.
Consulted with clients for establishment and implementation of quality assurance programs.
Certifled as Auditor per. ANSI N45.2.34/E 1.146, and level li Electrical Inspector per ANSI N45.2.6/E 1.58.
April 1977 _May 1980 U. S. Navy Nuclear Power Training thit - 836-Ballston Spa, New York 12020 Training Coordinator Manage the training of thirty students conmencing every three months; scheduling
- watches, maintenance of records and performance appraisals.
As Engineering Watch Supervisor, directed propulsion plant watchstanders during all normal and casualty operations; and radiological situations.
Reactor Controls Division Leading Petty Officer Supervised seven staff instructors in preventive and corretive maintenance of reactor controls and instrtmentat ion.
instructed students, both in the classroom and on an operating reactor plant, in reactor theory and operations. Directed the qualification e f forts of reactor operator students.
As a Reactor Operator, was directly responsible for safety of the reactor plant during operating and 'shutdom conditions.
__m
Barry S. Norris 3
May 1974 - March 1977 USS Sargo (SS(N) 583)
Pearl Harbor, Hawall 96601
~
Reactor Operator s Engineer's Administrative f
Assistant J
Stood watch as Reactor Operator during i
operating and shutdown conditions on a shipboard reactor plant.
Maintained train-ing and technical records for engineering department.
Engineer's assistant during a major shipyard
- overhaul, during. which runerous technical modifications and test were performed and all watchstanders were trained and requali fled.
August 1971 - April 1974 Naval Nuclear Program Trainee l
ASSOCIATIONS & CERTIFICATIONS:
1986 ffC High Quality Certificate 1983 American Nuclear Society (Marber) 1981 Anerican Society for Quality Control Treasurer of Iocal chapter Certifled Quality Engineer #E-8575 (12/81) 1981 U. S. Navai Reserve (Dri!iing)
EEUCATKN & TRAINitO:
1987 U. S. Nuclear Regulatory Cmmission Cmixist ion Engineering Systens Technology (80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> classroom and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> sirrulator)
Pre-Supervisory Orientation Course (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> setf-study & 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> classroam) 1986 - 1987 U. S. Nuclear Regulatory Cmmission Babcock & Wilcox Systens Technology (80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> c lassrocyn, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> sirrulator)
l s
Barry S. Norris 4
l l
EDLCATION & TRAINING cont.
1985 - 1986 U. S. Nuclear Regulatory Canmission Westinghouse Systems Technology (160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> classroom, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> sinulator) 1983 Houston Lighting & Pou r, Co.
South Texas Project.
Oral Presentation Skills (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> classroom) 1982 - 1985 University of Houston, Victoria Pursuing course of study towards f4R 1982 Dechtel Energy Corporation South Texas Project Field Welding Upgrade Course (60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> classroom) 1981 Gilbert /Cartmonwealth, Inc. Reading, PA Lead Auditor Training Course (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> classrocm) 1971 - 1974 Naval Electronics Technician School (10 months)
Naval W clear Power School (6 months)
Naval Nuclear Power ' Prototype (6 months) 1971 B.S.
Ursinus College Collegeville, Pennsylvania Major - Math 1
l l
A 4
l David M. Silk Operation Engineer (Examiner)
EXPERIENCE:
U. S. Nuclear Regulatory Cm mission King of Prussia, Pennsylvania 19406 August 1986 to Present Operations
- Engineer, Reactor
- Engineer, (Examiner)
As a certified Westinghouse examiner, served in the capacity of examiner or. chief examiner for examinations administered on seven examination trips.
Responsibilities included preparing, administering and grading of written, sinulator, and operating examinations, i
January 1986 to July 1986 Reactor Engineer (Examiner).
Prepared written examinations for three Westinghouse plants to meet qual i fication requirements.
Observed certified examiners conducting simulator and oral examinations.
Conducted simulator and oral examinations while under the observation of certified examiners.
Certified as an MC Examiner on Westinghouse technologies in accordance wIth MREG-1021.
August 1985 to Decmber 1985 Reactor Engineer (Examiner).
Prepared and administered written and oraI examinations for nonpower reactors.
Certified as an WC Examiner for nonpower technologies in accordance with NLREG-1021.
January 1985 to July 1985 Reactor Engineer (Examiner).
l On the job training at Sequoyah Nuclear Power Plant (TVA).
Observed the activities of Operations,
- Training, Maintenance, instrumentation and
- Control, Nuclear Engineering, Carpliance, Health Physics and Rad waste.
l January 1984 to Decmber 1984 Reactor Engineer (Examiner).
j indoctrination to WC.
4 I
1 WINING #O EDUCATION:
June 1987 - USMC Cmbustion Engineering Systen Technology (80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> classroom).
April 1987 - USMC Westinghouse Operator Licensing Refresher Sirrulator Course (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />).
July 1985 - October 1985 - USMC Westinghouse Systems Technology Series.
(160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> classroom, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> simulator).
October 1984 - USNIC RNR and EHR introductory Class (80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> classroom).
May 1984 - Graduated fram Pemsylvania State University with a Bachelor of Science degree in Nuclear Engineering.
a 80CKETED
.I USNRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 37 ET 14 P3 37 BEFORE THE ADMINISTRATIVE JUDGE TI Gk'Mk MAwi in the Matter of
)
)
ALFRED J. MORABITO
)
Docket No. 55-60755
)
(Senior Operator License for
)
Beaver Valley Nuclear Power
)
Station, Unit 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO SPECIFICA-TION OF CLAIMS" in the above-captioned proceeding have been served on the following by deposit in the' United States mall, first class, or as indl-cated by an asterisk through deposit in the Nuclear Regulatory Commis-sion's internal mail system, this 9th day of October,1987:
Charles Bechhoefer, Esq.
Alfred J. Morabito Administrative Judge 685 Tulip Drive Atomic Safety and Licensing Board New Brighton, PA 15066 U.S. Nuclear Regulatory Commission Washington, D.C.
20555*
Atomic Safety and Licensing Board Jay Gutierrez U.S. Nuclear Regulatory Commission Regional Counsel Washington, D.C.
20555*
USNRC, Region 1 631 Park Avenue Atomic Safety and Licensing Appeal King of Prussia, PA 19406*
Board Panel U.S. Nuclear Regulatory Commission Docketing and Service Section Washington, D.C.
20555*
Office of the Secretary U.S. Nuclear Regulatory Commission Dr. David L. Hetrick, Professor Washington, D.C.
20555*
Department of Nuclear and Energy Engineering University of Arizona Tucson, Arizona 85721 I
Colleen P. Woodhead Counsel for NRC Staff l
l
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