IR 05000341/2006301

Er 05000341-06-301(DRS); Fermi Power Plant, Unit 2; Initial License Examination Report

ML061210521
Person / Time
Site:	Fermi
Issue date:	05/01/2006
From:	Hironori Peterson NRC/RGN-III/DRS/OLB
To:	Cobb D Detroit Edison
References
ER-06-301
Download: ML061210521 (15)

Text

May 1, 2006

SUBJECT:

FERMI POWER PLANT, UNIT 2 NRC INITIAL LICENSE EXAMINATION REPORT 05000341/2006301

Dear On March 10,

2006, the NRC completed initial operator licensing examinations at your Fermi Power Plant, Unit 2. The enclosed report documents the results of the examination which were discussed on March 10, 2006, with Mr. K. Hlavaty, and other members of your staff. A subsequent telephone conversation was conducted on March 24, 2006, with Mr. R. Duke and other members of your staff to discuss the NRCs resolution of the written examination post-examination comment.

NRC examiners administered the operating test during the week of March 6, 2006. Members of the Fermi Power Plant Training Department staff administered the written examination on March 10, 2006. Five senior operator and four operator applicants were administered license examinations. The results of the examinations were finalized on April 5, 2006. Five applicants passed all sections of their examinations resulting in the issuance of four senior operator and one operator licenses. One operator and one senior operator scored less than 82 percent overall on the written examination; and, in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501.D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted.

In accordance with 10 CFR Part 2.390 of the NRC's Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

-2-We will gladly discuss any questions you have concerning this examination.

Sincerely,

/RA/

Hironori Peterson, Chief Operations Branch Division of Reactor Safety Docket No. 50-341 License No. NPF-43

Enclosures:

1.

Operator Licensing Examination Report 05000341/2006301(DRS)

2.

Simulation Facility Report 3.

Post Examination Comments and Resolutions 4.

Written Examinations and Answer Keys (RO & SRO)

REGION III==

Docket No:

50-341 License No:

NPF-43 Report No:

05000341/2006301 Licensee:

Detroit Edison Company Facility:

Fermi Power Plant, Unit 2 Location:

6400 North Dixie Highway Newport, MI 48166 Dates:

March 6 through 10, 2006 Examiners:

M. Bielby, Chief Examiner R.K. Walton, Examiner C. Zoia, Examiner Approved by:

H. Peterson, Chief Operations Branch Division of Reactor Safety

Enclosure 1

SUMMARY OF FINDINGS

ER 05000341/2006301(DRS); Fermi Power Plant, Unit 2; Initial License Examination Report.

The announced operator licensing initial examination was conducted by regional examiners in accordance with the guidance of NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9.

Examination Summary

• Nine license examinations were administered (four operator and five senior operator).

• Five applicants passed all sections of their examinations, one of these applicants was issued an operator license, and four were issued senior operator licenses. Two operator applicants failed the written examination and will not be issued licenses. Two applicants (one senior operator and one operator) scored less than 82 percent overall on the written examination; and, in accordance with the guidelines of NUREG 1021,

Operator Licensing Examination Standards for Power Reactors, ES-501.D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted.

(Section 4OA5.1)

REPORT DETAILS

4OA5 Other

.1 Initial Licensing Examinations

a. Examination Scope

The NRC examiners conducted an announced initial operator licensing examination during the week of March 6, 2006. The licensee used the guidance established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the examination outline and to develop the written examination and operating test. The NRC examiners administered the operating test March 6 through 9, 2006. Members of the Fermi Power Plant Training Department administered the written examination on March 10, 2006. Five Senior Reactor Operator (SRO) and four Reactor Operator (RO) applicants were examined.

b. Findings

Written Examination The licensee developed the written examination. During their internal review, the NRC examiners determined that the examination, as submitted, was within the range of acceptability expected for a proposed examination. Written examination comments developed during review by the NRC staff, and as a result of examination validation were incorporated into the written examination in accordance with the guidance contained in NUREG-1021.

One post-examination comment on the reactor operator portion of the examination was submitted by the applicants and station training department personnel on March 17, 2006. The results of the NRCs review of the comment is documented in Attachment 3, Post Examination Comments and Resolutions.

Operating Test The NRC examiners determined that the operating test, as originally submitted by the licensee, was within the range of acceptability for a proposed examination. The examiners validated the operating test and replaced or modified several items in the proposed operating test. Test changes, agreed upon between the NRC and the licensee, were made in accordance with NUREG-1021 guidelines.

Examination/Test Results Five applicants passed all sections of their examinations, four of these applicants were issued senior operator licenses. Two RO applicants failed the written examination and will not be issued a license. Two applicants (one SRO and one RO) scored less than 82 percent overall on the written examination; and, in accordance with the guidelines of

NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501.D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted.

During administration of Job Performance Measure (JPM) j, Shutdown of the Plant from the Relay Room, examiners identified a procedure enhancement and training issue based on observation of applicant performance. As part of a simulated event, the applicants were informed that the control room had been evacuated due to a toxic atmosphere with the plant operating at full power. A reactor scram had not been actuated prior to control room evacuation and the applicants were directed to perform Abnormal Operating Procedure (AOP) 20.000.19, Shutdown From Outside the Control Room, Condition B to scram the reactor from the Relay Room. Procedure Step B.2 required applicants to take any two operable Average Power Range Monitor (APRM)module mode switches out of OPER by using the APRM mode switch keys that were staged in the Relay Room Key Cabinet. The procedure step referenced Enclosure E for determining location of the cabinet in the relay room. During JPM validation, operators showed examiners that the keys were staged in a red break-glass-to-enter box located in the Relay Room near Panel C97P001 (approximately 3 feet off the floor). However, during the JPM administration, applicants were not able to complete procedure action Step B.2 because Enclosure E did not clearly indicate location of the cabinet. Some applicants identified that the term Relay Room Key Cabinet was the name of the key locker located in the Control Room Shift Managers office and believed the procedure was possibly in error. Previously licensed applicants also could not locate the keys, nor had knowledge of their existence in the Relay Room. Completion of the procedure was delayed while applicants performed a search of the Relay Room for the key cabinet.

However, based on the JPM initial conditions, applicants realized they were still able to scram the reactor because Step B.5 required them to trip the Main Turbine (MT) by manually actuating the MT Trip Relay. With reactor power above the minimum MT First Stage Shell Pressure (FSSP) equivalent to approximately 35 percent reactor power, a MT trip would cause a reactor scram. Examiners noted that if reactor power had been below the equivalent minimum MT FSSP for approximately 35 percent reactor power, a MT trip would not cause a reactor scram, and the applicants would not have been able to scram the reactor in accordance with Condition B guidance of 20.000,19.

The licensee entered these self-revealing issues into their corrective action program as Condition Assessment Resolution Document (CARD) 06-21197. The training issue for lack of knowledge of the Relay Room key cabinet and procedure enhancement for location of the key cabinet were considered minor in nature and not subject to enforcement action in accordance with NRC enforcement policy.

.2 Examination Security

a. Inspection Scope

The NRC examiners briefed the facility contact on the NRCs requirements and guidelines related to examination physical security (e.g., access restrictions and simulator considerations). The examiners observed the implementation of examination security and integrity measures (e.g., security agreements) throughout the examination process.

b. Findings

The licensees implementation of examination security requirements during examination preparation and administration was acceptable and met the guidelines provided in NUREG-1021, Revision 9 and no findings were identified. However, one significant incident occurred that resulted in a loss of the licensees original proposed examination.

Three examiners were onsite the week of January 23, 2006, to validate and incorporate comments into the licensee prepared operating test and written examination that had been submitted to the region. On the morning of January 25, 2006, the Chief Examiner's rental car and three other vehicles in the hotel parking lot were burglarized.

The examiner's portable laptop computer containing a disc with the licensees electronically submitted examination was stolen from the vehicle. Local county sheriff officers were immediately notified and later apprehended the suspects. However, the examiner's computer and examination disc was not among the recovered items. For clarification, the disc contained the original licensee submitted examination before subsequent incorporation of any NRC comments. Since the examination disc was not recovered, and not password protected, headquarters concurred with the region's decision to significantly modify the original licensee proposed examination.

Headquarters identified a precedence for changing 50 percent of the examination based on a lost examination in another region. Headquarters concurred with the region's decision to use the original licensee submitted outline and a maximum of 50 percent of the significantly modified original examination items. The licensee Training Department immediately commenced to develop an examination. During the week of February 6, 2006, two examiners worked onsite with the licensee to co-author and complete replacement of more than 50 percent of the original examination. During the week of February 13, 2006, three examiners performed an onsite validation of the examination.

The final NRC review of the examination was performed in the region the week of February 27, 2006. The examination administration start date was delayed one week from the original date.

4OA6 Meetings

.1 Exit Meeting

The chief examiner presented the examination team's preliminary observations and findings on March 10, 2006, to Mr. K. Hlavaty and other members of the Operations and Training Department staff. A subsequent exit via teleconference was held on March 24, 2006, with Mr. R. Duke following review of the site post-examination comment. The licensee acknowledged the observations and findings presented. No proprietary information was identified by the stations staff during the exit meetings.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Duke, General Supervisor, Operations Training

E. Kokosky, Manager, Operations Training

P. Smith, Director, Nuclear Assessment

R. Gaston, Manager, Licensing

D. Coseo, Shift Engineer, Operations

M. Doucet, Program Supervisor, Operations Training

T. Horan, Principle Specialist, Nuclear Quality Assurance

K. Hlavaty, Plant Manager, Nuclear Production

G. Strobel, Operations Engineer, Nuclear Production

L. Barker, Operations Training

NRC

None.

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened, Closed, and Discussed

None

LIST OF ACRONYMS USED

ADS

Automatic Depressurization System

AOP

Abnormal Operating Procedure

ADAMS

Agency-Wide Document Access and Management System

APRM

Average Power Range Monitor

CARD

Condition Assessment Resolution Document

CS

Core Spray

DRS

Division of Reactor Safety

DW

Drywell

FSSP

First Stage Shell Pressure

HPCI

High Pressure Coolant Injection

JPM

Job Performance Measure

MT

Main Turbine

NRC

Nuclear Regulatory Commission

PARS

Publicly Available Records

RHR

Residual Heat Removal

RO

Reactor Operator

RPV

Reactor Pressure Vessel

SRO

Senior Reactor Operator

SIMULATION FACILITY REPORT

Facility Licensee:

Fermi Power Plant

Facility Docket No.:

50-341

Operating Tests Administered:

March 6 - 9, 2006

The following documents observations made by the NRC examination team during the initial

operator license examination. These observations do not constitute audit or inspection findings

and are not, without further verification and review, indicative of non-compliance with

CFR 55.45(b). These observations do not affect NRC certification or approval of the

simulation facility other than to provide information which may be used in future evaluations.

No licensee action is required in response to these observations.

During the conduct of the simulator portion of the operating tests, the following item was

observed:

ITEM

DESCRIPTION

Main Turbine

Vacuum Recorder

The digital recorder quit working for every other scenario, but had no

effect on the operating test. Simulator operators could not identify a

reason for the intermittent operation. The recorder status was

identified to applicants as part of the scenario turnover.

Post-Examination Comment and Resolution

Question Number 50.

The plant was operating at 100 percent power when the following occurred:

Time

Event

1010

MSIV Closure and Reactor Scram, all rods fully inserted.

1011

All Condensate Pumps and CRD Pumps tripped.

1015

Reactor Water level is 130 inches (lowering at 2 inches per minute).

1016

HPCI initiated on a VALID signal.

1055

Div 1 AND Div 2 RHR are operating in Torus Cooling Mode.

1104

1D40, ADS RELAY ENERGIZED, and 1D44, ADS TIMERS INITIATED, alarms.

1105

The CRNSO inhibits ADS Logic A ONLY

1105

Reactor Pressure is 400 psig.

Which of the following describes the expected plant response?

A. The ADS valves WILL NOT open automatically.

B. The ADS valves WILL NOT open manually OR automatically.

C. The ADS valves WILL open automatically. The RHR system will inject FIRST to restore

level.

D. The ADS valves WILL open automatically. The Core Spray system will inject FIRST to

restore level.

Answer: D

Applicant Comment:

Depending on the number of RHR pumps operating in Torus Cooling, it may be plausible for

RHR to inject first. [Answer: C]

Facility Proposed Resolution:

Facility management agreed with the applicants statement and identified two potentially correct

answers based on two different assumptions for Residual Heat Removal (RHR) Torus Cooling

alignment. The question required two different knowledges. The first part of the question

required knowledge of Automatic Depressurization System (ADS) operation and the required

prerequisites to actuate the system. The question stem provided all of the initiating conditions

and required the applicant to identify them. In addition, the applicant had to recognize the

effect on ADS operation of bypassing one ADS logic string. The second part of the question

required understanding of the fundamental knowledge of fluid flow based on a comparison of

alternate flow path effects and relationship of different modes of RHR.

The stem of the question required the applicant to evaluate the High Pressure Coolant Injection

(HPCI) start on a valid signal. Since Reactor Pressure Vessel (RPV) low water level

(approximately 32 inches) had not been reached, the only valid start signal was high Drywell

(DW) pressure which would also start all RHR and Core Spray (CS) pumps. Injection

permissive to the RPV does not actuate until about 431 psi. Subsequent to HPCI starting on

high DW pressure at 1016, Divisions 1 and 2 RHR were placed in Torus Cooling at 1055 which

required the Core Spray Defeat switch to be placed in the DEFEAT position and maintain RHR

in Torus Cooling mode. Both the RHR and CS systems have similar pump discharge head

characteristics. RHR has a discharge head of 310 psi and CS has a head of 285 psi. The flow

rates differ, RHR has a volume of 10,000 - 14,000 per pump (2 pumps per division) and CS has

a volume of 7000-7500 per division (1 pump per division). Each Torus Cooling line is orificed to

only allow a maximum flow of 10,000 gpm per line.

As written, the examination identified that RHR Divisions 1 and 2 were operating in Torus

Cooling Mode at time 1055. The question author intended that applicants would assume a

normally directed (by SRO from flow charts) RHR Torus Cooling maximized alignment that

would include one RHR pump in each division operating, the divisional heat exchanger bypass

valve closed, and flow maximized. This would have required the operators to shutdown one of

the operating RHR pumps in each division that had started on the high DW pressure. At time

1104 a low reactor water level would send a second start signal to all CS and RHR pumps. In

addition, ADS would actuate and lower RPV pressure. After the low pressure permissive

actuation, and based on this RHR alignment to the Torus (with its large line and higher

differential pressure), actual RHR injection to the RPV would occur at a significantly lower

pressure (approximately 150 - 180 psi) than CS, and Answer D would clearly be the correct

answer.

However, during the written examination administration, one applicant questioned the proctor

concerning the RHR Torus Cooling alignment (Question 50). The proctor incorrectly confirmed

that all Division 1 and 2 RHR pumps were running in Torus Cooling. If this were the assumed

lineup made by the other applicants, then the RHR heat exchanger bypass valve would have

been open and all 4 RHR pumps would have been running. As a result, the RHR and CS

injection to the RPV would occur at approximately the same time and pressure, such that either

Answer C. or D. could be correct.

The licensee recommended the question be modified in the bank to add with flow maximized

through the [RHR] heat exchanger to clarify the condition of the RHR pumps in Torus Cooling

to make Answer D. correct.

NRC Resolution:

Upon regional examiner review of the question, the applicants comment, facility proposed

resolution, and discussions with the facility, the question was deleted from the written

examination.

The examiners reviewed the question as a whole. A portion of the distractor B. statement that

...ADS valves WILL NOT open manually... was considered implausible. The majority of the

question was focused on analysis of the stem and determination of which ECCS system would

first inject to the RPV, not on the stated KA 218000 A1.04, Ability to predict and / or monitor

changes in parameters associated with operating the ADS controls including reactor pressure.

The question stem did not clearly articulate the RHR Torus Cooling pump status. One applicant

questioned the RHR Torus Cooling alignment and was incorrectly informed by the proctor that

all RHR pumps were operating in Torus Cooling. However, that information was not

disseminated to the rest of the class. The question author had intended that applicants would

assume a normally directed RHR Torus Cooling maximized alignment that would include one

RHR pump in each division operating, the divisional heat exchanger bypass valve closed, and

flow maximized.

The examiners reviewed applicable RHR and CS pump curves to confirm the expected flows

and head pressures, and verified the RHR, ADS, and CS system alignment and response to

actuation signals, and concurred with the facilitys descriptions. The facility submitted an NRC

Examination Feedback Form that documented their response to the applicants post

examination comment. The response identified that the facility ran a scenario on the simulator

similar to the question which included Division 1 and 2 RHR in Torus Cooling after all 4 RHR

pumps started on high DW pressure. The RHR discharge pressure was identified as very

high due to all RHR pumps running with discharge flow at approximately 20,000 gpm. As a

result [when RPV pressure decreased below the permissive] CS injected about 2 seconds prior

to RHR at a RPV pressure about 3 psid higher. Specifically, the results indicated that CS would

inject to the RPV first; however, both RHR and CS would inject at approximately the same time

and pressure and it would be impractical to require an applicant to predict with any degree of

certainty, which pump would inject first under these conditions.

The examiners disagreed with the licensees resolution to accept two correct answers based on

the lack of question KA focus, implausible distractor, unclear stem conditions, and inability to

identify a clearly correct answer. As a result, the examiners removed the question from the

examination.

WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)

RO Initial Examination ADAMS Accession # ML061180435

SRO Initial Examination ADAMS Accession # ML061180444