ML20199D350

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

ML20199D350
Person / Time
Site:	05532442
Issue date:	01/15/1999
From:	Munro J NRC (Affiliation Not Assigned)
To:
Shared Package
ML20199D323	List: ML20199D318 ML20199D331 ML20199D350
References
99-753-01-SP, 99-753-1-SP, SP, NUDOCS 9901200076
Download: ML20199D350 (48)

Text

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l January 15,1999 l

UNITED STATES OF AMERICA

. NUCLEAR REGULATORY COMMISSION '

e BEFORE THE PRESIDING OFFICER l'

' Before Administrative Judge:

l Peter B. Block (Dr. Richard F. Cole, Special Assistant) 1 In the Matter of ) Docket No. 55-32442-SP

) .

SHAUN P. O'HERN ) l

) ASLBP No. 99-753-01-SP  !

l,. -(Denial of Application for ) i Reactor Operator License) )  !

i i

AFFIDAVIT OF JOHN F. MUNRO IN SUPPORT OF NRC'S STAFF RESPONSE TO SHAUN P. O'HERN'S WRITTEN PRESENTATION i

John F. Munro, having first been duly sworn, does hereby state as follows:

1

l. I. My name is John F. Munro. I am employed as a Senior Reactor 1

' Engineer - Examiner Qualified, in the Operator Licensing and Human Performance l

- Branch (HOHB), Division of Reactor Controls and Human Factors (DRCH), Office of

]

Nuclear Reactor Regulation (NRR), NRC Headquarters Office, in Rockville, Maryland.

I was the lead reviewer assigned to evaluate the results of the Region III office review i

- and the appeal board's independent evaluation for Mr. O'Hern's written examination

contentions (Hearing File Items 5,7, and 9). A statement of my professional 4

qualifications is attached hereto.

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PDR Exhibit 2

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2. This Affidavit is submitted in support of NRC Staff's response to the written presentation dated December 7,1998, submitted by Mr. Shaun P. O'Hern (Presentation), in support of his request for a hearing on the NRC Staff's proposed denial of his application for a Reactor Operator (RO) license for use at the Enrico Fermi Nuclear Station, Unit 2, operated by the Detroit Edison Company.

3. The initial review of Mr. O'Hern's May 29,1998 letter (Hearing File Item 5) by the Staff in Region III concluded that Mr. O'Hern's failing grade for the written examination should be sustained with no grading changes for any of the contested questions (Hearing File Item 7).

4. Pursuant to Examination Standard (ES) 502.D " Processing Requests for Administrative Reviews and Hearings After Initial Denial," of NUREG-1021 (Hearing File Item 18, page 4), the Chief, HOHB, convened a three-person appeal board to review Mr. O'Hern's documented contentions. On July 31,1998, the appeal board communicated its findings to the Chief of HOHB, NRR, and recommended that Mr. O'Hern's failing grade for the written examination be sustained (Hearing File Item 9; Exhibit 1). However, the appeal board recommended that Mr. O'Hern's written examination grading be changed for three questions (#17, #38, and #87) contested by him and two questions (#56 and #71) contested by other applicants.

5. Pursuant to ES-502.D of NUREG-1021 (Hearing File Item 18, page 5),

the Chief, HOHB, is directed to make a final recommendation to the Director, DRCH.

The Director, DRCH, will then consider the findings and recommendations of the

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s appeal board and make a final decision whether to sustain or overturn the applicant's i

license examination failure.

6. In August and September 1998, I conducted the HOHB Staff review of the appeal board's findings and recommendations, as well as the results of the Region III informal review. I proposed that the appeal board's recommendations be accepted for all but two questioas (#25 and #87). For question #25, I concurred with

- Mr. O'Hern's request that it be deleted and disagreed with both the Region III and appeal board recommendations that it not be changed. However, for question #87, I endorsed the Region III recommendation, that no grading change was warranted, in opposition to both the appeal board's conclusion and Mr. O'Hern's request that an additional answer choice be credited as correct. The Chief, HOHB, and Director, DRCH, agreed with my recommended disposition and Mr. O'Hern was notified in writing on September 7,1998, that his proposed denial was sustained (Hearing File Item 10).

7. During the review, I noted that the appeal board's recommendation to sustain the written examination failure, as detailed in the transmitting memorandum and the detailed analysis section of Attachment 2, was in variance with the tabulated summary of examination question changes for questions #35, #38, and #54 (Hearing File Item 9, memmandum and page 12; Exhibit 1). For question #35, the grading summary indicated either answer choice (b) or (d) was acceptable and an additional 1

o i point ~was credited to Mr. O'Hern. This conclusion was incorrect since Mr. O'Hern L

l selected answer choice (c) (Hearing File Item 3, last page). For question #38, the i

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grading summary was not changed and no additional points were credited. This conclusion was incorrect since the appeal board's detailed analysis preceding the summary page recommended the question be deleted from the examination (Hearing File Item 9, page 6; Exhibit 3, page 6). Finally, for question #54, the grading summary indicated that either answer choice (b) or (c) was acceptable and an additional point was credited to Mr. O'Hern. This conclusion was incorrect since the appeal board's detailed analysis preceding the summary page recommended concurrence with the Region III resolution and, therefore, no change to the question's original grade was

! warranted (Hearing File Item 9, page 8; Exhibit 3, page 1). I communicated my l

l fmdings regarding the inconsistencies noted in the tabulated summary of examination question changes to the appeal board chairman, Mr. John L. Pellet, Chief, Operations

Branch, Division of Reactor Safety, NRC Region IV, in Dallas, Texas. Mr. Pellet t

agreed with my fmdings and reconfirmed that the appeal board's recommendation was  ;

l that Mr. O'Hern's written examination failure should be sustained. He could not l

i ' explain the cause for the errors in the tabulated summary of examination questions, but I

affirmed that the detailed NRC analysis for these three questions accurately reflected the appeal board's findings and recommendations. I recalculated Mr. O'Hern's examination grade using the correct tabulation, for a final grade recommendation of 76/% = 79.2 percent. On January 13,1999, I again contacted Mr. Pellet and l

reconfirmed the accuracy of the analysis discussed above.

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8. I hereby certify that the foregoing is true and correct to the best of my  !

knowledge, information, and belief.

, e f t# ,r=> f John F. Munro, Senior Reactor Engineer - Examiner Qualified l

I Subscribed and sworn to before me

! thi day January 19 .

&AV) < Y Nordry Public" f ELVA BOWOEN FERe:Y 1

NOTARY PU3UC sity op f,y,nyan My commissiori expires /4 Comminien Ennes December 1,1p;p  ;

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i 1/15/99 JOHN F. MUNRO i i l

PROFESSIONAL OUALIFICATIONS OPERATOR LICENSING AND HUMAN PERFORMANCE BRANCH -

DIVISION OF REACTOR CONTROLS AND HUMAN FACTORS -

OFFICE OF NUCLEAR REACTOR REGULATION -

U.S. NUCLEAR REGULATORY COMMISSION i My name is John Munro. My business address is: One White Flint North,11555 Rockville Pike, Rockville, Maryland 20852. I am employed by the United States Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation as a Senior Reactor Engineer - Examiner

! Qualified, in the Operator Licensing and Human Performance Branch of the Division of i i Reactor Controls and Human Factors. 4 i

I graduated from Villanova University in 1974 with a Bachelor of Science degree in Mathematics. Upon graduation from Villanova University, I was commissioned as an officer in the United States Navy and entered the Naval Nuclear Power Program under the direction of Admiral H. G. Rickover. Following one year of intensive classroom and on-the-job training, I was certified as an Engineering Officer of the Watch (EOOW) responsible for supervision of the activities of a naval nuclear reactor.

i I served two years aboard the nuclear submarine USS LEWIS AND CLARK and

approximately another two years aboard the nuclear submarine USS BREMERTON in various supervisory engineering positions before resigning my commission in 1979 to join the General

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1/15/99 Electric (GE) Company. I was employed for a year and a half as a Start-Up Test Engineer with the GE Company. As such, I was assigned to the Nuclear Energy Business Group Office in San Jose, California, and in 1980 to the GE Start-Up Test Group at the Shoreham Nuclear l Station. My duties as a Start-Up Test Engineer included: review and testing for two GE supplied plant systems, preparation of pre-operational test procedures, preparation of i

responses to technical inquiries from site field offices, and participation as an advisor during l

l integrated system testing. I also completed the GE Boiling Water Reactor Training Center's -

l Senior Reactor Operator (SRO) training course and was certified as a SRO on the Dresden l

(- Nuclear Station by GE in 1979.

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l- In February 1981, I resigned from the GE Company and accepted employment with the U.S.

1 Nuclear Regulatory Commission (NRC) as an Operator License Examiner in Bethesda, l Maryland. In 1981, I certified as BWR examiner and Chief Examiner, and led license j examination teams at various BWR facilities nationwide. In 1983, I transferred to the i Region II Office in Atlanta, Georgia, and continued to serve as a Chief Examiner for the BWR i i

facilities in Region II. In 1986, I completed the NRC's Westinghouse Pressurized Water Reactor (PWR) Technology and Advanced Technology courses and certified as an Examiner on PWRs. Also, in 1986, I was promoted to the Section Chief for the Operator Licensing Section in Region II and served in that position through 1990. As such, I was responsible for j implementing the NRC's operator licensing activity in Region II. My collateral duties

. included assignment as a BWR technical advisor and designation as a potential NRC Incident  !

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Investigation Team member. Also, in 1989, I was assigned for six months as a Resident Inspector at the North Anna Nuclear Station. In December 1990, I transferred to the U.S.

NRC Headquarters Office in Rockville, Maryland, in the position of Senior Reactor Engineer -

Examiner Qualified in the Operator Licensing and Human Performance Branch.

Currently, I serve as one of two Examiner Qualified Senior Reactor Engineers in the Operator Licensing and Human Performance Branch in the Office of Nuclear Reactor Regulation (NRR). As such, I am responsible for the conduct of oversight reviews and other associated tasks related to the conduct of the operator licensing function as applied by the Regional Offices. I also continue to be designated as a potential NRC Incident Investigation Team j member. I l

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f.-.s%n umTro STATES NUCLEAR REGULATORY COMMISSION 5 ,

! . REGloNIV -

) 811 RYAN PLAZA drive. sutTE 400

.....g ARUNGToN TEXAS 76011 a064 July 31, 1998 l

MEMORANDUM TO: Robert M. Ga!!o, Chief Operator Licensing and Human Performance Branch .:

Division of Reactor Controls and Human Factors .

Office of Nuclear Reactor Regulation FROM: John L. Pe!!st, Chief, Operations Branch

)

Division of Reactor Safety, Region IV y $1 $ i l

Thomas R. Meadows, Reactor En'gineer/ Examiner Division of Reactor Safety, Region IV k, Joe D' Antonio, Reactor Engineer. Examiner 1

Division of Reactor Safety, Region 1 - ,

)

SUBJECT:

6 O'HERN,M (FERMI 2) APPEAL RESULTS i e

's, In accordance with ES 502.D.2, the written test informal review requests filed bM p at r ( candidates at the

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Fermi 2 Nuclear Plant, were independently reviewed by Region IV and by the three-member panel as listed.

The panel, John Pe!!et, acting as chairman, met by telephone on June 25 30,1698, to consider the informal appeals of the throa candidates for license in accordance with Mr. Gallo's '

memorandum of June 6,1998. The results of the panel's review and recommendaSons are presented below. Detailed analysis of each contention of each candidate is presented in the attachments to this memorandum as described below. .

In summary, the panel unanimously recommends two candidates 6 be

• evaluated as having passed the written examination portion of their license examinations and, should that have formed the sole basis for the proposed denials, that the individual proposed denial should be overtumed and the appropriate licenses issued.

The panel recommends, also unanimously, that the third candidate (O' Hem) be' evaluated as failing the written portion of the operating test and sustains the proposed denla!.

If you have any questions, piease contact me at (817)880-8159. i I

1 Exhibit 3

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ATTACHMENT 2 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442 QUESTION # 2 Whenever HPCI is to be operated for surveillance testing, torus cooling is required to be placed in service as a prerequisite to starting HPCI. Insuring that torus cooling is in servjce before operating HPCl: .,

s. Allows the maximum average suppression pool water temperature limit to be increased to 105'F.

b. Ensures adequate thermal mixing of the water in the suppression pool to limit stress on the torus shell due to differential thermal expansion.

c. Extends the operating time for HPCI testing before the maximum average temperature limit is reached and testing is required to be stopped.

d. Ensures that heat added to the suppression pool does not increase torus air space pressure to the point where the Suppression Chamber to Drywell vacuum breakers cycle.

Answer (eer answer kevi: (c)

Candidate's resoonse: (b)

Candidate's Contention:

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The candidate contends that this question has two possible correct answers, (c) and (b).

The candidate contends that there are no references to support why RHR is placed in Torus Cooling to support running HPCI. However, the candidate contends that, based on Heat Transfer and Fluid Flow Theory, answer (b) is correct.

NRC Analvsis: -

The region found that the ca ididate did not provide any documentation to support his statement about Heat Transfer and Flu d Flow. The candidate did not provide any documentation which states that RHR is placed in Wrus Cooling to limit stress'on the torus shell due to differential thermal expansion.

The panel rejected the candidate's contention. The panel found, from the references supplied, that cross mixing is a benefit of cross-dMsion RHR torus cooling, but is not the reason torus cooling is used.

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• ATTACHMENT 2 i

4 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442

QUESTIONS 7 From full power operation, a transient has occurred. The following annunciators were received'.

l 3D73, Trip Actuators A1/A2 Tripped 1 3D74, Trip Actuators B1/B2 Tripped .-

l 3D99, APRM Upscale Neutron /ThermalTrip .

f immediately after receipt of these annunciators, the following parameters were reported to the j NASS:

i Reactor Powerh8'nd stable RPV Level 164 inches, decreasing slowly l -G> Reactor Pressure 10_8!Imig, 8 increasing slowly 5

With these plant conditions, what is the first action thatg be performed, and which indication 4 must be observed to verify proper response?

l _ g a. Manually operate SRVs to stabilize pressure at less than 1050 psig; observe Div 1 and 2 l post accident recorders.

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! b. Place the SVLCV Bypass Valve Mode Switch in STARTUP, and verify RPV level is not

increasing.

I c. Initiate Afternate Rod insertion; perform OD-7 option 2.

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.-+ d . Place the Reactor Mode switch in SHUTDOWN; verify blue group scram lights are Off.

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Answer (eer answer kev)

(d)

Candidate's renconse: (a)

j. Candidate's Contention:  ;

! The candidate contends that this question has two possible correct answers, (d) and (a). The candidate contends that he would be directed by t!'oi SRO to maintain reactor pressure between 900 to 1050 psig.

NRC Analvsis:

The region found that the candidate did not answer the question as it was asked. Per 29.100.01 SH 1 A, "RPV CONTROL - ATWS," directions are given to stabilize RPV pressure less than E psig if SRVs are not cycling. Per the stem RPV pressure was M psig increasing slowly.

Thus, manually operating SRVs are not required at this time. Per 29.100.01 SH 1 A,"RPV CONTROL . ATWS," the first operator action on the Rx power leg is to " Confirm Rx mode switch in S/D."

The panel concurred with the region's response.

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ATTACHMENT 2 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442 QUESTION # 17 Following a transient at power, a Reactor Scram has occurred, and annunciator 3D12, Div 1/Il ON Gas Radn Monitor High High, has alarmed. Channel A and B ON Gas Radiation Monitors are both reading 1100 mr/hr Based on these indications, what action (s) must be performed?

a. Verify Radiation Protection surveys the Scram Discharge Volume. -

b. Ensure the scram is reset within 5 minutes.

c. Isolate the Main Steam Lines, and initiate MSIVLCS.

d. Verify or initiate a trip of the Mechanical Vacuum pumps.

Answer (eer answer kevi: (a) i

! Qandidate's resconse- (c) 1 l Candidate's Contention:

l The candidate contends that this question has two possible correct answers, (a) and (c). The i candidate contends that because of the large increase in radiation level, the FUEL CLADDING

! FAILURE (GROSS) section of 20.000.07,' FUEL CLADDING FAILURE," should have been  :

1 l entered. This section directs the closing of the MSIVs and the subsequent initiation of MSIVLCS l for gross fuel failure.

I NRC Analvsis:

j The region found that the candidate should have entered the FUEL CLADDING FAILURE

< (SMALL) section of 20.000.07. There are two sections of Procedure 20.000.02 "SMALL" s. -

" GROSS." There were no symptoms present that directed the candidate to use the *GROSf/

section versus the 'SMALL" section. The 'SMALL" section directed the candidate to perforut 20.000.21, " Reactor Scram," concurrently. There is a Caution listed in 20.000.21 'REACTC,R

! SCRAM" which states, 'If fuel damage is suspected, do not reset the Scram until Radiation j Protection surveys the SDV for excessive radiation levels.

l The panel accepted the candidate's contention that the symptoms stated were applicable to both sections. However, rather than two correct answers, the cited reference supports that answers

, (a), (c), and (d) are correct. Based on the guidance in NUREG-1021, Interim Revision 8, the panel deleted the question.  !

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1 ATTACHMENT 2 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442 j QUESTION # 26 4

The plant is operating at 14% reactor power. A non-selected control rod drifts in from position 4 36 to position 00. The Rod Worth Minimizer (RWM) has identified this control rod as an insert 4

error and produced a rod block. -

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Which one of the following statements describes the action which must be taken with regard to j

' the RWM in order to continue the startup?

l a. With only one rod inoperable, station a second operator to verify rod motion, bypass the

' RWM, and continue the plant startup.

b. Remove the control rod from the enforcement secuence by using the RWM control rod bypass function and restore the control rod to itw original position.

I c. The rod must be declared inoperable and substitute position data entered into the RWM until power has been increased above the LPSP.

4 d. Correct the cause of the rod drift, bypass the RWM, and restore the control rod to its original position using another operator for second verification.

i l Answer (Der answer kevh (d)

)

j Candidate's rescense: (c). i l

Candidate's Contention:

j The candidate contends that none of the responses as written are correct, and this question j should be removed from the examination. The candidate contends that IAW 23.608, *RWM,"  ;

f Section 5.0, Prerequisite 5.1.4 requires the current control rod pattom be in compliance with the  ;

approved pull sheets prior to bypassing the RWM. j l

NRC Analysis, j l

The region determined that, if the RWM is being bypassed simply to be removed from service, l then the candidate is correct. Per the stem, the RWM is being bypassed to correct an insert  ;

error. Under this condition the RWM can be bypassed at any time as long as the requirements j of T/S 3.1.4.1 are met. Per T/S 3.1.4.1, the RWM shall be OPERABLE in OPERATIONAL CONDITIONS 1 and 2*, when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER.

The panel concurred with the region's resolution.

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ATTACHMENT 2 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442 l QUESTION # 34 With the plant operating at full power, HPCI is running in full flow test for surveillance testing, and Div 1 RHR is operating in Suppression Pool Cooling. Annunciator 17D8, Div 11 Hydrogen Concentration High,'is received. The CRNSO reports hydrogen reading 2.5% on T50-CRE-RB068, Suppres,sion Chamber Atmosphere Analysis, and confirms that the Div 2 Atmosphere monitor is aligned to the Torus. Hydrogen Water Chemistry has been shut down for two weeks.

Which cne of the following actions is appropriate at this time?

a. Isolate the HPCI System to prevent further exhaust into t'.' suppression pool p3r 23.202.

b. Immediately shift Div 1 RHR lineup to provide torus spray per ARP 17D8.

c. If a chemistry sample confirms the high hydrogen concentration, isolate HPCI per 23.202 and shift Div 1 RHR to the torus spray moda per 23.205.

I d. Vent the torus per 29.100.01.

Answer (eer answer kevi: (d)

Candidate's resoonse: (a)

Candidate's Contention:

The candidate contends that this question has two possible correct answers, (a) and (d). l l The candidate contends that since the answer (d) does not include the word " Purge," l that it is not the only correct answer.

NRC Analysis:

The region found that the information that hydrogen water chemistry was shut down and, therefore, the reactor coolant system hydrogen chemistry will be low makes answer (a) incorrect.

The panel concurs with region's resolution. The panel a!so noted that the emergency operating procedure takes precedence, which further suppolis that (d) the only correct answer.

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. l ATTACHMENT 2 DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL i DOCKET NO. 55-32442 i QUESTION # 38 l

Following a Loss of Off-Site Power, the EDGs have started. With the EDG breaker closed, j which one statement below contains the additional signal required to start the EDG Digital Load ,

Sequencer? '-

a. EDG bus under voltage )

b. EDG bus lockout relays reset i

c. EDG bus under frequency.  !

d. EDG bus load shed relay deenergized  ;

I Answer (eer answer kev): (a)

Candidate's resoonse: (d)

Candidate's Contention:

The candidate contends that none of the responses as written are correct, and this question

. should be removed from the exam. The candidate contends that a Loss of Off-Site power (LOSP), EDG start, and EDG breaker closure is all that is required to start the EDG Digital Load .

Sequencer. The candidate contends that a LOSP would actuate the load shedding circuit and provide the under voltage signal to the load sequencer.

NRC Analvsis:

The region found that the answer is correct as written. The candidate's answer supports the fact

that (a) is the correct answer. The under voltage signalis a requirement to start the EDG Digital

Load Sequencer. Regardless of how the under voltage signal actuated it is still required for the 1 EDG Digital Load Sequencer.

The independent panel accepted the candidate's contention. From the references presented, the candidate is correct in that, for the conditions given in the stem, answer (a) already exists I and cannot be an additional signal required. The panel deleted the question from the exam.

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ATTAQHMENT2 DETAILED* ANALYSIS OF O'HERN INFORMAL APPEAL DOCKET NO. 55-32442 QUESTION # 45 The following conditions exist after a LOCA has occurredi

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-Torus Water Temp 130'F

-Torus Pressure 23 psig

-Torus Level +16 inches

- RPV Level 112 inches and rising

- RPV Pressure 24g psig *

- Drywell Pressure 25 psig Whbh one of the following concems requires action to be taken? , .,

s. ECCS pump suction vortexlimits

b. !oss of drywell to suppression chamber vacuum breaker function

c. RCIC high exhaust pressure trip on elevated {orus pressure .

d. plugging of low pressure ECCS suction strainers '

Answer (ser answer kev): (d)

Candidate's reseense: (c)

Candidate's Contention: -

The candidate contends that this question has two possible correct answers, (d) and (c). The candidate contends that based on assumptions, appropriate operator actions have occurred to prevent ECCS suction strainer plugging.

NRC Analvs's:

The region found that the question did not require any assumptions. The candidate's assumption is not a valid one because the stem asked,'Which one of the following concoms requires action to be taken?' The RCIC trip is an auto action that does not require operator action. The ECCS pumps will not trip on low suction pressure. Thus, manual actions are required to secure them or to prevent suction strainer plugging.'

The panel concur's with region's resolution.

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ATTACHMENI2 ,

DETAILED ANALYSIS OF O*HERN INFORMAL APPEAL DOCKET NO. 55-32442 QUESTION # 54 Heavy thunderstorms just caused a loadveject from 100% power. The reactor conditions are:

- APRM Power stable at 20%

- No indications of control rod position -

- Recirc pumps tripped

- All MISVs are open

- Reactor Level being maintained by feedwater

- Reactor pressure being maintained through Turbine Bypass Valves

, Mode switch in SHUTDOWN The NSO's first actions should be:

a. Initiate ADS

b. Initiate ARI .

c. Inject SLC

d. Drive control rods in Answer (eer answer kevi: (b)

Candidate's resnonse: (c)

Candidate's Contention: .

The candidate contends that this question has two possible correct answers, (b) and (c). The candidate contends that answer (c) is correct because based on a scenario that was run on the Fermi 2 simulator, if the SRV setpoint is reached (1135 psig) the ATWS/ARI setpoint of 1133 psig would also be reached. Thus, ARI would have already initiated.

NRC Analygig:

The region determined that the question should have been answered based only on the information p!ven in the stem. Per 29.100.01 SH 1A, RPV CONTROL - ATWS, the NSO's first actions should have been to manually initiate ARl, without regard to whether an automatic setpoint had been reached.

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The panel concurred with region's resolution. _

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, p ATTACHMENT 2 1

DETAILED ANALYSIS OF Q'HERN INFORMAL APPEAL DOCKET NO. 55 32442 QUESTION # 59 ff the Reactor Mode switch is in START / HOT STANDBY, which one of the following instruments is NOT required to be operable? ..

a. Reactor Vessel Level 1 for ADS

b. Reactor Vessel Pressure High for ARI -

c. Reactor Vessel Pressure for High Pressure Scram

d. Reactor Vessel Level 2 RWCU System isolation .

Answer (eer answer kev): (b)

Candidate's renconse: (a)

Candidate's Contention:

The candidate contends that this question has two possible correct answers, (b) and (a).

The candidate contends that if the assumption is made that reactor pressure is less than 150 psig, then answer (a) is also correct.

NRC Analvsis:

The region found that this question did not require any assumptions. All the instruments with the exception of reactor vessel pressure high for ARI are required in Modes 1 through 3. -

The panel concurs with reglen's resolution.

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i ATTACHMENT 2 l DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL .

DOCKET NO. 55-32442 I l

s QUESTION # B7 4

l The plant is operating at 96% power with the following indications on the A Recirculation Pump "

i Seal: j

' 2

- Seal #1 Pressure 980 psig

. Seal #2 Pressure 10 psig ,

1 l Annunciator 3D123, RECIRC PMPA STAGING SEAL FLOW HICH/ LOW is alarming

! Flowindication indicates 0.4 gpm Which one of the following seal conditions exist?

l a. Seal #1 has failed

b. Seal #2 has failed

c. #1 Seal Labyrinth is plugged  ;

d. #2 Seal Labyrinth is plugged 4

Answer (eer answer kavi: (c) j Candidate's reneense: (b)

Candidate's Contention:

4 i The candidate contends that this question has two possible correct answers, (c) and (b). The .

I candidate contends that both answers (b) and (c) will result in seal staging flow decreasing to less than 0.5 gpm

- NRC Analysis:

i The region found ;hn its candidate's contention is not supported by the documentation. Per Lesson Plan 315-0004 001, ' REACTOR RECIRCULATION SYSTEM," the following ere indications of the #2 sealfailure:

1. 2 seal pressure decreasiqi High seal flow alarm 0.9 gpm increasing .

The following are indications of the plugging of the #1 seal: ,

1. 2 seal pressure decreasing

1. 1 seallow flow alarm 0.5 gpm decreasing I

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ATTACHMENT 2 l DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL l DOCKET NO. 55-32442 '

l The stem states that the seal staging flow higMow annunciator was in alarm and that seal ,

l staging flow was 0.4 gpm. This lesson plan does not support the statement that both (b) and (c) will result in seal staging flow decreasing to less the 0.5 gpm.

The panel accepted the candidate's contention. The panel found that the stem does not give the necessary information to differentiate between answers (b) and (c). Specifically, knowledge Of the 3D121 annunciator for outer seat leakage high was required. The panel found it unreasonable to expect the candidate to assume the status of this indication and, therefore, revised the key to accept answers (b) and (c).

. l l

I i

I 1

11 l l

l

. A i'

! ATTACHMENT 2

DETAILED ANALYSIS OF O'HERN INFORMAL APPEAL 1

DOCKET NO. 55e32442 f PANEL CONCLUSION.

Summarv of Examination Ouestion channen Original Score: 76/100 Question # Changa Effect on Candidate's Grading Changein Score 2 none no change 76/100* ' "

7 none no change 76/100 ,

, r 17 delete question delete missed question 76/99 25 none no change 76/99 34 none no change 76/99 l 35* accept b or d change fromincorrect to correct 7749 l 38 none no change 7749  ;

45 none no change 7749 [

53* accept c or d no change 7749 54 accept b or c change fromincorrect to correct 78/99 56' delete question delete correct question 7748 .

59 none no change 7748 71* delete question delete missed question 7747 87 accept b or c change from incorrect to correct 78/97 = 80.4 %

• Question 35 (SRO 34) was contested by another candidata. The panel accepted 2 answers.

Question 53 (SRO 52) was contested by another candidate. The panel accepted 2 answers.

Question 56 (SRO 55) was contested by another candidate. The panel deleted the question.

Question 71 (SRO 71) was contested by another candidate. The panel deleted the question.

Mr. O'Hern's original grade was 76/100 (76%). Incorporating the NRC analysis previously discussed, the panel concludes that the written examination should be scored as satisfactory and the original denial overtumed.

Final Result: Overtum.

4 i

. ee, e i ** seneemm== = = = = = *

• e . ..e e *= + e ame e 12

3 .,.

t Applicants' questions during NRC Written Esam 4/6/98

\

1 RO M1/SRO N0 '

Q. If1 believe that something happens at the value stated in the question (6.8 psig), and it appears ac l one of the answers, does the 'next ibaction' include that value or abould it beibe next setpoint that l l

. Is in the answes?

A. He question should be answered whh the 'next' value beyond that value stated in the questic j 1

Q. Can I assume that anything that occurs at 6.8 psig has already occurred? .,

~'

A. Yes,you can assume that.

l i

2- RO N2 Q. Is that 1 or 4 MSIVs? l A. It would be a Miisolation.

3- RO #77 Q. Is the figure drawn in accordance with normal convention? Are allthe contacts shownin a de-

===i=i state?

A. Yes.

Q. Do the dotted lines on the figure represent the current state of the associated contact?

A. De dotted lines represent an interrelationship and is not indicadve of the contact state.

4- RO #97 i

Q. What can I assume about the E11 F0!0 valve position? It is normally based on NASS direcdos.

A. Does the question provide direct *on that operater action has occurved?

5 RO #12 Q. Should 'ialtiate' really be 'liect'?

A. He question h correct as written.

6 SRO #21 Q. HPCI use is not required. Should I assume SRVs are not availabic?

A. Based on the geestion and conditions, select which answer appears best.

~

7 SRO #33 Q. Does this huply under normal or abnormal conditions?

A. Undernormalconditions.

Exhibit 4

2 ,

- 1 , , .

e . wg 9, ,,.

r ,

8. RO #86 Q. Does the fire in 2PC3 17 resuh la a complets loss of 130vec?

A. 'Ibe Sre in 2PC3-17 result in aloss Ibe affected 130vde.

9- ROW 74 Q. ShouldI assume B Loop is at normal operating parameters?

A. B Loop is la service.

10- no #s2 Q. Are you asking what actions are required to termbate blowdown femt?

A. What actions are required to terminste blowdown at this time? ,

e Y

Anoendix E APPENDIX E p POLICIES AND GUIDELINES FOR TAKING NRC EXAMINATIONS i

Each examinee shall be briefed on the policies and guidelines applicable to the examination 4 category (written and/or operating test) being administered. The applicants may be briefed

individually or as a group. Facility licensees are encouraged to distribute a copy of this appendix to every examinee before the examinations begin. All items apply to both initial and requalification examinations, except as noted.

PART A - GENERAL GUIDELINES

1. [ Read VerbatimJ Cheating on any part of the examination will result in a denial of your

- application and/or action against your license.

2. If you have any questions conceming the administration of any part of the examination, do not hesitate asking them before starting that part of the test.

4. SRO applicants will be tested at the level of responsibility of the senior licensed shift position (i.e., shift supervisor, senior shift supervisor, or whatever the title of the position may be).

5. You must pass every part of the examination to receive a license or to continue 1 performing license duties. Applicants for an SRO-upgrade license may require remedial training in order to continue their RO duties if the examination reveals deficiencies in the required knowledge and abilities.

6. The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC management. Grades provided by the facility licensee are preliminary until approved by the NRC. You will be informed of  :

the official examination results about 30 days after all the examinations are complete.

PART B - WRITTEN EXAMINATION GUIDELINES

1. [ Read VerbatimJ After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.

2. To pass the examination, you must achieve a grace of 80.'00. percent or greater. Every question is worth one point. ,

3. For an initial examination, the time limit for completing the examination is four hours.

For a requalification examination, the time limit for completing both sections of the NUREG-1021 1 of 5 Interim Rev. 8, January 1997 Exhibit 5

1 l

Anoendix E examination is three hours. If both sections are administered in the simulator during a single three-hour period, you may return to a section of the examination that was already completed or retain both sections of the examination until the allotted time has expired.

4. You may bring pens and calculators into the examination room. Use only black ink to ensure legible copies. ,,

5. Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.

6. Mark your answers on the answer sheet provided and do not leave any question blank.

Use only the paper provided and do not write on the back side of the pages. If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change.

7. If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor only.

( 8. Restroom trips are permitted, but only one applicant at a time will be allowed to leave.

Avoid all contact with anyona outside the examination room to eliminate even the appearance or possibility of cheating.

9. When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work ic your own and that you have neither given nor received

. assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.

10. After you have turned in your examination, leave the examination area as defined by the l proctor or NRC examiner, if you are found in this area while the examination is still in l progress, your license may be denied or revoked.

l 11. Do you have any questions?

i PART C - GENERIC OPERATING TEST GUIDELINES (CATEGORIES A. B. AND C)

! 1. If you are asked a question or directed to perform a task that is unclear, you should not hesitate to ask for clarification.

2. The examiner will take notes throughout the test to document your performance, and sometimes the examiner may take a short break for this reason. The amount of note-NUREG-1021 2 of 5 Interim Rev. 8, January 1997

l l

Anoendix E l

taking does not reflect your level of performance. The examiner is required to document i satisfactory as well as less than satisfactory performance.  !

l

3. The operating test is considered "open reference." The reference material that is  !

normally available to operators in the facility and control room (including calibration l curves, previous log entries, piping and instrumentation diagrams, calculation sheets, l and procedures) is also available to you during the operating test. However, you should know from memory certain automatic actirms, set points, interiocks, operating characteristics, and the immediate actions of emergency and other procedures, as appropriate to the facility. If you desire to use a reference, you should ask the examiner if it is acceptable to do so for the task or question under consideration.

You may not solicit technicallnformation from other operators, engineers, or technical advisors.

4. You must not discuss any aspect of your operating test with any other examinee until after all the examinations are complete.

PART D - WALK-THROUGH TEST GUIDELINES (CATEGORIES A AND B)

1. The walk-through test covers control room systems, local system operations, and administrative requirements. The examiner will evaluate each of these areas using a combination of job performance measures (JPMs) and specific questions. j The initial walk-through consists of ten JPMs for RO and SRO(l) applicants and five for SRO(U) applicants. Seven of the JPMs (two or three for upgrade applicants) will be conducted in the control room or simulator and the remainder will be conducted in the plant.

The requalification walk-through consists of five JPMs total, with at least two in the control room / simulator and at least two in the plant.

2. The examiner is a visitor at this facility. When you enter the plant, you may be expected to escort the examiner and ensure that he or she complies with safety, security, and radiation protection procedures. I

3. You should not operate plant equipment without appropriate permission from the operating crew. Nothing the examiner says or asks will be intended to violate this principle.

4. Before beginning each JPM, the examiner will describe the initial conditions, explain the task that is to be completed, indicate whether the task is time-critical, and explain which steps are to be simulated or discussed. You should perform or simulate the required actions as if directed by plant procedures or shift supervision. Do not assume that NUREG-1021 3 of 5 Interim Rev. 8, January 1997

~

Anoandir E the examiner will accept an oral description of the required action unless the examiner indicates otherwise.

5. Time-critical JPMs have been validated by your facility and must be completed within the predetermined time interval in order to obtain a satisfactory grade for that JPM. You will be permitted to take whatever time is necessary to complete those JPMs that are not time-critical, provided you are making reasonable progress toward achieving the task standard.

6. When performing JPMs, you are expected to make decisions and take actions based on the facility's procedural guidance and the indications available. Some of the tasks that the examiner asks you to perform will require the implementation of an attema+Jve method directed by plant procedures.

7. As part of the initial examination, the examiner will ask follow-up questions at the end of each JPM to investigate your knowledge of the applicable system or task. Many of the questions will require you to use plant reference material, while others should be answered without the use of references. If you need to consult a reference to answer a question, ask the examiner if it is acceptable to do so. There is no specific time limit for any question, however, you may be evaluated as unsatisfactory on a question if you are unfamiliar with the subject or reference material and are unable to answer the question in a reasonable period of time. You will not be permitted to conduct unlimited searches of the plant reference material during the examination.

Although the requalification examination does not include prescripted follow-up questions, the examiner may ask questions as necessary at the end of any JPM to clarify your performance.

8. To facilitate the examination, please inform the examiner when you consider your performance of each JPM and your answer to each question to be complete.

9. If you need a break during the test, you should ask the examiner. l

10. Do you have any questions before we begin the walk-through test?

PART E - SIMULATOR TEST GUIDELINES (CATEGORY C)

1. Your primary responsibility is to operate the simulator as if it were the actual plant. If you believe that the simulator is not responding properly, you should make decisions and recommendations on the basis of the indications available, unless directed otherwise by the examiner.

2. If the examiner asks you a question, you should answer it only if doing so will not interfere with simulation facility operations.

NUREG-1021 4 of 5 Interim Rev. 8, January 1997

h i

~

Anoendix E

3. Teamwork and communications are evaluated. You can enhance the evaluation process by vocalizing your observations, analyses, and the bases for your actions.

Requalification examinations evaluate the crew's ability to safely operate the plant and the performance of both the individuals and the crew.

4. If you recognize but fail to correct an erroneous decision, response, answer, analysis, action, or interpretation made by the operating team or crew, the examiner may conclude that you agree with the incorrect item.

5. You should keep a rough log during each scenario that would be sufficient to complete j necessary formallog entries.

6. A designated facility instructor (or an examiner) will act as the auxiliary operators, radiation health and chemistry technicians, maintenance supervisors, plant management, and anyone else needed outside the control room.

i

7. The facility instructor (or examiner) will provide a shift tumover briefing before the l scenario begins. The briefing will cover present plant conditions, power history, equipment out of service, abnormal conditions, surveillances due, and instructions for the shift.

8. Control board switches may be purposely misaligned to enhance a scenario or transient where appropriate. You will not be required to locate misaligned switches as part of the evaluation. If a switch is misaligned, it will be tagged or otherwise highlighted as appropriate to the facility and will be noted during the tumover briefing. The examiners will not misalign switches during the scenario.

1

9. Time compression may be used to expedite the sequence of events in some scenarios, i but it wil! not preclude you from performing the actions that you would typically be  ;

required to perform in response to the events. If time compression is used, you will be so informed during and after the scenario.

10. You will normally be given about five minutes to familiarize yourselves with plant conditions before starting each simulator scenario.

11. The initial test will normally consist of two or three scenarios lasting a total of three to four hours. The requalification test will normally consist of two scenarios lasting about one hour each. You will be given a short break between scenarios.

12. Do you have any questions before we L ; gin the simulator test?

NUREG-1021 5 of 5 Interim Rev. 8, January 1997

~ ~ ~ ^ ~ ~

4 - ~

~~

COPY TO S. SHANKMAN, NRR/LHFB cn* 8/30/89

^

Detrolf r.rm 2 Edison EN" W g.n, May 27, 1987 NIC-87-0077 U. S. Ibclear Regulatory Comnission

, Attn Document Control Desk Washington, D. C. 20555 f

References:

1) Fermi 2 NIC Docket No. 50-341 NIC License No. IPF-43

- 2) Detroit Edison Letter EF2-67,194 dated March 14, 1984

3) 10CPR55 Operators' License

Subject:

Operator and Senior coerator Reaualification Trainina Procram Reference 2 transmitted the Detroit Eison Licensed Operator Recpalification Program for NIC review and approval. Since that submittal, the NIC has revised 10CFR55, Operators' Licenses which became final May 26, 1987.

Based on the NIC's enhancements to Part 55, Detroit Eison bm updated the Fermi 2 Licensed Operator Recpalification Program. This program was inplemented effective May 26, 1987 and supercedes our previously

. submitted aequalification Program.

Inplementation of this revised program not only conplies with NIC regulations but'ahincorporates a' systems' approach to tralr.ing7

^

. .Detroit Elson's Licensed OpeiatoOhnd" Senior Operatdf Scraining'p prograns were' accredited by the' Institute'of teclear Power Operption (ItPO) on Decenber 18,19857 'Ib obtain accreditation, the Detroit

! Mison Conpany has demonstrated to the National teclear Accrediting Board that its training programs are based on a systens approach to training and meet ItPO accreditation criteria. 'Ihe Licensed Operator and Senior Operator Training programs have been evaluated by ItPO to >

provide an acceptable method to maintain and enhance the performance and professionalism of all licensed personnel to achieve the high Exhibit 6

o--- --

. US'IC N

May 27, 1987 NIC-87-0077 Page 2 operational standard required .to ensure nuclear plant safety and reliability.

Please contact Mr. Steven R. Frost at (313) 586-4210 if you have any further questions.

Sincerely, w -

k~ . {,s .

F. E. Agost , Vice President Maclear Operations I

l ocs A. B. Davis E. G. Greemnnn W. G. Rogers J. J. Stefano USNIC Region III

)

. e,;

Fermi 2 Operations Department Instruction .

Detroit ODI-022 Revision J_0 Edison D=ft Reactivity Management Approved By: H. Garrett /s/ ,Dc:: Cchb4s/

Date: 9/26/97 .3/12/97 1.0 Purpose To define Management's expectations when performing reactivity manipulations. Properly implemented, the guidelines set forth in this document will ensure that reactivity changes are performed in a carefully controlled, deliberate manner while constantly monitoring nuclear instrumentation and other redundant indications of reactor power level. These instructions will also define the practices of controlling reactivity changes and conservatively applying principles of manipulation to maintain a safe operating margin. These guidelines prevent challenges to the Operators, fuel integrity, and plant systems as well as ensure operation within design limits, analysis limits, and Technical Specifications.

2.0 Description Per the requirements of 10CFR55, Licensed Operators control the reactivity and power level of the reactor. From this perspective, the Operations Department " owns" the core as it does all plant systems. From a System Engineering perspective, Reactor Engineers are the system engineers for the core; therefore, Reactor Engineering has ownership of the core in terms of system / functional responsibility. It is therefore the shared responsibility of Operations and Reactor Engineering to ensure proper operation and control of the reactor.

Reactivity Management extends beyond the core to include all systems that are used to control, affect, or monitor j reactivity. Ensure that positive control and monitoring of the reactor is maintained during Reactivity 1 Manipulations by employing / understanding the following  ;

OPERATIONS M4N,f GEMENT OVERSIGHT l

• It is the responsibility of shift operations management for controlling the reactor core. t =h,Ithe NASS l shall concur withc=u- Jtha have direct authority over, and provide immediate oersight for all evolutions l that could affect reactivity. The NASS shall be in the main control room "at controls area" when any i evolutions that affect reactivity are in orocress. _TThe NSS shall provide general oversight for reactivity manipulations while maintaining a broader perspective of overall plant operations.

• Reactor safety and core integrity take precedence over power production and all other associated activities.

• The NASS shall supervise the approach to criticality.

• Frequent breaks shall be taken during long periods of control rod movement such as during reactor startups or rod pattern adjustments. Frequent breaks / rest stops will allow the operator and verifier to change their focus, stretch, etc. thereby ensuring personnel remain mentally and physically alert. )

• Reactor Engineering shall be present for all reactor criticals and all significant control rod movement such as l reactor startup/ shutdown, rod pattern adjustments, scram time testing, etc.

• Reactor Engineering shall be notified of all unplanned core reactivity changes and abnormalities.

REACTOR MONITORING e Operators shall monitor periodic core performance parameters for Technical Specifications compliance (i.e.,

power level, thermal limits, proximity to the stability awareness regions, rodline, etc.). Monitoring frequency should be increased following changes in core power or flow, and during Xenon transients.

• Operators shall anticipate, control, and respond to plant parameters in order to maintain the reactor in the desired condition.

Exhibit 7

. s Operations Cenduct M:nurl ODI-22 Reactivity Men:gement Revisi::n J,0 l Page 244 of i

433 l

l CONTROL ROD MOVEMENT Written instructions shall be provided for all planned control rod movements. These are provided by either the

[ Station Nuclear Engineer (SNE) or the Shift Technical Advisor (STA) and may include the Pull Sheet Books,

! Post Low Power Setpoint Pull Sheets, maneuvering plans generated by Reactor Engineering, or procedures.

• The operator assigned to manipulate control rods shall have no other collateral duties while cor. trol rod manipulation is in progress.

• Whenever a planned control rod movement is being performed, a second licensed operator, STA, or SNE shall be present to perform the function of Rod Movement Verifier-k addh:=, Ped W c=n Vaif:=h=!d beed_ ,Wwhen inserting the control rodsGRAM-Anay for emergency power reduction. the reauirement for rod movement verifier may be waived by the NASS.-

• Sufficient time must be allowed between rod maneuvers to allow for reactor pressure and power response.

• Control rods that are moved past their intended position by one notch and identified immediately, although not considered a "mispositioned control rod, shall be communicated to Reactor Engineering and the NASS/NSS for tracking purposes.

• Control rods shall not be moved without oversicht by the NASS.

• If control rod movement results in achievine an tmolanned suberitical condition. the shutdown must be continued. Contml rods shall not be simpiv withdrawn to re-achieve criticality from an unnlanned suberitical condition.

CONSERVATIVE REACTH7TYMANAGEMENT

• 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 I include rod insertion to lower power, or a reactor scram without hesitation, whenever such unanticipated or

. unexplained behavior is encountered.

• Shift tumovers shall not occur during startups or shutdowns unless the plant is in a stable, non-transient state.

As such, careful consideration should be given to the starting time for reactor startups and shutdowns in order J to avoid being in a susceptible condition (e.g. very close to criticality, POAH, etc.) during turnover.

• Planned control rod movements in Mode 1 or Mode 2 shall be preceded by a shift brief. This brief will include a review of all individual responsibilities, the instructions to be used, expected reactivity response, any increased potential for plant transients to occur, and immediate actions to take if a transient does occur. Any changes to the original plan shall be approved by the Supervisor, Reactor Engineering / Delegate and the NASS prior to implementation.

• Distractions in the control room shall be minimized. If the NSO feels that other activities are distracting, control rod moven:ent shall be stopped until proper decorum is restored.

• Reactivity changes should be limited such that performance of only one activity at a time that could affect core l reactivity is permitted. Examples of activities that could affect reactivity include control rod movement, I changes to core flow, placing feedwater heaters in/out of service, pressure regulator testing, turbine valve  ;

surveillances, heater drain pump or heater feed pump starts or stops, etc.

• Positive control over core reactivity must be maintained by the operator at all times. i e The mechanics of effective communications are outlined in Operations Department Instruction ODI-006. ,

Clear, concise communication between the operator and verifier is imperative. If there is ever a question .

i regarding control rod movement or instrumentation response, the evolution s!dl be stopped and l communicated appropriately. Only after the situation is fully understood and resolved shall control rod l movement resume.

• All Licensed Operators are responsible for shutting down the reactor when they determine that the safety of

the reactor is injeopardy.

!

• All Licensed Operators are responsible for shutting down the reactor when reactor protection setpoints have been exceeded and automatic action has not occurred.

• Operation of reactivity controls shall be done only by Licensed Operators per 10CFR55, and only with the r knowledge and permission of the NASS.

1 I

l

!- v..

l Operati::s Cecduct M::url ODI-22 Re ctivityMra:gemext Revisisc10 Page4 of 123 Licensed Operators In Training may not operate reactivity controls unless they are directly supervised by a - l l Licensed Operator per 10CRF55.

                                                                                                                                             )

i i l l I I I I. i I r-l l I i l i < l s l l-l

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

s Opzrati:2s Cerduct Mcau:1 ODI-22 Rnctivity Menigemert Revisi:n le Page f34 of l 433 e When the General Operatinn Procedures for Load reduction and Reactor shutdown have been entered any manipulation that would add nositive reactivitv to the core must be authorized by the NSS and SNE/STA.

e Durine a reactor shutdown. reactor power shall be held at the noint of 10 to 15% bvnass valve position until

the decision has been made to shutdown the olant Once the decision has been made to shutdown the plant I

and the annropriate section of the GOP is entered. this is to be considered the point of no return and the l reactor shall be shutdovn":- 'he 0 ca! ^ c "ne P : Jure f " = - "-edc"  : !de" . =d

derren~c- "-- h= 'ren ='n

d. 'e R: , hue -- --: de -!=:

         "2 Ind= ry Enpri=:: h= der- th:             h= = p!=' :: in = u= tab!: cr upx: :: d!!:en, cdding ped v:

reatlivity :: :::bi!!= the p!=* is ra::!y :===fu!. P=i: v: := ti ::y addi ! = = = :n== cf:::b!!!:ing p!=t rend!!!:= deu!d be avcid:dOnerators shall anticinate criticality anytime reactivity is added to the core. Four countrate doublings have been determined to be an accurate indicator of oroximity to criticality. Fermi does not perform Estimated Critical Position (ECP) calculations due to uncertainties which make calculating an accurate ECP (e.c.. moderator temperature coefficient. Xenon reactivity and distribution. fuel bundle reactivity as a function of exposure) difficult. The ECP could unduly bias an operator's decision making process which should be based on reactor response and instrumentation indication., I I I 1 l { l i 1 I

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

A

     %y Operatic:s C:nduct Marual                                                                                 ODI-22 Re ctivity Manrgeme:t                                                                                     Revisi:n le           -

Page 534 of 19

• Fer-! dcc no: pe-fer: E::!::::d Cr! !=! %:::ica (ECP) =!=!::!c= f=
• c/c == =: 1) ":= := ; numb .'

ef un::=i:-:!= c hich m h: =!=!::ing an :=um:: ECP d!5=!: (:.g., mcda::= :: .p =:un =:5:i=:, X=c = ti ::y =d di :r!bu !cn, fu:1 bund!: =:ti.::y = f=::!cn -f =p==); =d 2),^ n ECP =n unduly h!= = cp:=:c '; d=i;ic: making p===; cS!:h :heu!d b: b=:d en ==:= repc=: =d i-===:::!cn ied+eadens:

• Differina ceinions regardine reactivity manipulations shall be broucht to the attention of the NSS/NASS and resolved by the shift team before continuine with the manipulation. -
• Industry Experience has shown that when a plant is in an unstable or unset condition. adding positive reactivity to stabilize the olant is rarely successful. Positive reactivity additions as a means of stabilizine olant conditions should be avoided.

                '=:=d cf ECP , fc= ==* =:: dcub!!ng: h= 5:= :fhti=!y u=d ic ind!=:: :!=c p:9ity :: c :::c!!:y.

Thi: m :hed, H:/.n c, h=!d nc: p=:!ui the cp=:= f== =:!:!p 'ing ==:= =::i=!ity. : y " c. 3.0 Review Conservative decision making with respect to reactivity management is an important aspect in maintaining the health and safety of the public as well as providing for safe and efficient operation of Fermi. Reactivity manipulations shall be conducted in a formal manner with rigorous attention paid to communications, self checking and peer checking by the verifier. It is essential that the operators and verifiers are allowed to focus their attention on the task at hand. By implementing the guidelines as outlined in this instruction, we are striving to minimize / eliminate personnel errors, and provide a means ofidentifying adverse trends before problems occur. l l 4

i i l} Fermi 2 j 3} Operatians Dep2rtment Instructi::n o l ] Detroit ODI-007 Revision 2 l Edison i Command and Control l 1 1 Approved By: D. Cobb /s/ Date: 04/29/97 i l l 1.0 Purpose j Command and control, within the confines of the operating crew, is intended to provide oversight and direction  ! during shift routines and evolutions. Command and control will ensure that the shift crew is functioning as a team, and that each of the team members is cognizant of his responsibilities. By establishing and utilizing these guidelines on a daily basis and in a training setting, the crew will be successful in the normal operation of the facility and in mitigating unexpected plant transients. 2.0 Command and Control Structure The command and control function starts with tip Nuclear Shift Supervisor (NSS) and continues with the Nuclear Assistant Shift Supervisor (NASS). These two Senior Reactor Operators are responsible for the safe operation of the reactor and for the implementation of procedures and instructions to ensure compliance with all operating and regulatory limits. Normally, the Nuclear Assistant Shift Supervisor is tasked with personally directing activities in the control room and in-plant operations. The Nuclear Shift Supervisor establishes an oversight role, monitoring activities, establishing priorities, bringing to bear the necessary resources to support the control room, notifying plant management, and implementing the notifications required by the emergency plan. For complex or infrequent evolutions, or compound abnormal conditions (multiple AOP's etc.), the Nuclear Shift Supervisor must provide the prioritization of activities to the Nuclear Assistant Shift Supervisor (and crew) which will then be directed by the Nuclear Assistant Shift Supervisor. The Shift Technical Advisor supports the Nuclear Shift Supervisors and the Nuclect Assistant Shift Supervisors. The Shift Technical Advisor is not to provide direction in the operation of the facility, but may relay information and requests to supporting groups at the request of the Nuclear Shift Supervisor and Nuclear Assistant Shift Supervisor. During transient conditions, the Shift Technical Advisor remains cognizant of plant trends and decisions made by the senior operators so as to provide engineering related advice. When required to implement the emergency plan, the Shift Technical Advisor performs the duties outlined in the Emergency Plan and provides a backup to the Nuclear Shift Supervisor relative to event classification. Directly responsible for the manipulation of control room controls and the supervision of field activities are the Nuclear Supervising Operators, CRNSO and P603 NSO. The Nuclear Supervising Operators (NSO's) need to communicate with (normally) the Nuclear Assistant Shift Supervisor and receive direction (normally) from the Nuclear Assistant Shift Supervisor. The P603 Operator is the individual that maintains direct monitoring and control responsibilities for reactor power, level, and pressure. The Nuclear Supenising Operators will direct or supervise the activities of the plant operators and those personnel assigned by the Nuclear Assistant Shift Supervisor to support the plant's activities. The Nuclear Power Plant Operators are the control room's hands, eyes, and ears out in the power plant. They have operating responsibilities that involve the effective operation and monitoring of plant equipment, remaining cognizant of activities (maintenance and testing) in progress within their assigned watch station, and for notification of the Nuclear Supervising Operators of out of specification conditions or items of concern. l Exhibit 8

l? d Operati:ns Departme:t I:structi:n ODI-007 h . Ccmmard crd C= tral C !&h Revisiale l . Page 2  ! l 3.0 Communications I The mechanics of effective communications are outlined in Operations Department Instruction ODI-006. The l communication triangle is the concept that ensures the flow of communication mimics the command and control structure presented previously. This process is pictorially represented in Attachment 1 of this instruction. Attachment 1 is a guideline to illustrate the normal flow of shift communications. This guideline may be modified by the Nuclear Shift Supervisor, after careful consideration, to meet the needs of a planned evolution or the status of the power plant. 4.0 Briefs Briefs are an effective tool used to ensure crew members are aware of plant and equipment status as well as their individual responsibilities for evolutions to be performed or in progress. Three types of briefs used at Fermi 2 are:

• IPTE for " Diagnostic, Special and Infrequently Performed Tests or Evolutions", the highest structured brief performed in accordance with MES31
• Crew Briefs performed for unplanned, event based transients and evolutions e Pre Job briefs for all planned jobs that require exrdination, change plant conditions, or present personnel safety hazards.

1. IPTE Briefs When a planned evolution involves any of the following, refer to MES 31:

• Complex systems or components important to safety or affects plant capacity.
• Scope crosses a sensitive interface that could compromise plant safety or pose a hazard to public health and safety.
• Test or evolution has never been performed and could have impact on plant personnel, plant capacity or public safety, e Infrequent periodicity (normally considered to be once per fuel cycle) and personnel or equipment used have significantly changed or are no longer available.

     . Governoring procedure (troubleshooting, surveillance, SOE, AOP, EOP) has been changed more than three times while activity is in progress.

• Previous procedure changes made while activity was in progress are changed again.
• Plant Manager directs that IPTE be used to employ heightened management oversight.

2. Pre-Job Briefs - refer to ODI-37 for specific guidance A Pre-Job Briefis a preparatory action conducted before performing a planned operating, maintenance or testing task. Each brief will contain the essential elements as a minimum to ensure successful completion of the task.

Normally, a pre-job briefis conducted with assigned team munbers, but for a one-person task it may simply be a written or mental checklist applying the essential elements before starting ajob. ! All planned, non-routine jobs that require coordination, change plant conditions, or present personnel safety hazards will require a pre-job brief. A checklist (att 2)is provided to ensure that the essential elements of a brief l are discussed as a minimum. Attachment 2 for pre-job briefs has been coverted to a placque to replace the evolution brief placque in the Main Control Room and the Simulator.

A jj Operati=s Depirtment 1:structim ODI-007 i I p l Cimmad c:d Centr:! C !&!! :: Revista 20 ' Page 3 l 1

3. Crew Briefs - Use the placque in the Main Control Room and the Simulator for Crew Briefs Following the start of an unplanned, event based transient or evolution, Crew Briefs are used to I communicate plant status and identify priorities. During transients or unplanned evolutions, briefs are necessary to ensure the operating team understands present plant conditions present and the course of l

i action that will be taken. l Crew briefs can be requested by any crew member and are usually initiated by the Nuclear Assistant ShiR Supervisor. Important characteristics of a crew briefinclude: The announcement of the brief to focus the crew member's attention. The conduct of the brief by the Nuclear Assistant Shift Supervisor who provides pertinent information in a clear, concise format. An opportunity for input or questions from the other members and clarifica' ion ofitems by the Nuclear Assistant Shin Supervisor or Nuclear Shift Supervisor.

• A clear cut end of the brief and resumption of crew activities Generally, no specific action should be directed from the brief setting, rather once on station the guidelines for operational communications should be used to conduct the evolution.

1 i i l

Plant Technical Procedure - Fermi 2 22.000.02

 'A   General Oper:ti:n Procedure                                                            Revisi:n 39 5                                                                                                    Page1 l                                 PLANT STARTUP TO 25% POWER                                             l Revision Summary

1) Revised section 7, to allow warm up of MSRs with N30-F006 welded open.

l l Implementation Plan

1) This procedure goes into effect upon approval.

2) "A summary of this revision will be placed in Operations Required Reading Package.

3) No further training is required.

l l- Continuous Use

This procedure SHALL be performed as written. Each step shall be read by the user l before performing that step; performed in the sequence given; and when required, signed off as it is completed before proceeding to the next step.

L l f Attachments 1 082395 N/A Step Comment Form Enclosures A 072897 Recommended Feedwater Inlet Temperature information and Procedures , DSN Revision DCR# DTC File t' l 22.000.02 39 97 2198 TPNPP 1703 02 IP Code Date Approved Released By Date Issued Recipient 1 10/20/97 T Cox/s/ 10/20/97 l l Exhibit 9

22.000.02 O Revision 39 i Page 6 3.0 PRECAUTIONS AND LIMITATIONS 3.1 The following Precautions and Limitations apply at all times when this procedure is in use: 3.1.1 Use EMERGENCY IN mode ofinserting Control Rods only if emergency rapid power reduction is required.' Unusually high hydraulic forces may be developed in the Control Rod drives due to the settle function of the timer being bypassed. 3.1.2 With one reactor coolant system loop not in operation with THERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify and log the following differential temperature requirements are met within no , more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:

1. Less than or equal to 145 F between reactor vessel steam space coolant and bottom head drain line coolant, and NOTE: The following two steps do not apply when the Recirculation Loop not in operation is isolated from the Reactor Pressure Vessel.

2. Less than or equal to 50 F between the reactor coolant within the loop not in operation and the reactor coolant in the reactor pressure vessel, and

3. Less than or equal to 50*F between the reactor coolant within the loop not in  ;

I operation and the operating loop. 3.1.3 Due to Reactor Recire Pump Speed Controls drifting, during power ascension,  ; positive speed control should be closely monitored while Reactor Recirc Pump l Speed is maintained below Limiter #1. 3.2 The following Precautions and Limitations apply during Reactor Startup/Heatup and RPV i pressurization: 3.2.1 Startup following a scram from previous high power operation may result in unusually high Control Rod Notch Worths due to Xenon concentrations. 3.2.2 If a sustained Reactor Period ofless than 50 seconds is indicated, take immediate action to correct this condition by inserting Control Rods until a Reactor Period of greater than 50 seconds is indicated. Contact the Nuclear Shift Supervisor and the l Station Nuclear Engineer prior to resuming withdrawal of Control Rod (s). ( l 3.2.3 Limit the blowdown rate through Reactor Water Cleanup (RWCU) to prevent the RWCU Filter /Demineralizer Inlet Temperature from exceeding 130 F.  ! l l I

, 22.000.02 Revision 39 I Page 7 3.2.4 The Reactor is administratively limited to s 90'F/hr. heatup rate to avoid exceeding the s 100'F/hr heatup rate of Technical Specifications, Section 3.4.6.1. 3.2.5 If this is the first Reactor startup following core alterations or a Reactor shutdown of greater than 120 days, perform 54.000.03, " Control Rod Scram Insen Time Test," prior to exceeding 40% power. 3.2.6 Following maintenance or modification to a Control Rod or Control Rod' System which could affect Scram Insertion Time, perform 54.000.03, " Control Rod Scram Insert Time Test." 3.2.7 In order to allow continuous monitoring of Reactor Power during startup, do not withdraw all SRM Detectors simultaneously until the point of adding heat is reached. 3.2.8 With the MSIVs closed,24.202.02, "HPCI Flow Rate Test At 165 Psig Reactor Steam Pressure," and 24.206.04, "RCIC System Automatic Actuation And Flow Test," cannot be performed. If MSIVs are closed and either of these surveillances are due, do not raise Reactor Pressure above 130 psig. 3.2.9 During the approach to criticality when the SRM count rate has doubled four times, Control Rods at Positions 00 through 24 shall be moved in the Notch Mode - only until the point of adding heat. 3.2.10 44.080.501, "Off Gas Hydrogen Monitoring System - Channel Functional Tests And Channel Calibrations," shall be performed prior to starting Off Gas System, or enter Technical Specifications, Section 3.3.7.12, Action Statement. CM 3.2.11 Prior to RPV pressure exceeding 100 psig, RPV water level shall be controlled using the lowest indicated level on the RPV Narrow Range Water level instruments (160 to 220 inches) on H11-P603. 3.2.12 Whenever the Reference Leg Back Fill System is in service, the Back Fill Flow must be monitored shiftly to ensure it is between 20 and 35%. 3.2.13 IF the Reference leg Back Fill System is in service, an I&C technician must be available at B21-P408A and B to adjust the back fill flow during RPV pressurization to maintain the back fill flow rate at 20 to 35%. This will prevent Reactor Water level and pressure transients inadvertently induced by back fill flow.

(' 22.000.02 Revision 39 Page 8 g 3.2.14 The Reactor Mode Switca shall net be placed or maintained in RUN if the minimum number of operable APRM channels cannot be maintained above their i downscale setpoints. 3.3 The following Precautions and Limitations apply during Main Turbine Startup and Power  ! Increase to 25%: 3.3.1 Following each power change exceedint 25% in one hour, request Chemistry to perform Startup Gaseous Effluent Power Change surveillances in accordance with 74.000.18, " Chemistry Shiftly,72 Hour, and Situational Surveillances," Attachments 12 and 15. 3.3.2 Failure to maintain Reactor Flow Limiter greater than Reactor Power, could result ) in a High Pressure Scram. q I 3.3.3 Transformer 2A and 2B 345 kV Disconnects CI-A or CI-B will not close i electrically unless Tie Breaker between Generator 2 and Brownstown-Enrico  ; Fermi 3, CM and Generator 2 Breaker CF, are open. j 3.3.4 Feedwater temperature entering the Reactor must be 2 245 F prior to exceeding l 25% Reactor power. At power levels 2 25% the Feedwater temperature entering l the Reactor must be greater than or equal to the value shown on the curve in Enclosure A. 3.3.5 Do not exceed 25% Reactor Power prior to verifying 3D Monicore Core Monitoring Programs required at 2 25% Power are running properly. 3.3.6 Reactor Vessel level control should remain on SULCV until operating Reactor Feed Pump Suction Flow is greater than 35%. This will prevent oscillations of the Reactor Feed Pump Minimum Flow Valve and Reactor Water Imel. 3.3.7 Operating with a deviation between the reactor heat balance and its validators outside of approved guidelines requires:

1. Nuclear Shift Supervisor and Supervisor, Reactor Engineering to be notified j

2. Power ascension to be terminated pending investigation of the deviation l

3.3.8 . Minimize the time that the Bypass Valves are open between 30 to 45%. This region causes high vibration that leads to high stress. L 3.3.9 Minimize tir, ; Reactor Power is between 20 and 45%. This region causes

l. increased ste am line loop vibration resulting in increased stress to steam loop cotaponents.

l.

22.000.02 Revision 39 ? 4 Page 9 3.3.10 The Turbine Bearing Vibration Trips are automatically defeated until Main Turbine Generatoris synchronized. 3.3.11 Low Pressure Turbine casings are limited to temperatum changes of 150 F/hr. If temperature changes are approaching 120*F/hr, stop power / flow changes and allow the Turbine to soak. Failure to soak the Turbine could result in a rub causing high vibration. END OF SECTION a-- _

22.000.02 Revision 39 , Page 35 l3 RPV PRESSURIZATION Step No. Initial /Date CAUTION l , The Reactor Mode Switch shall not be placed or maintained in RUN if the minimum number of operable APRM channels cannot be maintained above their downscale setpoints. ~: I 6.2.28 When Reactor Pressure is stable at 944 to 949 psig, perform the following to enter Mode 1:

1. Review Mode Change Report to ensure all required O surveillances have been performed to enter Mode 1.

2. Review 24.000.02, "Shiftly, Daily, Weekly And O Situation Required Surveillances," and perform any requirements for Mode 1 not already performed.

3. Check Powemet WST Program for LCO Associated O Work Codes and verify there is no outstanding work that would prohibit entry into Mode 1.

4. Review LCO Sheets to ensure there O are no outstanding Limiting Conditions for Operation preventing entry into Mode 1.

l 5. Verify Reactor Power is between 5% and 10% as follows:

a. Request the Station Nuclear Engineer to verify O each operable APRM is reading greater than or equal to the powe'r level associated with the Power vs. Average Bypass Valve Position Graph in the Reactor Engineering Data Book.

b. Request the Station Nuclear Engineer to perform O 54.000.%, "APRM Calibration," to adjust any APRMs that do not satisfy step 6.2.28.5.a.

i l'

22.000.02 Revision 39 it Page 36 RPV PRESSURIZATION Step No. Initial /Date

6. Verify downscale alarms are clear on the operable APRMs. O l

7. Verify the following indicators on the MSIV Isolation O Mimic Status Display (H11-P601) are clear: ,

a. MAIN STEAM LINE LOW PRESSURE CHANNEL A O .

l 1

b. MAIN STEAM LINE LOW PRESSURE CHANNEL B C I

c. MAIN STEAM LINE LOW PRESSURE CHANNEL C O

d. MAIN STEAM LINE LOW PRESSURE CHANNEL D C I

8. Pig e Reactor Mode Switch in RUN. O l

9. Verify Annunciator 3D87, MN STM LINE ISO VALVE O CHANNEL TRIP BYPASSED,is clear.

10. Place Reactor Parameter Display Switch in 1 PWR OP. O

11. Verify Mode Status on ERIS is RUN. If Mode Status is displayed O as unknown, manually input the correct Mode Status.

                                                                                           /

6.2.29 Notify Chemistry to perform Startup Gaseous Effluent Power Change in accordance with 74.000.18, " Chemistry Shiftly,72 Hour, And Situational Surveillances."

                                                                                           /                i 1

I i e

h 3 7.6.1.17.6.2 operator Information s Readout instruments are provided in the main control room to display and record the Division I and II control air pressures. Recorders register the automatic initiation of the control air system compressors. 7.6.1.18 Alternate Rod Insertion 7.6.1.18.1 Eauinment Identification ,, The alternate rod insertion (ARI) components of the CRD system are designed to mitigate the potential consequences of an ant:,cipated transient without scram (ATWS) event. The ARI components are redundant to the RPS. 7.6.1.18.2 Eauinment Desian 7.6.1.18.2.1 Initiatina circuits l3 There are three initiating signals used for the ARI logics, namely:

a. Reactor dome high pressure

b. Reactor low water level 2

c. Manual initiation in the main control room Any one of the above signals can initiate the divisional ARI logics as shown in Figure 7.7-3, Sheet 4. Additional immediate response to the initiation signal.s includes the recirculation pump motor generator field breaker trip (see subsection 7.7.1.2.3.1).

7.6.1.18.2.2 Legig Two divisional ARI logic systems are provided: Division I, consisting of logic channels A and C, and Division II for logic channels B and D. The signal to insert the control rods is generated in two separate divisions on two-out-of-two logic channels in a given division. 3 The ARI logic receives reactor done pressure and water level signals from the nuclear boiler system. The logic causes automatic energitation of the ARI solenoid valves when either the reactor high-pressure trip set point or low-water level 2 set point is reached. The ARI logic can also be initiated manually from the main control room. Each ARI logic channel is provided with a disarmed / armed pushbutton switch. Both pushbutton switches in a given division must be depressed to energize the ARI logic and initiate control rod insertion. The ARI initiation signals are ( designed to seal in the initiation logic to ensure completion of the ARI function until it is reset manually. A reset pushbutton f per division is provided in the main control room to clear the 7.6-75 REV 5 3/92 l Exhibit 10

a ARI logic. A timer is used in each of the ARI logic channels to , inhibit the reset function for approximately 30 seconds after the ) initiation signal is received. ,A 30-second time delay is selected to ensure completion of the ARI function before the logic can be reset. The initiation of the two' separate ARI logics results in the energization of eight class 1E de solenoid valves (four per division). Two of these, F160A and B, vent the scram air supply line just downstrema of the F110A and B backup scram valves.. (Refer to Figure 7.6-36). These ARI valves also act to block the supply of air to the scram header. Check valves F161A and B provide an air-flow path around the F160 valves in the event one or more of them fails. Four additional ARI valves, F162A, B, C, 3 and D, vent the A and B scram header to the atmosphere. As the header Hepressurizes, the scram valves at each hydraulic control unit will spring open scramming the rods. Two ARI valves, F163A and B, vent the scram air header to the scram discharge volume drain and vent valves, closing these valves and isolating the scram discharge volume. All eight ARI valves are normally deenergized. 7.6.1.18.2.3 Annunciation and Indication The manual initiation pushbutton switch in the main control room activates an annunciator window whenever it is placed in armed position. A separate annunciator window is activated upon

                                                                                                       )

initiation of the ARI logic circuits. The open and close position of the ARI solenoid valves are also indicated in the main control room. 7.6.1.18.2.4 Ientability Four separate ARI initiation logic channels are provided to. permit maintenance, repair, test, or calibration of all circuit devices (at power) up to but not including the final trip devices (ARI solenoid valves). Each ARI logic channel is provided with'a test jack and indicating lights to verify logic activation in any given l division. 7.6.1.19 Safetv/ Relief Valves 7.6.1.19.1 System Identification The nuclear pressure relief system is designed to prevent over-pressurization of the nuclear system that could lead to the failure of the reactor coolant pressure boundary. 7.6.1.19.2 Safetv/ Relief Valve Eeuinment Desian Safety / relief valves (SRVs) are dual-functioning types: automatic self-actuating and solenoid operated. The valves are self- j actuated when reactor pressure exceeds spring set pressures that are adjustable in range. The SRVs are divided into three spring-set-pressure groups. The first group consists of five valves set 7.6-76 REV 5 3/92 l

unranour,rmAnmn>w .nnrmw wmom cu: . - ,, . 5 w Exaniination Outline Form ES 201-2 ES-201 1 Quality Assurance Checklist Date of Enemination: 08/28JOS Feoany: Fernd 2 Teek Deecription infusie  ; Item ' a bi e t e VerWy that the outline (s) fit (s) the ;; ,,-; te rnodel per ES.441. M [f[

                $        b. Assess whether all als 'J.d: ';;: and four abluty ostegories are appropriately esmpled.
                                                                                                                             'M k0g' i

s. Assoas odiother the outino overemphostaes any systems, evolutione, or generic A toples. 'M [AJ pUr

d. Assess whether the .;; _ - from previous seeminetton outlines le escoeLive. 74 g,03 g O 8- e. Using Form EH016, we number of normal evoldiens, tiet the proposed soonerto este cover the required ment and component feitures, and maior trenoiente.

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                 '        b. Assess whether there are enough esenerlo este (and spores) to test the projected
                                                                                                                 ""n and u         number       end mk of appiteente in ecoordance with the expected crew e.licant een be rotation schedule v4the64 compromising enem Integrity; ensure each app tooted using et lesst one new scenerlo end noeneries will not be ropeeted over p         p suomessive days.

ro me e.,ioni ,cesa i., see.e. whevi.r ihe ousne(s) .ordoym in Appendk D. and guentitettwo orderte spee4 fled on Form ENC 14 and s8ee with v.e .uemeu , g g E e. VerWy that the eutilnt(s) contain(e) the required number of control room and in plant w toeke and verWy theit no more then 80% of the test tratertel le repeated from the lost NPtC p'$ g ', ewemination. 8 b. V that sie tanske are siistributed among the oefety function groupinge se spoolfled m y ki E 1; one teek shall re unre e low. power or shutdown cond6 tion, one or two shall respairs the oppilcant to en ettemete path ,a: : M, and one should require k3 d g. entry to the IRCA.

c. Vertly that the r'equired adminletrouve topice are oevered, with emphaele on p g3 p p."-.

                                         ;; '-- ' eotivitise.

d. Determine if there are enough different outlines to test the d number and mie l of oppueante end ensure that ne enore then 30% of the llame are on 1A g)! I i

sucessolve days.

• e. Assess whether plant opoelfte prierttaes (ineluding pftA and IpE insights) are soworod p j e in the appropriate seem escuen.

8 b. Anoses whether the 10 CPft $5.41l48 and 86.46 sempling le appropriate. 9 8.oJ g gEnsure that K'4 importance retings (eacept for plant opeoffic priorftles) are j p'5 {g et least [

                  "        d. Check for dupli etion and overiep amont exam sectione.                                            M sp! d

• e. Chook the entins seem for balance of severage. M 4AJ M L f. Assees whether the seem flte the appropriate job level (IW or SftO). Y pd 'd Pvt.t.e Iteme ielenet.ro sets i

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