ML20237F199

Exam Repts 50-282/OL-87-03 & 50-306/OL-87-01 on 871102-12. Exam Results:Nine Senior Reactor Operator Candidates Passed Exam

ML20237F199
Person / Time
Site:	Prairie Island
Issue date:	12/14/1987
From:	Burdick T, Hare S, Lennartz J, Shepard D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20237F186	List: ML20237F184 ML20237F199
References
50-282-OL-87-03, 50-282-OL-87-3, 50-306-OL-87-01, 50-306-OL-87-1, NUDOCS 8712290435
Download: ML20237F199 (83)

Text

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'U.S. NUCLEAR REGULATORY COMMISSION REGION'III-Reports No. 50-282/0L-87-03; 50-306/0L-87 Docket No.~ 50-282 L'icense No.:DPR-42 Licensee:

Northern States' Power Company 414 Nicollet-Mall Minneapolis, MN. 55401 Facility Name:

Prairie Island.

Examination Administered At:

Prairie Island Examinations Conducted:

Senior Reactor. Operator Examiners: %kASMk l ~t-t 4 +7 5.

M.- Hare 4

Date W

' l 2 - t 4 - V ')

J. A. Lennartz*

Date Yt.Utc0$ck W L Z t 4 - V7 D. E. Shepard D

'Date Approved.By: M,uddd

1 L -/ 4 - O T.-M.

Burdick, Chier

Date Operator Licensing Section Examination Summary Examination administered on November 2-12,1987 (Report No. 50-282/0L-87-03) -

Areas Inspected:

Written and operating exams were administered to eleven

. senior reactor operator candidates.

Results:

Nine senior reactor operators passed the. examinations.

8712290435,871386 PDR ADOCK 05000282 V

PDR O :

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.t DETAILS a

L.

Persons' Contacted i

1 Licensee i

+*L Eliason, General Manager, Nuclear Plants

• E. Watzl, Plant Manager a

+*L.-Waldinger,- Manager, Production Training

+*T. Amundson, Superintendent of Training

• R. Lindsay, Assistant Plant Manager

+*M. Sellman,-Superintendent of Operations

• K. 'Beadell, Su)erintendent, Quality Engineering

• P. Valtakis, Slift Manaaer-

+*D. Reynolds, Operation Training Supervisor NRC

'*S. M. Hare, Chief Examiner-

• J. A. Lennartz, Examiner

+D.

L. Shepard, Examiner J. E. Hard, Senior Resident Inspector.-

M. M. Moser, Resident Inspector

• W. S. Roesener, EG&G Contract Examiner

+N. C.,Venson, EG&G Contract Examiner

+F. S. Jagger, EG&G Contract Examiner

• Denotes persons attending the preliminary exit meeting of November 6,.

1987.

+ Denotes persons attending'the exit meeting of November 12,'1987.

2.

Examination.0 observations During the administration of the operating exams, The examiners noted a disparitybetweentheway.thecandidate'sescortedexaminers(visitors) and the requirement for escorting visitors as contained in-procedure' 5AWI 5.1.1, step'6.2.8.

The procedure requires the escort use their_ badge to open-the Vital Area door, and then let the visitor place their badge.

into the card reader and enter the Vital Area while the' escort'is holding the door open.

Contrary to this requirement, the escorts were allowing l-the visitors to badge in first, then place their badge into the card reader-and enter the Vital Area.

Additional information regarding their failure to follow this AWI is. contained in Inspection Report Nos. 282/87016 and-l 306/87015.

L 3.

Written Examination-1 After the SR0 written examination was administered on' November 3, 1987, a copy of the examination questions with answer key (see Enclosure 2) was supplied to the facility's training. staff for their review and comment.

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i f-Subsequent to their review, the licensee transmitted their comments on the written exam in a letter from Mr. E. L. Watzl to Mr.' G. C. Wright, dated November 9, 1987 (see Attachment'l to this report).

Each comment on specific questions has been reviewed in depth and was evaluated on its merits.

A resolution to each of these comments is contained as Attachment 2 to this report.

The following paragraphs discuss the format of your letter commenting on the' exam questions and your generic comment regarding True/ False questions.

a.

The format of the comments on the written exam did not adhere to the requested format.

This format was s)ecified in Attachment 4 of our July 31, 1987 letter from Mr. W. C. iehl to Mr. C. E. Larson, which formalized the examination schedule.

This Attachment will also be contained as Attachment 3 to this report.

b.

Your objection to the number of True/ False questions and their applicability toward a comprehensive exam is noted.

However, your remark about 27 True/ False items is somewhat misleading because True/ False questions only comprised 13.5 percent of the exam.

This percentage is well below the 25 percent allowable by the examiners standard NUREG 1021 ES-202.

4.

Exit Meeting The examiners met with licensee representatives (denoted in Paragraph 1) on November 6, 1987, and on November 12, 1987, at the conclusion of the examinations.

The examiners observations regarding candidate performance on the operating exams (denoted in Paragraph 2) were summarized.

The-licensee representatives acknowledged this information.

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QUESTION 5.16 Prairie Island Comment:

Assuming a rod worth of 200 pcm, the prompt drop will be approximately 0.25 10 8 amps.

This drop is not readily observable to the operator.

Consequently, full credit should be given for:

Reactor power will decrease and level out'at a lower power level in the source range.

A discussion of prompt drop should not be required for full credit.

NRC Response:

Concur.

Full credit will be given proposed answer.

Prompt drop deleted from the answer key and points redistributed.

QUESTION 6.01(a)

Prairie Island Comment:

The answer key should be expanded to include additional inputs into the auto start annunciator.

Depending on selector switch )ositions, there may be an additional two (2) answers which are included witlin the answer number 1.

If the air compressor start selector switch is in "1st standby," then relay"2nd numbers 42X112-6 with 43/1st standby will actuate annunciator and if. in standby," then 42X/112-6 with 43/2nd standby will actuate annunciator due to a start from low pressure.

Also, the alarm will actuate if the thermal overloads cause the motor to trip.

The additional acceptable answers are:

5.

Running due to low air pressure and in "1st standby."

6.

Running due to low air pressure and in "2nd standby."

7.

Compressor trips on motor thermal overload.

References:

Logic NF-40313-1, Attached Schematic NE-40008-20, NE-40011-34, Attached C-47 47022-0101, Attached NRC Response:

Partially Concur.

" Compressor trips on motor thermal overload," has been added to the answer key as an acceptable response.

However, " Running due to low air pressure," will only be accepted as one correct response.

QUESTION 6.03 Prairie Island Comment:

1 The answer given in the answer key is taken from a Prairie Island quiz pool.

However, upon further review, we believe the question as written will not solicit the response given in the answer key.

A better response is:

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1-ATTACHMENT 2 SENIOR REACTOR OPERATOR EXAMINATION COMMENTS AND RESOLUTIONS QUESTION 5.01(b)

Prairie Island Comment:

i At Prairie Island, some but not all startups result in initial criticality occurring in the intermediate range.

Consequently, there are two equally j

valid answers for this question.

For a startup where criticality occurs in the intermediate range, an overcompensated IR would over-predict criticality (non-conservative).

This is caused by the first data reading being too high and subsequent values not~

increasing as fast as actual neutron flux.

For a startup where criticality occurs in the source range, the overcompensation would have no effect and the answer key is correct.

Full credit should be given for either no change or.over predict, since both are correct and the cuestion specifically stated Iliat candidates should limit their answers to :

lnder predict criticality, over predict criticality or no change.

Reference:

ICRR Data Sheet, Unit I, Cycle 9, 12/17/82, Attached.

NRC Response:

l Concur.

The answer key has been modified to accept over predict or no change for Part b.

QUESTION 5.03(a)

Prairie Island Comment:

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The question asks for a brief explanation, but the answer key requires additional detail not solicited in the question.

We believe full credit should be given for:

Decreases, because Pu239 Concentration increases, Which has a smaller beta bar than U-235.

NRC Response:

Concur.

Answer key has been changed to eliminate Beta Bar Effective and redistribute the points.

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j output signal is dependent on plant condi'tions and may be between e'.'- 100%, ion depending on previous conditions.

Full credit'snould be givat fcr'a discuss oftheconditionsthatwouldcausetheoutputnietettdbe1 in a :: articular position.

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We have submitted a change to our training material,to ccrrect the answer key jc in our quiz pool.

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NRC Response:

Questiondeleted.

'i QUESTION 6.05 Prairie Island Comment:

The answer to this question should be TRUE.

Altbrugh the reference that is k

stated (83-5) indicates that there is not a bypass for.the seaf injection

) 7 filters, Flow Diagram X-HIAW-1-39 does indicate'that tb ve is a manual bypass a

aroundthesealinjectionfilters. Plan Page 10) simply states that there are: teu (2)renc The othei refe filters it parallel and does not address the filter bypass.

References:

X-HIAW-1-39, Attached P8172L-001, Page 10 (CVC$ Les:.on PaA Attached NRC Response:

Do not concur.

Two filttrs in parallel allows the seal injection filter to be changed without interrupting the seal injection flow, not the manual bypass valve.

Therefore, the correct answer is FALSE.

QUESTION 6.06 Prairie Island Comment:

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The question does not require setpoints to be given; however, setpoints appear

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to be required for full' credit.

The correct setpoint is -0.4 psi as opposed to a'

3 the setpoint listed of 4 psid.

Setpoint should not be required for full cred t since the question did not solicit the need to include setpoints in the answ3r.

Reference:

NF-40764-2, Attached a

NRC Response:

Concur.

The setpoint will not be raquired for full credit.

However, the utility is urged to revise their lesson plan from which this question was taken to reflect the correct setpoint.

Additionally, the question has been modified to solicit the need to include setpoints for futura v ams.

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kUti$ TION' 6. 09

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?rairie Island Comment:

a yD e answer should be TRUE.

'We w ference stated (DG Lesson Plan, Page 19) does elot contain the answer list.d in the arswer. key.

Page 18 of the same reference does state that 01 'is normally supplied by 11 DC Panel or 21 DC Panel as the

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backup via a mantal transfer.

Page 19 of the reference material is regarding D2, not D1, as is asked in the question.

Meference:

P8186L-004, Page 18 (Diesel Generator Lesson Plan), Attached o E C Response:

Concur.

The.3nberkeyhasbeenmodifiedtoreflectthis.

ROMION6210 Praicie Island Comment; The question does not solicit the extensive explanation that is required for full credit.

The question asks the candidate to "briefly explain the selection arocess>>

. (Include in your anwer ALL the POSSIBLE sources of power to ESF buses 15 and 16)." lhe answer key goes into great detail explaining what happens when each of the power supplies are searcied.

A brief explanc, tion of the Voltr.ge Restoring Scheme and a simple list of the power supplies should be all thW 1s requited for full credit.

Also, there seems to be some confusion

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among the candicates as to the assumptions regarding what power sunly was wsumed to have been initially supplyiryg the ESF bus 16 and what 0

SITE power supply was lost.

A general clarification was not made, and some candidates were told to assume IR was lost only when they questioned the proctor, and others assumed that the Cooling Tower buses along with IR were lost causing a TOTAL ioss of offsite power.

Depending on the candidate's assumption, the preferential os der of searching the buses would channe for ESF bus 16.

CT-11 and IR may be'.intercNnged depending on the assumed initial conditions,date and since the question does not imply which one(s) are/were lost, the candi may provide an answer with CT-11 and IR interchanged.

Since a clarification was not made to the entire group, either order of tne above stated search should be accepttd for full credit.

Also, the point distribution should be reconsidered since the question does not imply the detail of thr explanation as provided in the key.

NRC Response:

Partially Cencur.

The question specifically states that Unit 1 experiences a lo,ss of all offsite power which dould imply that IR and CT-11 was lost.

Also, the quesU Un stated that with the exception of D1 supplying bus 26, all electrical lineups were normal, which means thEt IR would be initially supplying Bus 16 since it is the normal powcr supply.

Also, the voltage restoring scheme does not "sench for" IR as. t power supply to the bus, so it i

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cannot be interchanged with CT-11, because CT-11 is always looked at first.

The answer key has been modified to reflect the correct order of power supplies to Bus 16 and the point distribution has been changed to recuire no explanation of additional powar* supplies, only a simple list as proposec.

QUESTION 6.11(a)

Prairie Island Comment:

- Prairie Island does not have a Channel Current Comparator.

There is a Channel Comparator and a Detector Cwrent Cotparator.

Some of the candidates may have been confused by the terminology, and credit should be givera for a correct response for either device.

The correct response for a Detector Current Comparator is:

The Detector Current Comparator is used to detect radial flux tilts by comparin average of the upper (g the input from each upper (lower) detector with thelower) detectors i

than 1.01 (2%).

l Also, the question did rot solicit more than one purpose.

Full credit for l

Channel Comparator shocid be given for either:

To detect radial flux tilts y to detect faulty PR channek.

References:

6-9, Page 19, Attached LP-8184L-002, Pages 42, 43, 48, and 49, Attached NRC Response:

Partial ty concur. When questioned by the Candidates that were confused by I

the terminology, the Proctor clarified that the question was referring to the l

Channel Comperator, 30 no credit will be given for a description of the Detector 1

Currer,t Comprator.

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Full credit will be given for either purpose as proposed.

Also, the question will be modified for future eums.

QUESTION 6.12 Prairie Island Comment:

ThequestionimpliesthatinputsfromeitherT@V9eyar@V9 ort or T Auctioneer are acceptable responses.

The answer listed in th Auctioneer l

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only.

The responses for T are:

avg L

OTAT 2.

0YhT 3.

Low T for feedwater isolation 4.

Low-L8wT for stem aump interlock and steam line isolation 9

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High T,ygagyarm These answers should be added to the answer key.

Reference:

LP-P8184L-003, Page 6, /!ttached 5

NRC Response:

Proposed answers were added to the answer key as acceptable answers.

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-QUESTION 6.16(b)

Prairie Island Comment:

The question does not solicit setpoints as part of the response.

The setpoint should not be required as part of a full credit response.

NRC Response:

Do not concur.

In order to explain and to show a complete' understanding of how the rod bottom lights would function to indicate this stuck rod, the setpoint for when rod bottom lights would be on, must be included in answer to receive l

full credit.

QUESTION 7.03 Prairie Island Comment:

The answers listed in the key are correct; however, other indications may be symptomatic of a failed high steam flow channel such as steam generator level deviation alarm or decreasing T if the failed channel is controlling and feedwater control is in auto.

8flsiderationshouldbegiventoother plausible answers.

NRC Response:

The questions specifically asks for initial sym3 toms (those that are immediately present) of the failed steam flow clannel, and the proposed answers would be symptoms that would be seen a short time after the channel failed, and not initially.

Therefore, the proposed answers will not be accepted.

QUESTION 7.07(b)

Prairie Island Comment:

Although an accumulator check valve leakage of 3 gpm is acceptable under TS 3.1.c.5, this amount of leakage presents problems in the accumulator. 'These valves are tested during startup, and the maximum acceptable leakage is 2.8 gpm as indicated in SP-1269.

Also, a 3 gpm in-leakage would require draining the accumulator approximately every two hours.

Frequent draining of the accumulators would dilute the boron concentration and reguire adjustments to 4

be made.

Both of these corrective actions would be required by TS 3.3.A.1.b.

l The correct response is "yes."

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References:

SP-1269 and TS 3.3.A.1.b., Attached 6

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NRC Response:

Do not concur.

The question specifically stated to treat each case independently, and that all other operability requirements are uet. Therefore, since the leakage by itself does not violate Technical Specifications, the correct response is no.

QUESTION 7.19 Prairie Island Comment:

The answer given in the key is correct.

However, the backc;round information contained in the Westinghouse Generic ERG's under generic issues also discusses the need to trip RCP's to minimize inventory loss.

This information is taught in our training, and credit should be given for this response also.

References:

Westinghouse Background Information for ERG's, Generic Issues, Page 8, Attached.

LP-P8197L-012, Page 19, Attached NRC Response:

Concur.

The answer key has been modified to accept either response for full credit.

QUESTION 8.04(a)

Prairie Island Comment:

The blocked or locked components are also identified in a table located in SWI-0-3 and in a monthly surveillance that verifies the proper attachment of-blocks and locks.

Copies of the SWI-0-3 table and SP-1210 are attached.

Full credit should be given for either of these responses.

Also, dates.the word " identified" in the cuestion ma,y have misled some of the l

I candi This seems to imply:

Fow can an individual tell that a component has a lock or block attached to it? Locks and blocks have a red identifying tag attached to them.

Some candidates may respond:

By observing the red tag attached to the block or lock.

Consideration for partial credit should be given to such a response.

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Reference:

SWI-0-3, Table 1 and SP-1210, Attached NRC Response:

Purtially concur.

" Component Blocking and locking Log (Table 1) in SWI-0-3" will also be accepted as a correct response.

However, SP-1210 will.not be accepted as this only references SWI-0-3.

The answer key has been modified to 1

reflect this.

Also, partial credit will be awarded for the proposed response of " observing the red tag attached to the block or lock."

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QUESTION 8.07 Prairie Island Comment:

A requirement to review the list of temporary memos does not exist if a red stamped copy or a controlled manual is used, because a red stamped or controlled manual would already contain the temporary memos.

The list of temporary memos needs to be consulted only if the copy is compared to the section index.

The correct answer is:

1.

Use a red stamped copy.

2.

Compare the copy with a controlled manual.

3.

Compare the copy with the section index and review the list of temporary memos.

Reference:

5ACD 4.1., Page 20, Attached NRC Response:

Concur.

The answer key has been modified to reflect this.

QUESTION 8.11(a)

Prairie Island Comment:

Add as acceptable answers for determining that 4160 v ITE Bkr is racked in properly:

DC Control Power Knife Switch Closed (This will be indicated by green / red lights on)

Locking tang inserted into racking screw slot (Can be determined visually or by physical movement of racking screw)

These are simply expansions of the answer key list with derivations of acceptable answers.

Reference:

C20.15 NRC Response:

Concur.

The proposed answers will be added to the answer key as alternate correct responses for keyed answers number (a) four and five.

QUESTION 8.15 Prairie Island Comment:

The detail required in the answer is too detailed.

An acceptable answer should be:

To limit tritium releases.

Water is transferred from the CVCS monitor tank to the Spent Fuel Pit for reasons other than tritium releases.

An INPO review in 1985 recommended that 8

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Prairie Island reduce the silica content of the Spent fuel-Pit.

To accomplish this, a reverse osmosis device was temporarily installed to process the Spent Fuel Pit water to remove the silica.

To make up borated water lost in the process, CVCS holdup tank water is transferred to the Spent Fuel Pit via the CVCS monitor tanks.

Candidates may respond to this question with a discussion of silica or other chemical contaminants.

Consideration should be given to these responses also.

Reference:

INP0 1985 E and A Report, Page 29, Attached PI Operations Note, Attached NRC Response:

Partially concur.

The answer key has been modified to accept "To limit tritium releases" as an acceptable answer.

However, alternate responses will-not be accepted since the question specifically states that the reason for transferring water from the CVCS Monitor Tanks to the Spent Fuel Pit is to curtail liquid releases, and not to make up for water lost due to the silica removal process.

QUESTION 8.18 Prairie Island Comment:

The answer given in the answer key is the preferred solution for this particular problem.

However,ajudgementmustbemadewhetherornot containment failure is imminent.

The information presented in the ?roblem is not absolutely clear on the trend of containment pressure, and t1erefore ajudgementmustbemade.

Itisnotunreasonabletojudgethatcontainmentfailureisnotimminentin this case, and therefore consideration should be given to-the following response:

1.

Evacuate all sectors out to five miles.

2.

Evacuate FGH sectors out to ten miles.

3.

Shelter remainder of plume EPZ.

4.

Activate PANS.

Reference:

F3-8.1, Page 8 of 12, Attached NRC Response:

Partially concur.

The question, as stated, indicates that containment failure is imminent.

However, since it does come down to the judgement of the Emergency Director, if it is stated in the answer that the candidates' assumption made is that containment failure is not imminent, the proposed answer will be accepted for full credit.

No changes to the answer key are necessary.

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ATTACHMENT 3 l

REQUIREMENTS FOR FACILITY REVIEW 0F WRITTEN EXAMINATION' i

1.

There shall be no review of the written examination by the facility

. staff before or during the administration of the examination.

Following the administration of the written examination, the facility staff shall be provided a marked-up copy of the examination and the answer. key.

2.

The facility will have five (5) working days from the day of the written examination is given to provide formal comment submittal.

The submittal will be made to the responsible Regional Office by the highest level of corporate ~ management for plant operations, e.g., Vice President for Nuclear Operations.

A copy of the submittal will be forwarded to the chief examiner, as appropriate.

Comments not submitted within five (5) working days will be considered for inclusion in the grading process on a case by case basis by the Regional Office Section Chief.

Should the comment submittal deadline not be met, a long delay for finalization of the examination results may occur.

3.

The following format should be adhered to for submittal of specific comments:

a.

Listing of NRC question, answer, and reference.

1 b.

Facility comment c.

Supporting documentation.

NOTES:

(1) No change to the examination will be made without submittal of complete, current, and approved reference material.

(2) Comments made without a concise facility recommendation will not be addressed.

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.s ATTACHMENT 1 Northem States Power Company -

P.I.N.G.P.

1717 Wakonade Drive West Welch, MN 55089 November 9, 1987 Mr. Goeffrey C. Wright, Chief Operations Branch Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Re: COMMENTS ON NRC ADMINISTERED SRO WRITTEN EXAMINATION

Dear Mr. Wright:

Prairie Island has reviewed the Senior Reactor Operator written examination administered on November 3, 1987. General and specific comments on the examination and answer key are listed below.

Should you have any questions or comments, please contact us at (612) 388-1165.

We are pleased that the emphasis of many questions was on the basis of proce-dures and administrative requirements, rather than memorization of procedural steps not required by administrative procedures.

In addition, several system questions tested problem solving or analysis skills as opposed to testing rote memorization of facts.

We believe further improvements could be made in the written examination process.

27 questions in the examination required a true or false response.

We believe this form of questioning has limited value in a comprehensive examination, and discourage its use on our comprehensive examinations.

Our second concern is with the scope of section 8 of the exam. Although we do not take exception to most of the questions, we noted the absence of basis questions for those technical specifications that are rated the highest in NUREG-1122, Knowledge and Abilities Catalog for Nuclear Power Plant Operators:

Pressurized Water Reactors (KA catalog). We believe some questions should be selected from the highest rated knowledges in all examinations.

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During the' exit meeting with the examiners on November 6..1987, we discussed the philosophy of allowing equivalent.vording as acceptable responses to those listed in the answer key for all basis-type questions. We agreed that equiva-lent wording will receive full credit if the response is reasonable in'the judgment of the examiner. NSP is not writing specific comments on each of the basis-type questions that do not have equivalent wording in the answer key.

' Specific comments on. questions and the answer key are listed below:

Question 5.01 - (b)

At Prairie Island, some.but not all startups result in initial. criticality occurring in the intermediate range. Consequently. there are two equally valid answers for.this question.

For a startup where criticality. occurs in the intermediate range, an overcom-pensated IR would over-predict criticality.(non-conservative).- This is caused-by the first data reading being too high and subsequent values not increasing as fast as actual neutron flux, For a startup where criticality occurs in the source range, the overcompen-sation would have no effect and the answer key is correct.

Full credit should be given for either no change g over-predict since.both are correct and the question specifically stated that candidates should limit their answers to: Under predict criticality, over predict criticality or no change.

Reference:

ICRR Data Sheet, Unit 1, Cycle 9, 12/17/82, Attsched.

Question 5.03 (a)

The question asks for a brief explanation, but the answer key requires addi-tional detail not solicited in the question. We believe full credit should be given for: Decreases, because Pu e n entration increases, which has a 239 smaller beta bar than U-235.

Question 5.16 Asqming a rod worth of 200 pcm, the prompt drop will be approximately 0.25 x 10 amps. This drop is not readily observable to the operator.

Consequently, full credit should be given for: Reactor power will. decrease and level out at a lower power level in the source range. A discussion of.

prompt drop should not be required for full credit.

Page 2 of 7 L_--___-

O Question 6.01 (a)

The answer key should be expanded to include additional inputs into the auto start annunciator. Depending on selector switch positions, there may be an additional two (2) answers which are included within the answer number 1.

If the air compressor start selector switch is in "Ist standby", then relay numbers 42X/112-6 with 43/1st standby will actuate annunciator and if in "2nd standby", then 42X/112-6 with 43/2nd standby will actuate annunciator due to a start from low pressure. Also, the alarm will actuate if the thermal over-loads cause the motor to trip.

The additional acceptable answers are:

5.

Running due to low air pressure and in "Ist standby".

6.

Running due to low air pressure and in "2nd standby".

7.

Compressor trips on motor thermal overload.

References:

Logic NF-40313-1, Attached Schematic NE-40008-20, NE-40011-34, Attached C-47 47022-0101, Attached Question 6.03 The answer given in the answer key is taken from a Prairie Island quiz pool.

However, upon further review, we believe the question as written will not solicit the response given in the answer key. A better response is: The output signal is dependent on plant conditions and may be between 0 - 100%,

depending on previous conditions.

Full credit should be given for a dis-cussion of the conditions that would cause the output meter to be in a particular position.

We have submitted a change to our training material to correct the answer key in our quiz pool.

1 Question 6.05 The answer to this question should be TRUE. Although the reference that is stated (B3-5) indicates that there is not a bypass for the seal injection filters, Flow Diagram X-ElAW-1-39 does indicate that there is a manual bypass around the seal injection filters. The other reference listed (CVCS Lesson Plan Page 10) simply states that there are two (2) filters in parallel and does not address the filter bypass.

References:

X-HlAW-1-39, Attached P8172L-001, Page 10 (CVCS Lesson Plan) Attached Question 6.06 The question does not require setpoints to be given; however, setpoints appear to be required for full credit. The correct setpoint is -0.4 psi as opposed to the setpoint listed of 4 psid.

Setpoint should not be required for full credit since the question did not solicit the need to include setpoints in the answer.

Reference:

NF-40764-2, Attached Page 3 of 7 w ___________

Question 6.09 The answer should be TRUE. The reference stated (DG Lesson Plan, Page 19) does not contain the answer listed in the answer key. Page 18 of the same reference does state that D1'is normally supplied by 11 DC Panel or 21 DC Panel as the backup via a manual transfer. Page 19 of the reference material is regarding D2, not D1, as is asked in the question.

Reference:

P8186L-004, Page 18 (Diesel Generator Lesson Plan), Attached Question 6.10 The question does not solicit the extensive explanation that is required for full credit. The question asks the candidate to "briefly explain the selection process..........(Include in your answer ALL the POSSIBLE sources of power to ESF buses 15 AND 16)".

The answer key goes into great detail explain-ing what happens when each of the power supplies are searched. A brief explanation of the Voltage Restoring Scheme and a simple list of the power supplies should be all that is required for full credit.

Also, there seems to be some confusion among the candidates as to the assumptions regarding what power supply was assumed to have been initially supplying ESF bus 16 and what 0FFSITE power supply was lost. A general. clarification was not made, and some candidates were told to assume IR was lost only when they questioned the proctor, and others assumed that the Cooling Tower buses along with IR were lost causing a TOTAL loss of offsite power. Depending on the candidate's assumption, the preferential order of searching the buses would change for ESF bus 16.

CT-11 and IR may be interchanged depending on the assumed initial conditions, and since the question does not imply which one(s) are/were lost, the candidate may provide en answer with CT-11 and IR interchanged. Since a clarification was not made to the entire group, either order of the above stated search should be accepted for full credit. Also, the point distri-bution should be reconsidered since the question does not imply the detail of the explanation as provided in the key.

Question 6.11 (a)

Prairie Island does not have a Channel Current Comparator.

There is a Channel Compriator and a Detector Current Comparator. Some of the candidates may have been confused by the terminology, and credit should be given for a correct response for either device. The correct response for a Detector Current Comparator is: The Detector Current Comparator is used to detect radial flux l

tilts by comparing the input.from each upper (lower) detector with the average of the upper (lower) detectors, and alarming if the ratio is greater than 1.02 (2%).

Also, the question did not solicit more than one purpose.

Full credit for Channel Comparator should be given for either: To detect radial flux tilts cy; to detect faulty PR channels.

References:

B-9, Page 19, Attached LP-8184L-002, Pages 42, 43, 48 and 49, Attached Page 4 of 7

, s*

.c Question 6.12 (a)

The question implies that inputs from either T or Auctioneer are

-The answer listed in th!"Eey aTr$"for T Auctioneer acceptable responses.

avg only. The responses for T,y are:

1.

OTAT 2.

OPAT 3.

Low T for feedwater isolation 4.

Low-LS?rT for steam dump interlock and steam line isolation E

8 5.

High T afarm avg These answers should be added to the answer key.

Reference:

LP-P8184L-003, Page 6, Attached Question 6.16 (b)

The question does not solicit setpoints as part of the response. The setpoint should not be required as part of a full credit response.

Question 7.03 The answers listed in the key are correct; however, other indications may be symptomatic of a failed high steam flow channel such as steam generator level deviation alarm or decreasing T if the failed channel is controlling and feedwater control is in auto. doEsideration should be given to other plausible answers.

Question 7.07 (b)

Although an accumulator check valve leakage of 3 gpm is acceptable under TS 3.1.c.5, this amount of leakage presents problems in the accumulator.

These valves are tested during startup, and the maximum acceptable leakage is 2.8 gpm as indicated in SP-1269. Also, a 3 gpm in-leakage would require draining the accumulator approximately every two hours. Frequent draining of the accumulators would dilute the boron concentration and require adjustments to be made. Both of these corrective actions would be required by TS 3.3.A.1.b.

The correct response is "yes".

References:

SP-1269 and TS 3.3.A.1.b, Attached l

Question 7.19 The answer given in the key is correct. However, the background information contained in the Westinghouse Generic ERG's under generic issuer also discusses the need to trip RCP's to minimize inventory loss. This infomation is taught in our training, and credit should be given_for this response also.

References:

Westinghouse Background Information for ERG *s, Generic Issues, Page 8, Attached LP-P8197L-012, Page 19, Attached i

Page 5 of 7

(

.m.

...-.................,a

• s s i

I Question 8.04 (a)

The blocked or locked components are also identified in a table located in SWI-0-3 and in a monthly surveillance that verifies the proper attachment of blocks and locks.

Copies of the SWI-0-3 table and SP-1210 are attached.

Full credit should be given for either of these responses.

Also, the word " identified" in the question may have misled some of the candidates. This seems to imply:

has a lock or block attached to it?How can an individual tell that a component Locks and blocks have a red identifying tag attached to them. Some candidates may respond:

By observing the red tag attached to the block or lock.

Consideration for partial credit should be given to such a response.

Reference:

SWI-0-3, Table 1 and SP-1210, Attached Question 8.07 A requirement to review the list of temporary memos does not exist if a red stamped copy or a controlled manual is used, because a red stamped or controlled manual vould already contain the temporary memos.

The list of temporary memos needs to be consulted only if the copy is compared to the section index.

The correct answer is:

1.

Use a red stamped copy.

2.

Compare the copy with a controlled manual.

3.

Compare the copy with the section index and review the list of temporary memos.

Reference:

SACD 4.1, Page 20, Attached Question 8.11 (a)

Add as acceptable answers for determining that 4160 v ITE Bkr is racked in properly:

DC Control Power Knife Switch Closed (This vill be indicated by green / red lights on)

Locking tang inserted into racking screw slot (Can be determined visually or by physical movement of racking screw)

J These are simply expansions of the answer key list with derivations of

]

acceptable answers.

.{

.{

Reference:

C20.15 Question 8.15 The detail required in the answer is too detailed.

should be: To limit tritium releases.

An acceptable answer 4

i Page 6 of 7 L

,yi

'e*~> a

=

Water is transferred-from the CVCS monitor tank to the Spent Fuel Pit for reasons other than tritium releases. An INP0 review in 1985 recommended that Prairie Island reduce the silica content of the Spent Fuel Pit. To accomplish-this, a reverse osmosis device was temporarily installed to process the Spent Fuel Pit water to remove the silica.

To make up borated water lost.in the process, CVCS holdup tank water is transferred to the Spent Fuel Pit via.the CVCS monitor tanks. Candidates may respond to this question with a discussion of silica or other chemical contaminants. Consideration should be given to these responses alro.

References:

INPO 1985 E and A Report, Page 29, Attached PI Operations Note, Attached 1

Question 8.18 The answer given in the answer key is the preferred solution for this particular problem. However, a judgment must be made whether or not contain-ment failure is imminent. The information presented in the problem is not I

absolutely clear on the trend of containment pressure, and therefore a judgment must be made.

It is not unreasonable to judge that containment failure is not imminent in this case, and therefore consideration should be given to the following response:

1.

Evacuate all sectors out to five miles.

2.

Evacuate FGH sectors out to ten miles.

3.

Shelter remainder of plume EPZ.

4.

Activate PANS.

Reference:

F3-8.1, Page 8 of 12, Attached We request that representatives of our training staff meet with the examiners at your earliest convenience to discuss specific exam content issues.

We are willing to travel to your offices to facilitate this meeting.

If such a meeting is acceptable, please contact T. E. Amundson at (612) 388-1165 to determine a meeting date.

Sincerely, 4

i

. L. Watz1 i

Plant Manag Prairie Island Nuclear Generating Plant Attachments cc: Jim Hard (w/o attachments)

ELW/1jw Page 7 of 7

h U.

S.

NUCLEAR' REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION 1

FACILITY:

_ER81BIE_I@LOND_1h2______

REACTOR TYPE:

_PWR-WEC2________________

i DATE. ADMINISTERED: _@ZZ11/9}_________.______

EXAMINER:

_LENNARTZz'J.

CANDIDATE:

INSIBUCIlggS_Ig_CGNDID81El Use separate paper for the answers.

Write answers'on one side.only.

Staple question sheet on top of'the answer sheets.

Points for each question are indicated in parentheses after'the question.

The passing grade requires at least 70% in each category and a final' grade of at least 80%.

Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY

% OF CANDIDATE'S CATEGORY

__Y8LUE_ _IQIOL

___ECQBE___

_26LUE__ ______________C81EGQBy_____________

_2Es99__ _2Ez12

________ S.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS J

_23125__

23:3}

________ 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_25z99__ _29112

________ 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_2 Ergo __ _29z12

________ S.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_2212D__

Totals Final Grade All work done on this examination is my own.

I have neither given nor received aid.

1 Candidate's Signature i

1 i

J i

______-_______D

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Curing the administration of this examination the following rules apply:

1.

Cheating on the examination means an automatic deni al of your application and could result in more severe penalties.

2.

Restroom trips are to be limited and only one candidate at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoi d even the appearance or possibility of cheating.

3.

Use black ink or dark pencil gnly to facilitate legible reproductions.

4.

Print your name in the blank provided on the cover sheet of the ex ami nati on.

5.

Fill in the date on the cover sheet of the examination (if necessary).

6.

Use only the paper provided for answers.

7.

Print your name an the upper right-hand corner of the first page of eac_b section of the answer sheet.

8.

Consecuti vely number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.

9.

Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least threg lines between each answer.

11. Separate answer sheets from pad and pl ace f inished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility liter.ature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE DUESTION AND DO NOT L. EAVE ANY ANSWER BLANK.

16.

If parts of the examination are not clear as to intent, ask questions of the egaminer only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.

This must be done after the examination has been completed.

l 1

____-_________D

O 1b. When you complete your examination, you--shal1:

a.

Assemb1e your examination as fal1ows:

(1)

Exam questions on t o p_.

(2)

Exam aids - figures, tables, etc.

(3)

Answer pages including figures which are part of the answer.

b.

Turn in your copy of the examination and all pages used to answer the examination questions.

c.

Titrn in all scrap paper and the balance of the paper that you did not use for answering the questions.

d.

Leave the examination area, as defined by the examiner.

If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

i 1

1 l

s

EQUATION SHEET f = ca v = s/t Cycle efficiency = (Net work out)/(Energyin) 2 w = og s = V,t + 1/2 'at 2

E = ac g,g,,4 2

KE = 1/2 av

,,gyf, y )fg g,in c

PE = agn.

Vf = V, + at w = e/t A = an2/t.1/2 = 0.693/t1/2 if2*ff=[(tin)(t )3 2

t s

v., g.

,o A*

[(t1/2)*II))

4 b

AE = 931 am

,y A

~D av '

I = I,e

~

Q=$Cpat 6 = UA A T I = 1 e~"*

o Pwr = W ah I*I 10,377yg f

o TYL = 1.3/u sur(t)

P = P,10 HVL = -0.693/u P = P e'l SUR = 26.06/T SCR = S/(1 - K,ff)

CR, = S/(1 - K,ffx)

CR (1 - K,ff)) = CR II ~ eff2)

SUR = 26s/t* + (s - s)T j

2 T = (t*/s) + [(s - s yIo]

M = 1/(1 - K,ff) = CR /CR,

j T = a/(o - s)

M = (1 - K,ff,)/(1 - K,ffj)

T = (s - o)/(Is)

SDM = ( -K,ff)/K,ff o = (K,ff-1)/K,ff = AKeff/K,ff 1* = 10 seconds I = 0.1 seconds ~I o = [(t*/(T K,ff)] + [s,ff (1 + IT)]-

/

Idjj=1d 2,2 2 P = (t4V)/(3 x 1010)

Id gd jj 22 2

t = eN R/hr=(0.5CE)/d(meters)

R/hr = 6 CE/d2 (feet)

Miscellaneous Conversions Water Parameters 1 gal. = 8.345 lem.

1 curie = 3.7 x 1010dps I ga;. = 3.78 liters 1kg=2.21lbm 1 fte = 7.48 gal.

1 hp = 2.54 x 1 8tu/hr Density = 62.4 lbqi/f t3 1 nw = 3.41 x 1 Stu/hr i

Density = 1 gm/ene lin = 2.54 cm i

Heat of vaporization = 970 Stu/lem

• F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm

• C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-1bf 1 ft. H O = 0.4335 lbf/in.

2

'5 __IMEggy_gE_ NUCLE 88_EgWEB_ELONI_9EEBBIlgN _ELUJp$3_@ND PAGE 2

3 t

IUEBUppyN8MJCS QUESTION 5.01 (1.00)

What effect will EACH of the following have on a 1/M plot?

Limit your answer to UNDER-PREDIC'l criticality (conservative), OVER-PREDICT criticality (non-conservative), or NO CHANGE.

(Consider each case

~

separately.)

a.

Time interval between rod pulls on a startup is changed from 1 minute to 10 seconds.

b.

Intermediate range nuclear instruments are overr opensated.

c.

Source strength immediately prior to the startup changed from 4 cps to 10 cps.

QUESTION 5.02 (1.00)

A relief valve (to the atmosphere) on a pipe opens at 885 psig.

The temperature of the exhausted steam is 310 degrees F.

The temperature of the fluid (water or steam) wi thi n the pipe is appr ox i matel y :

(Choose the ONE correct response.)

a.

212 degrees F b.

310 degrees F c.

400 degrees F d.

540 degrees F QUESTION 5.03 (1.50) a.

Does Beta Bar Effective INCREASE, DECREASE, or REMAIN THE SAME from BOL to EOL?

(Briefly expl ain)

(1.0) b.

For equi val ent positive reacti vi t y additions to a critical reactor, will the SUR be the SAME, LARGER, or SMALLER at BOL as compared to EOL?

(No explanation necessary)

(0,5)

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

5 __ISEQBY_QE_NUC6E88_EQWEB_E6@NI_QEEB8IlQN2_E6UlDS _8ND PAGE 3

t 1

ISEBMQQYN8MICS QUESTION 5.04 (2.25) a.

At what axial location in an operating PWR core is the critical heat flux at the MAXIMUM value?

(Limit your answer to TOP, MIDDLE, or BOTTOM)

(0.75) b.

How does the MINIMUM critical heat flux value change as the following parameters are DECREASED.

(Limit your answer to INCREASE, DECREASE, or NO CHANGE.

Consider each case' independently.)

(1.5) 1.

Tave 2.

RCS pressure 3.

RCS flow OUESTION 5.05 (2.50)

Answer EACH of the following TRUE or FALSE:

a.

Equilibrium SAMARIUM concentration is-a function of power.

(0.5) b.

Equilibrium XENON reactivity is dependent upon core burnup.

(0.5) c.

XENON concentration initially increases after a reactor trip.

(0.5) d.

Equilibrium SAMARIUM concentration after a trip is a function of the previous power level.

(0,5) e.

Equilibrium XENON concentration after a trip is a function of the previous power level.

(0.5)

DUESTION 5.06 (2.00) 15.0 E 6 lbm/hr of 428 degrees F feedwater is being delivered 'to the steam generators at 1010 psig.

CALCULATE the core thermal power in MW that is being produced by the reactor.

(Show your work)

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

h __ISEQBY_QE_ NUCLE 88_EQWEB_E68MI_QEEffGIIONz_E6QlDSz_@ND PAGE 4

IUEBdOQXNGblCS QUESTION 5.07 (1.00)

During a xenon-free reactor startup, critical data was inadvertently taken two decades BELOW the required Intermediate Range. (IR) level.

Assuming RCS temperatures and baron concentrations were the same, the critical rod position taken at the proper IR. level _ _ _ _, _ the critical rod position taken two decades below the proper IR level, a.

Is Less lhan b.

Is the Same As c.

Is Greater Than d.

Cannot be Compared la OUESTION 5.08 (2.00)

Explain why a dropped rod, (wi th all other control rods out), is worth approximately 200 pcm while a stuck rod, (with all other control rods in),

is worth about 1000 pcm even though the same rod could be considered in both cases.

QUESTION 5.09 (1.00)

If the RCP's are tripped following a LOCA, and the break has been isolated, which ONE of the fallowing situations provide the LEST subcooling?

)

PZR Pressure HOT LEG Temperature COLD LEG Temperature a.

600 500 480 b.

800 530 520 l

c.

1000 540 530 j

I d.

1200 575 565

(*****

CATEGORY 05 CONTINUED ON NEXT PAGE

• • • • • )

j

5.___IHEQBX_QE_NUCLEGB_EQWEB_E68N1_QEE8@llQN _E6QlDS _GND PAGE b

1 1

IUEBMQDYNGdigS QUESTION 5.10 (1.75)

Answer EACH of the f ollowing questions concerning Centrif ugal Pump Characteristics:

a.

What happens to system fl ow rate (INCREASE, DECREASE, or REMAIN THE SAME) when a second centrifugal pump is started in series with a running pump?

(0.2D) b.

What happens to pump head (INCREASE, DECREASE, or REMAIN THE SAME) if pump speed is doubled?

(State the magnitude of the change, if any.)

(0.5) c.

Deiine pump runout in terms of pump head and system flowrate.

(0.5) d.

List FOUR indications of pump cavitation.

(0.5)

QUESTION 5.11 (1.25)

To what val ue must steam generator pressure be adjusted to in order to maintain a 200 degree F subcooling margin in the RCS, when RCS pressure is reduced to 1600 psi g?

Show all work.

(Specify answer in units of psig.)

OUESTION 5.12 (1.50)

How will EACH of the following affect a calori metric power cal cul ati on?

Limit answer to CALCULATED LOWER THAN ACTUAL, CALCULATED HIGHER THAN ACTUAL, or CALCULATED SAME AS ACTUAL.

(Consider each case separately.)

a.

Measured feedwater temperature is 10 degrees lower than ac'tual feedwater temperature, b.

Measured steam generator pressure is 30 psig lower'than actual steam generator pressure, c.

Measured feedwater flow is 1ES lbm/hr. lower than actual feedwater flow.

1 DUESTION 5.13 (1.25)

What is Shutdown Margin, as defined in Technical Specifications?

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

j 1

)

5.__IBERBY_QE_UUCLE@B_EQWEB_EL@NI_QEEB@IlgNg_ELQlDS _@ND PAGE 6

z IUEBdODYN@dlCS i

i QUESTION 5.14 (1.00)

Which DNE of the f ollowing statements concerning the power defect is correct?

a.

The power defect i's the difference between the measured power coefficient and the pred ted power coefficient.

b.

The power defect increases the rod worth requirements necessar y to maintain the desired shutdown margin following a reactor trip.

c.

Because of the higher baron concentration, the power defect is negative at beginning of core life.

more d.

The power defect necessitates the use of a ramped Tavg program to maintain an adequate Reactor Coolant System subcooling margin.

QUESTION 5.15 (1.00)

The Delayed Neutron Importance Factor at Prairie Island is _ _ _ _ _ than (Choose the correct response) one because delayed neutrons _ __, _.

a.

Less; are less li kely to leak from the core.

b.

Less; do not cause f ast finsion of U-238.

c.

Greater; are less li kely to leak from the core.

d.

Greater; do not cause fast fission of U-238.

QUESTION 5.16 (1.50)

I Assuming a Xenon-free reactor startup, while critical with power leveled out at 10-8 amps for critical data, Tave = 546 degrees F and RCS pressure 2235 psig, rod D-4 (control bank D) drops to the bottom.

Describe reactor power, RCS temperature, and RCS pressure transients caused by the. dropped rod.

(End discuscion at stable plant conditions.)

(***** CATEGORY 05 CONTINUED ON NEXT PAGE

• • • • • )

E r._ _IBE Q B y _QE _N Q C L E BB _ EQWE 8_E L ONI_Q BE 8811Q N g _EL QlD S1_ Gt]D PAGE 7

ISERdQDYNGdlCS QUESTION 5.17 (1.50)

Fill in the blanks to complete EACH of the following statements concerning subcritical multiplication.

(Limit answers to INCREASE, DECREASE, or REMAIN THE SAME.)

a.

If a neutron source strength increases, the rod height for criticality will (0.5) b.

If a neutron source strength is decreased, neutron level will.

(Assume original neutron count rate to be 100 cps.)

(0.5) c.

The period of time for a constant neutron level to be reached following a positive reactivity insortion at Keff = 0.8 will

_ _ _ _ _ f rom an equal reactivity insertion at Keff = 0.9.

(0.5) 1

(***** END OF CATEGORY 05 *****)

l I

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,6,___P(GNI_@VSIEd@_ DESIGN 1_CQNIB061_@ND_INSIBUdENI8IlON PAGE' B

QUESTION 6.01 (2.00)

Answer EACH of the f ollowing questions concerning the Instrument Air System:

a.

Air Compressor 121 " COMPRESSOR AUTO START-STOP" annunciator is energized in the Control Room.

What are THREE conditions / signals that would cause this alarm to ener gi n e?

(Annunciator Test is not a valid response)

(1.5) b.

TRUE OR FALSE Unit 1 Instrument Air Containment Isolation Valves will aut omati c al l y i

I close if a containment isolation signal is received for that unit.

( 0. 5 )

DUESTION 6.02 (2.50)

How do EACH of the following valves fail if c on t r ol air is lost?

(Limit answer to CLOSED or OFEN) 1.

CV-31202, Letdown Heat Exchanger CCW Flow Control Valve 2.

CV-31203, Letdown Pressure Control Valve, (PCV-135) 3.

CV-312OO, Bl en d er to VCT Outlet, (FCV-110B) 4.

CV-31201, Blender to VCT Inlet, (FCV-110C) 5.

CV-31199, Boric Acid to Bl ender, (FCV-110A)

QUESTION 6.03 (1.00)

What will the PZR Pressure Controller output meter on the Main Control Board read if the controller is in AUTOMATIC and actual RCS pressure equals the automatic pressure setpoint?

(Give a brief explanation)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6t__P66NI_SYSIE,dS_ DESIGN _CONIBObt_6ND_lNSIBudENI6110N PAGE 9

1 QUESTION 6.04 (1.50)

Unit 1 is in Cold Shutdown with.12 S/G Narrow Range level at 16%. 12 S/G is being drained dry for maintenance at this time.

The RO has just completed an operability test of 12 Motor Driven AFW pump and placed its Selector Switch in the " SHUTDOWN AU'(O" position. L.at er that shift, t.h e 12 AUX FW PUMP LOCKED OUT" annunciator energized.

EXPLAIN the events which caused this annunciator to energine, AND WHAT should have been done to prevent it? (Assume all system parameters / components meet operability requirements for Cold Shutdown.)

QUESTION 6.05

(.50)

TRUE or FALSE.

The RCP seal injection filter is equipped with an external bypass line which allows the filter to be changed without interrupting the seal injection flow.

l OUESTION 6.06 (1.00) l Explain how the Vacuum Dreakers will AUTOMATICALLY function to pr event excessive containment vacuum if the air operated butterfly valves go shut due to a Containment Isolation Signal. (Include any applicable setpoints)

OUESTION 6.07 (1.00)

While at 98% power, maintenance had to repair a leak in the Instrument Air System on Unit 2.

This required hooking up a temporary supply to CV-31939, (CA I sol ati on Control Val ve to Containment Spray Pump Suction),

using a nitrogen bottle which resulted in overpressurizing the air regulator to CV-31939 and causing it to fail.

How would this af f ect CV-31939, and what would be the adverse affects to the system?

(+++** CATEGORY 06 CONTINUED ON NEXT PAGE

• • • • • )

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6 __E63NI_gYgIEUS_DESl@N _CgNIBgb2_SUD_INSIBUDENISIIgN PAGE-10 z

3 QUESTION 6.08 (2.00)

Answer the f ollowing questions concerning the Emergency Diesel Gcnerator (EDG):

When performing the monthly load test of the EDG in accordance with SP 1093, the Speed Droop Control Knob on the Engine Governor is manually set at 40.

a.

WHAT is the Speed Droop Control Knob NORMALLY set at?

( 0. 5 )

b.

Why is the Speed Droop Control Knob-set at 40 during the monthly load test of the EDG?

(1.5)

QUESTION 6.09

(.50)

TRUE or FALSE?

If DC Panel 11 is out of service f or maintenance, Emergency Diecel Generator D1 will not be able to start upon receipt of a start signal,

~

unless power to D1 is manually transf erred to DC Panel 21.

l OUESTION 6.10 (2.00)

Unit 2 ESF bus 26 is being supplied f rom the Emergency Diesel' Generator D3 (all other electrical line ups are normal), and Unit i experiences a loss of all offsite power.

Bri efl y expl ai n the sel ec ti on process that the Voltage Restoring Scheme will go through.in restoring power to'ESF buses 15 and 16.

(Include in your answer ALL the POSSIBLE sources of power to ESF buses 15 AND 16.

Assume both units were initially in mode 1.)

QUESTION 6.11 (2.00)

Answer EACH of the f ollowing questi ons concer ning. the Nuclear Instrumentation System:

a.

WHAT are the purposes of the Channel Comparator, AND HOW does it perform these functions?

(include any applicable setpoints in your answer.)

(1.0) b.

If Reactor Power is at 15%, and the RO inadvertently turns both SR (Source Range) Block Switches to RESET, will the SR instruments reenergize?

(Ex pl ai n why or why not)

(1.0)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE

• • • • • )

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'/

(

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1 QUESTION 6.12 (1.50)

Answer the f ollowing questions concerning the Reactor Prg:ess

. Instrumentation System:

[,

f 8,,

~,

• ,'I*

Tavg/Tavg Auctib.,\\t-er input to?

a.

What are FOUR CONTROL circuits that VA.) d (1.0)

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b.

TRUE or FALSE Ny,s All four Delta T circuits feed into a Delta T AUCTIONEER which selects the HIGHEST Delta 7~ input as an output.

(0. 5) -

QUESTION 6.13

(.50) 4 TRUE or FALSE failsHIGHwithreacker If PT 485 (Turbine 1st Stage Pressure 1ransmitt r) power at 100% and rod control in AUTOg then the rods will step out duej 't Tavg less than Tref.

(Assume Tavg and Tref were matched prior to fy failure.)

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OUESTION 6.15

(,50)

'f t'

TRUE or FALSE I

i All Radiation Monitoring System channels will. certirue to read UPSCALEj' regardless of how high a radiation field becomec.

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')i DUESTION 6.16 (2.50)

Answer EACH of the f ollowing questions concerning the Rod Posi tion

,3 Indication System:

a.

Is it possible to have a Control Bank C rod stuck on the bottom I

without the " ROD BOTTOM / ROD DROP" alarm env gized with the Control l

Bank C bank demand position at 28 steps?

h* x pl ai n )

(1.5) l l

b.

Could you use the rod bottom indicating lights to detect the stuck l

rod in part a?

(Explain)

(1.0) i f

1 e'

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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, h_.ZLBUL.diSIEMQ_DESlgN t_ CONI 606,3_9ND_IN@IBUMENIGIlgN PAGE 12 1

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DUESTION 6.17 (1.25) f 1

J r

i Answer EAOJ of the following questiond boncerning the Reactor Vessel Level Indi cating System (RVLIS):

a.

What are THREE inputs into Rblis,+ hat provide for density compensattili of the indicated Yesse' l evel readings?

(.75) f i

b.

'; RUE or FALSE

/'

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g : ',

v' Reactor Vessel UPFER RANGE reading is'envalid if the Reactor 7,, 'h

, 't Coolant Pumps ar.z operating.

(O.Si s

/

'/

,s DUEST[ON, 6.1 E' k,00)

(

Indicate !hether EACH of the f ollowing situations will ARM only, ARM and ACTUA E, or HAVE NO EFFECT on the Steam Dump Syy, tot..

(Assume condenser inter Mk as met.

Cc s der each situatico independently.)

)s ',i

's N9 s

p b

gif a.

Hot zero powerg Tavg/= 350 degrees F, steam dumps in STEAM PRESSURE

' mode wish 1005,qsig set into the Steam Pressursr Controller.

)\\

h b.%

80% power, 2%/th n! /A.>p decrease in turn ne load for 3 minutes, Tavg greatpr than Tref by d s%grees F, steam' dumps in Tavg mode.

\\

Turbine tg)pg Tavg = S38 degrtMI F.( steam dudsas in Tavg mode.

+/

+

c.

/

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+

r

.\\

t 7 5 rap decrcgoa,in turbinb load for 3 minutes, Tavg d.

50% power, t*..af[y;/./ mi ny' % i degrep [, si,t eam dumps in Tavg greater y

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/

O' JEST I ON 7.01 (1.00)

St.up 14, " Der essurize RCS," of ES-0.3,

" Natural Circulation Cooldown,"

directs you a wait 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after establish 2ng RCS pressure at 1200 psig I

and RCS t emperature at 350 degrees F,

without CRDM fans running.

WHN does this proccdure require you to wait 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> before continuing on?

DUESTION 7.02 (1.00)

A CAu f 10t' in E-3.

'S/G Tube Ruptur e, " staten that if no intact S/G ic available, the ruptured S/G with the lower level should be used for RCS c ool ri oon.

WHAT is the bases for this?

OLlESTION 7.03 (1.50)

In accordance with Instrument Failure Guide IC51.3, what would be FIVE initial symptoms, (expected plant response), of the 12 S/G STEAM FLOW, I F--4 74 (Blue), failed HIBH?

(St at e ai,y assumptions made)

OUESTION 7.04 I.1., 0 0 )

A note in C3, Sectic.#

4.0, "RCP Emergency Startup," states that the seal leakoff flow meters are not environmentally qualified, and that other indications may have to be used to determinetthat seal leakoff is established.

What are TAO of these indications?

l 1

OUESTION 7.05 (1.50)

A precaution in C 3., " Reactor Cracl ant Pump," states that RCP Mi seal bypass "shall" remain clos 9d unless certain conditions are met.

One of the conditions is that RCF. pressure muet be greater than 100 psi qi but less than 1000 psig.

What are the bases for these 2 pressure limits for opening the #1 ueal bypass valve?

(1.5)

(**nu CM EGOryv O / CONTINUED ON NEXT PAGE **++ *)

Zs__EBQGEQQBEE_ _UOBd86t_6EUQBd661_EMEB@EUGY_8UQ PAGE. 14 689106991086_GONIB96 DUESlION 7.06 (1.50)

A' CAUTION in C12-8.4, " Emergency Boration of the RCS," states that concentrated boric acid flow to the charging pump suction must be limited to less than 75% of the total charging pump flow.

a.

What system design limits this boric acid flow rate?

(.75) b.

What is the reason for limiting the boric acid flow?

(.75)

DUESTION 7.07 (2.50)

Indicate whether EACH of the f ollowing parameters /si tuations will requi r e corrective actio'n as dictated by Technical Specifications.

(Consider each case independently.

Assume all'other operability requirements are met.

Limit answer to YES or NO) a.

Containment pressure =

1. 75 psi g b.

Accumulator check valve leakage 3 gpm

=

c.

One Unit 2 PZR Safety Valve is declared inoperable while at 1D%

power d.

11 Boric Acid Storage Tank ' level of 2100 gallons e.

21 Condensate Storage Tank level of of 10,000 gallons DUESTION 7.08 (1.00)

The NOTE before Step 8, "Ini ti ate RCS Cool down," in ES-0.3,

" Natural Circulation Cooldown," says that when using the S/G PORVs for cooldown, STOP the fl ow through the valves for a period of 5 minutes at every 25 degrees F interval.

WHAT is the bases for doing this?

DUESTION 7.09 (1.25)

What are the IMMEDI ATE acti on steps of E12, " Loss of RCP Component Cooling and Seal Injection?"

(***** CATEGORY 07 CONTINUED ON NEXT PAGE ***+*)

l 1

_____=---__ _-__ -

1

=_

PROCEDURES - NORMAL AE4 NORMAL _EttgRGENgy_9ND PAGE.

15.

7.

2 3

BODIOLOGIC66_CgNIBOL.

QUESTION 7.10 (1.00)

A genera 1' precaution in C14, " Component Cooling System," warns:that flow from ONE Component Cooling Pump during continuous operation, should not go bel ow 230 gpm or above 4000 gpm.

WHAT are the reasons for these restrictions?

o DUESTION 7.11 (1.25) l.

A general precaution in C15, " Residual Heat Removal System," states that I

both CC pumps should be oper.ated when the RHR system is above 225 degrees F and both RHR heat exchangers should have CC flow through them.

WHY is this required?

QUESTION 7.12 (1.00)

A CAUTION in ECA-3.2, "SGTR with Loss of Reactor Coolant: Saturated Recovery," states that feed flow SHOULD NOT be initiated to the RUPTURED S/G if it is also FAULTED.

WHAT is the bases for this caution?

l QUESTION 7.13 (1.00) l l

In 1E33, " Failure of $11 RCP Seal," if the 441 seal leakoff indication should peg high, the procedure directs you to immediately go.to Section 2.2, "RCP Seal Leakoff Greater than 6.5 gpm."

WHAT is the 6.5 gpm limit based on?

DUESTION 7.14

(.50) i TRUE or FALSE In ES-3.1,

" Post-SGTR Cool down Using Backfill" while controlling the ruptured S/G level, it is intended for the operator to allow the ruptured ~

j S/G to drain to a low narrow range level and then refill it with feed flow as opposed to continuously f eeding.

J

(****+

CATEGORY 07 CDNTINUED ON NEXT PAGE

• • • • • )

i

,22.__EBggEQUBES_;_NQBM L _8BNgBM L _gMEgGgggy_9yp PAGE_ 16 68D19LQGlg66_QgNIS96 QUESTION 7.15 (1.00)

FR-S.1,

" Response to Nuclear Power Generation /ATWS," step 2 has you verify that the turbine has tripped.

WHAT actions should be carried out by the operator if the turbine has not tripped?

(Include all contingency substeps contained in the response not.obtained column in your answer )

QUESTION 7.16 (1.00)

A general precaution in C20.7, " Diesel Generators," states that if the Diesel Generator automatical1y started via a saiety injection - signal, then the MCA relay on Panel G-1 must be reset bef ore stcpping the Diesel Generator.

WHY do you have to reset the MCA relay?

OUESTION 7.17 (3.00)

Step 5 of the IMMEDIATE actions in E-0,

" Reactor Trip or Safety Injection," has you verifying Safeguards Component alignment.

WHAT are SIX indications / components that you have to veri f y to complete thi s step?

(Include in your answer the required condition / position of each component /2ndication.)

DUESTION 7.18

(.50)

TRUE or FALSE If you have an ORANGE condition on 1 F R--C.1,

" Response to Inadequat e Core Cooling," during execution of ECA O.0,." Loss of All AC Power," you should l

l first check the other status trees f or RED conditions, and if no RED conditions exist, IMMEDIATELY implement the ORANGE path procedure, l

QUESTION 7.19 (1.50)

Step 16 of E-0,

" Reactor Trip or Safety Injection," has you STOP both RCPs if you have at least one SI pump running, and RCS pressure is equal to or j

less than 1200 psig.

WHAl is the reason for NOT running the RCPs BELOW this pr essure?

(Assume normal containment con-ditions exist)

)

1

(***** END OF CATEGORY 07 ***++)

l i

i

92_dBPMINISIBBIIME_EBgCEDUbES3_C9BplIlgNS _8Np_6]d1I8IJgNS PAGE 17 3

QUESTION 8.01 (1.50)

.WHAT is the Technical Specification BASES for restricting the quantity of radioactive material contained in outside temporary tanks to 1ess than or equal' to 10 curies?

QUESTION 8.02

(.50) 1 RUE or FALSE The Shift Supervisor that 2s required per the minimum shift crew composition is allowed to assume the operator's duty on either unit in.the case of an unexpected absence.

QUESTION 8.03

(.50)

TRUE or FALSE The Unit 1 Shiit Super vi sor becomes the Emergency Manager during-an emergency condition on Unit 2 until properly're.lieved by the Plant Manager.

(Assume an Alert Condition was declared on Unit 2) l l

DUESTION B.04 (2.00)

Answer EACH of the the f ollowing questions concerning SW1-0-3, "Saf eguards Hold Cards and Component Blocking or Locking":

a.

How is a component identified as requiring a Block or Lock? (0.75) b.

Whose permission i s required to remove a Block or Loc k ?.

(0.75) c.

TRUE or FALSE When a Bl ock or Lock devi ce i s removed, it'is returned to the plant Mechanical Maintenance Foreman.

. (0. 5)

DUESTION B.05 (1.50)

According to Technical Specification BASES, WHAT is the reason that.no more than 45 recently discharged fuel assemblies shall'be located in-the j

small fuel pool (pool number 1)?

1 l

(***** CATEGORY 00 CONTINUED ON NEXT PAGE'*****)

'1 a

8.

GDMINISIBSIIyg_BBgggDUBgS _CQUD1110932_SUD_61MlISIlgNS_

PAGE 10 3

l QUESTION 8.06

(.50)

TRUE or FALSE When using procedures that are applicable 'to both' units, the valve numbers, equipment numbers, and parameters that are NOT applicable MUST be deleted by drawing a line through them.

QUESTION 8.07 (2.00)

In accordance with 5ACD 4.1, "Pl ant Operations Manual," WHAT MUST be done to' ensure that the portion of the Operations Manual -you are using is current?

(List 2 possible methods.)

OUESTION 8.08 (1.50)

Answer FACH of the f ollowing questions TRUE or FALSE concerning 5AWI 5.1.1,

" Security Policies and Procedures,":

a.

If more than one operator is entering a vital. area to hang hold

~

cards, each operator must insert their badges into the card reader, but only the first operator has to enter the reason for entering the area on the cipher pad.

(0.5) b.

The maximum number of vicitors that any person can escort into a vital area is FIVE.

(0.5) c.

When escorting visitors into a vital area, the visitors will place their card into the card reader, and then'the escort will une his/her-badge to open the vital area door.-

(0.5)

DUESTION 8.09 (1.00)

Answer EACH of the following questions TRUE or FAL.SE concerning 5ACD 3.9,

" Bypass Control"t 1

a.

The Shift Supervisor i n the onl y indi vi dual who can' authorize the use of a bypass.

6.

For bypasses that are installed f or troubleshooting purposes, independent verification is NOT required on installation or removal provided the' jumper is returned to the control room upon' completion of the troubleshooting.

)

i

(+****

CATEGORY 08 CONTINUED ON NEXT PAGE +****)

]

91__BDd1NigIB6IIVE_PBQGEpuBEg2_G9991IlgNS _6NpiLJdlIGIJgUS PAGE 19-2 QUESTION 8.10 (2.00)

Technical Specifications require 'that at - l east' 23 f eet of water be

~

maint ai ned above the reactor vessel flange during refueling operations.

However, the water level may be lowered below 20 feet for upper internals removal / replacement and for' latching and unlatching RCCA drive-shafts..

WHAT are TWO Technical Specification BASES that allows this?

OUEST10N 8.11 (3.00)

Answer each of the following questionc concerning SAWI 3.1.0,1, '" Met hod s of Performing Independent Verification":

a.

In a case where operation of~a 4.16 kv breaker is NOT possible due to plant conditions, the independent verification that the breaker is in the CONNECT position i s done. vi suall y.

WHAT.are FOUR visual checks that have to be made to perform this independent verification?

(Assume you have approval f rom the General Superintendent of Plant Operations.)

(2.0)

~

b.

WHAT actions must be perf ormed to independently verif y that. a manually operated rising stem. valve is in the OPEN position?

(1.0)

QUESTION 8.12 (1.50)

Answer EACH of the f ollowing questions TRUE or. FALSE concerning 5AWI 3.10.3, "Use of NSP Safety Tags":

/

a.

HOLD or SECURE Cards may be temporarily removed and equipment temporarily restored provided that the cards are released by the Shift Supervisor.

(0.5) b.

SECURE Cards can be used in place of UNSAFE Cards, but UNSAFE Cards i

cannot be used in place of SECURE Cards.

(0.5) c.

Authority to order attachment and removal of SECURE Cards f or substation switching shall be the responsibility of the System.

Di sp at c h er.

(0.5)

)

I OUESTION 8.13 (1.00)

J WHAT is the Technical Specification BASES for a maximum temperature differential of 44 degrees F between the average containment and annulus air temperatures?

q I

(*+***

CATEGORY OB CONTINUED ON NEXT PAGE **+**)

. ; 91___G901NISIBBIIYLEBQCEpyBEL_,CQNplIlgNL _@Np_LINII@IlgNS PAGE 20 QUESTION 8.14 (1.00)

.In accordance with SWI-0-22, " Tech. Specs. Rel ated Shutdown Requi r tments,," -

how long do you have to initiate a. unit' shutdown if the RWST is BELOW its Tech. Spec. required 1evel of 200,000' gal 1ons?

QUESTION 8.15 (1.50)

WHY has, a pl an been proposed to curtail liqui d releases by transferring the CVCS Monitor Tanks to the Spent Fuel Pit?

QUESTION 8.16

(.50)

TRUE or FALSE In accordance with F6, "Chemi cal Leaks and Spills," the bect method of combatting a caustic spill is to NEUTRALIZE it.

QUESTION B.17 (1.50)

In accordance with F2, " Radiation Safety," WHAT are THREE precautions thz.t' MUST be taken whenever entries are made into an oxygen deficient atmosphere?

(***** CATEGORY 08 CONTINUED ON NEXT PAGE..*****)

191__BDd1NJ3IB611ME_BBQCEDUBES _CONp]IlgyS _GNp_6]M11GIlpNS PAGE

21. i 2

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QUESTION 8.18 (2.00)

One hour after a Unit i reactor trip, a GENERAL EMERGENCY was' declared ori F3-2, Initiating Condition No.

6, Case 1,

loss of' clad,.LOCA, and a high-potential for loss of contai nment on Unit 1.

(Containment pressure was 29 psig and decreasing when additional failures resulted in 3 fan coil units and 2 containment spray pumps f ailed. )

Radiation monitor readings are as follows:

1R-48 21000 R/hr (alarming)

IR-49 20000 R/hr (al armi ng )

IR-22 30 cpm 2R-11 40 cpm IR-50.2 mr/hr 2R-50.3 mr/hr Other plant conditions:

Wind from 315 degrees at 5 mph Stablity Class D Temperature 80 degrees F Using the attached procedure F3-8.1, WHAT are the Protective Action Recommendations, (i f any), based on the above given information?

1 i

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(***** END OF CATEGORY 08 *****)

J

(************* END OF EXAMINATION ***************)

.s PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 Reviewed by:

Supt Rad P ot crion Effective Date:

5~sp/-5*57 Approved by:

Title:

• *"* hk"8'#

RECOMMENDATIONS FOR OFFSITE OC Review

• C-N ~7 PROTECTIVE ACTIONS FOR THE ON SHIFT EMERGENCY DIRECTOR /

SHIFT SUPERVISOR 1.0 PURPOSE The purpose of this procedure is to provide guidelines to establish the basis upon which protective action recommendations i

are to be made to the offsite authorities responsible for i

implementing such actions.

2.0 APPLICABILITY This instruction shall apply to the on-shift Shift Supervisor /

Emergency Director.

3.0 PRECAUTIONS 3.1 Initiation of protective actions for offsite areas is the responsibility of the State of Minnesota and the State of Wisconsin.

If it is determined, by the Emergency Director, that, immediate protective actions are' required, and the State EOC's are not activated, the Emergency Director SHALL authorize such recommendations to be made directly to the local authorities.

Once the State EOC's are activated, all protective action d

recommendations SHALL be made to th'e State EOC's.

3.2 The protective actions in this procedure are limited to protective actions for minimizing the exposure of the public within the 10 mile plume exposure pathway, to external and internal radiation exposure from the passage or inhalation of the radioactive plume.

Other protective actions for minimizing the exposure of the population within the ingestion exposure pathway should be determined and implemented by the appropriate state authorities.

4.0 RESPONSIBILITIES l

1 4.1 The Emergency Director has the non-delegatable j

authority to authorize protective action recommendations.

(

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l IBM PAGE 1 OF 12

1 EMERGENCY PLAN IMPLEMENTING PRAIRIE ISLAND NUCLEAR e

GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 5.0 DEFINITIONS 5.1 Plume Exposure Pathway, The 10 mile radius around Prairie Island defined for the short term " plume exposure".

The principal exposure sources from this. pathway are:

(a) whole body external exposure to gamma radiation from the plume and from.

1 deposited materials and (b) inhalation exposure from the The duration of principal passing radioactive plume.

potential exposures could range'in length from hours to days.

5.2 Plume Emergency Planning Zone (EPZ)

A defined area around:the plant to facilitate emergency planning by state and local authorities, to assure that and effective actions are taken to protect the prompt public in the event of a release of radioactive material.

It is defined for:

Plume Exposure Pathway a.

i A 10 mile radius around the plant where the principal exposure source.is: (1) whole body exposure to gamma from the plume and from deposited material; and (2) inhalation exposure from the passing radioactive plume (Short-Term Exposure).

b.

Ingestion Exposure Pathway A 50 mile radius around the plant where the principal exposure would be from the. ingestion of contaminated' water or foods such as, milk or fresh vegetables (Long Term Exposure).

5.3 Projected Dose An estimate of the radiation dose which affected individuals could potentially receive if protective actions are not taken.

5.4 Protective Action

)

An action taken to avoid or reduce a projected dose.

(Sometimes referred to as protective measure).

\\_

PAGE 2.0F 12 IBM

PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING j

5

-GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 i

5.5 Protective Action Guide Projected absorbed dose to individuals in the general population which warrants protective action.

5.,6 Public Alert and Notification System (PANS)

(1)

The Public Alert and Notification System (PANS) is used to alert the public within the 10 mile Emergency Planning Zone, of'an emergency-condition at Prairie Island.

Once alerted, the public should then turn to local commercial broadcast messages fgr specific protective action instructions.

The Public Alert and Notification System surrounding Prairie Island consists of the following:

a.

Fixed Sirens for'100% coverage throughout the 5 mile zone =and in population centers in the 5-10 mile zone.

b.

Emergency vehicles with sirens and public

~

address in the 5-10 mile areas not covered by fixed sirens.

c.

National Oceanic and Atmospheric Administration (NOAA) L:tivated tone alert radios in institutional,. educational,.and commercial facilities.

d.

The-Emergency Broadcast System (EBS) which has access to television and' radio stations.within 4

the area.

1 (2)

Activation of the Public Alert' and Notification System should be recommended if any protective action such as sheltering or evacuation has been recommended for the public within all..or a portion of the 10 mile Emergency Planning Zone.

5.7 Designation of Affected Areas The designation of the area requiring protective actions is going to depend on the nature and extent of the 1

incident and existing meteorological conditions.

1 Generally, the affected area should resemble a keyhole, consisting of a 360* area surrounding the facility, 7

out to a distance of about two to'five miles and I\\

u IBM' PAGE 3 OF 12 I

9 e

4 PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING.

GENERATING PLANT-PROCEDURES KTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1

\\

continuing in the downwind direction, which should.

include'one-sector on either side of-the affected sector, out.to a distance, as. determined by the Protective Action Guides.

Initial efforts for protective actions-should'be in the downwind direction.

O PROCEDURE 6.1 Protective Action Recommendations 6.1.1 If a " GENERAL EMERGENCY" has been declared,.

refer to Figure 2, " GENERAL EMERGENCY PROTECTIVE ACTION GUIDELINES".for immediate protective action recommendations.

NOTE:

DO NOT DELAY protective action recommendations during General' Emergency conditions.

Figure 2 contains protective action' recommendations for General Emergency conditions' based on Control Room' indications and requires no dose proj ec tions ~.

6.1.2 Obtain the offsite dose projection: data, s

utilizing F3-13, "Of fsite Dose Calculations".:

6.1.3 Using current meteorolo'gicalLdata, determine-the plume direction and windEspeed. ' Evaluate the potential for wind direction shifting..

NOTE:

Weather-forecast information may be obtained by calling the National

" Weather-Weather Service (See F3-13.6,,he Forecasting'Information" for t numbers to call).

6.1.4 Using MIDAS, obtain the projected dose to the offsite population.

6.1.5 Determine'the appropriate. Protective Action Recommendation by comparing the projected offsite dose with'the Protective Action Guides listed in. Figure.1.

In addition,.

recommend activation of the Public Alert &

Notification System if any protective action such as sheltering or evacuation has been

.{

recommended within all, or a, portion of the 10 mile Emergency Planning' Zone.

]

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PAGE 4 OF 12 <

BM

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PRAIRIE ISLAND NUCLEAR _

EMERGENCY PLAN IMPLEMENTING GENERATING. PLANT.

PROCEDURES NORTHERN STATES POWER COMPANY Number:~F3-8.1 Rev:-1

'l s

6.1.6 Document all Protective-Action.

Recommendations and justifications for such

~ Protective Action: Recommendations on the

" Emergency Notification Report Form", Figure 5 and in the' Operations Log.

6.1.7 The Emergency Director shall make recommendations.for appropriate protective action to State' authorities identifying the affected area by(e.g.er.designationLand by-lett radius in miles.

" Evacuate within_a'2.

mile radius and in sectors C, D and'E out.to 5 miles.

Seek shelter in all other sectors out to 10 miles").

NOTE:

Prior to activation of the' State EOC's, protective action recommendations shall be issued to state and local authorities.- Once the State EOC's are activated, protective action recommendations shall only_be transmitted toistate authorities.

6.1.8 The Emergency Director SHALL then authorize-the " Emergency Notification Report Form" (PINGP 577), Figure 5, to be transmitted to the Minnesota Team Coordinator and-the Wisconsin State Radiological Coordinator via the TSC telecopier.

6.1.9 If, as a result of continuing assessment, dose projection results or meteorological:

conditions change significantly, re-evaluate 1 the recommended protective action and if necessary update the' initial recommendation.

6.1.10 Verification should be made with, state authorities,.regarding actual protective actions being implemented and the affected populace.

?

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IBM PAGE 5 OF 12

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PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 s

FIGURE 1 l

EPA Guidelines for 1

Recommended Protective Action to Avoid Wole Body and l

Thyroid Dose from Exposure to a Gaseous Plume

{

Projected Dose (Rem) to The Population Recommended Actions Comments Whole body (1 No planned protective actions.

Previously Issue an advisory to seek shelter recommended Thyroid (5

and await further instructions. protective Monitor environmental radiation actions may levels.

be reconsid-ered or terminated.

Whole body 1 to (5 Seek shelter as a minimum.

If constraints Consider evacuation. Evacuate exist, spe-unless constraints make it cial consider-Thyroid 5 to (25 impractical, ation should Monitor environmental radiation be given for levels.

evacuation of Control access.

children mod pregnant women.

Whole body 5 and above Conduct mandatory evacuation Seeking Monitor environmental radiation shelter would levels and adjust area for be an alter-Thyroid 25 and above mandatory evacuation based on native if these levels, evacuation Control' access were not I

immediately possible.

f s

Projected Dose (Rem) to Emergency Team Workers Whole body 25 Control exposure of emergency Although res-team members to these levels pirators and Thyroid 125 except for lifesaving missions.

stable iodine (Appropriate controls include should be used time limitations, respirators, where effee-and stable iodine.)

tive to con-Whole body 75 Control exposure of emergency trol dose to members performing a emergency team lifesaving mission to this workers, thy-level (Control of time roid dose may exposure will be most effec-not be the tive).

limiting factor for j

lifesaving 8,

missions.

W 7

1 J

PAGE 6 OF 12 IBM

PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 I /

s$

FIGURE 2 FLOW CHART FOR GENERAL EMERGENCY OFFSITE PROTECTIVE ACTIONS RECOMMENDATIONS The following situations require urgent actions by offsite l

officials.

Conditions are based on Control Room indications with no dose projections required.

The following protective action recommendations in this flowchart shall be made within 15 minutes.

This flow chart should always be entered at Step 1.

1.

Control Room Staff Detects GENERAL EMERGENCY.

A.

If "YES", then: (1) RECOMMEND SHELTERING WITHIN A 2 MILE RADIUS AND OUT TO 5 MILES DOWNWIND IN AFFECTED SECTORS.

ACTIVATE PANS.

(2) Continue with Step 2.

B.

If "NO",

then:

Continue to evaluate plant conditions 2.

Loss of Physical Control of the Plant to Intruders.

A.

If "YES", then:

(1) RECOMMEND EVACUATION WITHIN A 2 MILE RADIUS AND OUT TO 5 MILES DOWNWIND IN AFFECTED SECTORS.

SHELTER REMAINDER OF PLUME EPZ AND ACTIVATE PANS.

(2) Continue with step 7.

B.

If "NO", then:

Continue with step 3.

3.

Substantial Core Damage in Progress or Projected (release of 20% GAP from core).

(See Figure 3).

A.

If "YES", then:

Continue with step 4.

B.

If "NO", then:

Continue with step 7.

4.

Large Fission Product Inventory in Containment

(> 100% GAP activity).

(See Figure 3).

[s A.

If "YES", then:

Continue with step 6.

B.

If "NO",

then:

Continue with step 5.

IBM PAGE 7 0F 12

f.~

s I

EMERGENCY PLAN IMPLEMENTING PRAIRIE ISLAND NUCLEAR GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1 t

FIGURE 2 j

(Continued)

{

FLOW CHART FOR GENERAL EMERGENCY OFFSITE PROTECTi'VE ACTIONS RECOMMENDATIONS 5.

' Imminent Projected Containment Failure and Core Damage or Release Underway.

A.

If "YES", then: (1) RECOMMEND EVACUATION WITHIN A 2 MILE RADIUS AND OUT TO 5 MILES DOWNWIND IN AFFECTED SECTORS.

SELTER REMAINDER OF PLUME EPZ AND ACTIVATE PANS. SHELTER AREAS THAT CAN'T BE EVACUATED BEFORE PLUME ARRIVAL.

CONCENTRATE ON EVACUATION AREAS NEAR THE PLANT. PROMPTLY RELOCATE THE POPULATION AFFECTED BY ANY GROUND CONTAMINATION FOLLOWING PLUME PASSAGE.

(2) Continue with step 7.

B.

If "NO", then: (1) RECOMMEND EVACUATION WITHIN A 2 MILE PADIUS AND OUT TO 5 MILES DOWNWIND IN AFFECTED SECTORS.

SELTER REMAINDER OF PLUME EPZ AND ACTIVATE PANS. PROMPTLY RELOCATE THE POPULATION AFFECTED BY ANY GROUND CONTAMINATION FOLLOWING PLUME PASSAGE.

(2) Continue with step 7.

6.

Imminent Projected Containment Failure or Release Underway A.

If "YES", then: (1) RECOMMEND EVACUATION WITHIN A 2 MILE RADIUS AND OUT TO 5 MILES DOWNWIND IN AFFECTED SECTORS.

SELTER REMAINDER OF PLUME EPZ Ah3 ACTIVATE PANS. SELTER AREAS THAT CAN'T BE EVACUATED BEFORE PLUME ARRIVAL. CONCENTRATE ON EVACUATION OF AREAS NEAR TE PLANT. PROMPTLY RELOCATE TE POPULATION AFFECTED BY ANY GROUND CONTAMINATION FOLLOWING PLUME PASSAGE.

(2) Continue with Step 7.

B.

If "NO", then:

(1) RECOMMEND EVACUATION WITHIN A 5 MILE RADIUS AND OUT TO 10 MILES DOWNWIND IN AFFECTED SECTORS. SELTER REMAINDER OF PLUME EPZ AND ACTIVATE PANS. PROMPTLY RELOCATE TE POPULATION AFFECTED BY ANY GROUhT CONTAMINATION FOLLOWING PLUME PASSAGE.

(2) Continue with step 7.

G4 Continue with dose assessmen:. throughout the emergency situation.

7.

Recommend Protective Actions in accordance with the EPA PAG's.

'l PAGE 8 OF 12 IBM

EMERGENCY PLAN IMPLEMENTING PRAIRIE ISLAND NUCLEAR PROCEDURES e

Rev: 1 GENERATING PLANT Number: F3-8.1 NORTHERN STATES POWER COMPANY FIGURE 3 i a CONTAINMENT DOSE RATE VERSUS TIME 7

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PAGE 9 0F 12 IBM

~

- ~-

~

l.

1

-PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING GENERATING PLANT PROCEDURES NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1

\\

FIGURE 4 EVACUATION TIME ESTIMATES

• Evac. Zone Distance Evacuation Time (Minutes)

Map.

Sectors (Miles)

Population Best Adverse Confirmation P-1 Q-F 0-2 238 10 13 26 P-2 G-P 0-2 111 10 13 26 P-3 Q-B 0-5 370 22 29 58 P-4 C-F 0-5 610 22 29 58 P-5 G-K 0-5 2302 22 29 58 P-6 L-P 0-5 378 22 29 58 P-7 Q-B 0-10 2153 42 55 110 P-8 C-F 0-10 4355 42 55 110 P-9 G-K 0-10 14016 76 100 200 s

P-10 L-P 0-10 1790 42 55 110 E

E 0-10 713 8

11 22 F

F 0-10 2238 12 16 32 G

G 0-10 10057 43 56 112 H

H 0-10 2617 8

11 22 l

i Taken from Evacuation Time Estimates, dated March 1981, Rev. 3 and Evacuation Time Estimates for 4 worst case sectors, dated March 1981

\\-

IBM PAGE 10 OF 12

EMERGENCY PLAN IMPLEMENTING

' PRAIRIE ISLAND NUCLEAR PROCEDURES GENERATING PLANT Number: F3-8.1 Rev: 1 NORTHERN STATES POWER COMPAtlY FINCP 577, sev. 5 FIGURE 5 Retention: Life Page 1 of 2

"' I." ROLNCY NOTIFICATION RIPORT TORM N

1. Notify State / Local authorities within 15 minutes, Instructions:

with information contained in Part 1.0.

2. Part 2.0 to be transmitted on verification calls.

l l

l PART 1.0 l

l l

1.1 Plant Identification

, Shif t Imergency Connunicator at This is the Prairze Island huclear Generatzng Plant.

1.2 Event Classification (a) Notification of Unusual We have:

_ (a) Declared a(n)

Ivent

_ b) Alert

(

a(n)

JA ovngt n)

Area Imergency General Imer2ency

_,,(d) Termina (date)

At o

1.3 Release inferna

%W (a) DOES NOT involve a radioactive release The emergency:

radioactive release (b) DOIS involve a 11 quad /airporte 1.4 Protective ' Action Reconcendation The protective action recommended at this time is:

_ (a) None sectors out to siles

_ (b) Shelter in sectors out to alles sectors out to miles

_ (c) Ivacuate sectors out to miles

_ (d) Activate the Public Alert and Rectification System PIT.AII PT AI TEIS IKTOP2'ATION TO T012 IERGINCT ORGAKIIL'IOK 1.5 PERSOKKT.J..

IERGENCY DIEICTOR APPROYAI.

hame Date/Iame PAGE 11 OF 12 IBM

PRAIRIE ISLAND NUCLEAR EMERGENCY PLAN IMPLEMENTING

)

GENERATING PLANT PROCEDURES 4

i NORTHERN STATES POWER COMPANY Number: F3-8.1 Rev: 1

\\

FIGURE 5 (CONT'D)

PINCP $77, hev. 5 i

Page 2

• • • DERCENCY NOTITICA!!ON RIPORT TORM ***, Cont'd I

i l

PART 2.0 l

i I

.2 Ivent Descrittien The initiating event causing the emergency is:

O 3

m a uv' CTDep"M men

\\W5%wmWS

~

A gww" g%ss" Q

(

2.2 Meteerelerical Ief FCIE: Use 15 r.imute average meteorological data from the los sensor.

Present Meteorological data is:

k h

j (a) Wind Speed eph

%,, 3

. d.,

/#

C (b) Wind direction (from)

Je.,** O g /

• j j

r, N \\h *4

's e

(c) Temperature

• T Pi e,"T s%

f

'I (d) Precipitation 8

s s J

Nt t g

(e) Stability Class

\\\\\\

,e

/F (f) Affected sectors

,e

~$f e

1, t

a ca s :T DIEI: TOR AFPE0%I.

hame Late /Isme s

PAGE 12 OF 12 IBM J

1

p L._IBEQBY_QE_yUCLE88[EQWE'S1E(@yI_QEEB8119%_ELUlQSi_@NQ 22.!

PAGE

- ISEBd9DYN8dlQEn t-ANSWERS -- PRAIRIE ISLAND 1&2-

-87/11/03-LENNARTZ,.J.

jM ANSWER 5.01

'( 1. 00 )

d

)

f-a.

Over predict (non-conservative)

'b.

No Change or Over-Predict c.

No Change f

(0.33 pts each, 1.0 all three)

REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, pgs. B-20 to 6-47 l

015000K506

...(KA*S)

~

ANSWER 5.02 (1.00) d.

(1.0)

REFERENCE Westi nghouse -Thermal-Hydraul i c Pr i nci pl es, ~ pgs. 10-67 through 10-71 193OO3K125

...(KA'S)

ANSWER 5.03 (1.50)

JW F'u 239 concentration increases 4+=69.(while U 235 a.

Decreases L.253 concentration decreases) which has a smaller Beta Bar than U 235 miou11 m.

L ct.

Ci'.i_ c _ :

.c ; P " _.

D,-

Et'r"J, -d _ ^

i!:,. m i u.

u Ahg-

- 6 "

M-n n r - _- C. ' s a..

a u,. uc u v e la pi ugui u..

b.

Smal l er SUR.

(0.5)

REFERENCE Westinghouse Fundamentals of Nuclear Reactor. Physics, pgs. 7-26 to'7-30, 7-51 to 7-58 192OO3K104 192OO3K106

...(KA'S) l l

l' k:

's h __THgDRY'OF NUCLEAR POWER PLANT OPERATION _E!yJph _SND.

PAGE ' 23' 3

ISEBOOpyNSdlC@

ANSWERS -- PRAIRIE ISLAND 1&2

-87/11/03-LENNARTZ, J.

4 ANSWER 5.04 (2.25) 1 a.

Bottom (0.75) b.

1.

Increase 2.

Decrease 3.

Decrease 1

l (0.5 pts each)

REFERENCE Westi nghouse Thermal-Hydrauli c-Principl es, pgs. 13-18, 13-24,'13-34 193OO8K106

...(KA'S)

ANSWER 5.05 (2.SO) a.

False b.

False (concentrat i on changes with life but not reactivity) c.

True d.

True e.

False (0.5 pts each)

REFERENCE Westinghouse Reactor Core Control, pgs. 4-31, 4-18, 4-21, '4 - 33, and 4-22 OO1000K534 192OO6K105 192OO6K107

...(KA'S) i i

L]

-I 4

_m.______.i__

______m

)

5.

THEORY OF NUCLEAR POWER PLANT OPERATION f-lUlDS _AtjD' PAGE 24 t

2 IHEBMODINGMlCS g

ANSWERS -- PRAIRIE ISLAND 102

-87/11/03-LENNARTZ, J.

ANSWER 5.06 (2.00)

Q-dot = m-dot delta-h (0.5)

M-dot = 15.0 E 6 lbm/hr.

h-steam = h g at 1025 psia =.1192 BTU /lbm (0,5) h-feed = h-f at 428 degreen F = 405 BTU /lbm OR

= h-f at 1025 psia - (c-p x delta 1) 402 BTU /lbm

= 546 - (1.2 x (548-428)

=

(400-430 BTU /lbm acceptable)

(0.5)

Therefore:

0-dot = 15 E 6 x (1192-402) ~ BTU /hr 1.2 E 10 BTU /hr (0.3)

=

Converting to MW 0-dot = 1.2 E 10 BTU /nr x (1 MW/3.41 E 6 BTU /hr)

= 3470 MW (0.2) (3350-3500 MW acceptable)

REFERENCE Westinghouse lhermal-Hydraulic Principles, pg. 5.4 193OO3K125 193OO7K108

...(MA'S)

ANSWER 5.07 (1.00) b.

Is the Same As (1.0)

REFERENCE Westinghouse Reactor Core Control for Large PWRs, pgs. 9-16 to 9-17 192OO8K105

...(KA'S)

E,.__ISEQBy_QE_NLJQLESB_EQWEB_CLONI_QEEBBIlONx_ELMIQQ2_6NQ -

PAGE 25 ISEBdQQYN8dlQQ ANSWERS -- PRAIRIE ISLAND 1&2

-87/11/03-LENNART2, J.

ANSWER 5.08 (2.00)

Rod worth is a f unction of the,mlativa f1un difference between the 41un at the rod tip and the core average flux.

(1.0)

When a rod is stuck out with all other rods inserted, that rod " sees" a much higher flux than the rest of the core, and could be worth 1000 pcm (0.5)

If a rod is dropped, that rod depresses the flux in the area it is in relative to the rest of the core, and would be worth only about 2OO'pcm.

(0.5)

(Similar wording wi)1 be accepted)

REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, pgs. 6-26-to 6-32 192OO5K105

...(KA'S)

ANSWEh 5.09 (1.00) c.

(1.0)

REFERENCE Steam Tables 193OO3K125

...(KA*S) 4

5t__ISE98Y_RE_Nyg(g68_EQWEB_ELON1_QCg8@llQUg_E(ylpS _QUQ PAGE 26!

t

-ISEBdgDYU8dlGE ANSWERS -- PRAIRIE ISLAND'1&2-

-87/11/03-LENNARTZ, J.

ANSWER 5.10 (1.75) a.

Increases [0.253 b.

Increases CO.253 by a' factor of 4 CO.253 c.

Head approaches O psi at' maximum system flow rate (0.5) d.

1.

Excessive noise 2.

Fluctuating discharge. pressure 3.-

Fluctuating motor current' 4.

Reduction in-pump capacity' 5.

Excessive vibration (Similar answers acceptable)

( Ar y 4 0.125 pts each)

REFERENCE Westinghouse Thermo-Hydraulic' Principles,'pgs. 10-32 to 10-48 191004K109 191004K110 191004K113 191004K114 (KA*S)

ANSWER 5.11 (1.25) 1600 psig = 1614.7 psia (.25)

Saturation Temp. for 1614.7 psia = 606.1_ degrees F (interpolated) (.25) 200 degrees F subcooling = 406.1 degrees F (.25)

Saturation Press. for 406.1 degrees F = 264.97 psia (interpolated) (.25) 264.97 psia = 250.26 psig (250 psig +/- 3 accepted) (.25)

REFERENCE Steam Tables OOOO74K101 035010A102 035010K109 193OO1K101 (KA'S)

ANSWER 5.12 (1.50) a.

Calculated Higher Than Actual b.

Calculated Higher Than Actual c.

Calculated Lower Than Actual I

(0.5 pts each) i REFERENCE-l Westinghouse Thermal-Hydrauli c Principles, Chapters 4 and 5 OO2020K501 193OO7K106 193OO7K108

...(KA'S)

'j i

i i

a

'N 5.___ISEQBY_QE_UQCLE88 EQWEB_E68NI_QEEBBIlON _E691DSz_8ND PAGE 27 3

IEEBdggyN8dlC'y

/

l ANSWERS - PRAIRIE ISLAND 1&2

-87/11/03-Li NNARTZ,

J.

t i

1 ANSWER 3.13 (1.25)

The amount by which the reactor core woul d be st&cri tical at hot snutdown I

conditions E.25] if all control rod assemblies were tripped E. 25 ].,

I assuming that the highest worth control rod assembly remained f ully i-withdrawn C.253, and assuming no changes in xenon C.25] or baron concentrations I.25].

(Equivalent wording accepted for full credit) l REFERENCE PI Technical Specification 3.10.A 192OO2K110

...(KA'S)

I ANSWER 5.14 (1.00) b.

(1.0)

]

REFERENCE Westinghouse Reactor Core Control for Large PWRs, pgs. 3-29 to 3-41 192OO4K113

...(KA'S) l l

ANSWER 5.15 (1.00) b.

(1.0)

REFERENCE Westinghouse Fundamentals of Nuclear Reactor F hysi cs, pg. 7-34,7 and PI Plant Information Summary 192OO3K107

...(KA'S) 1 (l

ANSWER 5.16 (1.50) i RCS temperatur e and RCS pressure unaf f ected by the dropped rod.

(.25 each)

Nf pcm

t.-2 p

c:. p t ^

.c p

!_~73, then reactor power will decrease EO.5], and level out at lower power level in the source range (as supported by subcritical multiplication) [ ~-'.

fT

'l

4r-

< __q

'j s,

3 t

PAGE 28' ~

5.

THEORY OF NUCLEAR POWEB_ PLANT _OPERATIgb _F662D,fja_GND M,

A ItLEBOODydSUI[S

/(.

j9

/03-LENNART2, J.

1 ANSWER!

-.s]'AIRIE IGLAND 4k2 I,

y E}? f*k N

\\,5 4 S r

.o f '. (, -(

/ (,' \\ s( N e 1

_N,

s >

s v

A N

\\

t F *EFERENCE I h j,,

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j~

i

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kss%inghouse Reactor Theory Tent

/

j

. 19'ENSK103v s -

1920CLK112)

...'t.'Al.'/1)

\\1 4

(

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(

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g

(

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h AhfSWEfi 17

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\\ Ase t

'/,

Dec(\\ase b.

A Det

_s c.

s o,

(0,$gots,each) g 1

k, '

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REFERENC' i

1 Westinghouse.Fundamenta)s of Nuclear 'f cactor Phynics,. Chapter 8, p)g s. 12-19 i

,e

...(KA\\-)

Q15 COOK 5%

19200t?r.104 i

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h_ hMNI_ SY SIE!1S E S LG h _ C O NIBOk BN12,'J U S1B yd EUI[tTI O N PAGE 29 ANSWERS -- PRAIRIE I St. AMD 1 &2

-87/11/03-LENNART7, J.

ANSWEF 6.01 (2.00) 1.

Comoresscr starts on low air pressure 2.

C.;wpressor stapped by unload timer 3.

'J.omp r es s or stopped by load rejection 4.

Coreressor stopped by low oil pressure 5.

Compressor tiipped on motor thermal overload (Any 3 9 0.5 pts ench) 6.

Falci.

(O zS l

REFERENCE PI Instrument ud Static,n Air Lesson Plan, Temp. Chenge D6-44 and pg.

19 078000G008 07800DK103

... O :: A ' S )

i AN9IJER 6.02 (2.50) 1.

Open 2.

Open 3.

Closed 4.

Cl osec) 5.

Open (0. 5 pt s; eatn)

REFERENCE

( ', B-14, Ct,mponent Cooling Water, pg.

17 P1 B12A, Onemical and Volume Ccatrol System, pg. 16 Figare B12A-2 004010A203-OOOO10A201 078000K302

...(KA'S) i ANSWER 6.03 (1.00)

Anywhere from 0-100%.

(.50)

Output will be the signal necessary to cer. tor e ' actual pressure to the setpoint.

(0.50)

REF ERENC

RI.F2:1 Pressure Control Syst em i.uuson Pl an Quiz Duestions $t19 and U2O 000027A21.G

,..(KO*S) i

)

-6 __PLONI_@Y@lgd@_DE@l@N2_QQNIBQ61_GMD_IN@IGQUENIGIlON PAGE 30 t

ANSWERS -- PRAIRIE ISLAND 1&2

-87/11/03-LENNART2, J.

ANSWER 6.04 (1.50) 12 Motor Driven AFW' pump auto started on 12 S/G lo-lo' level (13%).

(0,5) and then tripped due to low' discharge pressure (<500psig). (0.5)

To prevent this, the 12 Motor Driven AFW pump Selector' Switch'should have been placed iri Manual. (0.5) b REFERENCE PI LER 87-006-00 PI B-28, AFW p.

7 061000GOOB 061000K101 061000K406

...(KA'S)

J 1

ANSWER 6.05

(.50)

. False.

(0.5)

REFERENCE PI CVCS Lesson Plan, pg. 10 PI B3, Reactor Coolant Pumps, Figure B3-5 j

OO3OOOK103 OO3OOOK602

...(KA'S)

ANSWER 6.06 (1.00)

The air operated butterfly valve will reopen if delta P increases to -0.4 psid [0.53, allowing air to flow from the shield building annulus into' j

containment [0.53.

REFERENCE P1 Containment System Lesson Plan Temporary Change #87-23, pg. 1

'103OOOK107 103OOOK401

...(KA'S)

I 1

ANSWER 6.07 (1.00)

Failure of. the air regulator caused CV-31939, (CA Isolation Control Valve), to open [0.53 and contaminate the spray pump suction line with Sodium Hydroxide (NaoH)

[0.5].

j 1

g icant Operating Event Report, #P-SOE-2-87-1

)

078000K302

... (1K A ' S )~

l

.i 1

h __E66dI_EYSIEdE_QEglGh _CQNIQQ62_GNQ_lN@lBydENIGllON

..PAGE 31 ANSWERS'-- PRAIRIE ISLAND 1&2

-97/11/03-LENNART2, J.

i ANSWER 6.08 (2.00) a.

Zero

'( 0. 5 )

b.

During the monthly load test,.the EDG i s operat ed in parallel, IO.753 so speed droop is set at 40 to prevent largeiload swings-when. load i s adjusted [0.753.

REFERENCE PI Diesel Generator Lesson Pl an, pgs. 11, 25 064OOOK403 064000K406

...(KA'S)

ANSWER 6.09

(.50)

True. (0.5)

REFERENCE PI Di es el Generator Lesson Plan, pg. 18 PI D.

C.

Distribution Lesson Plan, Attachment 1 063OOOK301 064000K104

...(KA*S) i-O I

l l

1 i

l ll 4

2--_______.----

h_._ELONI_gygI[idg_95SJ E _QgNIBQL2_8ND_JUSIBydENIBIlgN PAGE 32-ANSWERS -- PRAIRIE ISLAND 1&2

-87/11/03-LENNARTZ, J.

ANSWER 6.10 (2.00) l ESF Bus 15

'1.

1R transformer E.1253 - undervoltage sensed by-voltage restoring scheme, will.not close breaker-(15-3) 0.1253.

2.

Bus tie to Bus 26 E.1253 - D1 supplying bus 26 so voltage restoring scheme will not close bus cross tie (26-8) E.1253.

l

3..

D1 (EDG) E.1353 - voltage restoring scheme will close D1 breaker 15-2 E.1253 (Auto / Manual switch for 15-2 in Auto).

1 Other possible cources of power to Bus 15 1.

1M Transformer (.25)

ESF Bus 16 1.

CT-11 Transformer E.1253 - undervoltage sensed by. voltage restoring

~

scheme, will not close breaker (16-0) 0.1253.

2.

Bus tie to Bus 25 E.1253 - vol tage restoring scheme will close breaker (25-8) E.1253.

Other possible sources of power-to Bus 16 l

1.

D2 (EDG) (.25) 2.

1R Transformer (.25)

REFERENCE PI 4160, 4GO, 120 VAC Inst. Safeguard Dst. System Voltage Restoring Scheme Handout l

PI Circuit Di agram #40020-1 062OOOk104 062OOOK403 064000K101

...(KA'S)

-l 1

1 l

l i

i

h __ELQNI_@y@ led @_RE@l@N2_CQNIBQl 1, GNO_IN@lB('dENI@IlgN PAGE 33 y

ANSWERS -- PRAIRIE ISLAND 18< 2

~B7/11/03-LENNARTZ, J.

ANSWER 6.11 (2.00) a.

The Channel Current Comparator is used to detect radial flux tilts

[.1253 and faulty PR channels E.1253 by comparing the highest PR Channel reading E.253 to the lowest f'R channel E.2SJ and alarming when there is a 2% deviation E.253.

b.

No EO.53, P-10 (2/4 PR.) 10%) is a backup to P-6 (SR block p er mi ssi ve at 1/2 1R > 10 -10 amps) and the SR Dlock Switches E.253 and will prevent the SR instruments from energining E.253.

REFERENCE P1 NIS Lesson Plan, pgs. 33, 48, 49 015000 GOO 7 015000K401 015000K603

...(KA'S)

ANSWER 6.12 (1.60) d Lblo Totve PLO Tscl' a.

1.

Steam dumps b @QTm % L#4-2.

Rod control f.M gg I

3.

PZR level 4.

Tavg - Tref devi ati on alarm 4

g7 Tavg deviation alarm I

v.

/0 0P OSt2L T (Any 4 G O.2D pts each) b.

False.

(0.5)

REFERENCE PI Reactor Process Instrumentation System Lesson Pl an, pg. 7 016000K402 016000K403

...(KA'S)

ANSWER 6.13

(.50)

False.

(0.5)

REFERENCE P1 Reactor Process Instrumentation L esson Pl an, pgs. 26-29 OO1000K602

...(KA'S)

1 6 __EL@NI_EYSIEUE_ DESIGNS _CONIBOL1_@ND_INSIBydENI@IlON PAGE 34 t

ANSWERS -- PRAIRIE ISLAND 10<2

~87/11/03-LENNARTZ, J.

ANSWER 6.15

(.50)

True. (0.5)

REFERENCE PI Radiation Monitoring System, B-11, pg. 4 072OOOA101 072OOOGOOB

...(KA'S)

ANSWER 6.16 (2.50) a.

Yes EO.53, the rod bottone bypass bistables will block the Rod Bottom / Rod Drop alarm CO.5J until the Control Bank C bank demand posi ti on reaches 35 steps CO.53.

b.

Yes [0.53, rod bottom lights function at all times EO.253 and will be on when the rod is below 20 steps EO.253.

(

REFERENCE P1 Rod Control and Rod Position Indication Lesson Plan, pg. 67 OO1000K401

...(KA'S) 1 1

ANSWER 6.17 (1.25) a.

1.

Wide range hot leg temperature 2.

Wide range RCS pressure 3.

Level transmitter (delta pressure cell) 4.

Reference leg temperature (Any 3 0.25 pts each) 9 b.

True. (0.5)

REFERENCE PI RVLIS Lesson Plan, pg.

3, and RVLIS-2.

OO2OOOK603

...(KA'S) o 9

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

6,._ _E66NI_@ y @lg d @,, DES l@ N 3_QQN1B QL3,_6N D _lN @lBQ U ENIGIlgd

.PAGE' 35 ANSWERS -- PRAIRIE, ISLAND 1&2

-87/11/03-LENNART2, d.

ANSWER 6.18 (2.00).

a ~.

. Arm and Actuate-6.

No Effect' c.

Arm Only d.

Arm Only (0.5 pts each)

REFERENCE PI Steam Dump and PORV Control ' Systems Lesson - Pl an,' pgs. ' 8, 16, 15 041020K105

'041020K418 041020K501

...(KA*S) 9 b

s 1

l

.1

.i j

.j

_1_1 _ __2 __.c_____

__._______._____u__.____________

_a

.Zu PRQGEDgRE@_ _NOBdALx_GBNQBdGLt_EdERGENQY_AND.

,PAGE 36-BGQ1969ElGG6_G9BIE96 ANSWERS'-- PRAIRIE ISLAND 18<2

-87/11/03-LENNARTZ, J..

i ANSWER 7.01 (1.00)

This allows'the upper head to cool (to 350 degrees F) [0.53, and-prevents void formation when depressurizing the RCS (to 400 psig) EO.53.

REFERENCE PI ES-0.3, " Natural Circulation Cool down, " Background Info, pg. 4 OO2OOOK514

...(KA'S)

ANSWER 7.02-(1.00)

Since tube rupture flow is directly related to the size of the rupture EO.5], the S/G with the smaller tube. rupture is. recognized by having the lower secondary water level end is considered as non-ruptured for subsequent steps EO.53.

(Alternate wording accepted for full credit)

REFERENCE PI E-3, "S/G Tube Rupture," Background Info, pg.~1 OOOO38 GOO 7

...(KA*S)

ANSWER 7.03 (2.50) 1.

Flow indicator (1FI-474) fails high 2.

High Steam Flow blue status light'on 3.

High-high steam flow blue status light on 4.

Feed flow less than steam flow blue status light on 5.

If feedwater control in Auto and blue channel selected, then 12 S/G FRV opens

{

I 6.

12 S/G flow mismatch annunicator 7.

12 S/G isolation. channel alert annunciator B.

12 S/G Hi-Hi steam flow annunciator (Any 5 0 0.5 pts each) i 1

y

u Zt__ESQGEQQBE@_;_NQBd@61_@pNQSd@Lx_gdE6@ENGI_@ND.

PAGE-37 6801069EIGOL_QQNIBQL' ANSWERS -- FRAIRIE ISLAND 1&2

-87/11/03-LENNARTZ, J.

1 l

REFERENCE l

PI) Instrument Failure Guide IC 51.3,.pg. 25 l

035000 GOO 8

...(KA*S) l; l

l l

.ANSWE 7.04 (1.00) 1.

Labyrinth seal delta P 2.

l' ump ' radi al bearing temperature L

3.

1. 1. seal leakoff temperature l-(Any 2 O O.5 pts each)

REFERENCE PI C3, Reactor Coolant Pump, pg. 5 OO3OOOA109 OO3OOOA201

...(KA*S)

ANSWER 7.05 (1.50)

> 100 psig is to prevent any backflow of dirty water or undissolved gases from the VCT CO.75].

i 1000 psig is to prevent cocking of the #2 seal EO.752.

REFERENCE PI RCP Lesson Pl an Temporary Change. 86-53.

003000G010

...(KA*S)

ANSWER 7.06 (1.50) a.

Emergency Boration Flowmeter Outlet Valves ar e throttled (to.1;2 gpm).

(.75) b.

Higher boric acid flows will plug the seal injection throttle valves.

(.75) i REFERENCE PI C12-8.4, " Emergency Boration of the RCS" TM 60 OOOO24 GOO 7 OO4010K609

...(KA*S)

1 -

7.-

-PROCEDURES'- NORMALg_ADNQRMALa_EMERGENQYiAND PAGE'.'38 BADigLQGig& _QQNISQL ANSWERS -

PRAIRIE 1SLAND 1&2

-87/11/03-LENNARTZ, J.

1 l

l.

ANSWER 7.07 (2.50) a.

No b.-

No i

l c.

Yes d.

No e.

Yes (0.5 pts each)

REFERENCE PI; Technical Specifications,.3.6.B, 3.'1. c. 5,

3.1.4.d, 3.2.B.3, 3.3.A.1.a OO2OOOG01.

OO6000G011 103OOOG011

...(KAfS) l ANSWER 7.08 (1.00)

It allows the stroke of the snubber [0.53 to equalize with the Main Steam System expansion movements E0.5].

REFERENCE PI ES-0.3, Natural Circulation Cooldown, Background Information, pg. 2-OOOO55 GOO 7

...(KA'S)

ANSWER 7.09 (1.25) 1.

Attempt to start any charging pump to restore seal injection.

(0.25) 2.

Prior to the lower bearing water temperature reaching 200 degrees F, perf orm the f ollowing s (0.25) a.

Trip the reactor-(0.25) b.

Stop the RCP's (0.25) c.

Close the #1 seal leakoff isolation valves for both pumps.

(0.25)

REFERENCE PI E12, " Loss of RCP Component Cooling and Seal Injection".

003000G014

...(KA*S)

)

]

l

Zn__BBgCBDQBES_ _NOBd862_GBUDBdG61_BUEBGENCy_GND PAGE 39 BOD 196001CGL_CgNIBQL ANSWERS - PRAIRIE ISLAND 1t<2

-87/11/03-LENNARTZ, J.

ANSWER

'e.10 (1.00) 1 230 gpm - prevent overheating the pump

( fumf b + MW 2.

4000 gpm - prevent pump run out (0.5 pts each)

REFERENCE PI C14, " Component Cooling System," pg. 4 0080000010

...(KA'S)

ANSWER 7.11 (1.25)

The flow control valves on the outlets of the RHR heat exchangers are not a zero leakage valve EO.53, and if the RHR system is above 225 deqirees F-boiling of the Component Cooling System EO.5], in the.out of service RHR heat exchanger can occur L.25] (resulting in water hammer and CC surge tank rising l evel ).

REFERENCE PI C15, " Residual Heat Removal System," pg. 3 OO5000G010

...(KA'S)

ANSWER 7.12 (1.00)

Feedina a ruptured S/G that is also faulted could increase the possibi1ity of an uncontrolled cocidown of the RCS EO.53 and may increase the possibility of S/G overfill CO.53.

REFERENCE PI ECA-3.2, SGTR with LOCA: Saturated Recovery, Background Information, pgs.

7, 8

OOOO38K306

...(KA'S)

ANSWER 7.13 (1.00)

The 6.5 gpm limit i s based on the thermal barrier long term cooling capacity in event of a loss of seal injection.

(This'is based on Westinghouse ex p eri men t al data with 95 degreen F CC water at 40 gpm.)

(1.0)

7.

PROCEDURES - NORMAL,_GBNgRMALLEt1EBGENQy,_ANQ -

.PAGE 40 BOE1960@lgGL_ggNIBg6.

^

ANSWERS -- PRAIRIE ISLAND 152

' -87/11/03-LENNART2, J.

REFERENCE

PI E33, " Failure of #1 RCP Seal," pg. 2 003000G010

...(KA'S)

ANSWER

7.14

(.50)

'True.

(0.5)

REFERENCE PI ES-3.1,." Post S/G Cooldown Using Backfill," Background Info, pg. 3 OOOO38K306

...(KA"S) l ANSWER 7.15 (1.00) 1.

Manually trip turbine 2.

Manually runback turbine 3.

Close main steamline -isolation valves -

4.

Close main steamline i so1'at i on bypass valves l

(..25 each)

REFEF.ENCE l

PI FF -S.1,

" Response to Nuclear Power Gener ation/ATWS," pg. 3 OOOO29K306 OOOO29K312

...(KA'S) 1 ANSWER 7.16 (1.00)

Failure to reset the MCA relay prior to stopping the Diesel Generator will i

result in another start of the Diesel EO.53 when speed decreases to the low speed relay setting (250 rpm) EO.53.

REFERENCE PI C20.7, " Diesel Generator," pg. 5 064OOOG010

...(KA'S0 l

Zi__EBgCgDyBES_:_NOBM06.t_8BNgBd861_EMEBGENGL,8NQ -

PAGEl 40' BODig60ElG8k_GONIBQL ANSWERS --- PRAIRIE. ISLAND 162

~-87/11/03-LENNARTZ, J.-

REFERENCE PI E33, " Failure of'#1 RCP Seal," pg. 2

..OO3OOOG010

...(KA'S)

ANS6JER

.7.14.

(.50)

True.

(0.5)-

REFERENCE PI ES-3.~ 1,. " Post S/G. Cool down Usi ng Bac kf i ll, " Background-Info, pg. 3 OOOO38K306

...(KA'S)

ANSWER 7.15 (1.00) 1.

Manually trip turtsine 2.

Manually runback turbine 3.

Close main steam 1ine isolation valves 4.

Close main steamline i sol ati on bypass valves

(.25 each)

REFERENCE PI FR-S.1,

" Response to Nuclear Power Generation /ATWS," pg. 3 OOOO29K306 OOOO29K312

...(KA'S)

ANGWER 7.16

'(1.00)

Failure to reset the MCA relay prior to stopping the Diesel Generator will result in another start of the Diesel EO.53 when speed decreases to:the low speed relay setting (250 rpm)

[0.5].

REFERENCE PI C20.7, "Di esel. Generator, " pg. 5 064000G010

...(KA'S.)

i Zt__EBgCggyBEQ_ _NQBdBL1_6BNQBd661_EdEBQgNCYiGNQ PAGE' 41 RADIOLOGICAL CONTROL i

AN'SWERS -- PRAIRIE ' ISLAND 1&2

-87/11/03-LENNARTZ, J.

ANSWER 7.17 (3.00) 1.

Category I doors - closed 2.

Check Ops log f or any ventilation openings that must be closed-within six' minutes.

3.

SI ready lights - OFF-4.

SI active ' lights - ON with. exceptions 5.

Containment isol ation lights ON with exceptions 6.

Spent Fuel Pool Normal and Rad Waste Building Ventilation Systems -

Shutdown

(.25 pts each indication / component,.25 pts each condition / position)

REFERENCE PI E.0,

" Reactor Trip or Safety-Injection," pg. 4 000007G010 OOOOO7K301

...(KA'S)

ANSWER 7.18

(.50)

False.

(0.5)

REFERENCE PI ECA O.0,

" Loss of All AC Power," pg. 3 000055G012 OOOO55K302

...(KA'S)

A

Z.e.__ BB Q C E D yBE S _;._N O Bd@63_8 pNQBd @ 61_ E dE B @E N C Y _BN D PAGE 42 BOD 10bOEICeL_C9BI696 ANSWERS -- PRAIRIE ISLAND 1&2

-87/11/03-LENNART2, J.

d ANSWER 7.19 (1.50)

Bel ow thi s trip pressure setpoint, steam voiding.is expected in'the reactor vessel in the event of a LOCA EO.753, and if pumps are left on,.

peak clad temperature of the f uel may actually increase LO.53 if RCPs are subsequently lost (secured) [0.253.

(Equivalent wording acceptod)

OR Prevent excessive depletion of RCS water inventory (0.5) which might lead to severe core uncovery (0.5) if the RCP's wero. tripped for any reason later in the accident (0.5).

REFERENCE PI E-D., Reactor Trip or Safety Injection, Background Information, pg. 3 OOOOO7K301

...(KA'S)

M.__8Ed1NIGIB6IlME_ESQCEQQBES _CQNgillgN@2_QNQ_LldlIGIlgNS-PAGE 43 3

ANSWERS

- PRAIRIE ISLAND 1&2

-87/11/03-LENNARTZ, J.

ANSWER 8.01 (1.50)

Provides assurance t hat in the event of an uncontrolled release of the.

contents of the tank CO.5] the resulting concentrations would bo less than the limits of 10 CFR 20 (Appendix B, Table II, Column 2) [0.53, in an unrestricted area EO.5].

REFERENCE PI Techni cal Specifications, T.S.

3.9-9 000059G004

...(KA'S)

ANSWER 8.02

(.50)

False.

(0.5)

REFERENCE PI Section Work Instruction, SWI-0-2, pg. 3 002000G001 194001A103

...(KA's)

ANSWER 8.03

(.50)

False.

(0.5)

REFERENCE PI F3-1, Onsite Emergency Organization, pg. 2-4 PI F3-4, Responsibilities During An Alert, Site Area, or General Emergency, pg. 9 194001A116

...(KA'S) 1 ANSWER S.04 (2.00) a.1.

Designated on the " Integrated Operations Ct_-klist" (by the words Block & Tag or Lock & Tag in Status Column).

OR 2.

Located in Table 1 of SWI-0-3 (Component Blocking and Locking Log)

(Any 1 at 0.75 each) b.

Shift Supervisor.

(.75) c.

False.

(0.5) i

'\\;

i 8 __epdlNISIB@IlyE_BBQCEDUBES _CQNDlIlgNS _GND_(ldlIGIlgNS PAGE 44:

L.

m 2

3

' ANSWERS -- PRAIRIE' ISLAND 102

-87/11/03-LENNARTZ,oJ.

I REFERENCE i

PI Section Work Instruction, SWI-O-3, pg. 2 194001K102

...(KA'S) 1 ANSWER 8.05 (1.50) l To provide assurance that in the event of loss of pool cooling capabilities CO.53, at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are available under worst case conditions to make repairs L0.53 until the onset of boiling [0.53.

REFERENCE PI Technical Specification TS.3.8-4' 033OOOGOO6

...(KA'S)

ANSWER 8.06

(.50)

True-(0.5)

REFERENCE PI Section Work Instruction SWI-0-10, pg. 4 194001A103

...(KA'S)

ANSWER 8.07 (2.00) 1.

Use a red stamped copy 2.

Compare the copy with-a controlled manual 3.

Compare the copy with the section index [0.53 and review the list of temporary memos LO.DJ.

(Any 2 0 1.0 pt each) i REFERENCE PI Administrative Control Directive, SACD 4.1, pg. 20 194001A101

...(KA'S) 4 L_________________..

)

By__GDd1N1@IB@llME_PBggEQUBEgx_gQUplIlgN@i_8NQ_bidllSilgN@

.PAGEt 45' ANSWERS -- PRAIRIE ' ISLAND 1842

-87/11/03-LENNARTZ, J.

ANSWER 8.08'

-( 1. 50 )

a.

False b.

True c.

False (0.5 pts each)

REFERENCE PI Administrative Work Instruction, 5AWI 5.1.1, pgs. 4-6 194001K105

...(KA'S)

ANSWER 0.09 (1.00) a.

True (0.5) b.

True (0.5)

REFERENCE P1 Administrative Control Directive, 5ACD 3.9, pgs.

6, 16 194001K102

...(KA*S) i

)

ANSWER 8.10 (2.00) 1.

Refucling cavity pool has sufficient l evel to allow time to initiate repairs or emergency procedures to cool the core.

2.

During l atching/ unlatching and upper internals removal / replacement, the level is closely monitored because the activity uses this level as a ref erence point.

3.

The time spent at this level is minimal.

(Equivalent wording accepted)

(Any 2 @ 1.0 pts each)

REFERENCE PI Technical Specification'T.S. 3.8-5 034000 GOO 6

...(KA'S)

O a

B,__8DululSIE8IlyE_EBQCEDUBES _QQUQlIlgNS3_6NQ 11dll@llgNQ PAGE-46 x

ANSWERS -- PRAIRIE. ISLAND 1842

-87/11/03-LENNARTZ,'J.

ANSWER 8.11 (3.00) a.

1.

Dreaker is in the connect position 2.

. Springs charged 3.

Manual trip button not depressed 4.

Red or Green light on:cubi,cle door. (DC Control Power Knife Switch closed) 5.

Ensure latching is complete. (Locking tang inserted into racking screw slat)

(Any 4 0 0.5 pts each) b.

1.

Perfarm physical check by:

(a)

Closing val've approximately 2 turnn in the clockwise direction and then reopening OR (b)

Attempting to open the valve by turning in-the counterclockwise direction.

(0.5 pts for either case) 2.

Mter physical check, perform visual check by observing the stem position above the handwheel.

(0.5)

REFERENCE PI Administrative Work Instruction, 5AWI 3.10.1, pp. 5,8 194001K101

...(KA'S)

ANSWER B.12 (1.50) a.

False b.

False c.

True (0.5 pts each)

REFERENCE PI Administrative Work Instruction, 5AWI 3.10.3, pgs. 12-19 194001K102 194001K107

...(KA'S) a 1

i

ec__GDULUISIBGllVE EBQQEQQBEQ3,QQNQlllQNQ1_@NQ_GldlI@Ilgd@

PAGE 47' l

ANSWERS -- F'RAIRIE ISLAND 1&2

-87/11/03-LENNARTZ', J.

1 1

ANSWER 8.13 (1.00)

To provide assurance that offcite doses in.the event of an accident reraain below those calculated in the FSAR.

(1.O)

REFERENCE PI Technical Specifications, TS 4.4-8 103000G006

...(KA'S)

ANSWER 8.14 (1.00) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(1,0)

REFERENCE PI Section Work Instruction, SWI-O-22, pg. 1 OO4000 GOO 5

...(KA'S)

ANSWER B.15 (1,50) l To limit Tritium releases (1.5) l REFERENCE PI C12, "CVCS," pg. 56 194OO1K103

...(KA'S)

ANSWER 8.16

(.50)

False.

(0.5)

REFERENCE PI F6, Chemical Leaks and Spills, pg. 17 194001K110

...(KA'S)

8 __0Dd1NISIBGIlyE_BBQQEQUBES _GQUQlligNS _8NQ_LldlIGIlgNQ PAGE 48 c

t S

3 ANSWERS -- PRAIRIE ISLAND 18<2

-87/.11/03-LENNARTZ, J.

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ANSWER 8.17 (1.50) j 1.

Standby person available to assist in an emergency.

2.

Respirator must be worn-(self-contained breathing apparatus).

3.

Communications' maintained between respirator wearer and standby person.

4.

Person working in the area should be equipped.-with a safety harness and saf ety lines or equivalent provisions.

(Any 3 0 0.5 pts each)

REFERENCE PI F2, " Radiation Safety." pg. 22 194001K113

...(KA'S) l ANSWER B.18 (2.00) 1.

Evacuate all sectors out to 2 miles 2.

Evacuate FGH sectors out to 5 miles 3.

Shc1ter remaining sectors out to 10 miles 4.

Activate the Public Alert and Notification System (0.5 pts each)

REFERENCE PI Lesson Plan, "Of f site Protecti ve Acti on Recommendati ons," Example No. 3 PI F3-8.1, "Recommendati ons f or 0 fisi te Protecti ve Acti ons. "

194001A116

...(KA'S)

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