Text

Exam Rept 50-361/OL-85-02 on 850730-0814.Exam Results:All Candidates Passed
	ML20198C458
Person / Time
Site:	San Onofre, 05000730
Issue date:	10/17/1985
From:	Morrill P, Pate R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	;
Shared Package
ML20198C433	List: IR 05000730/2008014; ML20198C428;
References
	50-361-OL-85-02, 50-361-OL-85-2, NUDOCS 8511120097
	Download: ML20198C458 (113)
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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Examination Report No. 50-361/0L-85-0A Facility Name: San Onofre Nuclear Generating Station Units 2 and 3 Docket Nos. 50-361 and 50-362 Examinations Administered at: San Onofre NGS, San Clemente, California from July 30 to gust 15. 1985 Chief Examiner:

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Mdrrill, erator Licensing Examiner Dath Signed

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ate, Chief, Operations Section Da/eS%ned

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Summary:

Examinations on July 30 through August 15, 1985 Written and operating examinations were administered to fcurteen initial candidates for R0 licenses. Written and operating examinations were administered to nine initial candidates for SRO licenses and a written examination was administered to one reapplicant for a SRO license. Eight candidates passed the written examination for R0 and six candidates passed the written examination for SRO. All candidates passed the operating examinations.

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REPORT DETAILS 1.

Persons Examined:

SRO Candidates:

R0 Candidates:

Ten Candidates Fourteen Candidates 2.

Examiners:

P. Morrill, RV D. Willett, RV J. Elit, RV

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R. Main OLB:HQ G. Strier, EG&G G. Johnston, RV J. Smith, PNL J. Upton, Jr., PNL

Lead Examiner 3.

Persons attending the exit meeting:

SCE H. Morgan, Operations Manager W. Kingsley, Units 2 and 3 Training Manager J. McKinnon, Quality Assurance M. Hyman, Training M. Metz, Compliance NRC l

R. Maines, OLB:HQ

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D. Willett, RV P. Morrill, RV:

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Written Examination and Facility Review:

Written' exams were administered as follows:

/e Fourteen Reactor Operators - July 30, 1985 Ten Senior Reactor Operators - July 30, 1985 During the written examination, the examiners observed that the facilities provided by the Utility appeared to meet NRC requirements, but may not have been in the best interest of the candidates, Specifically, the test room was near an outdoor swimming pool which became increasingly noisy in the afternoon, the lighting appeared poor in some areas, and some tables were small which prevented the candidates from spreading out their work.

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At the c'onclusion of.the written examination the facility staff reviewed the examination and ansser key. lecility comments were discussed with the facility staff and appropriate revisions to the master examination key were made byjthe examiners. The facility staff provided additional written comments in a letter from H. E. Morgan to J. B. Martin dated August 9, 1985.

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The lead examiner subsequently reminded facility management in both Operation and Training that written comments were due five working days after the written examination,.that delays in submitting the comments could result in delaying grading the examination or not considering the comments, and that many of the written comments appeared to have been

- dealt with during the post-exam review or were editorial in nature. The facility staff amended the August 9,1985 letter with another letter dated August 16, 1985. This second letter contained the same comments, but subdivided them into three categories:

(a) resolved at post-exam review meeting,'(b) for reference (for future exam questions /no action requested), and (c) additional comments.

The facility written comments (as presented in the August 9, 1985 letter)

are addressed in the enclosed attachment (1). Appropriate revisions to the master examination key were made by the examiners prior to grading the candidates' responses.

During the grading of the written examinations the examiners observed that most of the candidates' low marks were associated with heat transfer, fluid flow, and reactor theory (sections I and 5) as well as procedures and technical specifications (sections 4 and 8).

In general the candidates did very well on all the other sections of the examinations.

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Operating Examinations:

Simulator exams and facility walkthroughs were conducted July 31 through August 14, 1985.

No particular generic weaknesses were identified by the examiners during the course of the examinations.

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Exit Meeting:

On August 14, 1985, the NRC representatives met with licensee personnel.

Those individual candidates who clearly passed the aperating exam were

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identified. The NRC representatives discussed the overall performance of the candidates.

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RESOLUTIONS

4 1.0 -Question 1.02" '

t Facility Comments:

"The question' assumes that the.. reactor is critical below the point of adding heat.. The question'should. have given an initial reactor power

' level significantly below the point of adding heat, such as 5 x 10-4%

power.

This comment is provided as a suggestion for rewording to ensure clarity if used on future exams."

Resolution: This point was discussed with facility personnel during the review of the answer key following administration of the examination.

The examiner pointed out that the question stated a " stable reactor period is reached" and that a stable reactor period could not occur if heat was being added.

No' change to the answer key will be made.

2.0 Question 1.07 Facility Comments:

"Part (a):

SONGS does not have PORV's. Also if steam generator pressure was reduced by 50% a main steam isolation signal would occur, isolating the steam generators by closing all tha valves associated with the steam generator."

-Resolution: This point was discussed with facility personnel during the review of the answer key following administration of the examination.

The examiner pointed out that the question referenced a simplified steam generator system depicted on a figure for the candidates' reference. No change to the answer key willLbe made.

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Facility Comments

"Part (b):

Part (b) answer assumes the examinee will answer this

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part.to the answer the examinee supplied in Part'(a)."

Resolution: Comment accepted. No change to the answer key will be made, t

however candidates response to (a) will be taken as starting point for-grading (b) whether.(a) is correct or not.

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3.0 Question 1.08c and 5.01 Facility Comments:

"The assumptions for this question are not clear.

It is therefore

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i requested that the: answer key reflect the vagueness of the question in l

the following manner:

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Allow the problem to be solved assuming:

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constant Tavg, or 2)

a constant exit temperature of 610*F, or 3)

a constant core inlet temperature of 550*F."

Resolution: This point was discussed with facility personnel during the review of the answer key following administration of the examination.

The examiner explained that the answer would be graded based on the candidates stated assumptions.

4.0 Question 2.03a

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Facility Comments:

"The three' listed answers are:

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MSIS

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EFAS'1 2.

EFAS 2 EFAS I and EFAS 2 could easily be listed as one ESFAS Trip. The examinee

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should be allowed to list the steam generator low delta P as the third trip. This is correct and should be part of the answer key."

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Resolution: This point was discussed with facility personnel during the review of the answer key following administration of the examination.

The examiner stated at that time that the candidates had been instructed to consider each steam generator separately for the answer to the question. The examiner had also agreed to change the key to include the low delta-pressure trip. Additional review of the reference material indicates that the low steam generator delta pressure trip is part of the Reactor Protection System. Although the question was for parameters and signals associated with the Engineered Safety Features Actuation System.

5.0 Question 2.08a Facility Comments:

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"The examinee may state only the equalizing function of the switch. The switch's main function is to initiate an equalizaing charge. The switch is normally positioned to zero and thus the charger puts on a float charge. The operator may assume this and hence not put it down as part of his answer. The fact that 2 functions were not called for will only add further to the probability of this occurring.

The examiner commented that the answer key would be changed to delte the need to state the float charge function of the switch."

Resolution: Comment accepted.

6.0 Question 3.01c Facility Comments:

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"The question asks which signal "AMI or AWP" is generated by PPS.

Correct answer is neither - PPS generates CWP - not AMI or AWP.

Referenced answer source supports this statement."

Resolution: Comment accepted, key will be changed.

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7.0 Question 3.02 Facility Comments:

- Reactor power will return 'to approximately its original value vs. exactly its original value as shown.

Mr. Morrill agreed to change the answer key to reflect this possibility.

The question used the term " bank is shimmed in".

This phrase is non-standard terminology and would be confusing to the examinee. That is, he would be unable to determine whether.the rods would have been inserted or withdrawn.

Figure 3.1 " Rod Position" is contradictory to the question. Therefore, the answer key fi ures should be inverted.

F It is also recommen ted that, due to the vagueness of the term

" temperature fuel", the answer key permit the use of fuel centerline,

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pellet surface or a rerage fuel temperature.

Resolution: This qt estion has been dropped from the examination, due to conflicts between question statement and verbal directions provided to candidates during the examination.

8.0 Question 3.06a Facility Comments:

"A reactor tripped override (RTO) signal will cause the main feed pumps to ramp to minimum speed vs. a 5% flow demand speed as stated in the answer key. The answer key should be changed to reflect this."

Resolution: The Training System Descriptions, "Feedwater Regulating System" page 32' states:

" Reactor trip override (RTO) occurs when there is a reactor trip. When this' occurs the Feedwater' Control System (FWCS) closes the feedwater

regulating valve.(FWRV) and positions the feedwater regulating valve bypass (FWRV bypass) valve and the main feedwater pump (MFP) speed set point control to the 5% flow positions."

During a.re-review of the reference material the examiners were unable to find justification 'for the requested change.

Since the facility did not provide any additional reference for the comment no change to the answer key will be made.

Question 3.06b

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Facility Comment

" Addition of the term coolant system following reactor is also an acceptable answer."

Resolution:

Comment accepted.

9.0 Question 3.07c and Question 6.12(b)

Facility Comments:

"Another answer is:

At > 2 x 10 % power, neutron pulses pile up and produce an AC signal which fluctuates around a mean DC signal. The square of this fluctuation is proportional to power.

(Gammas have a smaller fluctuation about the mean and, since this fluctuation is squared, the gamma-to neutron current ratio is even larger, resulting in ef fective gamma discrimination.)"

Resolution:

Comment accepted.

10.0 Question 3.08b Facility Comments:

" Unit 3 excore startup channels are BF tubes; Unit 2 excore startup

channels are gamma-metrics fission chambers."

Resolution:

Comment accepted.

Question 3.08 d Facility Comments:

The gamma contribution to the DC current output of the UIC detector is negligible - essentially, only neutrons are detected.

Resolution:

Comment accepted.

11.0 Question 4.03 Facility Comments

"The question is too specific with regard to the overall procedure.

NUREG 1021, ES 202, Section B, 4, suggests that needing to know verbatim normal and administrative procedures is unnecessary."

Resolution: The question does not require verbatim knowledge of the procedures. The question requests four of five general items which an oncoming operator is required to discuss with an off going operator.

Since an operator must do this each time he/or she comes on shif t and that the test requested " general" knowledge as opposed to verbatim procedural material, no change to the answer key will be made.

12.0 Question 4.04

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Facility Comments

"This question is confusing. The answer key is accurate for the provided reference. However, S023-8-15 (Radioactive Gas Releases) would lead to a different answer than the one in the examination answer key.

In S023-8-15, the SRO Operation Supervisor must approve the initial alignment for the waste gas release and the Shif t Supervisor must approve the actual start of the release. Our recommendation is that the original answer as written or the answer as derived from S023-8-15 be acceptable on the a) part. As for part b), the same should apply. From an operations viewpoint, the Common and Unit Control Operators should also be aware of the release."

Resolution: Comment accepted.

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13.0 Question 4.07

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Facility Comt nts

"We believe this question to be ambiguous.

Part c) could be construed to include the time after the answer'in b).

We recommend the answer key for part c) accept the original answer or "1700 25 July"."

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Resolution: Comment accepted.

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I 14.0 Question 5.04 Facility Comments This' question is.potentially misleading. The question should have clearly stated that the_ example, reactor used in the question is not the licensee's reactor and that this is a theoretical question.

If the

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purpose of the question is to probe the candidates' understanding of the heat rate vs. temperature difference during boiling aloag an electrically heated platinum wire (which is the standard example used to describe the various heat transfer regimes), then the question should be stated as such.

The question implies that normal PWR operations at 100% power are at DNB.

This is not true.

i If a large percentage of the candidates do not answer this question correctly (as shown in the answer key), the question should be reevaluated and not counted.

The reference stated for this question's answer does not directly relate this curve to reactor operations. The reference relates this curve to the heat input vs. temperature difference required to transfer heat if critical heat flux is exceeded.

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"The answer for Part (a) is incorrect and the proper answer should read:

" Single Phase Convection" or "Non-Boiling Convection"."

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Resolution: The introduction provides the data and assumptions necessary to answer the' question. A review of the reference material did not identify.any error in the question'.or answer. However, due to the possible confusion single phase convection or non-boiling convection will be accepted as correct in part (a).

15.0 Question 5.05

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Facility Comments Part (b): There is no dif ference in SDM before or af ter a Rod withdrawal.

The only change which occurs is in the % shutdown. The answer

key must be changed to reflect this or the question should be j

deleted.

Resolution: Comment accepted. Question 5.05(B) has been deleted from

the examination.

16.0 Question 6.02b

Facility Comments

"The question does not indicate the need to include setpoints as part of the answer. The answer key should take into account a non-numerical type answer (e.g., a large rapid change in steam flow versus 15% over a 5 to

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10 second period)."

Resolution: Comment accepted.

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17.0 Question 6.03a & B Facility Comments

"The intent of this question is not clearly' stated, and alternate answers are possible. This question can be answered in three ways:

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Raise feedwater flow to match steam flow, or 2)

Reduce power by control rods or boron, or 3)

Reduce steam flow by manual control of steam bypass control" Resolution: Comment accepted.

18.0 Question 6.04b

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Facility Comment

" Shutdown cooling flowrate is not automatically maintained. Shutdown cooling heat exchanger bypass flow is manually adjusted by the operator.

The answer key should be changed to reflect this fact. The answer's

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reference did not' include the design change package that mide this i

change."

Resolution: Comment accepted. However, it is noted that the reference material provided by the facility did not include the design change.

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Question 6.04c Facility Comment

"The recirculation line bypasses (HV9353 and HV9358) are more commonly known as the warmup valves. The function of the ' warmup' valves are to warmup the Shutdown Cooling System and to equalize boron concentration between the RCS and Shutdown Cooling System. The examinee will probably state that these warmup valves are used when transferring to shutdown cooling operation. -These.are correct responses to the question and the answer-key should be changed accordingly. The examiner responded favorably to the terminology difference and stated he would accept that as part of answer. However, the difference in the wording as to the function of these valves is unresolved.,We recommend the answer key be changed as stated above.!'

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Resolution: Comment accepted.

19.0 Question 6.09b Facility Comment-

"The answer key implies that the stated pressurizer pressure transmitters are IE.

However, the study guide that was referenced states that the pressurizer pressure transmitters as well as the recorders are non-1E.

Therefore, the question is invalid as the signal could have been generated from either the transmitter or the recorder.

In either case there would not be an interconnection between a IE and a non-1E System.

The question should be deleted."

Resolution: Comment accepted. Question 6.09b has been deleted.

20.0 Question 7.02 Facility Comment The responsibility to determine if additional dosimety is reqaired is the responsibility of Health Physics. This question is inappropriate because it is beyond the scope of SRO training.

Resolution: Basic ALARA considerations are a SRO responsibility, therefore some knowledge in this area is required. The key will be modified to reflect required SRO knowledge.

21.0 Question 7.04 Facility Comment

"The question asked for restricted area exposure limits, but the answer is for a radiation area per 10 CFR 20.

The correct answer is 0.25 mr/hr or exceeding 0.6 mr/hr average. See the attached sheet."

Resolution: Comment accepted, question deleted.

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22.0 Question 7.05a Facility Comment

" Additional acceptable starting prerequisites:

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No reverse rotation is indicated (zero speed light is illuminated)

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Pump seals are properly staged 3)

CCW supply temperature to the non-critical loop is less than 95*F

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No abnormal alarms for the RCP to be started" Resoluticn: Comment accepted.

N 23.0 Question 7.10 s-Facility Comment

"CPC's do not input to COLSS to determine LilR, peaking factors, power tilt or DNBR margin. Therefore, this is an invalid question and should be deleted or allowances must be made in grading the candidates' answers due to tnis misleading statement."

Resolution: The facility comment is accepted. However, question 7.10 (a), (b), and (c) do not specifically address the relationship between COLSS and the CPC's.

The key to part (c) will be modified to conform to

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Technical Specification bases 3/4.2.1 and 2.1.1.

Candidates answers will be judged in view of the possible misleading question statement.

24.0 Question 7.12

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Facility Comment

"The CPS's and RPS do not reqeive inputs from COLSS. Therefore, the questiiu is invalid and should be deleted or allowances made in grading the ca'adidates' answers due to this misleading statement."

Resolut'.on: Facility comment accepted. Ilowever, question 7.12, a, b and s

c do nyg,ad. dress the relationship between COLSS and the CPC's.

Each candid. te's answer will be judged in view of the possible misleading questio}n statement.

25.0 Question 8.05(c)

Facility Comment'

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"This question is too complex for the required answer. The cited reference refers to an attachmJnt of the main procedure and would require in-depth memorization in order to correctly " identify the answer. The

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answer key did not have the correct answer from the table. The answer should be ISS, 2SRO, 3RO, 3AO, ISTA 9)ce ISS, 2SRO, 2RO, 2AO, ISTA. We recommend deleting this part of the qd ytion. Again, we believe this question to be too complex in nature to be able to respond from memory."

s Resolution: The key util be modified to reflect facility comment.

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26.0 Question 8.08(c)

Facility Comment-

"This question is vague. There are many different' members of.the facility staff (other than operators) who may be involved in the initial mitigating effects (i.e., H.P. & Chemistry personnel). As an Emergency

~ Coordinator (SRO), he would be expected to know what all thesa people are doing. The response which was expected by the answer key only wanted what the " Operations" people would do.

This part of this question should be deleted for these reasons."

Resolution: Comment accepted.

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U. 5. NUCLEAR REGULATORY COMMISSION

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SENIOR REACTOR OPERATOR EXAMINATION

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Facility: San Onofre Nuclear Generating Station Units 2 & 3 Reactor Type:

Combustion Eng.

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Date Administered:

July 30, 1985 Examiner:

D. J. Willett Candidate:

i INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers.

Write answers on one side oniv.

Staple question sheet on top of the answer sheet.

Points for each question are indicated in parenthesis 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.

Category

% of Candidate's

% of Value Total Scor:

Cat. Value Catecory

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23.5 25. Z'T 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamic s N.D 15, T)

6. Plant Systens Design, C3ntrol and Instrumentation 24.0 75.&l 7. Procedures - Normal, Abnornal, Emergency and Radiological Control 2l.T N'

8. Administrative Procedures, Conditions, and Limitations D'

/00 TOTAL 5 Final Grade

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all work done on this examination is my own; I have neither given nor receiveo 61C.

s Canc1cate's Signature i

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SONGS SRO EXAMINATION JULY 30, 1985 SECTION 5 TOPIC KEY

5.01 THERMODYNAMICS 4.00 5.02 XENON 4.00 5.03 RX KINETICS 1.00 5.04 HEAT TRANSFER 3.00 5.05 RX KINETICS 1.50 5.06 RODWORTH 2.00 5.07 S/D MARGIN 1.00 5.08 FLUID FLOW 3.00 5.09 CHEMISTRY 4.00 TOTAL 23.50 GRADE SECTION 5 25.27 SECTION 6

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6.01 F/W CONTROLS 2.50 6.02 STEAM BYPASS 1.50 6.03 OPERATOR ACTIONS 2.00 6.04 S/D COOLING 2.50

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6.05 CVCS 2.00 6.06 COMP. COOLING WATER 2.00 6.07 S/W COOLING 1.50 6.08 CST 3.00 6.09 PZR & SBC 1.00 6.10 PZR 2.00 6.11 COMP COOLING WATER 2.00 6.12 NUCLEAR INST 2.00 TOTAL 24.00 GRADE SECTION 6 25.81 l

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SECTION 7 7.01 ECW 4.00 7.02 DOSIMETRY 2.00 7.03 RADIATION LIMITS 2.00 7.04 AREA LIMITS

.00 7.05 RCP OPERATION 2.00 7.06 INADEQUATE COOLING 3.00 7.07 REFUEL STAFFING 1.00

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7.08 S/G TUBE RUPTURE 2.00 7.09 LIQ EFF RELEASE 3.00 7.10 CPC & COLSS

,'2.00 7.11 RCS OVERPRESS 1.50 7.12 RPS & CPC 1.50 TOTAL 24.00

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GRADE SECTION 7 25.81

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SECTION 8

8.01 FUEL LOADING 2.00

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8.02 TEMP CHANGES 1.50 8.03 IMED NOTIFICATION

.00 8.04 SS RESPONSIBILITIES 2.00 8.05 WORK PERIODS 3.00 8.06 FIRE SUPPRESSION 3.00 8.07 COOLANT ACTIVITY 3.00 8.08 EMERGENCY RESPONSE 2.00 8.09 INDEPENDENT VERIF 2.00 8.10 T.S. SURV & OPS 3.00 TOTAL 21.50 GRADE SECTION 8 23.12 DVERALL TOTAL S3.00 OVERALL GRADE 100.00

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Pwr = v an f

TYL = 1.3/u P = P,10 "'III

HVL = -0.693/u P = P,e /T t

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SUR = 26.06/f SCR = 5/(1 - K,ff)

T: 3 TT DI CR, = S/(1 - K,ff,)

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T = (1*/s) + [(s - a)/ o]

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

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

T = (s - e)/(as)

SDM = (1 - K,ff)/K,ff 1* = 10-5 seconds p = (K,ff-1)/K,ff = aK,ff/K,ff T = 0.1 seconds ~I p = [(1*/(T K,ff)] +,[T,ff (1 + $T)]

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jj=Id2,2 2 Id P = (r4V)/(3 x 1010)

Id gd jj

2 R/hr = (0.5 CE)/d (meters)

I = eN R/hr = 6 CE/d2 (f,,t)

Water Parameters Miscellaneous Conversions _

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I curie = 3.7 x 10 dps 1 ga;. = 3.78 liters 1 kg = 2.21 los

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3 Stu/nr I hp = 2.54 x 10 1 ft* = 7.48 dal.

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Density = 52.4'1bs/ft3 1 aw = 3.41 x 108 Stu/hr

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Density = 1 ge/cm Heat of vaporization = 970 Stu/los,

F = 9/5'C + 32

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Heat of fusion = 144 Si.u/les

C = 5/9 (*F-32)

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1 Ata = 14.7 psi = 29.9 in 1 STU = 778 ft-lbf i ft. H o = c.4335 lbf/in.,. Hg.

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Properties of Saturated Steam cnd Satur.t d Wct:r-endud:d

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Heat of Latent Hut Total Heat feswcafic Volume

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250.0 325.3 400.97 376.1

~s25.0 1201.3 0.01865 8 u3i7 255 0 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80n02

.

260.0 245.3 404.44 379.9 821.6 1201.5 091870 1.77416 265.0 250.3 406.13 381.7 820.0 1201.7 0.01873 1.74157 270.0 255.3 407.80 343.6 338.3 1201.9 0.01875 1.78013 l'02.1 0.01878 3.6797a 275.0 260.3 409.45

- 345.4 816.7

.

240.0 265.3 418.07

" 387.1 815.1 1202.3 0.01880 1.65049 285.0 270.3 412.67 388.9 513.6 1202.4 0 01882 3.62218

}90.0 275.3 414.25 390 6 812.0 1202.6 0 01885 1.59462 m.0 280.3 415.81 392.3 810.4 1202.7 0 01887 1.h835

.

300.0 285.3 417.35 394 0 808.9 1202.9 0 01889 8.54274 320.0 305.3 423.31 400.5 802.9 1203.4 0 01899 3.44801 340.0 325.3 428,99 406.8 797.0 1203.8 0.01908 1.h402 360.0 345.3 434.41 412.8 791.3 1204.1 0.01917 1.28910 380.0 365.3 439.61 418.6 785.8 1204.4 0.01925 1.22177 400.0 345.3 444.60 424.2 780.4 1204 6 0.01934 1.36095 420.0 405.3 449.40 429.6 775.2 1204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535 460.0 445.3 454.50 439.8 765.0 1204 8 0.01959 1.00921 440.0

5.3

2.82 444.7 760.0 3204 8 0.01 % 7 0.96677 500.0 445.3

7.03 449.5 755.1 1204.7 0.01975 0.92762 520.0 505.3 471.07 454.2 750 4 1204.5 0.01982 0.89137 540 0 525.3 475.01 458.7 745.7 1204 4 0 01990 0.85771 h0.0 545.3 478.84

3.1 741.0 1204.2 0.01998 0.82637 540.0 M5.3 482.57

7.5 736.5 1203.9 0.0200e 0.79712 600 0 545.3 446.20 471.7 732.0 1203.7 0.02013 0.76975 620.0 605.3 489.74 475.8 727.5 1203.4 0.02021 0.74408 640 0 625.3 493.19 479.9 723.1 1203.0 0.02028 0.71995

%00 645.3 4 %.57 483.9 718.8 1202.7 0.02036 0 69724 660.0 M5.3 499 86 447.8 714.5 1202.3 0.02043 0.67581 700.0 6&5.3 503.08 491.6 780.2 1201.8 0.02050 0 65550 720.0 705.3 506.23 495.4 706 0 1201.4 0.02058 0 t h30 740 0 725.3 509.32 499.1 701.9 1200.9 0.02005 0.61822 760 0 745.3 512.34 502.7 697.7 1200.4 0.02072 0.60097 780 0 765.3 515.30 506.3 693.6 1199.9 0.01080 0.58457 800 0 785.3 518.21 509.8 689.6 1899.4 0.02087 0.%B96 820.0 805.3 511.06 533.3 685.5 1198.8 0.02044 0.55408 540.0 825.3 523.86 516.7 681.5 1198.2 0.02101 0.53988 860.0 845.3f 526.60 520.1 677.6 1197.7 0.02109 0.52631 880 0 MS.3 f 529.30 523.4 673.6 1197.0 0.02116 0 $1333 900.0 885.3!

533.95 526.7 M 9.7 1196 4 0 02:23 0.50091 920.0 905.3 534.56 530.0

%5.8 1195.7 0.02130 0 48401

.

940.0 9 25:3 537.13 533.2 661.9 1195.1 0.02:37 0.47759 960.0 945.3 539 65 536.3 658.0 1194 4 0.02145 0.46662 980.0

%5.3 542.14 539.5 654.2 1193.7 0.02152 0 45604 8000.0 965.3 544.58 542.6 650.4 1892.9 0.02159 0.44540 1050.0 1035.3 550.53 550.1 640.9 1191.0 0.02177 0.42224 1100.0 1085.3 556.28 557.5 631.5 1189.1 0.02195 0.40058 1150.0 1835.3 MI.82 h4.8 622.2 1187.0 0.02214 0.38073 1200.0 1185.3 M7.19 578.9 613.0 1164 8 0.02232 0.36245 1250 0 1235.3 572.38 578.8 603.8 1182.6 0.02250 0.34550 1300.0 1245.3 577.42 585.6 594.6 1180.2 0.02269 0.32991 1350 0 1335J 582.32 592.2 585.6 1177.8 0.02288 0.31536 1400 0 1385.3 587.07 598.8 567.5 1175.3 0.0230; 0.30176 1450.0 1435.3 591.70 605.3 M7.6 1172.9 0.02327 0.28900

._

1500 0 1485.3 5 %.20 611.7 558.4 1170.1 0.02346 0.27714 1600.0 1585.3 404.87 624.2 540.3 1164.5 0.02387 0.25545 1700.0 1685.3 613.13 6%.5 522.2 1158.6 0.02428 0.2 h07.

1800 0 1785.3 621.02 648.5 503.8 1152.3 0.02472 0.21861 3900.0 18&S.3 628.56

% 0.4 485.2 1145.6 0.62517 0.10278 2000.0 1985.3 635.80 672.1 4%.2 II38.3 0.02h5 0.18831 2100.0 2085.3 642.7 6 u3.1 4u.7 1130.5 0.02615 0.17501 2200 0 2185.3 649.45 695.5 426.7 1122.2 0.02669 0.16272 2300.0 2255.3 M5.89 707.2 406.0 1813.2 0.02727 0.15133 24004 23&5.3 M2.Il 719.0 344.8 1103.7 0.02790 0.1407F *

2500 0 2485.3 M8.Il 731.7 361.6 1093.3 0.02&$9 0.13006 2600.0 2545.3 67J.91 744.5 337.6 1082.0 0.02938 0.12110 2700.0 26&5.3 679.53 757.3 312.3 1069.7 0.03029 0.11194 2800.0 2785.3 664 %

770.7 285.1 1055.8 0.03134 0.10305 2900.0 2885.3 690.22 785.1 254.7 1039.8 0.03262 0.09420 J000.0 2985.3 695.33 301.8 218.4 1020.3 0.03428 0 08500 3100.0 3085.3 700.28 824.0 169.3 993.3 0.0 % 81 0.07452 3200.0 3165.3 705.08 375.5 M.I 933.6 0.04472 0.05 % 3

'

-.

3208.2 3193.5 705.47 906.0 0.0 906.0 0.05078 0.05078

,

.

,

.. _,

.

5.01 ( 4.0 )

For the following questions refer to the attached steam table.

80 million Lbm / Hr. of reactor coolant at 2100 psi enters one of two steam generators at 610 deg.

F.

and leaves at 550 deg.

F.

Feedwater at 450 deg.

F.

enters the secondary side of the steam generator and leaves as saturated steam at 550 psi.

The S/G heat transfer process is represented by the following equations :

Q = m cp 6T

.

O "m Ah Where cp 1.1 BTU

=

,

_______

Q=UA 6T Lbm - F.

-

sa)

What is the total thermal power

,

in BTU / Hr of the reactor 7

,

(1.0)

(b)

What is the total steam flow rate from each steam generator ?

(1.0)

(c)

What would be the steam pressure if the total heat flux were reduced to 1/4 of it's initial value ( assume the same Tave.)?

(1.0)

(d)

What Tave would be required to maintain a

steam pressure of 550 psi at the reduced (1/4)

steaming rate 7 (1.0)

NSWER :

(a) b = $ cc 6F

= J $ co ( Th - Tc )

C e

' ' = ' ( 30

)

( 1.1 1 ( c10 F,

-- '. 5 0 F.

)

O = 10.56 T

10 BTU / Hr e

e

,

(b) '

=a Ah 1/2 U S.20

OTU / Hr a

=

b=m (1204 - (cp) (t))

(from steam tables)

b.23

9TU / Hr / (1204 - 1.1 450F) -7. 45

Lbm m =

,

b=Un (c)

Ai Un( Tave T ta >

Tave

=

Th + Tc

=

.

_ _ _ _ _. _ _

1 = UA initial ( 580 - 470 )

- - -.

_ _. -.. - -

-_-.-.-.-_

- _

Un final ( 580 -- Tstm )

4tm b54.

F.

deg.

=

ressure Steam - 1070 pai at S tid.

F.

d ea. ( f r ecn steam table)

(d) Tavc f i n o.1 478 * ( 500 -- 478 ) /

<!

= '50 7 l-deg.

=

.

' TEriE f !CE :

0.P.

t.t tc l ear t'o c n. Ecct.

E, Pg.

2-1 ~5 9 *- 141 L 2-155 i

'

.

5.02 ( 4.0 )

Refer to figure 5-2, Axial Xenon Concentration and Neutron Flux Profiles.

Graph

"A" depicts initial equilibrium power and Xenon profiles with the control rod withdrawn.

Assuming a partial control rod insertion at time zero, sketch the neutrcn flux and xenon concentration profiles at about 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the partial rod insertion and again at about 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after partial insertion on graphs

"B"

&

"C".

Your drawing should

,

indicate magnitudes of power and Xenon relative to initial equilibrium values.

ANSWER :

ATTACHED

_

,

CEFEHEinCE :

I-c.'.

P! !Y S I CS, Vol.-II, Chapt.

3, Sect.

D, Pa. 5-50 i

l t

b

.

R

k

'

\\

. -

N TOP g

\\

)

P0k'ER or @ FLUX CRAPH "A"

"

XENON I

Al

/

r l

l i

BOTTOM j

,

RELATIVE AXIAL MAGNITUDE EQUILIBRIUM PRIOR TO ROD INSERTION

^

.

R l

O I

TOP l

t CRAPH "B"

,

'

i i

BOTTOM

__

lI RELATIVE AXIAL MAGNITUDE ABOUT 6 HOURS AFTER ROD INSERTION N

TOP R

O g

D

,

t I

GRAPH "C" i

i e

i

.

BOTTOM l

RELATIVE AXIAL MAGNITUDE ABOUT 24 HOURS AFTER ROD INSERTION FIGURE 5-2

-

.

i

N R

O k

!

-

\\

n_.

TOP

.

g

\\

)

POWERor@ FLUX GRAPH "A"

XENON I

Al

/

r l

/lI i

I BOTTOM RELATIVE AXIAL MAGNITUDE EQUILIBRIUM PRIOR TO ROD INSERTION

@

I

,,

R

'ps O

6,l TOP

/g Xenon (1.0)

Il o

GRAPH "B"

~

l j

--Power (1.0)

/

/.

l

BOTTOM

,-

,

RELATIVE AXIAL MAGNITUDE ABOUT 6 HOURS AFTER ROD INSERTION N

e i

TOP R

T l

',

l

'

-

p s

g I

.

j

\\

'

- - -

--Power (1.0)

_,

\\

k l l

GRAPH "C" I

-

',3 -

--Xenon (1.0)

/l

-

.-

-

BOTTOM

,,,,,

,

l RELATIVE AXIAL MAGNITUDE

!

l ABOUT 24 HOURS AFTER ROD INSERTION FIGURE 5-2 i

,

-

--

-

.

-.

.

_

.

5.03 ( 1.0 )

E::pl ai n why the stable negative startup rate ( SUR

)

of the reactor cannot be shorter than Apro::. - 1/3 DPM.

ANSWER :

The shortest negative startup rato in-determined by the decay rate of delayed neutron crecursors. (The longest delayed neutron nalf-life = 5 5...a secondc.

Thi s yields a negative period of

- 80 seconds 1. 44 ': DT SUR = 26.06/~80 = -1/3 DPM )

=

.

3

}

l i

i l

FEFEPlirEE :

,

!

G..'.

PHYSICS, vol.-II, Cnapt.

5, Sect.

C, Pg. 5-39 l'

. _ _.

,

.

-

. _.-

. _ _ -.. _.

_.,

--._

.

~

.

5.04 ( 3.0 )

Refer to figure 5 - 1, A plot of Heat Flux VS Temperature Difference between Wall and Bulk Fluid, for two different system pressures (

P,

<

Pg).

At 5 % reactor power the heat flux is 1 x 10 BTU /Hr - Ft. At 100 reactor power the heat flux is 3 x 105 BTU /Hr - Ft.

,

(a)

At pressure P

and

%

reactor power what type of heat trknsfer is occuring between the wall and bulk fluid ?

(0.5)

(b)

At pressure P,

and

%

reactor power what type of heat transfer is occuring between the wall and bulk fluid ?

(0.5)

(c)

Which pressure will yi. eld a lower fuel temperature at 5 % reactor power 7 (0.5)

(d)

At 100 % power which pressure will yeild a greater fuel temperature ?

(0.5)

(e)

What type of heat transfer is occuring between the wall and bulk fluid at 100 % reactor power and pressure P

(0.5)

(f)

At 5 % reacter power, how will decreasing pressure affect the bulk fluid temperature ?

(0.5)

ANSWER :

(a)

Natural convection, sincie gnase con vec ti on, or non -bai1i ng convection.

(b)

Nucles.te boiling (c)

cseure 1 (d)

c'renrure 1 (e)

R di r nt haat trannfor steam blanket >

'

(fi the bulP Eluid temperature

'enains constant and in indepeod;nt cf prescure.

I j

HEFERENCE :

G.

P.

uclear F.c h.

cact, g,

pq, 2.-144 h 2-151 % O 159 % 2-164 a

_ _ -,

_

,

._

_.-

-

.

t

'

.

P (P g

y

10

3 x 10

--

r" C 'N

,

I

'

', _

'

,

'

i

-

%

'

v

/

s u

sc

=

Q I

a I

w P

P

/

E

2 m

),/

m a

l, D

<

x D

!

d

0

?

4

10 10~

10 Temperature Difference Between Wall and Bulk Fluid ( F FIGURE 5 - 1

,

%

_ --_

_,

- - - _ _ _ _, -,. - - - - -

.

-.,

--.._,_--

, _

,, -.. _ -

r. -

.

5.05 ( 1.5 )

Consider a subcritical reactor with a source strength of 20 cps.

After a rod withdrawl, the count rate levels off at 100 cps.After further withdrawl, the final count rate levels of at 200 cps.

Shgw calgulatiggs.

(a)

What is the final value of K eff ?

(1.5)

9.t 5NER :

(a)

ECP 5 /

t.

1 - Keff )

f equation (0.5)

)

(From formula sheet)

=

Keff - 1 -

S

- --

SCR

ef f 1 ~ 10 / 200 0.9 ( correct v a l ' u? incertion (0.5) )

==

=

t'.c f f final = 0.9 ( correct answer ( 0.5) )

REFERENCE :

G.P.

PHYSICO. Val -II, Chrot.

5, Sect. r4, Pg. 5-Da S

-5.06 ( 2.0 )

Differential rod worth is affected by various factors.

How will each of the following factors, if considered separately, affect rod worth ?

(a)

Radial or axial rod position (0.5)

(b)

The proximity of.other rods (0.5)

(c)

Temperature (0.5)

(d)

Baron concentration (0.5)

ANS!VER :

(c) Rod worth is a function of occiticn relativn to tne flux profile. A red pocitionec in the area of highest flux has a higher worth than the came rod in a area of lower flun.( O.5 )

(b) The prescnce of other red; in pro::imi ty to escn rtner results in increased competiticn and shiciding. The net effect is to reduce rod -zo r t h. ( O.5 )

(c) Increascd temperature increases rad warth cue tc a neutrons increased.mi gr at i on.tength, increacing the or cbabili ty that a neutran uill untar a od.

0.5

<

(d) The higher the coran concentration t h e l ower the rod worth

,

because of the ccmpetiticn between rodc and boren for neutrons. ' O.5 )

i riEFEREr;ce.

I

!

b.P.

PHYO!CS,

'/01. - I I, CHAPT. 4., Sect.

D, Pg. 4--108 l

l O

I

.

'

.

5.07 ( l.0 )

-

A startup is being performed 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months after a reactor trip using a

calculated estimated critical position for the time that the startup commenced.

How will each of the following events or conditions, occuring immediately prior to the startup, affect the actual critical rod position compared to the estimated critical position ? (HIGHER, LOWER, NO SIGNIFICENT DIFFERENCE )

(a)

A steam generator's level is increased significently, with feedwater at 450 deg.

F.

(0.25)

(b)

A reactor coolant pump is stopped.

(0.25)

c)

The startup is delayed for appro::imately two ( 2)

hours.

(0.25)

(d)

The steam dump pressure setpoint is increased.

(0.25)

ANSWER :

(a) Lcwor ib) No ni ani f i cent d3+ference (c) Lcwer (d) Higher

,

REFERENCE :

San Onofree Technical 00ccifications Danin 3/4.1.1.1

?..2

i

.

5.09 ( 3.0 )

Figure 5 - 3 shows a closed loop system with three (3)

identical parallel centrifical pumps. Initially all pumps are running.

+** EXPLAIN YOUR ANSWERS ***

(a)

If pump A

is stopped and isolated by closing

"

valve

"A-3",

what happens to Tank delta pressure ?

(1.5)

(b)

What happens to Loop B

flow when pump

"A"

"

is stopped and isolated by closing valve "A-3 "

?

(1.5)

.-:N S WER :

'

in) DECREASES ( 0.5 )

'otal col u net r i c i1ou is the cum uf the fIcw from all pumpn.

( O.5

The t an t.

differential or;,ssure is a eunction cf total volumetric flow squared ( Delta P =K V A)

As total flow is reduced by running on1N two pumen delta P 1 reduced. t 0.5 )

(b) INCDEASES (

5 As the ' a n !/ tiel t a d e c e r: a s or,. the dalta F in pumo

.

"

B line must increase to account for the pump

"

discharge pressure. ( O5 )

In order to increase the pump line delta P, the volumetric ficW rata must increase. ( 0.5 )

REFERENCE :

G.P.

ri!ELEAR TFCH.

'.'o 1 -- I I I, ::ection H, Pg. 2-237 and Sect.

D, M.

2-117 thru 2-131.

d

'

,

.

,~

.

A-2 A-3 n

'

A-1

_ _..

,

Ps!-

-M--

y A

><

W s=eny i

B-2 B-3 l

B-1 M

w D,

M

$

TANK M

C-2 C-3 ( esa ter-L_---

,-_.---

C-1 N

w

' - - - - -

,

l C

t_

_ _- 3 Hed N G

_

[

SqFAY

><

}

F-2 S T/A M

-

,

r FILTER,

220.0 i

s

/

200.0

'

f I

I 180.0 i

'

/

r 160.0 I

/

f f

140.0 l

x

,

M N-

/

'

e

's-r l

$

120.0 N

/

,

v

., t

/

%

t 100.0

'

X to

)

%

c

/

' s Y

/

'N

~

i 80.0

/

N

,

/

s

/

%

l 60.0 l

N f

N,

'

i

.

v

'

.,

%

40.0

-

'

f

>

-

_

,

20.0

-

/

a'

0.0

.

,

0.0 40.0 80.0 120.0 160.0 200.0 240.0 280.0

,

System Capacit.y (GPM)

-

Figure 5.- 3

. - _ _

....

.

.--

-

._.

.

-.

..

.

5.09 ( 4.0 )

-

~

(a)

Why is there a limit placed on dissolved oxygen in the primary coolant ?

(0,5)

(b)

Identify the limiting values and temperature associated with the dissolved oxygen concentration of the reactor coolant system.

(1.5)

(c)

What are two (2) sources of of dissolved oxygen in the primary coolant.

(0.5)

(d)

Explain how oxygen concentration is controlled during : shutdcWn / startup ; and normal operation.

(1.0)

(e)

Why isn't the same method of oxygen control employed during operation that is used during shutdown ?

(0.5)

a 4N3Wi k

.

(a)

Theru in a limit piecca an di ssol vec a :ygen in the primary to insure that ccerosion is minimized.

<0.5)

sb) Stead / = tate linit for di sco1 ved c+; v g an is 0.10 ppm ( Q.5 or =

-

Transient )init to or 1.00 ppm ( O.S )

The Steady State and frannient limit are not applicable with T ave s or =

to 250 F degrees.( O.5 )

(c) The sources of di csolved c::vqen are : (,ny twa O.25 ca.

)

1.

radiolytic decamposition

,

!

2.

p r i s.ar y make-up water 3.

dissolved in RCS during refueling i

l (d) During shutdcwn and startup, Hydrazine is added to the itCS to j

remove cuygen ( O.5

),

Dur2ng normal op er a ti on there ir a

Hydrogen over pressure on the VCT to cambine with any

!

oxygen in the RCS. ( 0.5 )

(e) Hydrazi n a which is used-during-anutdown, breaka down at

.

normal coeratica. temperatures. '(

O.5 )

DEFERENCE :

'

San Gnotroe tech spec'c. 3/4.4.6. & basis. Operating inst

'5023-3-2. 3 CVC3 Chemi cal addition tank operation.

!

.-

-.-

.

6.01 (2.5)

(a)

What is the purpose of the "reacto.- trip override feature" of the feedwater regulating system ?

(1.0)

(b)

How does the feature accomplish its function ?

(1.0)

(c)

How is this feature cleared '

(0.5)

ANSWER

.

t' n )

To prevent precsuriner

,oressure and level trancients which would result from overcooling reactor cool an t after a reactor trip.

(b)

On a trip, the icedwater ccntrol system conds a

zero denanu eignal to the main i ned reg, valves a r.d a inninum (5%) fl'u demand signal to the feed reg.

byoass val vtrs.

(c)

The 9TC f eature remainc in,ffect until actual m

feerwater flcw d e:a an d 12 lecs than the RTO c:nnimum (5%) demard signal.

i i

RE.CEFENCE :

Sonqn

$ 3 Training Syctem Oescription, Feedwater i

92qelating System, page G.

t l

.

we

,

.

_

_.

.

6.02 (1.5)

The steam bypass control (SBC) system has three (3)

allignment modes for operation.

(a)

Explain the rational for the " Quick-Opening Block" feature in the SBC system.

-( 1. 0 )

(b)

How is the " Quick-Opening" function of the SBC initiated ?

(0.5)

ANSWER :

(a)

The

"Ouich-Open:na 91cck" teature at the SbC system mir.imtres pressure ind temperature swings in the RCS.

(Prevente a n':e s c i s e ccoidown without cperator contral,'

(b)

The

Oul c k --Op en i n g " f unc t i on cf the GDC system is initiated i:henever,. l arge Icad decrease rate is detected (in encess o f-A L ". in 5 to

seconds)

(20% in 5 cec or 16% in i sec).

-

.

I REFERENCE :

i

'

Songc 2 & 3 Training Gyntem nne"rioti.00, rtoam Dypass Control

,ystem,Page 16 -23.

---

a

-

i

^

I I

--.

6.03 (2.0)

The plant is warmed-up and pressurized and the RCS is heating up to normal operating temperature, with the steam bypass system controlling pressure at or below 1000 psia.

(a)

Fcr the above conditions, if steam flow starts to exceed auxilliary feedwater flow, what should the operator do ?

(1.0)

(b)

During plant cooldown, at or below approximately 1

%

power with the steam bypass system controlling pressure, if steam flow rate starts to exceed auxiliary feedwater flow, what do ?

'

should the operator (1.0)

ANSWER :

'a'i Fai ca feedwater 410w to match steam flcw

.

or reduce pcNec by cr: tral ~odc

^r bcron er reduc 2 utcam flew by manual control o+

_ steam bypacs ccatrol fo)

Daer e rr z.n cul d t ak: - manual control of t h t-u t ca:n hypass alves.

REFERENCE :

Lanoc 2 L 2, Training Syctom Deccription, Eteam bypass control, r' ace 35.

m

-

.

-

6.04 (2.5)

The Shutdown Cooling System (SDCS) is designed to cool the RCS after RCS temperature drops below the point of significant steam production by closed loop exchange between the RCS and CCW system.

(a)

How does the operator control the cooldown rate ?

(1.0)

(b)

How is a

constant shutdown cooling flowrate to the core maintained ?

(1.0)

(c)

What is the function of the " Recirculation Bypass Line" (HV 9359 & HV 9353) in the SDCS

?

(0.5)

ANSWER :

(a)

The coolcown rate in controlled by manuall',

ac;uctinc the flcurate of I'C S through the EDC heat

_xchan Crc.

(b)

Shutdoun cooling h'at w: changer cypass

-- l ow

.s man u-21 l y adjucted.

(c)

The ': D.r C O s e of tha r;1 r c ul at i on S< pacs line Is-t c 1i<

t thcrmal cterzi curina stem warmua and squiline

^ r-r n

ccntr atica in piping and cre.ccnents & -g

,qtes int chutdown cooling.

.

REFERENCE :

Sonen 2 S 3, - Training Syutzm Description.

Safety Injection and Chutdaun Cooling, Page S.

.

.....,..

-

,

-

...

.

_

l j

'

d l

,

6.05 (2.0)

During normal operation, one of two letdown backpressure control valves, in the chemical and volume control system (CVCS),

is in service.

(a)

If the inservice backpressure control valve failed open, What potential adverse effects, if any, could be expected ?

(1.0)

(b)

What CVCS design considerations would limit or mitigate the consequences of the inservice backpressure control valve failing open, during normal operation ?

(1.0)

N!SWER -

(a)

Flashing of the RCE dcWnstream of the letdowm

,

ficw control v c.l ve s.

to)

The lotdown accumtilator is crevi ded to limit the the depressurinaticn

.7 n d pressure transient associated tntn a

failed open backpressure central valve.

.

I hEFEREi!:E :

'c nngs

&

3 Tr a i n i nt; Gyntem Description,. Chemical an Volume Cantrol 3ystem,Page 6.

E

.

.-

.

,

6.06 (2.0)

The component cooling water system is comprised of two critical loops and one non-critical loop.

(a)

For Unit #3, name four (4) components where the non-critical loop supplys cooling water besides the RCP motor coolers,RCP seal packages and CEDM coolers ?

(1.0)

(b)

What is the purpose of the nitregen blanket in the component cooling water surge tanks 7 (1.0)

ANSWER :

OnY t C Ltr 0.25 encn (a)

Gpant fuel pool heat c:: h an ;orn Wante ces compreuso.m Gac striperc(Loclars and c:ndensar )

Mioc. nasta evapcrator Rade uaste cancentrato return (RCR) c ei..o l e ; color Iai The nitrogen blanket i2 establi_ned to raintain a ccnctant nucticn prascure at the auct:an at the CCN s u.ro c.

J l

PEFERENCE :

Canyn 2 & ~? Trsining System Description, Component Cooling Water Syctam. Page 4 & 5.

..

-

,

.-

_ -

. - _ _.

.

__ _.- -

.._.

-__

6.07 (1.5)

The Saltwater cooling system (SWCS) is designed to operato during a

design basis accident or earthquake with an associated loss of off-site pcwer.

(a).

How is correct valve alignment assured, in the

'

(0.5)

SWCS, during a

loss - of off-site power

i (b)

What is the function of the cyclone separator within the SWC system ?

(0.5)

l (c)

What interface does the SWCS have with Nuclear

Service Water Turbine Plant Cooling Water or the

,

Service Water System ?

(0.5)

i

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6.08 (3.0)

The Condensate Storage and Transfer System design criteria is based on discharging steam to the atmosphere in mode 3,with a

l concurrent loss of off-site power.

The storage tanks are comprised of two tanks, T-120 and T-121 the smaller.

i (a)

What is the make-up source (s) for T-121 ?

(1.0)

(b)

What is the Backup source of make-up to T-120 ?

(1.0)

(c)

Should Tank T-121 empty and Tank T-120 rupture what would be the source of suction to the Auxiliary Feedwater System ?

(1.0)

BNSWER

-

(a)

The cource ci maea ao to i - i _' l is T-120.

i (b)

The backuc acur:2 cp w.ncr to T-- 120 i n the

' Tire Prot ecti on 'vstem.

ic)

The str uc t ur e currcunding T-120 ts seicmic

c a t a g er y-i.

and

an coctain tho. a ic 1 ty of the water e.cuid the tank - u p t ur '- t 2. n K.

.~h e structure can supal:

21 th r a ug:h a scianic-1 linc.

'

.

!

t

REFERENCE :

I Scngs, 2 S 3, Training dystem Deucription, Condencate Trancier

nd. Storage System, Paae 2-4.

1.

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6.09 (1.0)

The following questions pertain to the Pressuriter and its interface with the Steam Bypass System

,

A s

.3

-

s (s )

How is pressurizer pressure (f rom PT-100X & Y via

,

recorders PR-0100A

&

B and Gauge PIC-0100)

)

utilized in the steam bypass control system ?

(1.0)

'

,

A-

.

ANSWER:

~ * -

'

,

)

%,*

s.

(a)

A pressuriner pressure signal is used to bias the SBC system cetcoint so that the SBC 'nystem res. ponds more quicklev tq,dlsturbancer$go71ginatino

,

in the RCC gri, mary.

-

.

-

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=

.

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s.)

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i REFERENCE :

l Scnon 2 & 3 Training System Descriptionc,

Steam Bypass l:

Ccntrol System,Page 15.\\

l,

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-

6.10 (2.0)

There is a lag unit (aprox. 5 sec.) between the pressuri:er level controller and the letdown throttle valve controller.

(a)

What are the reason (s) for the lag unit between the pressuriner level controller and the letdown throttle valve controller ?

(1.0)

(b)

What precaution should the operator observe when manually controlling letdown ?

(1.0)

ANSWER :

(a)

The lag unit slows cown the cocea at unicn the

-

l e t d c '.c valve utrokes Rapid onening of the letdown tcntrol

.alve could

'

cause

rassure transients which the backcransure regulating valvt not ce Tale tc c or,t r ol.

n

1:ao l.j clasure of tne..:-t d own contral valvo could'

ceduce pressure encuch to oreduce flashina in tne letecwn 2ino.'resulting in a waterhammer When th n. -st:

..m cids ccllapce)

'c)

Th^

operator

'should chance the output of the

.

rav_reller clrNls. (oc-cauce the isa u' i t is before

'

the lend wn Lhrottic valvo controller.)

REFERENCE

-

Ecngs

C<

J

. Training Syntem Dectription, Proscuricor and Penscure Control System, Page 16.

.

_. -

-

..

.

%

+

,

s

.

6.11 (2.0)

i

't Interlocks are provided for CCW pump P-25 to prevent cross connection of critical loops

"A" and

"B".

(a)

How is.ESF train separation preserved when placing

" third-of-a-kind" CCW pump P-25 in service ?

(1.0)

(b)

How will the CCW cystem respond to receipt of a

low-low CCW surge tacik icvel ?

(1.0)

.,.

-n '.-,. !c.n. :

s

.;)

G " K i r t~ L' 7 y " i r;t cr 1 sec :

equiren both Iocked cloced intcrloc.:t in

-

tnlockad ccndition b.afore op er a t _ u.,

cf

_na w for s.vitch 4-:n i c h maintain;

,

i aloctrical z ep p a '; 1 cr..

bit c h.tn i c /41 cp.< rm t i

.o 1., maintain :d ny interlock rei ni. i ormn i n n 0, :t :t _. the Jucticn and diccharqa

/c l v es and si : r:

- } c v:

solatien valvec acccci; tad sich pu,p

.4 5.

'_*

.A Low-L:u CCi!...e I caci.;i zy r a l will icolate

,

'_ h e r. a n--c r 1 i c -

d urn au:o.

'surga tank

"'~

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Training Bypt ':m d eccri pti cn, Cc m. pan ~nt Cooling Ncter

..y a bem,. Papa 1i to 16.

'

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-

6.12 (2.0)

Reguarding the Ex-Core nuclear safety channel instruments :

(a)

At low power levels -(approximately 10-2 to 10-8

%

power)

how are neutron-interactions selected over gamma interactions ?

(1.0)

(b)

At power levels between 10-2 and 200

%

power, where individual pulses cannot be counted by normal methods, how is the power level si gnal generated ?

(1.0)

ANSWER :

(a)

A culse hight di nari mi nator is uneo to eliminate lower amplitude camma pulses frcm the higher amplitude neutren pulnes (b)

At greater than 0.02.

aawer, neutron pulses pile up and produce an elC 11 gnal which F1u;tucten around a

mean DC signal.

Th t'

aquare af this fluctuation is proporticnal to on-wr.

(i3camas have a

cmaller f l uctua ti cn about the man ano, nince this fluctuation

's 2cucred.

th= camma-to-neutron

.

current ratic is even l ar ger,

resulting in cH e:tive gc.mma dist_r einaticn.)

REFERENCE :

' Songs Training System Deucription, En-Core Neutron. Monitoring System.

i l

'

l

.

e q---e-

.

,

.

7.01 (4.0)

With only one Emergency Chilled Water System (ECWS)

operable, Technical Specifications 3/4.7.10 Action Statements address

Hour, 8 Hour, and 24 Hour requirments.

(a)

What verification does the one (1)

hour ACTION require ?

(1.0)

(b)

If Unit 2's ESF power suppl y enters a

degraded mode what precautions should be taken for the ECW system ?

(2.0)

(c)

Why should two (2) Emergency Chiller Units not be operating,or stopped simultaneously ?

(1.0)

ANSWER :

(a)

Verify that the normal HVAC syctem :s providing cpace ccling to tha vi tal power distribution rccms

on t ai ni r n

e emergency oattery chargers and inverters that decend on the inoperable ECW loco.

( b,'

Thn urce o+

power thould be the alternate unit

=r 3 ( 1. (D. Aacitional" the ocwer cource of i9us ses GS anu 3C, snicn suppl. the ECW a u:' : tiaries ( p-160 &

16D. aust also be can,1do md (i.O),

ic)

If the ECW unitc ce 4to ped.

the oil pump is oreventeo c ram ocing restarted fer a or c::.

minutes to allow precsure c. c a y.

Theref ore if a less of off-sita pcuer or manual stoppino of the units occured, whi l e the were both running.

ECW ccoling would be last for this time.

REFERENCE :

l l

'lon q c Daeratino Inntructions, S023-1-3.1. REV.

7, Training system acccriotion. Harmal and ECW system, and Technical Joecifications l

'.

-

<

.

7.02 (2.0)

Under certain conditions, additional Personnel Dosimetery must be worn.

(a)

When must additional dosimetery be worn on the head, legs, or gonads

?-

(1.0)

(b)

When must additional dosimetery be worn on the extremities, hands, forearms, feet, or ankles ?

(1.0)

ANSWER :

ta)

Additional dosimeterv is required Nhen localized whole badv Jane ir exnected to excced the trunt:.

(b)

Add i t i cr.a1 docimitry is required unen expecten done exceeds "he coce to the trun!' (by a factor of

or more, Or wnen the quarter 1v exposure to the extremoti25 in i 1 e : coed 4./ rem.)

REFERENCE :

.

I

?onas 123-VII-4.1.2, E':ternal Radicttion Danimetery,rev 2, page 4 L 5 of 14.

>

u e

-,

'.

-

'

.

7.03 (2.0)

(a)

What are the quarterly Administrative l i mi t's for radiation doses to the :

Whole Body ; Extremitie_

and Skin ?

(1.0)

(b)

Administratively, how far can the quarterly limit for Whole Body dose be extended ?

(0.5)

(c)

What is the Administrative limit for exposure to airborne radioactivity ?

(0.5)

ANSWER :

(a)

Whole bcdy = 900 or

0. 33 ).

E::tremi ty = 4.7 rem (0.33)

Skin

= 3.75 rem (u.33)

tb]

2250 inr/Otr tc)

30 MPC hourc in any 7 day oeriod.

9EFERENCE :

Langn 123-VII-4-0 Rev.

E, pope 7 and 9.

.

...

. -.

.

.__

.

7.04 (0.0) Question deleted i

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.

7.05 (2.0)

Conditions are : atarting a normal Heat-Up following a refueling, a

bubble is in ;he pressurizer and the RCS has been vented.

The check-off list for the Reactor Coolant Pump's electrical an valve line-up has been completed.

In order to start the first Reactor Coolant Pump certain initial conditions must be established or verified.

,

(a)

Besides adequate NPSH, identify four (4) initial conditions which must be ratisfied prior to starting the first RCP.

(1.0)

i (b)

Identify the three (3) precautions which should be followed prior to starting the three remaining RCP's.

(1.0)

e>

u.;-

-

u 9.25 cccn icr

.3ny-ou

L'

.. cn np :

.

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Opper anc

- ier b a.7 ac

,1.

.ne'.cire <aust be

'vil 2) Mermal d,W

law

> t. t *

.k c d to the octor c a c i. <2r s

.3 n d xcat * : r.t ar o :.t

r,aer.

).ontrol1ad bl eedof -

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<-

-12

. m cil 11 h p u c.p

.

- veruc rct2 tion

r

r'

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n 7 t an cj b'/ )

5)

!o-nvarce

' otat: -

.. n a i c at c d (nero cpced 1 1 t;h t-is illuminatca)

c) Pump r.eal s ar e procer l y it :e3 rd 7) CCW aupply tempera re

'.a the ncn-critical loop n 1 m.s than 95 F.

I 9) No abnarmal nia* rs f r:r

-

n-

+o be c t o r t e ri.

(b)

9.33 each for the f el m:. r

.

o not attempt #c

t.

+

n 9CP if e'

ccn

+

at;.tien is i ndi catac+.

T t: "-t cnly onc

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r0 n 5tartC)

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.

7.06 (3.0)

Initial conditions are

A LOCA has been sustained, RCS subcooling is less than 10 degrees, and RCS pressure is less than 1900 PSIA.

Identify the significence and reason for each of the following :

(a)

All RCP running and RCP AMPS low (< 500 each)

(1,0)

(b)

All RCP running and S/G DELTA P low ( < 20 PSID)

(1.0)

(c)

30 minutes after the LOCA, the source range count rate is slowely increasing.

(1.0)

ANEWER :

,

(a)

Indicatec

/oldina and inadequate core cooling.

Densi,

cf L:atcr in tha RCS is decrcasing below r;crmsi.

(b)

Indicates inadequate core cooling.

Pumus are not cume:nq much Flute cr crced ficw is inaaequate.

(c)

Indicato- 'naceaut,te core cooling.

Cauned by the

.

buildup af /aida.

!

REFERENCE :

Ccnas mi ti aating cera darr. age handout.

S O23-3--2. 30,

inadequate ccro cooling.

.

-

,-~s

._

-

.

7.07 (1.0)

Can the duty Shif t Supervisor directly supervise Core Alterations during refueling ? Explain your answer.

ANSWER :

r No Ccre alterations cha11 be observed and directel y supervined

,

by either a licensed SRO or SRO l i:ni cod to Fuel Handling, who has no etner concurrent responsibil i ti es ducira tnia operation.

.

$

}

PEFERENCE

-

Technical Cipacification S. 2. 2

..

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7.0S (2.0)

Procedure 5023-12-4,

" Steam Generator Tube Rupture",

Step 5b states

5) Establish Pressurizer pressure at Less Than 1000 PSIA 6) Maintain Pressuriner pressure Greater than 100 PSI above the

affected Steam Generator and as low as possible.

What are the basis for these precceding two steps 5) & 6) ?

!

ANSWO -

7^'-

bacis - cr iL ta reduce the oreccurre di ? '2ron t i c.i between

-

th2 RCG a,n d

_<G to minim za prinary to n ecend ar./

1.2kagc to

~

s t.ca cph er e.

Tht Sasic f m-

.i in t r-inut r ? that RCS pr*:curn

's higher than

,

-

rfG pr7GCurO

': 0 inSura

't?

dilutiCn 30C0 not

?CCur,

th Qreb'f

" D T.p r OUli C l h fl thO ~ h L1 ! C O! " 4

! 1 t' O i n.

,

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f i

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m

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'f r T1 C.. :r ~ ' i n c; i r ~ *. r ' '. L i c n. *

-12' 3

" T,'

pnge 5.

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.

7.09 (3.0)

The following questions pertain to procedure SO123-III-5.11.23,

" Liquid Effluent Release".

(a)

Identify four (4) sources of liquid effluent, besides the S/G blowdown processing system neittralization sumps (BDP),

for which a batch or continuous liquid effluent release permit would norn all y be generated (1.0)

(b)

If liquid effluent monitor-2/3-7813 is inoperable, what steps are necessary for effluent release of other than the BDP neutralization sumps

?

(1.0)

(c)

With less than the required minimum number of liquid effluent monitoring channels operable, what two (2) steps must be taken prior to initiating a release ?

(1.0)

ANSWER :

.25 each any f out- (4)

L)

1) Primary Plant Makeup 5tcrace Tants, 2) Radwaste

.

Pr:veary TanPc, 3) Radwaste 6econdary fanks, 4)

~4 i n c. veste Cond ncata Manitor fanks, 5) Other (b)

For all cources cacept UPS If monitor 2/3-7E13 is

-

inoperabl2, andnpendant Jample, analysis and erleace rate calculation verification is required.

(c)

1) At least two independent samnles are an alycod 2)

At least two ' techni call y quali f i ed nambers of the Facility staff independentiv verify the release rate calculations and diccharge line valving.

.

REFEREffCE' :

's en g s

'l ach Spec.

3.3.3.8, tabl e 3. 3.12 Acti on 23, and proceduro 10122-1II-L 11.23.

.

7.10 (2.0)

The COLSS calculated core power provides an adequate margin for normal power transients and equipment failures.

A minimum core power of 20 % of rated Thermal Power is assumed by CPCs in its input to COLSS for determining Linear Heat Rate, Planar Peaking Factors, Armuth Power Tilt and DNBR Margin.

(a)

What is the reason for maintaining the 20 %

Rated Thermal Power threshold ?

(0.5)

(b)

What is the reason for the increase in DNBR penalty factor as fuel exposure increases ?

(0.5)

J What relationship does the linear heat rate limit of 13.9 KW/Ft have with the safety limit of 21.0 KW/Ft ?

(1.0)

'79WER

.

t 2.'

The

_10 *. ihermai user ru ni mua a-.eumed threnhcid is d _t e to tr.e 7eut cn lur dctoctcr avctaa being inaccurate aelow.70

. ;ouer.

'

( b

The die R W en tr i i..e actcra increase t i.h fuel

.

e ';; o w r 0

'?

nuna r c -!

b ;;w incre aes with a ver a :;o

.:urrup

,: cr en c 2d by

' hat a s mtru l y.

(c)

Both l i >* : i c vct: t u.ctor care from fuel damage frcm nich tanperOture during can t :. c i p at ed operational cccur;nces 2nd/or design b asi s accidents.

.

I REFEFWNCE :

Nchm t ;u luccicic;tt2 nc 3/4.2.1. bases z.2.1

l l

,

-y

7.11 (1.5)

Below 235 F

deg.,

the reactor coolant system overpressure protection system must be operable.

(a)

Under what conditions is this requirement applicable in mode 6 ?

(0.5)

w(b)

What is the lift setting of the Shutdown Cooling System relief valve ?

(0.5)

(c)

If one of the relief isolation valves was inoperable, what alternate method of assuring compliance with the overpressure protection requirement is acceptable ?

(0.5)

]

AN3WER :

(c) With tne veccel head on (b) 106 + cr - 10 P910 (4ar temo'a 120 tr )

( c.'

RCG dauressurized with an 7CG vent of at least 5.6 Square inchen.

.

REFERE.NCE :

7,: c h '?,p e c. J. 4. 3. C. '.

._.

7.12 (1.5)

The Reactor Protection system and Core Protection calculators with inputs from COLSS functions to initiate protective action to assure that acceptable RCS and fuel -design limits are not exceeded during Anticipated Operational Occurances.

(a)

What two reactor trips associated with the RPS are initiated by the CPC's ?

(0.5)

(b)

How do CPC auxiliary trips differ from trips within the CPC " Operating Space"

?

(0.5)

(c)

What is the significence of a positive (+)

" Quality Margin", as used in the CPC ?

(0.5)

LWSWER :

'. 2 )

1) low droarturn from DNBR c.tio 2) high local pcwer canaity

'b)

Uar suniliary trian no dc>t ai l ed Dr&R cr iJ-D calcul.1tions are perfcrmed.

(c)

Duelity margin indicaten uhearner Fhe water ccnditi n ax;tirq

"h7 cero is cubcooled cr not.

.

(Note:

Per O t'f 6/

m,.t or i al

+ f ?M 1mdicaten

. ot cubcoolad but + 0 *' ~

LPC dic,clav indicates reactor

colant :s cubccolua)

.

REFEFENCE:

S onet s Trainion Gyctem Detacripticn for RPF

.n d CPC system.

l 'd..

.

8.01 (2.0)

The following questions pertain to Technical Specifications for fuel loading and refueling :

.

(a)

What is the Basis for mantaining a minimum water elevation of 23 feet above the top of the fuel in the reactor vessel and in the spent fuel storage pool ?

(1.0)

(b)

What is the Basis for restricting the movement of loads, in excess of 2000 pounds, over the fuel assemblies in the storage pool ?

(1.0)

3. NEWER :

<a)

The 23 feet in 2.vai1able ta removo 99% of the 10%

.

Iodine

,c t i vi t y 2n the foul gap, annumed to be released in tno nvent f

dmocco and damaued fuel assembi".(1.9)

(b)

In the eeent

i d r c_ u ; _*a icad the avtivity e

release will bc.inited to that cantained in a

,

ningle Fuel

550'00lv,(0.b)

and any cossi b t 's

.

j distortion ce fuel in the.udi stcrace racks will not ecult 1 cri' cal arrz v.

.. G ;'

,

.

,

i

1 PEFERtil4CE :

TccFnical Specificaticn J.0 r

- - -

.

- -

-

. - -

. -

- - -

-

. -

- - -

-

.

..

-

.

-

.

- -

..

-

.

8.02 (1.5)

What are the three requirements which must be met in order to make a termocrary change to a refueling procedure ?

ANSWER :

(2)

Intent of th e cri cti n a t urocedure is not al tr rod.

to.5)

,

a

{b)

ar] c r e v e d by

"""hcrs c?

,2 : an t manocement staff at

'

least are ev

eh om Mrs a ERO lic2rme 7n tr 7 afFacted uni t. (0. 5)

')

The cnante is d ocu;nen t ed,

reviewed, and acoteved bs the 9tation Manaccr or his designea within ii f

davs of i ma l i ment at i o.. (.0, *13 j

s

!

J s.

-

REFE1UENCE :

,

T cnnic21 ;;>uc i f i r.a t i on 6. 9. :?

o m

,,

,p

,,#.

.... -

+-

e v-

--

-

"

' ' ' " ' " ' " -

!

.

8.03 (O.0) Question deleted.

.

..

8.04 (2.0)

Shift manning procedure 5023-0-14 states that : -During energency conditions the shift supervisor will assume overall responsibility for the operation of both units.

(a)

If both units' are affected by an emergency condition and only one SRO Ops Supervisor is in the control room, explain the control room command function, and primary responsibility fer directing each units operations.

(1.0)

(b)

Under what conditicns can the Shift Supervisor absent himself from the protected area, while in mode 4 o (1.0)

(c)

Ex p l ai n the responsibi.lities of the individual with ccntrol room command function, during normal and ' emergency operating conditions, reguarding maning a

watch station for rollet purposes. (deleted)

72WEF

-

'u If both unitt v r-

+ t +ere nd r em ri-

<:

ah _1 :

.- [

thO O p Ot'J i~, r n 9

O Ct"ri ~ C" 1 ",'i U r ~ e

".

cO nn )n d

.

-

fur.CtiOn On1' ~~ rl!'

'

U U.p C's. : O ""

..

^ " " C. ":t r 2 i

'" C 011, thr?

F.-

luCCr 1 1 7 '-

it

!

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"! :, r O C - i r a c:O cr 7 t i c, * r-m'

'7 f1 m ' a r, i 7,

i

I,*

!""

".hD

,120 C 20 0 *.

'it p ;" t * ? r ' '

iCCOCD31bliit, DF 1 F Oc t i r. ]

' h f'

ithOr i

Fd

"

.,

thO Chift Supur;.CJr han

/Grali P O U p C n 7 i b l 1 1 L',,

fOr bOth unlDC.

,

t )

d tF i r, Q

- f r t'

Or Ot h t?r ir:C r g e n c y ID

'M Z'.f G G

'h)

I f".p d C t Jf C O n t i r.*J M d C D C r a t i c n 3.

2) 13r i. e f huainens nuit to t.i', n. U. S. Suilding.

FEFEli'!O:C.

[

C QD Q r,

> l ~~,. *r ? ? i n(1 t51'f ' r it *_ G n c. 0 2.J. " O 1. 4, C h i t' h flann l lig, POQQ b.

a i

k'.

.

8.05 (3.0)

Scheduled work periods, for the minimum shift crew, should n o.t

,

exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I (a)

What are the maximum hours an individual of the minimum shift crew is allowed to work in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, considering travel, meal and turnover times ?

(1.0)

(b)

When is an individual deemed to be operating the controls of a nuclear facility (as defined in 10CFR55) ?

(1.0)

(c)

What is the total minimum shift crew composition, for units 2 and 3, if unit 2 is in mode 3 and unit

3 is in mode 6 moving fuel ?

(1.0)

,

!

MN5WER :

'. 2 )

16 hourn in 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, 24 nearn in a SS hour period. 0:< c l ud i n g travel, meal. ana turncver time.

B)

nn individual is deamed to cpr: rate t he controls if he dircctly mnipulaten cr directa e,n c t h er to

' nan t oulat a tre control s.

i ()

1 95. 2 ";RO, 3 PC, I AQ, 1 GTA,

"

'une tabl e onlcu cc a p p r op r i c..t e )

Dull ;

U'31I 3 3liEM61SO E6900 10I06 G5

0

1 9RO

0

2 RO

1

3 AD

1 e

!

STA i

0

.

NEFEPENCE :

Congn, 'Jo er at i o n Inntruction 8023-0-14 attachment 3.1.

.:

I

i

,_,

.,.

,

m.

--

_

-.

.

8.06 (3.0)

What are the three (3) operability requirements for the Fire Suppression Water System ?

.

ANStiER :

(a)

No electric nottr-driven K'. E ) and one d i e s e l --

(1.0)

driven eirc cump t s) (0. 5).

(b)

nio s oc ar ',t e wat cr supp12 es.

(Fach with a minimum (1.0)

volumc cf !.00,000 gallan5.)

(c)

An opere.ble clew path.

Capable e-f r a L:i n g suction (1.0)

.

from aach water suppl',-

2nd tran2foring tro v:ater throuch di stri ottti en pipinq./

r.EFERE! !CE :

Technit.t1 Opec:i2 cation I/4.7.D.1

_____________ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _

.

8.07 C. O)

Technical Specifications contain limits for activity within the Primary Coolant.

(a)

What is the limit for specific activity of the primary coolant ?

(1.0)

(b)

What is the basis for the specific activity of the primary coolant ?

(1.0)

(c)

What is the Basis for the Technical specification action statement which allows continued power operation for less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , when the specfic activity e::ceeds the limit

?

(1.0)

ANGWER :

( m. )

1.0 microcurie / gram Dose liqui v a l ent I-131 and Less than cr ocual to 100/E microcurics/ gram.

2)

The banis in to limit the specific Octivity of primary coolant to ensure that the resulting dacas at the qite boundry will not e:< c ee d an appropriately mall part of the Part 100 limits fol?owing a 7 team generatcr tuce rupture.

(ci The basic 17 tr ac cmidato passible ladino coi!.;ing 9EFEfiENCE :

' echr: tal Q;ec: f i ca t t ers J./4.4.7 b

..

-

.

8.08 (2.0)

The following questions pertain to procedure SO123-VIII-10; Emergency Coordinator Duties :

(a)

For a Unusual Event (UE), Alert (A),

or Site Area Emergency (SAE),

under what conditions is a

precautionary evacuation warranted

'

(1.0)

(b)

Under what circumstances are a " Plant Evacuation" and a " Site Evacuation" mandatory "

(1.0)

(c)

During a UE, A,

SAE or GE, when an individual can be released from mitigigation efforts, to which post should he be assigned

?

(deleted)

ANSWER

.

(c)

If conditicro indicate that hazarnc _ ;: 10 :

ohicn nay cnd.nger porconnel safety.

(b)

Di an t.

evacuation in man t! a t or y at dite area E.r.er c on c y.

Site evacuatinn i.s mandattm ctt Cieneral Emergencv.

i l

t

!

F.EFERE' ICE :

Conan urocedum E0123--VI I I-10; Emeroency Ccordtnator Duti es,

,

'

nago G G

r-

.

8.09 (2.0)

Identify two (2)

circumstances for which independent verification, by a

second qualified operator, of systems important-to-safety need not be performed.

ANEWER : 1.0 each f :.r

,r y t >o of the follcNing.

(1)

T"o se porti ons of s cyutam within a ra it ation area

.

w:ii c h wi:1

<ecult in areater than 10 nr uncle body dc su.

'. 2 )

Wnen componentc are Accated in a ecsted airborno contaminaticn area.

(3)

Wbren thr*

amconent in not noutred to be indecendently :ertifled.

i 'l )

If functional t e r.t i n g can be certcrmed without camprcmicinc :afety.

and can prove all equipment,

'< a l v e c, and twitches involved are correctolv 21 l i gned.

"< EFEFeer:E :

'.innon 2.23-0- M. rev.

, Control of nyctem alignments, page 8.

i h

l

.

, _,

. _. - -.

--m--

d

"

r a

-

,

..,

8.10 (3.0)

The following questions pertain to Technical Specification

" Limiting Conditions of Operation and Surveillance Requirements" and their BASE 3 :

(a)

How is the OPERABILITY requirements for a limiting condition of operation affected by a failure to perform a

Surveillance Requirement within the specified time interval 9 (1.0)

(b)

Can entry into or through an OPERATIONAL MODE be made if the limiting conditions of operation cannot be met ? EXPLAIN (1.0)

(c)

What is the BASES for the following Technical Specification 3.0.3 which states

When a

"

Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour, action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it as applicable, in :

1) At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

')

At least HOT SHUTDOWN within the following

hours 3) At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

(1,0)

rNSWER :

'a)

Failure to

. cr f er n.

the e c aui r 'rt

r ; 1 1 1 m c r.

ccnatitutc.;

F a i i u;n?

0

~. c u t

o crai:111 ty recuirencntc.

(b)

Pascage thrr.ucr rr into 2n ' n rr ',t i on: l noun iu acceptable tc r: :mp! v ni th AC TI Ol' '~;u i r mn q tc.

(c)

The D T,r:5 i c:c

.0,"

i, to del!niatr t-he

? a nur e r; to

?O t-

~n EGr cli P um 7t Jr. : c 9 tiot

ji r e c t q l y pro'.idPd t. '

1:1 lI:'2 ( i,'T I O d

nt<a gni

, ;ne

,j ; o r, g

' c.~ '. t J. f i :J hCuld

/ I. '2 i a O 62

' h.:

' [1 t on t, a{

0'J1 l * ~'. i t i c f l.

I E T[$Eli T.

r'

noic !

'L

. i

_ t t i <: n > >,

".O.2,

_

6. t a n;j no ' t ' of 3.O,7

..

-

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..

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,

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~

i U.S. Nuclear Regulatory Commission i

Reactor Operator License Examination

$

-

l Facility: San Onofre Nuclear Generating Station Reactor Type":"" # ""

Combution Engineering

Date Administered:

July 30, 1985 Examiner:

P.J. Morrill

-

Candidate:

~

INSTRUCTIONS TO CANDIDATE:

'

Use separate paper for the answers. Write answers on one side only.

Staple i

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

<

Value Total Score Value Category

25 25.W 1.

Principles of Nuclear Power Plant Operation, Thermodynamics,

Heat Transfer and Fluid Flow l

25.r,V 2.

Plant Design Including Safety and y,g g,g Emergency Systems

!

M

i.

3.

Instruments and Controls l

25 4 4.

Procedures - Normal, Abnormal,

!

ps Emergency, and Radiological Control

,

l g

100 TOTALS l

i Final Grade

%

'

All work done on this examination is my osn.

I have neither given nor received aid.

,

Candidate's Signature

.

,

.- _ - - _ _

.. - - _.. _. _. _ _

. _. -. _ _ -. _ _ _ _ - _. _. _.. _. _ - _ _,.... _. _ _..

.-

-_.

EQUATION SHEET

'

.

f = ma v = s/t Cycle efficiency = (t:etwort out)/(Energy in)

<

.

-

,. og s = V,t + 1/2 at

-

I E = ac KE = 1/2 av a=(Vf - V,)/t Afin A = A,e'

PE = gn

-

Vf = V, + at w = e/t t = s'n2/t1/2 = 0.693/t1/2 1/2'## * EI*U')I*S)3 t

-

g

,y,

[(c/2)*(*b)3

'

aE = 931 as

'

I = I,e' *

'

,

Q = stpat Q=UA4T I = I,e~"*

I I = I,10"* N

'

Pwr = W an

l TYL = 1.3/u P = P,10 "'III

HVL = -0.593/u 9 = P,e /T t

.

SUR = 26.05/f SCR = S/(1 - K,ff)

T * I YT DI

,

CR, = S/(1 - Keffx)

SUR = 25s/t* + (a - e)T CR (1 - K,ffj) = CR II - *eff2)

j

l T = (1*/s) + [(a - s)/Io]

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

'

l T = s/(s - s)

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

l T = (s - e)/(as)

SDM = (1 - K,ff)/K,ff

p = (X,ff-1)/K,ff = AK,ff/K,ff t' = 10 second T=0.1 seconds"j

!

p = [(t*/(T K,7f)] +,[T,ff (1 + [T)]

/

.

Idjj=Id 2 =2 2 P = (I4V)/(3 x 1010)

Id Id j

E = eN

!

'

R/hr = (0.5 CE)/d (s:etars)

R/hr = 6 CE/d2 (feet)

,

Water Parameters Miscellaneous Conversions

.

) gal. = 8.345 lem.

I curie = 3.7 x 1podps

,

!

I ga;. = 3.78 liters 1 kg = 2.21 los

.

!

I f t-

= 7.48 (a1.

I hp = 2.54 x 103 Stu/hr

-

Density = E2.4'1bm/ft3 1 me = 3.41 x 10' 8tu/hr Denstiy = 1 gm/cm3 lin = 2:54 cm

Heat of vaporization = 970 Stu/lom,

F = 9/5'C + 32 Heac of fusion = 144 8tu/lbe

C = 5/9 ( *F-32)

-

1 Ate = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-lbf I ft. M 0 = 0.4335 lbf/in.2

.

r.,- - -, - - - -, -, - -

,

.-

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i-s 1.01 (2.0)

,

A steam generator relief vgive and a pressurizer relief valve are each leaking to atmospheric pressure.

The steam generator is operating at 1000 psia and the pressurizer is operating at 2300 psia.

s, l

.t a.

Show the leakage process for each valve or, the attached MOLLIER CHART (Figure 1.1) (Clearly lable each valve!)

- c, (1.0)

b.

For each valve, i s the leakage steam Saturated or Superheated?

(1.0)

,

s

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STEAt1 TABLES

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FIGURE /, /

1000 1.335 1.650 1965 2.300 l.0 l.2 1.4 1.'6

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/ 1.6 1.8 2.0 2.2 1.0/

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1000 1.335 1650 1965 2 200 ENTROPY, s, 8tv/lbm. R

,

6. E Y FIGURE /,/

ID00 1335 1650 1965 2.300 yggg'0 l

1.2

!.'4

1.'8 ' ~ YO~~ ~~3~~~

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1000 1335 1650 2 300

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ENTROPY, s, 8tv/4. R

.

.. _

1.02 (4.0)

Refer to Figure 1.2 which shows an instantaneous, positive reactivity insertion into an already critical reactor core (at time t=0),

f ol l owed by a removal of this positive reactivity after a stable reactor period is reached (at time t=1).

Assuming no source neutronst a.

Show the resulting reactor Startup Rate as a

function of time for this reactivity change.

(1.0)

b.

Show the reactor power level as a function of time for this reactivity change.

(1.0)

c.

Explain the shape of the reactor power response at a time IMMEDIATELY AFTER t=0.

(1.0)

d.

If the Keff was made equal to 1.0020 at t=0, what startup rate would result as time approached t=1?

(Assume g = 0.0072 and A

= 0.1)

(1.0)

j a & b attached c.

" PROMPT JUMP" in total neutron flux due to an increase in prompt neutron production.

d d.

/

= Keff - 1 / Keff = (1.0020 - 1.0000) / 1.0020

/

= 0.0020

/$ - /

d>

T=

( f or / <~ j hf (0.0072 - 0.0020)

T

=

= 26 sec.

_

(0.0020)(0.1)

SUR = 26.06 / T 1 dpm

=

Equation sheet, Neutron Kinetics l

l l

.

_ _.

-

-.

-

.-,

.,A-

-s

_

a

--

a L

a

-

%

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1.03 (1.0)

Explain why the stable negative startup rate (SUR) of the reactor cannot be faster than approximately -1/3 dpm after a

reactor shutdown.

The shortest negative startup rate is determined by the decay rate of del ayed neutron precursors.

(The longest delayed neutron precurser half-life is 55.3 seconds yielding a negative period ot 80 seconds or a -1/ 3 cpm SUR. )

G.P. Physics, V:: II, Chapt 5 l

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1.04 (4.0)

Differential rod worth is affected by various factors.

How will each of the following factors, if considered separately, affect differential rod worth ?

'

a.

Axial rod position.

(1.0)

b.

Radial position close to other control rods.

(1.0)

c.

Temperature (increase in Tave)

(1.0)

d.

Boron Concentration.

(1.0)

a.

Rod movement in an area of higher flux will yield a greater differential rod worth than the same movement in an area of lower flux, yielding a greater differential rod worth in areas of greater flux.

b.

Shielding by other rods resul':s in a l ower differential rod worth for rods located in clo se proximity to each other.

c.

The increased neutron migration length at higher temperature will cause differential rod worth to increase due to the increased probability that the neutron will reach the rod.

d.

Shielding by the boron will l owe r the di f f eren t i al rod

.

worth.

!

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G.P. Physics, Vol II, Chapt 4 l

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1.05 (2.00)

A startup is being performed after a reactor trip using an estimated critical position calculated on a startup commencing at 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months after the trip. How will each of the following events or conditions, occur i ng i ndepetiden tl y pr i or to the startup, affect the actual critical rod position as compaired to the estimated critical position?

(Higher, Lower, No Change)

b a.

A steam generator's level is increased significently with feedwater at 450 deg. F.

(0.5)

b.

A reactor coolant pump is restarted.

(0.5)

c.

The startup is delayed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(0.5)

d.

The steam dump pressure setpoint is decreased.

(0.5)

a.

Lower b.

No change c.

Lower d.

Lower Tech Spec 3/4.1.1.1&.2 l

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1.06 (4.0)

Figure 1.3 is a sketch of Reactor Power vs Time in hours.

Prior to t=0 hours the reactor has been operating for several days at 50% power.

At t=0 the reactor power is increased to 100%

where it remains for the next several days.

a.

Sketch the Xenon concentration response in the reactor core from this power transient.

(1.0)

b.

Sketch the Iodine cencentration response in the reactor core from this power transient.

(1.0)

c.

At approximately what time will the Xenon concentration be at equilibrium again?

(1.0)

d.

Will the 100% power equilibrium concentration of Xenon be:

(1.0)

(1)

double the 50%

power equilbrium concentration.

(2)

more than double the 50%

power equilibrium concentration.

(3)

Less than double the 50%

power equilibrium concentration.

a,b,& c attached d.

(3)

less than double the 50%

power equilibrium concentration.

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1.07 (4.0)

Refer to Figure 1.4, a sketch of a centrifugal pump curve and a system characteristic curve for an auxiliary feeowater pump system using a centrifigual pump, for the foll owing quest i ons:

a.

Show how the curve (s) will change as the PORV opens and reduces the steam generator pressure by 50%.

(1.0)

b.

Show the changes to the curve (s)

as the pump discharge valve is partialy shut.

(1.0)

c.

Explain how the net positive suction head of the pump would be affected by an increase in the temperature of the water in the storage tank.

(1.0)

d.

Explain how a flow restriction in the suction line from the storage tank to the pump would affect the net positive suction head of the pump.

(1.0)

a&b attached.

c.

Increasing the water temperature increases the saturation pressure of the water which decreases the NPSH.

d.

Under f l ow conditions there would be a pressure drop across the flow restriction decreasing the actual suction pressure thereby decreasing the NPSH.

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1.08 (4.0)

For the fo11owing questions refer to the attached steam l

table.

i

million Lbm / hr of reactor coolant at 2100 psi enters one of two steam generators at 610 degrees F and leaves at 550 degree F.

Feedwater at 450 degree F enters the secondary side of the steam generator and leaves as saturated steam at 550 psi. The S/G heat transfer process is represented by the f ol l owi ng equationt:

e Q = m cpAT where cp = 1.1 BTU / Lbm-F

.

G = m Ah

.

G=UAAT a.

What is the total thermal power,in BTU'hr of the

/

reactor?

(1.0)

b.

What is the total steam flow rate from each steam generator?

(1.0)

c.

What would be the steam pressure if the total heat flux (power) were reduced to 1/4 of it's initial value in steady state?

(1.0)

d.

What Tave would be required to maintain a steam pressure of 550 psi at the reduced (1/4 initial)

stemming rate?

(1.0)

-

.

i

.

l a.

0 = m cpa T = 2 m cp (Th - Tc)

T

= 2 (80 x 106)(1.1)(610-550) = 10.56 x 10 BTU /hr

.

,

l b.

0 = m ah 1/2 0

= 5.28 x 10 BTU /hr (part a)

=

'

d=m (1204

- (cp>(T))

(from steam tables)

m = 5.28 x 109 BTU /hr / (1204 - (1.1)(450)

l m = 7.45 x 10k Lbm/hr 6 = UA A T

= UA (Tave - Tstm)

where Tave = (Th + Tc)/2 c.

UA initial (580 -478)

Tstm = 554 degree F^

=

UA final (580-- Tstm)

P stm = 1070 psig (steam table)

d.

UA (580 - 478)

=

UA (Tave - 478)

Tave = 503 degrees F G.P. Nuclear Tech

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Properties of Saturated Steam and Saturated Watee--concluded

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Pressure Temper.

Heat of Latet Heat Total H,.:

Specshc Volume Lama, per Sq. Iri.

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250.0 335.3 400.97 376.3 825 [

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265.0 340.3 402.72 378.0 822.3 3203.3 0.01868 3.80h02

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270.3 412.67 388.9 813.6 1202.4 0.03882 1.62218

}90.0 275.3 414.25 390.6 812.0 1202.6 0 01885 1.59462

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m.0 280.3 415.81 392.3 810.4 1202.7 0 01887 1.56835 300.0 285.3 417.35 394.0 808.9 1202.9 0.03889 1.54274 320.0 305.3 423.31 400.5 802.9 1203.4 0.01899 1.44801 340.0 325.3 424.99 406.8 797.0 3203.8 0.01908 3.36405 360.0 345.3 434.41 412.8 793.3 3204.3 0.01917 1.28910 380.0 365J 439.61 418.6 785.8 3204.4 0.01925 1.22177 400.0 385.3 444.60 424.2 780.4 3204.6 0.01934 1.36095 420.0 405.3 449.40 429.6 775.2 3204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535

,

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460.0 445.3 458.50 439.8 765.0 1204 8 0.01959 1.00921 480.0 465.3 462.82 444.7 760.0 1204.8 0.01 % 7 0.96677 500.0 445.3 467.01 449.5 755.1 1204.7 0.01975 0.92762 i

520.0 505.3 473.07 454.2 750 4 1204.5 0.03982 0.89137

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540.0 525.3 475.01 458.7 745.7 1204.4 9.01990 0.85771 560.0 545.3 478.84 463.1 741.0 3204.2 0.0lM8 0.82637 580.0 565.3 482.57 467.5

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736.5 1203.9 0.0200e 0.79712 600.0 585.3 486.20 471.7 732.0 1203.7 0.02013 0.76975

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605.3 567.6 1172.9 0.02327 0.28900 l

1500.0 1485.3 596.20 611.7 558 4 1170.1 0.02346 0.27734 l

1600.0 3585.3 604.87 624.2 540.3 II64.5 0.02387 0.25545 1700.0 1645.3 613.13 636.5 522.2 1158 6 0.02428 0.23607.

1400 0 1785.3 621.02 648.5 503.8 1152.3 0,02472 0.21861 1900.0 1845.3 628.56 660.4 485.2 1145.6 0.62517 0.20278

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2.01 (4.0)

There are four direct piping system connections between the Chemical and Volume. Control system and the Reactor Coolant System

,

(RCS).

(a)

List these f our connections.

(1.0)

+

l-(b) ~ Explain the use of each of these four connections.

(3.0)

(Each explanation may-be brief, but should relate

-

the major design purpose (s) of the CVCS to the connection listed.)

,

!

.

(a)

0.25 for any tour of the following.

,

i.etdown line (o+f the loop 1B cold leg).

Charg.ng line to loop 1A cold leg

_

.

Charging lino to loop 2A cold leg Au::i l i ary pressurizer spray-line to the p r es sur l : er.-

Reactor Coolant Pump ncrmal seal leakoff.

,!

(b)'

O.75 for each of the following.

t

.

Letdcwn line for normal letdown flow to control s

l RCS water volume, and coolant chemistry.

.

j Charging line for normal-charging to control-RCS I

water volume, and coolant chemistry.

i-

.

j Auxiliary spray-line f or ccntroling pressure (on

'

cooldown)- when Reactor Coolant Pumps are not j

operatino.

i

'

Coolant Pump seal leakoff, for ceal cooling and to

. equalize pressure across the seals.

,

f REFERENCE:

i fraining System Description,- Chemical and Volume Control System,

,

pages 1,^2, 60, and;61

,

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i 2.02 (3.0)

Two signals from the Engineered Safety Features Actuation System

(ESFAS)

provide automatic control of components in the Chemical and Volume Control System.

(a). List the two signals.

(1.0)

,

(b)

Describe which CVCS components are effected by (2.0)

each of these signals and how they are affected.

(a)

Containment Isolation Actuation Signal (CIAS)

Safety Injection Actuation Signal (SIAS)

(b)

CIAS chuts the lotdown isolation line and the i

controlled bleedoff line isolation valve.

SIAS starts the charging pumps, shuts the letdown isolation valves inside containment, and actuates baration.

.

'

.

REFERENCE:

Training System Deccriptions,

' Chemical and Volume Cc.itrol System page 62,-

and Engineered Safety Features Actuation System pages'5 and 7.

-_

_

.

_

,_.

.

i 2.03 (4.0)

,

The Engineered Safety Features Actuation System (ESFAS)

uses carameters associated with P.he steam generators to automatically generate three different safett si gnal s.

(a)

What are these three signals ?

(1.0)

(b)

What ccmoination of steam generator parameters (3.0)

generate each signal ?

(Note:

Candidates advised during test to consider each staem generator separatly)

a)

Mnin Steam Line tool.>t or i c n.s i 1fiI S )

e mergency Feedwate octuat :n./ stem C'.F A S )

for stecm generator nuraber i Emergency Feedwate Ac tu. ri ct

-< tem (EFAS)

for s t er.m generator,uncer -

Steam generator icw delta P 4: lcu crimary flow)

is e.n acceptable a l b er r. at i x 3 i

h7 tream generator-IF05 signata 3.r o

" c a :, r. a d 20 an-'

nicnal by the

_

candidate.

(b)

M' IE is c;en t.r a t ;d v 1:w s t c -e rn generator pressure n cither s ta a:a ;cnercter.

EFAS for eteam cenerator numoer

(S/G-1)

is generated bv icw level in number 1 steam generator coincident with uit.her S/G-1 preocure greater than S/G-2 pre:suec cr 3/C-! pressure not law.

EFA5 Fc-

, t cn genorator n urr.b er

(S/G-2).

is generated by law level in number 2 steam generator ccincident with olther S/R-2 pressure creater than 3/G-1 prerisure or S/G-2 presuure not low.

I

.

9:EFERENCE:

Trainirg System Descriotion.

Engineered Gatety Features Actuation

ystem paces 9 and '1

-

-..

_

.

-

.

._. -

. _ - -

-2.04 (3.0)

You are beginning a normal plant heat-up following refueling.

A

<

. bubble is in the pressurizer, the RCS has been vented, and the check-off list f or-the reactor coolant pumps'elet:trical and valve line up has been completed.

In order to start the first reactor coolant pump certain initial conditions must be established or verified.

(a) ' List four of the five initial conditions which (2.0)

must be satisfied prior to starting the first reactor coolant pump.

(b)

List two of the three additional precautions which (1.0)

should be followed prior to starting the three remaining reactor coolant pumps.

(a)

v.J each for any four cf the following:

,

t;PcH requirement s must te met Upper and lower bearing oil reservoirs must be f e.tl l i

l Normal ccmponent cooling water flow must be established to the motor coolers and the seal water.5 eat exchanger.

Centrolled bleedoff Flow must be established.

.

..

I

An cil lift pump and an anti-reverse rotation device pump must be operating (redundant pumps in

'

standby)

(b)

0.5 each for any two of the following:

.

!

Do not attempt tc start a RCP if roverse rotation

,

is indicated, i

.

Start only one RCP at a ti me (5 mimutes between-

,,

starts).

Do not start the fourth RCP until PCS temperature is above 500 F.

REFERENCE:

Tenining System Description, Reactor Coolant System page 45 and crocedure 5023-3-1.7, Reactor Coolant Pump Operation 2.

,

_

~

e

.-

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--

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._ -

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- - -

2.05 (2.0)

,

!

The controls for the component cooling water system include START, STOP, and OVERRIDE pushbuttons.

(a)

What does depressing the START pushbutton initiate ?

(1.0)

(b)

What is the purpose of the OVERRIDE pushbutton ?

(1.0)

,

(

(a)

The START pushbutton starts the CCW pump and initiates the start sequence for the associated Saltwater Cooling System critical loop.

(b)

The OVERRIDE pushbutton allows the operator to shutdown one of the two CCW pumps that start on SIAS if it is not required.

i i

i

.

f i

!

l

REFERENCE:

!

Training System Deccription, Component Cooling Water System, page

5 l

l

'

, _.

-

.

... -

_.

. _ _

_

.

..

.-

2.06 (4.0)

During shutdown of the plant an inadvertent Safety Injection

,

Actuation Signal. occurs.

The emergency diesel generators start automatically.

(a)

Aside from engine overspeed and low lube oil (1.0)

pressure what automatic protective function will trip one of the diesel generators ?

(b)

Before resetting the SIAS, how does the control (1.0)

room operatar stop the diesel generators ?

I (c)

What signal would restart the diesel generators (1.0)

after they have been stopped by the contal room operator but prior to reset of the SIAS ?

(d)

If one of the diesel generators had been running (1.0)

in the normal mode of operation and was in parallel with off-site power.for testing, what

,

would that diesel generator do upon receipt of an SIAS ?

(a)

Generator differential will trip the diesel

,

generaters.

f (b>

To stop the diesel generator with an SIAS present the operator must depress two SIAS OVERRIDE STOP pushbuttons simultaneously.

,

(c)

Loss of Voltage Signal (LOVS)

(d)

The diesel generator breaker'would trip and the l

diesel would remain running in standby.

l

,.

i l

i REFERENCE:

Tr ai ni ng System Description, Emergency Diesel Generator ' System, pages 27, 28, and 29

l i -

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. - -.

.

%

2.07 (3.0)

Regarding the Core Protection Calculators-(CPCs):

(a)

What are the the two reactor trips initiated by (1.0)

the CPCs associated with the RPS ?

(b)

What events do these two trips protect the reactor (1.0)

,

from ?

(c)

How do the CPCs identify when one control element (1.0)

deviates f rom its' subgroup ?

,

(a)

Low departure from nucleate boilina ratio and high local power density, (b)

Cladding failure and fuel melting (respectively)

,

I (c)

Control Element Assembly Calculators receive inputs from all CEAs and send penalty factors to the CPCs based on rod deviations in a subgroup.

l

.

I l

l I

!

.

i

!

REFERENCE:

Training System Description,- Reactor Protection System and Core

,

Protection Calculators, pages 19 and 22 i

l

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2.08 (2.0)

Regarding the Class IE Electrical system battery chargers:

,

(a)

A FLOAT-EQUALIZE toggle switch is located on the (1.0)

front of each battery charger. What is this switch used for ?

l (b)-

When placing a battery charger in service why is (1.0)

"

the DC breaker. closed before the AC breaker 7

.

(a)

The switch is used to recharge the battery when in

'

the ECUALIZE position.

(It normally is in the FLOAT position and just maintains buss voltage.)

i (b)

Closir.g the DC breaker first alicws the charging i

cauacitors to be charged by the battery and avoids l

damaging the rectifier or blowing fuses, t

,

i

!

i

i i

,

d

!

l

,

REFERENCE:

l Training System Description, Class IE. Electrical System, pages 30 and 57

i t

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.

RO3 3.01 (4.0)

The Control Element Drive Mechanism Control System (CEDMCS)

receives Automatic Moti on Inhibit (AMI) or Automatic Withdrawl Prohibit (AWP) signals from~other plant systems.

(a)

List the signal (s) (AMI or AWP) from the Reactor (1.0)

Regulating System to the CEDMCS and describe the plant parameters or conditions which cause the signal (s).

(b)

List the signal (s) (AMI or AWP) from the Steam (2.0)

Bypass Ccatrol System to the CEDMCS and describe the plant parameters or conditions which cause the signal (s).

(c)

List the signal (s) (AMI or AWP) from the Plant (1.0)

Protection System to the CEDMCS and describe the plant parameters or conditions which cause the si gnal (s).

w ^ CTOi REGULh ii N3 SYSTEM iut:matic Withdrawl Frchibit signal is generated n

acassi m deviation between T(ave)

and

.

firef)

(6.8 c) or by high cold leg temperature

'559 n)

<b)

ITEM 1 WPOSS C'WTROL SYSTEM Autcmatic Motion Inhibit sicnal is generated when reactor power s below 1 5 */.

AND Automatic Withdrawl Prchibit signal is geners.ted when there is 2 comand for s t ear.3 bypass control valves te open (c)

PLnNT oROTECTIG"1 SYSTEM Neither, the Plant pr ot ec ti on system generates a CWP signal.

or Control El e:nelt Assembly Withdrawl Prohibit signal 17 geners.ted by pretrips from high reactor preusure, high D FT, or low DNBR. (Hicn kW/FT and Icw DNBR nratrips are not unabled until

!. 0 -i

".

power)

HCFERENCE:

Training System Description CEDMCS, pages 50 and 5i i

e

. _ -..

_

(b'Oh 3.02

5t*

_

The reactor is c-ting at 60 % reactor wer when the control bank is shimmed in O steps.

Sketch (o the attached Figure 3.1)

reactor power, (fuel), T(hot), T(col and T(steam) versus time

,

until new fi 1 values are reached Numerical estimates are not required, b

carefully show th elative relationships of peaks and trends f these parameters Assume no reactor trip occurs.

0W See attached Figure 3.1 (0.5 points per variable)

l l

[

'iEFERENCE:

i Academic Program for Nuclear Power Plant Personnel, Volume III,

Nuclear Power Plant Technoloay, Chapter 3, Section C

.

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3.03 (4.0)

Refer to Figures 3.2 for questions (a) through (d)

bel ow.

Plant initial conditions are:

Steady state 75 % power All control systems in automatic mode One feedwater pump trips prior to t = 0 At time t = O the operator decreases turbine power to 50 %.

t (a)

Why would the operator decrease power from 75 % ?

(1.0)

.

(b)

REFERRING'TO THE ROD POSITION RESPONSE:

(1.0)

'

What plant conditions or signals caused the

,

control elements to insert ?

,

(c)

REFERRING TO THE PRESSURIZER LEVEL RESPONSE:

(1.0)

.

After the intial pressurizer level transient, j.

why is the final steady state level lo,wer than the

steady state level b'efore the power reduction ?

(d)

REFERRING TO THE CHARGING / LETDOWN FLOW RESPONSE:

(1.0)

What plant conditions or signals caused letdown flow to first decrease and then return to its'

original.value ?

,

!

ta)

The capacity of cne feedepump is 65

%,

,therefore

'

to maintain steam generator levels the operator nust decrease pcwer celow this level.

(b)

Control rod insertion was caused by T (ave) 7(ref)

- missmatch.

(Automatic sequential mode is not normally used'at SONGS 2/3)

i (c)

The level depends upon the level demanded byL the

level contal program which is based on T(ave).

.

Since T(ave)

is lower after the transient the

'

level control program satpoint and steady ~ state

- level are lower.

(d)

Letdown decreased because pressuriner l evel was

.

below the level control program er level setpoint

was decreasing in responce td power decrease.

It.

.

returned to normal when actual level and level

control program matched.

.

!

PEFERENCE:

Training System

. Descriptions, Reactor Regulating System, Pressuriner'Contal System, Feedwater Regulating System, Chemical j

pd Volume Contrcl System l

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i 3.04 (3.0)

Refer to Figure 3.2 for questions (a) through (c)

below.

Plant initial conditions are:

Steady state 75 % power All control systems in automatic mode One feedwater pump trips prior to time t = 0

'

At time t = O the operator decreases turbine power to 50 %.

(a)

REFERRING TO THE STEAM GENERATOR LEVEL RESPONSE:

(1.0)

What caused the initial steam generator level decrease ?

(b)

REFERRING TO THE FEEDWATER FLOW RESPONSE:

(1.0)

Why did feedwater flow initially exceed steam flow ?

(c)

REFERRING TO THE STEAM FLOW RESPONSE:

(1.0)

What caused the steam flow to decrease ?

(a)

On a

power decrease, steam voids in the steam generator decrease which icwers the level in the cteam generator ( tih r i n k ).

(b)

Stcam generator level was less than programmed level.

(c)

Decreasing the turbine power decreased the steam flow.

REFERENCE:

Training System Descriptions, Reactor Regulating System, Pressuriner Contal System, Feedwater Regulating System, Chemical and Volume Control System

_ _.

__

_ _

3.05 (2.0)

Regarding the containment spray system:

)

(a)

What automatic si gnal (s) start the Containment (1.0)

Spray pumps ?

(b)

What automatic signal (s) open the spray control (1.0)

valves ?

(a). Safety injecticn actuation signal.

.

(b)

Containment spray actuation signal.

.

REFERENCE:

Training System Description, Containment Spray and Combustable Gas Control Systems, page-32..

o

.

, -.,,

3.06 (4.0)

The feedwater regulating system has two abnormal conditions, Reactor Tri p Override (RTO) and High Level Override (HLO).

(a)

What, automatic actions occur as a result of a RTO ?

(1.0)

(b)

What is a RTO designed to acccmplish or prevent ?

(1.0)

(c)

What automatic actions cccur as a result of a HLO ?

(1.0)

(d)

What is a HTO designed to accomplish or prevent ?

(1.0)

(a)

RTO closes the feedWater reQulating valves, and n asi ti onn the

eeduater regulating valve bypass 2nd F eed p u;r: 0 speed setpoint control to the 5%

f l ei s occitions.

(b)

The purocse ci the reactor trip override (RTO) is tc cravent pressurizer pressure and level transients resultina from overcooling or the reactc; c. calan t af ter a reactor trip.

(c)

HLO cleces the feedwater regulating valve and its bypass and nets main feedpump speed setpoint to itu mi rn mum.

(d)

HLO prevents moisture carryover to the main turbine.

l l

,

i l

l REFERENCE:

Iraining System Description, Feedwater Regulating Syctem, page 32 i

b

,

-

- - -

3.07 (2.5)

-

Regarding the ex-core nuclear safety channel instruments:

(a)

What kind of detector is used ?

(0.5)

(b)

.At low power levels (approximatly 10-2 to 10-8

%

(1.0)

power)-how are neutron interactions selected over gamma interactions ?'

q (c)

At power levels between 10-2 and 200 % power where (1.0)

individual pulses cannot be counted by normal methods how is the power level signal generated ?

(a)

4Tission chambers (b)

a culse height discriminator is used to eliminate l eezer 2mplitituce gamma pulses from the higher rmolititude neutron pulses.

(c)

At

-

10-2

% power neutron pulses pile up and produce an-AC signal which fluctuates around a

nean DC 91gnal.

The square of this fluctuation is procortional to power.

(Gammas have a

smaller

_ fluctuation about the mean and, since this

+1uctuation is squared, the gamma-to-neutron current ratio is even larger, resulting in e+ + ec t i v re gamma discrimination.)

..

REFERENCE:

Training System Descriptioni Excore Neutron Monitoring System-

-

_

-

.

3.08 (3.0)

For the radiation detection devices listed below (a through f),

state the kind of detector used and what type of radiation is

.

detected.

(a)

Containment area radiation monitor (0.5)

(b)

Excore startup channel detector (0,5)

(c)

Plant vent stack airborne monitor (0.5)

(d)

Excore centrol channel detector (0.5)

,

(e)

Low range' main steam line monitor (0.5)

(f)

High range main steam line monitor (0.5)

(a)

Scintillation counter, gamma

-

-(b )

Unit 2 fission chambers / Unit 3 BF3 tube, neutrons

(c)

Scintillation counter, gamma

.d)

Uncomcennated lonization chamber, neutron-and gamma

(c)

G-M tube, gamma i

.(f)

Uncompensated icnization chamber, gamma-t a

!

.

,

!

,

1 itEFERENCE:

,

Trai ning ! System. Descriptions, ' E;; core Neutron Flux Monitoring System pages 2-& 3,

' Process anc Ef. fluent Monitoring Systems pages

?.6 & 40, Area Radiation Monitoring Gyctem-pages.3 & 6

,

.

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RO4 4.01 (3.0)

While the plant is in hot shutdown you are operating the auxiliary g

feedwater system for testing.

Following the test one of the pumps needs to be valved out to complete minor maintenance.

The maintenance is completed in a few minutes and the system is returned to a normal line-up.

(a)

One of the precautions to be taken during (1.0)

feedwater additions is to maintain steam generator level above 50 % narrow range. What is the purpose of this precaution ?

(b)

When the pump is isolated for maintenance the (1.0)

manual discharge valve is recuired to be closed prior to closing the manual suction valve.

Why is this done ?

.(c)

Why is it important to shut the auxiliary (1.0)

feedwater isolation valves to steam generators

'and 2 (HV-4715 and HV-4730) after the testing is completed ?

(a)

Keeping level above 50 % ensures that the feed ring will not be uncovered.

Uncovering the feedring can lead to water hammer and damage to the steam generator and piping.

(b)

Closing the valves in this cequence prevents overpressurizing the auxiliary feed pump suction piping.

(c)

Shutting the isolation valves precludes a steam line break scenario concurrent with the loss of

,

one 125 VDC busses, which would result in feeding a ruptured steam generator.

.

'

REFERENCE:

~

Ocerating Instruction,-

0023-2-4,

' Auxiliary Feedwater System

Coeration, pages 4 and 5 1'

.

-

--

.

_..

i

,

4.02 (4.0)

While in mode 3 one of the auxiliary feed water pumps requires

maintenance.

The following questions relate to taking this pump out of service.

^

i (a)

Who has the authority to release the auxiliary (1.0)

feedpump for maintenance ?

J (b)

Who can perform or independently verify safety (1.0)

'

system alignments ?

(c)

Who must approve the repositioning of components (1.0)

(i e.

valves or breakers) ?

(d)

What are two of the three circumstances which (1.0)

eliminate the need for a

second independent

.

i verification of a velve position ?

i (. a )

An on-shift SRO can release the pump for maintenance.

(b)

Only qualified operatorc can verify system alignments.

(c)

Control operator approval is required.

hj)

0.5 for any two of the following.

,

Valve located in a posted airborne contamination

,

area.

I Verification would result in greater than

mr whole body dose.

Functional-testing

., n be safeAy done and. can prove correct alignments.

,

i

!

1.

-

,

i.

!

REFERENCE:

,.

..Operatino Instruction, 5023-0-36.

Control of System Alignments,

.

pages 3 and 8

.

-

-.

,

,-

-

...

-

4.03 (2.0)

The Unit Control' Operator comming on shift is required to discuss plant conditions with the off going Operator.

The discussion is documented by use of shift relief logs and includes five general items. List four of the five general items.

0.5 each for any four of the following general items.

Operations in progress or performed during the proceeding shifts.

Equiprent status.

Op2 rating or maintenance activities that are to

-

occur auring the next shift.

'

New or revised operating instructions.

Changes in Radiation Monitor background levels.

,

a

.

}

,

I REFERENCE:

f'

Operatir.g Instruction 5023-0-3, Unit Control Operator's Responsibilities and Duties, page 5 i

!

,

"

I

,.

,

,_

.. -.,.. -.

--

4.04 (2.0)

A waste gas decay tank release is to be made.

(a)

What individual on shift cust ce notified by the (1.0)

Radwaste plant Equipment Operaters before the release is made ?

(b)

What other individual (s) on shift should be aware (1.0)

of the release ?

'

(a)

The On-Shi't SRO Operationn Supervacar, and/or the Shi't Supcrintendent nust be no:ified.

$

(b)

~ne Unit Centrol Operator

m d

'amman

.antrol cper2 tar -runt ;t i sa be 0. ware of the reicana.

REFERENCE:

(1)

Operating Instruction 2023-0-4, Common Control Operator'u Responsibilities and Lutlea, oage 7 (2)

Operating Instruction S023-H-155, Radwante Oas Dischargo, Attachments 1 and 2

_ _ - _ _ _ - _ _ - _ _ - - - _ - - - - - - - - - - - - - - - - - - - - - - - - -

Y 4.05 (4.0)

Following a reactor trip.you are in the. process of completing the standard post trip actions.

(a)

What are the two RCS inventory control criteria ?

(1.0)

(b)

What is the RCS pressure control criteria ?

(1.0)

(c)

What-are the-three care heat removal criteria ?

(1.0)

(d)

What are the three containment isolation criteria ?

(1.0)

(a)

Pressuriner level must be controlled between 10

%

and 70-Z and RCS naturation margin must be greater than 20 F.

(b)

Pressuriner pressuriner must be controlled between 1800 and 2380 PSIA.

(c)

At least one RCP must be operating with its'

associated steam generator, core delta T must be less than.10 F, and core exit saturation margin must be greater than 20 F.

(d)

Containment pressure is less than 1.5 PSIG, containment area radiation monitors are not alarming, and steam plant radi ation monitors are not alarming.

1

.

REFERENTE:

Emergency Op ratino -Instructi on S023-12-1, Standard Post Trip Actions,'page.8

.5

.-.

- 4.06 (2.0)

- Following the plant trip you determine that the plant has had a

loss of forced circulation.

You are told to verify that natural circulation has been established in at least one loop per the

'

emergency procedure for loss of forced circulation.

What are the

,

four criteria which should.be used to verify that natural circulation has been established and is being maintained ?

I (a)

Operating loop delta T less than 59 F.

(0.5)

(b)

T(c)-and T(h) not rising (0.5)

$_

(c)

Reactcr vessel level (pl enum) preater than 82 %

(0.5)

(d)

Operating loop and REP CET within 16 F.

(0.5)

,

i

.

,

REFERENCE:

' Emergency Operating. Instruction, 5023-12-7, Loss.of Forced Circulation, page 5

--

- -

-

- -

-

- -

-

- -

.

-

.

4.07 (4.0)

A review of plant operations and shutdown margin determinations of the plant shows the following secuence of events.

0800 21 July --

plant in made 3 1200 21 July --- snutdown margin determination

'

1600 22 July --- shutdown margin determination 1400 23 Julv --- reactor critical reactor in made 2 1500 23 Julv --- shutdown margin determination 2300 22 July --- shutcown maroin determination 0400 24 July --- reactor in mode 1 1400 24 July --- shutdown margin determination Refer to the attached Technical Ececification requirements for snutdown margin determinacion 14.1.1.1.1) to answer the following questions

.

(a)

Eased on the secuence of events described above, (2.0)

descrice any Tecrnical 5pecification violations which have occured cccurec ?

(b)

When is the next snutdown margin determination due ?

(1.0)

(c)

When is the la: esc that the next shutdown margin (1.0)

determination can be mace without violating the rechnical Soecifications ?

ta)

Co 23 lu~y.-

n:..; c u n marcin cetcrmination should have been c ar g 1 ;; $ :cJ uithin 4 hourn of criticality (i e: O ct a enen 1030 nd 1400).

tb)

0200

~'5 J u l -- '

t' c )

0500 25 Jul"

.ot3 owed by 1700 23 July)

REFERENCE:

Tec hni c:t1 Specificationn l

.

.-

._-

. -.

_ _.. _

..

.. _....

... <..

_......

-

,

(

3/4.1 REACTIVITY CONTROL SYSTEuS-3/4.1.1 00 RATION CONTROL SHUTD0k'N MARGIN - Tav3_ GREATER THAN 200*F

.

LIMITING CONDITION FOR OPERATION 3.1. l.1 The SHUTDOWN MARGIN shall be graater than or equal to 5.15% delta k/k.

APDLICASILITY: MODES 1, 2*, 3 and 4.

ACTION:

With the SHUTDDWN MARGIN 1ess than 5.15% deltr. k/k, immediately initiate and

-

continue beration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN Mt.RGIN is restored.

-

SURVEILLANCE REOUIREMENTS

,,

'

4~1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 5.1E% delta k/k:

Within one hour after detection of an inoperable CEA(s) and at least a.

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.while the CEA(s) is inoperable.

If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable cr untrippable

~

CEA(s).

MI I

b.

When in MODE 1 or MODE 2 with K ss atse %an or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by vpIy.ng +phaiCE1groupwithdrawalis within the Transient ion Li s of Specification 3.1.3.6.

V

.

When in MD,E d i h K ess than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months crior to c.

critical CEs);gtq,r t+1Iality by verifying that the precicted achieving a

fosition is within the limits of Specification 3.1.3.6.

"

See 5::eciai Iest Exception 2.10.1.

.

L'

.

SAN ONOFRE-UNIT 2 3/4 1-1

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-

REAC*IVITY CCN ROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

.

- d.

Prior to initial operation above 5% RATED THERMAL P0h*ER after each fuel leading, by consideration of the factors of e. below, with the CIA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e.

When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at least the following factors:

1.

Reactor coolant system boron concentration, 2.

CEA position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and

.

6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall ye T$c:o red to predicted

. values to demonstrate agreement within + 1.0% delt#t(k*j dt le'ast once per 31 Effective Full Power Days (EFPD). TIiis com:far1Qg Thall consider at least.

.

those factors stated in Specification 4 9, above. The predicted

.

reactivity values shall be adjusted (n

.

to correspond to the actual a

core concitions prior to exceed a.fu rnup of 60 Effective Full Power

,

Days after each fuel loadin (,

,

-

..

~

.

e

.

SAN CN1FRE-UNIT 2 3/4 1-2

.

..

-

.

,

-

.

+

4.08 (2.0)

,

The following questions relate to primary coolant system leakage as defined in the Technical Specifications.

t (a)

Describe the three sources of Identified Leakage.

(1.0)

(b)

Define Unidentified Leakage.

(1.0)

I

.:

(a)

Leakage into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank:

Leakage into the containment ate.;osp here from

.

sources that are both specifically - located and

!

.known either not to interfere with'the operation

of ' leakage detection systems or not Pressure Beundary Leakage:

Reactor coolant system leakage through a

steam

.

generatcr to the secondary system.

,

it)

All leakaoe which is not Identified Leakage is Un i

identified Leakage.

3 -

,

!

t a

!

t i

!

.

>

i

,

. REFERENCE:

'

c

,,

Technical Specifications. Definitions,.pages 1-3 and 1-6 i

!

O

-

-. -

-

-..

-

- -

- -.

.

..

4.09 (2.0)

/

Regarding Radiation protection limits:

(a)

What are the quarterly Administrative limits for (1.0)

radiation doses to the whole body, extremities, and skin 7 (b)

Administrative 1y, how f ar can the quarterly limit (1.0)

for Whole Body dose be extended 7 (a)

Whole body = 900 mr.

Extremity.= 4700 mr.

Skin = 3750 mr (b)

2250 mr/Qtr.

f

,

REFERENCE:

CONGS Procedure 123-VII-4-0 Rev.

3, pages 7 and und

-

b

v t e Inspection Report - - 2008014
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2008 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 14(0) Security Operator Exams Emergency Planning Other

IR 05000730/2008014

Text

REGION V

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