Text

Exam Repts 50-373/OL-86-01 & 50-374/OL-86-01 Administered on 860602-05.Exam Results:Two Senior Reactor Operators & One Reactor Operator Passsed Exams
	ML20206S243
Person / Time
Site:	LaSalle
Issue date:	06/30/1986
From:	Burdick T, Cliff W, Sly G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20206S213	List: ML20206S207; ML20206S243;
References
	50-373-OL-86-01, 50-373-OL-86-1, 50-374-OL-86-01, 50-374-OL-86-1, NUDOCS 8607070315
	Download: ML20206S243 (54)
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r U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-373/374-OL86-01 Docket Nos. 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle Nuclear Plant, Units 1 and 2 Examination Administered At: LaSalle Nuclear Plant and Production Training Center Examination Conducted: Senior Reactor Operator and Reactor Operator Examiners:

hVb G. A. Sly 4

IM Date C Wi98Y' Date Approved By:

h44 lT.M.Burdick, Chief ((lfk Operator Licensing Section Date Examination Summary Examination administered on June 2-5, 1986 (Report No. 50-373/374-OL86-01)

Results: Written and operating examinations were administered to three senior reactor operator candidates. Section 7 of the senior reactor operator examination was administered to one additional candidate. A written and plant walk through examination was administered to a reactor operator candidate.

Two senior reactor operators and the reactor operator passed the examinations.

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PNL REPORT DETAILS

1. Examiners G. A. Sly - PNL C. W. Cliff - PhL

2. Examination Review Meeting N/A

3. Exit Meeting Discussed the facility comments for R0/SR0 written exams. No generic deficiencies were noted. R. Crawford, R. Armitage, and S. Harmon represented LaSalle; Gary Sly and Bill Cliff represented PNL.

4. Facility Comments Facility comments were to be mailed to Chief Examiner five working days after the administration of written exams. Comments by the facility and their resolution are attached.

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Question Number 1.05 Comment Question has a good concept, however, the use of the phrase " fuel pin" is extremely misleading. Ceco does not use this terminology, but more accurately refers to either the fuel pellet or the cladding. As the question is worded, it could lead the examinee to believe that the corrosion is on the outside of the pellet.

In this case, the answer would change to Increase the fuel temperature and the cladding and the coolant would remain the same. " Fuel Pin" more closely describes the actual fuel pellet so credit should be given for the above stated answers.

Resolution Comment reiterates answer provided in answer key - no action taken.

1.06.b Comments Condensate Depression is in reality a design consideration and does not change much, due to changing mass flow rate of steam being dumped to the condenser. It should also be considered that reducing steam flow would also raise vacuum by as much as 1.5 - 3.5" Hg. This change in vacuum would also tend to drop the saturation conditions in the condenser which would change condensate depression.

i Resolution s Deleted question.

1.07.b Comment I A change from 50% power to 100% power would have very little i effect on moderator temperature since a BWR operates at saturated l conditions. LaSalle even teaches that moderator changes very little, see attached Page 122 of Physics Review Lesson Plan.

Should also accept as full credit that the effect of moderator coeff. changes little from 50% to 100% power.

Resolution Comment accepted. Cue to refererce, answer key was changed accordingly. Alpha m - no change 1.08.a Comment Answer key is wrong as the question is presently worded.

l Question states that Xenon comes frcm decay of iodine-139, please see from attached Page 210 of Physict Review Lesson Plan also 3

reference chart of nuclides. The proper Iodine isotope is 135, therefore, as worded the answer False. It is the contention of the reviewers that the candidates should not be responsible for memorizing specific isotope numbers, but should be familiar with the nuclide from which the decay takes place. It would be the suggestion of the reviewers that this question be dropped or accept both True and False as full credit. Drop only Part a.

Resolution Deleted question.

1.09.b Comment Question is not clear. Is the first rod pulled in Group 7 after having pulled Groups 5 and 6? In this instance, the Group 7 rod ,

is obviously worth more due to higher flux (higher power). If the question implies that a Group 7 rod is pulled instead of a Group 5 rod, it is not clear what rod worth would do. Assuming the Rx. goes critical in Group 4 (this is actual data) then Group 5 (or if mistakenly Group 7) is still in the heating range.

Since we cannot expect operators to memorize rod sequences, there is no way to tell if this is a center rod or peripheral rod. There is not enough information available to answer this question, suggest that question be eliminated.

Resolution Deleted question.

1.10 This question is poorly worded for the concept the answer key indicates was being tested. First of all, the scenario that was made can not happen on a large power reactor. The noise experienced by the SRM period meter just due to control rod motion makes the period meter botr e erratically, there would be no way to stop rod motion at exactly infinite on the period meter.

Secondly, we deal with large rod notches, six inch increments, and have no way to " fine control" rod motion to this extent. Third, tne statement in the answer key that this would make the Rx

" exactly critical on prompt neutrons" is a false statement. This statement is the definition of promat critical, when reactivity equal to or exceeding the value of 3 is added to core, which wouldcauseanextremelyrapidunconN11ablepowerrise. Please see attached Pages97-112 of Rx Physics Review Lesson Plan.

Fourth, the c Rfept implied that while the rod y moving in, it is absorbing o from the short lived delayed o prgjursors,but afterrodmotionstopstherodwillnotjbsorbtheo 's now emitted from the longer lived delayed o n precursor. This is essentially " black" to all thermal neutrons. If the rod was inserted to drive a Supercritical Rx to a subcritical condition by inserting enough-reactivity to overcome the positive period, then it m_ay be argued that the Rx would end up subcritical, '

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" exactly" critical or supercritical. Strongly suggest that this question be removed from both SRO and RO exam. A completely hypothetical question that can not be proven without access to a l test Rx with sensitive metering and fine rod motion control.

The concept of the " latent effect of delayed neutrons" can be explained on a scram since all rods are in, explaining why power '

doesn't immediately drop to zero. Trying to tie this concept to exactly critical is confusing at best and far enough removed from the "real world" concepts of understanding the practical aspects of delayed neutrons to make the reviewers ask that this question be removed.

Resolution Deleted question.

1.11 Reviewers believe that the question is good in thought provoking concepts, but would suggest that answer key accept both Range 1 and Range 2 of the IRM's. Operational limits could be included to mean the procedural limits of LOP-NR-02 (see attached Page 3)-

which state maintain indication between 25 and 75 on the even rages and between 8 and 24 on the odd ranges. Since 33.5 is above the procedural limitation of 24/125 s of scale, credit should be given for saying Range 2 as well as Range 1.

Resolution Comment accepted if comment was not clarified for candidate.

Answer key modified to consider Range 1 or Range. 2.

2.03 Comment S.D.V. hi level bypass switch is a single switch not plural as ,

indicated in the answer key. '

Resolution Comment accepted. The "es" was dropped from the word switches..

2.07 This question asked how gases are removed from the primary containment. Full credit should be given for saying venting either by Standby Gas or Primary Containment Ventilation..

Resolution Comment accepted. Answer key changed to require venting from SBGT and Primary Containment Ventilation if radiation' levels-are low enough.

2.10.a Comment Full credit should be given for just saying the sensors are in the exhaust hoods.

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Resolution Comment accepted.

2.10.b Comment We do not stress to operators to memorize the opening setpoint of the turbine exhaust hood spray valve. Please see attached annunciator procedure for this valve. The alarm would come in as soon as the valve opens and operators are trained to use the annunciator procedure for actions and setpoints. It is not unreasonable to expect the operator to know the turbine trip setpoint since this would be a major component trip and plant transient. In light of this, the reviewers believe that full credit should be given for the turbine trip OR the point value of the question be reduced by .5 and the regiiirement for the 120 F setpoint be deleted.

Resolution Comment on turbine trip accepted. No change to answer key.

Comment on hood spray initiation accepted if candidate refers to initiation of exhaust hood spray. Exact setpoint not required for full credit.

2.11 No problem with the question, good operational oriented concept.

, The reviewers believe the answer key should also accept the start of the D/G cooling water pump. This is a major piece of equip-ment that starts every time the D/G starts. Please see attached drawing from lesson plan, this was also verified by Schematic 1E-1-4223AR.

Resolution Commend denied. The D/G cooling water pump start is independent of HPCS operation and as per facility comment

" starts every time the D/G starts."

3.01.c Question is confusing - If the '0' D/G was closed on to U1 before the ECCS condition on U-1, (i.e., per the LOS) then when U1 gets an ECCS the '0' D/G output breaker opens and the D/G will be run-ning unloaded. When the U2 bus experiences an undervoltage and an ECCS, then the D/G will load onto Unit 2. This would make the answer key wrong. At least one examinee asked for clarification on this and was told that the D/G was in fact manually closed on unit one before the ECCS, as per above scenario. Since this clarification was not given to all examinees, the question could be interpreted to mean that the 'O' D/G output breaker was closed onto Unit 1 after the ECCS condition occurred on Unit 1. In this case, the answer key is correct in stating the 'O' D/G would remain loaded on the Unit 1 bus. Since the clarification was given that contradicts the answer key, full credit must be given 6

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for saying the '0' D/G would close on the Unit 2 bus and that the

'O' D/G would remain loaded on the Unit 1 bus.

Reference 1E-04121AA and AB - attached.

Resolution Comment accepted. Since clarification given to candidate was consistent with working of question, answer key was changed to reflect correct answer " trips on Unit 1, closes in on Unit 2."

3.02.b Comment RWCU isolation of the (F001), the answer key should reflect that loss of power to the leak detection system is also an acceptable answer, reference LaSalle License System Description, Chapter 49, Pages 20 and 21 (attached).

Resolution Comment accepted. " Loss of power to the leak detection system" was added to the answer key.

3.02.c Comment Although the question asked for loss of service air, LaSalle has the service air and instrument air systems cross-tied. All scenarios at LaSalle are taught and discussed simply as a loss of air. Since instrument air also feeds the air operated valves on

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the RWCU system and these valves all fail closed on loss of air -

credit should be given if examinee states the RWCU flow would be lost through the filters which would trip the RWCU system pumps.

Resolution Comment accepted. Due to the lack.of a instrument air / service air isolation valve, the answer key was modified to include 4

isolation of filter demineralizer, loss of RWCU flow, and subsequent isolation.

3.03.c Comment Answer key should not require reference to the min. flow valve ,

for full credit. Could also expect to see comments concerning RCIC turbine trip due to overspeed, this would be a "real world" :

i situation and a true statement. For the individual to know that the same AP cell feeds the flow controller and the min. flow

control is asking the individual to memorize the P&ID's and C&ID's.

Resolution <

Comment on minimum flow valve accepted and reference to it l was deleted from the answer key. Answer key was changed to i accept that the "RCIC turbine trips due to overspeed" rather-7 i

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than " injects at maximum flow rate" due to the speed control setpoint dictating the maximum _ rate of injection as per Figure 41-5, RCIC Logic.

3.04.a Comment

1. Answer key should also accept Gen. frequency or D/G speed, they are the same. (LaSalle exam bank accepts either statement).

Resolution Comment accepted. Generator frequency is considered the same as generator speed. The answer key was modified.

3.06.a Comment There was a recent modification done to Unit 1 only, M-1-1-84-036, on the ADS system. If the. question is answered per the new modification, then the answer key should reflect that Manual ADS via the arm and depress pushbuttons, no longer requires a divisional low pressure pump running for the valves to open.

Reference attached Required Reading package, Pages 13, 14, and 15.

Clarification was given to one SR0 to answer the question as before the modification, since the clarification was not given to all examinees the answer key must reflect responses to either before or after the modification.

Resolution Comment accepted. Modification was received at facility.

The answer key was modified to accept either answer if properly supported.

3.08.b Comment The question only asked for control and bypass valve positions, however, the answer key refers to possible Rx scram and other actions. This should not be required for full credit. This statement also applies to Part d. of this question.

Resolution Comment accepted. Answer key modified to delete the reference to Rx scram for full credit.

3.08.c Comment This question is confusing as to what actually fails. The question is worded "'A' pressure regulatory (transmitter) failed low.," this is not clear if it is the regulator that failed low or the steam line pressure transmitter tlat failed low. Please 8

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reference the attached copy of LOA-EH-01, "EHC Pressure Regulation Malfunction." As you can see by Step 2 under Symptoms, the regulator failing downscale is the same transient as the transmitter failing upscale. What this means is, if the examinee answered as if the regulator failed downscale, then the answer would be just opposite that of the answer key. Applying the EHC electrical diagram attached at the back you can see .that if the pressure regulator fails downscale (the 920# setpoint fails to 0) then there would now be a 950# signal from the steam pressure transmitter and a '0' signal from the pressure regulator causing a large (950#) signal to be passed through the high value gate (hvg) and the percent flow gain unit. This large signal would be passed on to the low value gate (lvg), where it is compared to the 105% signal from load set. The 105% signal would pass and tell the control valves to open to 100%. The large open signal would have also been sent to the bypass valve demand circuit. At the LVG in the bypass circuit the 105% going to the control valves would be compared to the large open signal from the bypass circuit. The end result would be the control valves being full (100%) open and the bypass valves open five percent.

Due to the confusion factor involved, it would be reasonable to accept the above answer as correct. This would still show that the examinee knows how the circuitry works and understands the concepts even though he started with an opposite failure. The reviewers believe that both answers should be accepted for full credit.

Resolution Comment accepted, if properly supported with explanation.

3.09.a Comment The reviewers believe that this question is not operationally significant. Due to vendor recommendations, operationally the station air compressors are never operated in the modulate plus twc mode of operation. They are left in the modulate mode. If question is not discarded then credit should be given for discussing the possibility of surging or improper compressor loading during low air demand conditions.

Resolution Comment accepted. Answer key was modified to give credit for the discussion of surges and loading concerns.

3.10.c Comment Manual scram could also be by placing the mode switch to shutdown position - in this case, the scram is bypassed for ten seconds. Reviewers believe this is acceptable answer also.

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Resolution Comment accepted. Because the 10 second TD is not a fixed bypass, no credit was given or taken for its inclusion or non-inclusion. No action on answer key.

3.11.a Comment This is only a general comment, no action required. As can be seen by the attached NED0-21231, Page 3-1 and LaSalle License System Description RSCS, Chapter 19, Page 7, the limits specified as N1-N3 and M1 are not set limits and are subject to change.

We do not expect operators to memorize rod withdrawal limits, all rod pulls are done in accordance with approved rod sequence packages.

Resolution No action. Facility provides only generic comment, not a recommendation.

4.01.b Comment The way this question was worded may have led the examinee to put down the immediate actions out of the Scram Procedure LGP 3-2.

This should be acceptable since it is still testing the examinee on procedural knowledge. Please find attached, a copy of the LGP 3-2, the reviewers believe any steps listed under the procedure portion of LGP 3-2 should be acceptable for full credit.

Resolution Comment accepted. Answer key was modified to include only those procedural steps that dealt with the verification of the scram, not scram procedural steps.

4.02.b Comment The answer key states the recirculation mismatch is not allowed due to jet pump vibration and uneven core flow. Our Technical Specification Bases section for this mismatch discusses only the effect on recirculation pump coastdown during a LOCA condition.

Therefore, this answer should be accepted for full credit.

Please find attached the appropriate page from our TS Bases.

Resolution Comment accepted. Answer key was modified to include the effect of recirculation pump coastdown as per TS Bases 3/4.4.1.

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r 4.03 Comment Considering that these actions fall into subsequent Operator Actions (LOA-NB-09), should not require five methods. Three (3) of five (5) would be more realistic.

Method 5 on the answer key should also accept for full credit if the examinee says venting the control rod. This is the same as removing "overpiston" pressure.

Resolution Comment denied due to the safety implications for reacter safety. Facility coment to Method 5 was taken into consideration during grading.

4.04.a Coment Reference to LPA-GP, General Precautions is wrong in the question.

The LGA's have the General Precautions which could confuse the examinee, so credit should be given if the examinee discusses that conditions are stable and under control, or when continued operation would worsen the plant conditions. Action should be taken after review and approval of SRO imediately available.

These actions are discussed in LAP 1600-2, Conduct of Operations.

4.04.b Full credit should also be given if the examinee discusses monitoring relevant parameters by a licensed operator to assure safe operation of the plant as precautions, while the system / ,

component is in manual control. This action is implied in the LGA Precaution No. 11, and actually discussed in LAP-1600-2,_ :

Conduct of Operations.

Resolution Comment accepted. Answer key was modified to incorporate facility coments.

4.05.a Coment The technical specifications concerning idle RR loop startup list only the 145*F ^T, and the two 50*F ^T's. The reviewers believe that the answer key should not include the 100*F/hr heat up rate on the Rx and RR loop since that is a different TS. We suggest dropping the 100* limit and reducing the question point value by the appropriate amount. Please find attached appropriate page from the technical specifications. The answer key references .

LOA-RR-03, this is incorrect procedure number. '

Resolution Coment accepted. Part 3 of Part a. was deleted and points reassigned to Parts 1 and 2 of Part a.

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3 4.06.a Comment

1. If the examinee talks about a Type II Radiation Work Permit, then full credit should be given for stating four hours to go to the higher limit on a Type II RWP.

2. If student assumes current Form 4 on file, then full credit should be given for stating 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months . Note the R0 examinee was told by examiner to assume a Form 4 already on file, so that test will not show the assumption.

Resolution Comment accepted. Answer key was modified.

4.07.c Comment Full credit should be granted for only stating Moderator or Coolant temperature and Mode Switch position. The setpoint of 212 F not required for full credit.

Resolution Comment accepted. No action to answer key.

4.08.b Comment Please find attached copy of LOA-NB-02 that lists all the symptoms for a stuck open SRV. The answer key only lists two parameters. The reviewers believe that any of the parameters listed under the symptoms should be accepted for full credit.

Resolution The answer key was altered to accept any other reasonable responses.

4.09.b Comment Since this question referenced a step (c.8.a) in a procedure that was not given to the examinee and it was stated this step is to "Depressurize the RPV," it is logical to assume this was an ADS step. If this assumption was made by the examinee, then the General Precautions state that if ADS has initiated do not stop it, but allow it to go to completion. This would mean not going back to the beginning of the LGA. The reviewers believe that this response should also be accepted for full credit.

Resolution Comment accepted. Candidate must support facility comment withproperjustification.

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4.10.b Coment Since the policy at LaSalle is to not allow anyone to enter the plant buildings (i.e., leave the service building) without signing all the generic RWP's, an acceptable answer to this question would be "any time a person enters the plant."

Resolution Coment denied. Answer key was not modified.

4.11.a Coment Generic coment again asking R0 to memorize a technical specification action statement.

Resolution No action. Generic coment on question content.

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5.01 Comment This question makes reference to the Core Thermal Hydraulics section which addresses " CRITICAL POWER." The question takes the CRITICAL POWER concept one step further to ask about the MINIMUM CRITICAL POWER RATIO (MCPR). This was probably done to make the question a bit more realistic. Numerous complications are encountered, however, when concepts of CRITICAL POWER are directly applied to OPERATING MCPR LIMITS. i.e.:

CRITICAL POWER is determined in a test environment where the heat transfer elements are electrically heated. The water functions solely as a coolant (heat transfer medium).

However, in the plant water functions BOTH as a coolant which removes heat AND as a moderator which controls the power level thus heat transfer rate. The dual function of water in a reactor complicates simple conceptual accuracy in predicting the effects of a parameter change on MCPR.

The actual MCPR operating limit changes whenever core flow is less than rated 100% flow. The limit INCREASES or becomes more conservative due to thefK multiplier. (SEE Figure 20, Page 70 of Core Thermal Hydraulics L.P.)

When dealing with MCPR in an operating power reactor, automatic system responses may counteract events that might change primary / secondary system conditions.

Depending upon how the examinee interprets conditions, limits, and systemic interactions, the answers may differ even though logical, correct thought processes were used.

5.01.a Comment If the decrease in Rx power due to adding heaters is igaored or judged to be insignificant, the answer " closer to" is correct.

However, the part of the key which is in parentheses "(increase CP)" is incorrect. Given the above, addition of a heater string will decrease subcooling, thus decrease CP.

If, however, the decrease in reactor power due to warmer feedwater (docrease in subcooling) is judged to override the decrease in CP, the examinee might logically predict that the addition of a heater string will put the core farther away from the MCPR limit. There is some support for this line of thinking in that a loss of FW Heaters will decrease actual CPR. See FSAR Transient Analysis.

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5.01.b Comment As flow in the test facility decreased, CP decreased. However, it is complicated in a nuclear reactor because of the moderating effect of water and the K, multiplier which adjusts the MCPR limit flow less than rated flow conditions.

If the decrease in Reactor Power was judged to be MORE significant than the decrease in CRITICAL POWER, then the answer

" FARTHER AWAY" is correct in that actual MCPR would decrease.

However, as core flow changes, so too does the MCPR limit due to the K multiplier. The K multiplier would tend to keep the distance 7from the MCPR limit fapproximately the same. Since the distance from the MCPR limit was asked for, UNCHANGED should be allowed for full credit if question is not deleted. If Rx power is judged to be less significant than CRITICAL POWER, then explanation could be justified to get " CLOSER T0."

5.01.c Comment If it is assumed that reactor pressure decreases and stabilizes at a lower value, then CRITICAL POWER would increase causing the Rx to be " FARTHER AWAY" instead of " CLOSER T0" the MCPR limit.

However, when an SRV opens, EHC will throttle the control valves to maintain Rx pressure constant (SEE EHC Schematic) thus the effect on MCPR limit would be UNCHANGED.

SUMMARY

5.01 is confusing as evidenced by conflicting answers in exam key. Reviewers suggest that question be deleted, or else credit be given for any logical direction.

Reference LaSalle Thermal Hydraulics - additionally Pages 68 and 70, LSCS - UFSAR, Table 15.0-2.

Resolution Deleted question 5.04 Comment Realistically, the leaks size would have a greater effect on appearance of vapor or liquid with a leak of this type (smaller leaks being rapidly condensed to liquid).

Because of conditional variables involved, full credit should be allowed for selecting correct choice (saturated steam / liquid leak) supported with a reasonable explanation.

Calculations should not be required.

Q5.04 Resolution I Comment denied. Answer key was modified to include that "the enthalpy of the liquid (544 btu /lbm- F) is between saturated liquid / vapor for a 14.7 psig."

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5.05 See R0 Comment 1.05 5.06 See R0 Comment 1.06 5.07 See R0 Comment 1.07 5.08 See R0 Comment 1.08 5.09 See R0 Comment 1.09 5.10 See R0 Comment 1.10 ,

5.11 See R0 Comment 1.11 SECTION 6 6.01 Comment The control room has indication of main steam flow which comes from the restricting orifice so the credit should be allowed if this is listed as an interface.

The question states that reactor power is at 60% and ask for systems with direct interface. At this high power level, the examinee may not list RWM as an answer due to the fact RWM is not used above 30% power. Also, the question states only one orifice fails which may further confuse the examinee. Based on this confusion listing, three of five would be appropriate.

The five choices being RWLC, Process Computer, RWM, PCIS, and Control Room Indication.

Reference LaSalle System Description, Chapters 2, 18, and 21.

Resolution Comment accepted. " Control room indication" was added to the answer key with only four (4) required for full credit.

6.03 SEE R0 Comment 2.07 6.04.a Comment Credit should not be subtracted if examinee mentions that when SBLC is used in conjunction with the ATWS LGA's that the reducing power concern overrides the chugging concern, and when possible two SBLC pumps will be run.

Reference LPLGA-ATWS Lesson Plan.

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4 Resolution 4

Commer.t accepted; no action taken.

6.04.b Full credit should be given for knowing the reasons for increasing the boron concentration from 660 to 1075 ppe. Numerical _ values for the change are not important to understanding the design bases and operation of the SBLC system, therefore, they should 4

not be required for full credit to answer the question.

Resolution

! Comment not accepted. SR0 candidate should have a feel for i the approximate amounts of additional boron necessary to override positive reactivity _effect. Exact numbers are not required but an understanding of the relative amounts is necessary for full credit.

6.07 SEE R0 Comment 3.01 6.08 SEE R0 Comment 3.03 l

6.10 SEE R0 Comment 3.06 6.12 SEE R0 Comment 3.08 l

6.13 SEE R0 Comment 3.09 6.14.b Comment With the flow control valve operating in AUTOMATIC, the system flow and cooling flow should not change by closing down the drive pressure control valve to adjust drive water pressure. The flow control valve which is upstream of the drive water pressure control valve will sense the flow being decreased by the

, throttling of the drive water pressure control valve and open to maintain flow constant. If the examinee assumes the flow control valve is in manual, then credit should be given for answering the total system flow and cooling flow will decrease.

Reference LaSalle System Description, Chapter 8, Figure 8-2.

l Resolution Comment not accepted. Theifacility comment that " total .

system flow would not change"_is correct; but, due to the .

increase in system pressure, scram valve charging flow would increase and cooling flow would not change. The answer key has been modified accordingly. Automatic operation would have to be assumed to'be correct unless otherwise-stated by candidate.

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SECTION 7 7.01 SEE R0 Comment 4,01 7.02 Comment Full credit should be given if the examinee discusses open RPS breakers at the RPS distribution panel. The breaker numbers listed in the procedure are marked in the cabinet. Also, this is a step in the subsequent actions of the LOA's which are not required knowledge.

Reference LOA-RX-01 Resolution Comment accepted; no action taken. It should be reitercated that although abnormal procedure subsequent action needs not be memorized. The ability to discuss them is required knowledge.

7.03 SEE R0 Comment 4.03 7.05 SEE R0 4.08 7.07 SEE R0 Comment 4.05 7.08.a Comment The correct answer for the condition given is -11.5 feet.

Reference LGA-GS.

Resolution Comment accepted. The answer key was modified.

7.10 Comment Answer given is no longer correct. LGA-ATWS-01 now reads that if Suppression Pool Temperature and Rx Power reach the limit established by the Boron Initiation Temperature curve, that SBLC shall be initiated.

F.eference LGA-ATWS-01 Resolution Comment accepted. The answer key was modified.

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l 7.12 Comment ,

The intent of ALARA is to reduce the total dose to the Radiation Work Group without putting individuals at excessive risk. Since Group 1 accomplishes both goals, it is more in tune with ALARA.

However, many assumptions could be stated which would justify either answer as correct, (i.e., if one individual is closer to his administrative exposure limit, dose leveling becomes a strongerobjective).

Answer to this type of question is quite subjective, and depends on many possible conditions.

Resolution N Comment accepted. The answer key was wrong and the correct answer is Group 1, due to minimization of total dose.

SECTION 8 8.02 Same as R0 Question 4.04. See comments on R0 Review.

8.04 Comment Clarification provided during the exam for two different candidates, led each candidate to a different conclusion concerning the extent of Scram Discharge Volume Operability.

Credit should be given for answers in response to action for RPS instrument level switch failure (in accordance with

3.3.1-1, Action No. 1), as well as a total loss of Scram Discharge Volume Operability.

Resolution Comment accepted. The answer key was modified. The

. question should be reworded to alleviate future problems.

8.06.a Comment Question relates generic GSEP terminology (non-station specific).

LaSalle terminology for " Emergency Director" is Station Director or Acting Station Director. This individual is normally the Shift Engineer or other qualified on-shift Acting Station Director.

8.06.b Comment Station specific terminology needs to be reflected in the answer key. The Acting Station Director (normally Shift Engineer) is later relieved by the Station Director; normally the Station Manager or other qualified Station Director.

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Reference LZP 1110-1, Page 3 Resolution Comment accepted. The answer key was modified. No GSEP manual was provided in training material.

8.07.a Comment Answer key references both LAP 1600-2 and Technical Specification 6.1-3. Technical Specification Table 6.1-3, Pages 6-13 and 14 permit one Shift Foreman as a minimum requirement. Station procedural requirements are for two Shift Foremen which exceeds technical specification requirements.

Since question asks for technical specification requirement, answer key should reflect only one Shift Foreman.

Resolution Comment accepted. The answer key was modified. Only "1 S.F. required as per TS."

8.07.c Comment Bulk of credit should be allowed if candidates demonstrate an understanding this condition would result in a violation of technical specifications because technical specifications requires a full compliment of crew manning. Details of the exception to this rule need not be committed to memory.

Reference TS 6.1-3, Pages 6-13, 14.

Resolution Comment not accepted; no action taken.

8.09 Comment Technical Specification, Page 3/4 3-3, which was provided to the candidate, shows an instrumentation operability action matrix.

Action steps are number coded and specific action steps are delineated on Page 3/4 3-4 which was not provided in reference material.

Some confusion exists among candidates concerning the extent of detail required for a correct answer in response to technical specification questions relating to reference material provided.

During the pre-exam briefing (by the examiner) candidates were instructed not to write out full wording for a technical specification item, but to reference the correct number, thus indicating an understanding of which specification applied.

With this understanding, a correct answer should be considered Action No. 7 from Table 3.3.1-1.

20

Reference TS Table 3.3.1-1, Page 3/4.3-3, 4 Resolution Comment not accepted. During initial briefing, candidates were directed to provide a brief synopsis of actions. This being a.one hour action item lends itself to a testable knowledge and is required for full credit.

8.10.b Comment Full credit should be allowed for an explanation that states "the minimum operable channels per trip system is not met because three are required by technical specifications and only two are available in Trip System A."

Resolution Comment not accepted. Facility comment does not address specific answer to question but the content of the facility response if already included in the answer key. No action taken.

8.11.b&c. Comment Change made to Part b.and c during exam to correct division numbers.

Resolution Comment accepted. The answer key has been modified.

8.11.c Comment Full credit should be allowed for answers referring to action required by 3.0.5. Division 2 does not have its emergency power supply and Division 1 LPCI is a redundant system. Therefore, action required based on 3.0.5. is; Within two hours action initiated to place the unit in; (1) at least startup within next six hours.

(2) at least Hot Shutdown within following six hours.

and (3) at least Cold Shutdown within subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Reference TS 3/4 0-1 Resolution Comment accepted. out candidate must refer to "the intent of the procedure has not changed" for full credit. Answer key has been changed to show facility comments.

21

8.12 Comment Question is confusing. It was originally written referencing LAP-240-6 which is Temporary System Change Procedure. During the exam, the procedure title was changed to " Temporary Procedure Change" but the Reference LAP-240-6 stayed. Temporary Procedure Change is LAP-820-4. This confusion factor coupled with the report that not all examinees were informed of the change justifies giving credit for either the correct answer for LAP-240-6, Temporary System Change or for LAP-820-4, Temporary Procedure Change.

Correct answer for LAP-240-6:

1. Safety evaluation (50.59) review complete.

2. Change reviewed by two SR0's, one with an engineering degree.

3. Authorized by the Shift Engineer.

Correct answer for LAP-820-4:

1. Reviewed by two SR0's including the

2. Shift Supervisor NOTE: The answer on the key gives 0.5 points credit for "the intent has not changed." The review by the SR0's is to verify that the intent has not changed. The question asks for what requirements must be met. The review answers this question.

Reference LAP-820-4, Attachment A.

Resolution Comment accepted. But candidate must refer to "the intent of the procedure has not changed" for full credit. Answer key has been changed to show facility comments.

8.13.a Comment Answer key references the wrong technical specification (i.e.,

ECCS Specification 3.5.1). This is not the most limiting specification for this condition. The most limiting specifi-cation is 3.6.2.3, Suppression Pool Cooling which states; "with one suppression pool cooling loop inoperable, restore to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

8.13.a&b Comment Time required to Cold Shutdown should not be required to fully answer the question asked (maximum time for continued operation J

22 l

[.

in Condition 1). l Reference TS, Page 3/4 6-21.

Resolution l l

Coment accepted. The answer key has been modified.

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1 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: LASALLE 1 REACTOR TYPE: BWR-GE5 DATE ADMINISTERED: 86/06/03 EXAMINER: SLY, G.

APPLICANT: Maf/es .6[O INSTRUCTIONS TO APPLICANT:

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

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF APPLICANT'S CATEGORY

_VALUE TOTAL SCORE VALUE CATEGORY

/'f6

-25.no VW-lOf 25.00 5. THEORY OF NUCLEAR POWER PLANT OD kM 25.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 f 25.N(0 00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL

{f' CONTROL 25.00 25.00 8. ADMINISTRATIVE PROCEDURES,

- CONDITIONS, AND LIMITATIONS h

a nn nn 100.00 TOTALS FINAL GRADE %

All work done on this examination is my own. I have neither given nor received aid.

APPLICANT'S SIGNATURE

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

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

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

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

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one Ude of the paper, and write "Last Page" on the last answer sheet.

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

10. Skip 'at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations orely if they are commonly used in facility literature.

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

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

l l

1

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc. l' (3) Answer pages including figures which are a part of the answer.

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

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

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

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_ . , . . - ~ ^ '

h.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 THERMODYNAMICS QUESTION 5.01 (1.50) h f

)L i

LaSalle Technical Specifications has established MCPR limits t preclude the onset of transition boiling. This limit can effected i by a number of variables. For each event listed bel ATE whether an operating facility would be operating (CLOSE , FARTHER AWAY FROM, or UNCHANGED with respect to) their ned MCPR limit. ,

S a. Addition of a feedwate ater string. (0.5) 0" b. Recirculatio mp decrease from 100% flow to 90% flow. (0.5) 6i c. Sta reactor pressure following an inadvertent opening an SRV (no scram and SRV remains open). (0.5)

% Ett%

QUESTION 5.02 (2.00)

Concerning General Electric's Preconditioning Interim Operating Management Recommendations (PCIOMR):

a. According to LGP 3-1, the PCIOMR threshold power level is h, reduced at a fuel exposure of 4.5 GWD/T from 11 kW/ft to 8 kW/ft. This reduction is primarily due to (SELECT one):

l /e (1.0)

1. Burnup of U-235 causing a reduction in nodal power.

2. Increased embrittlement of the fuel rod inner surface.

3. Increased ductility of the cladding due to neutron irradiation.

4. Fuel pellet densification increasing the fuel rod helium gap and inner cladding surface stress.

b. Starting with the fuel at a threshold of 11.0 kW/ft, a maximum ramp increase is begun at time 0000 and the final desired power is to be 13.0 kW/ft. SELECT the minimum time which would be required to raise power to 13.0 kW/ft, assuming the maximum ramp rate is used. (Ifanswerisnot exact, select next highest value). (1.0)

1. 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

2. 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months

3. 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months

4. 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months

(***** CATEGORY 05CONTINUEDONNEXTPAGE*****)

[

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 THERMODYNAMICS i QUESTION 5.03 (1.00)

During the last refueling outage a center bundle was inadvertently positioned in a peripheral bundle location. The reactor was then brought to 100% power. FILL-in-the-blanks with (LARGER THAN, SMALLER THAN or THE SAME AS) to indicate the correct response for the thermal hydraulic conditions in the misplaced bundle.

4

a. The misplaced bundle flow rate will be the adjacent peripheral bundles. (0.5)

b. The power level in the misplaced bundle will be if the bundle had been placed in its correct central location. (0.5)

QUESTION 5.04 (2.00)

During your shift, a leak develops in the outboard isolation valve of the RWCU system inlet piping. Given that the reactor is operating at 100% power condition (1000 psia), CALCULATE whether you would direct your in-plant personnel to: (SHOWyourworkand assumptions.) (2.0)

a. look for a superheated steam leak.

b. look for a saturated liquid leak.

c. look for a saturated steam leak.

d. look for a saturated steam / liquid leak.

QUESTION 5.05 (2.25)

A fuel in, over a period of time, has a uniform coating of corrosion products out 0.001 inches thick on its surface. Assuming that power generation within the fuel pin remains constant during the time of the buildup, would you expect the following temperatures to (INCREASE, DECREASE, or REMAIN THE SAME) during the buildup? EXPLAIN each answer.

a. Fuel temperature. (0.75)

b. Cladding temperature. (0.75)

c. Coolant temperature surrounding the lower portion of the fuel pin (prior to the onset of boiling). (0.75)

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

l.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 THERMODYNAMICS j l

l QUESTION 5.06 (2.25)

Following a normal reduction in power from 90% to 70% with recirculation flow, HOW will each of the following change (INCREASE, DECREASE, or REMAIN THE SAME) AND WHY?

a. The pressure difference between the reactor and the geq h turbine steam chest. (0.75)

_ Ccaden:ctc d; pre 55ien et the cxit ef the cv..denscr. (G.75)

~~

c. Feedwater temperature. (0.75)

QUESTION 5.07 (1.50)

For the situations listed below STATE HOW each of the reactivity coefficients (void, moderator, doppler) would respond. (i.e. more negative, less negative, no change).

a. E0L versus BOL. (0.75)

b. 50% versus 100% power operation (0.75) f i I I

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

I.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 THERMODYNAMICS 1

QUESTION 5.08 (2.50)

STATE whether the following statements concerning fission poisons are T UE or FALSE.

The laraest nrnd e t hr. a n iun vi unun is tne 9

(Q a.

p aoloactive decay of Iodin (0.5)

b. A 25% power reduction from 100% power would have a larger xenon peak than a 25% power reduction from 50% power. (0.5)

c. A rapid reactor shutdown from 100% power would have the same resultant xenon peak as a reactor scram from 100%

power. (0.5)

d. Since the production and removal of samarium is a direct function of thermal flux, a reactor that has operated at M

k\

ONLY 50% capacity will have the same equilibrium samarium concentration than one that has operated at 100% capacity. (0.5)

e. Upon restarting the reactor following a 6-month outage, the samarium concentration will decrease to its 100% full power concentration. (0.5)

QUESTION 5.09 (3.00)

STATE WHY the following situations would or would not change differential control rod worth. (Also INDICATE whether rod worth would INCREASE, DECREASE, or NOT CHANGE).

[y a. A rod is withdrawn from notch 06 to notch 10. (0.75)

. Pulling the first rod in non r,rnon 7 inetaad nf the first

, fu in KUD Group 5. (0.7'5)

c. Localized voiding of a region not previously voided. (0.75)

d. A change of the size of the control rod reducing the surface area while maintaining the same boron volume. (0.75) i 1

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

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 5.

THERMODYNAMICS

]

QU :STION 5.10 (2.00)

The reactor is critical at 10E+6 cps. A eriod of 60 seconds is achieved. If rods are insert nuously until the period ,

O drops to infinity and t e rod insertion is immediately stopped,

.\ WILL the reactor RITICAL, SUPERCRITICAL, or SUBCRITICAL) in the time foll the rod stoppage? EXPLAIN. (2.0)

QUESTION 5.11 (2.00)

Five (5) minutes following a reactor scram from 100% power, reactor g power is 15 on IRM Range 4 and decreasing. WHAT is the minimum IRM 10 \ Range that you could go to two (2) minutes later without violating U any operational limits? SHOW calculation and EXPLAIN any assumptions made. (2.0)

QUESTION 5.12 (2.00)

The reactor is suberitical with a Keff of 0.95 and a SRM countrate of 200 cps. The control rods are withdrawn and the new countrate is 300 cps.

{ a. HOW much reactivity was added? (1.0)

b. If the same amount of reactivity, determined in Part a.,

were added again, would the new stable countrate be (LARGER THAN, SMALLER THAN, or EQUAL TO) 450 cps? Give a justification for your response. (I.0) 1 I

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)

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

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 5.

, THERMODYNAMICS QUESTION 5.13 (1.00)

You are currently operating at 100% power BOL when you lose partial feedwater heating. If the SCRE tells you that reactor ,

coolant temperature decreased by 10 deg. F, voids decreased by 2%,

WHAT would be the corresponding temperature change to the fuel temperature? (ASSUME no rod movement, recirculation flow changes k and the reactor reactivity returns to zero.) (1.0) l yr Jh, 1. increase by 30 deg. F j 2. decrease by 30 deg. F I 3. increase by 300 deg. F

4. decrease by 300 deg. F

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

7 I_ ._ _ . _

O

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 1

QUESTION 6.01 (2.00)

While operating at 60% power, the pressure tap on the MSIV flow

@y' restricting orifice fails low on one (1) main steamline. For this 9,

f p, jg6 situation WHAT four (4) interfacing systems would be directly effected from the loss of signal? (2.0)

QUESTION 6.02 (2.00) p/ LIST the four (4) signals which will cause an automatic recirculation pump to downshift from fast speed. (Actionswhich

{}h g4 cause a unique signal condition are considered to be one signal.) (2.0)

QUESTION 6.03 (1.50)

DESCRIBE HOW combustible gases are removed from the primary containment in the event that both hydrogen recombiners are

,\%p inoperative following a LOCA event? (1.5) v QUESTION 6.04 (2.50)

The Standby Liquid Control System was designed around both injection j time and boron concentration limitations.

t\

a.

WHAT injectionistime the basis of having(a limitation? PROVIDEminimum as well the values as a maximum (1.5) for each.)

b. If the maximum amount of boron concentration necessary to shutdown the reactor from a full power condition is 660 ppm, WHY is the total design concentration 1075 ppm?

(Include numerical values). (1.0)

, (***** CATEGORY 06CONTINUEDONNEXTPAGE*****)

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 QUESTION 6.05 (1.00)

Concerning the relief valve low-low (LLS) set function:

y a. HOW does the LLS perform its purpose? (0.5)

C

b. DOES the LLS actuate on manual operation of the relief valves, automatic operation of the relief valves, or on either one?. (0.5)

QUESTION 6.06 (1.50)

The Inboard Shutdown Cooling Containment Isolation Valve (E12-F009) is normally powered from a Division II switchgear. An alternate power supply is provided from Division I. Concerning using the alternate power supply, INDICATE whether the following precautions are TRUE or FALSE. (1.5) y a. The Outboard Shutdown Cooling Isolation Valve (E12-F008)

O must be closed when using the Inboard Valves alternate power supply,

b. There is no protective circuitry to prevent feeding the valve motor operator from both the normal and alternate power supplies.

c. The F009 valve will NOT automatically isolate when the valve is powered from its alternate power source.

QUESTION 6.07 (1.50)

For the following situations STATE the actions of the Diesel Generator "0" output breaker if:

a. there is an ECCS condition on Unit 1, and a simultaneous ECCS and UV condition occurs on Unit 2. (0.5)

' JJ b. it is closed onto Unit 2 (241Y), as a result of

,hi undervoltage on 241Y, and an ECCS initiation occurs on Unit 1. (0.5) r (c it is manually closed on Unit 1 (141Y), an ECCS initiation occurs on Unit 1, and an ECCS and UV condition is sensed on Unit 2. (0.5) l l (***** CATEGORY 06CONTINUEDONNEXTPAGE*****)

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 i

QUESTION 6.08 (2.25)

ASSUME the RCIC system receives an initiation signal, all system components function properly, except the items listed below. Each i failure is present prior to the initiation signal being received. j DESCRIBE the RCIC systems response for each of the following and JUSTIFY your answer. Consider each item separately.

a. The turbine exhaust valve (F068) is stuck shut. (0.75)

b. The Ramp Generator portion of the RGSC (Ramp Generator Signal Converter) has failed and is producing a large j signal. (0.75) 7N h The D/P cell, for the RCIC flow control element, has a perforated diaphragm. (0.75)

QUESTION 6.09 (1.00)

WHAT would be the result on water level (INCREASE, DECREASE, or y NO CHANGE) if, the FWLC Setpoint Setdown logic initiated 0 sporadically (K11 contact energized) at a full power condition.

WHY? (Assume 3-elementcontrol.) (1.0)

QUESTION 6.10 (2.00)

ANSWER the following questions concerning ADS initiating logic:

a. With the plant at 100% power, the channel Al and channel A2 manual initiation push buttoms are rotated and depressed. WILL the ADS function occur? WHY? (1.0)

./

I,[/ b. If ADS manually initiated (and all initiation conditions are met), WILL the ADS SRV opening be delayed by the i 105 second timer? (Yes/No) (0.5)

c. If an ADS blowdown is in progress, with all initiation signals still present, WILL depressing the ADS logic reset pushbutton switches reinitialize the 105 sec. timer and close all ADS valves. (Yes/No) (0.5)

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

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 QUESTION 6.11 (1.50)

Given the following data for APRM Channel C:

LPRM Level: A B C D i Number of LPRMs assigned: 6 5 5 5 4 0 0 I Number of LPRMs bypassed: 3

a. If APRM Channel C selector switch on the local (back) panel i was placed to the COUNT position, WHAT would be the expected meter reading? (SHOW calculations.) (0.5)

b. Based on the above data, is APRM Channel C operable:

ANSWER YES or NO and EXPLAIN WHY. (1.0) 1 QUESTION 6.12 (2.00)

The reactor is at 100% power with the generator synced to the grid.

Electrohydraulic Control (EHC) load set is 105%. By using the attached EHC diagram, EXPLAIN WHAT would happen to the control valve and bypass valve in the following circumstances:

, a. load limit potentiometer reduced to 95%. (0.5)

b. maximum combined flow limit potentiometer reduced to 95%. (0.5) h "A" pressure regulatory (transmitter) fails low. (0.5)

d. failure to two (2) bypass valves full open. (0.5)

QUESTION 6.13 (1.00)

The Unit i station air compressor is lined up in the 'ON' and the

-g ' MODULATE' mode of operation. If system demand is less than 60%

(hf.

, capacity, you are required to switch the mode switch to

' MODULATE +2 STEP' position. WHY must this action be taken? (1.0)

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

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 QUESTION 6.14 (2.00)

Concerning the CRD Hydraulic System:

a. The reactor operator is going to increase drive pressure (q to the HCU. WOULD you as the acting SR0 direct him to j- OPEN or CLOSE the drive water pressure control valve? (0.5)

/N b. EXPLAIN how your action in part a. has changed the following flow rates (INCREASE, DECREASE, NO CHANGE.) (1.5)

1. scram valve charging flow 2 CRD total system flow p se'j3),.

3 cooling flow QUESTION 6.15 (1.25)

Concerning the RHR System:

a. WHAT is the reason for the interlock between the (0.5)

1. shutdown cooling suction valve and the test return valve?

2. pressure control valve bypass valve and Rx Pressure?

b. If a LPCI auto initiation function (high drywell) were over-ridden to realign the RHR system to the shutdown cooling mode and another LPCI signal (low level) were to come in, WOULD the k RHR loop realign from the shutdown cooling mode to the LPCI mode? EXPLAIN. (0.75) i i

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(*****ENDOFCATEGORY06*****)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGEN'CY AND PAGE 13 RADIOLOGICAL CONTROL QUESTION 7.01 (3.50)

A total loss of AC power has occurred. ANSWER the following questions concerning LOA-AP-08 Total Loss of AC Power.

a. WHAT SRM and reactor water level indication (s) are available in the control room following this event? (1.0)

M' b. WHAT three (3) immediate actions are required following the scram? (1.5)

c. WHAT are the conditions for initiating the GSEP and WHAT classification would you recommend? (1.0)

QUESTION 7.02 (2.50)

During your shift the 1H13-P603 panel experiences an electrical short and a major panel fire occurs requiring you to evacuate the control room. Due to the fire you were not able to perform any of the immediate operator actions on Unit 1 but all on Unit 2 prior to exiting the control room.

a. WHERE would you expect your shift personnel-to assemble following the evacuation? (0.5) h With regards to Unit 1, WHAT is the method described in LOA-RX-01, Control Room Evacuation for (1.0) 11 1. scramming the reactor?

2. determining reactor level and pressure?

c. If you transfer shutdown control to the Remote Shutdown Panel, RCIC will not isolate on high water level. (TRUE or FALSE) (0.5)

d. If you transfer shutdown control to the Remote Shutdown

. Panel, N0 control is possible of those effected systems from the control room. (TRUE or FALSE) (0.5) l

(***** CATEGORY 07CONTINUEDONNEXTPAGE*****)

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 7.03 (2.50)

WHAT are the five (5) methods described in LOA-NB-09, Alternate Rod Insertion, to insert control rods still out following an automatic ,

, scram? (2.5)

'( d I QUESTION 7.04 (2.25)

According to LGA-ATWS-01, ATWS Power Control, WHAT are the three (3) conditions in which a cooldown rate above 100 degrees F/hr may Of be exceeded? (2.25)

QUESTION 7.05 (2.75)

The plant is operating at power when an SRV inadvertantly opens.

As per LOA-NB-02, Stuck Open Safety Relief Valve, the operator cycles the SRV control switch from AUTO to OPEN and back to AUTO.

a. If this action does NOT close the SRV, WHAT other method

l can be performed in an attempt to close the stuck open t

valve? (0.5) fi h WHAT control room indications would the operator have if the valve closed? (0.75)

c. WHAT three (3) conditions would require the operator to manually SCRAM the plant if the SRV remained open? (1.5)

QUESTION 7.06 (1.00)

Procedure LOA-NB-11, Preventing ADS Auto Actuation, is entered into from only two (2) other procedures.

a. WHAT are these two (2) procedures? (0.5)

b. The ADS system has been defeated by the lifting of leads in accordance with LAP-240-6. EXPLAIN what effect lifting these leads would have on automatic and manual initiation? (0.5)

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 7.07 (2.50)

You have been operating at 60% power when one (1) recirculation loop trips. You have been requested to restart the idle loop.

J h' According to the Recirculation Procedure, WHAT are the Technical Specifications temperature limits that apply to the restart of an idle loop? (1.5)

[t

b. If the idle loop cannot be restarted, COULD you return to 60% power operation with only one (1) recirculation loop for an extended period of time, (i.e., greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )? EXPLAIN. (1.0)

QUESTION 7.08 (2.50)

Concerning procedure LGA-03, Primary Containment Control:

a. Using the attached figures (curves LGA-G1, LGA-G2, LGA-G3 and LGA-G4), DETERMINE the minimum and maximum allowable suppression pool levels for:

p. 1. Reactor Pressure = 1000 psig j 2. Suppression Pool Temperature = 150 degrees F (1.0)

b. If all three (3) of the following conditions existed, you would be required to depressurize the Reactor Pressure Vessel? (TRUE or FALSE)

1. Reactor Pressure = 1000 psig

2. Suppression Pool Level = 0 ft

3. Suppression Pool Temperature = 180 degrees F (0.5)

c. WHAT could be the possible consequence and brir.71y EXPLAIN how this consequence could develop when inittai.;ng drywell spray when conditions are on the unsafe side of curve LGA-G3. (1.0)

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 i' RADIOLOGICAL CONTROL QUESTION 7.09 (1.50)

The LGA's refer you to LOA-RD-07, Simultaneous Operation of Both 9, CRD Pumps, as a way to minimize injection into the vessel. If you o are using the CRD Pumps to maintain level, SHOULD you reset the scram? EXPLAIN. (1.5)

QUESTION 7.10 (2.00)

.b F N.N If all rods did not insert following a Reactor Scram, WHEN would p you be required to inject SBLC7 (2.0)

QUESTION 7.11 (1.00) c4 h DEFINE a High Radiation Area. (1.0)

QUESTION 7.12 (1.00)

According to the philosophy of the LaSalle ALARA program EXPLAIN WHICH group you would identify for the following task: (1.0)

Group 1 - one (1) person would receive 150 mrem total exposure and take 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to perform the task (150 mrem for job)

OR Group 2 - two (2) persons taking two hours with each worker receiving 100 mrem total exposure (200 mrem for job)

(*****ENDOFCATEGORY07*****)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 QUESTION 8.01 (2.00)

You just performed a P-1 edit and you see that the computer calculated MFLCPR is 1.02. (Use attached Technical Specifications.)

a. IS this a Technical Specification Violation? (Yes/No) (0,5)

Given the following information IS a MFLCPR of 1.02 a valid

[ b.

value? (SHOW all work for full credit.) (1.5)

, MCPR = 1.42 Core flow = 71% of Rated Ave. control rod scram time = 0.83 sec.

I loop-recirculation flow controller in flux manual E0C-RPT operable QUESTION 8.02 (2.50) ,p According to Procedure LOA-GP, General Precautions:

a. WHAT precautions must be taken PRIOR T0 placing an ECCS system in manual? (1.5)

b. WHAT precautions must be taken WHILE an ECCS system is in manual? (1.0)

QUESTION 8.03 (2.00)

a. Other than "AS directed," WHAT are the entry conditions to LGA-01, Level / Pressure Control? (1.0)

. b. If you are executing step C.8.a "Depressurize the RPV" in procedure LGA-01 and another (not the initial one) entry 0}g condition to procedure at LGA-01 the initialisstep?

met, ARE y(ou required Yes/No) PROVIDE atoreason restart for the your response. (1.0) l

(***** CATEGORY 08CONTINUEDONNEXTPAGE*****)

. . _ . _ m

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 QUESTION 8.04 (1.50)

With the plant at 75% power, you are informed by the I&C foreman that the scram discharge volume (50V) level floates have failed the C( channel function test. STATE your actions and the reason WHY you must take action. (1.5)

QUESTION 8.05 (1.00)

WHEN may an unlicensed individual manipulate the controls of a reactor? (1.0)

QUESTION 8.06 (1.00) a.

In the event of an emergency WHO initially becomes the emergency (0.5) director?

Q

b. WHO later assumes the role of Emergency Director? (0.5)

QUESTION 8.07 (2.50)

Concerning combined Unit 1 and Unit 2 shift staffing:

WHAT is an adequate Technical Specification shift complement O

1 @ (limit answer to only licensed personnel)? (1.0)

b. WHO is responsible for assuring that a shift reporting for work is adequately staffed? (0.5)

C@

You report to work and find that a NSO is missing. Both units are at 100% power. ARE you in violation of Technical Specifications if you accept shift duty with one (1) man missing? (Yes/No) WHY? (1.0)

QUESTION 8.08 (1.50) 7 Under certain circumstances it is recognized that operation outside l the Technical Specifications may be required. WHAT three (3)

[f-situations allow you to violate Technical Specifications? (1.5)

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 1

QUESTION 8.09 (1.00)

During a surveillance on the RPS system with the plant in hot shutdown, it is discovered that the reactor mode select switch scram y function is not totally operable. STATE the action required. (1.0)

QUESTION 8.10 (1.00)

You are in operational condition I when you notice that IRMs A, B, and C, are inoperable. According to Technical Specifications WOULD.

you violate an LCO by:

a. staying in CONDITION I? EXPLAIN. .

(0.5)

O b. placing the mode switch in CONDITION 27 EXPLAIN. (0.5)

QUESTION 8.11 (3.00)

The Division 2A Diesel is operating and is 30 minutes into a surveillence test when the air starting system fails. The maintenance repair team estimates a 2-day minimum repair time.

'Ute the attached Tech. Spec. to explain your answers and reference tiie applicable technical specifications in your answer.)

a. Is the Diesel Generator inoperable according to Technical Specifications? EXPLAIN. (1.0)

b. ARE all the Division 1N (Unit 2) ECCS systems inoperable because of the Diesel Generator problem? EXPLAIN. (1.0)

c. IfatthesametimetheDivision$LPCIpumpisoutof bk service WHAT added implications does this have on your Technical Specification position? (1.0)

, QUESTION 8.12 (1.50)

, The reactor operator is performing a surveillance on the backshift ~

and due to a system modification a procedural step becomes impossible to perform. The decision is made to issue a temporary change such that the surveillence can be completed on schedule.

According to LAP-240-6, Temporary System Changes, WHAT requirements must be met prior to issuance of the temporary. change? (1.5)

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20 l

1 QUESTION 8.13 (2.00)

Using the attached Technical Specifications, DETERMINE the maximum time that the reactor may continue operation in condition 1 given the following malfunctions. Reference the sections of Technical Specifications used in determining your answer.

a. It is discovered that RHR pump 1A is inoperable. (1.0)

b. Subsequent to the malfunction in Part a., it is found that valve F048B (RHR "1B" Heat Exchanger Bypass) is failed open and cannot be closed. (1.0)

QUESTION 8.14 (2.50)

The RCIC outboard isolation valve (F008) steam motor controller has failed in the deenergized postion and the valve won't shut.

Maintenance is currently attending to the problem. By using the attached Technical Specifications:

h, a. STATE which Technical Specifications apply to this problem. (0.5) i STATE whether RCIC is OPERABLE or IN0PERABLE and GIVE ANY necessary action statement (s) required. (2.0)

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

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 5.01 (1.50)

a. Closer to (i D? S CP) [+0.5]

ce b b. Tarthe r:: y (decrease Cp) [+0.5]

pp, g. -Clc5cr4e (increase Cp) [+0.5]

REFERENCE

1. LaSalle: Core Thermal Hydraulics, pp. 56-62 of 107, TP0:Sc, 5d.

ANSWER 5.02 (2.00)

a. 2 - increase embrittlement [+1.0]

b. 3 - 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months [+1.0]

REFERENCE

1. LaSalle: LGP 3-1 and Exam Questions 5-LS-56.

2. LaSalle Core Thermal Hydraulics, pp. 80 and 90, TP0:8a, 8d.

ANSWER 5.03 (1.00)

a. Smaller [+0.5] (2 phase flow larger due to higher power, orifice fixed)

b. Smaller [+0.5] (2 phase flow larger due to smaller orifice)

REFERENCE

1. LaSalle: Core Thermal Hydraulics, pp. 20 and 21.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSUER 5.04 (2.00) 100% Rx power and 1000 psia implies 544.58 degrees F and h(x) = 542.6 BTU /lbm [+0.5]

assume isenthalpic process: [+0.25]

for 14.7 psia and h(x) = 542.6 BTU /lbm h(steam) = 1150.5 h(fluid) = 180.17

% steam = [h(x) - h(fluid)](100)/[h(steam - h(fluid)] [+pg]5

% steam = 37.35% vapor (fwy sw# WM-dg-F) 4 Ws Therefore d. saturated steam /liqu leak [+0.75 gp g REFERENCE

1. LaSalle: Thermodynamics and Steam Cycles, pp. 50 and 52, TP0:15b.

Provide Steam Tables.

ANSWER 5.05 (2.25)

a. Fuel temperature would INCREASE [+0.25] to get the needed delta T to transfer the heat to the coolant. The corrosion layer will require some delta T across it to transfer heat [+0.5].

b. Cladding temperature would also INCREASE [+0.25] because the pin temperature increased and the cladding is now transferring heat to the corrosion film instead of the coolant. [+0.5]

c. Coolant temperature REMAINS THE SAME [+0.25] since it is a function of pressure, which is maintained constant by the EHC system. [+0.5]

REFERENCE

1. LaSalle: Fluid Flow and Heat Transfer, pp. 76 and 78, TP0:II.B.5.

4

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 5.06 (2.25)

a. Decreases. [+0.25] There is less steam flow, therefore, less pressure drop through the main steam lines. [+0.5]

b. Increases. [+0.25] With the same amount of cooling water through the condenser and less of a heat load, condensate depression will increase. [+0.5]

c. Decreases. [+0.25] Less extraction steam from the turbine to heat the feedwater. [+0.5]

REFERENCE

1. LaSalle: Fluid Flow and Heat Transfer, pp. 34-40, TP0:C.I.

2. LaSalle: Fluid Flow and Heat Transfer, pp. 78-88, TP0:8.3, B.4.

3. LaSalle: Thermodynamics and Steam Cycles, pp. 78-86, TP0:24, 25.

ANSWER 5.07 (1.50)

a. 1. alpha m becomes less negative ;+0.25;

2. alpha v becomes less negative ,+0.25,

3. alpha d becomes more negative +0.25

b. 1. alpha m becomes in : ..o g a i. i ve ;+0.25; n' C T b"1 IA 7)

2. alpha y becomes less negative ,+0.25,

3. alpha d becomes more negative +0.25 REFERENCE

1. LaSalle: Reactor Physics Review, pp. 124, 138, and 164, TP0:16e, 17d, 18b.

2. LaSalle: Reactor Physics Review, pp. 122, 132, and 162, TPO:16e, 17d, 18b.

ANSWER 5.08 (2.50) fi/4 (_T-ir)

a. I.sua ' 0. 5'

+

b. True '+0. 5'

c. False '+ 0. 5'

d. True '+0. 5'

e. True l+0.5l

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

REFERENCE

1. LaSalle: Reactor Physics Review, pp. 208, 220, and 224, TP0:21, 22.

ANSWER 5.0" (3.00)

a. Rod worth increases, ;+0.25;duetohigherflux. [+0.5]

b. Rod worth increases, _+ 0.25 due to higher flux.

[+0.5]

c. Rod worth decreases, [+0.25] due to decrease in thermal neutrons.

[+0.5]

d. Rod worth decreases, [+0.25] due to reduced surface area seen by flux. [+0.5]

REFERENCE

1. LaSalle: Reactor Physics, pp. 184, 188, 190, and 198, TP0:19.C.

2. General Electric Reactor Theory, p. 5-13a.

ANSWER 5.10 (2.00)

Supercritical . [+0.5] When the period reaches infinity, the reactor is exactly critical on prompt neutrons. [+0.5] After the rod insertion stops the delayed neutron precursors which were formed in previous generations and at a higher power level tend to pull power back up. [+0.5] Therefore, the reactor is still supercritical due to the latent effect of delayed neutrons. [+0.5]

-De- (ps & 2 & pfuo45M & Q &

REFERENCE g O

1. LaSalle: Reactor Physics Review, pp. 102-112, TP0:7a, 12b.

2. NMP-2 Operations Technology, Module 1, Part 11, pp. I-11-6.

1 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

i ANSWER 5.11 (2.00)

Using P=Poe**(-t/T) [+0.25]

= 15 e**(-120/80)

= 3.35 on Range 4 [+0.25]

Therefore Rang i t owest Range [+0.5] -00 k A ljf MM Assumptions: On a down power transient, with large negative reactivity insertions, the stable decay period is determined by the longest lived half-life. [+0.5] For this example, it is as';omed to

+

be -80 seconds. [+0.5]

j REFERENCE

1. LaSalle: Reactor Physics Review, pp.80-112, TP0:12b, 13d.

Provide Equation Sheet.

ANSWER 5.12 (2.00)

a. CR1 (1 - Keff1) = CR2 (1 - Keff2) [+0.25]

200 (1 - 0.95) = 300 (1 - Keff2)

Keff2 = 0.9667 [+0.25]

delta p = Keff2 - 1)/Keff2 '(Keffl - 1)/Keffi [+0.25]

delta p = 0.9667 - 1)/0.9667 - (0.95 - 1)/0.95 delta p = -0.03448) - (-0.0526) delta p = 0.018 [+0.25]

b. Part b. will be graded independently on part a.

delta p = (Keff3 - 1)/Keff3 - (Keff2 - 1)/Keff2 [+0.25]

0.018 = 1 - (1/Keff3) - (-0.0345)

Keff3 = 0.984 CR3 = CR2 (1 - Keff2)/(1 - Keff3) [+0.25]

CR3 = 624.37 cps Therefore LARGER [+0.5]

i OR Larger [+0.5] because countrate is proportional to the inverse of total reactivity and as total reactivity approaches zero, countrate approaches infinity. [+0.5]

REFERENCE

1. LaSalle: Reactor Physics Review, pp. 66-76, TP0:15a,b.

l l

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 26 THERMODYNAMICS ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

I I

ANSWER 5.13 (1.00) ;

I

3. [+1.0]

REFERENCE

1. LaSalle: Reactor Physics Review, pp. 122, 132, and 162, TPO:16e l 17d, 18b.

J

'l i

6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 27 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

l t

ANSWER 6.01 (2.00) .

1. Reactor Water Level Control i

2. Process Computer

3. RWM

4. PCIS l

S. Co&Ak & g)

[+0.5] each (t/ 5 REFERENCE

1. LaSalle: System Description, Chapter 21, pp.19-22, TPO:1, 4.

ANSWER 6.02 (2.00)

1. Steam line (or dome) to pump suction temper 4ture difference is (10.1 degrees delta T.

2. Total feed flow (30%.

3. TSV or TCV closure with power >30% of rated (EOC-RPT).

4. Reactor water level (12.5". (AlsoacceptRxlowlevel.)

[+0.5] each REFERENCE

1. LaSalle: System Description, Chapter 5, pp. 70, 72 and 80, TP0:12c.

ANSWER 6.03 (1.50)

Nitrc$r s-admitted te Ge cent +bment-(to4Hute th: 92 =d 1.

Oih m tentrattert) via the containment vent and purgey system q M *" M/ A d

[+0.75] _g~

2. The Standb mixture. +0.75]

[y Gas Treatment System is used to remove the gaseous REFERENCE

1. LaSalle: System Description, Chapter 50, pp. 38-44, TP0:1f,h; 3a.

2. LaSalle: System Description, Chapter 51, p. 22, TPO:6b.

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 6.04 (2.50)

a. 1. Minimum - 50 minutes [+0.25] to ensure even mixing without ,

power chugging. [+0.5?

2. Maximum - 125 minutes [+0.25] to inject prior to the decay ,

of fission products providing additional positive reactivity. [+0.5]

b. Additional 25% poison (165 ppm) to allow for imperfect mixing

+0.5; and 250 ppm addition to accommodate for RHR dilution.

+ 0.5 REFERENCE

1. LaSalle: System Description, Chapter 10, p. 30, TP0:2b, 5b.

ANSWER 6.05 (1.00)

a. By changing the opening and reclosing pressures that the valves associated with the LLS operate at. [+0.5]

b. This will function if valves are opened manually OR automatically. [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 21, pp. 37-38, TP0:2.

ANSWER 6.06 (1.50)

a. False ;+0.5; l

b. True +0.5

c. True '+0.5 REFERENCE

1. LaSalle: Exam Bank, Question 2-LS-1.

l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29

-86/06/03-SLY, G.

ANSWERS -- LASALLE 1 i

ANSWER 6.07 (1.50) ,

f

a. closes on Unit 2 [+0.5]

b. remains closed on Unit 2 '+0.5'

c. rh_on-Uni t-4 '+0. 5' QonL M C,c W eJoU~:iA l REFERENCE

1. LaSalle: System Description, Chapter 47, pp. 37, 42, 43, 54, and 55, TP0:6c.

ANSWER 6.08 (2.25)

a. RCIC will not initiate [+0.25] the RCIC steam stop valve (F045) will not open if the exhaust valve (F068) is not full open.

[+0.5]

b. The turbine will trip on overspeed, [+0.25] the ramp generator is usually the low signal which controls the turbine on quick starts with this signal high the turbine is up to speed before sufficient oil pressure is available to the governor valve to close it. [+0.5] j p,. uc9ud

c. RCIC will tid;e6 o, : x'-"= rate end th: =i" fl:a valv; aiF1 Ret c;p^ d, (=ill .smo.n vemuj [+0.25] the flow signal is at minimum due flow from the to the RCIC zero d/p[+

system. 0.5] sensed therefore demanding Max.

de e d REFERENCE

1. LaSalle: System Description, Chapter 41, pp. 16, 21, 24, and 25, TP0:26, 3a.

2. WNP-2 System and Procedures, Vol. III, RCIC L.P., pp. 11, 12 and 13.

ANSWER 6.09 (1.00)

Decrease, [+0.5] because the level setpoint would be reduced to i 18" regardless of the setpoint tape setting. [+0.5]

REFERENCE c

1. LaSalle: System Description, Chapter 31, pp. 53, 56, 63, and 64, TP0:10.

l 1

I 1

l PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30 6.

ANSWERS -- LASALLE 1

-86/06/03-SLY, G.

I i l

ANSWER 6.10 (2.00)

a. No ADS response, [+0.5] due to no low pressure pump at pressure. "

[+0.5] - ok ye>Oa0 cha Ts m Q (e- /..Jy owau- W (Wsad d.J5ag,EjQ,;/ t b.

c. No [+0.5))

Yes [+0.5 E REFERENCE

1. LaSalle: System Description, Chapter 37, pp.14-16, TP0:6.

ANSWER 6.11 (1.50)

a. 70% [+0.25), 5% (volts) for each LPRM not bypassed [+0.25]

b. No [+0.5], there are fewer than two (2) operable inputs on Level B [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 14, pp. 24 and 62, TP0:4, 6.

ANSWER 6.12 (2.00)

a. control valves close 5% [+0.25], open one bypass valve [+0.25]

(or similar control answer valves on dia close 5%+0.25 [ gram)], reactor scram probable due to b.

increasing pressure since bypass valves will not be open. [+0.25]

c. 'b' controls, C.V. closes then, reopens to 100% as 'b' takes over. Bypass does not respond. [+0.5] oCe-

d. control valves close to 90% [+0.25] to maintain Rx pressure at 920 psig. [+0.25]

REFERENCE

1. LaSalle: System Description,, Chapter 25, 26, 73, 74 and 76, TP0:4, 5.

2. NMP-2 Operations Technology, EHC, Rev. 1, pp. 2, 5 to 9 of 14, Student Learning Objective Nos. 5, 6, 8, including EHC Diagram.

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 6.13 (1.00)

Because the air operated blow off valve will not open in the 8

' MODULATE' mode [0.5] to relieve the excess pressure [0.5]

REFERENCE C f"" *"

1. LaSalle: System Description, Chapter 68, pp. 15, 18, 19, and 20, TP0:3b, 6b, 7a.

ANSWER 6.14 (2.00)

- Cl * [+

. . r.. b.. .. . .,',70P. , ,-

l

2. no change +0.5

3. decreeqe l+0.5[

REFERENCE

1. LaSalle: System Description, Chapter 8, pp. 20 and 33, TP0:2e, 3 and Figure 8-2.

ANSWER 6.15 (1.25)

a. 1. prevent inadvertent draining of the vessel. [+0.25]

2. prevent exceeding RHR design pressure. [+0.25]

b. No [+0.25], the LPCI suction and injection valves will be shut in the shutdown cooling mode and will not automatically-realign [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 39, pp. 24, 25, 28 and 35, TP0:4, 7.

l l

-)

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 7.01 (3.50)

a. SRM - SRM meters only on panel 603 [+0.5]

b. Level

1. - Narrow ARM Range and depress theB&C meters Manual Scramon panel 603 [+0.5][+0.5]

pushbuttons

2. Place Mode Switch in SHUTDOWN [+0. 5]

3. Check Control Rod Position by performing a OD-7 option 2

[+0.5]

c. 1. if all 4160 V ESS buses are lost for more than 15 minutes

[+0.5]

2. site emergency [+0.5]

REFERENCE

1. LaSaile: LOA-AP-08, pp. I and 2.

ANSWER 7.02 (2.50)

a. in the Unit 1/2 Aux. Electric Rm. [+0.5]

b. 1. Open CB2A/CB2B RPS subchannel Logic Power Supply Breakers

[+0.5]

2. at the Remote shutdown panel following transfer of instrumentation [+0.5]

c. TRUE '+0. 5'

d. TRUE l+0.5 REFERENCE

1. LaSalle: Control Room Evacuation, pp. 1-4, and 8.

ANSWER 7.03 (2.50)

1. Method 1 - removal of fuses for scram solenoid power [+0.5]

2. Method 2 - drive rods with CRD system [+0. 5]

3. Method 3 - reset the scram and initiate a manual scram [+0. 5]

4. Method 4 - individually scram the pontrol rods not inserted

[+0.5]

Method adi) from the overpiston side of the 5 - remove the pressu(re 5.

CRDunit(F102 valve) [+0.5]

REFERENCE

1. LaSalle: LOA-NB-09, pp. 1-4.

j 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G. .

ANSWER 7.04 (2.25)

I

1. Conserve RPV water inventory [+0.75]

2. Protect containment integrity [+0.75]

3. Limit radioactive release to the environment [+0.75] l REFERENCE

1. LaSalle: LGA-ATWS-01, p. 9.

ANSWER 7.05 (2.75) l

a. Pull the fuses for the affected valve. [+0.5]

b. Control of fuses switch

[+0.25]valve indication or tailpipe [+0.25] followingl temperature. [+0.25 - replacement e-

a 1 01 wA eata c. 1. Four attempts to cycle valve

2. Pool temperature reaches 110 deg [+0.

F 5] [+0.5]

3. Two minutes have elapsed [+0.5]

REFERENCE

1. LaSalle: LOA-NB-02, pp. 2 and 3.

ANSWER 7.06 (1.00) a.- LGA-ATWS-01 [+0.25] and LGA-04, Level Restoration [+0.25]

b. 1. auto-timer defeated '+0.25'

2. taanual-still . active l+0.25 REFERENCE

1. LaSalle: LOA-NB-11, pp. 1 and 2,

2. LaSalle: System Description, Chapter 31, p. 57.

i j 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 7.07 (2.50)

I

a. 1. Steam dome space to bottom drain less than or equal to 145 deg F [+0.5] [+c 7Q

2. Idle loop to operating loop less than or equal to 50 deg F [+0.5] 070 l tmum he cooldoyn-rate for the Reactory or

3. @RectFc., System is 0 deg F/ho W O.1] V

b. No [+0.5], due to a maximum imposed limit of 50% power [+0.5]

l REFERENCE

! 1. LaSalle: l0A-RR-03, p. 3; LOP-RR-06, pp. 5-9.

! 2. LaSalle: Technical Specifications 3/4.4.1.1, 3/4.4.1.4, j 3/4.4.6.1.

i I

l ANSWER 7.08 (2.50)

a. 1. minimum = ft [+0. 5]

l 2. maximum = 9-ft [+0. 5]

b. TRUE [+0.5]

c. Insufficient noncondensibles in the suppression pool creating a negative pressure greater than the capacity of the vacuum

breakers [+0.5] resulting in the potential destruction of the containment. [+0.5]

REFERENCE j 1. LaSalle: LGA-03, all, (provide figures).

1 2. General Electric: E0P Fundamentals Bases.

l ANSWER 7.09 (1.50) l No [+0.5] (the scram should not be reset), because doing so would close the scram inlet valves causing a greater restriction to flow

! in the vessel. [+1.0]

REFERENCE i

1. LaSalle: LOA-RD-07, LGP-3-2, Exam Bank.

1 l _. - - . . . , _ . . . _ , _ . . . _ _ _ _ . _ . _ . . . _ _ _ . , . _ , . _ , . ~ _ _ _ . . . . _ , , . . _ . _ - . . - - , _ . , _ _ _ - _ . . . . . , . _ . - , ,

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 7.10 (2.00.)-

~ ~ ~

_ _ _ _ . . _ _ . c

} f.' S y pu~w.p

1. , If five or more adjacent rods are not inserted to at least 2.

notch 06 [+0.5] OR f g9,67 ,V4 g "a thirty or more rods are not inserted to at least notch position Ax QS,,ng, j 06 [+0.5] AND

3. j water level cannot be maintained above +12.5 in. [+0.5] or d "' 8' *.. b ' A % !d.

{ suppression pool temperature reaches 110 degrees F. [+0.5]

W j4 y n 46,I M ~ f REFERENCE M. ' (g

-f5 d 6 & c ,p / w ' g

1. LaSalle: CO nA-NB-09, p. 2.

jc,4- Ar.& of g g,Y ANSWER 7.11 (1.00)

High Radiation Area - Any area accessible to personnel in which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 100 millirem. [+1.0]

REFERENCE

1. LaSalle: LRP-1000-1, Definitions.

ANSWER 7.12 (1.00) jygj ggy Group f, [+0.5] because you try to minimize,Ddiv4 3tgal dese [+0.5]

REFERENCE

1. LaSalle: LRP-1000-1, ALARA.

2. 10 CFR part 20.

t

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

9

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 8.01 (2.00)

a. Yes [+0.5]

b. MCPR limit = (1.25 + 0.01) (1.1) = 1.39 [+0.5]

MCLCPR = MCPR limit /MCPR = 1.39/1.42 = 0.98 [+0.5]

1.02 - not valid [+0.5]

REFERENCE

1. LaSalle: Technical Specifications, pp. 3/4 2-4 to 3/4 2-6.

Provide Technical Specifications.

ANSWER 8.02 ('56)

a. Do not secure or piace an ECCS in MANUAL mode unless, by at least two independent indications [+0.5] (coaatd omA,k woA v>-* c c,4..G)

1. misoperation in AUTOMATIC mode is confirmed 0.5] OR

2. adequate core cooling is assured C+0.5 [m[+udSkJ,1Q

b. If an ECCS 1;#p aced in MANUAL mode, it will not lnitiate automaticalFf.v}Makefrequentchecksoftheinitiatingor controlling parameter [+0.5].(Whenmanualoperationisno longer required, restore the system to AUTOMATIC / STANDBY mode if possible){*Gr5-]

REFERENCE

1. LaSalle: LGA-GP, p. 2, Precaution ill.

2. General Electric: E0P Fundamentals, Specific Caution #10.

ANSWER 8.03 (2.00)

a. 1. Boron has not been injected [+0.25] and any of the following

2. RPV water level (12.5 in. [+0.25]

3. RPV pressure >1043 psig [+0.25]

4

4. drywell pressure >1.69 psig [+0.25]

(Setpoints not required for full credit)

b. Yes, [+0.25] to ensure that the intent of steps previously performed would not be altered due to the new entry condition.

[+0.75] REFERENCE

1. LaSalle: LGA-01, p. 2.

2. General Electric: E0P Fundamentals Bases for E0Ps.

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 8.04 (1.50) The SDV is inoperable, [+0.5] you have to assume more than 8 control rods are inoperable, so the plant must be placed in Hot n "." Shutdown within 12 hours. [+1.0] dead uret ofw/-Q o# M ' d' '~

S)dck REFERENCE (wcLuh f.o.Q inmf f.< h ,4c A.is / sh W ste yrkg y#. 3N?pf

5. y./-/

1. LaSalle: Technical Specifications LCO, Reactivity Control Systems, Control Rods.

Provide pp. 3/4.1.3, 3/4 1-3 to 1-5, 3/4 3-1 to 3-4, and OCRPS 3/4.3.1. ANSWER 8.05 (1.00) An individual may manipulate the controls of the reactor as part of his training [+0.5] (as a student under the direction) and in the presence of a licensed operator or senior operator. [+0.5] REFERENCE

1. LaSalle: LAP-1600-2, pp. 12 and 21.

2. 10 CFR 55.9.

ANSWER 8.06. . s pjy ShJ.:.p.ana O(.1.00) n L ,7 - u ">

a. GSE [+0. 5]

b. G e ral (" % tendent [+0. 5]

p t.e, p. & tv (s kJ.C 1%'; ' REFERENCE

1. (CAF) No, GSEP Manual provided.

                                          ~             -

22- /' ///S-ij /$ _f l

8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

i i t ANSWER 8.07 (2.50)

a. 1 SE [+0.25] j 1 SCRE [+0.25]

SF f(3NSO[+0.25] [+0.25]

b. Station Shift Engineer (SE) [+0. 5] l

c. Yes, [+0.25] because the time limit acception does not permit any shift crew position to be unman _ed upon shift change.[+0.75]

a REFERENCE

1. LaSalle: LAP-1600-2, p. 2.

2. LaSalle: Technical Specifications, 6.1-3.

ANSWER 8.08 (1.50)

1. prevent injury to the public or company personnel. [+0.5]

2. prevent release off-site above technical specifications limits

[+0.5]

3. prevent damage to equipment, if damage could adversely effect public health and safety. [+0.5]

REFERENCE

1. LaSalle: LAP-1600-2, p. 5.

ANSWER 8.09 (1.00) Verify all insertable control rods are inserted within one hour. [+1.0] REFERENCE

1. LaSalle: Technical Specifications, LCO, Instrumentation, Reactor Protection System Instrumentation - Scram Table 3.3.1-1, pg.

3/4 3-4. i

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39

ANSWERS -- LASALLE 1 -86/06/03-SLY, G. ANSWER 8.10 (1.00)

a. No, [+0.25], IRMs are not required in Condition 1 and you may stay there [+0.25].

b. Yes [+0.25], unless you had the RPS trip System A in the tripped position. [+0.25]

REFERENCE

1. LaSalle: Technical Specifications pp. 3/4 3-1 to 3-4.

ANSWER 8.11 (3.00)

a. Yes [+0.25], due to failure of surveillence 4.8.1.1.2.7, air pressure greater than 200 psig. [+0.75] -or- (attendent systems are not operable.)

b. No, [+0.25], due to specification 3.0.5 which states you can be without emergency power source if you have everything else [+0.75] 5

c. You would be in violation of Specification 3.0 5 [+0.5], and must perform the action statement 3 a 1 L=e [+0.51 3,0 6 REFERENCE

1. LaSalle: Technical Specifications, pg 3/4 0-1, 8-1 to 8-7.

Provide T.S. 3.0.1, 3.8.1.1, 3.6.5.3, 3.6.6.1, and 3.7.2. ANSWER 8.12 (1.50) A safety evaluation is completed. [+0.5] - 04 +M 5 # ' 1.

2. The change is reviewed by two (2) Senior Reactor Operators

[+0.5], (at least one (1) of whom has an engineering degree. [+0.25]) f.e gsh or.S

3. The change is authorized by the Shift Engineer. [+0.25] 46' l b"'"

REFERENCE

1. LaSalle: LAP-240-6, p. 3.

1 i i I .- - -- . - . . _ . - . . _ . . _ , _

.1

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40

  . ANSWERS -- LASALLE 1                              -86/06/03-SLY, G.

ANSWER 8.13 (2.00)

                          .) .L 23 . c-                             6

a. T.S. Action-3.5.1.b.2 states restore in 7ddeys or be in hot shutdown in 12 hrs (and cold shutdown in 24 hrs.) [+1.0]

b. T.S. Action 3.6.2.3.b states with both suppression pool cooling modes inoperable, be in at least hot shutdown in 12 hrs (and cold shutdown in 24 hrs.) [+1.0]

REFERENCE

1. LaSalle: Technical Specifications, 3/4.5.1 and 3/4.6.2.3.

Provide T.S. pg. 3/4 5-1 to 5-3, 3/4 6-20 to 6-21. 't ANSWER 8.14 (2.50)

a. T.S. 3.6.3, Primary Containment [+0.5]

b. Operable [+0.5], but you violate Primary containment and must adhere to Action 3.6.3.a [+1.5] -or- close the RCIC inboard valve [+0.5], then RCIC is INOPERABLE [+0.5] and you must adhere to Action 3.7.3.b [+1.0]

REFERENCE

1. LaSalle: Technical Specification, 3/4.6.3 and 3/4.7.3.

Provide T.S. pg. 3/4 6-22 to 6-28 and 3/4 7-7 to 7-8. l l l

j.- TEST CROSS REFERENCE PAGE 1

4 QUESTION VALUE REFERENCE , 05.01 1.50 SLY 0000676 { 05.02 2.00 SLY 0000677 05.03 1.00 SLY 0000678 ! 05.04 2.00 SLY 0000679 05.05 2.25 SLY 0000680 1 05.06 2.25 SLY 0000681 , l 05.07 1.50 SLY 0000682 i 05.08 2.50 SLY 0000683 05.09 3.00 SLY 0000684 l 4 05.10 2.00 SLY 0000685 l 05.11 2.00 SLY 0000686 i 05.12 2.00 SLY 0000687 i 05.13 1.00 SLY 0000782 i 25.00 06.01 2.00 SLY 0000688 06.02 2.00 SLY 0000696 06.03 1.50 SLY 0000697

06.04 2.50 SLY 0000698 06.05 1.00 SLY 0000699

! 06.06 1.50 SLY 0000700 1 06.07 1.50 SLY 0000701 06.08 2.25 SLY 0000702 06.09 1.00 SLY 0000703

06.10 2.00 SLY 0000704 I

06.11 1.50 SLY 0000705 06.12 2.00 SLY 0000706 06.13 1.00 SLY 0000707 l 06.14 2.00 SLY 0000722

06.15 1.25 SLY 0000723 i 25.00 1

! 07.01 3.50 SLY 0000689 07.02 2.50 SLY 0000690 07.03 2.50 SLY 0000691 l -07.04 2.25 SLY 0000692 ! 07.05 2.75 SLY 0000693

07.06 1.00 SLY 0000694 1 07.07 2.50 SLY 0000708 i 07.08 2.50 SLY 0000711 07.09 1.50 SLY 0000712

! 07.10 2.00 SLY 0000713 l 07.11 1.00 SLY 0000714 j 07.12 1.00 SLY 0000715 25.00 08.01 2.00 SLY 0000695 , i i

9 TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE j ________ ______ __________ 08.02 2.50 SLY 0000709 08.03 2.00 SLY 0000710 08.04 1.50 SLY 0000716 08.05 1.00 SLY 0000717 . 08.06 1.00 SLY 0000718 08.07 2.50 SLY 0000719 08.08 1.50 SLY 0000720 08.09 1.00 SLY 0000721 08.10 1.00 SLY 0000724 . 08.11 3.00 SLY 0000725 08.12 1.50 SLY 0000726 1 08.13 2.00 SLY 0000727 08.14 2.50 i SLY 0000728 I 25.00 100.00 l l t 4 I 4 6

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

l [% fAC.ln o' LASALLE SRO WRITTEN EXAMINATION HANDOUT

1. EQUATION SHEET

2. EHC DIAGRAM

3. FIGURES: LGA-G1(HCTL),LGA-G2(RSL), LGA-G3(SIPL), LGA-G4(HCLL)

4. TECHNICAL SPECIFICATIONS:

3/4.0 APPLICABILITY pp. 3/4 0-1 3/4.1.3 CONTROL ROSS PP. 3/4 1-3 to 1-5 - 3/4.2.3 MINIMUM CRITICAL POWER RATIO pp. 3/4 2-4 to 2-6 t 3/4.3.1 REACTOR PROTECTION SYSTEM pp. 3/4 3-1 to 3-3 , 3/4.5.1 ECCS<0PERATING pp. 3/4 5-3 to 5-3 3/4.6.1 PRIMARY CONTAINMENT pp. 3/4 6-1 3/4.6.2.2 SL'PPRESSION POOL SPRAY pp. 3/4 6-20 to 6-21 3/4.6.3 PRI!!ARY CONT. ISOLATION VALVES pp. 3/4 6-22 to 6-28 - 3/4.6.5.3 STANDBY' GAS TREATMENT SYSTEM pp. 3/4 6-40 to 6-42 3/4.6,6 PRIMAM CCNT. ATMOSPHERE CONTROL pp. 3/4 6-43 i 3/4.7.2 CONTROL ROOM FILTRATION SYSTEM pp. 3/4 7-4 to 7-6 !

                                                        ;                                                              i 3/4.7.3              RCIC SYSTEM                           pp. 3/4 7-7 to 7-8                           j 3/4.8.1              A.C. SOURCES-GPERATING                pp. 3/4 8-1 to 8-7 P

f l l l l i i i

  ,                                                         EQUATIult SilEET Where ml = m2 (density)1(velocity)1(area)'1 = (density)
     . . . . . . . . . . .. . . ._ _ _ . . .._ _ . . . . ... . . . _ _. . . 2 ( v e l oc i ty ) 2 ( a r e a ) 2 KE.= mv2                  PE = mgh        PEl +KEl +P V 7                                               1 1 = PE +KE  2    +P 2 V22          where V = specific volume i

P = Pressure 3 4= de p( Tout-Tin) Q = UA (Tave-Tstm) 0 = m(ht-h2 I .i P = Po 10sur(t) p . p oe t/T SUR =_26.06 I I delta K = (Ke rr.1)/Ke rr CR 1 (1-K e rri) = CR 2 II-Keff2) 14 = (1-Kerrg) SDit = (l.Kerr) x 1007, . ll-Keff2) K eff

                                                                                                          ~"~          ~

decay constant = inih) = 0693 k=k,Udecaycob$tabii'(t) g x t 1/2 t1/2 Water Parameters itiscellaneous Conversions 1 gallon = 8.345 lbs I gallon = 3.78 liters 1 Curie = 3.7 x 1010 dps 1 kg = 2.21 lbs

. 1 ft3 = 7.48 gallons                                                   I hp = 2.54 x.103 Btu /hr                                 "

j , i i Density = 62.4 lbm/ft3 1 ilw = 3.41 x 106 Blu/hr i Density = 1 gm/cm3 lleat of Yaporization = 970 Btu /lbm 1 inch = 2.54 centimeters Degrees F = (1.8) x (Degrees C) + 32 lleat of Fusion = 144 Btu /lbm 1 Btu = 778 f t-Ibf ^ 1 Atm = 14.7 psia = 29.9 in lig g = 32.174 f t-lbm/lbf-sec2 _ O i l

         .                                                                                                                                                 l
                                                                                                                                                          }

e s

                       .                    y                                                                                                   .                                                                                                                                                             >

TURBINE TRIP Acti_t.. S E t.*

                                                                                                       -                                             SPEED RROR OPEN BIAS ypo.j" Od INTERCEPT

[ +1Ej

                                                                                                 ,E                                                                                                                                                :                                 :     VALVE
                                                              -*             d d'                                               -

F 8 IV REGULAllON DEMAND

                                                                                                                                                % FLOW                                                           ,

mg ~ AA ' SPEED -

                                                                                                                                                                                                                ~.T I             CV REG SPEED                     -                         LV                                                                               T IV REG                                                               .

SELECT ' ' LINE SPEED MATCHER

PoucitoAO . . TIR81NE tussALAaser ' RE.740TE t ut / OG C. ST.fa . . g ggg p

                         .SeEED                                         _y,                               ,
                                                                                                                                                                            ,            o                                              i
                                                                                                                                                                                                                                                                                                                   ~
                                       .                .-+ d dr                                 ,E                      [        -

A -- 2 H DEC . .

                                                                                                     *                                         % FLOW .
                                                                                                                                                                                     '     ~

LOAD SELECTOR , NC Ac.LE L S a t..

MASTER FLOW i CONTROLLER -10' MAPAJAL ,

EG*% 4"Wu" NW? i O U TO RECIRC AUTO , FLOW = O\ (D

                                                                                                                                                                                                                                      .,          RUNBACK SYNC SPEED NOT SELECTED CONTROL                                 ,                      ,

GAIN

RUNBACK ON LOSS OF STATOR COOLING a se lsami *
STEAM 7tmOTTLE

                                                                                ,Q1 E

i lPSFF g ' FROM SPEED SELECTOR =A D 2 NO 2 STOP VALVE DEMAND . PRESSURE .

                                                                                                  ~                                                                                                                                                             t                                                                                        .

A 2~* * "la R**48 TURBINE TRIPPED CHEST WARMING Opsi hO CONTROL gigg 33 ; VALVE INC h- MOT ' =

DEMAND

                                                                                                                               '                                         O           "

Pae ss PRESSURE SET , T_ MAxiuuu COuulNED .[E r k DEC h E - LOAD FLOW

                                                                                                                                                         .              TUR8 8                     -         e tOpsi                                                     TRIPPED
                                                                                                                                                                                                        ~

8tAS * ,- BYPASS STEAM. + *

                                                                                                                                                                                                      =[ET                                                               r/LVG               VALVE l                          THROTTLE                                             - E                                                                                                                                                                                                        DEMAND i

PRESSURE ,V! ~

~ .

SMALL i

                                                                                                                                    -                    A                                                         CLOSE             BYPASS JACK-              Om              LOW VACUUM
                                                                                                                             '                                                                                             '                                                   7" Hg.

d

LGA- GI (HCTL? HEAT CAPACITY TEMPERATURE LIMIT .. l Y4 % % M M V f d %VA!VIVAfANMMMMN

         *'              !MMMMMVMMM%%%%MGM
                         !%%%%%%%%R#9M%%% VAM
                         !5MMMMMMMVM GVAM%%
                         !M5GMMMMMMMMMMMMM
         **       l      iMMMMMM%#MMMMMM
                         !sMMMMMMWWWWtf# AGE c                    fGWMWiW/#4M% PING!MP#' 4W t                 IwnmMMMM5%%279MANM
    =

I VMM%%MMMS@#AME ii l l Yn$MGMMVA R MMMM s 422WWARM58W#

AGE l &M d V$/VYYYV4YAh$WkVV M ~ 5* III YdMMMMSWRMFAME I v&WMWA%MVM$6t#MVA

    $                                 ^@@MMMMMMM
    ;                                     MnnMMMMMMM
    ?                                    'llM5V4MMMMMn 5
          'S I                I      I       X"#ViMMMM
    %                                                  'V4%MMMM                 -

NMMMi ll l%3%M

           '8                                                    ' s, 170 135 200       400          600 -   800    1000 1076 RPV PRESSURE (psig)

Rev. O LGA-ATWS-03

 , 09/17/85                                  2 of 21

LGA- G2 (RSL? m RPV SATURATION TEMPERATURE LIMIT 550 ' MMMMMMMMSP 500 MMMMMV 450 m MMMF7

MEF

    =   400     --

350 Mtf e W

300 y x gf d 250 t 200 10 0 300 500 700 900 1076 RPV PRESSURE (psig) Rev. O LGA-ATWS-03 09/17/85 6 of 21

s LGA-G3 (SlPL?

DRYWELL SPRAY INITIATION PRESSURE LIMIT 3U WMM M M MMM MMWWESMMMMMMME

            $$MRMMMMMMM@$MMMGMBR ENMMGMMMMMMMMMMMMM MEMRMMMMMMMMMMMMMW 250 MMMMMMMMMMMMMV                                            i MGMMMMMMMMM3V                                             I c         M@MMMMMMMF                                              II 2-        MMMh4MMWMSVI                           I                  I g   200

MWMMMMMPf a l I -

l

  @         MEMMMMW                     L                             I 5         WL% WSMM7                            l                    l l         &%%%MMM                                                   l-     -
       '50 E         b1MMLD            1 l            ll     l     ll s         MMMMW                                                   II 3         MMMMf1                                                     i E         SMMMR@#                                        '

I II b l00 B MGMW I II R$$$F/ a l II MMMW MMM I 50 MM' MWW l l 0 0 10 20 30 40 50 60

SUPPRESSION CHAMBER PRESSURE (psig) Rev. O LGA-ATWS-03 09/17/85 4 of 21

l LGA- G5 (HCLL) ^ HEAT CAPACITY LEVEL LIMIT 0 i i i i i i i i

         .,         i            l                         l                                    l Ilo             /                          /l                                  /     I       I I/O         le                          /                                   /               I
         ~*
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I/ W -5 1/ I/ l l l 3 j V

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5 ~6 Il /l n / I E 'l I /! l / I I g _7

               /             I/                   8                                         /                I o  -e I                     I                /                      I 5          i        /              le/                  I           /                           I       I/

W I / 'g/ s l / M

      $           I./                 /             ,           /                                  Y          I
        -10
                    !               !                   /                                     /\
                 /             /                      /                                 /                     I
                          /                    /                                /                             l
         ~

_;g I (/

g t 2RREBEREBERREMBERRE 135 200 300 400 500 600 700 800 900 1000 1076 RPV PRESSURE (psig) NOTE: MAINTAIN LEVEL AB0VE EXISTING PRESSURE TEMPERATURE POINT ._

   ,$Ui7)as                                                Is Ifi$

. LGA-G4 [SPLL? SUPPRESSION. P0OL LOAD LIMIT 8 /4 MilillllRHB 1888118188H1 188R558R8 I V8Bifflit

       !'                                     188E8N1 i                                        WINRW
       =                                          WRER        -

WBM i'

YRRt Ylil

El 1H BRR 0 250 500 750 1000 1250 RPV PRESSURE (psig)

    ,$9)i7 85 8of2N

K . l( t - i 3/4.0 APD'.ICABILITY ' j. * 'i LIMITING CCNOITION FOR OPERATICN i

   !                      3.0.1 Compf f ance with the Limiting conditions for Operation centained in the j                     succeeding Specifications is required during the OPERATICNAL CCMOITIONS or other
    ;                     conditions specified therein; except that unen failure to meet the Limiting
    ;                     Conditions for Operation, the associated ACTICH requirements shall be met.
           .              3.0.2 None:moTiance with a Specification shall exist when the requirements of i                     the Limiting Candition for Operation and associatad ACTICH requirements are not met within the specified time intarvals. If the Limiting Candition for Operation fs restored prior to expiration of the 1:ecified time intarvals, c:mpletion of the ACTICN requirements is not required.

3.0.3 When a Limiting Condition for Cperation is not set, except as provided -- in the associatad ACTION requirements, within one hour action snall be initiated ~ to place the unit in an OPERATICNAL CCNOITION in which the Specification does not apply by placing it, as applicable, in: - l 1. At least STARTUP within the next 6 hours,

2. At least HOT SHUTCCWN within the following 6 hours, and l ,

! 3. At least COLD SHUTC0hN within the subsequent 24 hours. Where corrective seasures are ecnoleted that permit coeration under the ACTICN requirements, the ACTICM may be taken in ac:sedancs with the snecified time j limits as measured from the time of failure to meet the Limiting Condition for Oceration. Exceptions to these requirements are statad in the individual Specifications.

      .                   This specification is not applicable in CPERATICNAL CCHOITICN 4 or 5.
      !                   3.0.4 Entry into an OPERATIONAL CCp0ITION or other specified condition shall not be mace unless the conditions for the Limiting Concition for Operation are
      !                   set without reif ance on provisions c:ntained in the ACTICN recuirements. This                                                        l' l                   provision snall not prevent passage through OPERATICNAL CCN01TIONS as recuirac                                                        l l                   to comoly with ACTION requirements. Exceotions to these reouirements are                                                              l

, stated in the individual Specifications.

I i 3.0.5 When a systas, subsystes, train, component or device is detemined to j be inoceracie solely because its emergency power source is inoperacle, or -

       ;                   solely because its normal power source is inoperanle, it may be considered l                  OPERASLE for the purpose of satisfying the requirements 'of its applicable                                                  -
       !                   limiting condition for Operation provided: (1) its corresponcing normal or j                  emergency power source is OPERABLE; and (2) all of its reduncant system (s),

i subsystam(s), train (s), component (s) and device (s) are OPERABLE or likewise I satisfy the requirements of this specification. Unless both conditions (1) i and (2) are satisfied, within 2 hours action shall be initiatec to place the unit in an OPERATIONA'L CONDITION in wnic.'i the scolicable Limiting Condition { f r Cparation ::ei cet rely by piscir; it. Is at:i':n:2. i n: -

1. At least 3TA.1 ;;' ithin t .e ev: 5 :.rs, V 2. At least HOT SHUTCOWN within the following 5 hours, and

3. At least COLD SHUTCOWN within the sunsecuent 2t hours. .

         .                 This specification is not appifesble in OP'ERATIONAL CCN0!TICN 4 or 5.                                                                l
                      ,    LA SALLI - UNIT 1                          3/4 0-1

REACTIVITY CONTROL SYSTEM 3/4.1.3 CONTROL RODS . CONTROL RCD OPERABILITY . LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE. AP'LICABILITY: OPERATIONAL CONDITIONS 1 and 2.

                         .       ACTICN:

a. With one control red inoperable due to being inunovable, as a result of excessive friction or mechanical interference, or known to be untrippable:

1. Within 1 hour: l a) Verify that the inoceracle control red, if withdrawn, is separated from all other inoceracle control esds by at least two control cells in all directions.

b) Disarm the associated directional control valves" either: l

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves. .

c) Comply with Surveillance Requirement 4.1.1.c.

2. OtheMse, be in at least HOT SHUTDOWN within the next 12 hours.

3. Restore the inoperable control red to CPE.WA8LE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours, b ,. With one or more contro'2 rods trippable but inoperable for causes other than addressed in ACTION a, above:

1. If the inoperaale control rod (s) is withdrawn:

a) Immediately verify:'

1) That the inoperable withdrawn control red (s) is separated l from all other inoperable withdrawn control rod (s) by at least two control cells in all directions, and

2) The insertion capability of the inoperable withdrawn

                                                                         . control rod (s) by inserting the control rod (s) at least one notch *Jy drive water pressure within the nonnal                                                     ~

operating range"". l b) Othe M se, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves" either: l

1) Elect.rically, or

2) Hydraulically by c' losing the drive water and exhaust water isolation valves "May be reensed intensittently, under administrative control, to permit testing l associated with restoring the control rod to OPERABLE status.

j **The inoperable control rod may then be withdrawn to a position no further l withdrawn than its position when found to be inoperable. l LA SALLE - UNIT 1 3/4 1-3 Amendment No. 18

REACTINITYCONTROLSYSTEM LIMITING CCNDITTON FOR OPERATION (Continued) - ACTION (Continued)

2. If the inoperable control rod (s) is inserted:

a) Within I hour disarm the associated directional control lig valves

either: .

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

b) Otherwise, be in at least HOT SHUTCCWN within the next 12 hours.

3. The provisions of Specification 3.0.4 are not applicable.

ls

c. With more than 8 control rods inoperable, be in at least HOT SHUTUCWN within 12 hours. .

SURVEILLANCE 'RECUIREwENTS 4.1. 3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERA 8LE by:

a. At least once per 31 days verifying each valve to be open**, and l id

b. At least once per 92 days cycling each valve through at least one l complete cycle of full travel.

l 4.1.3.1.2 When above the low power setpoint of the RWM and RSCS, all withdrawn control rods not required to have their directional control valves disar.ned glectrically or hydraulically shall be demonstrated OPERABLE by moving each , , control rod at least one notch: . l

a. At least once per 7 days, and I

b. At least once per 24 hours when any control rod is immovable as a result of excessive friction or sechanical interference. .

4.1.3.1.3 All control rods shall be demonstrated OPERA 8LE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.5 and 4.1.3.7.

              *May be rearsed intermittently, under administrative control, to permit tasting l gg associated with restoring the control rod to OPERABLE status.

~

             *"These valves may be closed intermittently for testing under administrativa ls control.                                                                                                                                    .

LA SALLE - UNIT 1 3/4 1-4 Amendment No. 18 er & --.~ ..... ........r. ._ ... . l

   .                                                                                                                                                                               i
.C REACTIVITY CONTROL SYSTEM SURVEILLANCE REQUIREwENTS (Continued)

The scram discharge volume shall be determined OPERA 8LE by 4.1.3.1.4 demonstrsting: a. The scram discharge volume drain and v4.-2 valves OPERA 8LE, when control rods are scram tested from a nomal control rod configura-tion of less than or equal to 50% RCD DENSITY at least once per l 18 months" by verifying that the drain and vent valves:

1. Close within 30 seconds after receipt of a sigt.a1 for control rods to scram, and

2. Cpen after the se' ram signal is reset. -

D. Proper float response by performance of a CHANNEL FUNCTIONAL TEST . of the scram discharge volume scram and control rod block level instrumentation after each scram from a pressurized condition. t "The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 provided the surveillance is performed within 12 hours after achieving less than or equal to 50% RCD DENSITY. I

                                                                                                            .                                                                      l

( .' 1 l l Amendment No. 18-3/4 1-5 l

                . LA SALLE - UNIT 1                                                                                                                                             l
      -- . . -     .    ~ .....-           .

i

  .-         .                                                                                                                                  1 POWER DISTRIBUTION l.IMITS POWER DISTRIBUTION LIMITS                                                                               -
                                                                                                                                     ~

3/4.2.3 MINIMUM CRITICAL PCWER RATIO LIMITING CCNOITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit detamined fres Figure 3.2.3-1 times the Xfdetermined free Figure 3.2.3-2 for two meirculation loop operatio:1 and shall be equal to or greater than the MCPR limit detamined from Figure 3.2.3-1 + 0.01 times the Xf detemined fres Figure 3.2.3-2 for single recirculation loop operation provided that the end-of-cycle recirculation pump trip (ECC-RPT) system is OPERABLE per Specification 3.3.4.2. APPLICABILITY:

           . OPERATICNAL CCNDITICN 1, when THEMAL POWER is greater than or equal to 25% of RATED THEMAL POWER.

ACTION

a. With the end-of-cycle recirculation puso trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided taat, within 1 hour,

( MCPR is datamined to be equal to or greater than the MCPR limit shown in Figure 3.2.3-1 ECC-RPT inoperable curve, times the Xfshown in Figure 3.2.3-2. ,

~~

b. With MCPR less than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2, initiata corrective action within

                                 -15 minutes and restore MCPR to within the required ifmit within 2 hours or reduce THEPAL POWER to less than 25% of RATED THEVAL POWER within the next 4 houp.

SURVEILLANCE RECUIREMENTS 4.2.3 MCPR, with: a. t"** = 0.86 prior to perfomance of the initial scram time sensurements for ttie cycle in accordance with Specification 4.1.3.2, or

b. t,y, determined within 72 hours of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be detamined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2;

a. At least once per 24 hours,

b. Within 12 hours after cesoletion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and I( c. Initially and at least once per 12 hours when the reactor is operating L with a LIMITING CONTROL RCD PATTERN for MCPR.

LA SALLE - UNIT 1 3/4 2-4 Amendment No. 18

     = = =
                       --                    . r- ,-,   ,        - - - , , - - . - - - . - - . - , -   ,-------w            -    -   - - -,

  ~          _    _ . . . . . . _                      _                  _                            __                                                                    _
                                                                  +

e. , :

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1.40 '" I 3 i a !! 2: sa H -4 H 3 e

                                                                                                                                                          ,,,, -         ij;
                                                                                                                                                   ~~,e**                C

] 1.36 , _ d

                                                                                                                                     ' , ,                               o Z

EOC-RPT inorierable ,,,,,-

                                                                                                                '                                                        [

5 . 1.30 y ,,,,=' 1.30 s E .-

o p

e lE sn ... EOC-RPT Operable

                                              ...-                                                                                                        /
'                                                                                                                                               /           ,

1.26 125

                                                                                                                   -        /

1.20 1.20 Y. I S .736.74 .76 .76 .77 .78 .79 .81 .82 l

        #                                                                                          .80                           .83       .84         .85     .88 i

g T u Figure 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T AT RATED FLOW On t , e

j

                             .  .. 0 .                ..     ..... ..              ..                                                                                      -                                                              *          ^*

(, i i a

                            ,                  i .n .

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e- N n e- . 1 se 3 , .in N - - - - -- - '

                                                                                                                                                                                                                                                                                   'll e

c , ;u i. 5 lc flow Control K gCurve o . a .

                                                                                                          / - Automat          .                                                                                    .

5

e . o

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I su l c-s . 9 .

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                                                                                                                                                                                                                                                                                   <n 3.?u                                                    -

s., w - N y 2 Hanual f, low Control K g Curve #

u. .

g,go . . . . . . . .. - . - . . . . - - - - _ > - . - . . . . . . . . l i I

                                                                                                                                                                                                        \        s
                                                                                                                                                                                                                                         \
                                                                              "                                                                                                                                       s N                .,

I.no *

                                                                                                                         - - - -     - - - - - - - - - -     ---             - - -     - - - - - - - - - - - -            - - - - - - - - - -       --  'w 8                                                                                                                                                 .
                                                                        .                                                                                                           e
                                                                                                     -l

g *

                                                                          .                                                        s                       e j                                           n .su ?il                         in                      40        ,

Sti dio-- 71 an s- = - - - - - - - , 9n lisci Core Flow, 1 of pateil Core flow Ng l'AC10R I . fIgore 3.2.3-2 f

                                                                                                                                                                          ..                     .                                  8            4 4

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION IMITING CONDITION FOR OPERATION 3.3.1 As a ninfaua, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESP 0.4SE TIME as shown in Table 3.3.1-2. APPLICABILITY: As shown in Table 3.3.1-1. ACTION:

a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channels and/or trip system in the tripped condition" g within 1 hour. The provisions of Specification 3.0.4 are not applicable;

b. With the the number of OPERABLE channels less than required by the Minimum CPERABLE Channels per Trip System requirement for both trip systems, place at least one trip systes** in the tripped condition within 1 hour and take the ACTION required by Table 3.3.1-1. '

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERA 81.E by the' perfomance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1 1. 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS ar.d simulated automatic operation of all channels shall be perfomed at least once per 18 months.*** 4.3.1.3 The REACTOR PROTECTION SYSTEM R'ESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 sonths. Each test shall include at least one channel per trip systes such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system. a With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the noperable channel shall be restored to OPERA 8LE status within 2 hours or the ION required by Table 3.3.1-1 for that Trip Function shall be taken. na - If more channels are inoperable in one trip systes than in the other, select that trip system to place in the tripped condition, except whenithis would cause the Trip Function to occur. l[ ***The specified 18-month interval may be waived for Cycle 1 provided the

\_              surveillance is performed during Refuel 1, which is to commence no later than October 27, 1985.

LA SALLE - UNIT 1 3/4 3-1 Amendment No. 30

                           ,    ,     -   , -         , - - - . , - . - - - ~ ,,,_.,,,             ,---.nm_,   ..,,,,..,-__v     ,--.,-,a   , - , . , _ , - , , , ,

f q. . (' - ( i I t TABLE 3.3.1-1 . C g REAC10R PROTECil0N SYSIfM INSTRilHENIATION . h . APPLICABLE MININUM GPERABLE

           ,                                                                           OPERAll0NAL          CllANIKIS PtR c       futiCI.I_nflAL UNIT
                          . .                                                          CONii_l_liatiS      1RIPSYSitM(a)   ACTION 2

U l. Intcomediate Range Monitors: ,

         "               a.         Neutron flux - liigli                              2                          ,3          1 3, 4                        2          2 5(b)                        3          3               .
!                        Is.        Inoperative                                        2                           3         ~1                                          -

3, 4 2 2 u i 5 3 3 1

2. ' A<crage Power R.snge Monitor:(c) u ... Neutron flux - 1119:. 8 Seldown 2 2 1 g '~~

3 2 2

         <;a                                                                           5(b)                        2          3 N               h.         Flow Blased Simulated Thermal Power-upscala                                      1                           2          4

c. Fixed Neutron flux-HI0li 1 2 4 i d. Inoperative 1, 2 2 1

                                                       .                               3                           2          2 5                           2          3
                                     ~

3. ke.n: tor Vessel Steam Some

! l'ressure - liigh 1,2(d) 2 1

4. Re.ictor Vessel Water level - Low, level 3 1, 2 2 I

5. Hain Steam Line Isolation Valve -

l Closure II *I 4 , 4 i'

6. H.iin Steam tine Radiation - 3 liigh l.

       .                                                                               1. 2(d)              y 2           5 i
 ,                                                               i i

D............---. , D. - - D-

                                                  )                                                     .        t                                              s
                                      ,._                                                            1 Ante 3.3.1-1 (Continued)                                        -

2

u. RfACIOR PRolfCIl0N SYSifM INSIRUNfMIATION N

s- APPLICAntE MINIHilH OPERABLE HPIRAll0NAL CilANHLLS PLR , fuMCipilAL UNIT COHillIIONS TRIPSYSl(H(a} , ACTION

7. I'rimary Containeen Pressure e liigli ' 1, 2 III 2I 'II ,

1

                                     -.      8.       teram Olscharge Volimie Water l evel - liigh                              I                          2                1              .-       !

5gg'3, 2 3

9. lurbine Stop Valva - Closure I III 4 III 6

10. ~lurhine Control Valve fast Closure, Valve Trip System all Pressure - Low Igg) 2g j) 6 ,

s" 11. R...ctor Mode Switch Shutdown Position I, 2 l 1

7 . . 3. 4 1 7 u 5 1 3

12. fl.uiual Scram I, 2 1 1 3, 4 1 8 i 5 1 9 e .

1 l . 4 s I 1

C 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 ECCS divisions 1, 2 and 3 shall be OPERABLE with:

a. ECCS division 1 consisting of:

1. The OPERA 8LE fow pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

2. The OPERA 8LE low pressure coolant injection (LPCI) subsystes "A" of the RHR system with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

3. At least 6 CPERA8LE** ADS valves.

b. ECCS division 2 consisting of:

( 1. The OPERA 8LE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR systes, each with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

2. At least 6 OPERABLE ** ADS valves.

1 l c. . ECCS division 3 consisting of the OPERA 8LE high pressure core spray (HPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel. APPLICABILITY: OPERATIONAL CONDITION 1, 2*# and 3*.

                                                                                ~
                  "The ADS is not required to be OPERA 8LE when reactor steam dose pressure                                                                                                 .

l 1s less than or equal to 122 psig. l **5ee Specification 3.3.3 for trip system operability. y

               #see Special Test Exception 3.10.6.
            , LA SALLE - UNIT 1                                                     3/4 5-1                          Amendment No. 29 P
                    - - _            _. - _ , , ~ . , . . - . . ~ ~ _ . . ,       ,       _.,.,,__,._,ym.,._._ _- _ , _ _ _   r. mn . , , . . , , . , , , , , , , , , . ,  y_,gw,,,,,,__ym-

_ ..m.. n... ..._u. . x...~. _._ m .. _, EMU GENCY CORE CCOLING SYST B S N LIMITINGCONDITICNFdROPGATICN(Continued) r

     .                      ACTION:                      ,

s. For ECCS division 1, provided that ECCS divisions 2 and 3 are OPE.98LE:

1. With the LPCS systas inoperable, restore the inoperable LPCS

    .                                        system to CPE.U8LE status within 7 days.

2. With LPCI subsystem "A" inocerable, res*Jre the inoperable LPCI subsystem "A" to CPE.USLE status witnin 7 days.

3. With the LPCS systas inoperacle and LPCI subsystas "A" inoperable, restore at least the inoperable LPCI sucsystem "A" or the inoperacle LPCS system to OPGABLE status witnin 72 hours. -

4. Otherwise, be in at least HOT SHUTCCWN within the next 12 hours and in COLD SHUTCChN within the following 24 hours.

b. For ECCS division 2, provided that ECC5 divisions 1 and 3 are OPERA 8LE: ,

1 1. With either LPCI subsystas "B" or "C" inonerable, restore the

_ inoperable LPCI subsystas "B" or "C" to OPGASLE status within 7 days.

2. With both LPCI subsystams "S" and "C* inoperable, restore at least

      ,                                      the inoperable LPCI suesystas "B" or "C" to CPE. V 8LE status within 72 hours.

3. Otherwise, be in at least HOT SHUTCCWN within the next.12 hours

                                            ,and in COLD SHUTC0hN,within the following 24 hours".

c. For ECCS division 3, provided that ECC5 divisions 1 and 2 and the

      ,                                RCIC systaa are OPE.UBLE:

L With'ECCS division 3 inocerable, restore the inoperable division .

to CPGA8LE status within 14 days.

2. Otherwise, be in at least HOT SHUTDOWN wi, thin the next 12 hours --

and in COLD SHUTC0hN within the following 24 hours. . t ~

       !                    "Whenever two or more 2HR subsystams are inoperable, if unable to attain COLD l                     SHUTDChN as required by this ACTION, maintain reactor coolant temperature as -                    -

'- low as practical by use of alternata heat removal methods. I r , .

( i LA SALLE - UNIT 1 3/4 5-2 .

  -m-     -
                ,                                + - ,

            .          ..                   /          .                   .

EMERGENCY CORE CCOLING SYSTEMS I LIMITING CONDITION FOR OPERATION (Continued) , I ACTION: (Continued)

d. For ECC5 divisions 1 and 2, provided that ECCS division 3 is OPERABLE: !

i

1. With LPCI subsystas "A" and either LPCI subsystas "S" or "C" inoper- i able, restore at least the inoperable LPCI subsystem "A" or inoper .

able LPCI subsystas "S" or "C" to CPERA8LE status within 72 hours.

2. With the LPCS systas inoperable and either LPCI. subsystems "B" or "C" inoperable, restore at least the inoperable LPCS system or inoper- !

able LPCI subsystem "B" or "C" to OPERA 8LE status within 72 hours.*

3. Othe mise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTCOWN within the following 24 hours".

e. For ECCS divisions 1 and 2, provided that ECCS division ~3 is !

OPERAELE and divisions 1 and 2 are othemise OPERABLE:

1. With one of the above required ADS valves inoperable, restore the '

i inoperacle A05 valve to CPERA8LE status within 14 days or be in at least HOT SHUTDCWN within the next 12 hours and reduce reactor I staan dose pressure to f,122 psig within the next 24 hours.

2. With two or more of the above required ADS valves inoperable, -

be in at least HOT SHUTCOWN within 12 hours and reduce reactor staan does pressure to < 122 psig within the next 24 hours. ( f. With an ECC5 discharge line % eep filled" pressure alarm instrumenta-tion channel inoperacle, perform Surveillance Requirement 4.5.1.a.1 at least once per,24 hours.

g. With an ECCS header delta P instrumentation channel inoperable, restore the inoperable channel to CPERA8LE status within 72 hours or detarsine ECCS header delta P locally at least once per 12 hours; otherwise, declare the associated ECCS inoperacle.

h. With Surveillanca Requirement 4.5.1.d.2 not perfomed at the required intarval due to low reactor staan pressure, the provisions of Specifica-tion 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequata to perform the tast.

i. In the event an ECCS system is actuated and injects water into the

Reactor Coolant Systas, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.6.C within 90 days describ-ing the circumstances of the actuation and the total accumulated actua- -

i tion cycles to date. 'The. current value of the usage factor for each l affected safety infection nozzle shall be provided.in this Special .. Report whenever its value exceeds 0.70.

j. With one or more ECCS corner room watertight doors inoperable, restore all the inoperable ECCS corner room watertight doors to OPERABLE status within 14 days, otherwise, be in at least HOT SHUTDOWN within the next

                                 ,12 hours and in COLD SHUTDOWN within the following 24 hours.

l "Whenever two or more RHR subsystems are inoperable, if unable to attain COLD ' l f SHUIDOWN as required by this ACTION, saintain reactor coolant.tamperature as ( low as practical by use of alternate heat removal methods. LA SALLE - UNIT 1 3/4 5-3 Amendment No. 13

                                         '                                                                                                                                        ~~~
                                                                                                                         .                           - -- - --           .~~~

v - 3/4.6 CCNTAINMENT SYSTEMS

                                                                 - ' ~ ~

3/4.6.1 PRIMARY CCNTAINMENT - PRIMARY CCNTAINMENT INTEGRITY LIMITING CCN0! TION FOR OPERATICM

3. 6.L 1 PRIMARY CCNTAINMENT INTEGRITY shall be maintained.

APOLICABILITY: OPERATIONAL CONDITIONS 1, 2,* and 3.# I ACTION:

                     'dithout PRIMARY CCNTAINMENT INTEGRITY, restore PRIMARY CCNTAINMENT INTEGRITY

within 1 hcur or be in at least HOT SHUTCChN within the next 12 hours and in COLD SHUTUChN within the following 24 hours. ,

i SURVEILLANCE REOUIREMENTS l 4.6.L1 PRIMARY CCNTAINMENT INTEGRITY shall be demonstrated: . ( v.

a. After each closing of each penetration subject to Type 8 tasting, excant the primary containment air locks, if opened following Type A or 8 test,-by leak rata tasting the seal with gas at Pa, 39.5 psig, and verifying that when the seasured leakage cata for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.L2.d for all other Type 3 and C penetrations, the connined leakage rata is less than or equal to 0.60 La.

b. At least once per 31 days by verifying that all primary containment penetrations ** not cacaole of being closed by OPERABLE c:ntainment

     ,                            automatic isolation valves and required to be closed during ac:ident
     .                            conditions are closed by valves, blind flanges, or' deactivated i    -                       aut:matic valves secured in position, except as provided in Table 3.6.3-1 of Specification 3.6.3.                                                                                            .-

c. By verifying each primary containment air lock OPERABLE per

     #                            5pecification 3.6.L3.
        .                      d. By verifying the suppression chamber CPEPABLE per ' Specification                                                                                                              _

3. 6. 2. L l i

      .                "See special Test Exception 3.10.1
      .               **Except valves, blind flanges, and deactivated aut:matic valves which are
                .        located inside the conttinsent, and are locked, sealed or othemise secured in the closed cesition. These pe s*. rat- s s.'all te ::rffino ci: sad duri ;

l each CCL3 SHUT:C*C: a::c2:t sucn veH f f:-- :n nede act a perf: ac nen - e ! pri rar/ centai..::n: ..:s net been cei.uru sic:a t.a ::s: verifica .icn :r . s more often than once per 92 days. l (,

                       #5ee Special Test Exception'3.10.7.                                                                                                                            -

LA SALLE - UNIT 1 3/4 6-1

   -               -                      . . -,                                   _ _ _ - - , _ _ - _. , . _ , , - _ _ _ . , - - . - .. - , __,. _ - ...,,.                           w .., - -,,y.--m, , , , - - -

CONTAINMENT SYSTEMS SUPPRESSICN POOL SPRAY ' LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of: 3

a. OneOPERA8LkRHRpump,and .

b. An 0PERABLE flow path capable of recirculating water from the suppression chamber.

' APOLICABILITY: OPERATIONAL CCNDITIONS 1, 2, and 3. . ACTION:

a. With one suopression pool spray loop inoperable, restore the incoerable loop to CPERABLE status within 7 days or be in at least HOT SHUTDCWN within the next 12 hours and in COLD SHUTDOWN'within the following 24 hours,

b. With both suppression pool spray loops inoperable, restore at least one loco to CPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN" within the C following 24 hours.

SURVEILLANCE RECUIREMENTS

4. 6. 2. 2 The suppression pool spray mode of the RHR system shall be demonstratad CPERABLE:

l

a. At least once per 31 days by verifying that each valve (manual, l

j power operated or automatic), in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b. By verifying that each of the required RHR pumps develops a flow of at least 450 gpa on recirculation flow through the suppression pool -

spray sparger when testad pursuant to Specification 4.0.5. i l l "Whenever octn AHR subsystems are inoperable, if unable to attain COLD SHUTDOWN l as required by this ACTICN, maintain reactor coolant temperature as low as l practical by use of alternata heat removal methods. l. LA SALLE - UNIT 1 3/4 6-20 . Amendment No. 18

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

CONTAINMENT SYSTEWS SUPoRESSION POOL COOLING . LIMITING CONDITION FOR OPERATION 1 3.6.2.3 The suppression pool cooling mode of the residual heat removal (RNR) system shall be OPERABLE witt two independent loops, each loop consisting of:

a. One OPERABLE RHR pump; and

b. An CPERABLE flow path capable of recirculating water from the suppression chamber througn an RHRSW heat exchanger.

APoLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:

a. With one suppression pool cooling loop inoperable, restore the inoperable loop to CPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

With both suppression pool cooling locos inoperable, be in at least b.

f HOT SHUT 00WN within 12 hours and in COLD SHUTDOWN" within the next 24 hou s.

l( SURVEILLANCE REQUIREWENTS 4.6.2.3 The suoeression pool cooling mode of the RNR system shall be demonstrated CPERABLE:

I

a. At least once per 31 days by verifying that each valve (manual, l power operated or automatic), in the flow path that is not locked, sealed or othemise secured in position, is in its currect position.

 .                              b:       By verifying that eactr of the required RHR pumps develops a flow of at least 7200 gpa on recirculation flow through the RHR heat exchanger and the suppression pool when tastad pursuant to                                        .

Specification 4.0.5.

                        %nenever notn AHA subsystems' are inoperable, if unable to attain COLD SHUTDOWN as recuired by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

LA SALLE - UNIT 1 3/4 6-21 Amendment No. 18

                                                                       - - - - . -                       . . . - . ,            .--..n-=     -. , .-, , _ . , - , . - .

1

                                                                                                            .                                     \
                     .                                                       .                                                                    i C                            CONTAINMENT SYSTEMS                            '

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES I LIMITING CONDITION FOR OPERATION 3.6.3 The primary containment isolation valves and the reactor instrumentation line excess flow check '.'alves shown in Table 3.6.3-1 shall be OPERABLE with isolation times less than or equal to those shown in Table 3.6.3-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:

a. With one or more of the primary containment isolation valves shown in

                                       -    Table 3.6.3-1 inoperable:

1. Maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours either; ,

a) Restore the inoperable valve (s) to OPERABLE status, or b) Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position," or . c) Isolate each affected penetration by use of at least one ! closed manual valve or blind flange." The provisions of Specification 3.0.4 are not applicable d) pro.vided that within 4 hours the affected penetration is isolated in accordance with Action a.1.b) or a.1.c) above, and provided that the associated system is declared inoper-able, if applicable, and the appropriate action statements for that system are performed.

2. Otherwise, be in at11 east HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

b. With one or more of the reactor instrumentation line excess f1'w o I check valves shown in Table 3.6.3-1 inoperable:

1. Operatio'n may continue,and the provisions of Specifications 3.0.3 and 3.0.4 are not applicable provided that within 4 hours either:

a) The inoperable valve is returned to OPERABLE status, or b) The instrument line is isolated and the associated instrument is declared inoperable.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

                                  " Isolation valves closed to satify these requirements may be reopened on an intermittent basis under administrative control.

l 1

                   ,           LA SALLE - UNIT 1                                       3/4 6-22           Amendment No. 26 i

                                                                                                                          ....a     .
                                                                                                                                                                                .. .:. . . : a . .r: . a V
                      ]                                   CONTAINMENT SYSTEw6 t

j $URVEILLANCE RECUIREMENTS i. 1 4.6.3.1 Each primary containment isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERA 8LE prior to returning the valve to service after.mainte-

             '                                            nanca, repair or replacament. Work is performed on the valve or its associated

actuator, control or power circuit by cycling the valve through at least one complete cycle of. full travel and verifing the specified isolation time.

4. 6. 3. 2 Each primary containment automatic isolation valve shown in .

Table 3.6.3-1 shall be demonstrated CPERA8LE during C::LD SHUTUChN or RE.:UELING at least onca per 18 months by verifying that on a containment isolation tast signal each automatic isolation valve actuates to its isolation position. 4.6.3.3 The isolation time of each primary containment power operated or . automatic valve shown in Table 3.6.3-1 shall be detarsined to be within its - limit when tasted pursuant to specification 4.0.5. 4.6.3.4 Each reactor instrumentation Ifne excess flow check valve shown in

               !                                          Table 3.6.3-1 shall be demonstrated CPE.4A8LE at least once per 18 months by                                                                        .

j verifying that the valve checks flow. . 4.6.3.5 Each traversing in-core prece systas explosive isolation valve shall v be demonstrated OPERA 8LE:

a. At 7 east onca per 31 days by verifying the continuity of the explosive charge.

I b. At Isast once per 18 months by removing the explosive squib from at least one explosive valve such that the explosive squib in each

;                                                                       explosive valve will be tastad at least once per 90 months, and 4                                                           initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the
                   .                                                    one fired or from another batch which has been certified by having j                                                       at least one of that batch successfully fired. No explosive.Jguib
;               g shall remain in use beyond the expiration of its shelf-life and i         ,

operating-life.

,               1 i
,                I l           .                                                                                                                                                                        ,

(j I 1 LA SALLE - UNIT 1 3/4 6-23

    -_-.y-_           . . - _ - - . , . _ , , _ . . , _ .                      . , , , , _ . . . _ _ _ - .         _                  .-,___,,,,,,__.,,,,,,_.,.~.,_.m,                               ,    .m...,,_   _.,          ,,m,

O *

                                                                                             ;       D                                         D     -

O ) TABLE 3.6.3-1 3 5- . p . PRIMARY CONTAINMENT ISOLATION VALVES J r m MAXIMUM , 1 a ISOLATION TIME q g VALVE FUNCTION AND NtHIER , VALVE GROUP (a) (Seconds)

                          '                                 ~

g a. Automatic Isolatlon Valves

1. Main Steam Isolation Valves .

1 5* l l 1821-F022A, B, C, D . 1 1821-F028A, B, C D

2. Main Steam Line Drain Valves 1 l

1821-F016 < 15 1821-F019 . 2 15 I 1B21-F067A,B,C,D(b) , 7 73 l g 3. Reactor Coolant System Sample 1 D Line ValvesI 'I - 3 15 l ] m . 1833-F019 - i e's 1833-F020

4. Drywell. Equipment Drain Valves 2 1RE024 < 20 1RE025 i 20 1RE026 2 15 1RE029 2 15

5. Drywell Floor Drain Valves 2 7 20 IRF012 1RF013

6. Reactor Water Cleanup Suction Valves 5 5 30 1G33-F001(d) i 1G33-F004

7. RCIC Steam Line Valves 8

{ [ ~ IE51-F008 I 'I ) h. $ 20

                                                                                                                         < 15 IE51-F063 1E51-F064 III                              -

1 15 g

.o IE51-F076 1 15
= -: f. O O '; g. .

5 ] TABLE 3.6.3-1 (Continued) s

PRIMARY CONTAINMENT ISO'ATION L VALVES i;;

MAXIMUM i e ISOLATION TIME 1 5

          -4 VALVE FUNCTION At:D NUMBER                               VALVE GROUP (a)                                  (Seconds)

H Automatic Isolation Valves (Continued) ! 8. Containment Vent and Purge Valves 4 l IVQO26 - -

                                                                                                                           < 10*8 IVQO27                                                                                      7 10**

IVQO29 7 10** . IVQO30 7 10** 4 IVQO31 7 10** IVQO32 75

        ' w                    IVQO34                                  ~

7 10**

          )                    IVQO35                     .

75

          .                   IVQO36                                                                                       7 10**

4 IVQ040 7 10** IVQ042 7 10**

IVQ043 7 10**

IVQ047 . 75 IVQ048 75 IVQ050 75

                         ' IVQ051                                                                                          75 IVQ068                                                                                        75

9. RCIC Turbine Exhaust Vacuum Breaker 9 N.A.

Line Valves 1E51-F080 - '. 1E51-F086

10. LPCS, HPCS, RCIC, RHR Injection Testable Check Bypass Valves I9) 2 N. A. l g IE21-F333 g IE22-F354 .

IE12-F327A, 8, C g IE51-F354 IE51-F355 .

SI 1

O i .

TABLE 3.6.3-1 (Continued) g . PRIMARY CONTAINMENT ISOLATION VALVES r-E MAXIMUM s ISOLATION TIME VALVE GROUP I ")

g VALVE FUNCTION AND NUMER ,. (Seconds) [ Automatic Isolation Valves (Continued)

11. Containment Monitoring Valves 2 -<5 :

ICM017A,8 ICM018A,8 - ICM019A,8 ICM020A,8 ' I ICM0218((h) ICM022A(h) i ICM025A h) R* 1CM026B(h) ICM027 T

ICM028 ,

M 1CM029 ICM030 ICM031 1CM032 . ICM033

ICM034: 12. Drywell Pneumatic Valves 2 ,

11N001A and B < 30 11N017 7 22 IIN074 7 22 11N075 2 22

i. '

IINO31 -75 (

13. RHR Shutdown Cooling Mode Valves IE12-F008 6

                                                                                                               < 40                        '

I 1E12-F009 5 40

      =           IE12-F023                                   -
                                                                                                               < 90 1E12-F053 A and'B                                  .                                         I 29 1E12-F099A and B I9)I'}
                                                 ;                                                             -< 30 i      M                                                                             -

i 1 , l

p - m m. . TABtE 3.6.3-1 (Continued) g PRIMARY CONTAlldiTiill50LATION VALVES HAXIMUM ISOLATION TIME h VALVE FUNCTION AND NUMBER VALVE GRollPg,) (Seconds) Autematic Isolation Valves (Continued) e '

14. Tip Guide Tube Dall Vahes (Five Valves) 7 N.A. l IC51-J004

15. Reactor Guilding C1nsed Cooling Water System Valves 2 1 3d IWR029 IWR040 IWR179 ,

Ik'RIGO

15. Primary Centainment Chilled R

Water Inlet Yalves 2 ; IVP113 A and B < 90 T IVP063.A and B - . 7 40 ti 17. Primary Containment Chilled Water Outlet Valves 2 IVP053 A and B < 40 l IVP114 A and B 7 90

18. Recirc. llydraulic Flow control Line Valves IG) 2 55 IB33-F338 A and B 1833-F339 A and B 1833-F340 A and B 1833-F341 A and B ,

IB33-F342 A and B E IB33-F343 A and B E 1833-F344 A and B

  &           IB33-F345 A and B
   $     19. Feedwater Testable Check Valves                    2 N.A.

1821-F032 A and D

I?, I sus er - 3

b. O O '

v. .

1

\                           .                                                                                                                                       .

I j TABLE 3.6.3-1 (Continued) j 5 l m . PRIMARY CONTAINNENT ISOLATION VALVES l > F ^ m MAXIMUM e ISOLATION TIME e VALVE FUNCTION AND NUMBER , VALVE GROUP (a) (Seconds)

5. .

b. Manual Isolation Valves e

1. 1FC086 N.A.: 2. 1FC113 N.A.

! 3. 1FC114 - N.A. i 4 .~ 1FC115(g) N.A.

5. N.A.
6. 1HC027(3)

IMC033 N.A.

7. ISA042 N.A.

8. ISA046 N.A.

M s -

W

. 2 5 i 1 , . 4

CONTAINMENT SYSTEMS STANOBY GAS TREATMENT SYSTEM . LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.# - APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. , I ACTION:

a. With one stancby gas treatment subsystem inoperable, restore the inoperable subsystem to CPERABLE status within 7 days, or:

1.' In OPERABLE CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN l within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.-

2. In Operational Condition *, suspend hand 11ng of irradiated fuel in the secondary containment, CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel. The ~

provisions of Specification 3.0.3 are not applicable. l( b. With both standby gas treatment subsystaos inoperable in Operational Condition *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potantial for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIRE 5aENTS , l 4.6.5.3 Each standby gas treatment subsystem shall be demonstratad OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorcers and verifying that the subsystem operatas for at least 10 hours with the heaters OPERABLE. _

I "When irraciataa fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. I The nomaI or emergency power ' source may be inoperable in Operational Condition 8 , LA SALLE - UNIT 1 3/4 6-40 Amendment No. 18

CONTAINMEAT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory

. Positions C.S.a. C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm 210L

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

3. Verifying a subsystem flow rate of 4000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation by verifying .

within 31 days after removal that a laboratory analysis of a f representative carbon sample obtained in accordance with Regulatory 1' Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than or equal to 8 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm 2 101

2. Verifying that the filter train starts and isolation dampers open on each of iM following test signals:

a. Reactec huilsing exhaust plenum radiation - high,

b. Dry e~i prt iure - high, -

l

c. Reactor vessel water level - low low, level 2, and -

d. Fuel pool vent exhaust radiation - high.

 ~
                                 ~3. Verifying that the heaters dissipate 23 t 2.0 kw when tested in                                                                            l accordance with ANSI N510-1975. This reading shall include the i

appropriate correction for variations from 480 volts at the bus. LA SALLE - UNIT 1 3/4 6-41 Amendment No. 21

CONTAINMENT SYSTSS SURVEILLANCE REOUIREMENTS (Continued)

     !                                e. After each completa or partial replacement of a HEPA filtar bank by verifying that the HEPA filtar banks remove greater than or equal to j                                   99% of the 00P when they are tasted in-place in-accordance with ANSI i
              .                          N510-1975 while operating the systas at a flow rata of 4000 cfm 210%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorters remove greatar than or equal to 99% of a halogenated hydrocaroon refrigerant test gas when they are tasted in-place in accorcance with ANSI N510-1975 while operating the system at a flow rata of 400.0 cfm a 1C%.

 . f' Q

( .

e I

s r l -

l l i 4 l 1 l.

      }
               /     _                                                                                                                        l
                                                                                                                   .                          \

LA SALLE - UNIT 1 3/4 6-42

                                                                                  - .-     <.. - . .         .                                1 l

CONTAINMENT SYSTEMS 3/a.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL DRWELL AND SUPPRESSION CHAM 8ER NYOROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent drywell and suppression chamber hydrogen recombiner systems shall be OPERA 8LE. APPLICABILITY: OPERATIONAL CONDITIONS I and 2. . ACTION: With one drywell and/or suppression chamber hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REOUIREMENTS 4.6.6.1 Each drywell and suppression chamber hydrogen recombiner system shall be demonstrated OPERABLE:

a. At least once per 92 cays by cycling each flow control valve and

                   .                recirculation valve through at least one completa cycle of full                                         '

travel.

b. At least once per 6 months by verifying, during a recombiner system

functional tast: ,. ,

'(                                  1. That'th'e heaters are OPERABLE by determining that the current in each phase differs by less than or equal to 5% from the other phases and is within 5% of the value observed in the original acceptance test, corrected for line voltage differences.

2. That the reaction chamber gas tamperature increases to 1200 2 25'F

      .                                    within 2 hours.

1

c. At least once per 18 aanths by:

1. Performing a CHANNEL CALIBRATION of all recombiner operating instrumentation and control circuits.

2. Verifying the integrity of all heater electrical circuits by

             .                             performing a resistance to ground test within 30 minutes following
                     -                     the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 100,000 ohms.

d. By measuring the leakage rite: ,

1. As a part of the overall integrated leakage rate test required by Specification 3.6.1.2, or

2. By measuring the leakage rate of the system outside of the containment isolation valves at P , 39.6 psig, on the schedule required by Specification 4.6.1.2 and including the measured

leakage as a part of the leakage determined in accordance with ' Specification 4.6.1.2. l LA'SALLE - UNIT 1 3/46[Og Amendment / I g

                *                                                       --         --                --   ~ ~~ "" "

1 1

                                            . . . .           .                  ~ . -        -                         -         '-
                                                                                                                                                                                                                              - =-

C - PLANT SYST215

3/4.7.2 CCNTROL ROCM AND AUXILIARY ELECTRIC EQUIPMENT RCCM 98ERGENCY j EILIRAi10N 5f5i-;1 j LIMITING CCNOITION FOR OPERA ION 3.7.2 Two independent control roos and' auxiliary electric equipment roos , emergency filtration systes trains shall be OPERABLE.# APPLICABILITY: All CPERATIONAL CONDITIONS and *. ACTICN:

a. With one e'mergency filtration system train inocerable, restore the inoperable train to OPERABLE status within 7 days or:

1. In OPERATICNAL CCNDITICNS 1, 2, 3, be in at least HOT SHUTDCWN within the next 12 hours and in COLD SHUTDCWN within the '

following 24 hours.

2. In OPERATIONAL CONDITION 4, 5 or *, initiata and maintain '

operation of the OPERABLE emergency filtration systas. in the pressurization acce of operation. ,

b. With both emergency filtration system trains incoerable, in

 .v                                          OPERATIONAL CCNDITION 4, 5 or *, suscend CORE ALTERATICNS, handling of irradiated fuel in the seconcary c:ntainment and operations with a potantial for draining the react:r vessel.

c. The provisions of Specification 3.0.3 are r.at applicable in Operational Condition ".

I SURVEILLANCER$0UIRE58ENTS i

    !              4.7.2 Each cartrol room and at.xiliary electric ecuioment room emergency
    !              filtration systas train shall be demonstrated GPERA8LE:

b a. At least once per 31 days on a'. STAGGERED TEST BASIS by initiating, from the control room, flow througn the HE?A filters and charcoal adsorbers and verifying that the train operates for at least i

    ;                                         10 hours with the heatars OPERABLE.                                                                   -

I i

                        "wnen irraciar.ac fuel is being handled in the secondary containment.

The noraal or emergency power source say be inoperable.in OPERATIONAL

    ;                        CCNOITICN 4, 5 or 3

( .:

               ,   LA SALLE - UNIT 1                                                                 3/4 7-4                                                                                -
           ,        , , - .         ___.._y          , _ _ _ _ _ . _ . _ , _ .         ,                  _ _ _ -         _ . _ . _ . _ . , . . . _ , , , , , _ , , . . - . . , _ _ - . .             ,_.._--.,--.,-,_,m.

            ,b          , PLANT SYSTE.us SURVEILLANCE REGUIREMENTS (Continued)

b. At least once per l8 months" or (1) after any structural maintenance on the HEPA filter or charcoal adsorter housings, or (2) following 4 painting, fire or chemical re? nase in any ventilation zone communicating with the train by:

l ,

1. Verifying that the train satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a. C.S.c ~and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rata is 4000 cfm 2 1C%.

2. Verifying within 31 days after removal that a laborator/

analysis of a representative caroon sample obtained in. - accordance with Regulatory Position C.6.b of Regulatorf

Guida 1.52, Revision 2, March 1978, seets the lacoratory . tasting critaria of Regulatory Positien C.S.a of Regulatory

Guide 1.52, Revision 2, March 1978.

     ,                                                                                 ~

3. Verifying a train flow rata of 4000 cfm + 10% during subsystem

( ' c. operation when tasted in accordance with~ ANSI M510-1975. After every 720** hours of charcoal adsorter operation by verifying u within 31 days after removal that a laboratorf analysis of a representative carton sample obtained in accordance with Requiatory Positan C.6.h of Regulatory Guide 1.52, Revision 2, Maren 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guida 1.52, Revision 2, March 1978.

d. At least once per 18 months Iby:

1. Verifying that the pressure drop across the combined HEPA

                              "                    filters and charcoal adsorcer banks is less than 8 inches Water                  i i                                            Gauge while operating the train at a flow rate of 4000 cfm                       '

2 10%.

2. Verifying that on each of the.below recirculation mode actuatinn -

test signals, the recirculating charcoal filter automatically switches to the recirculation sede of operation and the isolation - j dampers close within 6 seconds:

       !                                           a)     Chlorine detection, and b)     Ammania detection.
                                                                                                                    ~

"7his survefilance shall incluca :ne recirculati g ensrcoal f!1:er. "ecor aatar,"

,( .N in the nar 41 control room supp,.y filter train using 4d; N510-1375 as a quice ( d , to veri fy >, 70% efficiency in removing freon test gas.

                               **Except that recirculating charcoal filter samoles shall be removed and analyzed at least once per la months.

LA SALLE - UNIT 1 3/4 7-5 -

PLANT !iYSTEMS SURVEI. LANCE REQUIREMENTS (Continued)

      .                      3. Verifying that on each of the below pressurization mode actuation test signals, the emergency train automatically switches to the pressurization mode of operation and the control room is maintained at a positive pressure of 1/8 inch W.G. relative to the adjacent areas during emergency train operation at a flow rate less than or equal to 4000 cfm:
           ,                       a)     Outside air smoke detection, and b)     Air intake radiation monitors.

4. Verifying that the heaters dissipate 20 2 2.0 Kw when tested in accordance with ANSI N510-1975. This reading shall include the appropriate correction for variations from 480 volts at the bus. l

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP,when they are tested in place in accordance with ANSI H510-1975 while operating the system at a flow rate of 4000 cfm .

                       . 2 10%.

( f. Aftep,each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm 1 10%.

This surveillance shall include the recirculating charcoal filter, " odor eater," in.the normal control room supply filter train using ANSI N510-1975 as a guide to verify > 70% efficiency in removing freon test gas.

(<- - l

                                                                                                        .                          l l

l LA SALLE - UNIT 1 3/4 7-6 Amendment No. 26 l l x __ - _______ _ ___ ._ _ o

T.

                     , Pt. ANT SYST OS 3/4.7.3 REACTOR CORE ISOLATION CCOLING SYSTEM 1

LIMITING CONDITION FOR OPERATION -

3. 7'. 3 The reaci:or core isolation cooling (RCIC) systas shall be CPSA8LE with i an OPDA8LE flow path capable of taking suction free the suppression pool and j transferring the water to the reactor pressure vessel. i APPt.ICABILITY: OPERATIONAL CCNgITIONS 1, 2, and 3 with reactor staam dome pressure greater than 150 psig ACTICN:

a. With a RCIC discharge line "kaeo filled" pressure alars instrumenta-tion channnel inoperacle, perfors Surve111anca Requirement 4.7.3.a.1 at least once per 24 hours. .

b. With the RCIC systas fr.coerable, operation say continue provided the l HPCS systes is OPERA 8LE: restore the RCIC systas f.o CPERA8LE status '

within 14 days or be in at least HOT SHUTC01WN within the next I2 hours . and reduce reactor staan dome pressure to less than or equal to 150 psig within the following 24 hours. - _, SURVEILLANCE REQUIR S ENTS 4.7.3 The RCIC system shall be demonstrated OPERA 8LE:

a. At least once per 31 days by: 1. Varifying by venting

                                            ' piping from the pumo,at    the highvalve discharge. pointto vents    that theisolation the systas       system valve is filled with water,

2. Perfomance of a CHANNEL FUNCTICNAL TEST of the discharge line j " keep fillea" pressure alars instrumentation, and

   ;                                   3. Verifyirig that each valve, manual, power operated or automatic
   ;                                         in the flow path that is not locked, saaled or otherwise secured j                                         in position, is in its correct position.                                                         .__

4. Verifying that the pump flow controller is in the correct - -

j position. .

                                                       ~
   !.                          b.      At least onca per 92 days by verifying that the RCIC pumo develops a flow of greater than or equal to 600 gpa in the test flow path with a i                                   system head corresponding to reactor vessel operating pressure when j                                   steam is being suop11ec o the turcine at 1000 - 20. - 20 psig."
         ^                                                                                                              '

aine provisions or 5pecificationse.0.4 are not applicaole proviced tne

/ surveillance is perfomed within 12 hours after reactor steam pressure is i.( adequate to perform the tests.

l , #5ee Special Test Exception 3.10.7. . I l . LA SALLE - UNIT 1 3/4 7-7

l PLANT SYSTEMS , SURVEILLANCE REOUIREWENTS

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve in the flow path actuates to its correct pos1 tion, but any exclude actual infection of coolant into the reactor vessel.

2. Verifying that the system is capable of providing a flow of I greater than or equal to 600 gpa to the reactor vessel when staan is supplied to the turbine at a pressure of 150 2 15 psig using the test flow path."

3. Performing a CHANNEL CALIBRATION of the discharge line " keep filled" pressure alarm instrumentation and verifying the icw pressure setpoint to be > 62 psig.

d. By demonstrating MCC-121y and the 250-volt battery and charger .

lle OPERA 8LE:

1. At least once per 7 days by verifying that:

a) MCC-121y is energized, and has correct breaker alignment, indicated power availability from the charger and battery, . and voltage on the panel with an overall voltage of greater f than or equal to 250 volts. A b) The electrolyta level of each pilot call is above the plates, c) The pilot call specific gravity, corrected to 77'F, is greater than or equal to 1.200, and d) The overall battery voltage is greater than or equal to 250 volts.

2. At least once per 92 days by verifying that:

a) The voltage of each connected battery is greater than or equal to 250 volts under float charge and has not decreased more than 12 volts'from the value observed during the

original test, b) The specific gravity, corrected to 77'F, of each connected cell is greater than or equal to l'.195 and has not decreased more than 0.05 from the value observed during the previous 1

test, and , i c) The electrolyte level of each connected call is above the - i j plates. . .

                                                                                                                                                                                                                                             ,        l

3. At least once per 18 months by verifying that: l j a) The battary shows no visual indication of physical damage or abnormal deterioratiori, and

,                                                                b)       Battery terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.
                                   "TheprovisionsofSpecification4.0.4arenotapplicablyprovidedthe                                                                                                                        '

A ( surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the tests.

                      .                                                                                                               .                                                                                           I is LA SALLE - UNIT 1                                                    3/4 7-8                                                                Amendment No. 18 l                                                                                                                                                               .
          =. .                    . . - . . . . . . - . .             - . . . - - .                                  .             - - . . ~ . . . .                         ......
   ,---,---,-,-,.n.-,     ,---.-,-,,a               - - , - , ,        --     -     ,,,n..-, - - .
                                                                                                               - - ,   - - - - ~ -             - - - - . - - -      -------,----,.,,---.,-,,.--.n-,--,,,y                    .-,      ,-,nw,   --   n

      ~ . . . . .   . - -

j. .

3/4.8 ELECTRICAL PChER SYSTEWS . , i 3/4.8.1 A.C.SOURfES A.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATI0ft

3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be l OPERA 8LE:

a. Two physically independent circuits between the offsite transmission l networtt and the onsita Class 1E distr.ibution systes, and l

b. Separate and independent diesel generators 0, IA, 2A and 1B with: l

1. For diesel generator 0,1A and 2A: i a) A separate day fuel tank containing a sinimum of l 250 gallons of fuel. ,

b) A separate fuel storage system containing a minimum of l i 31,000 gallons of fuel.

2. For diesel generator 18, a separate fuel storage tank / day tank l containing a sinimum of 29,750 gallons of fuel.

3. A separate fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITICNS 1, 2, and 3. ACTION: j a. With either one offsite circuit or diesel generator 0 or 1A of the above required A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. sources by performing Surveil-lance Requirements 4.8.1.1 la within 1 hour, and 4.8.1.1.2a.4, I for one diesel generator al a time, within eight hours, and at least once per 8 hours thereaftar; restore at least two offsite circuits and diesel generators 0 and 1A to OPERA 8LE status within 72 hours or be in at least HOT SHUTDOWN.within the next 12 hours and in COL 3 SHUTDOWM within the following 24 hours. . . b.- With one offsite circuit and diesel generator 0 or 1A of the above required A.C. electrical power sources inoperable, demonstrate the ._ QPERASILITY of the remainir.g A.C. sources by performing Surveillance Requirements 4.8.1.1.la within 1 hour, and 4.8.1.1.2a.4, for one 1 diesel generator at a time, within six hours, and at least once per 8 hours thereafter; restore at least one of the inoperable A.C. -

sources to CPERA8LE status within 12 hours or be in at least HOT SHUTDOWN within the next 12 hours and in. COLD SHUTDOWN within the following 24 hours. Restore at least two offsite circuits and diesel generators 0 and 1A to OPERA 8LE status within 72 hours from the time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the follcwing 24 hours. ' LA SALLE - UNIT 1 3/4 8-1

Amendment No. 18

                                ,-    -w ,           . - . - ~ - - , - - ,    --.-.,.,.,-r,_ . - , . , , , - - , - - , - - . , , . ,                     -.-,.--.t , , - , .     .,,,_,-,.c,,.,

i i I I ELECTRICAL POWER SYSTEMS , i j LIMITING CONDITION FOR OPERA ION (Continued) , ACTION (Continued)

c. With both of the above required offsite circuits inoperable, demon-strate the OPERA 8ILITY of the remaining A.C. sources by performing

Surve111anca Requirement 4.8.LL2a.4, for one diesel generator at a . time, within eight hours, and at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite circuits to OPERA 8LE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours.

! With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to OPERA 8LE status within 72 hours from time of initial loss or be in at least HOT SHUTD0'dN within the next 12 hours and irr COLD SHUTDOWN within the following 24 hours,

d. With diesel generators 0 and IA of the above reouind A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remain ~-

)                                            ing A.C. sources by performing Surveillance Requirements 4.8.LLla within 1 hour and 4.8.LL2a.4, for one diesel generator at a time, within four hours and at least once per 8 hours thereaftar; restore at least one of the. inoperable diesel generators 0 and 1A to                                                        .

0FERABLE status within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD. SHUTDOWN within the following 24 hours. lf( Restore both diesel generators 0 and 1A to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN = within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. i

e. With diesel ger.erator 18 of the above recuind A.C. electrical power sources inoperacie, demonstrata the OPERASILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.L1.la within 1 hour, and 4.8.L1.2a.4, for one diesel generator at a time,.within six hours, and at least once per 8 hours thereafter; restore the inoperable diesel generator 18 to OPERA 8LE status within 72 hours or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.L: f. With diesel generator 2A of the above nquired A.C. electrical power sources inoperable, demonstrate the OPERASILITY of the remaining -

A.C. sources by performing Surveillance Requirements 4.8.1.Lla and 4.8.LL2.a4, for diesel generator IA, within one hour, and at . least once per 8 hours thereafter; restore the inoperable diesel generator 2A to OPERA 8LE status within 72 hours or declare standby i

gas treatment system subsystes 8, Unit 2 drywell and suppression chamber hydrogen recombiner system, and control room and auxiliary, '

i electric equipment room emergency filtration system train B inoperable and take the ACTION required by Specifications 3.6.5.3, 3.5.6.1., and 3.7.2; continued perfomance of Surveillance Requirements - 4.8.LLla. and 4.8.LL2a.4 for diesel generator 1A is not required . provided the above systems are delcared inoperab}e and the ACTION of their respective specifications is taken. ( l LA SALLE - UNIT 1 3/4 8-2 Amendment No. 18

                                                                                       . . . . .                -- ~ ..- ...            . .             ..

E---- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _

  ..                          ELECTRICAL POWG SYSTD5
         ~

SURVE!LLANCE REQUIREMENTS ! 4.8.1.1.1 Each of the above required indecencent circuits between the offsita l transmission network and the onsite Class 1E distribution systae shall be: l

a. Determined OPERAELE at least once per 7 days by verifying correct breaker alignsents and indicated power availacility, and j Demonstrated CPERA8LE at least once per 18 months during shutdown by ii I tr.

annually transferring unit power supply from the normal circuit to 1 i the alternate circui* i 4.8.1.1.2- Each of the above required diesel generators shall be demonstrated j l OPERABLE: ) i a. In ac:ordance with the frequency specified in Tacle 4. 8.1.1.2-1 on a l STAGGERE3 TEST BASIS by: l L Verifying the fuel level in the day fuel tank. l ) 2. VeHfying the fuel level in the fuel s.orage tank. i 3. Verifying the fusi transfer pumo starts and transfers fual from the storage systas to the day fuel tank.

4. Verifying the diesel starts from ancient condition and ac:aleratas l to 900 rps

SE, -2% in less than or equal to 13 seconds.* The l l generator voltage and frequency sna11 be 4160 2150 volts and i 6a
3.0, -L2 Hz within 13 seconds" after the start signal. l

[ 5. Verifying the diesel generator is synchroni:ed, loaded to greater A than or equal to 2500 kw within 60 seconds,' and operatas with l this load for at least 60 minutes.

6. VeHfying the diesel generator is aligned to provide stanchy power tar the associated emergency busses.

7. Verifying the pressure in all diesel generator air start receivers ta be greater than or equal to 200 psig. -

b. At least onca per 31 days,and after each oceration of the diesel where the period of operation was greater than or equal to 1 nour by checking for and removing accumulated water from the day fuel tanks.

c. At Toast once per 92 days and from new fuel oil pHor to addition to the storage tants by veHfying that a sasole obtained in ac:ordance .

! with ASTM-0270-1975 has a water and sediment content of less than or i equal to 0.05 volume percent and a kinematic viscosity 9 40*C of greater than or equal to 1.9 but less than or equal to 4.1 when tasted in accordance with ASTM-0975-77,. and an isourity level of less than 1 ag. of insolubles per 100 al. when testad in accordance . wita A57M-02274-70. .I. ,

"These oiesel generator starts from ambient conditions shall be parformed at least once per 184 days in these surveillance tasts. All other engine starts for the purpose of this surveillance tasting shall be preceded by an engine .

i prelute period and/or other warmuo procedures reconnended by the manufacturer I so that sechanical stress and wear on the diesel engine is' minimized. , )

                                                                                                                                                                                                             .       j 4

LA SALLE - UNIT 1 3/4 8-3 1 Amendment No.16 i * , 1

( ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. At least once per 18 months

during shutdown by: l

1. Subjecting the diesel to an inspection in accordance with pro.cedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesei generator capability to reject a load of greater than or equal to 1190 kw for diesel generator 0, greater than or equal to 638 kw for diesel generators 1A and 2A, and greater than or equal to 2381 kw for diesel generator 18 while maintaining engine speed less than or equal to 75% of the difference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is less.

3. Verifying the diesel generator capability to reject a load of 2600 kw without tripping. The generator voltage shall not exceed 5000 volts during and following the load rejection.

4. Simulating a loss of offsite power by itself, and: .

a) For Divisions 1 and 2 and for Unit 2 Division 2:

          '                                 1)   Verifying de-energization of the emergency busses and Icad shedding from the emergency busses.

2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently

( connected loads within 13 seconds, energizes the auto-connected loads and operates for greater than or equal to 5 minutes while its generator is so loaded. After j energization, the steady state voltage and frequency 1 of the emergency busses shall be maintained at 4160 2 i

150 volts and 60 1 1.2 Hz during this test. f , b) For Division 3:

1) Verifying de-energization of the emergency bus.

l 2) Verifying the diesel generator starts on the auto-start - signal, energizes the emergency bus with its loads with-in 13 seconds and operates for greater than or equal to . 5 minutes while its generator is so loaded. After energization, the steady state voltage and frequency of the emergency bus shall be maintained at 4160 2 150 volts and 60 2 1.2 Hz during this test. -

5. Verifying that on an ECCS actuation test signal, without loss -

of offsite power, diesel generators 0, IA and 18 start on the auto-start signal and operate on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 + 416 -150 volts and 60 + 3. 0, -1. 2 Hz within 13 seconds i after the auto-start signal; the steady state generator voltage and I frequency shall be maintained within these limits during this test. i i !/ "The specified 18 month interval may be waived for Cycle 1 provided the l( surveillance is performed during Refuel 1. l i LASALLE-UNIT 1 3/4 8-4 Amendment No. 24 l x , - - - _ _

i l i l ( ELECTRICAL POWIR SYSTEMS , i SURVEILLANCE REQUIREMENTS (Continued)

6. Simulating a loss of offsite power in conjunction with an ECCS I actuation test signal, and:

a) For Divisions 1 and 2: '

1) Verifying de-energization of the emergency busses and

load shooding from the emergency busses.

,                                                            2)       Verifying the diesel generator starts'on the auto-start j                                                                     signal, energizes the emergency busses with permanently connected loads within 13 seconds, energizes the l

auto-connected emergene/ loacs througn the load sequencer and operatas for greater than or equal to { . 5 minutes while its generator is loaded with the l ! emergency loads. After energization, the steacy i state voltage and frequency of the emergene/ busses shall be maintained at 4160 2 416 volts and 60 2 1.2 Hz ' during thi.s test. . b) For Division 3:

1) Verifying de-energization of the emergency bus. .: 2) Verifying the diesel generator starts on the auto-start

![ signal, energizes the emergency bus with its loads lA within 13 seconds and operatas for greater than or { equal to 5 minutes while its generator is loaded with i the ese gency loads. After energization, the steady { state voltage and frequency of the emergency bus i shall be maintained at 4160 2 416 volts and 60 2 1.2 Hz during this test. l I 7. Verifying that all diesel generator 0,1A and 18 automatic trips l except the following are automatically bypassed on an ECCS l actuation signal: 1 j a) For Divisions 1 and 2 - engine overspeed, generator differential current, and emergency manual stop. 4

,                                                     b)      For Division 3 - engine overspeed, generator differential .

er overturrent, and emergency manual stop. .. 4

8. Verifying the diesel generator operates for at least 24 hours. l During the first 2 hours of this test, the diesel generator i shall be loaded to greatar than or equal to 2860 h and during the remaining 22 hours of this test, the diesel generatcr shall be loaded to 2600 h. The generator voltage and frequency shall
be 4160 + 420, -150 volts and 60 + 3.0, -1.2 Hz within 13 suconds
after the start signal; the steady state generator voltage and frequency shall be taintained within these limits during this tast. Within 5 minutes after completing this 24 hour test. -

y' . pe ' arm Surveillance Requirement 4.8.1.1.2.d.4.a).2) and b).2).* i.ASALLE-UNIT 1 3/4'8-5 Amendment No.18 1 , ' = = - .. . .. _ . . . . . . _ . . . . . . . . . . i 1

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREwENTS (Continued) Verifying that the auto-connected loads to each diesel generator i 9. do not exceed the 2000 hour rating of 2860 kW. , I

10. Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, i b) Transfer its loads to the offsita power source, and l c) 8e restored to its standby status. , 11. Verifying that with diesel generator 0, IA and IS operating in a l test mode and connected to its bus: a) For Divisions 1 and 2, that a simulated ECCS actuation signal overrides the test mode by returning the diesel

generator to stancby operation.

! b) For Division 3, that a simulated trip of the diesel generator overcurrent relay trips the SAT feed breaker to 3 bus 143 and that the diesel generator continues to supply i I nomal bus loads.

12. Verifying that ther automatic load sequence timer is OPERABLE I~

[ with the interval between each load block within

10% of its A design interval for diesel generators 0 and 1A. ,

13. Verifying that the following diesel generator lockout features l prevent diesel generator operation only when required:

a) Generator underfrequency. b) Low lube oil pressure. c) , High jacket cooli,ng tamperature d) Generator reverse power. a) Generator overcurrent. f) -Generator loss of field. i g) Engine cranking lockout. \ ) i "If Surveillance Aequirement 4.8.1.1.2.d.4a)2) and/or b)2) are not satisfactorily ccepleted, it is not necessary to repeat the preceding 24 hour tast. Instead, the diesel generator say be operated at 2600 kW for 1 hour or ,until cperating . l tamperature has stabilized. , L .

LA SALLE - UNIT 1 3/4 8-6 Amendment No. 18 4 I .. - . . . . . .

 - . . . - . _ . . _ . _ . _ _ . _ . _ .                _ _ _ . , _ - _ _ . _ - _ _ _ _ .              . .,_       ,_     . _ , . _ _ _ _ _ _ _ . . _ _ _                  _ _ . .       _....______.m_..._.         _._ ..

r

     .               ELECTRICAL POWER SYSTEMS r
        - . - -      SURVEILLANCE REQUIREMENTS (Continued)

e. At least once per 10 years or after any modifications wh h' could affect diesel generator interdependence by starting diesel gener-ators 0, IA and IB simultaneously, during shutdown, and verifying ,

that'all three diesel generators accelerate to 900 rpm + 5. -2% in less than or equal to 13 seconds.

f. At least once per 10 years by: ,

g

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and

2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND, of the ASME Code in accordance with ASME Code Section 11, Article IWD-5000.

4. 8.1.1. 3 Recorts - All diesel generator failures, valid or non-valid, shall be reported to tne Commission pursuant to Specification 6.6.C within 30 days. l Reports of diesel generator failures shall include the information recommended

in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is greater than or equal to 7, the report shall be supplemented to C incluce the additional informatiori recommer.ded in Regulatory Position c.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

TABLE 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE

                                          .                   i Number of Failures in Test Frecuency Last 100 Valid Tests *
                                  <1                                    At least once per 31 days 2

At least once per 14 days _ 3 At least once per 7 days -

                                  ,4
                                   )                                    At least once per 3 days
                             " Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the last 100 tests are determined on a per nuclear unit basis. With the               '

exception of the semi-annual fast start, no starting time re-f Quirements are required to meet the valid test requirements of l ( Regulatory Guide 1.108. 1

    -                                                                                                                    l 3/4 8-7                   Amendment No. 23               1
                  . LA SALLE - UNIT 1

 ,                                                                                                                 1 i

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: LASALLE 1 REACTOR TYPE: BWR-GE5 DATE ADMINISTERED: 86/06/03 EXAMINER: SLY, G. APPLICANT: //#54# Mev -4/O '

INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

                                                                                                               ~
                                                         % OF CATEGORY % OF                     APPLICANT'S CATEGORY VALUE          TOTAL              SCORE        VALUE                   CATEGORY
                       -?.                                      1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00            0 .00                                   2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00
                       $<0YQ M                    '                   3. INSTRUMENTS AND CONTROLS 25.00                                                    4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL 96 S CONTR0t pp                                                                                                        -
      'nn nn           100.00                                   TOTALS                                           .

FINAL GRADE % All work done on this examination is my own. I have neither given nor received aid. , APPLICANT'S SIGNATURE 0

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

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

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

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

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one sTde of the paper, and write "Last Page" on the last answer sheet.

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

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION -

AND DO NOT LEAVE ANY ANSWER BLANK.

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

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

 %)                                                                                                                              P s

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top. (2) Exam aids - figures, tables, etc. J (3) Answer pages including figures which are a part of the answer.

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

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

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

) i l l v

~.

1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2

 ,               THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW N       QUESTION 1.01              (2.00)

List four (4) reactor conditions or characteristics which influence the point of criticality and the rate at which it is approached. (2.0) QUESTION 1.02 (3.00) Concerning Thermal limits:

a. WHAT are the three (3) peaking factors which make up the Total Peaking Factor? (1.0)

b. WHAT safety condition does limiting MFLPD to less than one (1) prevent? (1.0)

c. WHAT is the definition of MAPRAT7 (1.0) .

QUESTION 1.03 (2.00) During your Shift, an SRV inadvertantly opens from 100% power and 1000 psia. Use a Hollier Diagram or the Steam Tables to answer the following (ASSUME a saturated system and instantaneous heat transfer):

a. WHAT is the SRV tailpipe te'mperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization? (0.5)

b. If the Suppression Pool Pressure were to increase, would the Tailpipe Temperature INCREASE, DECREASE, or REMAIN THE SAME7 (0.5) _

c. If the reactor is depressurized when the SRV is opened, -

will the Tailpipe Temperature INITIALLY INCREASE, DECREASE, or REMAIN THE SAME7 (0.5)

d. At WHAT Reactor Pressure will the Tailpipe Temperature be at its MAXIMUM value (during the depressurization)? (0.5)

(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)

1- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (1.00) During the last refueling outage a center bundle was inadvertently positioned in a peripheral bundle location. The reactor is then brought to 100% power. FILL-in-the-blanks with (LARGER, SMALLER, or THE SAME) to indicate the correct response for the thermal hydraulic conditions in the misplaced bundle.

a. The flow rate in the misplaced bundle will be than the adjacent peripheral bundles. (0.5)

b. The power level in the misplaced bundle will be than if the bundle had been properly placed in its central location. (0.5)

QUESTION 1.05 (. _2.25) c., Af '# U o 9,..s , A fu. pin, over a period of time, has a uniform coating of corrosion produc s about 0.001 inches thick on its surfacc. A::uming that poner generation within the fuel pin remains constant during the time of the buildup, would you expect the following temperatures to (INCREASE, DECREASE, or REMAIN THE SAME) during the buildup? EXPLAIN each answer. f a. Fuel temperature. (0.75) N b. Cladding temperature. , (0.75)

c. Coolant temperature surrounding the lower portion of the fuel pin (prior to the onset of boiling). (0.75)

QUESTION 1.06 (2.25) N Following a normal reduction in power from 90% to 70% with

recirculation flow, HOW will each of the following change (INCREASE, DECREASE, or REMAIN THE SAME) AND WHY? a m a. The pressure difference between the reactor and the turbine steam chest. (0.75)

b. enna nute dep=::icr et. ihm!t of the wuanxi. (0.75); c. Feedwater temperature. (0.75)
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****): 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW s QUESTION 1.07 (3.00) For the situations listed below STATE HOW each of the reactivity coefficients (void, moderator, doppler) would respond. (i.e. more negative, less negative, no change). M37 a. EOL versus BOL (1.5) ( b. 50% power versus 100% power operation. (1.5) QUESTION 1.08 (2.50) STATE whether the following statements concerning fission poisons are RU' or FALSE. I 6 o~ a. The largest production contr)hution s of xenon is tha ~ raoica&c d:: y of I:dMe3 (0.5) r b. A 25% power reduction from 100% power would have a larger

            !           xenon peak than a 25% power reduction from 50% power.         (0.5)
            \
              \ c.

A rapid reactor sh'utdown from 100% power would have the same resultant xenon peak as a reactor scram from 100% power. (0.5) g d. Since the production and removal of samarium is a direct function of thermal flux, a reactor that has operated at ONLY 50% capacity will have the same equilibrium samarium concentration as one that has operated at 100% capacity. (0.5)

e. Upon restarting the reactor following a 6-month outage, the samarium concentration will decrease to its 100% full power concentration. (0.5) _

1 (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

b PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO1 PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.09' (3.00) STATE WHY the following situations would or would not change differential control rod worth. (Also indicate whether rod worth would increase, decrease, or not change.) h a. A rod is withdrawn from notch 06 to notch 10. (0.75)

                      . Pulling the first rnet h Sd Creup 7 b;te:d of the 7:t i viMn Rod Group 5.

_ _ . . (0.75)

                 $c. Localized voiding of a region not previously voided.             (0.75)
            &/   n_ d . A change of the size of the control rod reducing the surface area while maintaining the same boron volume.            (0.75) g"           ESTION 1.10            (2.00)

'th The reactor is critical at 10E+6 cps. A stabl o 60 seconds

   /        is achieved. If rods are inserted co              s y until the period g

A drops to infinity and then WILL the reactor nsertion is immediately stopped, CAL, SUPERCRITICAL, or SUBCRITICAL) in the time foll e rod stoppage? EXPLAIN. (2.0)

                                  ,                   i QUESTION 1.11              (2.00)

Five (5) minutes following a reactor scram from 100% power, reactor power is 15 on IRM Range 4 and decreasing. WHAT is the minimum IRM

    %q Range that you could go to two (2) minutes later without violating any operational limits? SHOW calculation and EXPLAIN any assumptions made.                                                                           (2.0) l l

l (*****ENDOFCATEGORY01*****)

2.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 s QUESTION 2.01 (2.00) WHAT are four (4) functions of the SBLC " Pipe within a Pipe?" (2.0) QUESTION 2.02 (1.50) 0 EXPLAIN HOW the pressure equalizing valves in the CRD Hydraulic System accomplish their purpose. (1.5) I QUESTION 2.03 (1.50) . WHAT two (2) conditions must be met to bypass the high level scram 4 U. on the scram discharge volume (SDV) following a reactor scram? (1.5) QUESTION 2.04 (2.00)

        ,4       LIST the four (4) signals which will cause an automatic v        recirculation pump to downshift from fast speed. (Actions which cause a unique signal condition are considered to be one signal.)    (2.0)
         ,     QUESTION 2.05          (1.50)

HOW is the 1A Diesel Generator kept warm in standby condition? (1.5)

QUESTION 2.06 (2.00) (b ANSWER the following questions concerning the vacuum relief lines between the drywell and the suppression chamber.

a. WHICH direction does the flow go? (0.5) 1

b. WHAT safety feature do they provide for or prevent? (0.75)

c. WHAT would be the consequence if a vacuum breaker line had failed in the open position? (0.75)

(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)

s

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 s

QUESTION 2.07 (1.50) i DESCRIBE HOW combustible gases are removed from the primary containment in the event that both hydrogen recombiners are inoperative following a LOCA event? (1.5) QUESTION 2.08 (1-00)

       , (i If the maximum amount of boron concentration necessary to shutdown the reactor from a full power condition is 660 ppm, WHY is the total by ' design concentration 1075 ppm?                                             (1.0)

QUESTION 2.09 (1.00) L'V Of the four (4) systems and indications listed below indicate WHICH ~ signal does NOT directly come from the main steam line flow restrictors? (1.0)

a. Feedwater Level Control System for steam / feed flow mismatch comparison.

I b. Process computer for steam flow, heat balance input.

,                    c. Rod Sequence Control System for steam flow, power level l                          indications.                '

d. Primary Containment Isolation System for steam flow, MSIV isolation.

(2.50) QUESTION 2.10

You are in the process of a turbine heatup and roll and have been requested to monitor turbine exhaust hood temperatures,

a. WHERE are the exhaust hood temperatures sensed? (0.5) bi b. For an increasing temperature in the hood, LIST any automatic action that occur and the temperatures they occur at (do NOT include alarms). (1.0) 1 0g c. WHAT protective function does the exhaust spray system provide? . (1.0) i

(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.11 (2.50)

When the HPCS ' system logic is initiated, LIST five (5) components (\ that receive direct signals and the type of signals they are (i.e., open, close, start, stop). gf, , ,,, , 4(pw ;,,7 (2.5) 1 QUESTION 2.12 (1.00)

      \        Concerning the relief valve low-low set (LLS) function:

a. .WHAT is the purpose? (0.5)

b. DOES the LLS actuate on manual operation of the relief valves, automatic operation of the relief valves, or on either one? (0.5)

()S QUESTION 2.13 (1.50) Concerning the RCIC system:

a. STATE the operating equipment that will trip when RCIC injects into the vessel. (1.0)

b. WHAT signals are used to in,itiate these trips? (0.5) l i

(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.14 (1.50) 0*

The Inboard Shutdown Cooling Containment Isolation Valve (E12-F009) , is normally powered from a Division II switchgear. An alternate ' power supply is provided from Division I. Concerning using the alternate power supply, INDICATE whether the following precautions . are TRUE or FALSE. (1.5) , The Outboard Shutdown Cooling Isolation Valve (E12-F008) a.- must be closed when using the Inboard Valves alternate , power supply. .

b. There is no protective circuitry to prevent feeding the valve motor operator from both the normal and alternate i power supplies,

c. The F009 valve will NOT automatically isolate when the valve is powered from it's alternate power source.

QUESTION 2.15 (2.00) , t The HPCS system can take a suction from either the Condensate Storage Tank or the Suppress. ion Pool. bl a. WHICH is the preferred suction source. (0.5)

b. WHAT condition (s) will automatically transfer the preferred suction source to the alternate suction source? (1.0)
c. If the automatic transfer condition, in Part b., were to i

clear the suction source WOULD automatically transfer back ' to the normally perferred suction source. TRUE or FALSE. (0.5) d I i l (*****ENDOFCATEGORY02*****) l \ l l 1

l

3. INSTRUMENTS AND CONTROLS PAGE 10 QUESTION 3.01 (3.00)

For the following situations STATE the actions of the Diesel Generator "0" output breaker if: 64 a. there is an ECCS condition on Unit 1, and a simultaneous ECCS and undervoltage condition occurs on Unit 2. (1.0)

        @' b . it is closed onto Unit 2 (241Y), as a result of under-voltage on 241Y, and an ECCS initiation occurs on Unit 1.     (1.0)
         . c. It is manually closed on Unit 1 (141Y), an ECCS initiation p%        occurs on Unit 1, and an ECCS and undervoltage condition is sensed on Unit 2.                                             (1.0)

QUESTION 3.02 (2.50) Concerning the Reactor Water Cleanup (RWCU) system: e a. LIST two (2) conditions that will cause the blowdown flow control valve (F003) to auto close. (1.0)

b. Other than manual isolation, WHAT are four (4) conditions a) that will close the inboard isolation valve (F001)? ,

(Any condition which causes the same isolation is considered to be one condition). [o /w t>. k & lu.%5 (1.0) ri c. HOW would a loss of service' air effect the RWCU system? (0.5)

                                        '     ~
                       $ v > .. .. . , o ."
                                            ~

I [,Ilti /b.a.> s t-) Ok n'i 0 l'* ~" J

                                             %D                                               _

9 (***** CATEGORY 03CONTINUEDONNEXTPAGE*****)

       ?. INSTRUMENTS AND CONTROLS                                                        PAGE 11 QUESTION 3.03            (3.00)

ASSUME the RCIC system receives an initiation signal, all system components function properly, except the items listed below. Each failure is present prior to the initiation signal being received. DESCRIBE the RCIC system response for each of the following and JUSTIFY your answer. Consider each item separately. 8 a. The turbine exhaust valve (F058) is stuck shut. (1.0) 0S

b. The Ramp Generator portion of the RGSC (Ramp Generator Signal Converter) has failed and is producing t. large signal. (1.0)

Q c. The D/P cell, for the RCIC flow control element, has a - perforated diaphragm. ( ,, , . pt ' (1.0) QUESTION 3.04 (2.00) When paralleling a diesel generator with the 4KV bus: p a. WHAT do you use th'e governor control for (1.0)

1. before the breaker is closed? b'.;v^d e

2. after the breaker is closed?

I pt b. WHAT do you use the voltage regulator adjust for (1.0)

1. before the breaker is closed?

2. after the breaker is closed?

QUESTION 3.05 (2.00) 1 Concerning the FWLCS Setpoint Setdown function: l l

a. EXPLAIN the function of Setpoint Setdown and HOW it is !

accomplished? (1.0) I b. WHAT would be the result on water level (INCREASE, if, the Setpoint Setdown logic DECREASE, or NO CHANGE)(K11 contact energized) at a full initiated sporadically power condition. WHY7 (Assume 3-elementcontrol.) (1.0) ' (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) j

3.. INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.06 (2.50) ANSWER the following questio'~ ns concerning ADS initiating logic:

a. With the plant at 100% power, the channel Al and channel A2 manual initiation push buttoms are rotated and depressed. WILL the ADS function occur? WHY? (1.0)

b. If ADS manually initiated (and all initiation conditions are met), WILL the ADS SRV opening be delayed by the 105 second timer? (Yes/No) (0.5) 2 c. If an ADS blowdown is in progress, with all initiation 8b*. signals still present, WILL depressing the ADS logic reset pushbutton switches reinitialize the 105 sec. timer and close all ADS valves. (Yes/No) (0.5)

d. EXPLAIN HOW a loss of Drywell Pneumatic Air ((160 psig) and the nitrogen bottles will not hamper the operation of the ADS. (0.5) .

QUESTION 3.07 (2.00) Given the following data for APRM Channel C: LPRM Level: A B C D Number of LPRMs assigned: 6, 5 5 5 Number of LPRMs bypassed: 3 4 0 0 et a. If APRM Channel C selector switch on the local (back) panel was placed to the COUNT position, WHAT would be the expected meter reading? (SHOWcalculations.) (1.0) ou b. Based on the above data, is APRM Channel C operable: ANSWER YES or NO and EXPLAIN WHY. (1.0) - ? , (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

l

3. INSTRUMENTS AND CONTROLS PAGE 13 l

QUESTION 3.08 (2.00) The reactor is at 100% power with the generator synced to the grid. Electrohydraulic Control (EHC) load set is 105%. By using the attached EHC diagram, EXPLAIN WHAT would happen (control valve, bypass valve) in the following circumstances: OL a. load limit potentiometer reduced to 95%. (0.5) O b. maximum combined flow limit potentiometer reduced to 95%. (0.5) q c. "A" pressure regulatory (transmitter) fails low. (0.5) et d. failure of two (2) bypass valves to full open. (0.5) QUESTION 3.09 (2.50) The Instrument and Service Air systems receive air from a common set of three (3) air compressors. *

    '   a. The Unit I station air compressor is lined up in the '0N',
             ' Modulate' mode of operation. If system demand is less than 60% capacity,. WHAT action, in reference to mode of operation, must be taken and WHY?                            (1.5)
    )u b. If Instrument Air were completely lost, in WHAT position would each of the following valves fail?                     (1.0)

1. scram inlet valve '

2. reactor water cleanup filter /demin, inlet and outlet valves

3. outboard MSIV valves

4. Turbine Building Closed Cooling Water Temperature -

Control Valve .. (***** CATEGORY 03CONTINUEDONNEXTPAGE*****)

3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.10 (2.50)

WHICH of the following Reactor Protection System (RPS) scrams can be bypassed? DESCRIBE HOW each is bypassed. (2.5) (,k a. APRM high flux or power

             & b. MSIV closure q c. Manualh.irah1-
                                          '\ '   "~

g d. Turbine control valve fast closure ti e. Main steam line high rad M f. Unit 2 low CRD header pressure QUESTION 3.11 (1.00) , f According to the Rod Sequence Control System (RSCS):

a. WHAT would be.the next withdrawal limit for a rod positioned at limit N2? (0.5)

1. 8

2. 12

3. 24

4. 48

                               .                  i
                               ~

b. Rod Group 9 can be selected for movement prior to Rod Groups 5, 6, 7, and 8. (TRUEorFALSE) (0.5) e<%

(*****ENDOFCATEGORY03*****)

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADICLOGICAL CONTROL QUESTION 4.01 (2.50) A total loss of AC power has occurred. ANSWER the following questions concerning LOA-AP-08, Total Loss of AC Power. QY a. WHAT SRM and reactor water level indication (s) are available in the control room following this event? (1.0) Q b. WHAT three (3) operator actions are required to verify the scram that should have occurred? (1.5) QUESTION 4.02 (3.00) . According to LOA-RR-05, Reactor Recirculation Flow Control Valve Lockout and Runback: Q a. WHAT three (3) recirculation flow control valve lockout conditions will cause an automatic transfer to Manual on the M/A station? (2.25) f b. WHY are you required to balance flow in the non-locked P( loop, following a recirculation flow control valve ,

       ,v ' "           lockout?                                                    (0.75)

QUESTION 4.03 (2.50) WHAT are the five (5) methods described in LOA-NB-09, Alternate Rod O Insertion, to insert control rods still out following an automatic scram? , (2.5) gAA) p QUESTION 4.04 (2.50) According to Procedure LOA-GP, General Precautions: og a. WHAT precautions must be taken PRIOR T0 placing an ECCS system in manual? (1.5) . p,b. WHAT precautions must be taken WHILE an ECCS system is in manual? 1 (1.0) l (***** CATEGORY 04CONTINUEDONNEXTPAGE*****)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.05 (2.50)

You have been operating at 60% power when one (1) recirculation loop trips. You have been requested to restart the idle loop. Qa. According to LOA-RR-03, Recirculation Procedure, WHAT are the Technical Specifications temperature limits that apply to the restart of an idle loop? (1.5) If the idle loop cannot be restarted, COULD you return to D( b. 60% power operation with only one (1) recirculation loop for an extended period of time, (i.e., greater than 8 hours)? EXPLAIN. (1.0) QUESTION 4.06 (2.00)

a. HOW long can an operator stay in a 25 mrem /hr radiation field

without exceeding: L"p-I ,I gA j . t v;d ,Lq1. a LaSalle administrative exposure limit? (0.5) pl' ' Q 2. a NRC quarterly exposure limit if you had already received one (1) rem this quarter? g ,,..v; ,r. f c. . ,, y...

                                                                                    . c p er s) (0.5)

b. You are working in a group of four (4) people; recharging CRD 04'\ accumulators when you drop your pocket dosimeter. The dosimeter is now reading full scale while the other three (3) members of your team are~ reading 5 mrem. According to LRP-1000-1, Radiation Protection Standards, WHAT is the most correct action (s) to this situation? (1.0)

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) i I

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL QUESTION 4.07 (2.50)

In reference to LGP-2-2, Unit Shutdown From Power Operation To Hot Standby:

a. HOW do you assure that an overlap of at least 1/2 decade exists between the IRMs and APRM7 (0.75)

b. Following the tripping of the main turbine you are cautioned to "not open the vacuum breakers to the condenser...,"

WHAT is the reason for this caution? (0.75) (j c. WHAT conditions determine when the plant is operating in HOT standby? (1.0) QUESTION 4.08 (2.50) The plant is operating at power when an SRV inadvertently opens. As per LOA-NB-02, The Stuck Open Safety Relief Valve, the operator cycles the SRV control switch from AUTO to OPEN and back to AUTO. d a. If this action doe.s NOT close the SRV, WHAT other method can be performed in an attempt to close the stuck open valve? (0.5) O '

b. VHAT control room indications would the operator have if the valve closed? (0.5)

Og c. WHAT three (3) conditions would require the operator to manually SCRAM the plant if the SRV remained open? (1.5) QUESTION 4.09 (2.00) - j

pL a. Other than "As directed," WHAT are the entry conditions to LGA-01, level / Pressure Ccatrol? (1.0)

b. If you are executing step C.8.a "Depressurize the RPV" in

g. procedure LGA-01 and another (not the initial one) entry condition to LGA-01 is met, ARE you required to restart the Q procedure at the initial step? (Yes/No) PROVIDE a reason for '

your response. (1.0) - (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 RADIOLOGICAL CONIROL QUESTION 4.10 (2.00)

"K

a. Procedure LRP-1000-1, Radiation Protection Standard, uses the words " Caution" and " Danger" to designate the major radiological conditions. DISTINGUISH when the word " Danger" and the word

            " Caution" would be used on a radiological sign?                   (1.0) g    b. WHEN would a Radiation Work Permit Type 1 be required?             (1.0)

QUESTI0!' 4.11 (1.00) p According to Technical Specifications WHAT must the operator - y immediately do when more than one withdrawn control rod has an inoperable accumulator? (1,0)

                              **** END OF CATEGORY 04 *****)

(********(*****ENDOFEXAMINATION***************)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 1.10 (2.00) Supercritical . [+0.5] When the period reaches infinity, the reactor is exactly critical e prc:pt ==utennr. [+0. 5] After the rod insertion stops the delayed neutron precursors which were formed in previous generations and at a higher power 1evel tend to pull power back up. [+0.5] Therefore, the reactor is still supercritical due to the latent effect of delayed neutrons. [+0.5] REFERENCE (3d, bd eM f.oSM N '

1. LaSalle: Reactor Physics Review, pp. 102-112, TF0:7a, 12b.

2. NMP-2 Operations Technology, Module 1, Part 11, pp. I-11-6. -

ANSWER 1.11 (2.00) , Using P=Poe**(-t/T) [+0.25]

               = 15 e**(-120/80)
               = 3.35 on Range 4 [+0.25]

Therefore Rang i t owest Range [+0.5] kott' hAS Assumptions: On a down power transient, with large negative reactivity insertions, the stable decay period is determined by the longest lived half-life. [+0.5] For this example, it is assumed to be -80 seconds. [+0.5] , REFERENCE

1. LaSalle: Reactor Physics Review, pp. 80-112, TP0:12b, 13d.

6 9

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 1.01 (2.00)

1. Xenon concentration

2. Moderator Temperature

3. Control rod position

4. Order of withdrawal

5. Core Exposure (4 of 5 required for full credit) )

REFERENCE

1. LaSalle: LGP 1-1, p. 6.

ANSWER 1.02 (3.00) 1

a. 1. Axial '+10.331 *

2. Local l+ 0. 33.'

3. Radial peaking factors [+0.34]

b. Exceeding 1% plastic strain [+0.75] between the cladding and pellet [+0.25]

c. APLHGR/MAPLHGR limit [+1.0]

REFERENCE

1. LaSalle: Core Thermal Hydraulics, pp. 38, 40, 42, 50, 52, TP0:3b, 6b, 7c. - '

ANSWER 1.03 (2.00)

a. [+0.5]

b. 295 degrees, Increase ,

                        +        0.5,F (+- 5 degrees    F)                            --

c. Increase + 0.5

d. 450 psia (+- 25 psia) [+0.5]

REFERENCE

1. Steam lables/Mollier Diagram., TP0:15b.

l l

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW )

ANSWERS -- LASALLE 1 -86/06/03-SLY, G. ANSWER 1.04 (1.00)

a. Smaller [+0.5] (2 phase flow larger due to higher power, orifice fixed)

b. Smaller [+0.5] (2 phase flow larger due to smaller ortftce)

REFERENCE

1. LaSalle: Core Thermal Hydraulics, pp. 20 and 21.

ANSWER 1.05 (2.25) .

a. Fuel ten.perature would INCREASE [+0.25] to get the needed delta T to transfer the heat to the coolant. The corrosion layer will require some delta T across it to transfer heat [+0.5].

b. Cladding temperature would also INCREASE [+0.25] because the pin

temperature increased and the cladding is now transferring heat to the corrosion film instead of the coolant. [+0.5]

c. Coolant temperature REMAINS THE SAME [+0.25] since it is a function of pressure, which is maintained constant by the EHC system. [+0.5]

REFERENCE

1. LaSalle: Fluid Flow and Heat Transfer, pp. 76 and 78, TP0:II.B.5.

5.06 ANSWER 1.06 (C.25)

a. Decreases. [+0.25] There is less steam flow, therefore, less ..

pressure drop through the main steam lines. [+0.5]

b. Increases. [+0.25] With the same amount of cooling water through the condenser and less of a heat load, condensate depression will increase. [+0.5]

c. Decreases. [+0.25] Less extraction steam from the turbine to heat the feedwater. [+0.5]

REFERENCE -

1. LaSalle: Fluid Flow and Heat Transfer, pp. 34-40, TP0:C.1.

2. LaSalle: Fluid Flow and Heat Transfer, pp. 78-88, TP0:B.3, B.4.

3. LaSalle: Thermodynamics and Steam Cycles, pp. 78-86, TP0:24,

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

25. ANSWER 1.07 (3.00)

a. 1. alpha m becomes less negative l+0.5?

2. alpha v becomes less negative ,+0.5;

3. alpha d becomes more negative ,+0.5;

b. 1. alpha m becomes ! r r wa't W'+0.51

2. alpha v becomes i = d bt!"es*I+0.5' no cb636 (71 N

3. alpha d becomes more negative +0.5 REFERENCE

1. LaSalle: Reactor Physics Review, pp. 124, 138, and 164, TP0:16e, 17d, 18b.

2. LaSalle: Reactor Physics Review, pp. 122, 132, and 162, TP0:16e, 17d, 18b. ,

ANSWER 1.08 (2.50) a. b. bTrue +0 5'

                      '+0. 5'

c. False '+0. 5'

d. True '+0. 5'

e. True '+0.5 , i REFERENCE

1. LaSalle: Reactor Physics Review, pp. 208, 220, and 224, TP0:21, 22.

ANSWER 1.09 (3.00)

a. Rod worth increases, '+0.251 due to higher flux. '+0. 5'

b. Rod worth increases, '+0.25 due to higher flux. l+0.5l

c. Rod worth decreases, l+ 0. 25' due to decrease in thermal neutrons.

[+0.5]

d. Rod worth decreases, [+0.25] due to reduced surface area seen by flux. [+0.5]

REFERENCE

1. LaSalle: Reactor Physics, pp. 184, 188,.190, and 198, TP0:19.C. ;

2. General Electric Reactor Theory, p. 5-13a.

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

1 l ANSWER 2.01 (2.00)

1. Sodium pentaborate injection

2. HPCS line break detection

3. Core plate delta P measurement

4. Jet pump delta P measurement

5. RWCU bottom head drain flow tap

6. CRD drive water and cooling water delta P tap (Four (4) of six (6) required for full credit at [+0.5] each)

REFERENCE

1. LaSalle: System Description, Chapter 2, p. 33, TP0:3h.

ANSWER 2.02 (1.50) , (Following a scram, the exhaust header will equalize to the scram discharge volume pressure across directional control valve 121.) Once the outlet valve is shut (when scram reset) the exhaust water header would remain at some low pressure causing a high differential pressure across the control rod mechanism if rod motion were attempted. [+0.5] The equalizing valves will open at approximately 70 psid between exhaust [+0.25] and cooling water header [+0.25] to facilitate a rapid re this problem. [+0.5] pressurization of the exhaust header to eliminate

                      ,                 i REFERENCE

1. LaSalle: System Description, Chapter 8, p. 30, TPO:2n.

~~

ANSWER 2.03 (1.50)

1. The reactor mode selection switch must be in shutdowr or refuel positions [+0.75]

2. TheSDVbypassswitch5Imustbeplacedinthebypassposition

[+0.75] REFERENCE

1. LaSalle: System Description, Chapter 8, p. 32, TP0:Sa. -

0

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 2.04 (2.00)

1. Steam line (or dome) to pump suction temperature difference is (10.1 degrees delta T.

2. Total feed flow (30%.

3. TSV or TCV closure with power >30% of rated (E0C-RPT).

4. Reactor water level (12.5". (Also accept Rx low level.)

[+0.5] each REFERENCE

1. LaSalle: System Description, Chapter 5, pp. 70, 72 and 80, TP0:12c.

ANSWER 2.05 (1.50)

1. Heater in jacket water cooling system cycles to maintain water

temperature (causing water to circulate through the engine).

[+0.75]

2. Water also circulates through the lube oil coolers which heats lube oil that is being , circulated by a pump. [+0.75]

REFERENCE i

1. LaSalle: System Description, Chapter 47, p.19, TP0:2b,c; 3.

ANSWER 2.06 (2.00)

a. from suppression chamber to drywell [40.5]

b. negative pressure in drywell to limit upward loading of the drywell floor [+0.75]

c. the flow would bypass the suppression chamber suppression ---

feature following a LOCA. [+0.75] , REFERENCE ,

1. LaSalle: System Description, Chapter 48, p. 10, TP0:4a.

1 0 9

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 2.07 (1.50)

1. -ftkr. ogen-44f&aumilied tetha coatikment (ttr t!11"+a +M H2 rd

        -02::::ntr:tica)viathecontainmentventandpurgesystem.           M UfA .l.Mk

[+0.75] -od.-

2. The Standby Gas Treatment System is used to remove the gaseous mixture. [+0.75]

REFERENCE

1. LaSalle: System Description, Chapter 50, pp. 38-44, TP0:1f,h; 3a.

2. LaSalle: System Description, Chapter 51, p. 22, TP0:6b.

ANSWER 2.08 (1.00)

(Additional 25% poison (165 ppm)) to allow for imperfect mixin and (250 ppm addition) to accommodate for RHR dilution. [+0.5]g [+0.5] REFERENCE

1. LaSalle: System Description, Chapter 10, p. 30, TP0:2b, 5b.

ANSWER 2.09 (1.00)

c. [+1.0]

REFERENCE

1. LaSalle: System Description, Chapter 21, pp. 19-22, TPO:1, 4.

M 1 I 9

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 2.10 (2.50)

a. Thermostats in the exhaust hood of the 'A', 'B', 'C' low pressure turbine. [+0.5]

b. 1. 120 degrees F - exhaust hood spray initiates [+0.5]

2. 225 degrees F - turbine trip on high exhaust hood temperature [+0.5]

c. Minimize distortion of the low-pressure hood and shell structure which could result in rubbing and improper bearing loading. [+1.0]

REFERENCE

1. LaSalle: System Description, Chapter 23, pp. 23-25, TP0:li, 3c, 5.

ANSWER 2.11 (2.50)

1. HPCS pump - start signal

2. CST suction valve (F001.) - open signal, if suppression pool suction valve (F015) is not full open

3. The injection valve (F004) - open signal

4. The CST flow test valve (F010, F011) - close signals

5. The suppression pool test flow valve (F023) - close signal

6. Division 3 (IB) diesel generator - start signal I

(Five (5) of the six (6) are required for full credit at [+0.5] each) REFERENCE

1. LaSalle: System Description, Chapter 36, pp. 10-13, TP0:4a, 5.

ANSWER 2.12 (1.00)

a. To minimize containment fatigue from duty cycles. (Also accept:

reduces relief valve cycling.) [+0.5]

b. This will function if valves are opened manually OR automatically. [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 21, pp. 37-38, TPO:2.

l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 2.13 (1.50)

a. Trips Main Turbine and both Feedwater Turbines. [+1.0]

b. Opening of Injection Valve (E51-F013) and/or Testable Check Valve (E51-F066) as indicated by valve limit switches. [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 41, pp. 28 and 29, TPO:3a.

2. 1E-1-4226BF, RCIC Electric Schematic.

ANSWER 2.14 (1.50)

a. False +0.5]

b. True '+10.51

c. True l+1.5j 0 ,

REFERENCE

1. LaSalle: Exam Bank, Question 2-LS-1.

ANSWER 2.15 (2.00)

a. Condensate Storage Tank [+0.5]

b. '

Low level in the CST [+0.5] High Suppress' ion Pool level [+0.5]

c. FALSE [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 36, p.15, TP0:4d e

9 0

3. INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 3.01 (3.00)

a. closes on Unit 2 [+1.0]

b. remains closed on Unit 2 [+1.0'

c. . smo i si. ch::d cr. Urdtd [+1.0' '

N 5os f dati i e.los u o n O J t a REFERENCE

1. LaSalle: System Description, Chapter 47, pp. 37, 42, 43, 54, and 55, TPO:6c.

ANSWER 3.02 (2.50) .

a. 1. Low pressure in piping upstream of the FCV. [+0.5] ;

2. High pressure in piping downstream of the FCV. [+0.5]

b. 1. Low Reactor Water-Level 2 ~

2. RWCUS Inlet and Outlet High Flow Differential

3. High Ambient Temperature in RWCU Hx Room

4. High dT across Hx Room ventilation ducts

[*, [ ss o[ ,RPg basu, j, gg,g j 3, ej,,fyj,,; ,y,4x Any four (4) [+0.25] each

c. Service air is used in the filter-demineralizer backwash operation of the precoat evolution. (Could not backwash.)

     -c)& -VC.5] Zaa h w ed A,)t. A u uti fo r' dok /.i = of /.74- de-t-en/.yk o-ci 41cw p:*e v. Sys km powW by a -d d.?* /<. ( **b aw 4' A.J/ WE Z%] cLL REFERENCE Jo .guvie die nof nahblu M Zuku-,s.} 4 x).

1. LaSalle: System Description, Chapter 9, pp. 18, 21, 22 and 34, TP0:2j, 4a, 6a.

W I . l t

                                               - - -           .     . ~    .- -           ----

9

3. INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 3.03 (3.00)

a. RCIC will not initiate [+0.25]; the RCIC steam stop valve (F045) will not open if the exhaust valve (F068) is not full open.

[+0.75]

b. The turbine will trip on overspeed [+0.25]; the ramp generator is usually the low signal which controls the turbine on quick starts; with this signal high the turbine is up to speed before sufficient oil pressure is available to the governor, valve to ,,

close it. [+0.75 P 4

                                                          ~~.'"'>^-  - ^ ='/ ^ ^ f N NM NM.

c. "TC i1T w::: :: maxt:5.DM"- w " '

                                                                 " ^ "" Aq " ,,c,uee
           $ F#C" ^^'^ ^ ^ ((+0.25]; the flow signal is at 4jy9,,v,g minimum due to the zero d/p sensed therefore demanding Max.

flow from the RCIC system. [+0.75] dJ REFERENCE

1. LaSalle: System Description, Chapter 41, pp. 16, 21, 24, and 25, TPO:26, 3a. h u ,e # 5, ge,c. b.S.I L*3.u.

2. WNP-2 System and Procedures, Vol. III, RCIC L.P., pp. 11, 12 and 13.

ANSWER 3.01 (2.00)

a. 1. diesel speedy( 6 ***g;

                                    .5    )

2. load control ;+ 0.5'

b. 1. voltagecontroi [+0. 5] *

2. VAR control [+0.5]

REFERENCE l

1. LaSalle: System Description, Chapter 47.

2. LaSalle: Exam Bank, 3-LS-66.

ANSWER 3.05 (2.00) l

a. Setpoint Setdown prevents vessel overfeeding after a scram i transient. [+0.5] ;

The Setpoint Setdown circuitry reduces the operated selected

          .setpoint by half when a low level trip occurs. [+0.5]

b. Decrease, [+0.5] because the level setpoint would be reduced -

to 18" regardless of the setpoint tape setting. [+0.5]

k

3. INSTRUMENTS AND CONTROLS PAGE 30 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

REFERENCE

1. LaSalle: System Description, Chapter 31, pp. 53, 56, 63, and 64, TP0:10.

ANSWER 3.06 (2.50) pH

a. No ADS response, [+0.5] due to no low pressure pump at pressur/e

[+0.5] -od.- VG l49 dut 4' eo Ip for /M rN f"f" GM M[6 d-

b. No [+0.5]

c. Yes [+0.5]

d. SRV valve pressure can still be supplied from Emergency Pressurization Station [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 37, pp. 14-16 TPO:6.

ANSWER 3.07 (2.00)

a. 70% [+0.5], 5% (volts) for each LPRM not bypassed [+0.5]

b. No [+0.5], there are fewer than two (2) operable inputs on Level B [+0.5]

REFERENCE

                                       ,          i

1. LaSalle: Syst'em Description, Chapter 14, pp. 24 and 62, TP0:4, .

6. s

                                                                                              -w
                                                                                          *=E e

W G

3. INSTRUMENTS AND CONTROLS PAGE 31

ANSWERS -- LASALLE 1 -86/06/03-SLY, G. ANSWER 3.08 (2.00)

a. control valves close 5% [+0.25], open one bypass valve [+0.25]

(or similar answer on diagram).

b. control valves close 5% [+0.25], (reactor scram probable due to increasing pressure since) bypass valves will not be open. [+0.25]

c. 'B' controls, C.V. closes then, reopens to 100% as 'B' takes over. Bypass does not respond. [+0.5] -ode

d. control valves close to 90% [+0.25] to maintain Rx pressure at 920 psig. [+0.25]

REFERENCE

1. LaSalle: System Description, Chapter 25, 26, 73, 74 and 76, .

TPO:4, 5.

2. NMP-2 Operations Technology, EHC, Rev. 1, pp. 2, 5 to 9 of 14, Student Learning Objective Nos. 5, 6, 8, including EHC Diagram. ,

i ANSWER 3.09 (2.50)

a. Manually switch mode se' lector switch to the " Modulate + 2 Step" position ;+0.5;, because the air operated blow off valve will

b. n b25 . j,,,)ll%h

2. shut +0.25

3. shut '+0.25'

4. openl+0.25[

REFERENCE

1. LaSalle: System Description, Chapter 68, pp. 15, 18, 19, and 20, TP0:3b, 6b, 7a. ._

. ANSWER 3.10 (2.50)

b. may be bypassed l+0.5', mode switch not in run I.+0.25]

d. may be bypassed .+0.5" , below 30% power [+0.25] first stage pressure

f. may be bypassed [+0.5], mode switch in run [+0.25] and shutdown -

[+ 0.25] J REFERENCE

1. LaSalle: System Description, Chapter 20, pp. 13 and 14, TPO:6.

~.

3. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 3.11 (1.00)

a. 2. (12) [+0.5]

b. TRUE [+0.5]

REFERENCE

1. LaSalle: System Description, Chapter 19, pp. 8 and 28, TP0:5. '

4 i i i

                                      ,                          i e

5 a0

e 4 e i

     .                                                                                                     1

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 l

RADIOLOGICAL CONTROL 1 ANSWERS -- LASALLE 1 -86/06/03-SLY, G. ANSWER 4.01 (2.50) .

a. SRM - SRM meters only on panel 603 [+0.5]

Level - Narrow Range B&C meters on panel 603 [+0.5]

b. 1. ARM and depress the Manual Scram pushbuttons [+0.5]

2. Place Mode Switch in SHUTDOWN [+0.5]

3. Check Control Rod Position by performing a OD-7 option 2

[+0.5] REFERENCE

1. LaSalle: LOA-AP-08, pp. I and 2.

ANSWER 4.02 (3.00)

a. 1. Individual M/A station input signal abnormal. [+0.75]

2. Drywell Pressure greater than 1.69 psig [+0.75]

3. Loss of 24 VDC to controller [+0.75]

b. Prevent jet pump vibration,p uneven core flow conditions,y[+0.75]

o A- aw t.ustd a n e M A ** [.r k s k lah,k p u (1 . REFERENCE

1. LaSalle: LOA-RR-05, pp. 1, 2 and 3.

2 AL. p. Raw. 314.4.1 , 4.03 Jgup W?c' 5 - 0 , ANSWER 4.03 (2.50)

1. Method 1 - removal of fuses for scram solenoid power [+0.5]

2. Method 2 - drive rods with CR0 system [+0.5] -

3. Method 3 - reset the scram and initiate a manual scram [+0.5] ._

4. Method 4 - individually scram the control rods not inserted

5. eh 5 - remove the pressur ro the overpiston side of the CRD unit (F102 valve) [+0.5] A REFERENCE LaSalle: LOA-NB-09, pp. 1-4.

1.

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

4 ANSWER 4.04 (2.50)

a. b Do least not twosecure or placeindicatipns independent an ECCS in[+0.5] MANUAL mode (gra unless,d #y at w.41 e M '9 i 1. misoperation in AUTOMATIC mode is confirmed. [+0.5] OR

2. adequate core cooling is assured 0.5] co d . G s M ,4.)

b. If an ECCS is laced in MANUAL mode, it et initiate automaticall ake frequent checks of the initiating or controlling parameter [+0.5]. When manual operation 4+-no longer required, restore the sy(stem to AUTOMATIC / STANDBY mode '

if possible).{60,6] , REFERENCE 4

1. LaSalle: LGA-GP, p. 2, Precaution ill.

2. General Electric: E0P Fundamentals, Specific Caution #10. .

ANSWER 4.05 (2.50) I

a. 1. Steam dome space to bottom drain less than or equal to 145

. deg F [+0."r] /vo 7f3 -

2. Idle loop to operating loop less than or equal to 50 deg F [+0.5-] feo.W3

    /[,,' g Jy          3.         Maximum heatup or cooldown rate for the Reactor Vessel or g-i             Recirc. System is 100 deg F/ hour [+0. 5]

b. No [+0.5], due to a maximum imposed limit of 50% power [+0.5]

REFERENCE , i

1. LaSalle: LOA-RR-03, p. 3; LOP-RR-06, pp. 5-9.

2. LaSalle: Technical Specifications 3/4.4.1.1, 3/4.4.1.4, l 3/4.4.6.1.

1 ANSWER 4.06 2.00)

                                                       -w           y,             ,& K /sh G l"*1$-

a. 1. e hours

2. .40 Jeentes [+[+0.5]0.5] 8 0 A *S o @

h ^_** *

b. Immediately leave the area [+0.5] and notify your immediate supervisor [+0.25] and Radiation-Chemir,try. [+0.25]

REFERENCE

1. LaSalle: LRP-1000-1, pp. 24 and 32. ,

t

r

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

ANSWER 4.07 (2.50)

a. Overlap is acceptable if no IRM is high prior to reaching APRM downscale alarms. [+0.75]

b. Prevent excessive loads on the turbine last stage buckets.

[+0.75]

c. Mode Switch in STARTUP [+0.5] and temperature >212 deg F. [+0.5]

REFERENCE

1. LaSalle: LGP-2-2, pp. 10 and 13.
2. LaSalle: Technical Specifications, Table 1.2, pp.1-9.

ANSWER 4.08 (2.50)

a. Pull the fuses for the affected valve. [+0.5]

b. Control switch valve indication following replacement of fuses

[+0.25] or tailpipe temperature. [+0.25] or a^1 A wa4 I t 65 W

c. 1. Four attempts to cycle valve [+0.5] /

2. Pool temperature r.eaches 110 deg F [+0.5] hp4c (2)

3. Two minutes have elapsed [+0. 5]

REFERENCE

1. LaSalle: LOA-NB-02, pp. 2 and 3.

ANSWER 4.09 (2.00)

a. 1. Baron has not been injected [+0.25] and any of the following

2. RPV water level (12.5 in. [+0.25]

3. RPV pressure >1043 psig [+0.25] -

4. drywell pressure >1.69 psig [+0.25]

Yet, [+0.25] to ensure that the intent of steps previously

b. performed [+0.75] -out- would po no(4 A %be altered due to the new entry conditlan. led "I k p e u. i ls u? "f noW w 6 sh pre 3 ass 'jou. us dko REFERENCE k:I %It . (jo.7Q

1. LaSalle: LGA-01, p. 2.

2. General Electric: E0P Fundamentals Bases for E0Ps. -

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND , PAGE 36 RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1 -86/06/03-SLY, G.

                                                                                                                )
                                                                                                                )

l l l ANSWER 4.10 (2.00)

a. Danger on only High Radiation Area [+0.5], Caution on all other radiological signs. [+0.5]

b. For ALL routine access to work in radiologically controlled areas where personnel are NOT expected to receive a whole body dose equivalent of 50 mrem / day. [+1.0] .

REFERENCE

1. LaSalle: LRP-1000-1, pp. 14 and 16.

ANSWER 4.11 (1.00) Immediately verify that at least one CRD pump is operating by inserting at least one CRD at least one notch [+0.5] or place the ~ mode switch in shutdown. [+0.5] REFERENCE

1. LaSalle: Technical Specifications, p. 3/4 3-1

                            ,                  i 6

m W 4 9 l

TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 01.01 2.00 SLY 0000765 01.02 3.00 SLY 0000766 01.03 2.00 SLY 0000767 1 01.04 1.00 SLY 0000768 01.05 2.25 SLY 0000769 01.06 2.25 SLY 0000770 01.07 3.00 SLY 0000771 01.08 2.50 SLY 0000772 01.09 3.00 SLY 0000773 l 01.10 2.00 SLY 0000774 01.11 2.00 SLY 0000775 25.00 l 02.01 2.00 SLY 0000740 02.02 1.50 SLY 0000741 02.03 1.50 SLY 0000742 - 02.04 2.00 SLY 0000743 02.05 1.50 SLY 0000744 02.06 2.00 SLY 0000745 02.07 1.50 SLY 0000746 , ! 02.08 1.00 SLY 0000747 02.09 1.00 SLY 0000748

02.10 2.50 SLY 0000749 02.11 2.50 SLY 0000750 02.12 1.00 SLY 0000751 .

02.13 1.50 SLY 0000752 02.14 1.50 SLY 0000753 02.15 2.00 SLY 0000776 25.00

03.01 3.00 SLY 0000729 03.02 2.50 SLY 0000730 f 03.03 3.00 SLY 0000731 03.04 2.00 SLY 0000732 03.05 2.00 SLY 0000733 03.06 2.50 SLY 0000734 03.07 2.00 SLY 0000735 -

03.08 2.00 SLY 0000736 . 03.09 2.50 SLY 0000737 03.10 2.50 SLY 0000738

03.11 1.00 SLY 0000739 i ______

25.00 04.01 2.50 SLY 0000754 i 04.02 3.00 SLY 0000755 04.03 2.50 SLY 0000756 04.04 2.50 SLY 0000757

        ,_          _j                u         a

a .- - - - a

  ",~ .

TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 04.05 2.50 SLY 0000758 04.06 2.00 SLY 0000759 04.07 2.50 SLY 0000760 04.08 2.50 SLY 0000761 04.09 2.00 SLY 0000762 04.10 2.00 SLY 0000763 04.11 1.00 SLY 0000764 25.00 100.00 W e I l ., - j l l 4 1

,}}

ML20206S243

Text

SUMMARY

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