ML20199F178

Exam Rept 50-456/OL-86-01 on 860312,0414-17 & 21-24.Exam Results:All But One Senior Reactor Operators Passed Written & Operating Exams & All Reactor Operators Passed Written & Operating Exams

ML20199F178
Person / Time
Site:	Braidwood
Issue date:	06/13/1986
From:	Burdick T, Isaksen P, Picker B, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199F161	List: ML20199F158 ML20199F178
References
50-456-OL-86-01, 50-456-OL-86-1, NUDOCS 8606240215
Download: ML20199F178 (100)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-456/0L-86-1 Docket No. 50-456 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Braidwood Nuclear Power Station Examination Administered At: Braidwood Nuclear Station and Production Training Center Examination Conducted: March 12 and April 14-17, 21-24, 1986 Examiners: T. Reidinger v'J T. Burdick 6 61

((

Date B. Picker [./- - 'l P. Isaksen

[t h Date' '

/ /

Approved By: T. Burdick, Chief /[

Operating Licensing Section Bate' Examination Summary Examination administered on March 12, 1986 (Report No. 50-456/0L-86-1)

And on April 14-17, 21-24, 1986 to 17 senior reactor operators and six reactor operators.

Results: All but one of the senior reactor operators passed the written and operating examinations and all reactor operators passed the written and operating examinations.

8606240215 860616 6 DR ADOCK 0500

REPORT DETAILS

1. Examiners T. Reidinger R. Higgins B. Picker P. Isaken

2. Examination Review Meeting Utility comment and their resolutions are attached.

3. Exit Meeting Observations and concerns addressed by the Chief Examiner included the following subjects. No copy of 10 CFR could be located in the Shift Engineers office, Center desk or reactor operator's desk, this hindered questions by the examiners relating to 10 CFR during the oral examinations, i.e. airborne release limits, shift manning etc. Observations noted during the simulator examinations include the following points, or recommendations, the candidates need to walk down the board (review each control panel) after a electrical loss of supply bus, the candidates would not notice the failure of motor operated valves or equipment or their indications until the plant (simulator) was in an extremis condition. Need to update the normalized currents to the nuclear instrumentation for the performance of the quadrant power tilt calculations during simulator performance. No procedure exists presently for shifting RCFC's from fast speed to slow speed. Simulator performance for the majority of the candidates were partially degraded in at least one specific area relating to realistic communications. The candidates prior to using the. telephones to communicate with the various simulated plant personnel during casualties would invariably look away from the control panels to look for the simulator operators. One recommendation made by the Chief Examiner for future examinations in August 1986, would be to enclose the simulator console with smoked glass or with one way mirror to enhance plant realism.

The majority of candidates exceeded the technical specification of i containment pressure or temperature during routine evolutions due to i inattention to plant parameters. Various simulator scenarios when run l uncovered major glitches relating to the poor modeling on equipment performances, these problems were discussed with Production Training Center's Manager, Jim Harris for possible future corrections. There l were no other generic weaknesses noted during simulator / oral examinations.

The candidates in general gave a strong performance during simulator examinations in all phases of routine, abnormal, and emergency evolutions and procedures.

2

1 BRAIDWOOD REACTOR OPERATOR AND SENIOR OPERATOR EXAMINATION COMMENTS AND RESOLUTIONS

l l

l R0 Exam Section 1.0 1.1: An additional method of hydrogen generation is discussed in our Mitigating Core Damage text Pages 1-8. In that text, it states that " additional small amounts of hydrogen may also be generated due to the high temperature metal-water reaction between stainless steel (Fe) and water." Therefore, an additional

, acceptable answer should be " stainless steel (or iron) and water

! reaction."

Examiner's Comment: In that same paragraph, it states ". . . The amount of hydrogen produced by stainless steel is determined to be negligible in comparison to the zirconium water reaction." Will not accept the utility suggested alternative answer.

1.2: An alternate correct answer should be as follows: "When in Mode 2 with Keff less than one, the shutdown margin can be verified by completing the shutdown margin surveillance (IBWOS 1.1.1.1.E-1). This will ensure the combination of core temperature, boron, rod position, xenon and samarium present in the core meets the minimum required shutdown margin."

Examiner's Comment: Will accept the alternative answer, with the conditions

, that for full credit the candidates state the shutdown margin surveillance will ensure the combination of core temperature boron, rod position, xenon and 4

samarium present in the core meets.the minimum required shutdown margin.

1.3: If the shutdown margin is inadequate when the reactor is in Mode 1, emergency boration is required until rods are above the low-low insertion limit, vice the low insertion limit, as stated in the answer key.

References:

IBw0A PRI-2 and

, T.S. 4.1.1.1.1.C.

1 Examiner's Comment: The typographical error of omitting " low-low" vice " low" I

will be corrected.

I 1.5: One pump draws higher amperage per motor versus four RCP's running in hot shutdown. Decreased flow resistance encountered in the reactor vessel and piping allows a higher flow rate through the single pump running and this results in higher amps.

Examiner's Comment: Disagree, four pumps running hot shutdown draw more per motor amperage than single pump running in hot shutdown. This data was supported during simulator examinations at the Production Training Center.

, The utility in subsequent phone calls indicate that presently the modeling on the simulator for motor amperage is now suspect and not accurate to the best of their knowledge. Will presently accept the utility response.

4 2

J

-y- n e ,- -,,ww + r , - - , , - - - g w,-- , - - - - - --~g -

T 1.7: A 2% band in accuracy is suggested to determine an acceptable _  !

numerical answer. If work is. shown, credit should be given for proper use of the steam tables.

1 Examiner's Comment: As explained prior to.the examination, it was agreed if

work is shown then credit will be given for defined work problem.

1.8: The answer key requires a two part answer, whereas the question

asks for "the primary reason." The answer should be considered

, as a single response.

Examiner's Comment: The answer, in reality, is a single answer with credit l given to the full explanation of the question. The examiner does note the l utility's concern however.

I 1.9: Use of the Mollier diagram provided with the exam yields answer j in the area of.285+/-5 degrees Fahrenheit. A band of i 280-300 degrees Fahrenheit should be considered as an acceptable range for answers.

Examiner's Comment: Agreed.

1.10: a. Fuel clad creep is an alternate term describing the gap reduction mechanism and should be considered as an acceptable answer.

4

b. Westinghouse Core Control text, Chapter 2, Page 45 states that the result of clad creep is a decrease in the gap i size...and a marked decrease in effective fuel '

temperature... The answer should be revised to include this concept as an acceptable answer.

Examiner's Comment: Agreed, these are equivalent acceptable responses. l 1.11: a. The correct answer is moderator temperature defect vice coefficient.

Reference:

Westinghouse Large PWR Core Control, Chapter 3, Page 41. j i

b. The single correct answer is that the decrease in critical r boron concentration causes the increased negative reactivity at EOL. The decrease in boron concentration is a result of i fuel burn-up and fission product buildup in the core center.

. These effects cause flux to shift to the' core extremities

] and leakage increases. Therefore, moderator temperature 4

coefficient is-more negative. Acceptable answers should

include boron concentration changes, fission product buildup and neutron flux shifts.

4 Examiner's Comment: a. The referenced text was not provided for review. The

, integral for moderator temperature coefficient is moderator temperature defect.

No candidate stated moderate temperature defect. b. The utility provided one I. additional response of neutron flux shifts, which is an acceptable alternative, f

3 I

1.17: The answer provided from the MCD text, Pages 2-19 and 2-20 is specifically for a " cold leg break of approximately two inches diameter." Since the question does not specify break size, a inore general approach to explain why cold leg breaks are more severe than hot leg breaks should be accepted. Per the referenced material, ECCS the RCS (Qg flow (M.

) versus hdSt) removed versus break (Q flow (Maretggt)andheatinto two key concepts duringallRCSbreaksthatdetermineEka)dtemperaturesfollowing the accident. With this concept in mind, then all cold leg breaks will result in higher clad temperatures as compared to hot leg breaks since the cold leg break will result in a longer time for M to equal or exceed M due to the loop seal in the cold legrd9trictingthepasssge8ytcore steam out the break (Q ).

Once Q Q >M U andcorerecoverywillpro8dhd.

UntilEHis> tide,,theM thecb9eisdbcoveringandclad(fuel) temperatures are increasing.

Examiner's Comment: The facility provides an interpretative revised narrative of the original answer; the question was of an intuitive concept as realistically cold leg breaks greater than two inches do not exhibit the relationship expressed in the question, i.e. cold leg breaks result in higher fuel and clad temperatures than hot leg breaks of the same size. Will accept the generalized version provided by the utility.

1.20: Per Westinghouse Reactor Core Control, Chapter 4, Pages 4-29, the reactor power level and the value of moderator temperature coefficient will dictate if the oscillation will converge or diverge after rod position is reestablished. An answer that discusses convergent or divergent oscillations based on the power level and value of the MTC should be considered as correct also.

Examiner's Comment: The text mentioned was not provided for review; the answer will not be revised.

4

Section 2.0 2.1: There are actually seven trips and the points should be spread equally among the seven. Answers 4 and 5 each contain two trips.

Reference:

BwGP 100-3, Page 7, note after Step 21.

Examiner's Comment: The examiner notes the facilities concern.

2.3: a. An additional acceptable answer should be in the "chromated drain tank," which is a subsystem of the auxiliary building equipment drains system.

Reference:

CCW System Description, Pages 17 and 18.

b. Since control switches are taken to the pull-to-lock position when equipment is taken 00S, it would only be necessary to have the breakers racked in to be able to start the pump. An acceptable answer would thus be:

(1) Control switch in pull-to-lock, OR (2) Given that the 1A pump is 00S (hence, its C/S is in P-T-L) only the breakers need to be racked in.

Reference:

CCW System Description, Page 12.

Examiner's Comment: a. Will amend answer key to included chromated drain tank.

b. Will accept the conceptual aspect of the facility response. For full credit the answer key will be amended to reflect the following answers:

(1) CCW pump (1A) C/S in P-T-L (2) 0-CCW pump breaker - racked in 141 (3) 0-CCW pump breaker on 383' racked in (4) 0-CCW pump C/S in after close position 2.5: An additional source of water for centrifugal charging pumps during an ECCS actuation is the containment sump via the RHR pump discharge.

Reference:

BwFR C-1 and ECCS System Description, Page 41.

Examiner's Comment: Will include the additional response to the list of possible acceptable answers.

2.9: Each RHR system suction relief valve is sized to remove the capacity of one centrifugal charging pump operating at the start of cooldown or to remove the capacity of two centrifugal 1

charging pumps under cold shutdown conditions. (

Reference:

RHR System Description, Pages 12-13.) Either of these answers should be considered as valid answers to the question and be given full credit.

5

1 P j

l I l l

Examiner's Coment: Will accept.the revised answer and the answer key will be l amended.

i 2.11: Although it is recognized that the PRT. is a correct answer, an.

t alternate acceptable answer should be RCDT since the seal water ,

can flow through the No. 2 RCP seal to its leak off to the-RCDT. '

Acceptable answers should thus be:

1 (1) PRT

] (2) RCDT i

] (3) A 1

I (4) C

Reference:

Braidwood Systems Training Manual, Chapter 13, Pages 17 and 18.

! Examiner's Coment: Will accept the alternative answer for this examination

! only. For future study and reference the facility is advised that the NRC will

! not accept the facility alternative answer of No. 2 RCP seal leak off to the.

, RCDT as an acceptable answer for this type question. The No. 2 RCP seal. leak i

off per facility reference material is three gallon per. hour (gph) as opposed

to RCP seal injection flow of 480 gph per RCP. Approximately 99.375% of the seal water flow will go to the PRT on Phase A isolation.

2.12: b. An additional signal that actuates the feedwater hammer prevention system is " Steam' generator, narrow range level-less than 5%."

Reference:

Condensate and Feedwater System l Description, Pages 61-64.

t l

Examiner's Comment: Agreed answer key will be amended to include additional

response.

2.16: Answer key shows 2 points for the question, whereas the question

sheet shows 1.5 points. It appears that the 1.5 points assigned i on the question sheet is the correct value.

1 Examiner's Comment: The examiner notes the comment.

I 2.14: Per 1BwEP-1, Step 14, ECCS transfer to cold leg recirculation i is initiated when RWST level is'less than 46%.

In addition, per 1BwEP ES-1.3, Transfer To Cold Leg Recirculation '

Unit 1, the entry condition from 1BwEP-1 discussed above is referred to as " low-low RWST' level." 'This is also stated on

! Page 79 of the PLS document, which shows the low-low level alarm l

] setpoint at 46.7% for the RWST. Any of the following answers 1 j should be awarded full credit: l

. (1) Less.than 46%

1 4

6 4  ;

I

(2) Below low-low alarm setpoint (3) Less than 46.7%

Examiner's Comment: The question and answer for this reference key was BWFR-C.3 which stated "less than 45%." Facility provided references as to "less than 46%," "less than 46.7%," and "below low-low alarm set point. In view of all the inconsistencies and inaccuracies in the various reference material, the examiner will accept all equivalent answers.

2.15: a. Step 33 of BwGP 100-5 (Page 13) is a major step in the procedure that calls for an accumulator line-up for cooldown between 800 to 1000 psig. An acceptable alternate answer to this question should be 800 to 1000 psig.

Examiner's Comment: Answer key will be slightly amended to include the higher range of 1000 psig.

2.17: It is agreed that manually shifting the charging and SI pump suctions during cold leg recirculation to the RHR heat exchanger discharge in a prescribe sequence will prevent sending contami-nated SI pump flow to the RWST. However, the primary reason for manually shifting the charging pump suctions during that phase is to verify automatic actions and realign the pumps' suction to the sump. Therefore, the correct answer should read " provide suction flow to the SI and charging pumps from the RHR pumps and contain-ment sump during cold leg recirculation."

Reference:

ES-1.3 Lesson Plan, Page 88.

Examiner's Comment: The facility did not provide the reference lesson plan.

Will accept the facility provided response in addition to the original answer key.

2.18: (2) Acceptable answers for this question should be expanded to include the C loop hot leg. Reference RCS System Description, Page 41. Since only one loop was requested in the question, credit should be given to either the A or C loop hot leg as an acceptable answer.

(4) & (5) The answers B--cold leg or A--cold leg should be acceptable for both 4 and 5 as long as the same answer is not given for both 4 and 5.

The PLS document Pages 58 and 59 and Bw0P CV-23, Page 2, direct normal and alternate charging locations to be alternated each fuel cycle.

Examiner's Comment: Will amend the answer key to include - C loop - hot leg if the candidate chooses either A or C loop hot leg.

i l

7

Section 3.0 3.1: a. Per the PLS document, Page 20, auctioneered low Tave feeds the C-16 control interlock. This an alarm function only.

An acceptable answer should be C-16.

b. Cold, over pressure protection does not use auctioneered low Tave. PORV-456 uses wide range T temperature while PORV-455Ausescoldlegtemperatuh8t Each has an associated pressure input. (

Reference:

Pressurizer Pressure and Level Control System Description, Page 46, Figure 14-13C.)

In addition on Page 31 of the PLS document, the auctioneered low RCS temperature referred to is not RCS autioneered Tave but wide range T and T c

The auctioneered low Tave circuit used narhN range R}ds. .

As worded, there is only one correct answer for the question. That answer is C-16.

Examiner's Comment: a. C-16 control interlock is equivalent to turbine loading stop and therefore acceptable. b. The RCS Cold Over pressure Mitigation System utilizes a auctioneered low RCS temperature vice a auctioneered low Tave; the answer key will be amended.

3.4: The two answers stated are correct. In addition, however, the following are also correct answers to this question:

(1) Control switch for CS pump must be either in after trip or after-close; (2) Educator suction valves for each train must be in the fully-stroked open position for the pump to auto start.

Since the question only asked for two of the possible four correct answers, any two of the four should be accepted for full credit.

Reference:

Containment Spray System Description, Pages 21 and 22.

Examiner's Comment: The additional answers are acceptable. The examiner notes that this revised reference material was supplied after the construction of the examination.

3.5 The question does not give information to lead one to believe that turbine power or nuclear power differ from each at the start of turbine and nuclear power decrease. As such, there should be no requirement to state "even though one may be higher than normal, there will be no error output to move rods."

Examiner's Comment: Notes comment.

l 8 ,

1

I 3.7: The correct answer for this item is that the steam generator blowdown isolation valves close on: (

Reference:

Liquid Radwaste System Description, Page 30).

(1) Containment Phase A signal (2) High temperature in the blowdown valve room (3) Manual l Examiner's Comment: The examiner notes that in the secondary sampling procedure it mentions that the blowdown isolation valves of the steam generators automatically close on a Phase "A" signal or a loss of power. Will accept the additional answers provided.

3.14: a. The answer key should read " low pressurizer pressure SI, less than 1829 psig."

Reference:

PLS, Page 9.

Examiner's Comment: The setpoint was amended.

3.15: In addition to the answers listed, PT-505 also provides the steam generator water level control system with an input for water level program.

Reference:

Main Steam System Description, Page 26.

Examiner's Comment: Additional answer is acceptable.

3.13: Answer B, containment purge process radiation high alarm, is not referenced in the System Description. The correct answer should be:

(1) Containment Atmosphere Area Radiation Monitor High (2) Phase A Isolation (manual)

(3) SI (4) Containment Spray Actuation (manual)

(5) SI Actuation (manual)

Reference:

Containment Vent and Purge System Description, Page 57.

Examiner's Comment: The examiner will delete the containment purge process radiation high alarm as it is not now described in the revised System Descriptions.

4 a

t 9

Section 4.0 4.4: It is agreed that the first two actions for a high RCP bearing temperature of 225 F should be to trip the reactor and then trip the affected pump. The answers elicited by the question are, however, contained in a caution of procedure that is not required to be memorized. Also, the caution has three actions listed with closed bullets which require all steps to be done in any order. A correct answer for this question would be to list any two of the three immediate actions specified in that caution without regards to order. Therefore, an additional acceptable answer to this question would be "Go to 1BwEP-0, Reactor Trip or Safety Injection."

Reference:

IBw0A RCP-1 Examiner's Comment: Will not accept facility response. The procedure specifically requires that in Step 8(a) that the required steps be performed in those sequenced steps, i.e. (1) Trip reactor, (2) Trip affected pumps (RCP),

(3) Go to 18wEP-Rx trip Safety Injection.

The facility's use of the caution step in allowing the operator to challenge the reactor protection system at least two times runs counter to good engineering and Westinghouse practices. As in this first example, the utility maintains it is acceptable for the operator when noting a RCP bearing temperature reaching 225 F the operator is then allowed to:

(1) Trip the affected pump (while at power)

(2) Trip the reactor (manually)

(3) Go to 18wEP-0 (Reactor trip or Safety Injection) Procedure or in the second case:

(1) Go to 18wEP-0 (Reactor trip or Safety Injection) Procedure (2) Trip the affected pump (while at power)

(3) Trip the reactor (manually)

Tripping the pumps (RCP) while at power challenges the reactor logic protection system.

4.5: On the answer key, Item 2 should read, " containment radiation greater than 105 R per hour" vice "10 R per hour." This appears to be a typographical error.

Reference:

1BwEP-0, Page 3.

Examiner's Comment: Typographical error was corrected to 105 R/hr.

10

4.7: It is suggested that only the underlined portion of the answer in the answer key be required to completely answer the question and receive full credit.

Examiner's Comment: Agreed.

4.11: The answer key has listed Items A through I of Paragraph E2 in BwGP 100-6. It is requested that the answer key also include Items J, K, and L under that same paragraph which are shown on the next page. Thus, Item J would be " loss of minimum AC electrical busses per T.S. 3.8.3.2," K " loss of minimum AC sources per T.S. 3.8.1.2," and L " loss of minimum DC sources per T.S. 3.8.2.2."

Another acceptable answer for suspending core alterations in the fuel handling building is "less than one fuel handling building exhaust ventilation plenum operable" per T.S. 3/4.9.12."

One additional condition that requires suspension of core alterations is inadequate boron concentration per T.S. 3/4.9.1.

In this case, core alterations are suspended if the Keff is greater than or equal to .95 or boron concentration is less than 2000 PPM.

j In summary, five additional answers are seen as acceptable responses to the question.

i Examiner's Comment: Will accept the additional responses listed in the first j paragraph.

Will not accept response listed in the second paragraph.

2 T.S. 3/4.9.12 "less than one fuel handling building exhaust ventilation plenum operable as the technical specifications state the applicability" whenever irradiated fuel is in the storage pool. The examination question specifically stated that Braidwood was under going a fueling operation, i.e. first clean core load not a refueling operation.

Will accept the alternative of boron concentration and Keff while noting that the refueling procedure specifies that these conditions apply in Mode 5 and the technical specifications conditions apply in Mode 6.

i 11 n- -w ,, , ,- - . , + - ~

BRAIDWOOD SR0 WRITTEN EXAM COMMENTS The following facility comments were made about the SR0 examination. Each facility comment is followed immediately by the NRC response.

Question 5.04A Facility Comment:

Although a control rod's worth is proportional to the square of the relative flux it sees, that relationship exists because control rod worth is proportional to relative flux times the importance factor.

Since the importance factor is proportional to local flux then the control rod worth becomes proportional to the square of the relative flux. Answers that discuss these additional concepts should be considered as correct and receive full credit. (

Reference:

Westinghouse Large PWR Core Control, Chapter 6, Page 15).

NRC Response: Agree. Reasonable answers were accorded full credit.

Question 5.048 Facility Comment:

If one rod is inserted adjacent to another rod, the effect of the second rod on the first rod will depend on the proximity of the second rod to the first rod. In one instance, the first rod's worth will be  !

lower and in another instance, it's worth can be higher, depending on their relative positions. Answers that state that the first rod's worth will be increased should be accepted provided a suitable explanation is given. (Reference. Westinghouse Large PWR Core Control, Chapter 6, Page 27).

NRC Response: Agree. Reasonable explanations were accorded full credit.

Question 5.07A Facility Comment:

Build up of xenon to equilibrium is generally discussed in terms of 40 to 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> vice 45 to 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br />. Answers in the range of 40-55 should be accepted. l NRC Response: Agree. The answer key was changed. I 1

Question 5.078 Facility Comment:

The change in Samarium is negligible. As a result, the weighting assigned to this part of the answer appears to be excessive.

NRC Response: Disagree. Samarium is an important fission product. The answer key was not changed.

v-w- - - - . e + -

Question 5.09A Facility Comment:

An explanation that discusses the opposite situation,.i.e. a larger shutdown margin, should also be accepted for full credit.

NRC Response: Agree. Answers which state that "the greater the shutdown margin, the longer time required to reach a stable count rate" were granted full credit.

Question 5.09B Facility Comment:

An explanation that discusses a lower initial count rate should also be accepted for full credit.

NRC Response: Agree.

Question 5.10B 2) Facility Comment:

i The answer should read, " Fuel temperature decreases as core ages, so the FTC will increase (become more negative)." Figure SNP-RF-11,

! titled " Doppler Temperature Coefficient Curve" from the Westinghouse

Large Pressurized Water Reactor Core Control Text shows that the fuel temperature coefficient becomes more negative at end of life at lower effective fuel temperatures, whereas at higher effective fuel temperatures, the reverse condition is the case. Thus, the answer would be dependent on effective fuel temperatures. For the Braidwood reactor, illustrated in FSAR Figure 4.3-27, which describes Doppler temperature coefficient at BOL and E0L for cycle one, it is seen that we operate on the part of the curve where an end of life Doppler temperature coefficient is more negative, and increasingly so at lower effective fuel temperatures. Westinghouse Core Control Text, Chapter 2, Page 42, discusses the reason for this and states that the E0L Doppler temperature coefficient becomes more negative at lower effective fuel temperatures due to the build up of Plutonium 240.

NRC Response: Agree. The answer was changed.

Question 5.13 Facility Comment:

From Braidwood's Technical Specification Basis, Page B 3/4 1-3, rod insertion limits ensure that:

(1) Acceptable power distribution limits are maintained, (2) The minimum shutdown margin is maintained, (3) The potential effects of rod misalignment on associated accident analyses are limited.

The above items should be considered as acceptable answers also.

NRC Response: The answer was modified to grant credit for either " misaligned rod" or " ejected rod."

2

Question 5.14 Facility Comment:

Due to two incorrect values used in the answer key, several parts of this answer are incorrect. The initial incorrect value used as the saturation temperature for 1,065 psia is actually 552.27 F vice 550 F. The second incorrqct value utilized is 587 F for Tave whereas the correct value is 588.4 F. Revising these two values yields new answers for several parts of the problem. The correct values in the order as they are shown in the answer key are 1,065 psia, 552.27 F, 1,190 psia, 566 F, 13.72 F, 527.7 F and in the last sentence of the answer key, it should read "If Tave rises to 572.7 F + 13.72 F at 50% power...." Reference for the Tave value is Technical Specification Table 2.2-1 (Page 2-8).

NRC Response: Agree. The answer key was changed to reflect the correct values.

Question 6.02 Facility Comment:

An alternate acceptable answer should be that "The undervoltage trip of a reactor coolant pump breaker, as a result of a reactor coolant pump bus undervoltage, is a direct trip of the reactor coolant pump breaker. (

Reference:

Reactor Coolant Pump Systems Description, Page 30). Also since only the term degraded voltage is used in this question and answer, the term undervoltage is used in all the reference materials to describe the actions of reactor coolant pump breakers, there appears to be an unnecessary confusion factor introduced because i there can be a degraded voltage condition on the 4160 volt ESF buses (91% voltage) that causes protective actuations after a time delay.

The confusion between the terms degraded voltage and undervol' age suggests that a broad range of acceptable answers be considered for full credit. Another factor is that the point value for this question (2 points) is inordinately high as compared to other questions of the same length and level of difficulty.

NRC Response: The answer was modified to also grant full credit for the response "undervoltage trip." The question's point value was not changed.

Question 6.05 Facility Comment:

Per BwAR 1-21-E8, the causes for a 125 volt DC battery charger trouble alarm are as follows:

(1) Loss of AC power supply, (2) A High DC Voltage trip of the AC input breaker, or electrical fault, 1

(3) Battery charger DC output voltage low, (4) Charger failure (AC output near zero amps),

(5) Blown fuses, 3

(6) Battery charger output voltage high, j (7) Charger output breaker trip.

NRC Response: Agree, the answer was modified to grant full credit for any three of the above responses.

Question 6.08 Facility comment An additional answer per Bw0P DG-3, Page 2, should be " Ensure no voltage indication exists on the diesel being secured."

NRC Response: Disagree. "No voltage indication" is not in and of itself proof that the diesel has shutdown after its cooldown is complete.

Question 6.09 Fic#'ity Comment Inasmuch as the manual / auto operation was not specifically identified, it can be construed that an appropriate answer to this question would be that if the controller was put in manual, then one could manually I control position of the letdown divert valve. An answer that addresses i this type of operation should be considered as an equally. acceptable  ;

alternate answer.

NRC Response: Agree. The answer was modified to also grant full credit for the response, "in auto, the valve returns to the VCT, but not in manual."

Question 6.15A Facility Comment:

The steam generator PORV controller does not have potentiometers which control the setpoint of the PORV. The only way to control the PORVs from the controller is to put the controller in manual and open and close them in that manner. The range of PORV opening is, however, 1115 to 1175 psig. The superseded System Description for Main Steam discussed this potentiometer setpoint range of 115 to 1175 psig as shown in the answer key. However, the new Main Steam System Description deletes any reference to potentiometer control of the setpoint.

Acceptable answers for this question should be either:

(1) There is no setpoint adjustment available (i.e. the " pot" has no effect), or (2) 1115 to 1175 psig, which is the range of PORV operation.

~

Electrical drawing EM-w 31MS19 includes the M/A station and shows no potentiom a er c.w. trol in the circuit.

NRC Response: Agree. The answer was changed to "1115 to 1175 psig, or no setpoint adjustment," either of which will be granted full credit. The facility is cautioned to improve the editing and the proofing of system descriptions.

4

i Question 6.16a Facility Comment:

Since this is a calculational type problem, partial credit should be given for recognizing portions of the calculational methods.

A suggested breakdwon is as follows:

(1) Recognize 50% steam dump capacity is 20% of rated steam flow.

(.2 points)

Reference:

Steam Dump System Description, Page 5.

(2) State that rated steam flow is 15.14 x 106 (or simply 15 x 106 )

lb /hr (.2 points)

Reference:

Main Steam System Description, Pa$e9.

6 (3) Calculations (.2) x (15.14 x 10 lb,/hr) = 3.028 l 6

j x 106 lb,/hr (or simply 3 x 10 lb,/hr) (.1 points).

NRC Response: Agree. The answer was modified to breakdown the point distribution so that partial credit could be awarded.

Question 6.17 Facility Comment:

Although many examinees may have recognized the event report as described in this question and knew from that the reason for the loss of RHR flow indication, the arrangement at Braidwood of the RHR flow elements, which are upstream of the cross-connect line, would invalidate the rationale behind the loss of flow indication and could lead one to either an erroneous or no conclusion regarding the cause.

Reference:

Braidwood P&ID, Page M-62 NRC Response: Disagree. Though the indications for RHR flow may be different than described in the question, the RCS depressurization should give the examinee enough information to diagnose the problem.

Question 6.18 Facility Comment:

a. The answer to this question is incorrect. If this question is not deleted, then two acceptable answers should be:

(1) The aux. lube oil pump does start automatically on auto initiation of an aux. feedwater pump; OR (2) The aux. lube pump is to be locally started prior-to securing the aux. feedwater pump to provide lube oil flow during pump coastdown.

Reference:

AFW System Description, Page 13 and Bw0P AF-3 1

5

k ]

4

b. The previous system descriptions submitted indicated the answer shown as outlined in 6.18.B. However, the current systems description submitted as reference material for-this exam did not reference right or left with regard to the flow lights.

The Aux. Feedwater System Description only mentions the value of 80 and 160 gpm as values at which the lights would actuate.

It is suggested then that the requirement to list right or left with the numbers 80 and 160 be eliminated. -Acceptable answers should then be simply: 80 and 160 gpm. An alternative acceptable answer that should be considered for at least partial credit is "to indicate flow conditions on the feed train."

c. The current Aux..Feedwater System Description does not mention any annunciator for the low flow condition described in your answer key.

It is suggested that this question be deleted. If it.is not deleted, -

alternate acceptable answers to the one listed in the answer key would be "by observing control room instrumentation (flow control

. valve position), controller.(potentiometer) setting and/or status display lights for the valves."

In view of the multitude of problems with this question, it is suggested that the entire question (6.18) be deleted.

NRC Response: Partially agree. The question was not deleted, but the following modifications were made:

J l 6.18.a The answer key was changed to grant full credit for either response-

" aux lube oil pump does start automatically" or " aux. feedwater pump to provide lube oil flow during pump coast down."

, 6.18.b References to right or left were deleted from the answer key.

I 6.18.c Answer key was modified to also accept the answer " valve lineup checks."

Question 6.20 Facility Comment:

a. In addition to the 4160/480 transformer as an answer,- full credit I should also be given to listing or stating the loads that come off i the transformer,.such as bus 131X or 132X.

Reference:

Braidwood i Systems Training Manual, pp. 46-48

b. Acceptable answers should be 50 seconds from safety injection

actuation or 20 seconds from receipt of containment spray l actuation signal, depending on how the question was.

Reference:

Diesel Generator and Auxiliaries System Description, Page 62.

NRC Response:

- 6.20.a- Disagree. The examinee should know and mention the 4160/480 v transformer.

I

6.

6.20.b Agree. The answer was modified to also grant full credit for "50 seconds after SI," "20 seconds," or "first start is missed and must wait for second start."

Question 7.01.b and c. Facility Comment:

Parameters not available at the Remote Shutdown Panel (RSP) due to malfunctioning or even because of its not being displayed on the RSP may be viewed in some cases locally as well. Acceptable alternate answers for B & C should include " local indications in the plant."

NRC Response: Disagree. The examinee should know the additional remote locations which are available from which plant parameters may be monitored.

Question 7.06.a Facility Comment:

An additional acceptable answer to "What three conditions determines whether a main steam line should be isolated per the immediate actions?,"

is that in the RNO column of Step 2 of the Reactor Trip or Safety Injection Procedure, it states that "If the turbine still will not trip, then initiate steam line isolation." Credit should thus be givea for any three of the possible four answers to this question.

NRC Response: Agree. The answer was modified to include " turbine has not tripped after having attempted to manually trip" as a correct response.

Question 7.09 Facility Comment:

Although D is recognized as a correct answer, i.e. establish ongoing communications between the control room and TSC in an emergency, ongoing communications are also necessarily established between the control room and the NRC when incidents are reported under 10 CFR 50.72. In fact, that document specifies that in addition to prompt telephone notification the " licensee shall also establish and maintain an open, continuous communication channel with the NRC operations center and close this channel only when notified by NRC." Since in the question, the term emergency was not clarified as to its intent, i.e. site emergency, general emergency, unusual event, etc., it is left open to the examinee to interpret the question from the standpoint of a Senior Reactor Operator. From that standpoint, it is realized that initial ongoing communications will be established with the NRC from the control room before control room and TSC communications are established. It is therefore seen that "A,"

i.e. establish emergency ongoing communications between the control room and the NRC, is an acceptable alternate answer.

l NRC Response: Disagree. "0ngoing conversation" implies a continuous i communications link. The control room will transfer to the TSC or EOF the  ;

communications link with the NRC as soon as the respective facility is '

operational. Therefore response "a," " control room and NRC," is incorrect.

7

Question 7.17 Facility Comment:

Answers that list those situations where the word " Caution" is used

should also be considered as valid answers. Caution is used in the following areas:

(1) Radiation area, (2) Radioactive materials area, (3) Airborne radioactivity area.

Reference CECO. Radiation Standards, Page 9.

NRC Response: Agree. Any of the following answers: " radiation area,"

" radioactive materials area," or " airborne radioactivity area" will be awarded full credit.

Question 7.18 Facility Comment:

a. Routine or Type 1 RWPs are valid for a maximum of one year from January 1.

Reference:

Radiation Protection Standards, Page 11.

b. Type 1 RWPs are required for all routine access or work in radiologically controlled areas where personnel are not expected to exceed a whole body dose equivalent of 50 mrems per day.

(

Reference:

Radiation Protection Standards, Page 11).

NRC Response:

a. Agree. The answer was changed to "one year from January 1."

b. Agree. The answer was changed to "if the daily dose will not exceed 50 mrem and the work is routine."

Question 7.19 Facility Comment:

b. Per BwFP FT-15, thimble plugs may be removed and installed in the following places:

(1) Fuel transfer system basket (containment upender),

(2) RCCA change fixture, (3) Spent fuel rack.

It is also true that thimble plugs are inserted at the manufacturer as well. Additional acceptable answers should thus include the spent fuel rack, and at the manufacturer's.

i 8

.- . - = . - ._, ,

c. Per BwFH-8, the spent fuel pit should also be an acceptable answer.

d. Per BwFH-14, an additional acceptable answer mentioned in Paragraph D.8. includes the verification that a slack cable light is illuminated. Also from Procedure BwFP FH-5, the requirement for 1/M data can be satisfied by subcritical data. Therefore, acceptable additional answers should be " slack cable light is illuminated" and "subcritical data."

NRC Response:

b. Agree. The answer was changed to grant full credit for any of the three following answers: " fuel transfer system basket (containment upender),"

"RCCA change fixture," or " spent fuel rack."

E Agree. The answer was changed to grant full credit for either "RCCA change fixture" or " spent fuel pit."

d. Agree. The answer key was modified to also grant full credit for " slack cable" or "subcritical data."

Question 7.20 Facility Comment:

An additional acceptable, clarifying answer to this question is that it must be left on the low scale so as to ensure proper operation of protective features since they are actuated by percent of meter reading.

(

Reference:

Fuel Handling System Description, Page 28).

NRC Response: Disagree. The lack of load cell limit circuit calibration on the high range is the overriding reason for not using the high range.

Question 7.22 Facility Comment:

Item C in the answer key states "... failure of one or more rods...,"

while the referenced procedure states "... failure of more than one control rod...."

Reference:

IBw0A PRI-2. Entry conditions from this procedure should be used as the correct answer.

NRC Response: Agree. The answer was changed to " failure of more than one control rod to insert fully after a trip."

Question 8.02 Facility Comment:

The observance of expected indications and plant system reactions do not appear to be mutually exclusive events. They in fact are intertwined to the point where it is not believed to be valid to require two separate items to answer this question. Observing plant system reactions is accomplished by observing expectcd indications. Therefore, the answers seem redundant and full credit should be given for having either answer.

NRC Response: The answer key was changed to accept either response "by observing expected indications" or " plant system reactions" for full credit.

9

Question 8.08 Facility Comment:

Partial credit should be given for correct penalty point assignment.

Suggested breakdown is as follows:

DATE TIME (0UT) TIME (IN) POWER (%) PENALTY POINTS CREDIT 10-31-85 0300 0318 85 18 minutes .3 points 10-31-85 1557 1633 65 36 minutes .3 points 11-01-85 0138 0300 45 41 minutes .3 points Time to increase power above 50% is 1614 on 11/1/85 (.6 points).

A band of +1/-0 should also be acceptable since at 1615, one is assured of being within the 60 minute limit.

Reference:

Braidwood Technical Specifications 3.2.1, Surveillance Req. 4.2.1.2.

NRC Response: Disagree. No partial credit will be given. The examinee should be able to calculate the exact time at which power can be raised above 50%.

L 10

MASTER COP _Y U.S. NUCLEAR REGULATORY COMMISSION l

REACTOR OPERATOR LICENSE EXAMINATION

)

FACILITY: @tgidwggd_______________

REACTOR TYPE Wgstinghgusg____________

DATE ADMINISTERED: Match 121_1286________

EXAMINER: Tz_Qt_Reidigggt_________

APPLICANT: ________________________

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%.

% of -

Category  % of Applicant's Category

_Yelue__ Igtel __gcgCg___ __yeiue__

__25_ __25_ _____ _____ l. Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow

__2g_ __25_ 2.

_____ _____ Plant Design Including Safety and Emergency Systems

__2g_ __25_ _____ _____ 3. Instruments and Controls

__25_ __25_ _____ _____ 4. Procedures - Normal, Abnormal, Emergency and Radiological Control

_19Q_ _1gg_ _____ _____ TOTALS Final Grade ________%

All work done on thi s exam i s my own , I have neither given nor received aid.

Applicant's Signature

a e

Section 1 - Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow 1.1 List four sources of hydrogen generation in containment after a loss of coolant accident . (2.0) 1.2 What method is used to verify shutdown margin when the reactor is in Mode 2 with Keff less than one. (1.0) 1.3 What action must be taken if shutdown margin is determined to be inadequate when the reacter is in Mode 1? (1.0) 1.4 The " secondary source" in a reactor refers to the neutron released as a result of radioactive decay of fission products. True/ False (0.5) 1.5 Choose and explain which condition requires higher RCP motor amperage per motor: all four pumps running in hot shutdown or only one pump running in hot shutdown? (1.0) 1.6 The enthalpy change (delta h) across a steam generator decreases as turbine power is increased. True/ False (0.5) 1.7 a. With RCS Tave equal to 290'F, what is the maximum allowable RCS pressure in psig, which does not violate the 320'F differential temperature limitation between the RCS and the pressurizer? (0.75)

b. What steam generator pressure must be established to ensure a 50*F subcooling margin when RCS pressure is 1000 psig? (0.75) 1.8 List the primary reason for the difference in shutdown l margin between Mode 4 and Mode 5. (1.5) l l

1 1.9 What is the temperature of the fluid downstream of a l

partially opened pressuri er PORV if the pressure in '

the PRT is 50 psig and the pressure in the pressurizer is 2000 psig? (0.5) 1.10 a. Doppler coefficients change from BOL to EOL because of three factors. Which factor is most dominant? (1.0)

6. Explain why this causes the doppler coefficient to be less negative over core life. (1.0) 1

"1 . 1 1 a. List the primary factor for causing the total power defect to be more negative at EOL than at BOL. (1.0) b.Give two reasons why this factor causes this negative reactivity at EOL. (1.0) 1.12 A pump experiencing runout conditions will have (high or low) flow, (high or low) discharge pressure and (high or low) motor amps. (Choose correct response) (1.5) 1.13 When the flow rate through a centrifugal charging pump increases, available NPSH (increases or decreases) and required NPSH (increases or decreases) . (Choose correct response) (1.0) 1.14 Explain how xenon is produced and removed at your reactor.

(List 2 methods of each) (2.0) 1.15 A reactor shutdown from 100% power inserts 480 PCM of negative reactivity due to a build up of samarium: Approximately how much reactivity would be inserted after a trip or shutdown from 50% power with equilibrium samarium? (1.0) 1.16 In the event of loss of forced and natural circulation boiling, the core may still be cooled by the " boiling /

reflux" mode of cooling. Describe this mode of heat transfer. (1.5) 1.17 Explain why cold leg breaks result in higher fuel and clad temperatures than hot leg breaks of the same size during a safety injection. (2.0) 1.18 Explain how the reactor can be on natural circulation, subcooled, with a steam bubble in the pressurizer, and still form a bubble in the vessel head. (1.0) 1.19 In a dropped rod recovery procedure with no scram the operator is directed to reduce turbine load. Will this action increase or decrease the departure from nucleate boiling ratio (DNBR)? (Choose one) (0.5) 1.20 What would happen if the reactor operator took no action other than to re-establish correct baron concentration when a xenon oscillation was caused by an inadvertant boron dilution? (1.0)

END OF CATEGORY 1 2

Section 2 - Plant Design Including Safety And Emergency Systems 2.1 List all the "at power" reactor trips automatically blocked as nuclear power decreases below the permissive interlock P-7. (2.0) 2.2 What limits SI pump runout if one of its injection lines had a pipe rupture into containment? (1.0) 2.3 a. To what location would the component cooling surge tank relief valve discharge in the event that RCP "A" had a RCP thermal barrier cooling coil rupture? (1.0)

b. What specific action must be taken in order for the "O" CCW pump to start on bus 141 if 1A CCW pump is OOS? (1.0) 2.4 A caution in the SI Termination Procedure states that

" Alternate water sources for Auxiliary Feedwater pumps will be necessary if condensate Storage Tank level decreases to less than 3%." List the alternate water source for auxiliary feedwater pumps. (1.0) 2.5 List two alternate sources of water for the centrifugal charging pumps during an ECCS actuation. (1.5) 2.6 What primary fission product (s) is/are designed to be filtered out by the containment ventilation charcoal filter units during a loss of coolant accident? (1.0) 2.7 All extraction steam air operated check valves to the heaters close on a turbine trip. What are the two primary reasons for the check valves to close? ()Be specific) (2.0) 2.8 What interlock prevents paralleling 120 VAC instrument bus inverter with the backup regulating transformer? (1.0) 2.9 The RHR system suction relief valve is to protect the RHR discharge lines from inadvertent over pressurization during RCS cooldown.

(a) What specific condition or accident is the RHR suction relief valve designed to protect against? (1.0)

(b) To what location does the RHR suction relief valve discharge? (1.0) 2.10 An RHR pump with its switch in the " pull to lock" position in Mode 1 would be considered operable per technical specification. True/ False. (1.0) 3

4 2.11 With no operator action on a phase A isolation signal, the Reactor Coolant pump seal water return automatically will go to the (Choose one) (1.0)

a. pressurizer relief tank

b. hold up tanks

c. reactor coolant drain tank

d. volume control tank 2.12 (a) What feed system valves are i sol ated when the feedwater water hammer prevention system is actuated? (1.0)

(b) What signal and setpoint will actuate the feedwater water hammer prevention system? (1.5) 2.13 List two reasons for maintaining a minimun pressurizer spray line flow during normal "at power" operations. (1.0) 2.14 When during a LOCA, must ECCS cold leg recirculation be initiated? (0.75) 2.15 (a) At what pressure should the SI accumulator be isolated from the RCS during a normal cooldown? (0.75)

(b) What would be the primary reason to isolate the accumulators? (1.0) 2.16 In the Braidwood procedure (BWFR 5.1) Response to Nuclear Power Generation /ATWAS a caution warns ngt to manually initiate saf ety injection during the ATWAS.

Why wouldn't you as an RO manually initiate a SI during an ATWAS? (1.5) 2.17 What is the primary reason for manually shifting the charging and SI pump suctions during cold leg recirculation to the RHR Heat exchanger discharge? (1.0) 3 2.18 Indicate where in each loop (A,B,C,D) and leg (hot or cold) the following systems penetrations are located in the RCS.

1) PZR surge line

2) RHR suction line (l i st one loop) (.20)

3) CVCS letdown line (.20)

4) Normal charging line (.20)

5) Alternative charging line (.20)

(.20)

END OF CATEGORY 2 l

4 l l

, Section 3 - Instrument And Control 3.1 List two systems / components which utilize auctioneered low tavg. (1.0) 3.2 According to technical specifications, the power range instrumentation is only required in power operations. In spite of this fact, no more than one channel can be ever removed from service during shut down modes. Explain why. (2.0) 3.3 The reactor is in hot shutdown mode and primary pressure is 1850 psig. Mr. Goodwrench wishes to conduct maintenance on the turbine first stage pressure instrument. Explain why you would or wouldn't advise him to do the maintenance. (2.0) 3.4 List 2 signals which must be present to allow automatic start of the containment spray pumps with ESF power available. (1.0) 3.5 Explain why rods in auto do not move when turbine power and nuclear power are being lowered at the same rate. (1.5) 3.6 What three parameters are used to generate the signal for feedwater pump speed? (1.5) 3.7 List three signals which would cause steam generator l

blowdown isolation. (1.5) 3.8 Match the following parameters relating to pressurizer actions / controlling functions (one answer each)

1. All heaters off a. 2185 psig

2. Sprays on full b. 2335 psig

3. Backup heaters on c. 2485 psig

4. Pressurizer PORV lift d. 1885 psig setpoint e. 2310 psig

5. Pressurizer safety f. 17% pzr level valve lift setpoint g. 2210 psig (1.5) i i

5 i

f

i 3.9 Which answer best utilizes the basic principle of operation of the Reactor Vessel Level Indicating System (RVLIS) at Braidwood.

(Choose best answer) (1.0)

a. The detection of a differential temperature between adjacent heated thermocouples.

b. The detection of a differential temperature gradient between adjacent unheated thermocouples.

c. The detection of a differential temperature between adjacent heated and unheated thermocouples,

d. The detection of a differential temperature between the unheated thermocouples and their heated separator tubes.

3.10 The following components, indications and controls is/are located or utilized on the remote shutdown panel (List true or false for each)

1. Auxiliary feedwater pump controls (motor driven only) (0.5)

2. Reactor containment fan coolers switches / controls (0.5)

3. Emergency boration valve and flow indicator (0.5)

4. Instrument air header pressure indications (0.5)

5. RC cold leg N/R temperature indications (0.5) 3.11 What two components are not sequentially loaded onto the 4160V ESF bus during a blackout without an SI signal but are sequenced onto the 4160V ESF bus during a blackout with an SI signal? (1.0) 3.12 If left in automatic control, in what position should PCV-131 (low pressure letdown valve) be found two minutes after a safety injection initiation? (1.0)

' 3.13 List four signals (manual / auto) which will isolate the containment purge system. (2.0) 3.14 A safety injection occurred in mode 5 when a technician

' cttempted to calibrate the white pressurizer channel to 2100 psig after he had tripped the red pressurizer channel b1 stables for testing.

c. List the two probable causes f or the saf ety injection actuation (include setpoints) (.75)

b. Explain how each S.I. signal was generated. -

(.75) 6 -

3.15 Turbine impulse pressure indicator (PT-505) ts ut11 ired in various primary control circuits.and interlocks. List four controls and/or permissives which are fed by impulse channel (PT-505). (List four) (2.0) 3.16 Explain how the failure of a VCT level controller would cause pressurizer level to decrease from its normal band. Include which controller (LT-112 or LT-185) would have to fail and the mode of failure (high or low) needed to cause this problem.

(assume unlimited water supply) (2.0) i I

1 END OF CATEGORY 3 7

i

\

i

Section 4 - Procedures - Normal, Abnormal, Emergency And Radiological Control 4.1 List three conditions which would allow operation outside the technical specifications, procedures or operating orders per administrative procedures. (1.0) 4.2 A caution states that certain interlocks and protective breakers trips would be defeated in the use of Local Emergency Control of Safe Shutdown Equipment procedure. What are the two plant conditions required to use this procedure? (2.0) 4.3 List the two criteria which would require emergency boration during refueling, Mode 6. (1.0) 4.4. In the Reactor Coolant Pump Seal Failure procedure a caution states "If RCP Bearing temperature at any time reaches 225'F then 3 immediate actions must be performed." List the first two immediate actions in specific order! (1.0) 4.5 What two criteria constitute adverse containment condition? (2.0) 4.6 List four parameters contained in the Red Path Summary for BWEP series procedures. (List setpoints) (2.0) 4.7 In the RHR precautions and limitations a statement is made that when the plant is water solid, the low pressure letdown control valve PCV-131 should be placed in manual prior to starting or stopping a RHR pump. Explain why. (1.0) 4.8 a.During the performance of ES-0.4, Natural Circul ation cooldown with steam void in vessel (without RVLIS), a reactor coolant pump (RCP) can be restarted when the pressurizer level is greater than 54% Explain the reason for this level. (1.0) b.If a RCP cannot be restarted in part a, the pressurizer level should be 20-25%. Explain the reason for this level. (1.0) 4.9 If reactor power goes below 10% and the P-10 light illuminates during a reactor startup, what two manual steps have to be rspeated in order for the plant to increase power? (2.0) 4.10 Emergency borate ____ ppm f or each rod not f ully inserted if ____ or more rods are not fully inserted following a reactor trip. (Fill in blanks in the caution statement) )

(1.0)  ;

O i

l l

1

4.11 List five Technical Specifications 1. imitations which if violated required suspension of core fueling operations. (2.0) 4.12 List two responsibilities of the Station Director which may not be delegated to other personnel under GSEP conditions. (2.0) 4.13 In the Nuclear Power Generation /ATWAS procedure the first immedigte_gttige if the reactor did ngt manually trip is that the RO will (Choose one) (1.0)

a. pull to lock both MG set generator / motor breakers

b. depress switch gear manual trip pushbutton for MG set generator breakers {

c. allow control rods to insert to less than 48 spm automatically then manually insert control rods i

l d. depress switchgrear manual trip pushbutton for the reactor trip breakers and reactor trip bypass breakers

[

4.14 When the control room is notified of a plant fire, what fire fighting responsibility is required of the reactor operator per Fire Fighting procedure? (Choose one) (1.0)

a. request relief and then go to scene of fire and assume responsibilities of Fire Chief

b. immediately sound the fire alarm, announce the fire location over PA and radio system

c. notify offsite fire department and security personnel

d. request relief and assume responsibility of a fire brigade member 4.15 Why does the Steam Generator Tube Rupture procedure BWEP-3 require that feed flow be maintained to the ruptured S/G until level is greater than 4*/. narrow range? (1.5) 4.16 BWAP 300-1 directs the reactor operator to continually monitor relevant parameters when a system is withdrawn from operation until conditions "of stable and under control" are reached. List the three " stable and under control conditions." (1.5)

I l

. 1

i

)

)

1 4.17 Routine independent verification of a system line-up may be conducted at the same time temporary alternations are being removed. True/ False (0.5) 4.18 Whose permission (s) is/are required to allow a unit i

NSO to leave his "at-the-controls" area to assist in en emergency at the other unit? (0.5)

)

END OF CATEGOTY 4 END OF TEST 3

4

)

2 I

)

l l

10

~

MASTER COPY Answers - Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer, and Fluid Flow 1.1 1. hydrogen dissolved in the RCS chemistry control pzr gas space (0.5)

2. zirconium - water reaction (fuel clad) (0.5)

3. (oxidation of aluminum or zinc) with h20 (0.5)

4. Radiolytic decomposition of water (0.5)

Ref.: Post Accident Cooling p.6, 7,8 ; Combustibles p41-31 1.2 Verify that the predicted critical control rod position is above the zero power rod insertion limit OR

1. 47 steps on bank C or

2. 162 steps on bank B (1.0)

Ref.: T/S 4.1.1.1.1.C p3/4 1-1 1.3 Emergency borate (.50) greater than 30 gpm until rods are above low low insertion limit (.50) or S/D margin is restored.

Ref.: T/S 3.1.1.1, BWOA-PRI-2 1.4 False Ref.: Nuclear Fuel, p. 14 1.5 one pump running draws higher per motor amperage because of flow resistance encountered in the reactor vessel and piping.

Ref.: Thermo/ Hydraulic Principles, p. 38 Reactor 11 1.6 True.

Ref.: Thermo/ Hydraulic Principles, p. 12 1.7 a. 1647 psig

b. 641 psig (psig +.1 pt)

Ref.: Steam Tables 11

h =

1.8 Primary reason is that a steam line break in Mode 4 at EOL would result in a cooldown and a reactivity transient (.75) but in Mode 5 no significant reactivity transient a

from a steam line break would occur (.75).

Ref.: T/S B 3/4 1-1 1.9 280-3OO'F Ref.: reactor theory 1.10 a. Fuel clad gap reduction is the most dominant effect (1.0) or clad creep.

b. More fuel clad contact due to fuel pellet swelling from neutron irradiation. Effective fuel temperature decreases so doppler coefficient becomes less_ negative gver_cgrg_iiie. (1.0) or less self shielding and less doppler broadening.

Ref.: W Theory, p. I-5.22 1.11 c. Moderator temperature coefficient

b. 1. fission product buildup or greater fuel burnup in core center which causes flux shift

2. baron dilution or boron concentration decreases Ref.: 1-5.27 pI5.10 1.12 high low high i Ref.: Thermo/ Hydraulic Principles, p. 10-43 1.13 decreases increases Ref.: Thermo/ Hydraulic Principles, p. 10-58 1.14 Xe is produced by direct fission (0.5) and as the decay product of the fission product chain with mass number 135 (Te-135 _ I-135 _ Xe-135) (0.5). Xe is removed by B decay (to Cs-135) (0.5) and neutron capture (to form stable Xe-136)

(0.5).

Ref.: Reactor Theory pI-5.69 1.15 240 pcm l l

Ref.: Rx Theory pI-5.78 12

1.16 The ref l u>: mode of core cooling steam produced in the core fills the hot leg and is condensed in the hot leg side of the S/G U-tubes (0.75). The condensate then drains along the bottom of the hot leg back to the core (0.75).

Ref.: Thermo/ Hydraulic Principles ,MCD p2-18 1.17 Because of the time delay of draining the loop seal to Establish a steam vent path (1.0) (more mass loss of inventory) cafety injection flow is less than break flow (0.5) resulting in a partial drain of the reactor core (0.50):

(reduced heat transfer causing higher fuel and clad temperatures)

Ref.: MCD, pgs. 2-19 1.18 If cooldown and depressurization is performed too rapidly (0.25),the bisb_tgeagtetute_gggleOt_iO_tbg unget_beed_wbteb_ bas _e_ slight _emeutt_91_E991109_fl9dz ceO_flesb_tg_stgem (0.75) when indicated loop temperatures are well below saturation.

Ref.: BWEP-ES-0.2

'1.19 increase Ref.: Thermo/ Hydraulic Principles 1.20 Oscillations would be convergent (self dampening) or stop.

Ref: reactor theory I-5.57 END OF SECTION 1 13

Answers - Plant Design Including Safety and Emergency Systems 2.1 1. Pressurizer high water level

2. Pressurizer low pressure

3. Low reactor coolant flow

4. Reactor coolant pump bus undervoltage or underfrequency

5. Reactor coolant pump breaker open and turbine trip Ref.: PLS, p. 4 (.40 pts each) 2.2 Branch orifice Ref.: ECCS/RH, p. 20 2.3 a. Auxiliary Building equipment drains or chromated drain tank

b. 1.The CC pump (1 A) powered f rom that bus must have its control switch in pull to lock.

2.0-CCW pmp bkr-rack:1 in 141 l 3.0-CCW pmp bkr on 383' racked in 4.0-CCW pmp C/S-after close Ref.: CCW System, p. 6 ,p17 2.4 ESW emergency supply Ref.: Figure 26-1 (Lesson Plans) 2.5 Rx make-up water VCT RWST RHR discharge (from containment sump)

Any 2 Ref.: BWFR C-1, p. 5 2.6 iodine OR bromine (halogens)

Ref.: Containment Ventilation, p. 12 2.7 1) Prevent overspeeding (0.5) of the main turbine after a turbine trip ,

by preventing steam trapped in the feedwater heaters from returning l to the turbine through the extraction steam lines (0.5).

2) Prevent water from entering the main turbine (.50) through l the extraction steam line if the f eedwater heaters are flooded (.50)  ;

I Ref.: Extraction Steam, p. 36-33 2.8 Mechanical bar interlock will not permit closing both breakers together.

Ref.: AC Electricalo @6 66/81

2.9 c. RHR valve sized to remove capacity of 1 CCP operating at cooldown start or 2 CCP's under cold shutdown conditions

b. Relieves to Hold Up Tanks Ref.: RHR System, pgs. 12, 13 2.10 False Ref.: BWAP 300-1, p. 15 2.11 PRT Ref.: M-64 2.12 a. Main Feedwater isolation valve - FWOO9 Preheater bypass valve FWO39 Temporary line isolation bypass valve - FW35 Feedwater isolation bypass valve - FW43 (.25 each)

b. S/G pressure < 600 psig (.75 each) or <5% s/g level narrow range Ref.: BWGP 100-5 , chap 25 p62 2.13 1. reduce thermal stress to spray line and spray nozzle

2. maintain pzr chemistry uniform with RCS

3. prevent excess cooling to spray system (any two)

Ref.: PZR 14-20 2.14 RWST level decreases to less than 45% or 46.7%

Ref.: BWFR C.3, p. 2 2.15 c. Before pressure decreases to between 1000psig and 800 psig

b. Prevent discharge of accumulators or injecting nitrogen into RCS Ref.: BWFR, p. 1 2.16 This would result in the possible loss of a heat sink (steam generators)

(.75) due to f eedwater isolation while the reactor is at power (.75).

Ref.: BWFR 5.1, p. 2 2.17 This sequence prevents sending contaminated SI pump-recirculation flow to the RWST (1.0).or verify proper alignment for pump suctions and discharge Ref.: ECCS/RH, p. 16 2.18 1.D-hot leg,2.A-hot leg,3.C-cold leg,4.B-cold leg,5.A-cold leg (C-hot leg)

Ref:RCS p12-41 M8LfolLK@JUYoTd_3

i Answers - Instruments and Controls 3.1 Turbine loading stop orc-16 Ref.: PLS, p. 34 3.2 If more than one PR-NI channel is de-energized,P-10 will block source

range high voltage causing a loss of source range (1.5). Source range NI is required in shut down modes per T/S (.50) 4 Ref.

T/S 3/4 p.3-1 3.3 Wouldn't allow him. Low pressurizer pressure trip is 1885 psig. If the turbine first stage pressure instrument is placed in test for maintenance, it will affect interlock P-7,this will unlock the low pressurizer pressure trip function and result in reactor protection Ref: RPS 60b-19 3.4 1. Containment pressure Hi-3 or 20 psig 2.ESF sequencer permissive signal present or diesel sequencer timing 3.C/S for cont. spray pmp -after close/after trip position

4. educator valves-fully stroked open Ref.: Containment Sprays, p. 12 3.5 The rate comparator looks at the rate of change between turbine power and auct high nuclear power (0.5). If both parameters are changing at the same rate (0.5), even though one may be higher than normal there will be no error output to move rods (0.5).

Ref.: Rod Control, p. 45 3.6 1. Steam header pressure

2. Feedwater pump header pressure 1

3. Total steam flow Ref.: Feed / Condensate, p. 28 (.50 pts each) 3.7 SI manual / auto containment phase a signal High temperature in S/G blowdown condenser room Ref.: S/G, p. 30 i

16

3.0 1. f 2. e 3. g 4. b 5. c

(.3 pts ea)

Ref.: Pzr, p. 36 3.9 c.

Ref.: RVLIS, p. 3 3.10 1. False

2. True

3. True

4. True

5. False Ref.: Remote Shutdown Panel, p. 11 3.11 SI pump RHR pump Ref.: ESFAS, p. 23 3.12 chut R2f: ECCS 15a-19 3.13 a. Containment atmos. area radiation high alarm b.SI auto

c. Phase A isolation (manual)

d. SI manual

o. Containment spray actuation (manual)

Ref.: Containment Purge System, p. 23 (.50 pts each) 3.14 a.1. low pressurizer pressure SI < 1829 psig

2. low stm line pressure SI <640 psig b.1. At >1930 psig(.10)-SI permissive block (.15) is now removed by the white channel greater than 1930 psig;the red channel was tripped so SI was initiated as steam pressure is less than 640 psi g (.50) .

2. At >1930 prig (.10)-SI initated as pressurizer pressure is less than SI setpoint(.50),(SI permissive unblocked) (.15)

Ref RPS p60b-20,61-34 3.15 Rod Control Steam dumps (Tref)

Feeds P-13 circuit C-5 circuit C-16 circuit s/g wics Ref.: Main Steam and Main Turbine, p. 42 17

, 3.16 LT-112 fail high (.20)

-full letdown divert (.20)

! -no makeup (.20)

) -VCT will slowly empty (.20)

-charging flow will cease,as VCT will be dry (.10)

-p r level decreases due to normal letdown,and no makeup (.10) a Ref.: CVCS p15a-30 l

i l

i i

3 END OF SECTION 3 1

e

]

1 i

18 i

[

Answers - Procedures - Normal, Abnormal, Emergency and Radiological Control 4.1 1. injury to the public or company personnel

2. releases off-site above technical specification limits

3. damage to equipment Ref.: BWAP 300-1,p13 4.2 This procedure is to be used only in an emergency when the cystem must operate for safe plant control or shutdown (1.0) and only if the redundant system is also unavailable (1.0).

Ref.: 1BWOA-ELEC-5, p. 2

, 4.3 < .95 Keff

< 2000 ppm boron concentration Ref.: Emergency Boration 1BWOA-PRI-2 4.4 Trip the reactor Trip the affected pump Ref.: BWOA RCP-1 4 4.5 1. Containment pressure > 5 psig

2. containment radiation greater than 10 (*5) R/hr.

Ref.: BWEP-0, p. 3

> 4. 6 1. Subcriticality - Nuclear pwr greater than 5%

i 2. Core cooling - core exit T/C > 12OO'F

3. Heat sink - all S/G < 4% NR

4. Total feedwater f l ow < 500 gpm available

5. Integrity - decrease > 100'F/60 min.

6. RC cold leg < 246'F

7. Containment - containment pressure > 50 psig (Any four)

Ref.: BWEP-1 l

19 l

4.7 PCV uses a P1D controller, the differential portion sees the etarting/ stopping gtgssutg_sutge_as_a_majgc_ttansiget (0.5) and gee 0Zelgse_eCv:131_tgg_fet_tg_stge_ceusieg_e_gtessute dtgeliocteese (0.5).

Ref.: RHR, p. 33 4.8 a.As RCP is started,the steam bubble will collapse and the pressurizer level will decrease rapidly to fill the void in RX vessel b.If the RCP cannot be started, a rapid cooldown will make the void larger, displacing water into the RCS causing an insurge into the pressurizer.

Ref.: BWEP ES-O-4, WESTINGHOUSE OWNERS GROUP p14,17 4.9 1. Block IR reactor trip by IR block switches

2. Block both power range low setpoint reactor trips using PWR RNG block switches Ref.: BWGP 100-3, p. 8 4.10 1. 100 ppm

2. 2 Ref.: BWGP 100-5, p. 7 4.11 Suspend core alterations if:

a. Less than one boron injection flow path (T.S. LATER).

b. With no charging pump operable (T.S. LATER).

c. Less than one borated water source (T.S. LATER).

d. Less than two nuclear monitoring channels are in service (T.S.

LATER).

e. Direct Communication is lost between control room and personnel in containment (T.S. LATER).

f. Less than one RH Loop in operation (T.S. LATER).

g. Less than 23 feet of water of vessel flange (T.S. LATER).

h. Activation of the containment evacuation alarm.

l i. Less than two nuclear monitoring channels have counting rates greater than or equal to 2 count per second.

j. Less 2 RHR loops operabir (Any five)

Ref.: BWGP 100-6, p. 2 4.12 1. Declaration of the emergency condition

2. Decision to notify and recommend protective actions to offsite authorities.

Ref.: BWZP 100-1 19

.g .a 4.13 c.

Ref.: BWFR S.1, p. 2 4.14 b. Sound the fire alarm and announce fire location over the PA and plant radio system.

Ref.: BWZP 300-7, p. 1 4.15 Promote thermal stratification to prevent the. ruptured stm generator depressurization OR To prevent the condensing of the steam on the S/G tubes in the steam generator as the RCS is cooled which would decrease S/G pressure resulting in RCS increased leakage to the steam generator..

Ref.: WOG p58-E-3 l

4.16 stable radiation levels, pressure, temperature in containment (0.5 ea) or if adequate core cools.;

stable RCS pressure, temperature and levels (0.5 ea)

Ref.: BWAP 300-1, p. 13 4.17 False Ref.: BWAP 300-1, p. 12 4.18 SCRE/ Control Room supervisor Ref.: BWAP 300-1, p. Q ,

1 f

• -END OF SECTION 4 I

.d-"

l k

e 20

~

J s r

^

,* . as / ),

,l l'

J i

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BRAIDWOOD 1 REACTOR TYPE: PWR-NEC4 DATE ADMINISTERED: 86/03/12 EXAMINER: T BURDICK APPLICANT: _________________________

INSTRUCTIONS TO APPLICANT:

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

Stople 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%. E:: amination Papers will be picked up six (6) hours after the examination starts.

% OF 3ATECORY  % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 0_______ ______ ___________ ________ ___________________________________

25.00 25 0 THEORY OF NUCLEAR POWER PLANT

________ ___I__0 ___________ ________ 5.

OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 PLANT SYSTEMS DESIGN, CONTROL,

________ ______ ___________ ________ 6.

AND INSTRUMENTATION 25*00 25*00 PROCEDURES - NORMAL, ABNORMAL,

________ 7.

EMERGENCY AND RADIOLOGICAL CONTROL 25 00 25.00 ADMINISTRATIVE PROCEDURES,

________ ______ ___________ ________ 8.

CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE _________________% '

All work done on this examination is my own. I have neither givGn nor received sid.

~~~~~~~~~~~~~~

APPLEC iI5'5EUU SURE

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDE, AND PAGE :

00ESTION 5.01 (1.00)

Placing an RTD loop isolation valve on its backseat during cold plant conditions may result in separation of the valve disc from its stesi after plcnt heatup. Briefly state why this happens.

QUESTIO.N 5.02 (2.00)

A large four loop plant similar to Braidwood experienced a momentary increase in turbine generator output (110 MWe) when feed water and conden-cate flow was reduced due to a cavitating condensate pump. At the same tis,e reactor power was constant. A few seconds later the unit tripped on low SG lovel.

a.

How did reduced feedwater end condensate flow affect the feedwater heater shell pressure and temperature conditioris?

b. What affect did reduced feedwater and condensate flow have on extractior, s t e e n, flow to the feedwater heaters?

c. Why didn't increased turbine power cause an increase in reactor power?

OUESTION 5.03 (1.50)

List the four parameters affecting Departure from Nucleate Boiling (DNB) and state whether the probability of approaching DNB is increased or decreased as these paraseter values increase.

QUESTION 5.04 (1.50)

Explain rod worth variation * [ consider only one factor at a time]

a. with its radial position in the core.

b. if another rod is inserted adjacent to it.

c. if the niodera tor teniper a tur e changes.

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

?

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 00ESTION 5.05 (1.75)

Freni the statements below, choose the most correct words or phrases (in parentheses) and write them on your answer sheet:

e. The ratio of peak to average value of power distribu n is known as (peaking factor, ther nial limit). (0.25)

L b. The value of the Heat Flux Hot Channel factor will (increase, rea.ain constant ) when power is decreased from 100 to decreasyt,d 50%. Lu (0.25)

c. The Heat Flux Hot Channel factor li m LL ( assur es , does not assure) that DNB will be prevented during nornial operation. (0.25)

d. Criculation of Enthalpy Rise Hot Channel factor assumes core power 2s (unifoini, not uniform) and flow through e hsnnel is (the same, different) throughout the core. (0.5) e . 2" ' ' ; ;p;- Irm ' .cl.e of c.ntosipy hiu Hv; C;..n ici fe;t- -

Ou -lume aw o s. i w n E. 6 v . . d i t i o r. ; _ ( S S )

GUESTION 5.06 (1.50)

a. How will differential boron worth vary with:

1. Increasing RCS teniperature ? (0.50)

2. Increasing Concentration ? (0.50)

3. Core age ? (0.50)

c. How do the tal power efect hange with vel b rnup ? .25)

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 GUESTION 5.07 (2.75)

A clean core is started up and taken to 50 % power, where it re-cains for 30 days:

e. Describe the reactivity changes the operator must compensate for due to fission product poisons. Indicate approximate times. (1 5)

b. After 30 days power is increased to 100%. Explain any further reactivity changes required. No times required.

(1.25)

No reactivity values are required.

QUESTION 5.08 (1.00)

Con.plete the following table by supplying the missing wor ds or phrases. Indicate direction, magnitvoe, and rate where applicable.

Rx. Per iod Rx. Power Response Start Up Rate

c. Seia11 positive Rapid increase ( )

b. ( ) Constant ( )

c. Large negative ( ) Small negative 00ESTION 5.09 (2.50)

a. For an operator taking data for a 1/M plot, how will the Shut-down margin (SDM) affect the time elapsed before a stable count rate can be obtained after withdrawing rods ? (0.75)

b. How will the initial count rate affect the count rate at crit-icality ? (0.75)

c. If the speed of the control rods were to somehow increase. What would be the effect be on:

1. Rod height at criticality ? (0.5)

2. Count rate at criticality ? (0.5)

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

5. THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AND PAGE 5 DUESTION 5.10 (2.25)

o. How and why will the magnitude of the Fuel Tesiperature Coeffi-cient (FTC) change as fuel temperature changes ? (0.75)

b. Explain the effect on the magnitude of the FTC due to :

1. Core power (0.75)

2. Core burnup (0.75)

GUESTION 5.11 (3.25)

Unit 1 is at 40% power when a single Reactor Coolant Pump (RCP) stops. Assume rod control in manual and no automatic or oper-Etor action occurs.

a. Desciibe what happens to the teniperature of the affected loop and explain. (0.75)

b. Describe what happens to the delta T across all steam generat-ors and the vessel or core. Explain. (1.0)

c. Describe what happens to unaffected steam generator pressure and explain. (0.75)

d. Explain the effect of stopping the RCP on individual loop and total coolant flow. (0.75)

GUESTION 5.12 ( .50)

Reactor power decreases from 1 watt to 0.1 watt in three minutes. The Startup Rate is____ DPM.

Choose the correct answer.

A.-3, E: . - 1, C.-1/4, D.-1/3, (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE *****)

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 DUESTION 5 13 (1.50)

What are the three bases behind the Control Rod Insertion Limits.

QUESTION 5.14 (2.00)

With reactor power remaining constant at 50% power, what value of Tave would cause the steam generator safety valve with the lowest lift setting to open? Assume that the normal steam ganarator pressure at 50% power is 1050 psig. Show your work!

(xxxxx END OF CATEGORY 05 xxxxx)

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 GUESTION 6.01 (1.00)

c. Which reactor coolant pumps are aligned to the SAT power supply?

b. Why are they aligned to the SAT?

GUESTION 6.02 (2.00)

Why will a reactor coolant pump bus degraded voltage condition cause a RCP trip?

QUESTION 6.03 (1.00)

For closure of a diesel generator output breaker there must be no over-current trip on the SAT supply breaker to the bus. Why?

OUESTION 6.04 (1.50)

Using the E:r aidwood electrical bus non.enclature, state the bus chstatter-istics including: voltage /AC or DC/ safety or nonsafety related.

a. 159

b. 144

c. 112

d. 123 OUESTION 6.05 (1.50)

The 125 vde battery charger trouble alarm on 1PM01J has four inputs. Name three.

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

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTA1 ION PAGE 8 OUESTION 6.06 ( .50)

The diesel generator fuel oil day tank must contain enough fuel to assure fud load operation for _______. [ choose the one correct answer]

a. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

b. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

c. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

d. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> GUESTION 6.07 (1.00)

How does the diesel generator overspeed trip ensure engine shutdown?

DUESTION 6.08 (1.00)

After the eniersency diesel gener ator has been stopped from the contr ol r oon-how does the operator verify froci the control room thtt the diesel has shutdown after its cooldown is complete?

DUESTION 6.09 (1.00)

The letdown divert valve can divert excess water to the holdup tanks either automatically or by operator manual. How does manual differ from automatic?

DUESTION 6.10 (1.00)

Wh:t is the purpose of the charging header pressure control valve, HCV-182?

OUESTION 6.11 (1.00)

What are the following cheniicals used in the reactor coolant system for?

a. hydra ine

b. hydrogen pero::ide

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

1

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 6.

GUESTION 6 12 (1 00)

Briefly describe the four possible boration paths via the CVCS.

QUESTION 6.13 (1.00)

State the design load follow program basis for the BTRS.

QUESTION 6.14 (1.50)

a. The suction valves from the RHST to the RHR pumps do not receive an SI signal. Why?

b. What is the normal position for RH 8716 A and B, crosstie valves, durin3 nors.a1 plant operations?

OUESTION 6.15 (1.00)

a. State the range of the SG PORV setpoint adjustn:ent.

b. State the total capacity for all four SG PORV's.

QUESTION 6.16 (1.50)

a. If the steam dump system is at 50 % capactity what is the flowrate in pounds-mass per hour? [.5]

b. How do the turbine trip and loss-of-load controllers differ in response to temperature error? E1.03 (war *> CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6. PLANT SYSTEMS DESIGH, CONTROL, AND INSTRUMENTATION PAGE 10 GUESTION 6.17 (1.50)

A plant similai to Braidwood was on RHR at 185 degrees and 350 psis with one RCP running. The operators were restoring RHR train A from a pump operability test. Upon opening crossover valve 8716 B the operator noticed B RHR flow go to zero and turned the pump off. 15 seconds later he noticed zero pressure in the RCS and turned off the RCP. It was later determined that no RHR reliefs had lifted and no leaks had occured. The problem was attributed to a procedure error in valving sequence.

o. Why did opening 8716 B cause a loss of RHR flow indication?

b. Why did the RCS depressurire immediatly afterwards?

DUESTION 6.18 (2.00)

2. When 1A AFW pump stsits automatictlly why does the auniliary 1pbe oil pump have to te locally started?

b. What do the frW train B flow lights signify?

c. How would the operator know that the AFW flow control valves are set too low when the system is in standby?

OUESTION 6.19 (1.50)

Byron station experienced personnel exposure due to the incore detector system.

a. What three component parts of the incore detector system are a source of radiation?

b. When do these parts become a radiation threat to plant personnel?

OUESTION 6.20 (1.50)

a. What load is not stripped from the 4160 VAC ESF buses during a station blackout?

b. How much time delay will there be until containment spray starts if it is required 30 seconds after SI actuates? Assume the DG is required to power the bus.

(***** END OF CATEGORY 06 *****)

7. PROCEDURES - NORMAL, AE:N O R M AL , EMERGENCY AND PAGE il

~~~~R5656E655EIE~56UTR6E-~~~~~~~~~~~~~~~~~~~~~~~

QUESTION 7.01 (2.50)

When the control room is inaccessible *

o. Where do the SE, SCRE and NSO report to?

b. If some or all of the remote shutdown panel indications are not functioning where can alternate indications be obtained?

c. Where may parameters, not displayed on the RSP, be viewed?

DUESTION 7.02 (1.00)

When performing a cooldown with the CR inaccessible *

a. Where is RCS wide range pressure indication obtained from?

b. Is it pr ef er able to continue cold shutdown using the stean. generators or transfer to the RH system?

DUESTION 7.03 (1.00)

The stuck or misaligned rod procedure establishes two criteria for alignment of a bank to its rod rather than the rod to its bank as usual.

What are those two criteria?

OUESTION 7.04 (1.50)

What are the three options for post-SGTR cooldown?

OUESTION 7.05 (1.00)

In the procedure for response to imminent pressurized thermal shock condition *

a. What is the alternative action if the accuniulators cannot be isolated per the procedure?

b. What is the cr iterion for de t e r niining if an RCS temperature soak is required?

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

7. PROCEDURES - NORMAL, AE:N O R M A L , EMERGENCY AND PAGE 12

~~ ~ ~Rd656L6556dE 66U5 RUE ~ ~~~~ ~ ~~~ ~ ~ ~ ~ ~~~~~~~ ~ ~~ ~

QUESTION 7.06 (2.00)

In the reactor trip or safety injection procedure:

o. What three conditions determines whether main steam lines should be isolated per the immediate actions?

b. How is adverse containment defined?

OUESTION 7.07 ( .50)

The Nuclear Accident Reporting System phone color is:

Echoose the one correct answer]

a. blue

b. yellow

c. steen

d. red GUESTION 7.08 ( .50)

Once an emersency classification has been siade by the Station Director:

Echoose the one correct answer 3

a. The NRC must be notified within four hours.

b. State and local authorities must be notified within 15 minutes,

c. The Shift Engineer relinquishes his responsibility for plant safety to the Station Director.

d. If NRC region III is not contacted via the ENS then they must be called by coms:ercial telephone from the site.

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

7. PROCEDURES - NORMAL, A E:N O R M AL , EMERGENCY AND PAGE 13 R

~~~~ 5656E655CdE'_EUUYR6E~~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 7.09 ( .50)

For an emergency on-going communications will be established between the:

Echoose the one correct answer]

c. control room and NRC

b. control room and EOF

c. control room, and CCC

d. control room and TSC QUESTION 7.10 ( .50)

The designated Station Director niey assume responsibility from the Shif t Engineer during an emergency by: Echoose the one correct answer]

a. stating he has taken charge in the Station Director los and signing his name.

b. conducting a turnover face-to-face and then signing the turnover sheet.

c. telephoning the control room from the TSC to inform the SE he has been relieved.

d. conducting a face-to-face turnover with the SE then announcing the transfer of r esponsibility on the plant page system.

GUESTION 7.11 ( .50)

The NARS system: Echoose the one correct answer 3

a. uses separate codes for each of the four emergency classifications.

b. uses a special code for general emergencies and another for all other classifications.

c. of f ers ready a ccess to local and state agencies for joint consultation regarding public evacuation.

d. will be used as a continuous communication link with local and state agencies in emergencies.

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

7. PROCEDURES - NORMAL, ADNDRMAL, EMERGENCY AND PAGE 14

--- EE5i6E 5iCEE 55RTRUE~~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 7.12 ( .50)

Onco communication is established with the NRC on the ENS during an G2ergency it siust be maintained until notified otherwise by the :

[ choose the one correct answer]

o. CCC

b. EDF

c. NRC

d. TSC QUESTION 7.13 ( .50)

The GSEP radio link in the control room persits communications with all other emer gency c enter s e:: cept the: Echoose the one correct answer]

s. EOF

b. TSC

c. CCC

d. OSC QUESTION 7.14 ( .50)

The GSEP siicrowave line is a gray phone which, when picked up, automati-cally initiates a call to the: Echoose the one correct answer]

a. TSC

b. CCC

c. EOF

d. OSC OUESTION 7.15 (2.00)  !

Individuals working in controlled areas must obtain information from the l radeon group prict to entering the controlled area. Nanie four of the five specific inf or s.at ion iten s .

1 DUESTION 7.16 (1.00) ggg The approval of the Rad Chem foreman or lead HP s.ust be obtained prioir to working in steas where whole body dose equivalent rates are greater than

______________. Efill in the blank 3 (marwx CATEGORY 07 CONTINUED ON NEXT PAGE *****)

4

7. PROCEDURLS - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15

- ~~E 65UL655EEE'_6UUT E L""~"~""~"~

QUESTION 7.17 (1.00)

When is the word ' CAUTION' used in regard to radiological signs and labels?

QUESTION 7.18 g(,J.00)

o. How long are regular RWP's valid?

b. When are they required?

DUESTION 7.19 (2.00)

a. Where does the insertion of a BPRA into a fuel assembly take place?

b. Where does the insertion of a thimble plus into a fuel assembly take place?

c. Wher e does the insertion of a RCCA into a fuel assembly take place?

d. What four verifications must be made prior to unlatching a fuel assembly in the core?

QUESTION 7.20 (1.00)

The refueling machine procedure cautions the operator NOT to operate with the Dillon load cell on high ran3e except in emergency or test conditions.

Why?

GUESTION 7.21 (1.00)

For a loss of offsite power in modes 3 or 4 the procedure cautions that if SG PORV's are being used then a cooldown should begin within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Why?

GUFSTION 7.22 (2.00)

There are six conditions requiring emergency boration. Name five of them per the procedure. l I

(***xx END OF CATEGORY 07 *****)

l l

1 l

l l

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16 GUESTION 8.01 (1.00)

How is notification of the OAS or duty OE documented?

DUESTION 8.02 (1.00)

How does the reactor operator verify local actions affecting plant operations according to his instructions whenever possible?

OUESTION 8.03 (2.00)

Name four of the five occasions requiring an independent verification of proper safety system lineup.

00ESTION 8.04 (1.50)

Whrt are the thr e e acceptable means of performing independent system lineup verification?

DUESTION 8.05 (1.00)

How are non-station personnel authorized to request the removal of plant equipment froni service? Include the authorizer's title.

QUESTION 8.06 (1.50)

For ready identification the binders for BWEP's, BWFR's and BWOA's are color coded ________________, ________________

and ___________________.

Cfill in the blanks]

OUESTION 8.07 (1.50)

Procedures have low level steps preceeded by letters, open bullets or closed bullets. State the meaning of each.

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

1 1

E. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 00ESTION 8.08 (1.50)

Assume that it is 0300 on 11-01-85 and the reactor is presently at 45%.

Considering the Delta-I penalty history listed below, when will you be ellowed to increase power above 50%?

DATE TIME (OUT) TIME (IN) POWER (%)

10-31-85 0300 0318 85 10-31-85 1557 1633 65 11-01-85 0138 0300 45 OUESTION 8.09 (1.00)

How often must the CSF status trees be scanned during an emergency?

DUESTION 8.10 (1.00)

What documents serve as the permanent record for the execution of general operating procedures?

OUESTION 8.11 (1.00)

The mechanical and electrical lineup forms have a double line at the bottom of each page. Why?

OUESTION 8 12 (1.00)

a. When a manual valve position is checked should it be checked in the OPEN or CLOSED direction?

b. Is a valve found on its backseat considered an important inadequacy that would require resolution before the lineup is acceptable for operation?

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 QUESTION 8.13 (1.50)

c. When taking heat traced systems DOS who is responsible for removing the heat tracitig from service?

b. How is equipment identified on 00S forms when an ESD input is required?

DUESTION 8.14 (1.50)

State the meaning of the single, double, and triple asterisk codes used on the daily equipnent logsheets preceeding certain parameters.

QUESTION 8.15 (1.50)

a. At what minitum power level is the green board concept expected to reptesent a notmal configuretion?

b. State the meaning of blue colored lights.

c. State the neaning of amber colored lights QUESTION 8.16 (2.50)

There are five general criteria that must be satisfied to establish containment integrity. Name them.

QUESTION 8.17 (1.00)

Surveillance frequencies are designated by code. Match the code letter with its appropriate frequency.

a. prior to each release 1. S

b. at least once every 18 months 2. P

c. at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3. O

d. at least once every 184 days 4. R

5. SA

6. T (arrar CATEGORY 08 CONTINUED ON NEXT PAGE *****)

1

1 I

I B. ADhINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 QUESTION 8.18 (2.00)

Tha minimum RCS temperature for criticality is based upon five criteria.

Name four of them.

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

(m*****x***xxx END OF EXAMINATION ****xx*ummmmmmm) i i

I._ ._

l es Table 1. Saturated Sleem: Tesaperalere Table  ;

aos Pren Serc lic wume Estaanpr le=0 Lt per Set

,. fent 50 en Lesee Esse Set Water Set besid fuse taper Set Set Est'ess

• Ret leme I tiend Imp 400s, fone i, ,

e es *s '8 he bee '#a e, aq, e, e ma- e 08859 0 0160:7 mes ss0s 7 .esl7s tens le7)S as e 0 0e60e 804C21 mais an61 s isot tett tiers us73 as-

. ~

g6 010 M S* ' 004C?C -N35 0 atM 0 amit 107 4 Ser64 essi 3.17u ueC?-- yr -

aul t il249 30712 14772 Set! 2 MSI 11732 38 004019 alm I 36MJ ES18 1972 1 8074 1 4 8127 ElW1 18463 38

-s

-<a*

., et 017143 9016c!t 2453 3edl e 0 077 3071 8 1979 0 SATU 2 5437 Ilsea e9 a

c8 e 13143 0 016019 2777 a 2777 8 le 635 19h9 8 1879 9 esa 084197 804019 2117 8 402C7 2 117S 2 IS27 52 et t 2117 8 82 0sl SE47 5307 0 0747 2 1787 2 HS9 esa SlHla 0040?0 9966 7 39(1 7 H es? BE76 tell S :287 Illil 21M3 at 40 8 9 86514 SS4C73 183C 0 8430 0 3621 3h64 Im23 8 6178 25m 1 8327 e8 90 I e 17796 8 Ol6c?3 IPGs 3 SI916S 370s 8 le OSa SES3 3m34 0 0361 1 0001 2IM7 mB

• 87 9 804074 3589 2 3605 7 30 057 30W 2 9942 0080C 2 0798 2 1197 074 WS S F06?S 9040M M8:4 Mai o hhj 27 058 Dell DNil 80s!T 2EM f ilM esa -Y",--

95 0 8 77183 GCIW8 13136 13tJ 6 2e 059 30Li t 5360 0 0s 74 2tH3 f lC*C Ma EJ3M3 GDAJ1 129JJ '3F1'1 M 50 SECA IM63 35 5

, 8054 2.8 51 188G MA 00 0 0 2M11 0016c33 17C? 6 62 8 C:74St S C403t I?c7 6 N OEC IE97 lee? ? 0 0'15 2BMI 2 hat 50 1175 7 1879 7

• es e 02m? 004C35 109t S 3C OSS letS leas t G M13 2 0?ll2574 t esti

• es . *

. . He $344 004043 set C 205t $ 32 058 1274 lett $ 0 0f 3? 20lE ass. _ . , -

- *f 84 8 SM68 b 90 4046 tri l 38 0M 50MJ 30K 4 00E 7C 2 &%t 2 076s *St ^

926 5 SM S 5 0be 1863 2 19912 8 070E . I M9t IS70s RI te s eJE?$? 0 016050 fi s s uut etnesa 368 3

$14 3 168 4 2 057 IEst 10973 0 0745 I Mit 2 054! 20 314 3 4C 0al 3052 9 IM3C C C783 1 96'A la s 8 4 tS! 0 040tt 764 i 2 051' 77 8 PW I 47 0st 10518 10538 1970f 2Dt?t 3s s e sas?C C C!E A 3  ?!? 4 717 a 0 0E ?. ha Me 8 474il 0C400 673 3 48 Od3 46 04; lot! 7 IM47 0 Of h6 1 tila 2 08'; la e 673 9 8086 5 IMS6 0 R 9' 42L7C 2 0sli 28 me 8 SD6t? O C40'? 633 3 63!3 82 8 C Saos! C 04C*7 $911 48 CU 10s44 leM a O M3; I 947t 2 0*?$ 58 St. S SC CD los? 3 es e E S 7; C 0:6X: $6C 3  %:3 105' 3 0 tM5 Il3D 2 C3:1 Ct On e 0 6:114 OCit d' 2211 577 5 57 0?S lost i 10512 0In I 9747 2 G744 es e es e 8 61551 0 016053 Sa t?L 10sS C IMtc 0 30a3 3 9151 2 0:ll 58 496 8 49t 4 M C;2 les3 9 10559 0 8C75 1 9060 2.0135 RA me O f9813 00 4099 463 1 ct 4 74!!? C Cillet aal l 46 ? ) SA C!! 104:7 11008 0111$ IS97C 280E8 58 4:13 GC Cls ICC 6 IIC16 0 117 ISM: 2ac33 53 se e 6 79X7 0 C16111 816 3 416 3 W CIC 96 8 O&C7 OC A ll ? less 5 1807 S O Its 137?? I3te? Os e 392 9 M25 64 0td 3039 3 IIC3 3 0 274 1870s IS9?t te es s $ 25316 0 016123 3709 370 5 u 0C3 Ic34 7 1104 2 0 76C 33617 19676 ha Ste t 098974 0 016130 SSC 4 lit o ts7 : $7999 M371 1105 3 01795 IS$: I tt'i l eCf 6 0 016137 331 3 331 1 M 995 Sc3S t 1105 9 me les 8 ten s 3 et 9(i 0 016144 all I till il M7 leM l 1106 8 81131 Sanaa 13775 Wa llH7 0 016151 796 14 2 % 38 73 95 9 1864 3 8358 33775 asI les t 12C30 0 0161S8 25028 2803; 3033 4 11076 0 14;7 13773 I M 'S Wa 75 96 IC32 $ IlW S 91437 13188 13626 Wa its 12750 00364% MS 37 MSM 112 8 IJS0t Mll St314 1109 3 81477 IJ10$ 19677 3 014:73 26337 251 38 75 90 IC3; 7 1310 7 tit a tie s l e?M 0 016180 234 71 234 7; GISC7 BAC21 1357s ttia S1 97 SC?t i 1111 0 lin g 35333 0 C16148 275 && 27585 SIW2 3 753t 1344C 344A gla s 3 6009 0041% 2142C fjell 83 9* 3077 9 till t 0 3577 1 70s4 89433 tes t 05 87 10M t 1112 5 E3611 4 7774 19 56 fata $.4.*.

@1?? 170 0 1 6977 0 016708 303 2S 2C3M 07 97 80756 1113 6 122 0 1 7891 0 016713 192 ta 167 95 MM 9874 5 line s 81Hf 1 493 13D9 115 8 tie s 1 990: 8 016771 18323 14324 816ac 17633 .lS?t3 W8 175 0 I MS9 0 016729 174 06 174 0$

91 % 3073 3 IllSJ G litt 11533 137a7 nos la s 93 96 16:2 2 11M 1 f l068 004234 I6$45 16Sa7 EM Olkt 1 9453 IS7C? M8 1071 0 1887 0 G l?tJ IJ374 13157 M8

• las s 2 2730 G en?e? 35737 157J3 97 tE 8019 8 1137 8 137 9 2 3*45 0 016? % M9W H9 66 8 1817 1 7795 13117 WB IN e 2 a'47 0 C1676$ 147 at 547al SE M 1613 7 tils 6 0 3851 1 7717 13068 W3 103 95 30173 1819 S IM t 2 Md 7 404774 13S SS 13S S7 183 M 1914 4 S158 L7140 lasca ses e lag e 2 7838 112c 3 3 1313 3 7063 ists:

8 0476a 179 09  !?t il IE M 30152 3121 3 GIE) satu 13137 3m 8 WJ tag e 28ts? 0 016'93 177 98 173 0C 507 % leit e ll?? C to t 3 0s il 0 0st 303 18711 0 1985 1691C 18995 le t *

. _ . . . *. ' . san g 3 1997 004117 181 74 81777 til ?6 10t % 3017 9 1177 8 8 20:8 16534 18837 tea

. - les e 3 3tt) Ill M 1013 7 3373 6 01051 14769 IJe tt tes t 0043?? 306 S8 let H 183 M 303C $ 1174 5 02sa Imaa IA769 _. i Mel 3 5341 904337 10168 101 70 tse e 185 95 000lJ B125 3 42317 la610 1A727 tsaa the t 3 Tita SO4M3 9725 9737

~' 157 9 3 905.5 9 016h3 97 W 317 M 900E 7 IIM I SJISC 5406 12604 BIS 37 68 819 % 30C10 alm 9 sta t elC?> 0016E3 08 50 4. S? Ifl M 81183 laa43 labet B78 146 s 4 3t48 8 014374 Sa M 4005 8 3177 ? E2716 3 &390 32006 Ws4

-* 04 S7 173 95 300s i 1178 6 19e 0 4 sit? 984364 80 87 00 U 825 %

82743 14318 IAM& Ws8 lE34 lift 4 SJ791 1878$ !&M BaJ 39 4 7418 00llMS M21 M29 IM M r

387 8 4 9777 001W06 73 tc M2 1927 82313 18174 1887 We 7397 179 M IOC10 8131 0 taa s 12174 4 016417 M 70 7C 72 131 96 SJMS 1A103 lated E70 tes e 54673 404a28 9918 IDie SJ3)F 14437 1A809 5s0 6767 6768 13357 SIB 6 1837 6 tes t 1 7723 Sana 40 64 78 64 80 SJ409 ISell 13371 MS 13597 M74 13134 42e83 13897 1A113 Ma lit e 1 9976 40445t U ta 67 0E 13797 gg6 2 Ilw 2 1728 &J736 00lW63 M43 M as IM et geb e 31350 SN73 13077 13795 333 934 4 6MM 0 016474 M 95 96 51 Mi te Stil 11358 137H 12758 mJ 81s s S W9C 00 4886 la st Wel M199 7 33Ma IA271 taas litJ 2J8so SalW94 9t? 6 1836 6 52t68 13616 Islas SEA 57JS 57J6 MS 99 931 4 1837 4 i SJt0C 13648 1A147 Sha j.

%a

. ..vg.L , ........ ... .. . . . . . , . , . . .

. ~

so Table 1. Satureted Steam: Temperature Table-Contineed A0s Ptess 50ecific toivner tatma tp, intsep, 10 pre $at $at 1st Sat $47 Set' 3emt 3eet Evet V0pm Lated test 90pv fen' lan 50 mi Lited IsoO tape tited he hg s, sla 81 8 l 3 to e sa '8 h eg j

95 e 23110 16610 9021 30 27 GC 2 Ills r 32s31 Sete latil ma i 7mc 16M7 5 872 30 189 01 0 8139 0 r?t67 ,Sals 337% 3r8 3rs e 40s a inn 8 237 4 749 01 78 1133' 38a6 laso 90s a me

$2C3 s%8 1697 as 33 as a00 82 S Ilec $ 77$ 1779 IP m8 m3 0 967, Inbt Sall 42 U8 43 J 8841 3 _

PM .S!!3 IJent _W 8_ _

93 : * - 93se egi&S77 ewi . m957 los es 80 0 lier 1 esm? laiot 17tw me * * * *

• 323 9 M7 681664) 3 137 S Ma 36C 05 523 ll42 9 16 ISS? 17WC me 9es e E n3 4816M4 37 38 37 824 367 0S OB16 1143 7 8 13017 1735 tes s ta l 18 6Ct telWil S ka 5 36a sta tt nos llea a 79 la967 1 7:31 30 llea 301M?4 as tb4 38 970 386 0s 0793 11452 62910 145s lJ798 m8 30 sel II S?6 0 016637 3 6?? SI6M 2009 977 9 3546 0 tJean l ag?4 IH64 se 12 5;7 001u68 Si nn 31151 17288 954 11475 8 3001 1 4697 I Me6 Ss0 ass 077 8 31490 SJOLI l eS71 1 7632 38 Esa 13 % I 001W91 30 0E7 2887t 876 14 212 8 M et- 0 016719 8 7t? M P99 30017 970 3 1150 5 83173 l eaa7 17Mt 213 8

. n n ?*

-

• 24 8 IS SC; $ 616747 3e 876 fe tta 18410 9670 llS2a SJ181 J ,4323 17SCS MS - -

226 8 17 184 0 0:6 '75 2313! 23 Itf 18823 985 ? IIS3 4 SJ781 3 4?C1 174a? 31e 3 374 0 IS %t 0C1663 21 W9 21 545 19727 062 6 119 9 &l10C laan: IJ38: Del 273 8 2C C;S 904634 20 0 % 2C C?3 194 31 960 0 31bi l 833 % TJh 1 7320 at 232 0 21 M

• GCaed 18 7C: 18 718 200 M 9574 11670 63417 IJW2 37M: II:s .

,': v.N 23236 GC1689a 17 4W 17473 20443 968 6 ID92 SM76 IJ72$ [72C3 353 e

-

• D68 38 ti! O C169?! 16 304 30046 9S21 116C 6 83133 11609 37147 3e $~

3ee 16 371 E3Hi 33aN . I ? Ort ans 8 -

pas t M E't 00169S4 IS Ps3 IS f60 217 50 989 S 1162 0 pas i 29 W SC169% 14 ?W 34 78; fit M 986 8 lit 38 SJW9 13375 I7 Cit . - 888 8 25: t 3C EH 0 C;7C:I 33 351 13 375 22C C Saa 1 Ilw 7 SJ?ot IJ?tt 149?2 21: 1 fu l 33 0t. C C170t! 12 5 C 32 S31 22a tn 941 4 Ilu l DJ76) IJlsa Ish? Ess set t MAN 00:70Fi Il 7a5 11 767 228 76 93E 5 1167 4 C3t:9 1J3? 1 68!! In:I Jes t J? f- Sti?!?! 11 C?) 110G 23?t? 9359 11667 031's 17913 1601 ses e 36t t 4M: S C1715! 3C 356 1C 37) 23t l . 9333 Il7C t 0353: IJC3 16?ts 385 377 t 4) 2a a D C.719J 9 735 9 75? Pat ti 93* 3 1171 3 03H' IJ7;) 3 6?;  ??: 0 21s t e6 14 ' G C;1224 t il: S il; 245 0f 927 6 1172 S 043) IJK IMS: 276 8 49 2?" 0 0!??u 0 677 0 6a4 24917 924 4 11738 88095 IJSCI 185*9 ml ges.

3s 8 t Sy s:4 0 0. 7E BIM B laS) 2$3 3 921 7 IIM O 04154 IJMS ISWE 3b8 8 3es s  % 795 0017M 76tM 1 68:? 2S74 918 9 Il76 7 04?CI 1229: 1 6891 38 3t2 8 M 3M 0 C1738 723Cl 7147S 26: S 915 9 18774 S a?t! IJits least It's 2et t U Oba 0 01741 6A2S9 6 6a33 265 6 913 0 1876 6 4 43;7 1206? 1 6s00 38

~

-- see t 57 005 0 01745 6 4443 EstSt M97 91CC 117S ? 0 4372 t il19 16351 3RI ses s 71 119 0 01749 609% 6100 273 8 9070 110C l 04a?t 118') 163:3 Sat mes FS 4D 0015) S 76 % 5 763C 276 0 904 C 316 C 04a ?9 I 1776 16?SE Set 212 8 29 953 00157 S4Mt 6474; Fl? ! Sci t 1183 1 04533 1 1676 16?ct 312 0 3163 So ut 0 0176: $ 1673 SIWS te6 3 M19 lik t G alat I IS?6. 14162 318 8 3pe t M 643 0 01766 40%) 4 9138 29C 4 Ste l 1105 2 046s: 1 1477 16115 See ass e es 6't 0 0177C 4 6418 46%$ fu l 0916 1186 7 0 4697 1 83'8 160') 348 373 8 30C ?si 0 01774 44c30 447ct 296 7 008 5 11677 Sa745 1 173; Ig:S 313 332 s le E7 0 03779 4 1788 419tt 3C' t 08 3 IIN 7 S e79E 3 1143 1591: SI:I ase Ill5 0 0 01783 3 9681 3943 3C71 0423 IIBS I O 44g 3ist 15934 388 tes t 117 957 0 01787 3 7699 3 78'8 311 3 0788 119C 1 449C7 1 999: 13097 SII 8 32e 43: 8 C1797 3 S8M 36c;3 315 5 GMS lit: 0 4499 1 08,8 IS849 ans t ansa 131342 S Cl?t' 3 4078 3 4751 319 7 872 2 till ) 030D4 5 0'95 ISSM Sef t t # _.*" A>.

@ asa s 362 3 13t:38 G Cll;; 324:3 32K3 373 9 06t 9 lit? ? O SCSI 18705 15743 325 348 HS 424 0 01006 34063 3 1064 Sil l OLS S 41536 8311C Init 33721 368 Wet In t:0 0 01811 29M? 2 9573 332 3 0671 llW e SS161 1 9617 55678 381 she e Mc E3 0 018:6 250C2 23184 SM S SSE 6 1195 7 O S?!? I k?4 IM37 ast Ess Ml li) 00108 2 H91 2 68'3 3aC B 0% I !!959 0 5213 1933 IS$95 mal 372 s 377640 0 0lt'l 2 Wil 2M33 SaS O OSI 6 IIM 7 O S114 1 9280 15 %4 372 1 15 517 0 01&31 2 4279 2 4402 3e9 3 Sa81 1897 4 SSMS lslag 13S13 37s e 376 8 Wes 395 779 t ellM 2 317C 2 3353 3S36 tad S 1998 0 0 Seil 10057 1 M73 mI mas 30$ 2W etik: 22 ire 2233 m79 esc 8 iiu 7 GWu O M.6 IWr met 3B8 8 2]S 770 0 016s? f ilM 2 1313 M22 8372 IIM 3 S M16 0 9f ?t 1 515 3a8 32s 3?SS16 00llS3 2 0144 2 0369 34% 8334 lite t $$MT 0 9784 I S!S? 3:t Set 3M 193 8 01864 1 929 19:17 370 0 8?97 1200 4 SMl? 8 9696 15313 38

-=*

411 8 3s'2S9 0 01864 I gaea 3 063C 375 1 0?S 9 17C10 0 Mt1 09E' 1 S??t 9 5717 Stul I S2M

=0 est 4088 2S8 7?S 8018 ?C 3 764D 3 7t?7 379 4 S?70 120: S 40s 1 27C 60C 80:4'S I 6877 170ma 383 8 011 7 17C19 0 S'M 99Ct I S195 etI

• * "

• 4128 m?SW 0 0:62. 361S2 5 634; 383 884 2 17C? 8 Sle:t 09MI ISIS? dit t ele g 395617 SalN? 1 5463 I MSI 32S 0102 12C2 4 0 $0M 89?S3 13118 416 8 etI NE 700 0Glets t 4000 3 4997 306 9 567 1703 1 0 Hll 8 986S I 90 2 asB ees 377391 00190C 3 4168 1 4374 a0. 3 GD2 2 1703 S SHea 0 9077 1304 ese ce8 33 463 00190t iMll Ilit? #S7 798 e !?03 7 8 6014 SSMC I Sept esI sit ab100 0 01913 13C264 IJ7179 410 1 193 9 1704 0 0 6063 009C3 1 4954 820 est e 3603 0 01919 IJ4447 IJea0L 414 6 709 7 !?04 2 0 6112 8516 5 857,8 4368 esei Sei Se eSita lu?6 12i6:7 4i90 30%4 l?ta a e6ili 60729 Iatte ese ass e 39? M 0 01933 3 14874 1 36006 423 S WIi 120s 6 0 6210 00W3 IasS3 eas e amt 4W ol 00iWC n07:2 n2m 4Me 2m? uk7 062w 00SS7 le:S die s e28 43114 00lts? 19576a 1 97711 43?S 772 3 120s 8 063C0 0 8471 1 4774 428 e64 4 # 73 8019W ISIS!I 84M72 4374 3 73 120s 8 863M 8835 1 4741 e68 a

, ~ ~. 4. s:- >;.. .

a .- ..

> :c .5 : : .. . .. . .: . ; . - .e . , .m *

.c.--- 1.~ ~:=7 <~l'..; ~~ ,--~ - .

T , .-

i E

Table 1. Satursted Sleem: Tomssersture Table-Confinarect letmatp? introp, ett Pien 50ent.c towme

$at $41 Set leg leg $41 $a1 $st Le See loop Sepo' Late a lose ease, tatec Es00 yppe. few fem se en titre t ag Og 6, 51 I I O 9, esa vg es Sea 481 5 Elf 8308 0 0 4401 $8M9 IA70s mes SIG eM S' 00lMI 8 97463 490s74 768 6 130s ) $6sb8 88213 IA067 est 40t W SGIMt 89350s SMSS7 ea61 G88 es8 eC7 fts e 8708 6 4 460? tal?? SM est tes t) tellM 80t84) Stim? SSJ Nf3 17883 il tes? S? 477 3

- 477 0 SN 67 0 01964 800a3 4a079 6136 leu alta GILA 085 18 401962 88M68 Stette ens Sees laaJ 585 199 6 .32081 Sept 8707I 3 518 _ g l ,_,

est 004 15 0 07000 8 79716 6 01717 04696 07705 8 as8: ausa

• es6 S8701 647009 8 70613 4 70627 etI 72 7 370 B as 4738 779 7 1203 5 0 6746 0 7700 l eada est 61C IC te7el? 0 7Ma7 0 ?MW 774 6 1703 3 067t3 0 76Ia i asC7 47e 613 C3 001C76 0 70738 8 77820 878 5 .- ,

. e26 e3J 739 3 1202 7 0 4042 47W8 14370 est eBS 866 61 StFGM 048L5 4 70100 80C tf 807D83 S&Wat R&7at? 573 714J 17072 testo 0 7443 3 4313 est e f

Wet e77 305 0 17C17 OM39 0 73$7 1479L mas tes t 705 78 0 02053 SW135 06stil e73 3827 u0n Sou? 0?m Icu eme i On 9 Fu e 007Er nMm unt? 9023 N62 1700 5 $ 70M O 7tti 8 A221 0173 617 8 757 72 G C2C?2 S ts?l8 06C799 470D 6 7006 1 4143 St&J 70' M 0C2081 0 3 5997 SSE79 007J N27 litLB 84J Sl? C 570 11M C 0 7113 .0 7013 1 Alas ses 678 9 012 53 0 07091 0 53868 0559M Os! 3 lite r 8 7182 06t26 8 410e 948 -w . g,.

4 """- -

• 574 8 081 08 0 0:302 0)lsla 0139a S C2112 046M3 66793 sit t Mla 675 % lit'3 0 7731 0 64'4 l ac70 GM 8 578 9 G PC U 0 77K . 06 ?S2 l eC37 087 3 H:0 000 M G C2173 S a7W7 ESOC7C SM8 till litt e 131 IF O L2.M 94L123 04L?S7 531 7 063 6 11934 0 7325 &&Mb IJn3 035 8 SM S *-

SM I $175 11W 3 0 7378 01577 . tse 0 M49 tC 75 0 C7846 0 44M' S C?lS1 0 470*7 S 4:513 0 44t M Wie SS13 193 1 0 7427 tas75 _1J 1 M,9 15 Onss . , ; , .,

tes t Ott 22 p5C 19: 9 0 7474 0 640C 137t Ost s tes t IC?t a9 G C24th 041M4 041:17 set t O c?lt; CJW 79 04M60 36? C UfS  :!K t 0 75?? 1 4311 1537 067 8

$12 3 10i.2 59 0 7576 0 &I22 1 3767 - Set s IM7 55 C C:lk CJ7%e 04C4C 557J U2C 11852 548 11!? 38 S C??t' C 3st!' 0 38734 M!d 6'53 Ill' ? 0 7tM 0 6132 IJ75' Sta t Mtt 0 7474 C63: IJ'll 06s 1 IIM IC SC:::: SJ73?C M't 6181 Illi l 36s t I?C' 77 C C 22$ C h?".

CJ374 0 3t?'S 5725 6:11 IlW S 0 7771 C 595: IJt's hat Wit

$77 8 I?st ?f C C:!ds 8 3?a5 0 34+ 75 5753 6041 lit: 7 0 7776 S$tti IkW 8724 CJM?t 30 7 S$ 7

• 118; 5 0 752$ CSMt IJS5; 878 4

!?ls 174: 74 CC2W C 311&*

0 C:275 CJef7? CJ27J6 SF91 Sf 9 5 115 C 0 78'E C 5t73 13%* Ose t Mrs 13M I? gc;a 117tl 0 75?? C 5585 13SC' tes t tes t 1W7 O C??tt 0 28'53 C 310st tes t Baltt e C:311 CJMai 0295:5 60. 1 574 7 1174 8 0 M78 C Sall IJs6a ses e Mat Oct ? $64 9 3372 9 0 0C3C 0 53 % I W7C ts2 0 teJ e 8813 3 SC237f 026ste 87at?;

S C2341 02bs?$ SJ777C 6144 S$8 8 ll?C 2 0 0082 0 5263 133M Ost e ten s lai?A S C?Ma Cla3sa C16747 ' 6!71 $6ei iii? ? 8.iu Giits iJi>C .

st8 15832 943 IWi? S C2367 0 23174 C21757 627 9 ta?? 1865 3 0 81f 7 05M7 IJ76a Gas t IU73 0 C;ar DJ73W C247ti eta W34 Ilu a 0 0;a3 0eM7 IJ233 ase Oss e 012 8 IMil GC7477 0 21842 0138t5 SM8 Wa7 in95 S t?W Getti IJitt 412 8 tila 37359 0 0244a 820t16 C22960 es&S S15 6 IIM 4 SAM 8 04?H IJ141 016J 07e 8 3795 9 0 C7468 C196:5 0170tl estl 90t 3 3153 7 0 0832 046ft 3J09: 970 8 E74 e ILM C 0 C74!v 0 18?37 0 21726 653 i eMi 1149 4 G esif 48583 IJ04: 874 8 Its!

• S chis 0 1788: 02C2W 659 5 esi f 1846 1 e t514 0 4474 32968 07. s 073 8

.lW'c S CMM 017ka OltSE? 061 5 eMa 1842 7 0 8171 04M4 1J938 523 637 8 038 8 20C 8 S CDM C 16?25 E18782 672 a 4657 1136 1 00L26 04h1 12871 036 3 Osl8 70M 9 00395 0 15427 010C71 079 3 4546 1133 1 00t0f $4134 82071 Gas t ess e 2118 3 G CMM 0 1464.: 017Mt 355 as3 3 Il?t t 0 8746 0 40:5 1176: essa ess e 2176 1 SCMi' G IM 76 0165M 329 431 3 Il74 C SN3t 0Jes3 IJtM Gas e 0 0766: 0 13174 0l$416 30cc 4187 1118 7 4 0868 03767 IJEM S20 Sof t 22252 36A 0 C2728 4 12387 S Mlll M74 457 tiill 0 0531 SJU7 IJM7 086 3 *3C. 7 m21 1107 C tsats 0Jsc7 lJets One "Idi.[.*7.5' Oss e tw57 Smu O!!st? 8 34a33 Fia 9 0 9068 SJMI IJan Sat 06s 4 N3:1 BC2818 S itN' 0 13757 722 9 377 7 110C 6 Pa981 S O?tS6 G IC??l 013C87 731 5 571 3053 5 8 9137 4J210 12547 580 SM S 8774 2564 4 0875a S et$1s $ 17474 PaCJ 3sh 7 3001 9 et?!? SJ0sa IJM6 677 8 874 8 7636 8 SA297C 0A8795 C llMS MlJ 379 3 30773 4R287 82812 12179 096 4 77088 GHC37 000ht" 0 11117 758 5 SIC I NHS StMS 82770 IJtet 300 ess t G Ma' 42S37 I ntes esas Gaa t 2 78: 1 003ila 0 C7345 C loat) 3bt 2 ftCJ 10ba s 368 7 1987 C StS35 01337 ! ! 8 72 M88 Ost e 29t's 003?OL 00llti S M 7tl 778 8 SEM 021C 1 8744 37e 79M ! 0233:3 0 06'97 D M110 78C 5 7438 3033 6 851 8 30;3 8 BCMh5 8 88116 0 06371 maa 212 3 10172 0 9741 tlki Illt! Sta 360 i 30W 3 30M&7 3 63857 0 C7519 0724 177 7 OttJ 0 Mel 8149C l13K NIe ,

m.._.- att 10:2 213t l Sc3874 0 03173 00f 95' ent laa 7 971 7 Imoi 017a6 1 875? 192 8 0 04100 S cill2 00L330 Ba2 807 0 GM 7 1949 0 0876 l leet ans

_..' met 31772 Wie 258 3Mt3 Gau?? G CJ 30s 80$733 073 f 614 eMa 1 8325 00t?7 IWW - - - - -

set t SC 90L C 33L12 0 0000 1512 Ele 7*

3Be ef 32C11 0 05C78 00000C S C$C78 1

- _ 1 e 5 r

% tat tem 0eratre s. .

g i

.e

.'  : ,7. . l . .  %; , ... . .;;. , ,, . . . g.; , % .p ,. ;; r ;w. . _

3, _, ,, _

~

t Table 2: 's'Sturtted Steem: Pretsure Tal>ie . C Specitst teleme isthelpy EntrepF h s hets 1eseg Sat Sat Sat Set Sat $st . ht hvis 10 50 la faln Esop esper ed Esot Wape' VeSoe , $W54 se

. , be.wt Lat.ed Ea.sp i

8 38 0 3 0,0 0 0 8 9 00- 0 Saleta 32 011 0016c?? 3302 a 33074 $mlC3 8075 5 307S 5 a000C  ; l877 ,'3372 Gass:

8 23 98 323 27Js2 'es?S  ;'0067 --SA - -

. 0A@C32 12JS S 123S5 30bc l test a 4csa?  ;

tae . . M SN 13 30134 Stauft 6all 6all 47 W3 lest e stet s 0 0gM 'saan ODO 0 38 0 016IM 313 St $33 6C H F3 ISM i 1058 .

I1306 San 978: la la 86224 0 036407 nS15 73 S3? 2C D t 131 3 i llps acta Saa! La

. MS 321 0 01459; 40s aK 126 3 143 3 i i ft M 3043 75 75 38 y,

ta M8 12 0C 0 016719 C 799 0C 17 70 3 160 5 < i3471 4447 Put tems t&J 1133 0 016726 74 40323 997 ISO S I l3137 4415 .7552 23 se t 727 M 90168M 30 070 N 087 89627 508 1154 3 03358 33962 1 7320 33 38 PSC W 0 017006 13 7?M 13 74M 218 9 Sat ? 1164 8 0 3682 IJ313 3 6995 as me 3E71$ 0 017151 80 4734 104MS IM I 0336 IIH B 0 3124 326aa 8 6765 as

. Is t MI C2 0 C1777a 0*M7 S S14C FSC 2 073 9 5174 1 0 4112 1 2474 1658c Sai .

Ge t 25 71 0C:7383 flM2 P l?M 2622 9154 16ss;

. 1s t 3C' 93 S C174U 01875 ~ 6705C 2727 0074 1177 4 ilk t 0 4273 t aall 12167-lita I Uli 38 23

i. A x '. -

unsucheC3 at 312 08 0 C17573 taSM 5 4718 - 282 1 WC9 1183 1 GaSM 31675 3 A206 as m3 37C28 0 017659 4 8779 4 9953 FK7 Skt lits 3 0 4643 11470 3 6113 33 1st t 327 3 O C:77a0 4 4133 4 4310 29t S Bai 118'2 0 8743 11768 3 6C?? te I ..

118 3 334 4 0 0178? 423X 4 0464 3ct s m!1 1384 5 t otu 1 1815 lug tis t 179 8 N:27 0 01789 3 7057 3724 312 6 0773 IIK 4 S allt 1 0960 134 't tal 7. . -

s 83s g 3d?J3 S C1791 3 a362 3894 Sit C 0728 lit; 7 e atst tests I St;) tal

'- tag e 330: 0 01833 32C10 3239C 3M C %tt lit 3 C O Sch I MII

• 1375: 400 .,.

ist : 3St a ? 0 0lbM 2 96SE 3 C:35 33; t 004 IIN I 0 5181 I CSSA I W tt ust 16: s M3 5' C C 815 28 SS 2833t 3M 1. 0%C liti ; S S?0t l ed3S 136A: let t

. 17t t 3mi C C Cll?1 2HM 2 6?3t Sai ? 054 8 319 t O S?tt I C3:2 I Sit; 1748 tt: t 373 Ct C C1877 2 S:25 2 5312 Sa' ? 05C 7 lita l 0 53?t ISils I Shai tes t 18e 8 37753 0 C4833 23847 2 4C3C 3M $ NL7 lit? E O $364 19113 13461 les 8 3st t 3f: t* C C18M  ? ?lts 2 78:3 3SS S N28 1111 3 0 5836 10 Cit 15 ass 30s:

27 t 31: 5: s Clu: 2 163?3 2ll?if 3H5 436 1 1195 C 0 SaK 0 95:3 3 5a:3 Fis 27: e 3tt 66 C C;tM 2 0t?76 2 08tM 364 2 sala lits t 0 554 Stua 1 5374 Dtt M17; CCitSS 197H: I t%:t 36f3 831 0 lit t 3at t 3C 35 B C;tM I A*lM 170C 4 C S$8e 0 1789 I S3M 238 1 91765 3723 078 8 170C 6 SMM 0 %4S 3 5 ?94 3at t 26s t a0C l? C L!sti la7sb2 1 48317 343 02S C 12C: 1 OM4 0 95f ? 1 5?64 Det saa a 40 da O Cit?O I 75S48 I 77818 349 821 ( IK $ 0 5722 0 9530 8 523C as8 lig t aC78: 0C:54 149.37 171Ci3 306 018 3 IK:6 0 5764 06433 I Sil? Ste t set t all C' 0C188: 1 63169 16$089 3e' l SIS I 17C ) 0 5805 OlH1 IllM mi 3es e al4JS SC1885 15597 ISWD 39C 4 012 C IK26 0 6844 0 1751 3 5135 sua ute 41735 0 01989 1 37383 314??4 398 0 Ett 370* 9 S tat?

anc : a3173 0 C1912 I Job 42 137%a 4046 0 9223 1 5105 mt 754 2 I?os 0 06M5 0 89M 3 at68 Sta t act e aas 6C 0 519 M l.84162 1 16095 474 2 P&C 4 120s 6 0W47 0 8637, 3 484I ses 8

- - . alt t 4M 't 0 Cl958 I C122a 1 03171 4373 7675 I?Os t 0 6360 0 8376 5 473r ast

68. 6 46 3 C. 00194 0908' C lDC 485 5 75S 1 1704 7 0493 0 8i88 l ain mg utt 476 W C Citra s 082183 0 82177 eM S 7433 170a 3 Guit 0 7tM . l 4bt? tet t ea:t a8i!; O C70:3 0 749t! 0 7HT5 4717 737 C 32C3 7 0 6723 0 7739 l ast! mt Est e aw fi S CK32 0 68811 0 70M3 48: 9 FM S 12C2 8 0 6876 0 75S2 3 438: as4 let t Sc3 t 6 S C2CSC 06h0S Cluk 4916 71C2 12C.J 0 6978 0 7377 3 430C M4 758 9 51t 6: O C2065 058880 0 Scosi 50C9 H5t 120t 7 0 7C27 0 7210 3 4212 asa St: t Sil!. S C?tif O bs8M 0 %Ik tot t 605 6 Ilts a 0 71tl 0 70S1 lalu att ess e 12h24 0 C:1 05 051167 C S3K: 518 e 645  !!9f C 0 71t? 06t9s 3 40 % ase

- Set t SJ19i 0 02123 S a ?t68 03aCs. 52t ? eH 7  !!M 8 0 7279 0 6753 l aC37 38 SSe t 13! !? 0 02181 t eboba 0 472C' S34 7 GM C llW 7 0 7358 G H;2 IJ57C est less s Saa St 0 C2155 0424 M C 409t $42 6 6SC a til?I 8 74M S 6474 1J91C tuta test t SSC S3 0 C2177 0 a 3047 0 42na SSC 3 64:6 lit; C 0 307 06344 IJtSI ans8 ..;:t4"[?.c p tme SM 29 007:t5 03780 04 acta 557 5 63: S I!n l O n7: S trit 1379 tiet s tint e Mi s: S Cnia 03 San C IK73 ssa t 6:: Illic 0 7647 Olon IJ7n tuat 8766 e M7 il 802:32 034c13 03ces 571 9 613 C 11643 0 7734 0390 IJ60 tal 17bt s 572 31 0 0225C 03230t D NSM $75 8 GC3 8 Il8? f 9 778: S SSSC 13t 3; last 134:8 $774! O C7?H SJ0722 0 3294: tal( 94s t IIK 2 0760 0 5733 1J577 tas3 13St t 58: 3: O C;288 015250 0 3153' S$7 3 Set 4 11773 0 7 tot 8MK IJS?S tast

. Mt: t M7C" S C23CT 027871 830176 Ses t SMS 1175 3 0 79M $ 5507 13a 74 tsun g leSt e H:7C 0 C7327 SJ6584 02t9:1 ect ) M?s 1172 8 0 3::?f, 0$M7 33a?! Met s 164: e 69t M S D2Mt 02S372 02773% 611 7 658 4 ll?C 1 O 3CES O3?tt 1 3373 test t Inst 60C SS 9 02 % 6 0242h 0266Ci 618 C Sale tilfe e sis? $$182 3 3328 Itas t 16et t 6688- 0 02M' 0131H 02SSas L?a ? $4C 3 Il6a 5 0 8196 05C76 13Ds tese s asSe 6C9 CS 0 C2a;7 0 221s3 024 % : 6304 S31 3 116: 6 0 8.4a 0as73 13:2t taas 17e6 4 6:3 13 SC74?t 0J:178 SJ3h0' 636 5 $222 lib' t 013C9 04M7 IJ:76 11m s

_ . .[ *

~

. 17Se t 61717 0 0?aSO 02C?63 0 7M13 682 5 5111 11 % 4 0830 0 87t$ 3!?t I?ns e . -

lett e 62; C: 8 02472 0 19?90 0 7156: 64t h

,ast e SC3 4 15: 1 O sal? Cau! 30 4 tats --

624 8! 0 0?a95 010M8 0 210S? 6ha t 4644 lat t 0 8470 04MB X3C ant e Set t 628 M S C?Sl? O 17761 020??t the e 40$ 2 185 t 0IS22 0 4459 2%; tutt i est t 63722 00?Ss! O 69 % 0 9 sac te6 3 4758 42 c 0 3 S 74 04354 ,Jtil als 8

• Set 4 th K $C?MS S ,6?M C 0838 672 i eM 2 38 3 0MM Sa?M Jeal

• 3: 8 met t 6a? 76 0 C7615 0 48tt C FSO: 633 8 4a6 7 8 K$ 0 87?? 0 aOS3 J7K Fies e IIst e 6avat 00?%9 0 .3603 0 UFJ 69s t 476 7 1.2: 2 0 8878 0 304s IJ64 Isen t Spas s 6M fi 0 027N 0 12436 $ l5133 7C7 2 406 0 til3 ? 0 8979 S ust IJW9' 331 paal t M2 il 80279C 0 31247 0.'4076 talt has llc 3 7 0 9C31 SJ4Jo RJe60 ses mee t Mi ll SA?tS9 010?0s 013sse 731 7 3ill 50933 0 9139 0320( IJMS Ms g l

Sies 9 673 9. 0 02938 0 MI72 0 82880 Paa S 3376 500 C 0 9247 Oft 77 ljt??S ans e i 21st s 071 53 0030?t 0 08165 S tilta 757 3 312 3 3866 7 0 93M 02341 IJM7 FIm e me a tt 0031W 0 07171 O IC3et 270 7 MSI leM t 0 9s68 02491 8 1958 me l Sust e 48'C 40 22 0 0326? 0061St S MsK 705 1 2Ss ? 10M 8 0 9588 02215  !!GC3 Mt j

mee t 095 33 S CM78 00SC73 000$0C WIS 213 4 B&2C 3 0 9728 Olps 1 8619 Wa s mes e 30C 28 0 0368: 0 03773 C eteb? 024 0 169 3 M33 0 9918 0 1460 t ilf3 Sin t Wee s 3M Dt 00a472 0 01191 00%U GPS S M1 03I 6 19M1 0 6482 1532 Wmg Wes 2' 30547 0 05078 00200 0 05474 M60 00 W60 8 0612 08000 Sal 2 M8 2*

WCntical perstere

• • ...i,.. ...

. . ? n.7 - * . e.--n.em.* u ~-- - -WN*N"**"**' w* "- - * * " " - " ' 1* ' I -

bem / Thermodynamics eI steam 4

33 I2 3.3 38 IS 16 3y gg 16 2 19 20 -21 23

/> l

'~ 8StamtTemperafwre >-F\/ n,U  ;

.5. jf4y7N / N //g ve ~ <- m A

,,./

x , , ,. -

gfgA / N///%

/ i

,5.

15m 7\ / / At/

N // %a j

/ ,f i45o -\/ l450

! / M/ /' ds 3400- / N U/r%j

"*! SgfNU7jy/4'j*'" '

l l h*/

I y

/

1350 13an 1250' a

///

l 7 n/wm %/

i

' ' / NW 1300 N 1250

'4 3200 ffffg /?? f/t D 1

I- //#b%' 77%i7gl yy  ; 1100 i

1050 1050

[

J000

- KL wmm<, // -

MMy,%g w(f/ .

J.amw -

850 300 JN%.T ,, ,,

Entropy, ago,g ,

l8 I9 2.0 23 1

yy g50 Mollier diagram (h s) for steam.

h .

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

ANSWER 5.01 (1.00)

The heatup causes the valve body and stem to expand E0.53 and the body oxpands at a faster rate resuting in stress on the stem and disc E0.53 REFERENCE Reactor Materials, page 18; Reactor Vessel and Internals, page 16; NUREG 1122, 3.2.1-002-000-K4.07 and K5.01 ANSWER 5.02 (2.00)

a. Feedwater heater shell pressure and temperature increased E.53 b.

Extraction steam flow was decreased E.53 c.

With extraction flow diverted through the turbine instead of the heaters E.53 turbine power was increased without a correspondin3 increase in reactor power E.53 REFERENCE PWR Experiences, 4961 Thermo I, chapter Si Thermo II, chapter 7 and 9 ANSWER 5.03 (1.50)

a. Reactor Power E.253 Increase C.1253

b. Main Coolant Temperature E.25] Increase E.1253

c. Main Coolant Flow E.253 Decreases E.1253

d. Primary Pressure E.253 Decreases E.1253 REFERENCE THERMO II, 13-23 t

6

,m ,,, __ _ _ _. . , -

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 q

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

ANSWER 5.04 (1.50)

a. A control rod's worth is proportional to the square of the relative flux it sees. Since flux density drops off as you move away from the core centerline, so does rod worth. (0.50)

b. Placing a second rod adjacent to the first will depress the flux in that area and decrease the worth of the first rod. (0.5)

c. The rod worth will increase as temperature increases. This is because the neutron migration length increases as temperature increases, allow-ing the control rod to see more neutrons. Since the moderator is less dense, a neutron will travel farther while slowing down and diffusing through the core increasing chances of encountering a control rod.(0.5)

REFERENCE Large PWR Control 6-22 ANSWER 5.05 (1.75)

a. Peaking Factor (0.25)

b. Increase. (0.25)

c. Does not assure. (0.25)

d. Not uniform, [0.25] The same E0.25] (0.5) 9.

i 3667 LV.4DJ TurLDEF owdy I a um LV.4eJ Uh

/ Oh_ -

A

)

swwci r Lv.4aJ C oser to L .ZDJ -

REFERENCE E h THERMO, 13-30 to 36 ANSWER 5.06 (1.50)

a. 1. Decreases

2. Decreases

3. Increases [0.50 each]

REFERENCE BWCB THERMO II, 12-36,37,

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

ANSWER 5.07 (2.75)

s. Build up of Xeno(,g) 05 2j~

to equilibrium in 45-55 hours [-Fv&&

end Samariu n 400-500 hours. (1.5)

E g}_

b. Xenon will irst decrease E.253 and then increase to a new higher value E.53. Samarium will first decrease E.253 and then return to the same value as before E.253. (1.25)

REFERENCE BHCB WESTINGHOUSE REACTOR THEORY REVIEW TEXT, PAGE I-5.78 AND 79 ANSWER 5.08 (1.00)

a. Large positive (0.25)

b. Infinite, zero (0,5)

c. Slow decrease (0.25)

REFERENCE Westinghouse Fundamentals of Nuclear Physics, P. 7-17 i ANSWER 5.09 (2.50)

a. The closer to criticality, (less SDM) the longer time required to reach a stable count rate. (0.75)

b. A higher initial count rate will result in a higher count rate at criticality. (0.75)

c. 1. Critical rod height is not affected. (0.5)

2. Critical count rate will be lower. (0.5)

REFERENCE Wastinghouse Fundamentals of Nuclear Physics, Pp. 8-55,58,59  ;

l

1

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23 AN5WERS -- BRAIDWOOD 1 -85/10/28-T BURDICK F

ANSWER 5.10 (2.25)

n. The magnitude of FTC decreases as fuel temperature increases [0.253 because the self shielding of the fuel decreases as the resonce peaks decrease in height. CO.53 (0.75)

b. 1. Fuel temperature increuses as power increases ,00.53 thus the magnitude of FTC decreases. [0.253 (0.75)

2. Fuel temperature decreases as core ages E0.53 so the FTC will A.--m p1 W CO.253 4 /p 81f8 fc.'# fey Ogh(( (0.75)

REFERENCE Westinghouse Fundamentals of Nuclear Physics, P.6-41 dM 7-fd ANSWER 5.11 (3.25)

c. Most of the coolant from the cperating loops flows through th~e core, but some will flow backwards through the idle loop (due to the delta P across the core). E0.253 The net result is that the idle loop temperature becomes Teold of the operating loops.

[C.53 (0.75)

b. Since power demand has not changed, the remaining loops must make up for power not being supplied by the idle loop. [0.53 With delta T across the idle steam generator essentially zero (may be slightly reversed), the delta T across the' operating S/G's and the core will increase. [0.53 (1.0)

c. With no heat transfer across the idle S/C, delta T (Tave.- Tstm) cost increase for the operating S/G's.. [0.53 This means that operating S/G tem'perature and pressure must decrease. [0.253 (0.75)

d. Total flow will stabilize somewhere above 75% E0.253 as the loss of 1 pump will decrease total head, increasing the flow rate of the operating pumps. [0.53 ,

(0.75)

REFERENCE Westinghouse Thermal-Hydraulic Principles, Pp. 12-15,16

\

o d

\

\

I i

B

_ . - r-

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

ANSWER 5.12 ( .50)

D.

REFERENCE FUNDAMENTALS OF NUCLEAR REACTOR PHYSICS 7-20 ANSWER 5.13 (1.50)

1. Sufficient negative reactivity must be available to achieve the required SDM at all times,especially in the event of a steam rupture accident

2. Minimize the positive reactivity which could be in ertey should a rod

• ejection accident occur yt C # jd A8# M

3. Provide an acceptable radial flux distribution to minimize peaking factors E.50 es.3 REFERENCE

'LARGE PWR CORE CONTROL 6-34 MY S/CC(

ANSWER 5.14 (2.00)

Normal steam generator temperature at 50% power is obtained from the steam tables. 1050 psis = 1065 psia E0.23 which corresponds to550degreesF.EO.23 Thhkt[h[m,generatorsafetyvalvewiththelowest lift setting will open at 1175 psis = 1190 psia E0.23 which corresponds to 566 degrees F. [0.23 The steam generator temperature must therefore be raised 566-550

= 16 degrees F. [0.43 7p_yp y N'orIb1 Tave at 50% power is (557 + 587)/2, which is 572 degrees F. [0.43 If Tave rises to 572 + 16 = 588 at 50% power the steam generator safety with the lowest setting will open. [0.43 fgggy REFERENCE BDWD PLS AND STEAM TABLES

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 6.01 (1.00) r

a. C and D

b. these are the spray loops CO.5 each]

REFERENCE MODULE 1, AC POWER, PAGE 50 ANSWER 6.02 (2.00)

A degraded bus voltage results in higher RCP motor current [1.0]. Higher actor current will cause the breaker t o t r i p 7 :;;.:! t i r.3 i r, t P. : 105: c' #1 -

[1.O3. U A10 d tAlbG P b6 'T21/' O(<- !T7 6 O ik REFERENCE MODULE 1, AC POWER, PAGE 48 ANSWER 6.03 (1.00)

An overcurrent trip on the SAT supply breaker to the bus implies a bus fault E0.5] which would adversely affect the diesel generator E0.53.

REFERENCE MODULE 1, AC POWER, PAGE 60 ANSWER 6.04 (1.50)

s. 6.9 kv/AC/NSR E0.33

b. 4160 v/AC/NSR E0.3]

c. 120 v/AC/SR AND 125 v/DC/SR EBOTH REQUIRED G 0.3 EACH3

d. 250 v/DC/NSR E0.33 REFERENCE AC POWER, PAGE 42 l

l l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 6.05 (1.50)

s. loss of ac power to the chargerde p

b. loss of de output voltage a

c. low charger output de current

d. igh charger output de voltage E3 required 0 0.5 each]

^~

kEF R CE 125 VDC POWER, PAGE 8 ANSWER 6.06 ( .50) c.

REFERENCE MOD 1, CHAP 9, PAGE 12 ANSWER 6.07 (1.00)

By closing the air inlet damper E.53 and shut off fuel E.53.

REFERENCE MOD 1, CHAP 9, PAGE 32 ANSWER 6.08 (1.00)

SX valve for DG cooling is indicating shut. Ethis is the only indicator 3 REFERENCE MOD 1, CHAP 9, PAGE 52 l

ANSWER 6.09 (1.00)

In auto the valve will divert letdown to the holdup tank in proportion to the VCT level error but manual is a complete diversion of all flow.

AHo,m u% adust. t&"nS % VCT bat- ns% merut.

i 1

I

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 6.

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK REFERENCE CVCS, PAGE 12 ANSWER 6.10 (1.00)

The valve is used to maintain proper RCP seal injection flow.

REFERENCE CVCS, PAGE 18 ANSWER 6.11 (1.00)

n. scavenge o:-:ygen

b. clean crud out of RCS CO.5 each]

REFERENCE RMS PAGE 10 AND 16 l I

ANSWER 6.12 (1.00)

c. Normal mode via the blender and BA pump.t/ldCUll#4[8

b. Emergency borate via 8104 and BA pump.

c. Via 8439 and BA pump. MMgt/h,wg,yp

d. From RWST via 112 D and E.

Evalve names are acceptable 3 C, Wh T-/d4 B745 Ya UCT' 'A/ N REFERENCE /J1 4 RMS PAGE 14

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK F

ANSWER 6.13 (1.00)

o. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at 100%

b. 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ramp to 50%

c. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at 50%

d. 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ramp to 100% E.25 each]

REFERENCE BTRS PAGE 2 ANSWER 6.14 (1.50)

e. They are normally open during plant operation. E.753

b. They are normally open. [.75]

REFERENCE RH page 8 and 14 (1.00)

ANSWER 6.15 a./115-1175 psis E.53 /L M g - 6/L

b. 11% [.53 REFERENCE MS page 37 l

ANSWER 6.16 (1.50)  !

y *l . A FPSF= IDI'J [b)

e. 3 x 10 pph E EJ j G,o g GL} l

b. TT has a slower response to error E.53 but LOL has a dead band E.53 REFERENCE SD page 8 - 10 l

l l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 6.17 (1.50) r

a. The flow was diverted through the crosstie to the RWST E.753

b. With the RWST at atmospheric pressure the RCS depressurized E.753 REFERENCE LER 483/84-16 and RH ANSWER 6.10 (2.00)

a. There is a bypass on auto start that takes the aux LO pump start inter-lock out. E.5]

b. e i c h t_ > 80 spm E.53 M :> 160 spm E.5]

c. Annunciator sounds when pot is set below 160 spm. E.53 OP Y Y REFERENCE M' #'

AFW page 11

> ANSWER 6.19 (1.50)

a. 1. detector

2. drive cable

3. thimble E.25 each] ,gf

b. when detector and cable are not in the vessel or stored E.e5 a when the thimbles are removed for refueling E.25] W M e , M, REFERENCE Byron / Zion LER's

@ (,&#)

i ANSWER 6.20 (1.50)

a. 4160/480 transformer E.753 ,

b. 20 seconds E.75] M .

REFERENCE Od O

- MODULE 10, CHAPTER 1,PAGES 21-23 p 1

- i I

7. PROCEDURES - NORMAL, ABNORMAL, EMERCENCY AND PAGE 30

~ ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~Rd656L66Y6dl 66UTR6L ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

ANSWER 7.01 (2.50)

a. SE-TSC C.53 SCRE-RSP AND THEN TSC [.53 NSD-RSP E.53

b. FIRE HAZARDS PANEL [.53 C. TSC from process computer [.53 REFERENCE 1BWOA PRI-5 pages 3,4,6 ANSWER 7.02 (1.00)

a. high rad sample panel OR TSC

b. steam generator E.5 each3 REFERENCE 1BWOP PRI-5, pages 21, 26 ANSWER 7.03 (1.00)

a. The misaligned rod is in a controlling bank E.53 AND

b. The misaligned rod is above the core bottom but below the rest of the bank. [.53 REFERENCE 1BWOA ROD-3, page 3 ANSWER 7.04 (1.50)

a. backfill faulted SG into RCS

b. blowdown faulted SG

c. steam dump faulted SG E.5 each]

l I

I

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31

~~~~R 5656[ 556d[~EUUTR6[~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK REFERENCE F 1BWEP ES 3.1, 3.2, 3.3 ANSWER 7.05 (1.00)

o. Vent the affected accumulator E.5]

b. If the cooldown rate in any cold les exceeds 100 degrees in any 60 minute period. E.5]

REFERENCE IBWFR-P.1, PAGES 8 AND 12 ANSWER 7.06 (2.00)

a. 1. any steamline pressure less than or equal to 640 psis

2. containment pressure greater than 8.2 psis

3. containment spray actuated E.5 each]

b. 1. containment pressure greater than 5 psis E.25] OR

2. containment radiation greater than.1x10E5 R/hr E.25]

REFERENCE ,

1DWEP-0, PAGE e

.i WER 7.07 ( .50) f l ce

'EFERENCE BWZP 400-3A1, PAGE 1 SWER 7.08 ( .50) b.

REFERENCE 300-1, page 2

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32

~~~~Rd656E66Y65E~C6UTR6E~~~~~~~~~~~~~~~~~~~~~~~~  ;

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r I ER 7.09 ( .50) do

'EFERENCE BWZP 100-1, PAGE 3 A ER 7.10 ( .50)

Co FERENCE BWZP-1, PAGE 2 A WI:R 7.11 ( .50) bo

'EFERENCE BWZP 400-3A1, PAGE 1 cWER 7.12 ( .50) l c.

FERENCE BWZP 400-3A2, PAGE 1 AN ER 7.13 ( .50) do ERENCE E.WZP 400-3A5 PAGE 1 I

I l

_- c

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 i

~~~~R 5656[65fEEE UUUTRUE~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK r

4 ANSWER 7.14 ( .50) b.

REFERENCE BWZP 400-3A3 PAGE 1 ANSWER 7.15 (2.00)

a. dose rates and monitoring requirements

b. required protective clothing

c. required respiratory equipment

d. timekeeping and dosimetry required

o. special precautions due to unusual radiological conditions E4 at .5 ea.3 REFERENCE BWRP PAGE 6 ANSWER 7.16 (1.00) 3 Rem / hour REFERENCE BWRP PAGE 6 ANSWCR 7.17 (1.00)

In all cases other than high radiation areas.

REFERENCE BWRP PAGE 9

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34

~ ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~Rd656EU65CdL C6 TR6L ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER;

, f

_ . . _ _ ,,i _ e. , a <__ _ _ _ _ . . _ ,

gi o3 j

b. 1. If daily dose wil e 50 mrem E 6 2 u n ,- ,ve l m , w e i g i,,3, Tiame cutting, grinding, sewins ui- iir e i, i n a

~ - . . . . . - . . . + E=

REFERENCE BWRP PAGE 12 ANSWER 7.19 (2.00)

a. spent fuel pool E.5]

b. containment vpender for new [.25] OR RCC A change fix)ure f or ie or used E.25] gffg 6& ft ,) p fgM on /M

c. RCCA change f i::tur e E.53 (72, f p /

d. core location, 2-axis tape reading, weight indicator 1 m data 4M E4 at .125 each] g ggfyg REFERENCE -

BWFP FH-5 PAGES 3-6 ANSWER 7.20 (1.00)

Because the load cell limit circuits are calibrated on the low range.

REFERENCE BWFP FH-14 PAGE 2 ANSWER 7.21 (1.00)

To ensure an adequate water supply.

REFERENCE IDWOA ELEC-4, PAGE 8 1

e

_ - . . _ _ c- _

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35

~~~~RI656E 65CEE C6NTR6E~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK F.

ANSWER 7.22 (2.00)

o. inadequate shutdown margin

b. control rods ow TS insertion limit New I

c. failure of 1 -- =a-e rods to insert fully after a trip

d. unexplained or uncontrolled reactivity increase

e. uncontrolled cooldown

f. inability to borate normally C4 at .5 each]

REFERENCE 1BWOA PRI-2 PAGE 1 l

1 l

l d

_. _ . . _ _ . . . . . _ . . - . . - . , _ . . . , _ , _ . _ - _- . . , _ , _ y __ ,_ - _ , . - - . . _ . . , , _ . - _ . - - .

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 8.01 (1.00) ,

SE los entry.

REFERENCE BWAP 300-1 PAGE 4 ANSWER 8.02 (1.00) gpN ,

By observins expected indications and plant system reactions L/Jf. bO REFERENCE BWAP 300-1 PAGE 11 ANSWER 8.03 (2.00)

o. After taking safety related equipment 00S.

b. Following the return to service of safety related equipment.

c. Following maintenance or modification on safety related equipment.

d During the performance of safety related surveillance.

o. Following the placement and removal of temporary alterations.

REFERENCE BWAP 300-1 PAGE 12 ANSWER 8.04 (1.50)

o. second qualified individual

b. outomatic status monitoring system

c. functional test REFERENCE BWAP 330-1 PAGE 7 l l

l

. ~ - - , , ,

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 8.05 (1.00) y By special letter E.53 signed by the plant superintendent [.53.

REFERENCE

BWAP 330-1 PAGE 7 l ANSWER 8.06 (1.50)

c. BWEP'S-RED

b. BWFR'S-ORANGE

c. BWOA'S-YELLOW REFERENCE BWAP 340-1 PAGE 6 l

1 ANSWER 8.07 (1.50) j o. Letters mean the steps must be done in sequence.

b. Open bullets mean only the steps that apply can be done in any order.

c. Closed bullets mean all steps can be done in any order. [.5 each]

! REFERENCE BWAP 340-1 PAGE 4 ANSWER 8.08 (1.50) 1614 on 11-01-85 REFERENCE TS 15.4.2.1.2 f

ANSWER 8.09 (1.00)

Continuously if any condition higher than yellow e>:ists E.53 AND every 10 to 20 minutes otherwise E.5].

l 1

f l

l

[

i . _ - -

4

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK REFERENCE BWAP 340-1 PAGE 10 e l ANSWER 8.10 (1.00) i ficw charts REFERENCE BWAP 340-1 PAGE 12 ANSWER 8.11 (1.00)

To indicate the page is complete.

REFERENCE 340-2, page 1 ANSWER 8.12 (1.00)

o. closed E.53

b. yes E.53 REFERENCE BWAP 340-2 PAGE 6 4

i ANSWER 8.13 (1.50)

o. The individual who requested the 00S. C.753

b. By placing an asterisk next to the equipment in the '00S BY' and

'R/S BY' columns of the 00S form. C.753 REFERENCE BWAP 330-1 PAGES 9 AND 10 l

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 8.14 (1.50) y

a. single - recorded to meet station commitment dbtf 2

b. double - running equipment only M^ '

c. triple - TS surveillance requirement E.5 each3 REFERENCE BWAP 350-5 PAGE 2 ANSWER 8.15 (1.50)

a. 30%

b. redundant equipment in standby

c. equipment in a tripped status E.5 EACH3 REFERENCE BWAP 380-1 PAGE 2 ANSWER 8.16 (2.50)

o. All penetrations required to be closed during accidents are either capable of beins closed automatically E.253 OR are manually isolated E.253

b. All equipment hatches are closed E.253 AND sealed E.253

c. Each air lock meets TS compliance requirements E.53

d. Containment leak rates are within TS limits E.53

o. Each penetration sealing system is operable [.53

REFERENCE TS 1.7 l

l 1

e 1

\

i l

. [

s ]( j l i

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40 ANSWERS -- BRAIDWOOD 1 -85/10/28-T BURDICK ANSWER 8.17 (1.00) r

s. 2 or P

b. 4 or R

c. 1 or S

d. 5 or SA REFERENCE TS TABLE 1.1 ANSWER 8.18 (2.00)

c. MTC within the analyzed ranSe

b. RPS instruments within operating range

c. pressurizer in operable status with a bubble

d. reactor vessel above minimum RT-NDT

o. P l ant above the cooldown permissive (P-12) E4 at .5 each]

REFERENCE TS B 3/4.1.1.4 l

O

,