ML20155K406
| ML20155K406 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/23/1986 |
| From: | Brockman K, Burdick T, Castro C, Mcmillen J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20155K390 | List: |
| References | |
| 50-461-OL-86-02, 50-461-OL-86-2, NUDOCS 8605280133 | |
| Download: ML20155K406 (76) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-461/0L 86-02 Docket No. 50-461 License No. CPPR-137 Licensee: Illinois Power Company ATTN: Mr. W. C. Gerstner Executive Vice President 500 South 27th Street Decatur, IL 62525 Facility Name: Clinton Nuclear Power Station Examination Administered At: Clinton Nuclear Power Station Examination Conducted: April 28-30, 1986 Examiners: .'i.c il en [h Id l
756avAtb y 27 K. E. Brockman Date /
/1bM
/ ?
J. F. Munro 28
&6t<4k i f C. Castro 2 fp, ,of /)wY Approved By: Thomas M. Burdick, Chief 67NM Operating Licensing Section Date Examination Summary Examination administered o_n_ Ap_ril 26-30,1986_(Repprt No. 50-461/0L_86-02)
Written examinations were administered to'~f'ive senior reactor candidates on April 28, 1986, Operating examinations were administered to four senior reactor and two reactor operator candidates on April 29 and 30,1986.
Results: Four senior operator candidates passed the written examination, and two operator and two senior operator candidates passed the operating examination.
8605280133 860523 PDR ADOCK 05000461 V PDR
REPORT DETAIL _S_
- 1. Examiners J. I. McMillen, Region III K. E. Brockman, Region II C. Castro, Region II J. F. Munro, Region II
- 2. Examination Review Meeting At the completion of the written examination, a copy of the questions and answer key was left with facility training personnel. They were requested to provide written comments on the examination and answer key within five working days. Those concents and the resolution of the comments is attached to this report.
- 3. Exit Meeting J. Munro, K. Brockman, and C. Castro met with M. Lyons of Illinois Power Company at the conclusion of the operating examination and informed Mr. Lyons that both steam flow annunciators would alarm when only one
. channel was down scale. This caused confusion among the operators.
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Senior Opera _ tor Examination Coments and Resol _utions Questions 5.4:
Facility Coment: Answer is correct; however, candidate may respond with critical point vs critical temperature.
Resolution Credit was given to responses that indicated an understanding of concept. The words " critical temperature" were not required for credit.
5.5:
Facility Coment: Answer is correct; however, the term " parameter" is misleading. Both power and neutron flux distribution are used by the process computer to calculate APF and RPF.
Therefore, core thermal power, neutron flux distribution and local peaking factor should be also considered as an acceptable answer. (Reference, G.E. SNE Manual).
Resolution: Credit was given for responses that indicated an understanding of concept. All candidates answered correctly.
5.10:
Facility Coment: The answer is partially correct; however, this effect on core-beta is negligible. The overwhelming effect on core-beta is due to the difference in delayed neutron fractions of U-235 and Pu-239. The difference in delayed neutron energies is negligible (0.432 MeV for U-235 vs 0.433 MeV for Pu-239). Therefore, the difference in leakace is also negligible. Beta (U-235)=0.006cvsBeta(Fu-239)=0.0021is the major reason for the change in core-beta over core life
(
Reference:
Glasstone and Sesonski, " Nuclear Reactor Engineering," Van Nostrand Rienhold Co. 1967 page 93).
~ Resolution: Comment accepted, but full credit was not given unless the concept given in the answer key was part of the answer given by the candidate.
5.15 Facility Coment: The answer for part "a" should be 3; e.g., as level increases in the reactor, the pressure differential across the level transmitter will decrease.
Resolution: Coment accepted. Answer key corrected.
5.17 Facility Coment: Answer correct; however, please consider adding the following to list of acceptable answers: 5. Increase in the amount of fission product poisons present (
Reference:
PCIOMR Implementation Procedures, Revision 5, NEDE-21493, February 1982, page 1.0-1).
Resolution: Coment accepted, but no candidate gave that as part of their answer.
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6.1 Facility Coment: a.1 The CRD FCV will fail shut on loss of air. (
Reference:
CPS 10P3214.015 Step 8.2.2.1.7). a.4 There is no TCV on VP chiller; however, student answer will probably be per answer key as the damper will close and system will shutdown. a.5 This valve will fail closed (Reference M05-1057 Sheet 1, Revision L).
Resolution: a.1. Comment accepted answer key changed, a.4. This part of the question was deleted since there is no such valve in system. Points were distributed among four answers.
6.2b Facility Coment: Answer is correct; however, candidates may list: Mode switch in refuel; one rod withdrawn, another (not the one withdrawn) rod selected.
Resolution: Comment accepted: Credit given for answers which indicated knowledge of system operation without using exact words in answer key.
6.3 Facility Coment: Answer is correct; however, CPS has no MSCV's or LPSCV's.
Resolution: Comment noted. All candidates answered correctly.
6.7 Facility Coment: Answers in blocks E, F and H are simultaneous interlocks and therefore, should be interchangeable. The s6me comment applies for blocks B and G.
Resolutions: Comment rejected. Chart was presented so that candidates did not have to memorize every step and should be able to fill in blocks in proper sequence.
6.8 Facility Coment: B or C should be considered as a correct response. It should be noted that the CCW pumps are not powered from the ESF buses. (
Reference:
E02-1AP03 Revision C).
Resolution: Comment accepted. Answer key changed to accept either answer.
6.9 Facility Comment: Answer is correct; however, students may respond with the following: RHR System A or B; Drywell Equipment Drain Sumps or Drywell Floor Drain Sumps vs Drywell Sumps; Cent.
Equipment Drain Sumps or Containment Floor Drain Sumps vs Cent. Sumps. (
Reference:
M05-1045sh12 Revision M).
Resolution Comment accepted. Answers given that indicated knowledge of sumps were given credit.
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6.10 Facility Coment: The key is incorrect. Procedure 3314.01 Revision 3 page 5 of 12, a note states, "An SLC pump will not start until its associated suction valve has cycled to the full open position. The suction valve will not automatically open unless the Test Tank Suction Valve is fully closed. . .
therefore, the correct response should be D. F001 does not open SBLC pump A does not start (Reference 3314.01 Revision 3 page 5 of 12).
Resolution: Coment accepted. Answer key changed.
6.11 Facility Comment: The correct response should be, slow in the clockwise direction.
Resolution: Comment accepted. Slow in the fast direction was also an acceptable answer. Answer key changed.
6.12a Facility Comment: The answer is correct; however, the RCIC system will isolate on High Room Temperature or Differential Temperature on steam escaping into the room after a period of time.
Resolution: Comment accepted and credit given for additional answers.
Key corrected.
6.14 Facility Comment: Answer 4 shculd reference H13-P634 or P639 (
Reference:
CPS 4009.0 Revision 3 Step 3.2).
Resolution: Comment noted. Panel numbers were not needed to receive full credit for answers, but if given had to be correct and candidates only gave panel numbers for SRV position indication.
7.1 Facility Comment: Procedure 4404.01 is titled Reactivity Control Emergency vice Reactor Scram. Imediate operator action is not specified; however, operator action is and the answer should include the following: (4)Ifmainturbineis on-line and MSIV's are open, then runback recirc flow to minimum; (5) activate the Backup ARI/RPT. (Reference 4404.01). Answer 4 should be prior to 3.
Procedure 4100.01 " Reactor Scram" immediate actions are:
- 1. Place Mode switch in shutdown.
- 2. Verify appropriate auto actions occur, manually perform any that do not.
- 3. If relief valves lift, or if lifting is iminent, evacuate containment.
- 4. Verify all control rods fully inserted.
- 5. Verify reactor power decreasing.
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- 6. If two feed pumps are operating and level is increasing, then secure one feed pump and control level in the normal band.
Resolution: Procedure 4404.01 refers to the Reactor Scram procedure and credit was given for answers that included those contained in the comments from the facility.
7.2 Facility Coment: Answer should say ". . 3 SRV's. . ." not 2 SRV's (Reference 4403.01, Revision 4).
Resolution: Comment accepted. Answer key changed.
7.3 Facility Coment: Answer is correct, however, 4403.01 cautions the "SRV operation in a sequence which results in uniform suppression pool heating per Frg 1. Then the procedure reference to MS Procedure 3101.01 which states to reduce pressure sufficiently to reduce the number of valve cycles.
Resolution: Comment noted. No change to answer key required since this was a multiple choice question.
7.4 Facility Comment: If the cperator at the remote shutdown panel transfers RCIC controls to the RSP the RCIC trips /isolations are by passed and RCIC may be run. (Reference E02RS99, SR104 and E02R199 sh 8, 13,501)
Resolution: This question was deleted since it was a multiple choice question and based on additional revicw of the reference material the examination did not contain a correct answer.
7.5 Facility Coment: CPS 0AP3302.01 RR states to take manual control of the FCV's and balance recirculation loop flows (Step 8.2.4) and Technical Specification requires the limits but CPS does not have CAF Procedures or Lesson Plans.
Resolution: CAF (Check at Facility) was a note to assure that facility personnel make sure the answer key was correct. All candidates answered correctly and the facility coment indicates that the answer key was correct.
'7.6-Facility Comment: Answer A and B are incorrectly stated, a 20% power reduction vice 30% is called for and if at high power and high flow conditions (100% rod line) directs power rod insertion to maintain APRM scram margins (does not mention
. . . monitored by APRM's) Reference 0AP4005.01).
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Resolution: Coment noted all candidates answered correctly. Answers A and B have been changed to 20%.
7.7a.
Facility Coment: This question does not apply to the Clinton Station
. . . Technical Specification Suppression Pool Level (3.6.3.1)are: 8'11" (12'8 condition 4 and 5) 19'5" Temperatures are: 95 maximum and Condition 1 and 2 105 Testing 110 Thermal Pwr 1%
120 MSIV shut after scram All procedures required observance of the above limitations. Recomend throwing out the question. There is no information regarding this question in any CPS Procedure or CPS Training Manual.
Resolution: Facility coment states that all procedures require observance of the Technical Specification limits and the basis for the Technical Specifications state the reasons these limits must be maintained. Answer key changed to agree with Technical Specification limits.
7.8 Facility Coment: Answer is correct, however CPS Procedure 3304.01, Revision 1, Step 8.2.4.2 states " connect a hose from valve 107 to a floor drain," so another correct response would be when no water is observed draining from the hose (Reference CPS 3304.01, Revision 1).
Resolution: Comment accepted. Additional correct response would be no water observed draining from hose.
7.9.a.
Facility Coment: Answer is correct. Some candidates may reply that maintaining a level to promote natural circulation is only a concern with RR secured. This should be acceptable (Reference CPS 0AP 3312.01).
Resolution: Comment noted. Additional information in answer, when correct, does not reduce the grade.
7.10 Facility Coment: Question states set points are not required. Answer 2 contains a typo. Should read from the generator with C02. Answer 3 should be with Service Air or SIA.
(Reference CPS 3111.01).
Resolution: Setpoints were not required. They were given in answer key for the benefit of the grader in case a candidate listed them in his answer. Typo in answer 2 corrected, answer 3 corrected. Credit was given to candidates for concept and sequence of events rather then for specific systems.
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7.11 Facility Coment: B should be 3 rem (vice 25 rem) (Reference RA-03, Revision 1, Step 4.12).
Resolution: Coment accepted answer key changed.
7.12 Facility Coment: Tolerances for setpoints were not called for.
Resolution: Coment noted. Tolerances in answer key were for graders benefit.
7.13 Facility Coment: Either of the following answers should be acceptable, both are correct. CPS 4100.01 does not address a failure to isolate, however, per 4001.02, Automatic Isolation, Step 3.2 says verify all appropriate automatic actions occur and perform any which did not occur. In addition, CPS 4401.01, Revision 6, Condition requiring a MSIV isolation is an entry condition and the operator action this: (1) Place mode switch in shutdown, (2) Sound the containment evacuation alarm. (Reference 4401.01, Revision 6 and 4001.02).
Resolution: Coment accepted. Credit was given to reasonable answers that indicated an understanding of the requirements to place the facility in a stable condition.
7.15 Facility Coment: A.1, When both CRD pumps are declared inoperable (CPS 5068.08), A.2. When more than one accumulator in inoperable and one of the drives associated with the accumulators is withdrawn (3304.01, Revision 1). b.
Delete ". . .and verify appropriate auto actions have occurred," This is part of 4100.01, vice 3304.01.
(Reference 5068.08, 3304.01, Revision 1).
Resolution: Coment accepted. Answer key modified.
7.16 Facility Coment: This question is misleading, as we normally use condensate booster pumps at this stage. If the question asked why we minimize flow demand, then the answer would be as stated, "Per 3002.01 minimize flow to minimize thermal duty on the RV feedwater nozzles and CPS 3103.01 has a caution regarding increasing flow causing positive reactivity additions.
Resolution: Candidates interpreted question per the 3002.01 procedure.
Credit was given for this answer. Question revised for future examinations.
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7.17 Facility Comment: This question could be interpreted in two ways; (1) If all rod position indication (all 145 nods) is lost, then the answer is True per Technical Specification 3.1.4.2; (2) If all rod position indication on one rod is lost the answer is False. (Reference Technical Specification 3.1.4.2 and 3.1.3.5).
Resolution: Candidates assumed that question related to all rods.
Credit was given for this answer. Question revised for future examinations.
7.18 Facility Conment: The answer should be: " Secure non-essential CCW loads, consideration should be given to transferring one or both FX HX's to SX cooling" (Reference 3203.02, Revision 2, Step 8.2.1.3.2). There are no immediate actions on partial loss of CCW. The abnormal operations section of the procedure states the following actions upon abnormal temperature on CCW Heat exchanger header outlet header.
- 1. Verify proper operation CCW heat exchanger temperature regulation valve controller. If necessary control temperature in manual or manual handwheel control.
- 2. Verify operating CCW pump-heat exchanger line up is consistent with plant heat loads.
- 3. If CCW heat exchanger outlet temperature remains high, proceed as follows:
- a. Vent heat exchangers
Consideration should be given to transferring one or both FC (fuel pool cooling) heat exchangers to Shutdown Service Water (SX) cooling.
A partial loss of CCW would most likely give indications Resolution:
that would place the operator in the abnormal operations section of the procedure. Required actions as stated in facility conments per procedure received proper credit.
4 Answer key wording revised.
8.4 Facility Comment: Answer is correct, however; some candidates may respond with viable methods / mechanisms listed in the EPG's to provide adequate core cooling; core submergence; spray cooling, steam cooling.
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Expanded in our Emergency Procedures, students also could respond.
- 1. Water level maintained above TAF
- 2. Reactor flooded as determined by
- a. level unknown: injecting with at least 3 SRV's open and RPV pressure not decreasing and _ 68 psig.
RPV pressure.
- b. Level known: injectir.g until RPV level is increasing.
- 3. Steam cooling in pregress
- 4. HPCS or LPCS in spray cooling or simply Level restoration; RPV flooding; alternate RPV flooding; steam cooling.
Resolution: Coment noted: Alternate answers by candidates that indicate adequate understanding of procedure were given credit.
8.6 Facility Coment: Answer is correct, however, some students may apply the definition of operability conservatively by concluding that since the valve failed the surveillance 4.5.1.d, that this would make the system inoperable and require Hot S/D in six hours and at least cold S/D within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Please consider this response to be acceptable as well. Additionally the requirement for steam dome pressure should be changed to 100 psig as penciled in an facility copy.
Resolution: Comment concerning change of pressure to 100 psig accepted. The other coment concerning the conservative answer is rejected, since the question asks for the action which most correctly details the allowance and/or limitation inposed by the Technical Specifications.
8.7 Facility Coment: Technical Specification total leakage as averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period per 3.4.3.2.c of attached Technical Specification supplied with exam is 25 gpm. Total average leakage is 25.4 gpm. This is in excess of the limit. (Key says 30 gpm), additionally the note to base the answer on the TS supplied with the exam is important as the 2 gpm increase from (0400-0800) is not included in recent set of Technical Specification.
Resolution: Answer key changed. Additional coment noted. No change necessary since answer was based on the Technical Specifications supplied to candidates.
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r 8.8 Facility Coment: Service platform hoist fuel - loaded, should be excluded from key as service platform hoist cannot be used for fuel (3.9.6.1).
Resolution: Coment rejected in part, Service platfonn hoist loaded is one of the interlocks that is a possible answer key changed to delete the word " fuel."
8.13 Facility Coment: The reference does not exist at CPS, however, CPS Procedure 1405.01, Revision 4, Performance of operational activities states: Verification of circuit breaker position shall include a check of the following items.
- 1. Physical position of breaker in its compartment / cubicle.
- 2. The spring charging toggle switch is on, if so equipped. (4.16 KV breakers do not have this switch, however, they do have a plug showing springs are charged).
- 3. The control power fuses are installed.
Other acceptable responses may include
- 1. Power available indication
- 2. Protective relay flags reset
- 3. Lockout relay flags reset.
Resolution: Reference was CAF (Check at Facility) since examiner could
-not easily locate answer in facility material so key was based on general knowledge. Coment accepted and key changed according to Procedure 1405.01.
8.17.b ,
Facility Comment: Before issuing at job site. Name SSN, Dose Margin; Initials. At the job site, however, time, exposure in and time, exposure out to the nearest 5 mr.
Resolution: Comment accepted. . Answer key changed.
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M STER COP _Y U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CLINTON REACTOR TYPE: BWR GE 6 DATE ADMINISTERED: April 28, 1986 _
EXAMINER: J. I. McMillen APPLICANT: $Md[h 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 Value Total Score Value Category ___ ___
25 25 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 26 26 -
- 6. Plant Systems Design, Control, and Instrumentation 23 A 24 7. Procedures - Normal, Abnormal Emergency, and Radiological Control 0
M 25 8. Administrative Procedures, Conditions, and Limitations
,10& 400* TOTALS Final Grade %
All work done on this exam is my own, I have neither given or received aid.
Applicant's Signature l
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- s SECTION 5: THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND THERMODYNAMICS 5.1 Select the diagram below (A or B) which indicates heat up stress. (1.0) ima, em i ua ama swa q swast t wm] swa g Z [\ s im s u s:
TENS 1LE \ TENSILE sTntss / sTaEss
/ smai mass . \
- ,e mm o- - --- -
o-E constssivE COMPatss!VE sTntss m ass g sTntss ,g (Cnoss-sECTION ,0F VESSEL WALL) (Cnoss-sECTION OF , VESSEL WALL)
A B 5.2 The convective heat transfer coefficient for boiling water is 300 - 9000. (Btu /hr ftsq F). From the below list choose the values that represent the convective heat transfer coefficient for film boiling. [ChooseeitherA,B,C,orD] (1.0)
Btu /hr ftsq F
- a. 5000 - 20,000
- b. 300 - 9000
- c. 50 - 3000
- d. 5 - 20 5.3 From the selection of answers in Column 2, choose the correct number identifying the horizonal and vertical axis of the curve on the next page of this exam. (1.0)
COLUMN 2 HORIZONAL = 1. delta temperature
- 2. delta pressure VERTICAL - 3. log (T[ clad] - T[ coolant])
- 4. mass flow rate
- 5. log Q
- 6. change in entropy
- 7. % void fraction I
gf NUCLEAR POWER PLANT T N SCIENCES polling Heat Transfer Reed Robert Burn .
September, 1994 Figure 9.3 Heat Flux Versus Temperature DiH erence Between Cladding and Coolant
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5.4 If a pressurized water system is operating at 705 F and ;
3400 psia, briefly explain what occurs in the system as )
temperature and pressure are increased at a rate of 10 units f per minute. Assume no physical restraints on the system. (1.0) 5.5 A representative value for a BWR TOTAL PEAKING FACTOR is 2.43.
What three parameters are used to determine the total peaking ;
factor? (1.0) 5.6 Using the following figures, choose the correct answer for each '
of the three (3) questions asked below:
les Peer __.
(
I O
flee di At I l2
~
Renon _ CN ll1 Ill C' Time
- a. What is the approximate time from Al to A2? (.1. 0)
- 1. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- 2. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />
- 3. 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> -
- 4. 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br />
- b. What is the approximate time frem B1 to B2? (1.0) :
- 1. 1-3 hours
- 2. 3-6 hours '
, 3, 6-9 hours
! 4. 9-12 hours I
- c. Why does Xe concentration decrease from Al to D1? (1.0)
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5.7 Answer the following statements about the Doppler Coefficient.
True or False
- a. Doppler coefficient becomes more negative ffoe O to 100%
power due to the inc~reased overlapping of resonance peaks at higher fuel temperatures. (1.0)
- b. Doppler Ccefficient becomes more negative over core life due to the buildup of Pu-240 and fission products with >
large resonances in the,epithersal range. (1.0) -
Which of the below best define, ' power density?"
5.8 (1.0)
- a. reactor p6wer in kw divided by the total surface area of the active fuel rods,
- b. reactor power in kw divided by uranium loaded in the core,
- c. reactor power in kw divided by core volume
- d. reactor power in kw divided by the active length of fuel
- pins in the core.
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5.9 The effective multiplication for a cold, xenon free reactor with its strongest control rod withdrawn is calculated to be 0.899. What is the reactors shutdown margin? Must include units in answer: (1.5) 5.10 Why does the presence of Pu-239 late in core life caasa beta effective to decrease? (1.0) m S.11 There are several intrinsic sour.ccs in a reactor. List three of these in crder of ' their contributions at the 80L. (1.5) 5.12 In which of the following situations is the control rod worth greater? EXPLAIN the_ reason for your selection. (Include comparison with the situation you did not select). (2.5) e SITUATION ONE -
- All control rods are fully inserted and 4 the center rod is then fully withdrawn.
- SITUATION TWO All control rods are fully withdrawn and -
l the center rod is then fully inserted.
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5,13 Dsfine the following terms:
- a. Void fraction (1.0)
- b. steam quality (1.0) 5.14 For the followir,g MATCH the cause of failure with its I associate.d limiting paremeter. (1.5)
CJuse of Faiiure Limiti5 8r**et P y
- 1. Fuel Pellet Expansion A. FLPD ;
B. CPR '
C. APLHGR
- 2. toss of Nucleat,e Soiling D. MARRAT around cladding E. LHGR ,
- 3. Occay laat and stored he8t following LOCA ;
5.15 D!ffprential pressure measurements can be uGEd to determine level, pre $$UTE, and flow. 700 GaCh of the following in COLUMN A, select the cppropri.3te typa of raiatiort. ship that
- exists, frorr, COLUMN B. (,1,0)
COLUMN A. (Item) C0ttfMN .B (Relatior.sbip)
Pre;.crt.ional t:0 differential
- a. i.evel 1.
pressure plus a constant
- o. Flow
- 2. Oraportiorial te differential ,
pressure alore 3, Proportional to the inverse differential presture -
- 4. Proportional to the square of differeftial pressure -
- 5. Proportional to the square root of differential pres'.spre.
7 5.16 Which of the followf ag is NOT CORRECT as applies to the impact ,
of de'J.ayed neutrons ort reactor operation? (1.0)
- a. When calculating reactor period, tha delayed neutron term may be coesidered INSIGNIFICANT if the reactivity addition is GREATER than Bata.
(*** QUE6 TION 5,16 CONTINUES ON NEXT PAGE ***)
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- b. The magnitude of the effective delayed neutron fraction (Beta-bar) is GREATER at EOL than at BOL.
- c. The delayed neutron fraction (Beta) is the RATIO of the number of delayed neutrons produced to the number of fission neutrces produced.
- d. The presence of delayed neutron causes the average neutron generation time (1-bar) to increase.
5:.17 The THRESHOLD power below which PCI failures do not occur is known to DEC.REASE with fuel burnup. State two (2) reasons for this decrease in the PCI threshold. (1.0) 5.18 A reactor tieat balance was performed (by hand) during the 0000 to 0800 shif t due to the ' process computer being out of service.
The GAF's wer.e computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. Which cf the following statements is true concerning reactor p6 war? (SELECT ONLY ONE ANSWER) (1.0)
- a. If the feedwater temperature used in the heat balance calculation was LOWER than actual feedwater temperature, then the actual power is HIGHER than the currently calculated power.
- b. If the reactor recirculation pump heat input used in the heat balance calculatien was OMITTED, then the actual power is LOWER than the currently calculated power.
- c. If the steam flow used in the heat balance calculation was LOWER than the actual steam ficw, than the actual power is LOVER than the currently calculated power.
- d. If the RWCU return temperature used in the heat balance calculatten was HIGHER than the actual RWCU return temp 6rature, then the actual power is LOWER than the currently calculated power.
(*** END OF SECTION 5 ***)
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s SECTION 6: PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 6.1 Consider an Off-Normal Event in which Instrument Air System pressure is lost.
- a. How will the following valves Fail? (CLOSED, OPEN, AS IS) (2.5)
- 2. RFP Minimum Flow Valve
chu2It 2 -Orywell Chillers Temperature Control Valves - 9
- 5. TBCW Make-up Valve
- b. EXPLAIN the cause of the potential High Radiation levels in the Of f-Gas Building. (0.5) 6.2 Describe the condition (s) which will generate EACH of the following indications on the operator control module.
- a. Channel disagree (0.5)
- b. Insert required (0.5) 6.3 The plant is operating at 100% RTP with Recirc Flow control in
" Flux Manual." An operator inaavertently DECREASES the " Pressure Reference Set" on the EHC Turbine Control System by 5 psig.
ASSUME: 1. No further operator action.
- 2. All other EHC control settings are normal.
- 3. Starting Parameters:
4
-TCVs (MSCV & LPSCVs) -100% Steam Flow Position
-BPCVs - 0% Steam Flow Position
-Rx Power -100% Rated Thermal Power
-Rx Pressure -1025 psig
-Load Demand / Load Limit -1310 MWe NOTES: All valve %s are in % Steam Flow Position.
See attached figures for information.
l Which of the following most accurately describes both the INITIAL
! RESPONSE and FINAL STATUS of the different parameters and components? (2.0)
(*** QUESTION 6.3 CONTINUES ON NEXT PAGE ***)
a b c d INITIAL RESPONSE
-TCVs l Partial l Partial lNo Change lNo Change IClose (<100%) lClose (<100% l l
-BPCVs lNo Change l Partial l Partial lOpen (>0%)
l lOpen (>0%) lOpen (>0%) l
-Rx Power l Increase lNo Change l Decrease l Decrease
-Rx Pressure l Increase JNo Change l Decrease l Decrease l l l l FINAL STATUS l l l l l l l l
-TCVs l$100% l Partial l0% (MSIV l$100%
l lClose (<100%) l SHUT) l
-BPCVs l0% l Partial l0% (MSIV l0%
l l0 pen (>0%) l SHUT) l
-Rx Power l>100% l>100% ls0% l<100%
-Rx Pressure l>1025 psig l>1025 psig lAs controlled l<1025 psig l l lby SRVs &RCIC l ONLY ONE ANSWER - READ ENTIRE COLUMN FOR BOTH INITIAL AND FINAL RESPONSES.
6.4 EXPLAIN the functioning of the Feedwater Control System "Setpoint Setdown Mode" feature from actuation to a reset condition. Ensure that your explanation addresses the following: (2.0)
- all applicable setpoint(s)
- specific effect(s)
- reset method (s) 6.5 Regarding the Control Rod Drive (CRD) and CRD Hydraulics
- a. Why is the hydraulic system Flow Control Valve mechanically blocked from going completely closed during a scram? (1.0)
- b. Scramming a CRD with the over piston flow path isolated (scram discharge valve closed or the area manual valve l
closed) will result in: (Select the best answer) (1.0) i
- 1. The CR0 staying at the position it was prior to the scram.
t 2. Extremely high pressures being generated in the over piston volumes.
- 3. Graphitor seal damage
- 4. High CRD temperature 2
6.6 The RHR-LPCI System has received a valid initiation signal. The system automatically initiated. The initiation signal is still present.
RHR-LPCI "A" flow is diverted to initiate Suppression Pool Cooling by use of the TEST RETURN LINE VALVE (F024A) MANUAL OVERRIDE function.
LIST the condition (s) that would defeat / inhibit this manual override signal to F024A. (1.0) 6.7 Consider the Recirc Pump Slow speed starting sequence logic depicted on the attached figure. List the nine (9) permissives that are left blank and lettered. (2.5) 6.8 The plant is operating at power with "A" and "C" CCW pumps running and the "B" CCW pump selected for STANDBY operation.
A Loss of Power occurs and the diesels start and tie in normally.
Which one of the following most accurately describes how the CCW system will respond during this transient? (1.0)
- b. Both the "A" and "C" CCW pumps can be started manually on ESF power after the buses are reenergized.
- c. The "B" CCW pump will not auto start, but can be manually started by the operator on ESF power after the bus is reenergized.
6.9 The post accident sampling system station located on the 737' elevation of the Diesel Generator building provides a central location for monitoring and grab sampling several fluid systems?
List three of the:e systems. (1.5) 6.10 SBLC System A is in a normal STANDBY lineup with one systematic deviation - the TEST TANK OUTLET VALVE (F031) is OPEN.
Which of the following most accurately describes the effects on the STORAGE TANK OUTLET VALVE (F001) and SBLC PUMP A of placing the SBLC Keylock Control Switch for Pump A to START. (1.0) i a. Valve F001 Opens - SBLC Pump A Starts after the valve reaches its Full Open position.
- b. Valve F001 Opens - SBLC Pump A Starts concurrently with the valve opening.
3
- c. Valve F001 does Not Open - SBLC Pump A Starts
- d. Valve F001 does Not Open - SBLC Pump A does Not Start 6.11 A diesel generator is the sole supply to an ESF Bus. When paralleling the Normal Power supply back to the Bus the synchroscope should be turning (a) in the (b) direction. (0.5) 6.12 For each of the RCIC (Reactor Core Isolation Cooling)
System component failures listed below, state whether or not RCIC will AUTO inject into the reactor vessel.
If it will not inject, state why, and if it will inject, provide one (1) potential adverse effect or consequence of system operation when the component is in the failed condition at the time RCIC receives the AUTO initiation signal. Consider each item separately.
- a. The Gland Seal Compressor fails to operate. (1,0)
- b. The minimum flow valve fails to AUTO open (stays shut) when system conditions require it to be open. (1.0)
- c. The RCIC pump discharge flow element output signal (to the RCIC flow controller) is failed at its maximum output. (1.0) 6.13 The reactor is operating at 80% of full power when a relief valve suddenly fails open. Recirculation flow control is in Master Manual.
- a. Will this result in a feed flow-steam flow mismatch? Explain. (1.0)
- b. What happens to MWe? Explain. (0.5)
- c. What is the initial response of reactor pressure? Explain. (0 5)
- d. Where will power end up relative to the power at the beginning of the transient? Explain. (1:0)
- e. How would your answer in part (d) differ if Recirculation flow control had been in Master-AUT0? Explain. (0.5) 6.14 While operating at 80% power, a Safety Relief Valve opens and remains open. List four (4) methods available to the operator to determine which SRV has opened. (2.0)
( *** END OF SECTION 6 ***)
4
L i'b-L J L Lisb L est.v A s,vi r.,uwe u CB.5 8' " I SYS. RR REV. O RR2-3 RTS3 1
DEPRESSED l LOW $ PEED SEQUENCE INITIATED ($EAL5 IN) lf INCOMPLETE y y y y SEOUENCE NOT -
OPERATED $ G Q If I CB 5 If I
,7 RACKED IN II gg, MOTOR TRIP g"~
CLOSE LOCKOUT CB-1 RELAY RESET CB-5 37 KVtFo60) 3p y
l IN MANUAL AT MINIMUM if CB-2 OPEN f
If PUMP SPEED SUCTION A NOT 8FTWEEN l DISCHARGE 20 AND 26% y y U VALVES ) 90% \
OPEN CB-2 GEN OUTPUT \
1 . b. CLOSED VOLTAGE NEAR/
Y RATED THERMAL SHOCx A r 5 y if y y if IN UMITS pggp CLOSE 10 SEC MOTOR TO I LOCKOUT RELAY WGH SPEED 4 CB-5 TIM 5 DELAY I I SEQUENCE RX POWER
( 30% AND y V 1 f NOT SYPASSED START do SEC RESET INCOMPLETE m NCOMPLETE y
- y SEQUENCE SEOUENCE TIMER TIMER Cs-1 A Cs-2 y TIMED RESET RACKED IN l OUT l 1p CLOSE CB 2 II LOW SPEED START SEOUENCE I
CB 5 NOT IN COMPLETE l "510P LOCK" i
U U TRIP CB 1 .
l AND CS-5 Low Speed Starting Sequence -
1
I SECTION 7: PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 7.1 A reactor SCRAM has occurred, but NOT all of the control rods have inserted to less than the 06 position. Reactor power is indicated as 3% on the APRM's. LIST the three (3) immediate operator action steps that are required per CPS 4404.01 " Reactor Scram." (1.5) 7.2 Assume that adequate core cooling CANNOT be maintained and
" Alternate Shutdown Cooling" must be established per 4403.01 DESCRIBE the RPV cooling water flowpath that should be established by this procedure. (1.0)
NOTE: INCLUDE IN YOUR DESCRIPTION THE SYSTEMS / COMPONENTS WHICH ARE USED.
7.3 Per 4403.01, "Cooldown Emergency," which one of the following most accurately describes how SRV operation should be used to control pressure, if needed? (1.0)
NOTE: ASSUME THAT THE INSTRUMENT AIR SYSTEM IS OPERATING F30PERLY
- a. Use numerous'SRV's, with short pressure reductions (* 50 psig) to equalize Suppression Pool heatup.
- c. Depressurize with a sustained SRV opening to maximize the emergency cooldown rate.
- d. Allow the SRV's to operate by mechanical actuation to ensure design pressure control and heat dispersion.
7.4 The Control Room is declared uninhabitable and evacuated. The immediate operator actions for " Remote Shutdown," are completed.
RCIC then ISOLATES. Level subsequently decreases to Level 2.
Restoration of level USING RCIC requires which of the following? (1.0)
ASSUME THAT THE CONDITIONS NEEDED FOR RESETTING AN ISOLATION,
" AUTOMATIC ISOLATION," HAVE BEEN MET.
- a. No Operator Action. RCIC will restart automatically.
- b. Operator Action. Close RCIC TURB FLO CONT in manual at minimum setting; Re-open RCIC TURB TRIP /THROT VLV and establish flow.
l 1
- c. Operator Action. Close RCIC TURB TRIP /THROT VLV; reset RCIC TURB TRIP logic; RCIC will now restart automatically.
- d. NONE OF THE ABOVE. RCIC cannot be restarted from the Remote Shutdown Panel after isolation.
7.5 " Reactor Recirculation," directs operator actions for an unexpected decrease in reactor coolant system flow rate.
FILL IN THE BLANKS (After the unexpected decrease), if both recirculation loops are still operating, transfer the FCV's to (a) .
Balance loop flows to within (b) at less than 70%
core flow, or to within (c) at greater than 70%
core flow. (1.5) 7.6 The unit is operating at 70% RTP; you notice power start to increase with NO CHANGE in recirculation flow or rod position.
You suspect a " Loss of Feedwater Heating." Which of the following is required / appropriate per CPS 4005.01 (1.0)
- a. A reduction in Recirc Flow, monitored by Recirc Flow indication.
A e */o
- b. A 30% Power Reduction, using Recirc Flow, monitored by APRM's.
- c. Insertion of Shallow Rods, to maintain proper flux shape, prior to reducing Recirc Flow.
- d. Insertion of Power Rods, to maintain proper flux shape, prior to reducing Recirc. Flow.
7.7 Procedures associated with operation of HPCS, LPCS, RHR and/or RCIC caution the operator to observe certain limitations on Suppression Pool Level and Temperature when operating these systems. (1.5)
- a. COMPLETE THE FOLLOWING:
Suppression Pool Level shall not be less than (1) .
Suppression Pool Temperature shall not exceed (2) during HPCS, LPCS, and/or RHR operation; it shall not exceed (3) during RCIC operation.
- b. STATE the basis for these temperature / level limitations on the Suppression Pool.
. 2
7.8 You are conducting a shutdown of the CRDH system, per 3304.01 you open drain valve 107 to drain the water accumulators. State the indication which should be used to determine that the accumulator is fully drained. (0.5) 7.9 Regarding the RHR Procedure, when operating in the shutdown cooling mode:
- a. You are cautioned to NOT allow reactor vessel level to decrease below 44 inches on the shutdown range. Why is this level of concern? (1.0)
- b. You are also cautioned to avoid opening the RHR test return line valve or the minimum flow bypass line valve.
Why must these valves remain closed. (1.0) 7.10 DESCRIBE the steps that must be performed in order to take the Main Generator from its normal operating status to a status where maintenance can be performed on the generator after a shutdown. (1.5)
NOTE: LIMIT YOUR RESPONSES TO THE GAS SYSTEMS REQUIRED TO EFFECT THE PURIFICATION. SET POINTS NOT REQUIRED.
7.11 FILL IN THE BLANKS; 1
Clinton Power Station Emergency Plan authorizes exposures to a MAXIMUM emergency dose to the whole body of (a) when taking measures to protect plant safety systems. Lifesaving actions which r.ay result in doses in excess of shall be (c) in nature and should not exceed (b)T-~
(d . (1.5) 7.12 List five (5) entry conditions for Containment Control-Emergency. (2.5) 7.13 A single MSIV closes and you determine that a high flow condition was reached in the other steam lines. Given that a Group 1 Isolation DID NOT OCCUR - STATE your Immediate Actions. (1.0) 7.14 Per the " Containment Combustible Gas Control," List two (2) conditions which require the operator to start the (1.5)
HYDROGEN IGNITERS.
3
7.15 Per "CRD Malfunction," if NO CRD Pumps are running and NO CR0 Pumps will restart:
- a. STATE WHEN immediate corrective action must be initiated. (0.5)
- b. STATE the Immediate Action (s) required. (0.5) 7.16 You are in the Reactor Heatup and pressurization phase of Procedure 3002.01. Reactor Pressure is 300 psig. Why should the use of a condensate booster pump at lowpeance be minimized? (1.0) 7.17 True or False 1 ch If all rod position indication is lost, rod insertion is p [##'* the only allowable rod motion. (i.e., Insert or Scram) (1.0) fs 7.18 Per procedure for Component Cooling Water System (CCW),
state the immediate actions required for a Partial loss of CCW. (2.0)
( *** END OF SECTION 7 *** )
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SECTION 8: ADMINISTRATIVE PROCEDURES, CONDITIONS AND LINITATIONS 8.1 FILL IN THE BLANK with one of the following TS terms:
"A shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips." (1.0)
- a. Channel Calibration
- b. Channel Check
- c. Channel Functional Test
- d. Logic System Functional Test 8.2 FILL IN THE BLANK FOR THE FOLLOWING:
In accordance with 10 CFR 55, "if a licensee has not been actively performing the functions of an operator or senior operator for a period of (1) months, or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operation and administration are satisfactory. (0.5) 8.3 Technical Specifications define SHUTDOWN MARGIN as. . .
" SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming. . . and the reactor is in the shutdown condition;. . ."
i LIST the plant conditions which complete the definition of SHUTDOWN MARGIN. (1.5) i 8.4 ADEQUATE CORE COOLING must be assured prior to securing an ECCS system that has automatically ini.tiated. LIST four (4) i plant conditions (per Level Control-Emergency) which will assure that Adequate Core Cooling exists. (2.0)
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l 8.5 During a Reactor Startup with the plant in Operational Condition 2, a Channel Functional Test on the E0C-RPT system is determined to be UNSATISFACTORY. The UNSAT condition affects no other TS systems.
- a. STATE whether it is allowable to enter Operational Condition 1. JUSTIFY your response. (1.0)
- b. DESCRIBE the physical phenomenon which necessitates the EOC-RPT system. (i.e., the Bases for E0C-RPT) (1.0) 8.6 The Unit is in Operational Condition 1, at 75% RTP, with one outstanding deficiency:
The Auto - swap of the HPCS suction upon receiving CST low level is determined to be UNSATISFACTORY. One channel of the swap-over logic is tripped, the suction is MANUALLY switched to the Suppression Pool, and the suction to the CST is IS0 LATED.
Which one of the following actions most correctly details the allowance and/or limitations imposed by the Technical Specifications in this instance? (1.0)
NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE
- a. . . .no new limitations or TS Operational Condition restrictions are initiated by this re-alignment.
- b. . . .be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. . . .be in at least HOT SHUTDOWN within six hours and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- d. . .be in at least HOT SHUTDOWN within six hours and reduce reactor steam dome pressure to less than or equal to W rpsig within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
/00 8.7 a. The following data was derived during a single day of operation at Operational Condition 1. The unit has been br in Operational Condition 1 for two weeks. Only FINAL DATA is presented; Preliminary data is not supplied. [,
SHIFTS i 00-04 04-08 08-12 b Floor Drain Leakage 2.52 gpm 4.58 gpm 3.75 gpm
(*** QUESTION 8.7 CONTINUES ON NEXT PAGE ***)
2
!. 0
Equipment Drain Leakage 20.91 gpm 20.58 gpm 21.00 gpa Total Leakage 23.43 gpm 25.16 gpm 24.75 gpm SHIFTS ,
12-16 16-20 20-24 Floor Drain Leakage 4.30 gpm 4.25 gpm 4.60 gpm Equipment Drain Leakage 22.25 gpm 24.33 gpm 19.33 gpm Total Leakage 26.55 gpm 28.58 gpm 23.93 gpm NOTE: THE DRYWELL LEAKAGE CALCULATIONS ARE THE TOTAL LEAKAGES WHICH WERE CALCULATED DURING THE INDICATED PERIODS. THUS, DAILY TOTALS WOULD BE ATTAINED BY ADDING THE 6 4-HOUR PERIOD TOTALS.
EVALUATE FOR EACH of the four (4) TS Leakage LC0 limits applicable in this plant condition whether the limit was exceeded, or not. (Disregard the Reactor Coolant System Pressure Isolation Valve Limit as defined in TS Table 3.4.3.2-1) (2.0)
- b. DEFINE " Pressure Boundary Leakage." (1.0) 8.8 With the Mode Switch locked in the Refuel position:
" CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless at least four associated Refuel position interlocks are OPERABLE for such equipment."
LIST three (3) Refuel Position Interlocks. (1.5) 8.9 All Fuel is removed from the core; however, Fuel Loading is scheduled to commence. TWO (2) Control Rods are removed from the core under the allowances of the Technical Specifications. /j, 0 Which of the following actions most accurately details the allowances and/or limitations imposed by the Technical Specifications in this instance?
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED FOR REFERENCE
- a. Fuel Loading may not commence until all Control Rods are inserted.
I
- b. Fuel Loading may commence and continue as long as the Shutdown Margin requirements of TS 3.1.1 are satisfied.
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- (*** QUESTION 8.9 CONTINUES ON NEXT PAGE ***)
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- c. Fuel Loading may commence - however the four fuel assemblies surrounding the removed Control Rods may not be loaded.
- d. Fuel Loading may commence AFTER one of the Control Rods is inserted. The four fuel assemblies surrounding the removed Control Rod may not be loaded.
8.10 Which of the following choices will correctly complete the blanks for the MCPR LCO listed below? (2.0)
The MCPR shall be equal to or (1) than (2)
MCPR(f) (3) MCPR(p) limits at indicated core flow and THERMAL POWER as shown in Figures 3.2.3-1 and 3.2.3-2.
NOTE: Figures 3.2.3-1 and 3.2.3-2 are enclosed for reference.
(1) (2) (3)
- a. greater; the smaller of the; or
- b. less; the larger of the; or
- c. greater; both; and
- d. less; both; and 8.11 The Unit is in COLD SHUTDOWN during a reactor startup with no outstanding deficiencies. Hydrogen Recombiner A becomes IN0P. It is anticipated that repairs will be complete within two (2) weeks.
Which of the following actions most accurately details the allowances and/or limitations imposed by the Technical Specifications in this instance? (1.0)
- a. Operational Condition 4 must be maintained (Entry into Operational Condition 5 is acceptable)
. b. Startup activities may continue; Operational Condition 3 may be entered, but not exceeded.
- c. Startup activities may continue; Operational Condition 2 may be entered, but not exceeded; 0xygen concentration shall be maintained < 2 v/o.
- d. Startup activities may continue; Operational Condition 1 and/or 2 may be entered, but the Recombiner must be returned to an OPERABLE status within 30 days.
NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE 4
30 t
(
8.12 The Technical Specification 3.4.4 established the following conductivity and chloride limits Plant Condition Conductivity Limit Chloride Limit 1 1 umho/cm 0.2 ppe.
2 and 3 2 umho/cm 0.1 ppa.
Per the Technical Specification basis, WHY is the chloride limit more restrictive at the lower steaming rate than at power? (1.0) 8.13 With the exception of breaker position, what THREE (3) items should an operator check on a breaker, if applicable during the performance of a system lineup checksheet per Control and Use of Operations Section Directives, 02-S-02-2?
Consider Local checks only, and a 4.16 KV I.T.E. Circuit Breaker as an example. (1.5) 8.14 The APRM Trip Setpoint Formula is ( 66W+48%)*T. Which of the following choices correctly details the definition of (f, o )
"T" AND when it is applied?
- a. T = FRTP/MFLPD; T applied if < 1.0
- b. T = MFLPD/FRTP; T applied if < 1.0
- c. T = FRTP/MFLPD; T applied if > 1.0
- d. T = MFLPD/FRTP: T applied if > 1.0 S.15 Per the Technical Specifications, COMPLETE THE FOLLOWING TABLE: (2.0)
MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 SS (a) (f)
SR0 (b) (g)
R0 (c) (h)
A0 (d) (i)
STA (e) (j)
.f
[
8.16 In OPERATIONAL CONDITION 1 or 2 a reactor water isotonic analysis foriodineisrequiredwhentheoffgaslevelatthe{alincreases by more than b in Oc during steady state operation at release rates {dlthan e. (Fill in the blanks.) (2.5) 8.17 Concerning Radiation Work Permits (RWP)
- a. Whose permission is required to commence work covered by an RWP? (0.5)
- b. What information is an individual required to enter on the RWP Access Log when entering / exiting the job site? (2.0)
( *** END OF SECTION 8 *** )
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. 2W 8 37894 0 017123 18 07S 11 082 23283 9359 1168 7 r 3876 12933 1 6404 2e4 8 288 8 40 $00 0 017157 10 358 10 375 236 91 9338 1110 0 0 3932 17823 16'M 268 8 272 0 43249 0 017193 9 738 9 75S 240 99 9303 1178 3 0 398' 11715 36702 272 0 278J 46147 0 017228 9 362 9 180 24508 921 5 1872 5 04043 IJ607 IHS 278 4 200 8 49 200 0 017264 4 627 S ka 24917 924 6 l173 8 0 8000 12Scl I H99 Ise t fu t $2414 0 01730 81280 0 3453 253 3 921 7 350 04154 12nS !6W8 isa s 2se t $$ 795 0 01734 7MM 7 68C 7 2574 til 8 176 2 0 4208 812% 3 Htt 2s4 e FIF 8 $9 M0 0 01738 72301 7247S 26t l 915 9 1774 04263 IJiu 1 64'9 297 8 tes t U 064 0 0174l 6 82S9 6 8433 2616 913 0 .178 6 04317 11082 1640: 296 8 3e0 8 67005 0 01745 6 4443 6 4658 269 7 910 0 .179 7 344 8 fl lit 0 01749 60%S 611$0 273 8 0 8377 1979 18351 3et t 300 1 71433 907 0 180 9 0 4426 1877 1 6303 0 C1753 5 76 % 5 7830 278 3 384 8 312 0 79 913 0 01757 $4%6 $ 4 '42 282 1 9040.' it3 l 9010 182 0 04479 1 776 162% 3es s 316 I M 688 0 01763 5 1673 5IM9 2063 1,184 1 0 4S13 IU6 3620S 3:23 097 9 0 GS86 . 8516 I 6162 318 0 3MI 09 W3 0 01766 4 8961 324 0 98826 0 03770 4 His 49138 290 4 894 8 lits 2 04W0 1 1477 3 6116 379 8 325 8 4 6595 29s6 893 6 1886 2 0 4492 l 1378 16c71 100 245 0 01774 4 4030 44208 324 8 332 8 105 9CF 0 01779 4 3788 298 7 088 5 11872 04745 ;1230 160*$ 328 8 41964 102 9 885 3 1888 2 04798 136 8 111 820 0017H 3 M81 3 9tSt 8113 I S9s t 132 8 3078 882 3 !!89 1 0 4850 ll 1086 8 5936 334 8 DI,' . Het 117 992 0 01787 3 7699 37878 311 3
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18 i Table 2: Saturated Steamt Pressure Table M .
58t. Sat 58t. 58t 5 81 Aes Press Aes Pres 8 fes, 58t.
vapor bewe Evap LHlwa Ev80 V8per th;50 in LD.50 in F8su b0ed Esso V40er 8g p I 'g hg hg 8g 8,g i f 't 'eg h eg 0 0000 2 1872 1872 esses 6 lasell 32 0t 8 0 016022 3102 4 3302 4 0 0003 :0M S 1075 5 0%7 en 8 21 59 323 0 016032 4235 5 42JS S 27382 060 1 10s7 4 0 0S42 2 0425 t he 19 Se6 0 016071 641 5 641 5 47623 048 6 0% 3 0 09?S .9446 2 0370 SM 18 108 74 0 016136 333 % 333 60 69 73 '036 1 05 8 0 1326 84 % i t18 f la 73 532 130 20 000 9 31 4 0 2349 6094 8443 10 SO 16224 0 016407 73 515 set 19321 34 420 942 8 43 3 O tSM $043 7879 10 0 0 016592 38 404 1612{ 4447 7%8 14 ess 14 8e8 212 00 0 016719 26 782 26 799 100 l r 970 3 SO S 0 3121 IS O 213 0J 0 016726 26274 26 290 18121 969 7 .150 9 0 3137 .4415 1 7552 118 set 22796 0 016834 20 070 20 087 196 27 960 1 11 % 3 0 33 % 1 3962 : 7320 NO 3e 8 2 % 34 0 017009 13 7266 13 74M 218 9 945 2 1164 1 O M42 3 3313 '.6995 att 10 496S 236 1 933 6 1:69 8 0 3921 12844 16765 40 8 40 0 26725 0 017151 10 4794 le t 281 02 0 017274 84967 0 $140 250 2 923 9 1874 1 0 4112 l2474 , 6586 tas
!!?7 6 0 4273 2167 ,6440 est 88 8 292 78 0 017383 Pl%2 717M 262 2 915 4 61s TS 6 20$0 272 7 9078 1100 6 0 4411 1905 1 6316 38 0 75 0 302 93 0 017442 54SM $4711 282 1 900 9 1183 1 0 4S 34 .167$ l 6208 a0 0 set 312 04 0 017573 na ..
894 6 1185 3 0 # 43 1 1470 l 6113 g* g g 98 0 32028 0 0176S9 4 4779 4 8953 290 7 1m 8 P-1m 8 32782 0 017740 4 4133 4 4310 298 S 884 6 11872 0 4743 1 1284 . 6027 -4~
4 0484 305 8 883 1 1180 9 0 4834 1 1115 ; 5950 tit t tisa 334 19 0 01782 4 0306 5879 tMO 178 8 34127 0 01789 3 7097 3 7275 312 6 877 8 !!90 4 0 4919 1 0960 130 0 34733 0 01796 34M4 3 4544 319 0 872 8 1198 7 0 4998 1 0815 $813 128 3S3 04 00L803 3 2010 3 2190 325 0 068 0 llSJ O 0 5071 10681 .5752 14e 8 184 8 2 9958 3 0139 330 6 063 4 1194 1 0 5141 1 0554 '.5695 IMI lH e 358 43 0 01809 16e e M3 SS 0 01815 2 81 % 2 4336 336 i 859 0 lits 1 0 $206 104n i % 41 las 0 368 42 0 01821 265% 2 6738 341 2 854 8 11 % 0 05269 1 0322 1 % 91 lie s 178 8 373 08 0 0:827 2 S L29 2 $312 346 2 850 7 11 % 9 0 5328 10215 I S$43 IN 8 IM O 198 8 377 53 0 01833 2 3847 2 4030 350 9 846 7 1197 6 0 1384 10!!3 1 5498 tes t 331 80 0 01139 2 2689 2 28?) 3%S 842 9 lit 0 3 05438 1 0016 I S4Se fee t 298 8 tra t 38591 0 01844 216373 2 18217 359 9 819 1 !!99 0 0 5490 0 9923 1 5413 218 e 389 88 0 018 % 2 06779 2 08629 3642 E35 4 199 6 0 % 40 0 9834 8 1374 228 e 229 0 39170 0 018 % .97991 99846 368 J 831 8 .200 1 0 % 84 0 9748 $3M 2M B 23s 3 248 8 39739 0 01860 89909 91769 3723 428 4 200 6 0 % 34 09%$ $299 240 0 40097 0 0186S 824S2 84317 376 1 825 0 201 1 0 % 79 09S45 1264 250 0 258 8 404 44 0 018'O ?%48 .77410 379 9 821 6 201 % 0 S722 0 9508 5230 298 8 Fee s 27s 0 407 80 0 018?S 69137 71013 383 4 818 3 201 9 0 5'64 09433 Sill 278 1
!se t 41107 0 01880 63169 6%A9 3871 815 1 202 3 0 5405 09361 SIM 201 0 414 2S 0 01885 57597 .59482 390 6 812 0 .202 6 0 5644 0 9291 . SIM 19s a 2ss t 41735 0 01889 I S2384 1 54274 394 0 009 9 1702 9 0 5882 0 9223 I SIOS 3es t 3es t ne0 431 73 0 Otti2 3 30b42 4 32 % 4 409 8 794 2 1204 0 0 6059 0 1909 84%8 35e 0 eso s 444 60 0 01934 1 14162 1160n 424 2 780 4 12046 0 6217 0 86J0 14M7 esal 4WO 4$6 28 0 019 % 1 01224 1 03179 4373 7875 1204 8 0 6360 0 8370 .4738 4M O H99 44701 0 019'S 0 90787 0 92762 449 5 MSI 1204 7 0 6490 0 8148 .4639 See s
$H O 476 94 0 01994 0 82:93 0 841?7 4,09 7433 1204 3 Gull 0 79 % -.4547 $$t t 600 I 486 20 0 02013 014h2 0 16975 4FI 7 732 0 1203 7 0 6723 0 7738 4461 tes 0 SWO 494 89 0 02032 0 68811 0 70843 481 9 ??c 9 1202 8 0 6829 0 75$2 4381 8te t FN 0 S0J 04 0 020$0 063S05 0 6M% 491 6 710 2 12018 0 69?S 0 7377 6 4J04 FN 0 750 0 $10 84 0 02069 0 58880 0 60949 500 9 699 8 1200 7 0 7022 0 7210 .4232 750 0 tes t $14 21 0 02087 0 54809 0 %s% S09 8 6896 1199 4 0 1111 0 7051 L 4163 ges 8 850 0 S2124 0 0210S 0 $1197 0 53302 518 4 679 5 1198 0 0 7197 0 68*9 40 % SSe s 988 I S3a 95 0 02123 0 41968 0 $0091 $26 7 669 7 11 % 4 0 1279 .4032 0 6753 ses 0 934 6 538 39 0 02148 0 45064 0 47205 , 534 7 660 0 '194 7 0 7358 OH12 3970 ms8 1000 0 $44 $8 0 02159 0424M 044%6 542 6 650 4 1929 0 1414 0 6476 3910 18es t 10$0 8 MO S3 0 02177 0 40047 0 47224 Mal 640 9 .191 0 0 7S07 0 6344 3858 inu t
- lite t 5 % 28 0 02195 0 37863 0 40054 557 5 631 5 189 1 0 7578 0 6216 : J 794 tint 11H g % I82 0 02214 0 358 % 0 38073 %48 622 2 ,1870 0 7647 0 6091 ..3734 Itu t N#9, 12N O % 719 0 02232 0 34013 0 3624S 571 9 613 0 1184 8 0 7734 0 $%9' L M83 1298 8 - -
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12S8 8 S12 38 0 02250 0 32306 034%6 578 8 6038 ilt? 6 0 7700 0 5850 MJo 123e 0 1300 0 57742 0 02269 0 30722 0 32991 58S 6 $94 6 ;180 2 0 7843 0 1733 3577 IMal itse t 582 32 0 02288 0 292 % 0 31537 592 3 585 4 l177 8 0 7906 0 % 20 3S25 13see 14e0 0 58707 0 02307 027871 0 30178 $98 8 576 S .175 3 0 79M 0 % 07 , 3474 1400 0 14M 8 591 70 0 02327 0 26 % 4 0 28911 60$3 $67 4 .172 8 O so?6 0 $397 .3423 Ism e 1900 0 $96 20 0 02346 0 ?$372 0 ??719 611 7 558 4 1170 1 0 8085 O S288 3373 tie t ISW 8 600 % 0 023M 0 24235 0 2 % 01 610 0 5494 l167 4 0 8l42 0 5182 3324 ttee t 1400 0 604 87 0 02387 0 23159 0 2M45 624 2 $40 3 164 5 0 8199 05076 3274 tees t 46400 609 05 0 02407 0 22843 0 24 % l 6304 $34 3 161 4 0 8254 lies 8 0 4973 4 322S tau 8 613 43 0 02424 021178 02M07 6MS S22 2 158 6 0 8Jol 0 4867 1 3876 Iles 8 1798 0 61712 0 02450 $20263 0 22713 642 5 5131 55 6 08M3 0 4765 31?8 1718 0 tese t 621 02 0 02472 0 19390 0 1861 648 5 503 8 S? 3 Otu?
law e 624 SJ 0 8417 3079 tesel g 0249 654 5 494 6 49 0 0 6470 04%I 3030 itse a stes t 428 % e 02Sl)r 008%8 .7764 0 0 10S20278 pos 485 2 45 6 0 8522 0 44 % 2981 toas t 19M 0 632 22 0 02541 0 6999 0 9540 6% ) 475 8 .42 0 0 8574 0 43S8 2931 tene t 20e0 0 635 80 0 02 % S 0 62H 0 8834 672 1 46 .38 3 0 86'S 0 42 % 2881 2 esse 21m g 642 76 0 02615 0 '4885 0 7501 683 8 446 {, 130 5 0 8127 0 405) 2180 2im e 2798 8 649 4S 0 02 % 9 0 .Mel 0 6212 695 5 426 7 122 2 08s?8 0 n48 2676 21ese 23se t 6 % 89 0 02727 0.2406 0 l$133 707 2 406 0 l13 2 0 8929 0 440 2%9 23es t lees t H231 0 02790 0 11287 0 .4076 719 0 384 8 l103 7 0 9031 0 3430 2 40 famt 2980 0 M411 0 028 % 0 10209 0 13064 731 7 M16 1093 3 0 9139 0 3206 l ?345 29e0 8 2eGe 8 67391 0 07938 0 09172 0 12110 744 5 3376 1082 0 0 9?47 0 2977 L 22?$ 2tes 8 270B 0 679 53 0 03029 0 08165 0 11194 7573 312 3 1069 7 0 9J% 0 2141 R 2097 27m t fles t 684 96 0 01134 007171 0 10305 7707 28S 1 10 % 8 0 6468 0 2491 19 % 2 ass a 2900 0 690 22 0 03262 0 06158 0 094/0 785 8 254 7 1039 8 09%8 0 2215 1803 !ses t 3eas e 69S 33 0 03428 0 05073 0 08500 801 8 218 4 1020 3 0 9728 0 1891 1619 3een s lies t 70028 O C?681 0 03771 0 074S2 424 0 1693 993 3 0 99 L4 0 1460 il373 Stes t 1700 0 705 08 0 04477 00a91 00%61 875 5 %I 931 6 1 0358 0 0482 : 0832 37es t 3200 l' 70$ 47 0 0$078 0 00000 0 0$028 906 0 00 906 0 106 L2 0 0000 1 0612 82ss 2*
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STEAM TABLE PHOPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) volume, ft'M Enthaley. Stum Entropy, Stum a F Y water Evan Steam water Evep Steam water "Eveo Steam Y hp, hg sg ' sq sg
't 'M 's he 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 32 , 0.08859 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 35 0.09991 0.01602 2948 2948 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 40 0.12163 0.01602 2446 2446 13.04 1068.1 1081.2 0.0262 2.1164 2.1426 45 45 0.14744 0.01602 2037.7 2037.8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 50 0.17796 0.01602 1704.8 1704.8 28.06 1059.7 1087.7 0.0555 2.0391 2.0946 60 60 0.2561 0.01603 1207.6 1207.6 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 70 0.3629 0.01605 868.3 30 633.3 633.3 48.04 1048.4 1096.4 0.0932 1.9426 2.M52 30 0.5068 0.01607 2.0086 90 468.1 468.1 58.02 1042.7 1100.8 0.1115 1.8970 90 0.6981 0.01610 1.8530 1.9825 100 350.4 350.4 68.00 1037.1 1105.1- 0.1295 100 0.9492 0.01613 0.1472 1.8105 1.9577 110 0.01617 265.4 265.4 77.98 1031.4 1109.3 110 1.2750 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 120 1.6927 0.01620 203.25 97.96 1019.8 1117.8 0.1817 1.7295 ~ 1.9112 130 130 2.2230 0.01625 157.32 157.33 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 140 2.8892 0.01629 122.98 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 150 3.718 0.01634 97.05 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 160 4.741 0.01640 77.27 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 170 5.993 0.01645 62.04 62.06 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 1.8111 leo 150 7.511 0.01651 1.5148 1.7934 190 40.94 40.96 158.04 984.1 1142.1 0.2787 190 9.340 0.01657 1.4824 1.7764 200 33.62 33.64 168.09 977.9 1146.0 0.2940 200 11.526 0.01664 1.7600 210 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 210 14.123 0.01671 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 212 14 696 0.01672 26.78 26.80 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 220 17.186 0.01678 23.13 23.15 19.381 198.33 958.7 1157.1 0.3388 13902 1.7290 230 230 20.779 0.01685 19.364 240 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 s 240 24.968 0.01693 16.304 250 13.819 218.59 945.4 1164.0 0.3677 13323 1.7000 250 29.825 0.01701 13.802 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 250 35.427 0.01709 11.745 11.762 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 270 41.856 0.01718 10.042 10.060 249.17 924.6 1173E 0.4098 1.2501 1.6599 280 280 49.200 0.01726 8.527 8.644 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 290 57.550 0.01736 7.443 7.460 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300
. 300 67.005 0.01745 6.448 6.466 1
280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 5.609 5.626 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 89.64 0.01766 4.896 4.914 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 340 117.99 0.01787 3.770 3.788 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 360 153.01 0.01811 2.939 2.957 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 380 195.73 0.01836 2.317 2.335 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.8444 1.8630 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 420 308.78 0.01894 1.4808 1.4997 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 440 381.54 0.01926 1.1976 1.2169 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 460 466.9 0.0196 0.9746 0.9942 464.5 739.6 1204.1 0.6648 0.7871 1.4518 480 480 566.2 0.0200 0.7972 0.8172 7143 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 0.6545 0.6749 487.9 l 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 520 812.5 0.0209 0.5386 0.5596 536.8 657.5 1194.3 0.7378 0.6677 1.3954 540 540 962.8 0.0215 0.4437 0.4651 562.4 6253 1187.7 0.7625 0.6132 13757 560 560 1133.4 0.0221 0.3651 0.3871 l 589.9 1179.0 0.7876 0.5673 1.3550 580 540 1326.2 0.0228 0.2994 0J222 589.1 l
550.6 1167.7 0.8134 0.5196 1.3330 600 600 1543.2 0.0236 0.2438 0.2675 617.1 5063 1153.2 0.8403 0.4689 13092 620 l .
620 0.0247 0.1962 0.2208 646.9 r 1786.9 0.8686 0.4134 1.2821 640 640 0.0260 0.1543 0.1802 679.1 454.6 1133.7 2059.9 1107.0 0.8995 0.3502 1.2498 660 640 0.0277 0.1166 0.1443 714.9 392.1 2365.7 310.1 1068.5 0.9365 0.2720 1.2086 640 640 0.0304 0.0808 0.1112 758.5 2708.6 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 700 3094.3 0.0366 0.0386 0.0752 906.0 0 906.0 1.0612 0 1.0612 705.5 705.5 3208.2 0.0508 0 0.0$08
--- _ , , - . . - - . - - - ,-----e
- -s
. = '
- f. -
EQUATION SHEET
~
f = ma y = s/t Cycle efficiency = (Network out)/(Energy in) w = mg s = Vg t + 1/2 at2 2
E = mc
~
KE = 1/2 mv a = (Vf - Vo )/t A = AN A = Age PE = mgh
'Vf=V o + at w = e /t x = an2/t1/2 = 0.693/tjjp t
NPSH = Pin - Psat 1/2 eff = [(t1/2)(t)3 b
[(t1/2) + (tb))
m a p AV AE = 931 am I=Ieo -Ex Q = [nCpa t Q = UAah I = I g e~"*
Pwr = Wfah I=I g 10-x/TVL TVL = 1.3/u 5
P=P o l0 "'I ) HVL = -0.693/u t
P = Po e /T SUR = 26.06/T SCR = S/(1 - Keff)
CR x = S/(1 - Kgffx)
SUR = 26p/t* + (8 - p)T CR)(1 - Keffl) = CRp(1 kdf2)
T = (1*/p ) + [(8 p )/Ap )
M = 1/(1 - Keff) = CR /CR j o T = t/(p -8) M = (1 - Keffo)/(1 - Keffj)
T = (8 - p )/(Ap ) SDM = (1 - Keff)/Keff p = (Kgff-l)/Kgf f = AKgff /K gff 1* = 10-5 seconds .
1 = 0.1 seconds-I p = [(1*/(T Kgf f)) + [8 eff /(1 + AT))
=Id 2p P = (z,V)/(3 x 1010) I)d)d 2 =Id Ij j p2 zs oN R/hr = (0.5 CE)/d (meters)
NPhi=Statichead-h -P sat R/hr = 6 CE/d2(feet).
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010dps 1 ga; . = 3.78 liters 1 kg = 2.21 lbm 1 f te = 7.48 gal. I hp = 2.54 x 103 Btu /hr Density = 62.4 lbq1/ft 3 1 mw = 3.41 x 106 Btu /hr Density = 1 gm/cv lin = 2.54 cm Heat of vaporization = 970 Btu /lbm *F = 9/5*C + 32 Heat of fusion = 144 Btu /lbm *C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
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,3/4.5 EMERGENCY CORE COOLING SYSTE_MS-3/4.5.1 ECCS - OPERATING I-
- LIMITING CONDITION FOR OF,EUTION I
I 3.5.1 ECCS divisions 1, 2 and 3 shall be CPERA6LE with: .
- 4. ECC3 division 1 consisting of:
'1. The CPERAELE low pressure core spray (LPCS) system with a flew path I capable of taking suction from the supprsssion pool and transferring the wetar lhrough the spray sparger to the reacter vessel.
- 2. The CPERABLE Icw pressure cociant injection (LPCI) subsystem "A" of the RHR systba with a ficw path capaole of taking suction frem the .
suppression pcol and transferring the water to the rea:t:r vessel. .
i 3, 7 OPERAELE ADS valves. .
- t. ECCS division 2 censtating of:
- 1. The OFERAELE low pressure coclant injection (LPCI) subsysters "B" and "C" of the AMA system, each with a ficw path capable of taking j suction from the suppressica pool and transferring the water to the reactor vesse1. *
) 2. 7 OPERA 3LE ADS valves.
- c. ECCS division 3 censisting of the CPERABLE his;h pressure core spray (HPCS) systes with a flow path capable of taking suction from the suppression .
pool and transferring the water thecugh the spray sparger to tha res:ter
. vessel. '
t .
I
! APPLICABILITY: OPERATIONAL CONDITICN 1, 2**# and 3*. *
~
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i "The AD5 is not required to be CPERABLE when reactor steam dose pressure is -
}
1ess than or equal to 100 psig.
- 5ee Special Test Exception 3.10.5.
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i 3 -
I CLINTON - UNIT 1 3/4 5-1 UA 191555 I,'
i
1 EMERGENCY CORE COOLING SYSTEMS D i LIMITING CONDITION FOR OPERATION ACTION:
- 1. With the LPCS system inoperable, rastore the inoperable LPCS system to OPERABLE status within 7 days. ,
1
- 2. With LPCI subsystes "A" inoperable, restore the inoperable LFCI sub-
. system "A" to OPERA 8LE status within 7 days.
- 3. With the LPCS system inoperable and LPCI subsysten "A" inopera::le, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ,
'4. Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. !
b' . For ECCS division 2, provided that ECCS divisions 1 and 3 are OPERABLE: ,
- 1. With either LPCI subsystem "B" or "C" inoperable, restore the ineper-able LPCI subsystem "B" or "C" to OPERABLE status within 7 days.
- 2. With both LPCI subsystems "B" and "C" inoperable, restore at least ,
the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ,
- 3. Othe mise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />".
- 1) With ECCS division 3 inoperable, restore the inoperable division to
- CPERA8LE status within 14 days.
- 2) Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and '
in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
r .
i "whenever two or acre RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor ccolant temperature as low as practical by use of alternate heat removal methods.
^
CLINTON - UNIT 1 3/4 5-2
~
AFA 1 S IS$1
..._. . .L . . :. . . .
EMERGENCY CORE COOLING $YSTEMS
./
LIMITING CONDITION FOR OPERATION (Continued)
! ACTION: (Continued) . ,
i i 1) , With LPCI subsystes "A" and either LPCI subsystem "B" or "C" inoper-
- able, restore at least the insperable LPCI subsystem "A" or inoper-i able LPCI subsystas "B" or "C" to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 2) With the LPCS system inoperable and either LPCI subsystems "B" or "C" inoperable, reste.e at least the inoperable LPCS system or inoperable
, . LPCI subsystas "3" or "C" to CPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. -
- 3) Otherwise, be in at laast HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and '
w in COLD SHUTDOWN within the folicwing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> **.
. divisions 1 and 2 are otherwise OPERABLE:
- 1. With ene of the above required ADS valves inoperable, restore the inoperable ADS valve tc OPERABLE status within 14 days or be in at 1 least HCiT SHJTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam does pressere to { 100 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i
- 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dcoe pressureto{ICOpsigwithinthenext24 hours.
- f. In the event an ECCS system is actuated and injects water into the Reactor i Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles i to date. The current value of the useage factor for each affected safety
! in ection ne ule shall be provided in this Special Report whenever its ,
' { va ue excaeds 0.70.
I i !
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.]- . "%nenever two or more RHR subsystems are inoperable, if unable to attain COLO
'. SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as
'.j I low as practical by us1 of alternate heat removal methods.
') . .. . .
CLINTON - UNIT 1 3/4 5-3 APR,1gggg3
-- m m. -
EMERGENCY CORE COOLING SYSTEMS ,
SURVEILLANCE REOUIREMENTS 4.5.1 ECCS division 1, 2 and 3 shall be demonstrated OPERABLE by:
- 1. Verifying by venting at the high point vents that the systes piping from the pump discharge valve to the system isolation valve is filled with water.
- 2. Verifing that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. Verifing that (when tested pursuant to Specification 4.0.5) (at least once per 92 days when tested each) each: .
- 1. LPCS pump develops a flow of at least 5010 gpm against a test line pressure greater than or equal to (119) psid.
- 2. LPCI pump develops a flow of at least 5050 gpm against a test line pressure greater than or equal to (119) psid.
- 3. HPCS pump develops a flow of at least 5010 gpm against a test line pressure greater than er equal to (490) psid.
- c. For the LOCS, LPCI and HPCS systems, at least once per 18 months performing a system functional test which inc.ludes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.
Actual injection of coolant into the reactor vessel may be excluded from this test. -
- d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the RCIC storage tank to the suppression pool
- on a RCIC storage tank low water level signal and on a .
suppression pool high watar level signal,
- e. At least once per 18 months for the ADS by:
- 1. Performing a system functional test which includes simulated auto-natic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
, CLINTON - UNIT 1 3/4 5-4 APR 1 s le5S
EMERGENCY CORE C00 LING $YSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2. Manually opening each ADS valve when the reactor steam done pressure is greater than or equal to 100 psiga and observing that either:
- a. The control valve or bypass valve position responds accordingly, or
- b. There is a corresponding change in the sensured stream flow.
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I, "The provisions of Specification 4.0.4 are not applicable provided the surveil-I
', , lance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate l j to perform the test.
, ~.
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CLINTON - UNIT 1 3/4 5-5 l
l . AFR 1 s ts;5 l \
L _
d EMERGENCY CORE COOLING SYSTEMS .
. 3/4 5.2 ECCS - SHt1TDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:
- a. The low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel,
~
- b. Low pressure coolant injection (LPCI) subsystes "A" of the RHR system with a flow path capable of taking suction from the suppression pool and trans-ferring the water to the reactor vessel.
- c. Low pressure coolant injection (LPCI) subsystem "B" of the RHR systes with .
a flow path capable of taking suction from the suppression pool and trans-ferring the water to the reactor vessel. .
- d. Low pressure coolant injection (LPCI) subsystee "C" of the RHR systes with a flow path capable of taking suction from the suppression pool and trans-ferring the water to the reactor vessel.
, e. The high pressure core spray (HPCS) system with a flow path capable of 1 taking suction from the RCIC storage tank or the suppression pool and transferring the water through the spray sparger to the reactor vessel.
l APPLICABILITY: OPERATIONAL CONDITION 4 and 5*.
ACTION:
- a. Withoneofk.heaboverequiredsubsystems/systemsinoperable,restoreat least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suspend i all operations that have a potential for draining the reactor vessel, '
i i
- b. With both of the above required subsystems / systems inoperable, suspend ,
t CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem / system to OPERA 8LE status ,
i within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish PRIMARY CONTAINMENT INTEGRITY within the !
next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
I j -
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"The ECC5 is not required to be'0PE LE provided that the reactor vessel head ,
is removed, the cavity is flooded, the reactor vessel to steam dryer pool gates are remove and water level s maintained within the limits of Speciff-cation 3.9.8 and .9.9. -
CLINTON - talIT 1 3/4 5-8 t .
-, . -. . . . - - _ , , - ___ ..-----.-.,----n., . _ - - . . . - - - , , -----,--.-- . ,.- .--,,,,- - n-- ,n.- --
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.
4.5.2.2 The HPCS system shall be determined CPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the RCIC storage tank required volume when the RCIC storage tank is required to be OPERABLE per Specification 3.5.2.e.
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- CLINTCH - UNIT 1 3/4 5-7 l AFR 1 3 1555 m --- .. :- --
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OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE.
- b. 5gpaUNIDENTIFIEDLEAKAGE.
- c. 25 gpa* total leakage (averaged over any 24-hour period).
- d. I gpm leakage at a reac+ar coolant system pressure of 1000 t
+ 10 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.? 1. .
- e. 2 gpm increase in UNIDENTIFIED LEAXAGE within any 4-hour period.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within ~
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3/ b.
With any reactor coolant system leakage greater than the Ifmits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. With any reactor coolant system pressure isolaticn valve leakage greater than the above limit, isolate tne high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least i two other closed manual or deactivated automatic valves, or be in at least ICT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the h 1 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. , ,
i d. With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than I 2 gpa within any 4-hour period, identify the source of leakage increase as
' not service sensitive Type 304 or 316 austenitic stainless steel within
. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in
- COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,
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. CLINTON - UNIT 1 3/4 4-8 ApA15 y h ,me . - im a me + e -i , . e
REFUELING OPERATIONS 3/4.9.10 CONTROL R00 REMOVAL -
SINGLE CONTROL R00 REPSVAL
. LIMITING CONDIT10N FOR OPERATION 3.9.10.1 One centrol rod and/or the associated control rod drive mechanise may be removerJ from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied.untti a control rod and associ-ated control rod drive sechanism are reinstalled and the control rod is fully inserted in the core. - j
- a. The reactor mode switch is OPERABLE and lockt1 in the Shutdown position
\
or in the Refuel posit, ion per Table 1.2 and Specification 3.9.1.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed;
- 1. May be assumed to be the highest worth control rod required to be assumed to be, fully withdrawn by the SHUTDOWN MARGIN test, and
. 2. Need not be assumed to be immovable or untrippable. '
- d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disamed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be reesved from the core and/or reactor vessel are removed from the core cell.
- e. All other control rods are inserted.
APPLICABILITY: OPERATIONAL CONDITION 4 and 3. ..
ACTION:
i With the requirements of the above specification not satisfied, suspend removal i of the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vesaal and initiata action to satisfy the above j requirements. -
- I' b +
CLINTON - (MIT 1 3/4 9-14 g yg g
/ '
REFUELING CPERATIONS
$URVEILLANCE REQUIREMENTS 4.9.10.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of removal of a control rod and/or the associated centrol rod drive mechanism from the core and/or reactor pres-sure vessel and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter until a control rod and associated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that: ,
- s. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1. -
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied per Specification 3.9.10.1.c. -
- d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control red drive mechanism to be removed from the core and/or reactor vessel are -
removed from the core cell.
- e. All other control rods are inserted.
3 CLINTON - UNIT 1 3/4 9-15 AFR La ISP.5
REFUELING OPERATIONS MULTIPLE CONTROL R00 REMOVAL .)
LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.
- a. The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Specification 3.9.1, except that the Refuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
- d. All other control rods are either inserted or have the surrounding four e fuel assemblies removed from the core cell.
- e. The four fuel assemblies surrounding each control rod or control rod drive $
mechanism to be removed from the core and/or reactor vessel are removed from the core cell. -
APPLICABILITY: OPERATIONAL CONDITION 5.
ACTION: .
With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
e e
e e
t .
i l
CLINTON - UNIT 1 3/4 9-16 APR.19 585 e
t -
l t
o REFUELING OPERATIONS i
SURVEILLANCE REQUIREMENTS l l
1 4.9.10.2.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of removal of control rods and/or.
control rod drive mechanises from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter until all control rods and control rod drive mecha'nisms are reinstalled and all control rods are inserted in the core, I verify that: '
. a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 <
or 4.9.1.2, as applicable, and incked in the Shutdown position or in the l Refuel position per Specification 3.9.1.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
- d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
- e. The four fuel assemblies surrounding each cent.rol rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from,the core cell. .
- f. All fuel loading operations are suspended unless all control rods are inserted in the core.
4.9.10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one rod-out" Refuel position interlock, if this function had been bypassed.
G CLINTON - UNIT 1 3/4 9- 17 APR 10 E35
4 1.7 -
4 1.6 -
1.5 -
c.'
a .
W
@ 1.4 =
c
6 w
- C 1.3 -
1
1.2 ~
l g
" OtfAC?M=t.13 i.
!, t .* =
l1 lt
- i. .
!)
, , , a '
a 1.o 100 ~ 120 m so 80 o 20 i
J l CORE FLOW (% of Parec)
- j t
t Cinton f.!CDRf (K() Vers::s Care Flow Figun 3.2.3-1 i
I. '
- 3/4 2-7 ._
- , --_ - --- - - _ _ _ ..n T e n -i.o. u r . _ _ _ , _
A 13 -
d
~
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1.5 S.
- E .
e.
c a t.4 -
c *
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. m 1.3 -
a f u -
b .
2 1.1 -
11
- l. , t ' '
18 sc too 1:
.~.
6 3 e so .
~4 c::RE Pcy.tM t% of Retso) l' I
l
- Cin:cn MCPRp (Xp) Ver.~.:s P:wer Figure 2.2.2-2 l <
l
.. ......_....r 1 _ 1/4 2-8 .
j
g 3/4.0 APPLICABILITY l LIMITING CONDITION FOR OPERATION . 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be l met. 3.0.2 Noncompliance with a Specification shall exist when the requirements of - the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for
~
Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided - in the associated ACTION requirements, within one hour actico shall be initiatec to place the unit in an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:
- 1. At least STARTUP within the next 6 hours, l
. 2. At least HOT SHUTDOWN within the following 6 hours, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours.-
) Where corrective seasures are compiated that permit operation under the ACTICN requirements, the ACTION may be taken in accordance with the specified time Smits as seasured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual Specifications.
- This Specification is not applicable in OPERATIONAL CONDITIONS 4 or 5.
f 3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are
- set without reliance on provisions contained in the ACTION requirements. This R
provision shall not prevent passage through or to CPERATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements. are stated in the individual Specifications. . j -
~
li - k. . i j . < e 1 1 l1 I! i ii S
# ~
[ .. . i, 1 CLINTON - UNIT 1 3/4 0-1 APR 1S;g ,
CONTAINMENT SYSTEMS , 3/4.6.7 ATMOSPHERE CONTROL CONTAINMENT HYDROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.7.1 Two independent containment hydrogen recombiner systems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With one containment and/or drywell hydrogen recombiner system inoperable, restore the inoperable syst2m to OPERABLE status within 30 days or be in at - least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.6.7.1 Each containment hydrogen recombiner system shall be demonstrated '- OPERABLE:
- a. At least once per 6 aonths by verifying during a recombiner system -v functional test that the heater sheath temperature increases to greater than or equal to (600)*F within (60) minutes. (Upon reaching (700)*F, increase the power setting to maximum power for (2) minutes and verify that the power meter reads greater than or equal to (60) kW. Maintain > (700)*F for at least (2) hours.) -
- b. At least once p,er 18 months by: -
~
- 1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits.
- 2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure; i.e, loose wiring or structural connections, deposits of foreign materials, etc.
- 3. Verifying the integrity of all heater electrical circuits by
. Performing a resistance to ground test following the above required -
functional test. The resistance to ground for any heater phase phase shall be greater than or equal to(10,000)ohns. , t 4 CLINTON - UNIT 1 3/4 6-64 APR 1 S 19S5
S
. I
\
1.7 = 1.6 - 1.5 - c' e. 8 1.4 - c 5
.u c
13 - 1.2 = Ot.M ?M =t..3 1.1 -
' ' '
- i 1.0 30 100 120 0 20 4 80 CORE FLGV (% of Rotect l
Cinton MC'Rf(Kfl Vers::s Care Flow
'** Figuts 3.2.3 1
- _:: Ten-urnT - 3/4 2-7 ..
O
J 1.7 = 1A = -
=
1.5
- e. .
C . a. O = r.a 1.4 c . ma
. E 1J =
- t 33 _
I' ( . . 1.1 - l i t' i . ! , t 1 ' ' IS E IU r 0 20 # - 1 I' c::n ecv.u m en.=> l . l-1
- c:me., ucra, txy vm::s p.wer -
- Figure 12:-2 L euvsras-gam? I 3/42-8
* ' Y Y k, ? Y, J Q ,
i k % *'. $ b b e- D. d. ANSWER KEY SECTION 5 l 5.1 B
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 5, page 5-10 5.2 D
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 5, page 5-29; Chapter 9, page 9-4 5.3 Horizontal 3 Vertical 5
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 9, page 9-6 5.4 Temperature can only increase to 705.5 F and maintain the water in a liquid state. The critical temperature of water is 705.5 F. (Candidate must identify change of state or refer to critical temperature to receive full credit.
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 3, page 3-13 5.5 Radial, Local and Axial peaking factors
Reference:
CPS Nuclear Power Plant thermal Science, Chapter 10, page 10-21 5.6 a. 3
- b. 2
- c. 3
Reference:
CPS Reactor Theory, pages 83-87 4
. r 5.7 a. False
- b. True
Reference:
CPS Reactor Theory, pages 68 and 69 5.8 c
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 10, page 10-4 5.9 SDM = 1-K(eff) K(eff) SDM = 1-0.899/0.899 = 0.112 AK/K = 11.2% AK/K
/,0 Woints for correct math)
$[(4,4-points
( for correct answer) 5.10 The delayed neutrons from Pp-239 are born at a higher energy then those of U-235 (0.5). Therefore, the slowing down length increases for delayed neutrons over core life, more neutrons leakout, and fewer delayed fast neutrons reach thermal energies (0.5)
Reference:
CPS Reactor Theory, page 49 ll 5 ,10 1. Spontaneous fission of U-238
- 2. Gamma n reaction with D20
- 3. Alpha n reaction with 0 18 (must be in proper order to receive full credit)
Reference:
CPS Reactor Theory, page 96 f 5.12 SITUATION ONE [0.5] Control rod worth is proportional to the square of local flux. [0.5] With all rods inserted and the reactor shutdown the average core flux is very small. If the center rod is fully withdrawn, the flux in the area of the l withdrawn rod increases substantially [0.5], and core multiplication increases. Because this rod causes the value of (local flux / average flux) squared to be large. Therefore, the worth for this rod is quite large [0.5]. 2
f.Q NUCt. EAR POWER PLANT THERMAL SCIENCES . Boiling Heat Transfer Reed Robert Burn . September, 1984 Figure 9.3 Heat Flux Versus Teeperature Di M erence Between Cladding and Coolant FILM SOILING ONSET OF TRANSITION BOLLING, C' D' cr 8 a - NUCLEATE SOILING b I I
&A+ 8 C DJ > ; E-SilIGLE- -
MIASE FLOW LOG (TCUO" COOLANT} Page 9-6
,-y -- _ - . _ - . . - . - - - - - -n, , ,. , . - - - - - - , _ - - ,
1 Inserting the same rod from the fully withdrawn position with all other rods fully withdrawn, the flux depression caused by the insertion will result in a small change in the value of (local flux / average flux) squared [0.5]. (2.5)
Reference:
CPS Reactor Theory, pages 80-83 5.13 a. void fraction is the volume fraction of steam in a steam-water mixture void fraction = volume of steam volume of steam + volume of water (1.0)
- b. steam quality is the weight fraction of steam in steam-water mixture steam quality = weight of steam <
weight of steam + weight of water (1.0)
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 9, page 9-20 5.14 1. E (LHGR)
- 2. B (CPR)
- 3. C (APLHGR)
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 12, page 12-2 5.15 a ,2 ,3 b-5
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 16 5.16 b
Reference:
CPS Reactor Theory, pages 48-53 3
5.17 1. Neutron embrittlement of the cladding
- 2. Thermally induced pellet growth
- 3. Inward motion of the cladding walls (creep down)
- 4. Clad weakening from (thermal) cyclic stress
Reference:
CPS Nuclear Power Plant Thermal Science, Chapter 10, page 10-16 5.18 b
Reference:
First Law of Thermodynamics CPS Nuclear Power Plant Thermal Science, Chapter 11, pages 11-6 through 11 ( *** END OF SECTION 5 *** ) i 4
ANSWER KEY SECTION 6 6.1 a. 1. fk! F C- .; O,1 Qq'* C
- 2. F0 .
f
- 5. .Fo F G shhlt fM ^*(N ,( t1 riu 0
-(0 5 e =+3
- b. Valve stem air to off gas system is lost (0.5)
Reference:
CPS No. 10P3214.01S 6.2 a. Indication that the RGDS (Rod gang drive system) finds disagreement between the signals received from the 2 RACS (Rod action control system).
- b. Indication that the withdrawn rod must be fully inserted before any other control rod can be moved. Mode switch (0.5 each) in refuel.
Reference:
CPS 3304.02 6.3 d
Reference:
EHC Lesson Plan; Recirc System Lesson Plan ,, d i 6.4 - Auto initiated at Level 3 (+8.9") (0.5) @
- Level signal increased to +55" (0.3) for 10 e p / (0.2)
- After 10 seconds (0.2), +55" replaced by 185 signal (0.3)
- No Reset until operator actuation of "Setpoint Setdown Reset" (0.5)
Reference:
Feedwater Control Lesson Plan 6.5 a. To minimize thermal shock to the CRD when the scram is reset. (a minimum cooling flow is maintained while recharging the scram accumulators)
- b. 2
Reference:
CPS Question 6.06 6.6 Containment Spray Initiation
Reference:
RHR Description, Page 13
6.7 See attached 6.8 c et D
Reference:
3203.01; 4200.01 6.9 1. Reactor water
- 2. RWCU system water
- 3. CRD system water
- 4. Drywell sumps
- 5. Containment drains
Reference:
CPS No. 1890.34 6.10 ft' Q I A s" ((>c k
Reference:
CPS No. 3314.01 6.11 a. Slow
- b. 4een krclockwise OrRi
Reference:
Standard Diesel Generator Operations 6.12 a. Will inject. Turbine seal leakage r ulting in pot t 1 airborne activity in the RCIC room, pqn g ST. ! b. Will inject. Pump overheating and seal damage result during low or non-flow conditions.
- c. Will not inject. Maximum signal from the flow element will result in the flow controller keeping the turbine speed at minimum.
Reference:
CPS Question Bank (6.14) 6.13 a. No. The SRV's are downstream of elbow taps (flow input to Feedwater Level Control). Therefore, all steam flow is sensed. , b. The generated megawatts will decrease a small amount ( because the Turbine Control System will close down the turbine control valves to maintain reactor pressure. l I l 2 l
- c. Reactor pressure will initially decrease due to increased steam flow until the Turbine Control System has a chance to adjust pressure upward by closing down on the TCV's.
- d. Reactor power will end up at approximately the same as when the transient started.
- e. Recirculation flow would increase resulting in a reactor power increase.
Reference:
CPS 4009.01 6.14 1. Discharge line temperature recorder on (1H13-P614)
- 2. SRV flow monitor on (1H13-P866)
- 3. SRV position indication on (1H13-P601, P642 or DCS display)
- 4. Suppression pool temperature recorder on (1H13-PkG ) 9-f 6 39)
- g. A c c e s , sc . m e u n c e s.
NOTE: Panel numbers not needed for full credit but if given must be correct.
Reference:
CPS 4009.01 ( *** END OF SECTION 6 *** ) 1 3
.=
c e d - L j t trad t. n r.v 1 t,* 8 *e m u8* G5 START SYS. RR REV. O RR2-3 RTS3 DEPRE55ED g LOW $ PEED SEQUENCE INtilAlID I!EAL5 NJ) _=.- INCOMPLETE 37 y y !f SEQUENCE f NOT -
/ PUMP PUMP $ PEED OPERATED CB 2 SPEED PUMP $ PEED BETWEEN 20 OPEN -
( 207 ) 957 AND 267, A 6 C O it _ [ , CB 5 1f , II RACKED IN lI PUMP MOTOR CI*I IP CLOSE LOCKOUT CLOSED i C81 RELAY RESIT CB-5 g 3p FCV(F060) y y l IN MANUAL pggp l AT MINIMUM MOTCR CB 2 l VOLTAGE ( 75
'I OPEN yoty$ pog 4 i PUMP SPEED SUCTION & NOT BETWEEN l DISCHARGE 20 AND 267. y if lf '
VALVES ) 907 OPEN CB-3 AND CB-2 GEN OUTPUT
. CB-4 SHUT CLOSED VOLTAGE NEAR U-g RATED THERMAL SHOCK dT5 1[ 1[ 1[ 1[ if IN UMITS PUMP HIGH SPEE 4 CB 5 ' TI E DELAY LOC O T EW SEOUENCE 1 RESET RX POWER
( 307. AND 37 lf lf NOT BYPASSED + START 4D SEC RESET cg,$ INCOMPLETE m INCOMPLETE
~ ' '
3 SEQUENCE SEQUENCE TIMER , TIMER . N ! CS-1 & CB-2 ! RACKED IN TIMED RESET )I OUT 1r CLOSE CB-2 1I LOW SPEED START SEQUENCE CB-5 NOT IN COMPLETE "5 TOP LOCK" Y If L K N TRtr CR-1 .
// A0"'N b RE AND CS-5 Low Speed Starting Sequence
ANSWER KEY SECTION 7 7.1 1. Place the Mode Switch in Shutdown
- 2. Insert a manual scram
- 3. Trip both recirc pumps (0.5 each)
Reference:
CPS 4404.01 b 7.2 Establish LPCS or LPCI flow from the Suppre ion Pool with injection to the RPV (0.5) and open SRV's to establish return flow to the Suppression Pool. (0.5)
Reference:
CPS No. 4403.01
- 7. 3 b
Reference:
4403.01 7.4 d
Reference:
4401.01 (CAF) RCIC License Review Manual 7.5 a. Loop Manual
- b. 10%
- c. 5%
Reference:
330.2.01(CAF)(4008.01)(CAF)RecircFlow Control System Lesson Plan 7.6 b
Reference:
4005.01 7.7 a. 1. li.5 feet , / 8. I
- 2. 212 deg P. f5 0 .
- 3. 140 t;-F Jg g* - )$$ fl 19 0,
- b. To ensure that there is adequate NPSH for the respective ECCS pumpsju Add <Mf . Aid M
Reference:
CPS 3310.01, 3312.01, 3313.01, T.S. 3.6.3.1
7.8 Gas Pressu're remains constant', ^+
Reference:
CPS No. 3304.01 7.9 a. This level promotes natural circulation in the reactor vessel.
- b. The valves should remain closed to prevent allowing reactor water to flow to the suppression pool.
Reference:
3312.01 7.10 1. Vent the hydrogen pressure (to 2-5 psig) (
- 2. Purge the hydrogen from the generationgCO-2 (to a CO-2 purity of 95%)
m y & o'b 630-
- 3. Purge t',ie C0-2 from the generator with instrum. cat eir (to a C0-2 purity of 0%)
Reference:
3111.01 7.11 a. 25 REM
- b. 25 re.T.. 3 A0wd '
- c. Voluntary
- d. 75 REM
Reference:
CPS Emergency Plan, Section 4.3.1.2, page 4-10 7.12 1. Drywell Pressure > 1.68 psig (+-0 psig)
- 2. Drywell Temperature > 135 deg F (+0 deg)
- 3. Suppression Pool Temperature > 95 deg F (+-0 deg)
- 4. Suppression Pool Level > 19 ft 5 inches (+-0 ft)
- 5. Suppression Pool Level < 18 ft 11 inches (+-0 ft)
- 6. Containment Temperature > 120 deg F (+-0 deg) (5 0 0.5 each)
Reference:
CPS 4402.01 4 2
7.13 1. Place the Mcde Switch in Shutdown (only 0.25 credit for scram the Reactor)
- 2. Close all Group 1 Isolation Valves
Reference:
4100.01 7.14 a. Reactor water level is unknown or cannot be determined to be above TAF and H2 concentration is less than 6% and 02 concentration is less than 5%.
- b. Drywell or containment H2 concentration is greater than 1% and less than 6% and oxygen concentration is less than 5%.
- c. Containment H concentration is greater than 1% and containment H 2 concentration2 is less than 6% or containment 0 conce;tration is 2
less than 5%.
Reference:
CPS License Review Manual, Combustible Gas Control System Qga n L % m'( 7.15 a. When both pumps are declared inoperable (0.25) or low (0.5) ecc N <w c % w.sleter[ A <t fer a withdrawn alarms _begin-apoearinat ressure ou .L% av.<b43,M g a < 4 /..,' y r
- b. Place the Mode Switch in Shutdown (0.35 f'or Scram the Reactor) and vHfy appropriate aulu actiomr h~ ave scen rrari. (0.5)
Reference:
4100.01 7.16 Due to the excess delta pressure developed across the Startup Level Control Valve level instabilities may result. mng ,.xhw J M G f u> m A ,
Reference:
Feedwater Control System Lesson Plan 7.17 False 271TW
Reference:
3304.02 7.18. 1. Verify the Standby Pump Starts 4g l 2. Isolate the Fuel Pool Heat Exchanger (0.25) (EW :,ide d MMW MY h f
- 3. Isolate RWCL' (0.25' (CCW side alsc) (0.25) l l
4. VL f& m f
&ni}taReac@turReviisuloticaSystems (0.5 each)
Reference:
3203.01 ( *** END OF SECTION 7 *** ) 3
ANSWER KEY SECTION 8 8.1 c
Reference:
TS definition, page 1-1 8.2 4 months
Reference:
10 CFR 55.31e 8.3 1. All rods fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn
- 2. Cold (68 deg. F)
- 3. Xenon free (0.5 each)
Reference:
TS Definition, page 1-7 8.4 1. Reactor Water Level maintained > TAF.
- 2. Core being sprayed by HPCS.
- 3. Core being sprayed by LPCS.
- 4. Reflooding flow of 1 LPCI pump injecting into the core with reactor water level high enough to produce 2 phase flow through the core.
- 5. Steam flow of (later) through the core. (4 @ 0.5 each)
Reference:
CPS No. 4401.01 8.5 a. YES (0.5) TS 3.0.4 does not preclude a mode shift since no Action Requirements are, or will, be relied on immediately upon shifting into Operational Condition 1. (E0C-RPT is not applicable until Rated Thermal Power is >- 40%) (0.5) (1.0)
- b. Void reactivity feedback due to a pressurization transient (0.5) could add positive reactivity at a faster rate than the Control Rods can add negative reactivity late in core life. (1.0)
Reference:
TS 3/4.3.4.2, 3.04
8.6 a
Reference:
TS 3/4.5.1 and 3/4.5.2 8.7 a. PRESSURE BOUNDARY LEAKRATE - Not Exceeded UNIDENTIFIED LEAKAGE (5 PM) . - Not Exceeded
- M Exceeded TOTAL LEAKAGE UNIDENTIFIED (SS-GPM)
LEAKAGE (2 GP1 INC A- Exceeded 6f'REASE) (0.5 each) (0400-0800) 0.1- ID 0.4- Eval)
- b. Pressure Boundary Leakage shall be leakage through a nonisolable fault (0.5) in a reactor coolant system component body, pipe wall, or vessel wall. (0.5) (1.0)
Reference:
TS 3.4.3.2 8.8 1. One rod out (aG A
- 2. Refuel platform Position
- 3. Refuel Platform Main Hoist Fuel-loaded
- 4. Service Platform hoist j)mq: - loaded
Reference:
TS 3/4.9.1 8.9 a
Reference:
TS 3/4.9.10 8.10 c
Reference:
TS 3.2.3 8.11 b
Reference:
TS 3.04., 3.6.7 8.12 Because the dissolved oxygen content of the reactor coolant is typically higher during low steaming rates (e.g., Startup or Hot Standby).
Reference:
TS 3.4.4 2
. . l 8.13 The following checks should be made: !
---Br:dg charainn_sp_rjage ;-d. - [3 , M (0.5) therging =otor-disconnect +4tch-en. . ' y --C' (0.5)
~J<$0 /
Cos@ci power 4h fM (O.5)
}
Reference:
CAF D AIA 0 I f /06 0 ( Mp g. 2.1. t 8.14 a
Reference:
TS 3.2.2 8.15 a. 1 f. 1
- b. 1 g. 0 (n/a)
- c. 2 h. 1
- d. 2 i. 1
- e. 1 j. 0 (n/a) (0.2 each)
Reference:
TS Table 6.2.2-1 8.16 a. SJAE
- b. 10,000 microcuries/sec - OR - 15%
- c. one hour
- d. less - OR greater
- e. 75,000 microcuries/sec NOTE: Answers to b and d must agree for full credit
Reference:
TS 3.4.5.c 8.17 a. Shift / Assistant Shift Supervisor
- b. Iirit?.i:r, uoLe, time, Dosimeter Reading'.J N Dett m Jjli
Reference:
CPS 1905.10 ( *** END OF SECTION 8 *** ) 3}}