IR 05000443/1985024
| ML20138J445 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/22/1985 |
| From: | Coe D, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20138J417 | List: |
| References | |
| 50-443-85-24, NUDOCS 8512170524 | |
| Download: ML20138J445 (45) | |
Text
{{#Wiki_filter:_ . ., n U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-443/85-24 (OL) FACILITY DOCKET NO. 50-443 FACILITY CONSTRUCTION PERMIT NO. CPPR-135 LICENSEE: Public Service of New Hampshire P. O. Box 330 Manchester, New Hampshire 03105 FACILITY: Seabrook 1 EXAMINATION DATES: October 1 - 4, 1985 CHIEF EXAMINER: // D. H. Coe, Reactor Engineer (Examiner) Date- ' ' REVIEWED BY // P Robert M. deller, Chief, Projects Section 1C Dr te a APPROVED BY: // J2/C Harry B. Kis%r, Chief, Projects Branch No.1 I Date SUMMARY: Five Senior Reactor Operators (SRO) were examined. One of the five SR0's took only the oral and simulator portions of the exandnation, having successfully passed the written portion during the preytaus NRC examination in March 1985.
" .R512170524'851210 DR ADOCK 050 4j3
- _ - - _ , . REPORT DETAILS TYPE OF EXAMS:. Initial X Replacement Requalification I l SR0 l l Pass / Fail l I I L Written Exam l 4/0 l ~l I r L Oral Exam l 5/0 l l
I ! l Simulator Exam l 5/0 l-l E I I i Overall l 5/0 l f.
.1.
CHIEF EXAMINER AT SITE: D. H. Coe (NRC) 2.
OTHER EXAMINERS: R. Keller (NRC) 8. Norris (NRC) 1.
- Summary of generic deficiencies noted on oral and simulator exams:
None 2.
Summary of generic deficiencies noted from grading of written exams: SRO Exam (four candidates) The following was an area of minor weaknesses: The ability to use pump and system curves to calculate' flow rate and pump head.
3.
INTERFACE WITH PLANT STAFF DURING EXAM PERIOD: The simulator instructors continue to perform their liaison duties in an outstanding manner, greatly enhancing the examination process.
-The simulator malfunction list continues to have a significant number of malfunctions with only a cursory description.
This was mentioned in the last Seabrook examination report (No. 50-443/85-04(OL)). NRC examiners and their contractors and consultants are infrequent users of the Seabrook simulator facility and, as such, require full and complete documentation of the simulator capabilities.
Presently, on-site changes to prepared scenarios are routinely made because additional ~information becomes available only.afte'r examiners-interface with the training staff and become more experienced through observation of the simulator during the course of the examination.
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _
.
., In order to reduce the need for last minute scenario changes, the follow-ing categories of information are required as appropriate for each mal-function: 1) Malfunction title and precise description (already provided).
2) Severity level options and ramp options with corresponding plant parameters, e.g. 0-100?; with 100?J=2000 GPM and can be ramped over five minutes.
3) Expected initial alarms and indications for a given initial-condition and malfunction severity.
4) Applicable procedures or references 5) The expected plant response for a given initial condition and mal-function ' severity, and assuming no operator action.
This should-include expected times to major events, e.g. five minutes to ESF actuation, be in chronological order, and include alarms.
This information will greatly assist in realistic scenerio development and will help acquaint the examiner with the specific plant transient response prior to arriving on site. Normally, only one initial condition (usually at 100*4 power) and malfunction severity level should be needed. 'Brief descriptions of how plant response would be altered for different initial conditions or severity. levels should be included where appropriate.
Strip chart recordings of significant plant parameters are highly desirable.
Certain malfunctions which cause minimal' transient response would not require this kind of documentation ~. 6) xpected operator action, if different from item 4 above.
7) Any options associated with this malfunction.
In addition to the above, information regarding the general capabilities of the' simulator are required. For example, what override options exist and what specific pumps, valves, and breakers can be overridden? Finally, the malfunctions should be categorized according to major system or some other logical order in order to facilitate their use.
In a 22 November 1985 phone call betwee'n D. Coe (NRC) and P. Richardson (Seabrook), the facility agreed to incorporate the above information in the simulator documentation prior. to the next Seabrook operator licens-ing examination.
(0 pen Item 85-24-01) 4.
Personnel Present at Exit Interview: NRC Personnel D. Coe, Chief Examiner B. Norris, Examiner D. Ruscitto, Resident Inspector %
_
- Facility Personnel-D. Moody, Station. Manager .. P. Richardson, Training Center Manager J. Gri l 1 o, As si s ta nt 0pe ra ti o n s.yan age.r____.- u- ~~',, ~~~~ D. Schreiner, Actingiratirnig~ Supervisor 5.
Summary of NRC Comments made at' exit interview: The. candidates' overall performance was characterized as'very good.
No specific preliminary results were presented.
The facility raised the issue of the validity of question 7.04 on the written exam. They agreed to make their comment in writing and this item-was left unresolved.
.6.
Examination Review At the conclusion of the written examinations, the examiners met with licensee personnel to review the exam and answer keys to identify 'any inappropriate questions relative to plant specific design, and to ensure that the questions will elicit the answers in the key and that they reflect the most current plant conditions. The following licensee per-sonnel were pr'esent: P. Richardson D. Schreiner . Attachments: 1.
Written Examination and Answer Key (SRO) 2.
Facility Comments on Written Examinations made after Exam Review and NRC resolutions .
_ . '. ' 'f t1rreclunen S / U. S. NUCLEAR REGULATORY C; lilSSION -
- ! SENIOR REACTOR OPERATOR L'ICEN.'.1 EYAMINA.T:C;i . _ -. ~ ~ ~ ~ ~ ~ J __ - / ~~' Facility: SEABROOK .- __.
' " ~ i Reactor Type: PWR-WEC4 Date Administered: 10/1/85 - Examiner: O.
W.
BURKE Applicant: INSTRUCTIONS TO APPLICANT: . Use separate paper for the answers.
Write answers on one side only.
Staple questions' sheet on top of the answer sheets.
Points for each question are indi-cated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
Category % of Applicant's % of Cat.
Value Total Score Value Category . 25.0 25.0 5. Theory of Nuclear Power Plant Operation, Fluids & Thermodynamics 25.-0 25.0 6. Plant Systems: Design, Control & Instrumentation ._ %.. 25.0 25.0 7. Procedures-Normal, Abnormal, Emergency & Radiological Control 25.0 25.0 . 8. Administrative Procedures, Conditions .. and Limitations 100.00 100.00 TOTALS Fin 61 Grr.de % ~ All work done on this exam is on my own, I have neither given nor received aid.
Applicant's Signature
. -_ w a..-- - .- - ~ ~ ^
, .- - . " v . .. ~ 5.
Theory'of Nuclear Power Plant Operation, Fluids,'and Thermodynamics'(25.0) Question 5.1 .(2.00) , Two identical reactors have identical shutdown reactivities. One has a neutron source which gives off twice as many neutrons as.the other. The reactors are simultaneously started up and rods are pulled to. criticality.
How will power l levels compare when. criticality is reached? Explain.
a.
b.
How will the critical rod heights compare when criticality is reached? Briefly explain.
- Question 5.2'
(1.00) -List two possible causes for axial flux deviation (delta I) being outside-the' target band (other than instrument error) during power operations.
~ Question.5.3 (4.00)
- A reactor whose reactivity change information is given in figures-5.1-5.3 attains. criticality'with all rods fully withdrawn except for rod bank D which is.100 steps withdrawn.' Assume that the boron concentration'is held constant at 900 ppm and that the reactor wa $ tia11y at BOL and hot zero oower (HZP) with all rods bottomed.
In your calculations use Q =. 007 and _ 2.= 0.08.
Specify the-figure number source for each number used in your calculations.
a.
What. was the initial Keff with all rods bottomed?- Show all work._.
(2.0) 'b.
What-would be the D bank' position required to obtain a stable 0.8 DPM startup rate (starting from a stable, just critical position of 100 ~ sters on D bank)7 Show all work.
(1.0) _ .c.
~If the reactor power is increased to 30% of ful1~ power using rod ~ Jmotion only, maintaining progranmed Tave, and. disregarding any long term poison effects, what will be the group D position at 30% steady power?' Show'all work.
(l.0) Category 5 questions continued on next page.
. . _ _ _ _ _ _ _ - _ - _ _ _ - . - _ _ - _
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- Question 5.4 (1.0)
a.
Delayed neutrons are effective than prompt neutrons in . causing fission. Answer either more, less, or equally as.
-(0,5) b.
What is the relationship between Beff and 57 ~ ' Question 5.5~ . a.
How does'the value.of MTC change over core life? (0.5) 'b.
' Explain the cause(s) for the MTC change over core life.
'(1.0) . Question 5.6 (2.0) ' Explain why a dropped rod could be worth 200 pcm and a' stuck rod could be worth 1000 pcm even.though the same rod is considered in both cases.
Question 5.7 (2.0) Explain the two mechanisms for the production and the two mechhnisms for the removal of xenon in the reactocetoTe.
... ""' -Question 5.8 (1.5) - .,-e> p.. Why.must the Shutdown Margin required by Technical Specifications be greater than.1.3% delta K/K for Modes 1, 2, 3, and 4 but only 1% delta K/K for Mode 5? - E ,- , Question 5.9 '(3.0) In order to. prevent overheating the fuel, the reactor is operated such that th> point of DNB is not reached (DNBR>l.3).
~ Since the DNBR is not directly measured, what four primary system-parameters do you, as the operator, monitor to assure that DNBR is greater than 1.3 and how would an increase in the value of each of these parameters individually affect DNBR-(increase, decrease,.or no effect)? Category 5 questions continued on next page.
.
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Category 5 questions continued.
Question 5.10 (2.0) Find the enthalpy change in an isentropic expansion of steam through a a.
turbine into a condenser (Note: Entering steam is 100% quality at 825 psia, condenser pressure equals 2-psia.) Explain how you arrived at your solution.
(1.5) b.
How would the change in enthalpy in part (a) be affected by a less than ideal turbine (i.e., some degree of inefficiency)? Select one of the following answers.
(0.5) 1.
higher 2.
same 3.
lower-Question 5.11 (3.0) Please refer to figure 5.4.
What will be the total system flow rate (G.P.M.)-and the pump discharge header pressure with all three pumps _ operating? Give a brief explanation of your reasoning process used to
- .
arrive at your solution. You may mark on the figure in doing your J 8F solution if you desir.e.
~- . ,u ,w ' Question 5.12 (2.0) Approxihately ten to fifteen minutes after the initial loss of forced flow in the primary system, natur,a1 circulation should be established. What indications should the operator check to determine whether natural circulation has been established? End of Category 5 questions.
-._ .. . . ,
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. . . U.S. NUCt. EAR REGULATORY COMMISSION l INTEGRAL CONTROL BANK WORTH VERSUS STEPS WITHDRAWN , ' WITHOUT OVERLAP (BOL AND HZP) , l 2200
2& %dl '
\\ \\ l900 L - 1 1600' .o
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i
0
40
80 100 120 140 160 180 200 220 240 Steps Withdeawn FIGURE 5.1 __ . . . - -
~ . . i .. . ' e . . U.S. NUCLEAR REGULATORY COMMISSION INTEGRAL SHUTDOWN BANK WORTH VERSUS STEPS WITHDRAWN (BOL AND HZP) 3600 320G j'8A - %m '5(' '2800 \\ t E \\ j 24m y fi \\ B t . So 2000 \\, o i !
\\ e 1( ei S 1600
\\ \\ \\ . \\ 1200 ~ MS \\ B 'Ml l \\ w' 3( 800 s N N 'N ' k & 400 % % % %.'N '% % % %, mu-.
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Plant Systems Design, Control-and Instrumentation (25.0) ! Question 6.1 ( 1. 0 ) ~- -What is the major reason'for installing the anti-reverse rotation devices
on the reactor coolant-pumps? . Question 6.2 (2.0) ' Suppose that a reactor coolant pump fails to start when the-switch is - placed in the. start position.- What four " logics" or interlocks would you checkLto determine the cause for.the pump's failure to start? Question 6.3-(1.0) What is the function of the J-tubes tha't are attached to the S/G feed ring?. . Q5e'stion 6.4 (3.0) Excluding the use~of the normal 80 RATION flow path, describe three methods of ~~ borating the RQS in case the normal EMERGENCY BORATION flow path is not . ~we-available. You may sketch on Figure 6.1 if you desire.
, ., i . Question 6.5- -(2.0) _ Refer to Figure 6.1.
Indicate the valve positions [open, closed, or modulated'(partially open)]- a.
of valves'110A, 1108, Illa and 1118 for the following positions of the' . makeup system mode selector.~ switch: 1.
Borate (.4) ~2.
Dilute (.4) 3.
Alternate Dilute (.4) b.
~1.
Why might you select ALTERNATE DILUTE over DILUTE to dilute the RCS? (.4) ' 2.
'Why should alternate dilution.be minimized? (.4) . Category 6 questions continued on.next page.
.
- =- . =. = _.
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. , ' . a . Category 6 questions continued.
Question 6.6 (1.0) ' Given the situation whereby a required change in RCS boron concentration can - be accomplished by either the BTRS or the reactor makeup system, which system would you choose? Give the major reason for your choice.
Question 6.7 (1.0) What feature of the pressurizer PORV control system prevents a given PORV.
(PCV-456A for example) from completely depressurizing the RCS if its pressure channel fails-high? . Question 6.8 (1.5) Motor operated block valve V122 is located upstream of PORVr PCV-456A.
If V122 ~1s closed and the control switch is in the AUTO position, the valve will open in' response to what operator actions or plant conditions? . Question 6.9 (3.0) _e A'swer the following concerning the nuclear instrumentation system: n a.
Explain the response of the reactor protection.' system upon a simultaneous failure of both IR channels (where both channel outputs go to zero) for each of the following two cases.
Briefly explain the ' differences, if any.
. 1.
The reactor is at 5% power during a startup.
(1.0) 2.
The reactor is at 20% steady state power.
(1.0) b.
Assume that while at 100% power one IR channel fails HIGH.
If a reactor trip occurs while in this condition, what additional action must be taken during the emergency procedures to ensure the proper operation of the NI system? Why? (1.0) Question 6.10 (3.0) Answer the following concerning the pressurizer level control system.
Assume that the plant is operating at 100% power.
a.
List the indications in the control room if the controlling pressurizer level channel fails low.
Do not include alarms.
(2.0) Category 6 questions continued on next page.
. ~
- ,-
- - - -,
~ -- .. _ . - r , - ., . + < .- , , . ', , - . Category 6 questions continued.
' Question 6110 continued " . hat could'cause a reactor trip as a result of the above failure if no-W 'b.
operator action were taken? Include any applicable coincidence and set point.
. Question 6.11 (2.5) If you see that the main steam dump valves are all indicating the fully a.
closed position immediately.or very soon after a reactor trip, what would
- you check to determine the reason that the dumps did not open? A minimum of three required.
(1.5) ~ ~ b.
What operational ~ difficulty could occur if the steam dump mode selector switch is not reset after a large load rejection? Why? (1.0) Question 6.-12' (1.0).
List the positive functions that will automatically trip a diesel generator during.a safety injection plus loss of power incident? ^-(-1.0) -Question.6.13-(.5) The diesel generator is normally.up to speed and voltage with its breaker . closed within after receipt of an undervoltage signal. Supply the . -tdpe. Question'6.14 (2.5) ' a- _ Answer the following questions concerning the Emergency Core Cooling System: _How are the CCp's protected from overheating after ddrmini-flow a.
recirc valves shut upon receipt of an "S" signal (Safety Injection actuation)? (.5) b.
What interlock conditions must be met before the SI pump mini-flow.
recirc valves can be opened and why is this. interlock necessary? Valve designations are not required as long as descriptions are ~ -complete. (1.0)
! What automatic action (s) occur during the shift from injection c.
-phase to cold leg recirc phase upon receipt of a RWST low / low alarm coincident with an "S" signal? Valve designations are not required . as long as descriptions are complete.
(0.5) d.
State two reasons for having throttling valves in each injection line which penetrates an RCS hot or cold leg.
(0.5) End of Category 6 Questions - .
- ._e.. - -. . ,
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Procedures--Normal. Abnormal. Emergency . and Radiological Control . (25.0) Question 7.1 '(4.5) , List the four immediate operator action steps of the emergency operating procedure FR-S.. RESPONSE TO NUCLEAR POWER GENERATION /ATWS.
For the first 3 steps only, include all substeps associated with each action step and.
'further actions to be :taken 'if the above actions fail. to achieve the i ' desired result.
.. . . Question 7.2 (2.5) 1The operator is in. Emergency Procedure E-0,' REACTOR TRIP OR SAFETY INJECTION- , when he finds-that SI is not actuated (immediate action 4).
List theifive conditions that would require manual initiation of SI.
Question 7.3_ .(1.0) ,,,, ,, uw.. _ _., . ~The following questions pertain to the emergency operating procedure FR-C.1,.RESFot4SE TO INADEQUATE CORE COOLING: .w.
What action must be taken if the RWST. level drops to-less than 23.5% a.
at any time during this procedure? (0.5) l b.
One of the conditions which' requires entry into this pYoceI'ure is a I single specified core parameter reaching a specified value. What is this core parameter'and its corresponding value? (0.5) ,. Question 7.4 (1.0) What condition requires entry into procedure FR-Z.3, RESPONSE TO HIGH a.
CONTAINMENT RADIATION LEVEL? (0.5) b.
What automatic action is initiated by the condition in part a? (0.5) l i - ' Question 7.5 (2.0) t [ Step 2.in procedure E-3. STEAM CENERATOR. TUBE RUPTURE, states " Identify [ Ruptured SG(S)." List four things the operator should check in an attempt ! to identify the faulted SG(S).
I Category 7 questions continued on next page.
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-- - -. -
. , . . . . ' -Category 7 questions continued.
Question 7.6 (1.5) - When critical and immediate action is required for work of short duration, Health Physics supervisory personnel or the shift superintendent may ~ authorize entry without an RWP if. certain conditions are met.
List these 3 conditions.
Question 7.7 (2.0) In order to maintain the plant at 100% power, work must be performed inside containment in a radiation field of 300 KREM/HR gamma and 60 MRAD /KR thermal and fast neutrons. The maintenance man selected is 30 years old and he has a lifetime exposure through the last quarter of 58 REM on his NRC form 4.
In addition, he has accumulated 1.0 REM so far this' quarter.
How long may the man work in this area without exceeding his 10 CFR limit? Show all work.
. - Question 7.8 (2.0) If, during startup, the reactor goes critical outside the allowable. band about the ECP, some actions are required.
a.
What is the allowed reactivity band about the ECP? (0.5) b.
What actions must be taken if criticality is achieved outside the band? Include actions if error is found or is not found.
(1.5) Question 7.9 (1.0) What are the required actions if a RCP motor upper radial bearing temperature exceeds its limiting value of 184 F while the reactor is operating at 100% power? Question 7.10 (1.0) With the reactor operating at 100% power and with control rod controller in AUTO, you observe that Tavg is rapidly dropping, nuclear power is rapidly dropping and the control rods are rapidly stepping out.
What is the most likely cause for this abnormal behavior? Category 7 questions continued on next pag _ .=m#,:.. c,s.mn.,. - - -
. .. . Category 7 questions continued.
'
Question 7.11 (2.5) . List five surveillance calculations that must be done manually in the event of a failure of the plant computer.
~ -Question 7.12 (2.5) The emergency plan lists 8 guidelines for the selection of individuals for emergency exposure during life-saving actions. Give five of these guidelines.
Question 7.13 (1.5) As a result of a fire in progress, the shift supervisor gives an order to evacuate the main control room.
In accordance with Operating Procedure S1200.02, Safe Shutdown and Cooldown from the remote Safe Shutdown Panel, what operator actions are required to establish remote safe shutdown conditions? Assume that these actions can be taken prior to control room evacuation.
End of Category 7 questions P W , . - _ _ _ _ _ __ __ _
_ - -.... e .. - s c.
. , ~ . -. + , n 8.
Administrative Procedures, Conditions.
and Limitations (25.0) - ' . uestion 8.01 (3.50) Q . a.
Assume that it is 0300 on 7-20-85 and that reactor power is presently .at 30%..Considering the AI history listed below, at what clock and calendar time pre you allowed to increase power above 50%? Assume - that-the A I. tech. specs. have not been violated during the given history and show all work.
(2.50) DATE TIME (leaving band) TIME (reentering band) POWER 07-19-85 0300-0308
~07-19-85 1747 1833
07-20-85-0148-0300
b.
Give the basis for maintaining AFD within Tech. Spec. limits.
(1.00) ' Question 8.02 (3.00) What conditions must exist for containment integrity to exist in accordance with the Technical Specification definition? Question 8.03 ,--(1.50) ** , . -. Under what conditions does the Shift Superintendent or Unit Shift Supervisor have the authority to issue changes to an UNTESTED procedure? Question 8.04 (1.00).
. hat constitutes the approval of a temporary change to a station operating W procedure? Give job titles.
. Question 8.05 (1.00)- During shift barnover, what should an operator do if his relief operator appears to be ceder the' influence of alcohol or drugs (he is not alert and coherent)? Category 8 questions continued on next page.
- r-
- - - - -. ~
- a- , - - - - ~ =, =. = =, - , " . - . ' . . Category 8 questions continued.
. Question 8.06 (1.00) Suppose that during routine plant operation a reactor operator becomes ill . and is sent home, leaving the minimum T.S. crew one man short.
a.
How soon must the operator be replaced.
(0.5) b.
Describe the routine used by management to replace the operator.
(0.5) Question 8.07 (2.50) Give the basis for having each of the following reactor trips: a.
Power range positive rate trip.
(0.50) b.
Power range negative rate trip.
(0.50) c.
Overtemperature delta T trip.
(0.50) d.
Overpower delta T trip.
(0.50) e.
Low-Low SG water level trip.
(0.50) ' . Question 8.08 (1.50) Give~the number of boron injection flow paths required to meet the limiting - condition for operation (LCO). Give a description of each flow path.
! Question 8.09 (2.50), ~~~ List the minimum A.C. electrical power scarces required to be operable to meet the limited conditions for operation.
Give any qualifying information pertaining to each source.
Question 8.10 (1.50) List the three reasons (bases) for ROD INSERTION LIMITS (RIL).
Category 8 questions continued on next pag > - ---.-~ y l
' . s . -
I - Category 8 questions continued.
- Question 8.11 (3.00) In some siLuations deviations from procedures are permissible without requiring a change to the procedure provided certain provisions are satisfied. For the following examples of procedure deviations, give the provisions that will allow their implementation without a change in the procedure.
a.
Procedure steps are performed out of sequence.
(1.00) . b.
Non-applicable steps are omitted.
(1.00) c.
A valve alignment is not completed as written because certain components are isolated for maintenance.
(1.00) Question 8.12 (2.0) Give the reasons for the following RCS precautions and limitations: a.
During a normal cooldown, at least one reactor coolant pump shall be operated.' (1.0)- -- . _., b.
During water solid operation, do not isolate the RHR suction lines" from the reactor coolant loops unless charging pumps are stopped.
.(1.0) ap ..- -" Question 8.13 (1.00) In acco.rdance with the Seabrook Emergency Plan, state authorities and the NRC must be notified within a specified time after an emergency is declared. What are the required maximum times for each' notification?' End of Category 8 End of Examination .
' EQUATIONS / DATA SHEET ,, ' 10 /o A(x) = A t-UX -
- - l1Ci = 3.7 x 10 d
o , ~ aD = - 1 x 10 5 AK/*F.
M = 1/(1-k) = CR1 , . K q - 1 x 10 3 AK/% voids N(t) = No e-AT a = y K a = (Lg+L-) (4 rod)2 r s aM m.- 1.0 x 10 4 AK/Z'F (4 avg) K n = v/(1 + d) op = -4.5 x 10 4 AK/% power K P = I 4 v/(3.7 x' 1010) I(t) = Io e-At t. (g_p)fxp - , Tl/2 = An(2)/A T - 1/p + (8 p)/Ap Cp = (CPbase) (Ks) (K ) T" 1/(P-6) A v=Vf + xvfg 2+b-h-ht=f+Z2+M pg+n+Vg 11 = xhg + (1-x) hf __ 2g
2g Y . S ~ *58 + (l~*) Sf (AP )3/2 ' gV --(Ap1) =
1 ist.p 2.54 cm ~^' ~
1 gal. = 3.785 liters . Q = MCp At 1 kg = 2.205 lb D 2gc N = pao /A ' f = 64/Re 17.58 watts = 1 BTU / min Ipsi =.Ji.895 Pa. p = k(eff) -1 1 psi = 2.036 * H (0 OC) K(eff) 1 psi = 27.68 " H 0 (@ 4C) 1_ = CR1 = 1-K(ef f )2 8 =.0071 M CR2 1-K(eff)1 I = 2 x 10 5 sec '
Q = MAh id = 12.5 . Q = UAAT RR = I 4th f A = 0.1 SCR = S 1-Keff- .
hg - kmV p=8 . P=P 10SUR(t) Ar + 1 o SUR = 26.06 T Reactor thermal power - (h -h ) x steam flow rate
t t e /r K=
P=P o 1p
. ,, 7--
. ..' s (- -.. 5.
Theory of Nuclear Power Plant Operation,
< Fluids,-and Thermodynamics t Answers L Answer 5.1 (2.0) The higher the source rate, the higher-the power level at criticality.
a.
(0.4).
For a given reactivity addition, subcritical multiplication results in a larger neutron population increase for the reactor with larger initial-neutron population.
(0.6) ~ l- 'b.
Will be the:same.
(0.4).
The critical rod position reflects the ^ positive reactivity necessary to' bring the reactor critical and is independent of source strength.
(0.6) y .Ref.
Seabrook Training material, Vol. 18. Reactor Theory Review, page HO-RTR-49.
. Answer 5.2 (l'0) . -1.
Xenon oscillation due to load changes.
2.
Rapid plant power reduction with automatic rod motion.
- 4.- - ! 3.
= Improper rod alignment.
~' (0.5 each, 2 required) 'Ref.
Seabrook Technical Specification bases, 3/4.2 Power distribution "" ~ limits, page B 3/4 2-1.
CAF . Answer 5.3 (4.0) l a.
Obtain and sum all rod worths when all rods are inserted.
Use figures 5.1 and 5.2.
Rod Bank Integral rod worth on bottom,'PCM A 1250 (0.1) B 2030 (0.1) C 1215 (0.1) , l D 1465 (0.1) S 3050 (0.1) A 1000 (0.1) Sg TOTAL 10,010 PCM (0.1) Category 5 answers continued on next page.
n ! . . M -. - ~ ~ " * -mg - -e +, - -, - -, -,,.,,_ .
.-
- .
, . , Category 5 answers continued.
- Answer 5.3 (a) continued At criticality with bank D 100 steps withdrawn. Bank D is worth approx. 500 pcm (figure 5.1).
(0.1) Therefore SDM = 10,010 pcm - 500 pcm = 9510 pcm (0.2) Keff-1/Keff the < Keff-1/Keff .0951 = 0.9132 (1.0) Keff = Ref. Seabrook Instruction material, Vol. 18, RTR, page HO-RTR-46-48.
b.
sur = 26/T 0.8 = 26/t; ,) = 32.5 sec.
(0.2) z. 0-rh = 32.5
- L(tho)
.007 - rho = 32.5 - - - - .08 (tho) rho =.00194 = 194 pcm (0.2) 500 pcm - 194 pcm = 306 pcm (0.2) - -- using figure 5.1, this would put group D at approximately 125 steps.
(0.4) Ref. Seabrook Instruction material, Vol. 18, RTR, Page HO-RTR-66.
c.
Using figure 5.3 at 900 ppm boron concentration, the power defect at 30% power is approx. 410 pcm.
(0.3).
k'e went critical with D at 100 steps (total rod worth approx. 500 pcm).
500 pcm - 410 pcm = 90 pcm rod worth at 30% power.
(0.2) From figure 5.1, group'D would be at approx. 165 steps.
(0.5) Ref.
Seabrook Instruction materials, Vol. 18. Reactor Theory Review, page HO-RTR-091.
Category 5 answers cor art page.
.
,- _ __ __
_. . , . .. Category 5 answers continued.
- Answer 5.4 (1.0)
a.
less (0,5)~ . h.
Quff equals 5 multiplied by the delayed neutron importance factor.
(0.5) Ref. Seabrook Instructional material, Vol. 18, Reactor Theory Review, Page HO-RTR-26.
Answer 5.5 (1.5) a.
The MTC becomes more negative over core life.
(0.5) b.
1.
Due to fuel burnup over core life the baron concentration must be reduced to compensate. Since boron tends to make the MTC more positive. its absence makes the MTC more negative.
(1.0) Ref.
Seabrook Instructional material, Vol. 18. Reactor Theory Review, page HO-RTR-76-78.
Answer 5.6 (2.0) Talen a rod is stuck out with all other rods inserted, that rod is exposed to a much higher flux than the rest of the core.
Because rod worth is a function of the relative flux difference between that adjacent to the rod and the core average, the rod could be worth about 1000 pcm, which is much more than normal.
- ( 1. 0 ) If a rod is dropped, while the rest of the rods remain out, the opposite of the above happens. Th'e flux is depressed adjacent to the dropped rod relative to the flux in the rest of the core and so the dropped rod could be worth about 200 pcm.
(1.0) Ref. Seabrook Instructional _ materials, Vol. 18, Reactor Theory Review (RTR), page 99.
Answer 5.7 (2.0) Xenon Production About 97% of the Xenon produced comes from the decay of radioisotopes: 135 B~ 135 B ~ pg135 8-g<o > Te 135 SB ( 6.7 hr.)
The remaining 3% is produced directly from fission.
Category 5 answers continued-on next pag *
. , . .' , Category 5 answers continued.
Answer 5.7 continued.
Xenon Removal Xenon is removed from the core by neutron absorption to form stable Xel36 which has a very low neutron absorption cross section. The other removal mechanism is the decay of Xel35 ~ l35 Xel35 6~ Csl35
- Ba
[9.2 hr.)
(2.3x10y)
Ref. Seabrook Instructional materials, Vol. 18, page HO-RTR-104 Answer 5.8 (1.5) ' ! i 1.3% SDM is required to control the reactivity transient during a main steam-line break.
In Mode 5 with Tave less than 200 F, this transient is minimal and a 1% SDM provides adequate protection.
! Ref. Tech Specs page B 3/4 1-1 i Answer 5.9 (3.0) - . -, ,# SystemParameters I.
Primary coolant temperatures (0.5) 2.
Primary system pressure (0.5) 3.
Primary coolant flow rate (0.5) 4.
Reactor power level (0.5) DNBR Effects 1.
Increasing temperature - DNBR decreases (0.25) 2.
Increasing system pressure - DNBR increases (0.25) 3.
Increasing coolant flow rate - DNBR increases (0.25) 4.
Increasing reactor power level - DNBR decreases (0.25) Ref. Seabrook Instructional material, Vol.17. Thermodynamics, Heat Transfer & Fluids Review, Objectives; also Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor II.
Westinghouse Electric Corporation Nuclear Training Services, Chapter 13, page 22.
Category 5 answers continued on next page.
- -. ,, .- ,,.. _ - _.. -,. _.. . ~ -.. _ _. _ _ _____.m.. . - _ _.. _. _ _ _ _ _ ._ __
-,- .- - .- - . . Category 5 answers continued.
- Answer 5.10 (2.0)
a.
1.
Solution using the Mollier chart. At the intersection of the '825 psia pressure line and the saturation line, read 1199 BTU /LBM enthalpy. From the above point, trace the constant entropy line down to its intersection with the 2 psia pressure line and read the enthalpy of 819 BTU /LBM. The change in enthalpy = 1199 - 819 = 380 BTU /LBM.
2.
Solution using steam tables.
Remember that the isentropic process is a constant entropv process.
Find in steam tables, 100% quality steam at 825. psia pressure, entropy = 1.4129 and enthalpy = 1198.7 dTU/LBM. From steam table, at 2 psia, entropy of sat, water is 0.1750 and Sfg = 1.7450.
Find the steam quality, X, at 2 psia and entropy 1.4129.
X = 1.4129 .1750 x 100 = 70.94% 1.7450 ,, The enthalpy at 2 psia = 94.03 + (1022.1 x.7094) = 819 BTU /LBM The enthalpy change = 1198.7 - 819 = 379.7 BTU /LBM.
Either solution 1 or 2 is acceptabic for 1.5 pts.
b.
3'(lower).
(0.5) Ref. Seabrook Instructional Material, Vol.'17. Thermodynamics. Heat Transfer and Fluids Review. Objectives; also, Thermal-Hydraulic Principios and Applications to Pressurized Water Reactor II, Westinghouse Electric Corporation Nuclear Training Services, Chapter 71 also ASME steam tables.
Answer 5 11 (3.0) . In the following manner, sketch the flow vs. pressure curve for all 3 pumps operating. Do this sketch on figure 5.4.
The points for the curve may be generated as follows: -System Pump #1 Pump #2 Pump #3 total Pressure flow flow flow flow
0
45
88
5
50
0
45
70
33
78
0
45
Category 5 answers continued on next page.
__
. . - . , . ' . . Category 5 answers continued.
! Answer 5.11 continued.
- Using the above points, construct a curve on figure 5.4 and note the point [ at which it intersects the system head loss curve (operating point). The pump discharge header pressure is 82 psig and the operating flow rate is 62 G.P.M.
Ref. Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor II. Westinghouse Electric Corp. Nucicar Training Services Chapter 10.
Answer 5.12 (2.0) Indications that natural circulation is established are: ! 1.
A constant or decreasing delta.T across the reactor core less than full load delta T.
(0.5) < 2.
Core outlet temperatures are constant or decreasing.
(0,5) 3.
A constant steam generator level with a constant auxiliary' feed rate and a constant or decreasing steam pressure.
(1.0) Ref. Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor II, Westinghouse Electric Corp. Nuclear Training Services, Chapter 14, page 27.
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Figure 5.4 .
c.
- . _
e . -
.. , ,
- .
6.
Plant Systems Design, Control and Instrumentation Answers i . ' Answer 6.1 ( 1.0 ) ! ~ .[Theanti-reverserotationdevicepreventsadeenergizedpumpfromrunning inthereversedirectionwhenthe-pumpsintheotherloopsarerunning.)
This permits the idle pump to be started without drawing excessive starting currents for prolonged periods of time and overheating the pump motor.
Ref..Seabrook Instructional Materials, Vol.-13. RCP, page HO-RCP-35.
Answer 6.2 (2.0) The four interlocks are: 1.
Lift oil pump running with a discharge pressure greater than 600 psig.
(Dv57 (c.H.)
2.
Pump breaker 86 relay reset. (0<57 (o. 6 6) -3.
P r elect,r4eal-supply-buc-SS= relay--eeset.
(O.5) 4.
No. I seal delta P greater than 220 psid.,(Or5T[0,67) Ref. Seabrook Instructional Material, Vol. 13 RCP, page HO-RCP-66.
L( E L.
D.m. 't 76.7 - pl - 9l c 98 2. - Gil A D.T h Answer 6.3 (1.0) The J-tubes insure that the feed ring remains full of feedwater when feed flow stops so that waterhammer will not occur when feed flow is resumed.
.- Ref. Seabrook Instructional materials, Vol. 13 SG, page HO-SG-26.
Category 6 answers continued on next page.
. 6- .
- , -~ _., - . '
'* . . . Category 6 answers continued.
Answer 6.4 (3.0) (Any 3 of the following 4 for 1 point each) 1.
Supply the suction of the charging pumps from the discharge of a running boric acid transfer pump through the manually opened isolation valve CS-V452 and open valve FCV-110A.
2.
Align the suction of the charging pumps directly to the boric acid storage tank (s) by opening two manual isolation valves in the line Icading to the cross-connect between the boric acid transfer pump suctions (Valves CS-V439 and CS-V442).
3.
' Align the charging pump suctions to the refueling water storage tank.
4.
With proper valve alignment, uso a centrifugal charging pump to discharge t' cert:nte cf the be en in '--* '- -- ' into the RCS.
t,, - A r, r J/> p c.4 k Note: Valve functional descriptions will suffice in lieu of valve numbers.
'Ref.
Seabrook Instructional material, Vol. 13. RMUS, page HO-RMUS-69 Answers 6.5 ,J4r0)' (:2, c ) FCV110A FCV110B FCV111A FCV1118 ~ a.
. 1.
modulated open closed closed ( '. 4 ) 2.
closed closed modulated open (.4) 3.
closed open modulated open (.4) , b.
1.
If immediate or prompt dilution of the RCS is desired.
.(G<5)' 'i' Y) 2.
To avoid the possibility ofintroducing oxygen into the primary by not sending all of the dilution water to the top of thn VCT. 10,5)(0,*/i) Ref. Seabrook Instructional Material, Vol. 13. RMUS, figure RMUS-5.3, page RMUS-62 and figure RMUS-5.4, page HO-RMUS-64; also pages llo RMUS-61, 63, 67 and 68.
Category 6 answers continued on next pag _ ! o i
' l ' . -
Category 6 answers continued.
- l Answer 6.6 (1.0)
The BTRS because the boron concentration change can be accomplished while creating less liquid radioactive wastes.
Ref. Seabrook Instructional materials, Vol. 13. BTRS. page IIO-BTRS-7.
Answer 6.7 (1.0) The PORV (PCV-456A) will not open unless the other press,ure channel (PT-458) is sensing a pressure greater than 2185 psig.
Ref. Seabrook Instructional material, Vol. 14, PPLC, page lio-PPLC-31.
Answer 6.8 (1.5) Any one of the following will open V122 when in AUTO.
1.
The PORV interlock circuit (PT 458) senses a pressure greater than 2185 psig.
(0.5) 2.
The control switch is momentarily placed in the OPEN position.
(0.5) 1.
Auctionected low loop cold leg temperature actuates (at approx.
3500F).
(0.5) Ref. Seabrook Instructional materials Vol. 14 PPLC, page 110-PPLC-33.
Answer 6.9 (3.0) a.
1.
The reactor will trip on SR high flux (0.5] due to P-6 dropping out and reenergizing SR nuclear instruments (NL).
(0.5) 2.
The reactor will not trip [0.5] due to P-10 which prevents reenergization of the SR NI even if P-6 drops out.
(0.5) h.
The SOURCE RANGE RESET switches must be used (0.5l in order to allow the SR NI to reenergize.
10,5) Ref. Seabrook Instructional Haterial Vol. 14. NIS.
a.
1.
page HO-NIS-39 2.
page llo-NIS-44 b.
page llo-NIS-74 Category 6 answers continued on next. page.
- - - .
___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ - _. . _., . ' .
\\ - . .- , l - Category 6 answers continued.
, . Answer 6.10 (3.0) a.
1.
Increased charging flow / charging pump speed.
(If'in AUT0) ' (0.5) 2.
Actual Icvel increase on other channels' (0.5) 3.
Letdown flow rate goes to zero indicating closure of letdown isolation valve.
(0.5) 4.
Pressurizer B/R heaters are off.
(0.5) b.
Pressurizer Icvel high [0.51, 2/3 [0.25), 92% (0.25].
Ref. Seabrook Instructional Materials, Vol. 14, PPLC, page HO-PPLC-47 and page H0-PPLC-51.
Answer 6.11 (2.5) , , i a.
1.' Check that at 1 cast one circulating water pump is in operation.
l (0.5) ' 2.
Check to see if condenser vacuum is established (t 5" Hg).
(0.5) Note: I and 2 above comprise C-9 interlock.
3.
Check to see if any of the steam dump Interlock selector switches are in OFF-RESET position.
(0.5) , b.
The dump will remain armed and a normal Tave/ Tref deviation or a high failure of any temperature channel input to Tave would cause dump ! actuation.
(1.0) Ref. Seabrook Instructional Material, Vol.14, SD, page H0-SD-16,17 Answer 6.12 (/ -) (146)~ .,. g f gg _ g,_V , ' 1.
Generator differential' current .(425) (Ji) 2,--. Generator-overcurrent - - (r25)- -3.
4,16-KV-AC-bu s-f aul t-(,2 5) - f,,4. Mechanical overspeed h2S7 (.1 O 27,
- ' / ' A-
/.. /.. t, s s w r e.
(,7+) . Ref. Seabrook Instructional Material, Vol. 19. E0e, page IIO-EDe-20.
v&C D w.
})c. '/ 71 7 - f.j - ( C 5' % '.2.
' l Answer 6.13 (.5) 10 seconds.
! l Ref. Seabrook Instructional Materials, Vol.19, E0e, page 110-E0e-35.
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l l -
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Category 6 answers continued.
, Answer 6.14 (2.5) a.
The thet:f' S d ef-the CCP'; (-2500 p:!g) f: g reatee-ean-the--- Prc;:ert: r ::fet!:: lift ::t p fat (2485 p;tg) :: c hang ing-f4ew-- eceld :: tine: :: 1:ng :: therd ::!:t: : P.C S. (.5) fl..a fle.v.u.' r c ee Le r-c < p en en /u /b a-p:tb tMth: fec f.I- ((> C) < . b.
The RHRS-to-CCP/ SIP suction isolation valves RHV35 and RH-V36 must be closed.
(0,5) This prevents contaminated containment recirc sump water from getting to the RWST when in ECC recirculation phases.
(0,5) c.
The containment recirc sump-to-RHRP/CBSP suction valves CBS-V8 and CBS-V14 open. (0,5) d.
(1) To balance the injection flows equally.
(0.25) (2) To prevent pump runout and damage after a large LOCA when RCS pressure drops containment atmospheric pressure.
(0.25) Ref: H0-ECCS-60-65, 128 y p, J )il 3 -(A ~ TOS 5'I A ' End of Category 6 Answers.
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. . 7.
Procedures--Normal. Abnormal. Emergency
and Radiological Control (25.0) Answers Answer 7.1 (4.5) ACTION / EXPECTED RESPONSE HESPONSE NOT OBTAINED 1.
Verify reactor trip .[0.25] Manually trip reactor (0.25] a.
ROD Bottom lights lit (0.25] If reactor will not trip, b.
Reactor trip and bypass then manually insert breakers open (0.25) control rods.
[0.25l c.
Neutron flux decreasing (0.25 2.
Verify turbine trip [0.25l Hanually trip turbine [0.251 a.
all turbine stop If no trip, runback turbine (0.25] valves closed [0.25] If no runback, then close HSIV's [0.25] . 3.
Check EFW pumps running [0.251 a.
HD pump running [0.25] a. manually start pump' [0.251 b.
Turbino-driven pump b. manually or locally running [0.25) open steam supply valves.
[0.25] 4.
Initiato emergency boration of RCS.
(0.50] Reference: Seabrook emergency procedurn FR-S.I. Rosponse to Nuclear Power Generations /ATWS.
Answer 7.2 (2.5) 1.
Pressurizer pressure ions than 1850 psig and decreasing.
(0.5) 2.
Containmont pressuro greator than 4.0 psig.
(0,5) 3.
Steamlino ptossuro loss than 585 psig and decreasing.
(0.5) 4.
RCS subcooling loss than 30"F.
(0.5) 5.
Pressuricor level loss than 5% and decreasing.
(0,5) Rof: Seabrook emergency procedure, E00, immodlato Action 4.
Category 7 answors cont inued on next page.
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. . Category 7 answers continued.
. Answer 7.3 (1.0) The SI system should be aligned for cold leg recirculation.
(0.5) r.
b.
Core exit thermocouple temperatures (0.25) exceeding 1200 F.
(0.25) Ref. Seabrook emergency procedure. FR-C.1 and Core Coolant status tree F-0.Z.
' Answer 7.4 (1.0) $.u6hrn D e 7' ContAli1 ment radiation level more than twice the backgrou a.
evel.
(0.5) 'N Ref.
Scabrook Conta t e F-0.5.
b.
Containment ventilation system ao tion valves are closed.
(0.5) Raf.
Seabrook emergoney procedure FR-Z.3. RESPONSE IIGil CONTAINMENT RADIATION LEVEL.
/ / / Answer 7.5 (2.0) 1.
An unexpected rise in any SG narrow range levol.
(0.5) 2.
liigh radiation f rom any SG steamlino.
(0,5) 3.
liigh radiation f rom SG sample.
(0.5)
liigh radiation from any SG blowdown lino.
(0,5) Roi.
Emergency proceduro E0-3. STEAM GENERATOR Tulle RUPTURE, stop 2.
Annwor 7.6 (1.5) 1.
The workorn are given verbal instructions and precautions.
(0.5) 2.
Continuous health physics coverago la providad.
(0.5) 3.
Tho job in documented on an HWP upon completion.
(0.5) llo f Hl'6.1. to<pionting and uso of RADI ATION W0ltK PERMITS AND STANDING RADIATION WORK PEIUtlTS.
C.stegory 7.in'tworn cont inued on noxt pano.
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Category 7 answers continued.
, Answer 7.7 (2.0) , 5(N-18) = 60 REM (0.25) Tokallifetimetodate=58+1=59 REM (0.25) Total lifetime available = 60 - 59 = 1 REM (0.25) Total this quarter available = 3 - 1 = 2 REM (0.25) Use lifetime limit since it is more restrictive (0.25) dose rate = 0.3 REM /HR gamma + (0.060 RAD /HR)(10QF) neutron = 0.90 REM /HR (0.25) 1 REH/0.90 REH/hr = 1.111 hrs. = 66.67 min.
(0.25) Credit for using a conservative quality factor.
( 0.'45 ) Ref.
10CFR20.
..r - Answer 7.8 (2.0) a.
ECP i 0.9% delta K/K rod worth.
(0.5) b.
1.
Insert rods as necessary to achieve 1.3% delta shutdown reactivity as determined from rod worth curves.
(0.25) 2.
Perform a shutdown margin determination.
(0.25) 3.
Recalculate the ECP.
(0.25) ' 4.
Verify boron concentration by sampling the reactor coolant system.
(0.25) 5.
If an error has been found and corrected proceed from the beginning of the startup procedure.
(0.25) ! 6.
If no error is detected notify the operations managor and reactor
engineer.
(0.25) l Ref. Seabrook station operating procedure No. OS1000.07 APPROACH TO [ l CRITICALITY. page 6.
i l ! Category 7 answers continued on nnxt pagn.
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- Answer 7.9 (1.0)
1.
Trip the reactor (0.5) 2.
Stop the RCP (0.5) Ref. Abnormal Proceduro OS1201.01. ABNORMAL RCP OPERATION AND S!!UTDOWN, page 6.
Answer 7.10 ( 1.0 ) A dropped rod.
Raf. Abnormal Procedure )S1210.05 DROPPED ROD, symptoms.
Answer 7.11 (2.5) 1.
axial flux difference.
(0.5) 2.
Group height.
(0.5) 3.
RCS water inventory balanco.
(0.5) 4.
NIS calibration to secondary heat balanco.
(0.5) 5.
Quadrant power tilt. ratio.
(0,5) ~ Eof. Abnormal Proceduro ON1251.01. I.0SS OF Pl. ANT COMPUTER, pago 5.
Answer 7.12 (2.5) 1.
Roscun personnot should be voluntoors or professinnal roscuo personnol.
2.
Rescuo personnot should be broadly familiar with the consequences of exposure.
1.
Women capablo of reproduction should not tako part in thoso actions.
4.
Othor things being equal, voluntoors abovo the ago of 45 should bo selected.
5.
Planned dono to the whoto body shall not excoed 75 R171.
Categorv 7 answorn cont inued on nokt pago.
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i Answer 7.12 continued 6.
Hands and forearms may receive an additional dose of up to 225 REM (i.e., a total dose of 300 REM).
7.
Internal exposure should be minimized by the use of the best available respiratory protection, and contamination should be controlled by the use of available protective clothing.
8.
Normally, exposure under these conditions shall be limited to once in a ! lifetime.
(Any five of the above utght at 0.5 each.)
Ref. Seabrook Faergency Plan, ER4.3, RADIATION PR(YfECTION DURING EMERGENCY l CONDITIONS, FIGURE 3.
Answer 7.13 (1.5) i i l 1.
Manually trip the reactor and turbine (0.25) ! Verify control rod insertion and turbine trip (0.25) ! j 2.
a.
Manually close MSIV's and MSIV bypass valves.
(0.25) b.
Close group A upstream steamline drains at MCB (0.25) 3.
Trip all RCP's (0.5) Ref. Seabrook station operating procedure) 51200.02, Safe Shutdown and Cooldown from the Remote Safe Shutdown Panel, pages 2 and 3.
l
End of Category 7 answers.
t l l l l ! ! I ! l
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Administrative Procedures. Conditions, and Limitations (25.0) Answers Answer 8.01 (3.50) Start back-counting the penalty minutes from 0300 on 07-20-85.
(0.50) At 30% power -- (0300 - 0148) x 0.5 = 36 penalty min.
(0.50) This leaves (60 - 36) = 24 penalty minutes available.
(0.50) At 50% power - (1833 - 24) = 1809 on 07-19-85.
(0.50) The power can be increased at any time 24 hours af ter the above time, or at 1809 on 07-20-85.
(0.50) Ref. Seabrook Technical Specification 3/4.2.1, page 3/4 2-1.
. b.
The limits on AFD assure that the Fo(Z) upper bound envelope of 2.33 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of Xenon redistribution following power changes.
(1.00)
( Ref. Tech. Spec. basis for Tech. Spec. 3/4.2.1.
r n Answer 8.02 (3.00) 1.- All penetrations required to be closed during accident conditions are either: Capable of being closed by an OPERABLE containment automatic f sola-a.
tion valve system (.5) or b.
Closed by manual valves, b11nd flanges, or deactivated automatic l valves secured in their closed positions, except as provided under ! approved specifications.
(.5) 2.
All equipment hatches are closed and scaled.
(0.50) l 3.
Each containment air lock must be operable with: l a.
Both doors closed except when the air lock is being used for normal l transit entry and exit through the containment, then at least one air , ! lock door shall be closed, (.5) ' , t + . l Category 8 answers continued on next pago.
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, , Category 8 answers continued.
Answer 8.02 continued 4.
Containment leakage rates must be limited.
(.5) 5.
The sealing mechanism associated with each penetration (e.g., welds, ' bellows,- or 0 rings) is OPERABLE.
(0.50) Ref. Seabrook Technical Specification definitions, also Vol. 16, ' Containment Structure, page HO-CONT'S-103.
Answer 8.03 (1.50) i When he is informed that it is evident that strict compliance with an UNTESTED procedure co'uld result in personal injury (0.50), or equipment damage (0.50) or the procedure does not accomplish the intended objective.
(0.50) Ref. Seabrook Standing Operating Order No. 85-006.
i ! Answer 8.04 (1.00) , ! Approval and implementation require the concurrence of both a Seabrook
- "
l Station staff supervisor knowledgeable in the area (s) affected, and either the unit shif t supervisor or the shif t superintendent on duty, i r Ref. Scabrook Admin. Procedure SM6.2, page 14.
, Answer 0.05 (1.00) The of f going operator should not allow the relief operator to asstune the shift.
Ruf. Seabrook Admin. Proceduro AQ10.003 Shift Relief and Turnover, pago 3.
l i r f Category 8 answers continued on next page.
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_..,_ - .,. [ Category 8 answers continued.
, Answer 8.06 (1.00) a.
The operator must be replaced within 2 hrs.
(0.50) b.
The shift superintendent calls in a qualified replacement from the spare shift.
If there is no qualified replacement available on the spare shift, a replacement is called in from among the qualified off-duty operators.
(0.50) Ref. Seabrook Admin. Procedure. AQ10.004.
Answer 8.07 (2.50) Insures that the criteria are met for rod ejection from mid powor.
a.
(0.50) b.
Provides protection against local flux peaking which could cause an unconservative DNBR to exist during rod drop accidents.
(0.50) c.
Providos core protection to prevent DNB for all combinations of pressure, power, coolant temperature and axial power distribution.
(0.50) - d.
Provides assurance of fuoi integrity.
(0.50) .. Protects the reactor from loss of he d sink in the event of a e.
sustained steam /foodwater flow mismatch resulting from loss of normal feedwater.
(0.50) Raf. Seabrook Safoty Limits and Limiting Safoty System Settings.
Answer 8.08 (1.$0) At least two of the following thren boron injection flow paths shall bo OPERABLE: (0.50) The flow path from the boric acid tanks via a boric acid transfor pump a.
and a charging pump to the RCS.
(0.50) b.
Two flow paths from the RWST via charging pumps to tho RCS.
(0.50) Rof. Soabrook Toch. Specs.. LCO'S, Roactivity.
, Category 8 answorn continued on noxt page.
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. , Category 8 answers continued.
, Answer 8.09 (2.50) As a minimum, the following A.C. electrical power sources shall bo OPERABLE: a.
Two physically independent circuits between the off site transmission network and the on site class 1E distribution system (0.50) and b.
Two separate and independent diosol generators (0.50), each with: 1.
Separate fuel day tanks.
(0.50) 2.' A separato fuel storage system.
(0.50) 3.
A separate fuct transfer pump.
(0.50 Raf. Seabrook Tech. Spec. LCO'S, T.S. 3/4.8.1 A.C. Sources.
Answer 8.10 (1.50) 1.
Ensure adoquate shutdown margin.
(0.5) '. Minimizo effects of a rod ejection (either for accident or limit the
potential offects of rod misalignment on associated accident analyses).
(0.5) 3.
Ensure that acceptable power distribution limits are maintained.
(0.5) Raf. Seabrook Instructional material, Vol. 14. TI, pago 110 TI-221 also T.S.3.1.3.6 and its basis.
Category 8 answors continued on next page.
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. . . . ' . . Category 8 answers continued Answer 8.11 (3.00) a.
1.
The intent of the procedure is not compromised.
(0.50) 2.
There is no note in the procedure stating that the steps mest be performed sequentially.
(0.50) b.
1.
The applicability is stated in the body at the procedure (U.$0) or 2.
The responsible department supervisor identifies and appropriately initials the non-applicabic steps that are not to be performed.
(0.50) c.
1.
The shift supervisor authorizes a deviation from the lineup by initialing the change and entoring "NA" in any sections that are not to be performed.
(0.50) 2.
The reason (s) for the deviation are indicated on the lineup.
(0.50) Ref. Soabrook proceduro SM6.2. Station Operating Procedure, page 12.
Answor 8.12 (2.00) 1.
To ensure that the temperaturo difference among the loops does not exccod 25 F.
(1.00) 1.
To assure that thorn in a rolluf path from the reactor coolant loop to the RilR suction lino roliof valvon when the RCS is at low pressuro (<500 psig).
(1.00) 4.
Ref. Soabrook. Vol. 13. RCS. pages 46 and 47.
Category H annworn continued on noxt pago.
o' . . ..
c . -. _,_ _ _, _, - , ' . ' .. . . Category 8 answers continued.
.
. Answer 8.13 (i.00) State authorities - 15 minutes.
(0.50) NRC - one hour.
(0.50) Ref. Seabrook Emergency Plan.
End of Category 8 answers.
End of Examination '.e ,w l .. - }}