ML20235W388
| ML20235W388 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/13/1987 |
| From: | Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235W361 | List: |
| References | |
| 50-416-OL-87-01, 50-416-OL-87-1, NUDOCS 8707230545 | |
| Download: ML20235W388 (173) | |
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l1 i I E_NCLOSURE 1 EXAMINATION REPORT 416/OL-87-01 Facility Licensee: System Energy Resources, Inc. Facility Name: Grand Gulf Nuclear Station Facility Docket No.: 50-416 Written and operating (oral and simulator, examinations were administered at the Grand Gulf Nuclear Station near Port Gibson, Mississippi. J Chiof Examiner: c.e , R :YW ? [ulv_/987
', ' cyg' r,a n Ken E. Date signed Approved by: .h I _f. [( 7 John F. Mun , Section Chief Date Signed Summary:
Examinations were administered on May 18 - 23, 1987. Written and operating (oral and simulator) examinations were given to eight Reactor Operators (ROs) and two Senior Reactor Operators (SROs). Seven ROs passed the written examination; eight Ros passed the a al e x -a mi n a t i on a r. d eight ROs passed the simulator examination. One SRO passed the written examination; two SROs passed the oral examination and two SROs passed the simulatcr examination. Based on the results described above, seven of eight ROs and one of two SROs passed the overall examination. Of che twelve questions on the written examination which required NRC resolution because of technical errors in the answer, five (42%) were due to ircdequate/ insufficient / inaccurate information in l the material provided by the facility for the examination l preparation. 8707230545 870715 PDR ADOCK 05000416 V PDR _______a
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- u REPORT DETAILS
- 1. Facility Emoloyees Contpcted:
*. C..R. Hutchinson. Site General Manager
- K. E. Beatty. GGNS. Training Superintendent
$: W. M,. Shelley, GGNS. Operations Training Supervisor D. Bottemiller, GGNS Simulator Instructor V. S* airs, GGNS Simulator Instructor J. B,ister, GGNS Simulator Instructor I
- Attended Exit Meeting
- 2. Examiners:
l 1: *. K. E. Brockman, Region II D. C. Payne, Region II F. W. Clark, Regior. III M. O. Bishop, EG&G, Idaho
- Chief Examiner
.3. Examination Review Meetinq:
At the conclusion of the written examination, the examiners provided your training staff with copies of the written examinations and answer keys for review. The comments made by the facility reviewers are included as Enclosure 3 to this report. The NRC resolutions to these comments are listed below,
- a. RO Examination
- 1) Question 1.01 NRC Resolution: Comment Accepted. The clarifying inform & tion provided by the proctor during the examination was, inaccurate, in that the mode of operation of the Recirculation System should have been Flux Manual, vice Flux Auto. Due to this error, the question will be deleted from the examination. Section and Total points will be adjusted accordingly.
- 2) Question 1.02 a.
NRC Resolution: Comment Acknowledged. While the term
" Gross Enthalpy" is not specifically referenced to in the Grand Gulf training material, it is not a technically unique term and should not have posed difficulty to the candidates. In the future, efforts will be made to ensure that plant-specific terminology is referenced; however, generic terms will continue to be used, as appropriate, in the examination construction.
- 3) Question 1.02 b NRC Resolution: Comment Not Accepted. The candidate can know the processes which takes place between the HP and LP turbines, while not knowing their effects upon the enthalpy of the_ cross-over' steam. Thus, the answer to the-(a)1part of the question is not inherent to being able to' answer the (b) part. Additionally, as.is the practice of Region II for questions such as.this, the concept oi " error carried forward" will be applied to the grading of the question. No change'to the exam or the answer key is required.
- 4) Question 1.03 NRC. Resolution: Comment Accepted. The diversity of effects described by the utility does support the potential for available NPSH.to either increate or-decrease, if integrated response is considered. The l answer key will be expanded _to allow for either response !
L (but not " remain the same") as a correct answer. 1
- 5) Question 1.04 l NRC Resolution: Comment accepted. Increased injection.
l time is an alternate, acceptable answer and will be l added to the answer key as a full credit response..
- 6) Question 1.09 b NRC Resolution: Comment not-accepted. This questjan is not considered to be double jeopardy, since the concept of " error carried forward" can, and will, be applied to +"a grading. Additionally, the question is specifically related to the utility's learning objectives. The utility requires that their students be H aole to identify the initiating cause of the displayed transient and be able to explain the reason for each of the parame tr ic responses. No change to the exam or answer key is justified.
- 7) Question 2.10 a L NRC Resolution: Comment accepted. The answer key will be modified to indicate the correct pressure setpoint of 125 psig. The utility training material should be upgraded to reflect the correct setpoint.
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- 8) Question 2.15 b ];
-l NRC Resolution: Comment accepted. The utility )
-centention Lonterning the requirement f or addressir.g the: !
potential; effects >on 3BGTS Train A.is accurate. The. question only elicits the actital-system response, not potential:enes. 'The answer key will-be changed to "GBGTS Train A will stop." The Total and Section poirts will be adjusted accordingly.
- 9) Question'2.16 b l NRC' Resolution: Commen t accepted. Setpoints'were not required as part of the. correct answer, The 100 gpm setpoint-for di'scharge flow will be deleted es required
'information. Additionally, based upon a post-examination review of the responses of the candidates, the plant specific terminology of " low dilution flow", was added to the answer key as an alternate. response for " Low PEW /CW Flow."
- 10) . Question 3.02-1
'NRC Resolution: Comment Partially accepted. The information provided by the facility is technically accurate. The requested additional responses are different specific signals which can actuate the ;
Recirculation Flow Control Valve Motion Inhibit 1
-Interlocks. Because of the potential for'the candidate to address the question in greater detail than was requested, the specific input signals will be accepted as a correct r e r,p o n s e , equivalent to the. respective interlock which-they actuate. The answer key will be modified as follows:
Alternate responses for Hydraulic Power Unit Failure will ben
- 1) Hydraulic Fluid Tank LoLo Level
- 2) Hydraulic Fluid Tank HiHi Temperature Alternate responses for Analog Control Circuit Failure will be:
- 1) Analog Circuit Demand High
- 2) Velocity Feedback Rate of Change High
- 3) Position Feedback Rate of Change High
- 4) Velocity Controller Oscillations
- 11) Question 3.03 a NRC Resolution: Comment accepted. The reference to the 15 minute Time Delay in the answer key will be removed.
Grand Gulf needs to change the training material to more accurately explain how this interlock functions.
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- 12) Question 3.05 b NRC Resolution: Comment accepted., The answer key will
.be changed to use the Grand Gulf terminology of -High Power Setpoint. Additionally, the alternate wording.for the correct response, as described in OP-CSI-.3-02,.will.
be added to the answer key. J3) Duestion 3.12 c NRC Resolution: Comment accepted. Per. Figure '7 of OP-C51-3-02,. response c.6 will be changed from
" Recirculation Flow Centrol Unit" to " Meter Selection Switch."
14.) Question 3.14 b NRC Resolution: Comment accepted. Typographical error corrected to: reflect the proper meter range as being from <
-160"'to +60".
- 15) Question 3.16 c lNRC Resolution: Comment accepted. The system operation described.by LP-LO-SYS-LP-N32-2 is specific in showing
- that the speed control circuit remains in the control scheme'when the turbine is tied to the grid. 'The answer key will oe modified to "None", _ indicating that no control systems are deleted-from the control scheme when the Main' Turbine Output Breakers are closed.
'16 ) Question 4.02 b NRC Resolution: Comment Accepted. The third action listed in the answer key is, in fact, not-unique to the conditions specified in the question. Instead, it is a c c,mmon action nr d , as such, is not elicited by the question, as stated. The answer key will be modified to reflect the deletion of this response; the Section and ,
Total points fer the exam'will adjusted accordingly.
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- 17) Uuestion 4.03 d NRC Resolution: Comment partially accepted. The alternate correct answer of " Inadequate Cooling of RCIC Lube Oil" will be added to the answer key. Grand Gulf needs to modify their training mater ial to include this i purpose in their lesson plans. '
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1Withfrespect to the utility-contention that this'part of the question is " double jeopardy" in that it depends'upon the candidate's-ability to answer parts a, b, and c, we do not accept the recommendation for deletion. The candidate does not need to.know the setpoints requested in' parts a a c to be able to answer what the basis for having the protection is. " Error carried forward" is not-
-required to be' applied by the examiner in this instance.
No change to the exam or the answer key.is required. 18)' Question 4.04 a NRC Resolution: The question does not'suggest that the containment pressure exceeds 17.25 psig. However, it is possible that the candidate could assume that._the point of the question was concerning such higher pressures. If the' candidate states that pressure conditions in excess of 17.25 psig will result in " containment venting", full credit will be given. Since this is in compliance with.the procedure as it is' written, this credit would have been given whether.the comment had been made.or not. If.the candidate does not specify his assumptions, then par tial- credit (50Y.) will be allocated for the response.of " containment venting."- The answer key will' be modified'to reflett'the alternate. response and its grading. criteria,
- 19) Question 4.04 b NRC Resolution: Cumment partially accepted. The specificity delineated in ,_.1e answer key is neither appropriate nor elicited by the question. The individual respunees which were first listed will be changed to "No assurance of adequate core cooling can maintained if'the-containment fails (0.5)." Total and Section points will be adjusted accordingly, it should be noted that the Grand Gulf training material lists the responses originally contained in the answer key. This material should be modified to reflect the appropriate level of specificity. >
The contention concerning double jeopardy is not accepted. The utility's recommendation is based upon the same logic as that in Question 4.03 id). For the same reasons as presented in the resolution to Guestion 4.03, the exam and answer key will remain unchaqqed. t-
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~20) Question 4.05
.NRC Resolution: Comment partially accepted. .The utility's-contention that the answer key is' inaccurate, as ;orapared to . the 'speci f ic action c'eps. listed.in procedure EP-14 is true. . However,.the lesson plan which .
teaches the candidates the way to apply the procedural steps does. list two' system's whose isolation must be bypassed.to use the Main Condenser'for. cooldown/depressurization. Due to the inconsistency between the utility. references and the fact that the question specifically asked for two bypasses to be listed, the question will be. graded. individually for.each candidate. 11,the candidate responds as per the lesson: plan (MSlVs.and IA) the question will be retained and-graded; 11.the candidate responds as per the procedure (MSIVs)-and then " searcher,".for an additional response, the. question will be deleted, so as'ntt to penalize him unjustly. Section and Total points for the exam will be
. adjusted appropriately.
- 21) Question 4.07 NRC Resolution: Comment accepted. The answer key will be modified to accept the plant specific terminology, per Procedure 03-01-01-1, 22)- Quest' ion 4.09 NRC Resolution: Comment accepted. The specific criteria implemented at Grand Gulf will be made the principal answer on the answer key; the BWROG guidelines, which provide the more generic responses to the question will be annotated as acceptable alternate answers. The comment concerning the answer to part (b) of the question requiring information not specifically requested is acknowledged. The specific statement "because they are being used to control reactor power" was not intended as required wording. The answer key has been appended to denote this fact more clearly.
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- b. SPO Examination. l 1 3 1
- 1) Question 7.10 i NRC' Resolution: ' Comment accepted.- If the candidate-specifies that-the radiation war kor may be accepted-if-the_ Radiation Protection Manager or the General Manager of the site approves.'his exceeding the administrative limits, full credit will be given. If these managerial approvals are not specified, only half credit will be-give7. The answer key will be modified to reflect the alternate / partial. credit responses.
- 2) Question 8.02 NRC Resolution: Comment accepted. The additional correct response will be added to the answer key, based
- upon the recent procedural change at the site. The Training Department should ensure that,.for future examinations, procedural-changes are forwarded to.the Regional staff as they prepare the wri tten examination so so as to prec.lude. future occurrences like this.
- 3) Question 8.03 NRC Resolutions Comment accepted. " Breaker fuses
-installed" will be added to the answer key as an alternate. response for " control. power a v a i l a b l e ." Based upon a Regional review of the candidate's answers,
" Breaker racked in" was added as a f ourth correc t I response to the question. Individual responses will be
-weighted to require three of the four correct answers at 0.50 points each.
- 4) Question 8.07 NRC Resolution: Comment accepted. Based upon the comments provided by the utility, there appears to be no correct answer (multiple-choice format) for the question.
It should be noted that this question was asked on the Requalification Examination given at Grand Gulf in December 1986. The question was reviewed for technica) and psychometrics considerations and was not commented upon; both Commission and utility staffs should work to improve their quality assurance reviews'of examination items. The-question will be deleted from the examination; Section and Total points will be adjusted accordingly.
- 5) Question 8.15 NRC Resolution: Comment accepted. The additional correct response will be added to the answer key.
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- 4. -Exit Meeting i
At the conclusion of the site visit, the examiners met with l representatives of the plant staff to discuss the results of the examination. There were three observations concerning the material
-submitted by .the utility for the preparation of the examination which were shared with the site management:
a) The training materials on Heat Transfer and. Fluid Flow, which were sent to the Region II and Region-III examiners, were'different. While this resalted in no difficulties in the exam deselopmenti it could be indicative of the trainees being' held to different standards within the same training program. b) No Reactor Theory or Thermodynamics l esson ' p'l ens were.provided to the' Region 11 examiner. . Text material was provided. c) At least two procedures (Conduct of Operations, Protective Tagging) were sent to the Region II examiner with only the even numbered pages included. Care needs to be taken to ensure that the material sent to the examiners is both complete and accurate. This is essential to ensure that the examiners can become as technically proficient on the plant as possible, and to ensure that the candidates are not penalized due to an erronecus omission of availab'le information. There were two items concerning the performance of the simulator during the examination process which were brought to the attention of the site management: a) When IC-17 initialized, Valve F-046 is OPEN vice CLOSED: CRD Cooling Water dP is approximately 23 #;
*BCCW Pump "C" is not in Standby, b) Many performance parameters (e.g., RCIC, LPCI) are not simulated in such a manner that the systems can satisfy the 151 procedural performance requirements.
While these items did not challenge the integrity or the validJty of the exams, they did produce additional situations for the candidates to deal with. Additionally, they extended the length'of the examinations as the candidates had to resolve these administrative problems. Care should be taken
to ensure,that such problems are identified and corrected; th
-cer ti f ica t ion process To' the. simulator that is required by the new revision to 10 CFR 55 chould correct these shortcomings.
There'Were three generic observations concerning.the performance of the-candidates which were made by the examination team: a) Three candidatos paralleled the Diesel Generators to' the' electrical busses with excessive voltage on che
. incoming line. In all cases, the KVAR meter;was pegged by the operation, as opposed to the small eactive load which should have been placed on the machine. The problem appears to be both performance (attention to detail) and equipment (incoming voltage meter resolution).
b) There was confusion displayed by many of the candidates'with respect to why there is a minimum temperature for the. Reactor Pressure Vessel. One
' candidate stated that the bottom of the RPV would fall off if the temperature decreased to less than 60 deg F. This indicates a weakness in their training in Material Sciences.
c) Communications were improved compared to past ; examinations conducted at' Grand Gulf. This applied
'to both intra-crew communications and to the )
communications with plant personnel. Since only'two SROs were examined during thin trip, there was not a j sufficient basis to determine whether the use of the Emergency ,
-Operating Procedure Flowcharts had significantly improved from the I l
previcas examination trip (Examination Report 50-416/OL-86-03). This performance indicator will. continue to be an important evaluation point for future examinations. The. licensee did not identify as proprietary any of the material provided or reviewed by the examiners. i i
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.E , , U. 8. NUCL,inik REGUL4iURY LOdMIb510N REACTOR OPERAI UR L ?. CENSE EXMHNA T ION FAC1L1TY: GR AN D _ G yld _ J,. _ . _._.. .._ _ I REACTOR TYPE: ..L M.t Gli6 ._...__ _ _._ _ ___ . DALE ADMINISTERED: 137 4 0Sz i,0_ _ _ _ . _ _ _ E uaM1NER: _CLAIL _F. CAND1DAIE: ;'_.RS $.lE._R _ _ _._ _ . .. WSTBUC1.lO[@_lO 3/R@) DOE: U s.e separat e paper for the answers. Wrlie answers on one side only. Staple quest)on' sheet on top of the answer sheets. Points for each question are indicated in parentheses after the questaon. The passing l grade requires at least 70% in each category and a fanal grade 04_ at least 80%. Examination papers will be oicked up six (6) hours after the examination starts.
% OF CAlEGORY % OF CANDIDAlE*S CATEGORY
.MAlU!i ._IOlOL __ _ S C O R E ., ,_ .. VALUE _. .l @I fl G O B L . , . __ ._ _ ._ __ _ _ __ .,..
16.5 0 PRINCIPLEG OF NUCLEAR POWER
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PLANT OPERATION, THERMODYNAMICS, HEAT IRANSFER AND FLUID FLOW 2~' OO
. 9.. ?S .] 2 - _ _ .._
- 2. PLANI DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS Cl OD
, . k, ,23,.73 _ _ _ _ _ _ _ _ . _ _ . , . . .
J. INGlRUMENTS AND CONTROLt?
'25 5D
_ .m..t.d - lhlZ _... _ ._._. _ _ - . _ . . . _ _ . . . .
- 4. PROCEDURES - NORMALS ADNDRMAL, EMERGENCY AND RADIOLOGICAL CONTROL 1 1 y . 2 3.. _ . _ _ . _ . , _ . . _ ...__._.__..% Tata1s Final Grade al1. wark done on th2s e :ami nat 1 on an my own. I have neither g1ven nor rece2 ved aid.
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Candidate's. Signature
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.'NRC RULES AND BUIDELINES FOR LICENSE EXAMINATIONS 7, t , Dur;ing the administration of this ex arna nat i on the to11cwang rules apply:
1, Cheating on the exarna nati on means an aut omat i c denaal of yG application and could result in more severe penalties.
- 2. Restroom tr: 1ps are to be limited and only one tandidate at a time may leave, You must avoid al] contacts with anyone outside the examination i
room to avoid even the appearance or possibility of cl.eati ng .
- 3. Use black ink or dark pencil onl,y to facilitate legible reproductions.
- 4. Print your name In the blank provided on the cover sheet of the.
' examination.
- 5. Fill in the.datn on t.he cover sheet at the examination _tif necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper rlqht-hand corner of the first page of eett!
settlon of the answer sheet. S.- Consecutively number each answer sh eet ,, write "End c C'at eg or y , , _ " as appropriate, start each category on a new page., wri te tjnly on one g1_de of the paceri, and wr i t e "Last Page" on the last answer sheet.
- 9. Number each answer as to categcry and number, tur example. 1.4, 6. 3.
- 10. Skip at least ttiree lines between each answer.
- 11. Separate answar cheets from pat and p l ac e f i ni shed answer sheets face i down on vour desk or taole. ..
- 12. Us? abbreviations only'If they are toomonly used in facility literature.
- 13. lhe point value for each question 15 indicated in parentheses after the ;
questien and can be used as a cuide f or the depth of enswer required
- 14. Show all c a l cul a t i on s ., methods. or assumptions used-to obtain en enswer to mathematical pr ablenis whether Indicated in the question or not.
- 13. Par ti al credit may be given. Therefore, ANSWER ALL Pi408 OF 1HE QUEST 1UN AND DO NOT LEAVE ANY ANSWER BL ANK ,,
lo. l 't parts of the e.c ami nat i on are not clear as to intent, ask questions of the eniemlrler on1y. 1 7 ., You must sign t htt statement on the cover sheet that i ndi c at es that the work is voor own and you have not received or been given assistance in compleL2ng the exem2 nation. Ih1s must be done a4ter the e: amination has 1 been comp]eted, i 1 i
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;18. When you' complete your examinaF_ ion, you shall:
- a. Assemble'9aur examination as fol1ows:
([) Exam.queJ,tions.on top. ; (2) Exam aids ~ fiqur.ws, tables, etc.
. (3)- Answer pages. including f a c,ures which are part'of the answer.
- b. 'Iurn in your copy .of the examination and al1 pages-used to answer 3 the . ex ami nati on questions,
- c. Turn in all scrap paper.and the balance of the paper that you did 3 not use'for answering the questions. I i
- d. Leave the examination area, as defined by the examiner. 'I f after I leaving, you are found in this area while the examination is still in. progress, ynur license may.be denied or revoked. 'l i
Ic.lBRGIELEDR NLM;(EGB_EOKB_ELeULgEEB811% PAGE 2 V ,- . IU?BUDDIUBt11Ch _UEBLISBMSEE B _SUD_E LU I D_E LOhi 4 r e n 9 1 ) t n i_n - n r
- 11 , ' I J b JI I el W n.'d 1 h . 40.swI [
INDICATE if the below ;11sted parameters will INCREASE., L REASE or REMAIN THE-SAME, i f, the fac112ty experiences a "J ET PUMP 4 is HEAD HOL.DDOWN' FAILURE" (ASSUME.'THE FAILED. JET PUMP 15 NOT utLY' INS 7RUMENTED)
- a.- . Flow for the Failed Jet Pump PC [
- b. Core D2 f f eren ti al Pree ~t r e
, _ c. Reactor (APP' - Power
- d. In . ated Core F1ow
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Loo DUESTION 1.02 (W) CHOUSE pither INCREASES, DECREASES, or-REMAINS THE:SAME to-+i11 the bl anks of the following statements 13etween the H. P. Turbine exhaust and the L. P. Turbine inlet, gross
- a,'
enthalpy of the steam (1) _ _ , , _ , . _ and specific enthalpy of..the steam (2) _ _ , , , , _ . _ . b ., JUSTIFY your answers to (1) and (2) above. 2.OC) QUESTION 1.03 (+M)
- a. Define Net Positive Suct1on Head (NPSH).
- b. Opening the Rec 2rc System Flow Control Val ve (FCV) will cause the evai1able NPSH for the Reca rc Pumps to (INCREASE, DECREASE, or REMAIN THE SAME)?
- c. Will the requ2 red NPSH for the Retarc Pumps (INCREASE, DECREASE, or 1 REMAIN 1HE SAME) when the Recire fumps are shiited from slow speed (15 Hz) to fast speed (60 Pz)7 i
DUESTION 1.04 ( .50) l - l S1 ATE whrt could happen if the Care Spray Pun ps started with the Joi_ key 1' umps shut down? l (***** CATEGORY v1 CONTINUED ON NEXT PAGE *****)- l- ! i
A__EBJUCleLgs;oE_uyCLEAR EQWE8_ELONLQCKBOIlOth PAGE 3
'- + - : IUEBt1QQ Y BOtil%_ U E GLIB 6t4S EEB_8U D_ E LVI D ,E L OW 1
f ouEs11oN. 1.os- -(1.00) A temperature instrument':with an out-of-date sticker on it is reading 400; degrees F. A recently calibrated pressure gage, senstn'g the same areai andicates 350 psig. CALCULATE the temperature which should, be indicated on the temperature instrument ? (SHOW ALL WORK)- ( A%% u fn E A ~a A7 O R.AT Cb T>N % T E.WQ QUESTION 1.06 (1.00) LIST fWO (2) . conditions that 3imit power operations per the Technital Specifications to 25% o~f rated thermal power?- QUESTION' 1.07 (1.00) STATE the~ mode (s) of heat transfer.for the f ollowing si tuations:
- a. Center of f uel pellet out to the pellet ~ edge '
- b. Across the Helium gap in the fuel rod.
c; . C1ad surface ta the center of the coolant channel.
- d. Gladr uurf ace to coolant under 4 21m boiling conditions.
UUE5110N 1.08- ( .50)- l
-GTATE the RELATIONSHIP between MAPRAT~& MAPLHOR? I i
l ( l i i 4 (***** CATEGORY Di CONTINUED ON NEXT PfiGE *****) 4 I w: . .
1[aC81NCIELES OE_.UUGLE80_EONEB_EL6NI_QEEB@lJOh PAGE' 4 1'- . . . IUghttODYNOt11Ch_UEOI_lBOUpEEB GliD_ELU.ID_ELgN
, 1 4 -
QUESTIUN':1.C9 (2.SO) The attached Figure #2 illustrates a transient that could occur at a'DWR. Given2 (1) No operator-act1ons are taken (2)- Normal functioning of instrumentation, control and-protection systems (3) ~ Normal functioning of the Reactor Protection System
- a. IDENTIFY the initial cause of this event.
- h. EXPLAIN the cause, and actuation s2 gnal (s) if appropriate, of the fallowing ~ recorder indications:
i (1) - Reactor pressure increase f rom time t =' -1.0 set to time t -= 3. 5 set (Graph 2) j (2)- Initi al decrease in neutron flux i ram' ti me t. == O to l time t = '" ' 2 . O s e c (Graph 1) i
.( 3/ Decrease 4n V'essel Steam flow 4 rom time't =
- 1.' s e e to time t = " " 1.6 sec-(Graph 3)
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(4) Increase in Vessel Steam flow from time.t = 3.4 set to time t = ' 3.75 sec (Graph 3) DUESIION 1.10 (1.00) i V#och of the f ol l owi ngj radi oacti ve i sot. opes if found in' the reactor '
. coolant, in'significant amounts, WOULD NOT indicate a leak through the fuel cladding?
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- a. Co-60 l 1
p.
- b. Xe-133 l
- c. 1-131 !
,d. Kr-87 1
i l I (44t#* CATEGORY 01 CON TINLIED ON NEXT PAGE *****)
h EBIUglELES,_gE,_Nyg6EGB,,,EgWEB_EL6NI,OEEBOIlgdt -PAGE 5 4'
'10EBUQDYt@NICh_UE61_IBONSEEB_6MD_ELVID_EL99 l'
DUESTION 1.11- -(1.00) WHICH ONE'of the followina is NOT' a characteristic of Subtritical Mul ti p l i c atii on?
- a. The subtritical' neutron level is directly' proportional to the neutron' source strength.
.b. Doubling the indicat ed count rate by reactivity cdditions will reduce the niargin to criL1cality by'approximately one-half.
- c. For .. equal reactivity addations, 1t takes longer for the new equilibrium count rate to be reached as Keff approaches unity.
d -. A single notch of rod withdrawal will produce an equivalent'
. equilibrium count rate increase whether Keff is 0.88 or.O.92.
-DUESTIOl= 1.12 (1.00)
.Ekeacti vi t y is defined'as WHICH of the fallowing?
a '. The ratio of ' the number of neutrons at some point in this l ~ generation to the number of neutrons at the same. point in t.he previous generation.
- b. The -fractional change.In neutron population per generation.
c.--The factor by which neutron population changen per generation. l
- d. The rate of change of reactor power in neutrons per second.
(***** CATEGDRY 01 CONTINUED ON N E X'l PAGE 4A***)
LIr _tBlWCJeLEs_gE jup6E68_POWEBIELGUI_QCEBGl] Qui - ' PAGE 6 4 i a sIDEBOODXUOMich _UE81_IBGUEEEB_8ND_ELLflp_ELQW-6
- DU'ESTION- ' 1.13 ' (l.00)
KDur ing a REACTOR START-UP, the reactor operator.is sequentially 2 withdrawing control rods to make the reactor critical. WHICH of the f ollowing statements correctly describes the SRMs-response?
~
- a. It will take a shorter length of time for the SRM count rate { '
to' slowly Increase to a new ctabl e count rate. b '. It' will take the name amount of time for the GRM count rate to slowly increase to a .new stable count r" a t e . ,
- c. It will take a longer length of time . f or the GRM.. count rate to slowly increase to a.new stable count rate. ..
- d. No change will be observed in the SRM count rate'until the !
reactor is' critical. { QUESTION 1.14 (1.00) The reactor trips f rani iul1 power, equi 1ibrium Xnnon conditians. Four p (4) hours l ater -the reactor i s' brought critical and power level is ! maintained on ranget (5) of the IRMs for several hours. WHICH of the f ollows no statements' is CORRECT concerning control rod motion during this. period?.
-i
- a. Rods will have to be withdrawn.due to Xenon build-in.
'b. Rods.will have to be rapi dly inserted since the critical
~
1 reactor will cause a high rate of. Xenon burn-out,
- c. Rods will have to be inserted since Xenon will closely follow its normal decay rate.
d.' Rods will approximately remain as is as the Xenon establishes 1tu equi 11brium value for this power 1evel. l l l l ll E l, h L L (***** CARGORY 01 CONTINUED ON NEXT PAGE *****)
1.!__CBlyglELES._OE_UUGLEGB_EOWEB_EL881_O[EESTIQUs PAGE 7 e + - IUEBOODXN6L11Ch_UE6 LIB 6NSEE B,_GUD_F LUID_ELOW DUESTION ' 1.15 . (1.00) Which DNE of the f ollowing correttly describes the format 1on and removal' processes at equilibrium atom density for Xe-13S?
- a. (FP + Iodine burnout) - (Xe decay + Xe-135 burnout)
- b. (FP + Iodine decay) -- (Xe decay + Xe-136 decay)
- c. (FP + Cesium decay) - (Xe decay + Xe-135 burnout)
- d. (FP + lodine decay) (Xe decay + Xe-135' burnout)
DUESTION 1.16 (1.00) Consider the equati on below and answer the following: S CR = count, rate of neutrons CR = - - - - - - S = source strength 1 -- K N == Keff
.a. WHICH term (s) betermine(s) the tict al neutron. production RATE?
- b. CHOOSE ONE. During a startup, with a Neff ( 1, succeeding
-generations of neutrons will EINCREASE/ DECREASE ~l
~
in population, I at a EINCREASING/ DECREASING] rate. 4 DUESTION 1.17 (2.OOs j Answer EACH of the following TRUE nr F AL.SE . ,
- a. Void Coefficient i n di rectl y proportional to core sace,
- b. The l ar ger the core, the lower the relati ve neutron leakage.
- c. The two isotopes Lhat make the l ar gest contributaan to the Doppler effect are U-2.55 and Pu-240.
- d. Durinq star t-up pha'es, s the effect of the Temperature Cort f i ci ent i s Insi gnificant in contral1ing power when compared t r- the effect5 of the Doppler and Void Coef f i ci ent s.
(*Atit UATEGORY 01 CONTINUED ON NEXT I' AGE 4/$44) I
li___ E81SQlCLf; SiOE _UQCljiOO,20WER_EL 901,. OEEBGI LOU 1 PAGE 8
- .- . L IUEEMODIU8dlcL20Es1_1 Bat 4;;Eng _eyD_ELulp_ELoy QUEST 1CN 1,18 (2,50)
Arwer EACH of . the f oll owing TRUE or FALSE.
- a. Xenon and Samar i um . concentrat2 ons i ncrease f ollowing a scram 4 rom high power oper at i on (within the farst five hours).
- b. Samarium has a higher microscopic absorption cross section-than Xenon,
- c. A reactor star t-up soveral days af t er a scram from extended high. power operation is consadered to be Xenon and Samarium ireu,
- d. The equilibrium concentration of Samarittm at 50% power is app r"ox i ma tel y the same as at 100% power.
- e. The equilibrium concentrate on of Xenon at'50% power is approximately one-half the equilibrium concentration at 100% power.
QUESTION 1.19 (1.50) MATCH each of the items listed in Col umn 1 to the most correct chart def)nition oiven in Column 2. COLUMN 1 COLUMN 2
- e. Prompt neutrons 1. Neutron population is-self-sustaining and constant.
- b. Shutdown margin
- 2. Occur i ndirectly f rom fi ssion thru
- c. K-excess fission fragment daughter decay.
- d. Del ayed neutrons 3, 1.443 times the Doubling Time,
- e. Reactor period 4. Measure of availability of f uel in the core above that needed for
-f. Critical initial criticality.
- 5. Occur directl y f rom fission reaction.
- 6. 1-Keif
(***** CATEODRY 01 CONTINLIED ON NEXT PAGE t4***)
1! .j]
- l'. ' PBLUGICLES_QE_ UUGLEGB,.20 WEB _PL6bil_QPgRGIlgljt PAGE' 9 C -* - LURBtlRRYUeb1Ch_UEGI_IBOb!gEEB OND_ELylp_ELOW' DUESTION '1.20 f1.50)
LIST;three (3) _ factors upon which a reactor's decay' heat generation r.a t e i s ,- dependent . QUESTION '1.21 (2.00) Use the 1/M plot and pred'ict the number of control-' rods. r equired to be withdrawn to, achieve c r i t i c al. i t y . NCirES . 1. .CR = Count Ra t.e
- 2. USE THE' FIGURE DELOW TO SKETCH'YOUR SOLLITION-CRO '= 40 cps- CR4 = 191 cps "'
CR1-= DO cps. CRS = 333 cpc CR2 = 89 cps CR6 = 800 cps l CR3 = 129 cps Each CR reading is recorded following a 5 rod withdrawal'with CRO representing 1007. rod density. 5 10 15 20 25 30 35 40 45 50 55 1,o;_.__ ;_- _; _j__-.:_._._.,;_ ___j _...;._.___;._.- _; __i. __;_..g'i,o l l .. 1" O . '9 - -O,9 O .' a - -0.a O.7- -0. 7
'1/ M O. 6-- -0.6 O.5- -0.5 O 4- -0.4 O. 3- -0.3 O.2- -0,2 O .1 - -0.1 o; . _ ~_ }',- _.-; _.;.-_..._; ___ ;_ ___ _; ._._._ ; -.__.__ ;-- - _._;_. .'1 ; _ __ _ ; _ _ ._ l 0 5 10 15 2
'0 25 30 35 40 45 50 SS' Control Rods Withdrawn
(***** END OF CrVIEGORY 01 *****)
- 2. PL AIJT DESIGN INCLUDIlJG GAFETY AND ENERGENCY '3 Y 5'1 E M'3 PAGE 10
' i
) DUEE110N 2.01 (1.00) WHICli of tho f ol l aeo rio is the orily normall y CLOSED val ve in the RCIC steam supply flow path in the at power ' standby lineup? l 1 I
- a. Steam Supply Va]ve ( Foil 5 )
l
- b. Outboard Steam 1 sol s t 1 on Va]ve ( i~ 06 4 ) !
- c. Iur b1 ne Trip 1hr at t1e Vaj ve
- d. T ur b i n re Governor Valve D U E 0 T,1 O N 2.02 (2.UO)
Consider an O f f - Nor mia l Event in which Instrument Air System pressure is 1 oc:t . tTr%% E' hcQ the followino valvet FAIL. ( C L.05 E D . OPEN, or AS IS)
- a. ' -l
- 1. CRD Flow Control Valve
- 2. RFF' l~li n i mu m Flow Valve
- 3. Scrum Valves
- 4. CCN burQC l a r i l-: l}dsl( D - u p valve (2.0)
- b. EXPLAIN the cause a -t- t i-s e potential Hloh Radaation 1evels in the Off-Gas En.t i,1 d i n gf . (0.5) 1,
- ' h *l ,s' l .\ t o '., '
) $ #
- Es.,_f'l GHl;OES I ON_ LOC (d) DIN @_ SOEETY,_6HD_ EUE60EUGy_ SySIEMS. PAGE '11 OUESTION .2. O'5 (2.00)
Concern 2nci the RHR System and its various modes of operation: a.'MA1CH the following actions, events, and/or interlocks in Column A with its 2nitiation SEIPOINT in Column B. (1.5) Column A Column D 4'15 p ig
- 1. Input to ADS 4u0 ps)o
- 2. Shutdown cool 2ng i sol ates 350 phig
- 3. Allows manual operati on of the 135 psag LPCI inboard i ru ec t i on valve J25 pelo 50 proig
- b. Wher e does the Shutdown Cooling made of RHR take it s sucti on?
(ASSUME NORMAL DPERATION) DOESTIDN 2.04- (2.00) 1he Standby Liquid Control (SLC) System as designed to provide enough reactivity compensation to shut down the reactor frrm rated t o ::er o power 2 ncludi ng shutdown margin and to all ow cooli ng the nuclenr wystem t.o 70 degrees F with control rods remaining withdrawn in rated power pattern. LIST four (4) additional reactivity gains that rutst be compensated for by SLC upon actuation as specified by the system's design bases. DUESTION 2.05 (1.50) LIST the t hr ee G) methods employed by the Correbusti bl e Eau Control System that will reduce the percent of H2 2n the CONTAINi1EN1 in an accident situation AND HOW each accompl2shes its iunction. (***** CATEGORY O2 CONTINUED ON NEXT PAGE. 44**4)
~ '- ~ "
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1- !), .s 2; 06 , ' "t ((I00) I- .OUESTION l 1--
' ' ?,%
The plant.is operatinN t 9:)% ~ power when.thestiperator' notices an 1
. INCREASE in.the #2 sen1 @reasure on Recirc Pump * % " The. 411 seal l
pr' essure -f or. the pump' has - NOT i nereased. These gbservati onn are f allowed shortly by a PUMP "4". SEAL'GTAGING FLOW IOGH/ LOW alarm.
- ., )p. _ .
f ; I L IST' the TWO mal f unctions which - give this a] arm for the given conditions, q r s ii , QUFST10N 2.07 (1.00) -I , \. I \ l { LIST four (4) conditions th at automaticallhdhuse,p DIRECT trip c( the( L l' Reactor Water Cleanup (RipC"l *g Q unps. (Se%!u'i nty lq1, required ) ,
- p. 3
. s -{J A t, i./ ,
4- - 9 s. t' f . s QUESTION 2.00 'Q . 00 ) . , (4' "
?
kj\ t f t rag, f uncto/[ns which ARE bypassed when f.a . LIST four (4) the DIVISION i III' Emergency p Wgel Generator i s operating on a LOCA initiation signal. (1.O} pl a,? t
- b. LIST three (34 v.ignals whi ch wi ll ' AUTOMAT,1CALLY start the Di vi si on I']I EDG cind'b ng it up to speed and voltape.
1 ,
<y.
l ,'.+ ..
'U l g- QLJESTIO!O 2.09 .yt ,1. SO )
l:
.' N . .
l LIST two criteran which q<iine an ESF Bus undervcitage NGUV) condition l' and DIFFERENTIATE between this condition and an EGF Dus Loss of r , Power (LDP). 1
?
n.a
'1 1;
I l V i s ( (4**** CATEGORY O2 CONTINUED ON 11 EXT PAGE *****-) d
[ T., . _.. . . . I
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY _jYGTgtJS PAGE' ' 13)
{
. i l
4
-OUESTION 2.10 .(1.50) 1 With regard ito the' diese1 ' dri ven FIRE PUMPS:
- a. LIST the discharoe-pressure at'which the "A" and "B" diesel driven fire pumps will AUTOMATICALLY start, i
'b.-
L.IST..the number of AUTOMATIC start attempts which EACH-diesel will make. ' i l
- c. If the ~ di esel goes through'all of'its attempts.without {
startinci, LIST: the action (s) which must be taken in order to ! try to start the diesel. BE SPECIFIC'AS.'TO LOCATION (S) AND CONTROL (S)~ TO DE OPERATED! 4 OUESTION 2.11 4(1.00) The Upper Containment Pool' serves different functions _ depending on the p3 ant's operating mode or condi ti on at the time. STATE,four (4) ! purposes /f unc tions of the Upper Containment Pool, i J i l QUESTION 2.12. (2.50) i The reactcr.Is operating with all des 1gn parameters within normal i limits when a HPCS initiation signal is received. Indicate whether each of the following valves'will receive an OPEN signal, a CLOSE signal or NO signal upon automatic start of the system. ( Aw u. tv\ E. A O c> R.m A L UAvt%V uusue)
- a. Test return to Condensate Storage Tank (CST) (F010, F011)
- b. Pump suction from Suppression Pool (SP) (F015)
- c. Suction f rom CST -(FOO1)
- d. Test return to SP (F023)
- e. Injection valve (FOO4)
QUESTION 2.13 .(1.00) The Recirculation Pumps dischcrge/ suction valves require approximately two (2) minutes to cl ose. EXPLAIN WHY thi s time frame is necessary. (4**** CATEGORY O2 COhlTINUED ON NEXT PAGE **W44) 3
- . J 2 4. fb6bl DE@l@b_lyCLUDIN@,,,,@@ Eely,,,8NQ ,EdERGEUCy_SYSIEUS PAGE 14 PUESTION 2.14 (1.00)
During reactor start-up under' Cold Conditions, the operator adjusts t h t? Control ' Rod Dr'I ve pressure control valve to maintain a +260 psid between CRD and reactor pressure. EXPLAIN HOW this: pressure di f f orenti al is AUTOMATICALLY maintained as reactor pressure increases
'during the encuing'Gtart-up. (IGNORE OPERATOR REQUIRED.' MINOR ADJUSTMENTS)-
\ .15 DUESTION 2.15- ( z . ~ :)
The Standby Gas Treatment System (GBGTS) is in operation with aniauto initiation signal present. EXPLAIN the effect(s) on SDGTS Train'"A" for the followingi conditions.
- a. The SDGTS Mode Select Switch is.In AUTO. The operator places SBOTG Div I " MAN 'INIT RESET SW" to RESET and then returns'to NORMAL, then he places the " ENCL DLDG RECIRC FAN "A"" and "SBGTS-FILTER TRAIN *A'" switches to STOP,
'b. The SbGTS Mode Select Swi tch is in STANDBY and the operator performo.the same switch manipulations as in part "a" abava.
QUESTION 2.16 (1.75)
Answer the f oll owing with respect to the Liquid Radwaste s
1 1
I
3b_lO31WdENIS_G[@_ Cots,llBOLS PAGE 16l l' . .
1~
4 DUESTION 3.01 (1.00) i I l l
- a. When the handswitch on Remote Shutdown Panel P-1D1 is placed in the !
"OFEN" position, the (A, D, or DOTH) sol enoi d valve (s) open the SRV'?
- b. The power supply to the ADS solenoid valves is (RPS Bus A and B, 250V DC or 125V DC)? 1 I
OUESTION '.02
. (1 00) l l
LIST the four interlocks that inhibi t reci rc FCV motion. (setpoints j not required) 'j i DUESTION 3.03 (1.00) i TRUE'or FALSE? !
- a. The LPCS-injection valve (E21-FOOS) can be opened by the handswitch as soon as reactor steam dome pressur e is loss than
'575 poig.
4..st t' . If the*handswitch for the LPCS injection valve (E21-FOOS) is in I the normal position, overload protection for its motor operator J is removed. DUEST1ON 3.04 (2.00) For each of the 'ollowing, state whether a ROD PLOCK or NO ROD BLOCK is generated for that conditron.
- a. APRM D Downstale, Mode Switch in RUN.
.b. 13 L.PRM . inputs to APRM C., Mode Switch 2n STARTUP.
- c. APRM D indicating 13. p ower , Mode Switch an STrRTUP.
- d. F1ow Unit A and D (; 108% flaw). Mode Switch :n RLiN .
(4**** CATEGORY 03 CONTINUED ON NaT PAGE ***4*)
>: 3. > INSTRUMENTS AND CONTROLS PAGE 17l <
QUESTION 3.05. (3.00)
- a. EXf: LAIN what condition will generate EACH o-f the follawing indications on the Operator Control Module (P680):
- 1. Data Fault i
- 2. Scram Valves
- 3. Channel Disagree
- 4. Insert Required T c w e.v- i b.- 1. Acove the High E- cc Set Point (HPSP), con tinuet. s wi thdr awal of a control rod i s ,l i mi ted to ___,______ nnt ch (es ) .
(Fill in'the blank)
- 2. The HPSP is determined an a function of what plant parameter?
I OUESTION 5.06 (1.00) List FOUR (4) systems that have components that can be operated or controlled from BOTH Remote Shutdown Panel s P150 and P.51. DULSTION 3.07 (1.50) With regard to the Nuclear Instrumentation System:
- a. During a reactor s h ut d owri . With.the mode switch in STARTUP, the IRM's.are readang 13 on range 5. You downscal e all IRM's to range 4.
STATE the expected level reading which would occur and L IST any automatic action (s) - JUS 11FY your answer.
- b. While operating at 100% power, you bypass APRM Channel A. STATE the effect. if any, that this has on the reactor recirculation system?
QUESTION 3.03 (1.00) The Area Radiation Monitors have installed check cources which, when act2vated or d e a c t 2. v a t e d , will provide an indicatson of an ARM's operability. BRIEFLY descritsc: HOW operab211tv is demonstrated wl:h the check source in B O'l H the actavated and deactivated condition. (4 tat
- CATEGORY 05 CONilNUED ON NEXT PAGE ***4*)
Ib_ _ _Ill@TBLJMENI@_ AUD,_CONIROLS . PAGE '18 QUESTION 3.09 -(2.00)
~ Assume the TEEDWATER LEVEL CONTROL SYSTEM is being operated in 7-ELEMENT control using reactor LEVEL. DETECTOR CHANNEL 'A.' Reactor power is.at 85%, STEADY STATE.
For each of the instrument or control signal. failures list ed below, STATE HOW REACTOR LEVEL.WILL INITIALLY RESPOND (increase, decrease, or remain constant) and ORIEFLY EXPLAIN WHY i n terms of .WHAT i s happening in the F eud wcit er Control System IMMEDI ATEL.Y AFTER THE- F AILURE. fFOR EXAMPLE,.your answers should iriclude the f ollowing detail, "Causes r eactor level to decrease due to a steam flow / feed flow er ror signal, steam flow < feed flow, resulting in a signal to increase t-ho speed of the reactor f eedpump ( s ) ,, " IF APPLICABLE.)
- a. Channel A REACTOR LEVEL detector signal fails LOW
.b. LOSS OF CONTROL SIGNAL to B Reactor Feed Pump Speed Lontro'11er i
DUESTION. 3.10 (1.25)
- a. The RCIC system (CAN DE RESET, CANNOT BE RESET), when a RCIC system ILOLATION occurs while operating the RCIC system at Remote Shutdown-Panel P150?
- b. L I GT -etre RCIC controls on Fanel P150.
W e e-. L O QUESTION 3,11 (3.00) LIST all actions (if any) that occur in the RPS and ISOLATION , circuitry- when: ! 1
- a. Main Steam Line Radiation Monitor HIGH VOLTAGE INOP a] arm or Channel. L C and an UPSCALE HIGH HIGH alarm on Channel D are recieved at the same time. q
- b. BOTH Chanr,e1 C and D are UPSCALE HIGH HIGH alarms. ,
- c. Assuming either (a) or (b) re,ulted in DOTH channels of the ISOL ATION LOGIC bei ng satisfied, LIST the components that would SHUT / TRIP.
1 (tt444 CATEGORY 03 CONTINUED ON NEXT PAGE ****4)
i
'3 nj UgIBU!gNIS.gSD_1OSIBOLS 'PAGE 19)
QUEST 10N- 3.12 (2.50)-
.m
. Answer ~ the f ol'1 owing questi ans concerning the L.ocal Power Range Monitors. !
l
'a. State what is used in LPRM Detectors to extend the, detect.ar
]
life. (0.5) 1 1
- b. LIST three'(3) indications that.the LPFWI Mode Selector Switch is in the LYPASS posi ti on. '(1.0) .j i
- c. ' LIST iour'(4) places where the' FLUX. AMPLIFIER sianal'is fed when the 1 Mode Selector Switch is.in OPERATE. (1.O) 'l 1
i OUESTION -3.13 ( 1. 50 ) . ! 1 Concerning'the Condensate. Filter Demineralized Hol ding Pumps. l i
- a. STATE under which conditions the pump will start and scop in.the 4 automatic mode of operation. (setpoints not requ2 red) l b ~. STATE what may occur if the holding pump failed to start when required by the automatic mode'of operation. l QUESTION 3.14 (1.50) 1 i
When. operating at 100% power. the wide range level instruments read lower i
.than the narrow range indication.
- a. STATE if thi s mi smatch i s a concern to the operatcr / ' EXPLAIN the i answer.
- b. STATE if the wi de rar>ge level instruments can be used to accurately indicate l evel at the TOP of active f uel' EXPLAlid the answer.
OUESTION 3.IU t2.00) When paral l el i ng the diesel generator with the 4KV bus: l a. STATE what the GOVERNOR CONTROL is used for; I
.1 . - BEFORE the output breaker is closed.
- 2. AFTER the output breaker is cl osed .
l b. STATE what the VOLTAGE'REOULATOR is used for: I
- 1. BEFORE the breaker ie closed.
- 2. AFTER the' breaker 19 closed.
l l l (****4 CATEGORY 03 CONTINUED ON NEXT PAGE $44f*) 1:
- 3. ~ -1NSTNUM(NIj3j,,Alj D; CONI!30lJ - PAGE ,20
- s. -. ' .
t
. - . \ .7 5 QUESTION- 3.16- . ( T . Z '3 )
Concerning the Electrohydraul2c Conteel System (EHC):
- a. LIET the.THREE (3) parameters sensed and evaluated.by the control system, .0.75)
( t.. STATE what contr o11 i n ce circuit is actually positioning the control valves. ASSUME 100% power and normal system operation. ( 0. 5) '
- c. STATE what contral circuit is out'of the- coistrol acheme, and WHY, When..the Main Generator is synchronized to the grid. (0.5) i 1
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l L {. Ef i i i 4 (4**** END OF CATEGORY 03 *****) L . _ _ ._ . _ _ _ _ _ _ _ _ . . _ . . . . . _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _
l' - <
- 4. PROCEDURES - NORMAL u ABNORiAli j Et1ERGENCy..AND' PAGE 21
+1 .- .- RADIOLOGICAL COfCROL l
l: DUELTION 4.01 (1.00) GTATE TWO (2) conditions which must exist to all ow c, NDN-LICENSED individual to manipulate the controls at the GONS
\ ,5 0 OUES TION 4.02 (W)
According to ONEP-05-1-02-I-2, "Turbi ne and Generator. . Tri ps, " ' 'Immediate Actions; I
- a. . STATE how to determine if the EHC speed demand circui t .is contolling turbine'speen. !SETPOINT. REQUIRED for. full. credit) b ,. '5 TATE the INIMEDIATE ACTIONS if the Generator has tripped and Turbine has not tripped.
/
OUESTION 4.03 (2.00) Centerring EP-3, Containment Control;
- a. STATE the MINIMUM Suppresni on Pool Level that will be maintained when operating RCIC. LFCS, HPCS and RHR in the LPCI mode,
- b. STATE the MAXIMUM Suppress 2cn Pool Temperature while operating RCIC.
- c. STATE the MAXIMUM Suppression Pool Temperature while operating HPCS, RHR in L.PC I mode,.and LPCS. ,
- d. STATE the BASIS for the temperature and level limitations, (i n a., b.. and c. above), for the Suppressi on Pool.
i (***** CATEGORY 04 CONTINUED ON NEXT PAGE dt***)
-3L.JROgF;DURES :+ NORMALLAEjgRMGL,,_EMERGElj gk_AtjD Pf.GE -22 6 ~.. . 60D10LggICe6_c9bilB969 4
L \ 1 :- 1 R 2.00 l- QUES 1 ION 4.04 (SE) 1.
- - Cor3cerning Containtant Pressure Control' per EP-3; I 1
- a. ' STATE what must'be accomplished whun' Containment pressure. reaches 17.26 psig.
- b. . STATE the BASIS for part a. above.
- c. STALE if Containment; Spr ay CAN DE INITIATED or CANNO1 M ' INITI ATED .
When containment temperature is 150 degrees F and Coni.ainment-Pressure i s 9 psig. (REFER to attached figure CN-T-1.) i
~d. EXPLAIN the DASIS for precluding contai nment spray actuation 1 (sh,aded area) per CN-T-3.
2.c o QUEST I Oid 4.05 FW) EP-14., '" Level /Fower Control," permits bvpassing two low level asolations;
- a. STATE the TWO (2) low 1. ev el isolations that may be bypassed,
- b. STATE the BASIS for allowing the bypassing of these 'i sol ati ons.
QUESTION 4.06 (2.00) LIST the entry conditions for EP-5. Secondary Containment C o n t r o l,. (INCLUDE SETPOINTS) (TA hLn N o T M Am EG D - 105t Th EW- TohC S ) QUESTION ,4.07 '( .50) FILL IN THE DLANKS l The transition from startup to run must be made with all APRMs beween' (1) _ ..,_.____._ _ , % and (2) _________._ _ % power. ' 2.Oo DUESTION 4.00. (M..>> Concerning EP-2. "RPV Control": LIST the entry c on d i t i on t., for EP-2. (INCLUDE SETPOINTL ; (**%tt CATEGORY 04 CONTINUED ON NEXT PAGE $$***) i _ = _ - - _ - _ _ - _ - _ _ _ - ,
A s _._ f B O C E D U B E S ,_ _ N Q O d h _ O D U g B U L . J U E B G E U G L O N D PAGE 23, 66DIOL901 COL _CQUILQL i
- 2. S O !
QUESTION 4.09 (;.1J) In Step L P-- 13 of EP-14, " Level / Power Con +.r ol , " lowering of rear: tor vessel level is directed.
- a. LIST the THREE (3) condition (s)/ level (s) at which the'lowerings of 1 vel is r.er mi n a t ed ,
- b. LIST AL.L. i rd ec t i on systems which are not terminated by Step LP"13.
OUEST10N 4.10 ( .50) P'rocedure SOI-04-1-01-C11-1," Control Rod Hydraulic Sytem Operation," CAUTION states to i sol _at e seal purge using valves EF026A(B)] prior to . shutting the reactor recirculation suction and discharge valveu and not.to startup the CRD Hydraulic system with the reactor ~ recircul ati on syst ero iso!ated. STATE the BASIS for thi s caution. DUESTION 4.11 (1.00) IMMEDIATE OPERA TOR #? TION step 4.5 of DNEP-05-1-02-I-4,-'" Loss of Offsite' Power," states to check that three specific DC oil pumps auto start or to manuallv start them. STATE whv the DC Generator Seal 011 Pump 1s needed and WHAT dangerous condi La on is prevented by operating the pump. I l l l l 1 I
\
(*4444 CATEGORY 04 COiJ .I NUE D ON tiEXT PAGE *****) i
$ _fBOGEDUBE Er NOBMG63._GQUgBU % _EMEBGENCLOND PAGE 24 e . .- - BGD196091CGL__gDNIBOL 1
OUESTION 4,12 (2.00) i i MATCH the following emergency cl assi ficati ons to thei r appropriate rief i ni t i on s. i
- a. Unusual Event 1. Tt'e occurrence of an event or events which involve actual .or likely major j
- b. Alert failures of the plant' functions needed i for'the protection of the oublic. !
- c. .Si tie' Area Emergency
.2. The occurrence of an event or even+s
'd.. General Emergency which . indicate a POTENTI AL degradatt ori f of the l evel of safety of the plant. 1
- 3. The occurrence of an event or events 6
-which involve actual or imminent..substan-tial core degradation or melting with the potential for loss of containment integrity and substantial. releases of large amounts of radioactive material off-site.
- 4. The occurrence of an event or events' which involve an actual or ootential SUBSTANTI AL degradati on of the level of safety of the plant. .
QUESTION 4.13 (2.00)
'a. ' STATE how Jong an operator can stay in a 25 mrem /hr radiation I field without exceeding: (."45 5 U M E y o u DO NOT have a NRC. FORM 4 ,
on' tile anr VOJR exposure this quar t er is 975 mrem) l'
- 1. A Grend Gulf administrative exposure limit.
I J
- 2. The NRC quarter 1y exposure l i mi t ' 'J l
b, YOU a-e working with L groep of four (4) operators recharaing CRC accumul ator s when you dr op your pocket dosimeter. The dosimeter 2s now indicating FULL SCALE. The dosimeters of the other three member s crf the group are indicating L mrem. According to Procedure , 01-8-08-2. "Ercosure and Contamination Control," STATE the MOST l; CORRECT act1on(n) for this si tuati on. j l l 1 1 1 I l (4**** M1 EGDfW 04 CONTINUED ON NEXI FAGE *****) l l l
&__ESQGEDUBES_;_[4QQM6Lg_80[JOBdh;EdEBGEdgLOND - PAGE_ 25.
i w 00DIOLOGICaL,cgNIeOL
> QUESTION ~ 4.'14 (1.50)
- a. STATE'why the DFFDAS system CHARCOAL ADSORDER beds are bypassed during. initial system start:up.
- b. STATE why a dry' air purge must be established for the RECOMBINER TRAIN:
- 1. PRIOR to STARTUP?
- 2. IMPEDI ATELY after SHUTDOWN 7 QUESTION 4.15 (2.00)
- a. GGNS Procedure 01-S-02-1, " Description and Use of the.5GNG Operations Manual," lists four requirements, one of which must be met prior to deviati_ng from approved procedures. LIST THREE (3) of these requirements.
t?. WHO'and WHEN must plant personnel notify after deviating from approved procedures. QUESTION 4.16 (1.00) Procedure ONEP 05--1-02-I-02, " Turbine and Gener ator Trips," provides the operator.with a CAUTION not to allow the 500 F:N Breaker to be OPEN with VOLTAGE applied to it for greater than THIRTY (30) minutes.
- a. STATE the BASIS for this precaution (li mi ti ng component)
- b. L.IST the action. required if the 500 KV Dreaker w)11 be open for more than THIRTY minutes. :(
I i i } { I (***** END OF CATEGORY 04 ***44) f (****At******* END_OF EXAMINATION ***************) { I t I _-_______-__-_--__2_____-_
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P = (seV)/(3 x 1010) 11d 1 2*1d22
=W R/hr = (0.5 CE)/8(meters)
R/hr = 6 CE/8 (feet) Miscellaneous Conversions Water Parameters 1 curie = 3.7 x 1010dps 1 gol . = 8.M5 lbs. 1 kg = 2.21 lhe 1 pl. 3.78 liters 1 hp = 2.54 x 1 St.u/hr 1 ft3 = 7.48 galt - 1 m = 3.41 x 1 Stu/hr i Bensity = 62.4 lbs/ft Density = 1 gn/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1ba *F = 9/5*C + 32
*C = 5/9 (*F-32)
Heat of fusion = 144 Btu /lbe 18TU = 778 ft-1bf
~
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l'-. LERINCIF' LEG OE_tMLgAFLEQWER_ PLANI _QPERATIQN 2 PAGE 26 J J.> . 'IHERt10DYNAtjICScHEAT TRANSFER AND_ELUID_E69kJ l l 1
. . ANSWERS' ~ GRAND GULF l' -07/05/10-CLARK, F. l Att9hER 1. I '1 N
- a. Dec ',se
- b. Decrease
-c. decrease j d.- Increase Q*< '
- e. Decrease g.'
l(5 & O.25 each) R5FERENCE C G N S '.O P - D 3 3 0 3 , L.O.-# 3.g.6 m ... < ,v-x< - e. ~ n e. , , e n mm .,<,:,,, w ,, 94 l ANSWER. 1.02' (1.50) I
- a. (1)- Increases (O.25)- 1 (2) Increases (0.25)
'b. (1)- Gross enthalpy increases dtm to the reheat portion _of the MSR.
(0.51 (2) Specific enthalpy increases due to the moisture removal in the MS Geccion EO.25] and heat added in the reheat secti on E O . 2 5 .1. REFERENCE. GGNS HT&F.F, Chapter 1,.L.O. #'s 7.1 and 7.2 OGNS - OP--N11-501/Rev. 1, L.O. #3 245000K408 ...(KA"S) 'l l l 2.00 1 ANSWER 1 . O '_'. (-1. 5 0) ;
- a. NPSH, the di f f er ence betweer the total. pressure at the eye of the q pump and the saturation pressure of the liqu2d, (1.0) j
- OR- )
NPSH = P actual - P sat. l b. Decrease (0.5) -co- \ n e" o-W O ' 6 ) 1
- c. Increases (0.5) l 1 lL REFERENCE j GGNS HT&FF, Chapter 6. L.O. #10 j 29';.OO6klO3- 293OO6K110 ... (KA'S) j i
l e .
I 1 , CBIURIELEB_QE_UUCLEGB_EOWEB,_ELBUI_DEEB8IlgNs
._ PAGE ~ 27l
.. - ISEOOQDXNOMICSt _bEOLIB6USEEB._QUQ_ELUID_ELQW
. ANSWERS -- PRAND' GULF 1 -07/05/18-CLARK..F.
ANSWER 1.04 ( .50)
' Water hammer- Hluid hammer) E0.25] tould cause severe system damage EO 25]) - o tt i n ec e.a se %_ g njec.bn b m e_ clu.v h>
t h =- b eI' A =- - b._q hh L (.o. ~2.5) REFERENCE 1 GGNS HT?aFF. Chapter 6 L. D. 4f 8 GONS OP-LO-bYS-LP-E21-03, L.D. #3 GGNS OP-E22-1-501/Rev. 2. L.O. ft3 :
.209001K603 209002K103 209002K401 209002K402 293OO6K105
...(KA"S)
ANSWER 1.05 (1.00) 350 psig + 14.7 psia = 364.7 psia (0.5) i. Daturation_ temperature for 364.7 psla: (444.6 degrees F - 431.73 degrees F) (14.7/50) + 431.73 degrees F = 435.5 degrees F (0.5) i EAnswer should be within + 2 degreees FJ l REFERENCE Steam Tables and GGNS HT&FF, Chapter 3 3 L.D. tt ' s 1 and 3 293OO3K123 ...(KA"S) l ANSWER '1.06 (1.00) Critical power operations with reactor pressure < 785 psig EO.5J .! or core flows ( 10% EO.5]. i i i REFERENCE I a GGNS HTbFF. Chapter 9 L.O. ft 5 i 293OOSK109 293OO9K105 ...(KA's) i l 1 l i
t
<- I l 1.;_EfLINCIFM S OEdyUCLEAR POWER PLANT OPERATION; 3 PAGE. 2G
-. IHE8MODydGtlICS1 ._,HEGLIBGtJSEEB;GND_EllMLELOW L . ANSWERS GRAND. GULF-1 -87/05/18-CLARK, F.
l l l l' l
' ANSWER 1. O'7 (1.00) l-
- a. Conduction: (0.25)
' b ,. Conduction (0.25) c .- Conduction (0.1) and Convection (0.15;
- d. ' Radiation. (0.25)
REFERENCE l GGNS.OP-HF-507/Rev. 1, L.O. t!1 j GGNS OP-HF-508/Rev. 1, L.O. #7 1293OO7K101 293OOOK112 ...(KA'S) ANSWER 1.08 ( .50) MAPRAT = APLHGR (i n a node)/MAPLHGR Limit H or that node.) (0.5) . REFERENCE GGN3 OP-HF-509/Rev. 1, L.D. #4 .i 293OO9K110 293OO9K113 ...(KA'S)
. ANSWER 1.09 (2.50)
- a. Closure of all MSIV's (0.5).
- b. (1) Pressure increase due to closure of all MSIV"s (0.5)
(2) Neutron flux drop due to reactor scram 00.253 caused by MSIV closure E O. 251 Iscram when three MSIV's less i than 94 % open) 1 (3) Vessel Steam fl ow decrease'due closure of the MSIV's (0. 5) (4) . Vessel Steam flow. increase due to 5RV's opening (0.5) i REFERENCE GGNS OP-DT-507/Rev. 1, L.D. fi s 1 and 2 l 202OO1K102 202OO1K105 202OO1K122 ...(KA"S) i 1
- ,; qn ;
l 11- .1 I'BJ NC J ELEg _OE_UUCLESB_COWEB,_ELGUI,_OCEB@T1001 PAGE
'~ * . . ItNBOODyy@OlCS._1]ESI;IBOUSEER AND FLUID,ELOW. 29]
3 1
. ANSWERS;'- OR ND GULF '1 -07/05/10-CLARK, F.
I I l l i q '; , . ANSWER 1.~ .10 . (1.00)-
.a REFERENCE i
' GE DWR. Academic Seri es, Reactor' Theory,. Chap. 6, LDti 1.1, 2.1 ]
292OO6K101- ...(KA"S) l 1 ANSWER 1.11 (1.00)
'l d-REFERENCE GE DWR Academic Ser'.i e s , Reactor Theory, Chap. 3, pp. S-12, LO# 1 292OO3K101 . .(KA"S)
ANSWER 1.12 (1.00)
'g .-'
L REFERENCE DPC,'FLindamentale of Nuclear Reactor Engineering, p. 96 GE BWR Academic Series, Reactor Theory, Chap. 1, p. 3D, L.0 4 6.1. 292OO2K111 ...(KA'S) ANSWER 1.13 (1.00) C REFERENCE GE BWR Academic Series,. Reactor ~I h eor y , ' Ch ap . 3, pg. 9, LD# 1.6 292OO2K107 292OO2K108. 292OO3K101' 292OO8K103 ...(KA'S) ANSWER 1.14 (1.00)
'a REFERENCE' GE DWR Academic Series. Reactor Theorv, Chap. 6, LON 2.5 2920' WK 107 ..,(KA'G)
L II.__EBJNQlELES OF NLICLEAR POWER PLANT OPERATION 3 PAGE 30-
'1
- D. .r Iu gBL1Q D yN 6L11CQ 3.,,,U E 6 LIBGL@E E 6_ G tJ D_ EL y lD _ E L Obl
. ANSWERS'-m GRAND ~ GULF ~~1 -87 / O'5/18-CL ARK . F.
ANSWER- l'.15 ' . (1.00) l 1 d REFERENCE'
- GE. - DWR Academi c Seri es, Reactor Theory, Chap.'6, L.O tt 2 . 2 ,, 2.3 292OO6K103 292OO6K104 ...(KA'S)
, ANSWER 1.16 (1.00)
- a. keff (0.5)
- b. increase,. decreasing (0.25 each)
REFERENCE GE DWR Academic Geri es, . Reactor Theory, Chap. 3, pp. 5-12, LO-# 1.5 292OO3K101 292OO3K102 292OO8K105 ...(KA'5)
-ANSWER 1.17 (2.00)
- a. False
. ti . True c . . Fal se
- d. Fal se (0.5 each)
REFERENCE GE BWR Academic Series, Reactor Theor y, Chap. 4. LO# 7.4 292OO4K101 292OO4K103 292OO4K1'13 ...(KA'5) ANSWER 1.18 (2.50)
- a. True (0.5)
-b. False (0.5)
- c. False (0.5)
/ d. True t0.5).
- e. False (0.5) 4 REFERENCE GE BWR Academ2c Series. Reactor Theory. Chap. 6, LOtr 2, 3 292OO6K110 292OO6K114 292OO6k121 ...(KA'S)
- i.
- v v'mp.% p , ;u ,
,g;
- 1. .C,. .EBINg1ELgS; OEiljugL E ARiPQWE6;EL Alq_0PgROT_lC}Ng ,
PAPE 31
> l.' . IUER!.10DXN9!1ICSi- HEAT -TRANSEEB_OUD_ELgIQ_ELOW
.: ANSWERS'- GRAND: GULF--1 -07/05 /10-CLARK,..- F.
A;
% , 4.6- -
r , e g t-
'\ .
C+ ,
.\
ANSWER 1.19 (1.50)
~
- a. 5-l b. b' 'a
- c. 4 '
,_ 'd.'2
- e. 3 1, f..-1 *-
-(0.25 each) l' "/ REFERENCE . .
I GE.DWR Acad>mic Series, Rtaat t or Theory, Chap. 1., pp. 33,35 LO#'4.1, 5.' 1 l and Chap. 3 ,.. p p . [5.71 L O # ".:. . ,1, 4 . 1 l 292OO1K107 292bOh107- 292002K109 292002Ki10' 292OO3K105
. . . ( K A ' 9)., ' '
- l. 1 s-i
^
- l. .. ;f
~
y.
' "E O '
' ANSWER- 1.20 ( 1., 50 )
- a. Power l evel .
.b.. Time at power
- l. c; Time since shutdown REFERENC$
l' GE DW9 Academi c. Seri en, Reactor Theory; Chap. 7, pp. 23-24, LO# D.3 292OOBK129 292OO8K130 ...(KA'S) l l- a 1: l1 l t . 1 1 it' I-
.4 s
. -f-f. . , .
'$3 1r _EBLUg]C[ES_OE_UUCLfiBBLE0158_ELGULDEEBOI.Lgth PAGE 32 7 -. .. IUEBt]Q DyU @[j [Q@ h.UE G I _IBeN SEEB_8SD_.EL UI D_ ELO W 1 ,
. ' ANSWERS -- GRAND GULF 1-
-87/OS/18-CLARK. f.
l 1 ANSWER 1.21 (2.00) Drawing-a stra2ght line between the last two **s predicts 34-36 l: control rods snust be withdrawn, l (0. 25 f or each point plotted, 0.50 for line and prediction) S 10 15 20 25 30 35 40 45 50 55 1 , o g _ .- _ _ ; ._ _ __ ._ _ .. _ ; ._ _ _. ; _ _. _ ._ ; _ _ . _ ._ ... _. t _ _ ; - _. _ ._ _ _ - g _ _ _ . ; . . - 1.0 O.9- -0.9 O . B -'- * -0.8-O.7- -0.7 1/M O.6- -0.6 O.S- -0.5 r- g -- 0;4- -- o . 4 l O.3- * -0.3 0 2- 4 -0.2 O.1- 4 -0.1 4 . O . O l - - - l --- - - - I - - - - l - - - - - - l - - - l -- - - ! - -- - X - -- - -- 1 -- - -- - l - --- -* i - -- - - ! - ---- ! O 5 10 15 20 25 30 35 40 43 50 55 Control Rods Withdrawn l REFERENCE 7
, , l. GE DWR Academic Series. Reactor. Theory, Chap. 3, pp. 33-15, l . Dit 2.3, 2.4
.H 292OO3K104 ...(KA'S) i 1
I
)
,' N. ._.PL8UI,_DEg1GU_INCLUDIUG _g6EEILOUD, EtlE3GGUCLSXSIEtjS PAGE- 33 ANSWERS -a GRAND. GULF 1 -87/05/18-CLARK, F.
h ANSWER 2.01- (1.00) a REFERENCE GGNS: OP-LO-SYS-LP-E51, LO tt4a 21~7000A212 ...(KA'S) ANGWER 2.02 ( 2. GO ) '
- a. 1. AI
- 2. F.O l (0. 6 e a.c.h )
- 4. I. v ,
i b , Valve stem air ta the Off-Gan System is 1ast providinct n potential ~l flowLpath for airborne contamination. REFERENCE GGNS: DNEP 05-1-02-V-9, OP-L.0-SY SJ LP-P53 -03, DP-N21--501, LO#6 OP-LO-SYS-LP-P42-02, LrA) Sd , OP-C71-501, L.OH 7b OP-LO-SYS-LP-C11-1A, LO flBa-27501'? AK2 O 25900]K106 223001K110 201001K109 ...(KA*S) 1 ANSWER 2.03 (2.00) l l a.1. 125 psig 1
- 2. 135 psig 1 J. 5. , . M'15 t>m 3 (3 ? O.5 each) I
- b. Takes suction from Recirculation Loop B. (0.5) j I
REFERENCE 1 CGNS: O P- L D- B Y E+- L P-E 12 - 02., LO #3d 205000A410 205000K100 205000K402 205000K603 ..(KA'G) 1 l l l' 4 4 l l
2.t. _f1691 DESIOLL.1NCLUDlyG.,58EEily_GUD,fdEfiOEUCy GySIEdG PAGE 34
- dNSWERS - GRAND GULF 1 -87 / 05/10 -C L ARK , F.
4 l ANSWER 2.04 (2.00) L 1 '. Decay af rated power Xenon Inventory l
- 2. Poni t.19e reac ti vi ty effects from eliminating'uteaia voids (or reduced neutron leakage iram boili ng to col d) 3.~ Reduced doppler effect
- 4. Drycreasing rod war th as rnoderator cools
- 5. Imper f ec, t mixing ( An y 4 .i) 0.50 each) fiEFERENCE
'GGNS: CP-LO- GYS ~L.P-C41, L.O #2 211000K301 ..(KA'S) l ANBWER 2.05 (1.50) a.. Containment purge system - by dilution (0.25 each) '
- b. Hydrogen riscombiner. -by the reaction 2H2 + O 2 ~ ~-- > 2H2O (0.25 each)'
- c. Hydrogen' igniters - by burning H2 at l ow concentrate ons (0.25 each)
REFERENCE GGNS: OP -Lor S YS--E61, LO43a 223001K404 223001K505 223001K506 223001 ...(KA'S) ANSWER 2.06 (1.00)
- a. Failure of Recirc Fump "A" til seal (O.S) ;
- b. Pl uggi ng of Retirc Pump " A 42 seal internal res tri c ti on /br eak down ;
or2 face (O. 5) i REFERENCE CONS: OP-LO-SYS~LP-D53-1-03, LO #3c 202001A1OG "'02001Al10 202001A210 202001A411 202001K605
. . . -( K A ' G ) 1 i
i i l i
. i 2E,_,_ PL6blL.yESIGO._ I_UCL Up,I U@,__gGEE Ty,._6Bp_EME RpENC Y_jYSTEMS PAGE - 35l ANSWERS - GRAIJD GULF 1- -87/05/3G-CLARK, F.
1 ANGWER 2.07 (1.OA)
- 1. CCW (pump) cooling-water temerature h2gh (195cF)
- 2. Pump suction 1ow f1oW (70 gpm)
- 3. Motor protection device !
- 4. Suction -flowpath valves not fully'open (<90% open) (4 & O.25 each) l 1
REFERENCE GONS: OP-LO-SYD-LP-G33/36, LO tt5a2 20400A204 '...(kA"5) ~{ ANSWER 2.08 (2.00) ; i
'a. 1. Low lobe oil pressure i
- 2. High j acket water temperature
- 3. L.os s of excitation (generator lockout).
- 4. Reverse power (generator lockout)
- 3. Generator overcurrent with voltage restraint
- 6. High trankcasu pressure (any 4 9 0.25 each)
- b. 1. Reactor vessel level low -41.6"
- 2. High drywell pressure -F 1. 39 prig
- 3. Manual HPCG System init2ation pushbutton
- 4. L. ass of normal power to ESF Bus 17AC f or 2.3 seconds (farst correct responso 0.04. others 0.33 each)
REFERENCE GGNS: OP-LD-SYS-LP-PDI, LO #4a 264OOOK402 264000K400 . . . (IT A ' S ) ANSWEfi 2.09 (1.50)
- 1. DUV 90% Bus UV for 9.O seconds 80% Dus UV for 0.5 seconds wi th LOCA u c)n.nl 70% Bus UV for 0,.5 seconds (any 2 Q O.5 each)
- 2. LOP - All ? 1:ines supplying the ESF Dus are deenergized. (0.5)
REFERENCE GGNS: OP-LD- SYS LP-R21 -03, LO #5a,b,c.e 264000K600 295003AA10 295003AA20 . . . O' A ' S )
2.. 2 11091, DEGirN a lGGLL.jDJNG ,G6EEIL,,6MD_E ttE13GENQY_S'(SIEtjS PAGE 36 ANSWERS - GRAND GULF 1 ~B7/05/18-CLARK, F. ANSWER' 2,10 (1.30)
\LS a, ti+ p s i g ' ( af t er preset ti me rjel ay) (0.5)
- b. Sin (6) . ( 0, 5 ) .
- c. Recet the starting circuit on the l ocal panel- (P154) (0.25) by placing the Dei sel Driven F 2 re Pump A/D Selec tor Switch to the DFF posi ti on (0.25).
REFERENCE 1 GGNS: OP-P64-D01, LO tt3b , c 286000A301 206000A304 286000A406 2G6000K402 ...(KA'S) ANSWER 2.11 (1.00)
- a. Prov2 des storage space for steam. separator, steam dryer assemblies and For some. fuel.
- b. Prov2 des shi el d i n() for fuel transfer.
- c. Providos shi elding when the reactor is in operation,
- d. Provides a post-LOCA source of Suppres si on Pool Make-up water.
(other purposes, as appropriate) (0.25 each) REFERENCE GGNS: OP--E 30- 501 LOn3a 223OO1K105 ...(KA'S) ANSWER 2.12 (2.GO)
- a. Clases i
- b. No
- c. Opens
- d. Closes !
- e. Opens REFERENCE GGNS: OP-E22-1-SO1, LO H3d 209002A201 209002A203 209002A210 209002A301 209002K405
...(KA'S)
j q
= ?t ._ E'l:6dld.DESJ GNa].NCld!D11{G_S6FETf_6ND .EMEFjGEUCY SySTEjdS PAGE_ 37q C ,e' .' i ANSWERS'"--GRAND GULF l' - 07/05/18- CLARK. F; I l
I ANSWER 2.13 (1 00) l To ensure that valve operation does not' interfere with pont-LOCA (0. 3 ) - Recirc Furro coast.down (O.7). REFERENCE
'GGNS: OP-B33-1-501, L.O 4t3e j 202OO1K101 . . . ( N (V S ) !
ANSWER 2.14 (1.00) The FCV opens up as reactor pressure iricreases rnaintaa ning a constant
.f l ow and therefore, a constant pressure to the PCV.
REFERENCE GONS: OP-LO-GYS-LP-C11-1A-03, LO #3e 201001A101 201001K400 ...(KA's) s .2 F ANSWER 2.15 (W)
- a. SDGTS Train "A" will continue to run (1.0)
' ' :1C T ;.,
-b. GDGTG Train "A" w.tl1 stop (0.25),
Tr+ ++ , """ ' 1, . f l a., .2?) c ;- T u-< i , ., ; . _u ; :. i . ., n
. a u c. , v . .,_ L .- u. w ,u;_
LLL g, a u. u ... s. _Li. REFERENCE GGhlG: DP-LO-S YS-L.P- T4 0. LO #3 261000A301 261000A302 261000K401 ...(KA'5)
- - - ___ ~ _ - _ - _ _ - _ - - _ - . _ , - _ _ - - . _ - - _ _ . ___ -___-____ _ _ - - _--- - _ - _ - _- h J ,f L O N I _.E E @ l @ d _ l y C L U D i U G _ S @ E E ] L 8 U Q , E b E 6 5 E U G L @ Y $ 1 E M S Pf6E 3D ANSWERSL -- GRAND ' GULF 1 -G7/05/18-CLARK, F. s ANSWER 52$16 (1.75)
- a. ;The three pra mar'y di vi si ans of 'the 'Li quio Radwacte .Gyntam a e:
- 1. Equipment Drains-
;? . Floor ~ Drains
- 3. Chemical Wae.tes (0.25'eath)
- b. 1. Inoperative
- 2. High, high radiation-
- 3. Downscal e
- 4. HighdischargeflowthroughF0355 valve (>100gpm) ~
- 5. Low PGW or circ. water flow-( <2500 gpm)
- (ot- " low ddwbon V W") (any : 4 - G) 0.25 each)
REFERENCE GGNS: OP-LO--S YS-LP-G17, LO; # 1,5 268000K106 268000K105 268000 GOO 4 . . .' OC A ' S ) 1 ANSWER '2.17 (1.00) l (When the turbine trips, pressure in the turbine decreases to condenser ' pressure.) The cc,ndensed. extraction s> team flashes back to stesm (0,5), and.backflows up the extraction. steam line to the Cross'Around f Steam Pipe and through the LP turbine (0,5), which can cause it to ~ overspeed. REFERENCE GONG: OP-N36-501, LO H3a,4,5 295005AK30 295005AK2O 239001K110 239001A303 ...(KA"S)- ANGWER 2.18 (1.50)
- a. The ECCS pump discharge prescure interlocks (at least one-of the respective L.P. pumps must be running). (0.5)
-b. F601 - the tall pipe prennure switch has ener gi z ed P631 -- (the "D") solenoid is enerq2 z ed (0.5 each)
REFERENCE l GGNG: OP-LO-S YS- LP-E22-2, LO #3d 8e c,a l- 21BOOOA301 218000K402 ...(KA"S) l l: I
3 .-._. t it EjTRl !LELJT S_AljD .LO_N._T R._DLS. .p ggg. ,9 ANGWERS - GRAND GULF 1 _ g 7 f g 5 f 3 g ...C L fir K . F. ANSWER 3.OJ s1.00)
- a. D (0.5)
- b. 12SV DC (0,5)
REFERENCE EG1JS OP E22- -02, L.D. 41 ' s ~.c and 6.b 210000F606 21aOOOKi<:5 , ,, , c g y g ) ANSWER 3.02 <1.00)
- 1. High Drywell Prensure (1.23 pslo)
~ oc. .) Aedeg crt %A High Anislog Contral Ca rt.ul i f ai' ur e d Y d * ' d y F=m Akvak. N ic3h 2.
d Tibbo, hb._ d C.k G" ;
*#" 1 3- Loss of control power
- 4. loydr aul i c Power Unit failure fe5L d Nj dtoAbc b d TC L""J k \
b) My d s M c b"d "t N' / vi
, , ,., v. 1.,
o .. pv_, eacn> ( REFERENCE GGNS OP D35 2--301, L.O. #-
'.c and 5.a.j 202OO2F. i 12 . . . I Km 9 ;
ANSWER 3.03 (1,ogy
- a. False (0,5) , _,3 i
- b. True (O.S)
REFERENi E GGND Of '- tT21 - U3., L O. U5.a 209001R601 20900JK40t .,,mg-5, ANUWER .;.. 0 4 (7,c;)
- a. ROD M OCK ( 0, 5 )
b ROD OtOCf. (<,.5)
- c. F'GD DLork (O 5;
- d. RO D BL'3L r'. (0. 5;
'Et _1.NGIBUL1EtJT,gJt4D'_QONTROt,S- a PAGE ,40 ANSWERS'--~ GRAND. GULF.1 ~O7/05/19-CLARK, F.
REFERENCE GGNS-. OP-CD1-4,1 L.D. #*s 3 and 4.b.
'GONS' Of-C11-2203, L.O. #7
~201005K403 ...(KA'5)
ANSWER 3.05 (3.00)
- a. 1. Indicates that at least'one rod has a defective position probe EO.5].
- 2. Indicates that all scram' val vee are not in the same position EO.5].
- 3. Indicater, that-the RGDS finds disagreement between ttie signals received from the RACS [0.53,
- 4. Indicates that tt.e sel ected rod must be fully inserted before ar.y other control rod can be moved CO.53.
- b. -1. 2 notches. ' ( 0, 5 ) 2 O TL ' \^*4C-h e b EM - ** "' -
- 2. Main turbine first stage pressure. (0,5)
REFEREldCE GGNS'OP-C11-2-03, L. O. ti~ s 6.a and 6.b 201005K601 20100SK103 20100SK102 ...(KA'S) ANSWER J.06 (1.00)
- 1. .GRVs or F,el i e f Valves
- 2. RHrs
- 3. GSW
- 4. CRD Pump (4 & O.25 each)
REFERENCE. GCNS OP-C61-OO, L.O. #3 295015K202 ...(KA'S) l l l l l l-p 1 N_ __ __ ._ .___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _
3 G lygI6LJtlEUIS OND;,COUIGOLS .PAGE '41 L. .... ;
'[ ANSWERS. -- GRAND GUL F 1
-07/05/18-CLARU, F.
- s. .
la
. hNSWER 3 . 0 7.' (1.50) on
.a. ' 130((0.53;tny: Rx % scram.(IRM'n > J20/125 of scale and mode switch.not in RUN) [0.53.
- b. Rx racirc.. flux controller auto switches. Lo APRf1 channel E. (0.5)
' REFERENCE G G N S ' O P - C 5 1 - 4 ,, L.O. #3 CGNdi DP-B33-2, L ., O . #4 !
GGNS OP-C51-2, L . O ., 41 5 _ q 215005K109 215003K101 202OO2K607 ...(KA*S)
' ANSWER. 3.08. (1.00)
When, activated - The source i s posi ti oned to arradiate the i detector caesing an upscale mater d ef l ec t i on ., ! (0.5) When deactivated - There is suf ficient leakage to cause a background 1evel reading- rio that a channel f allure would be indicated by a downstale. alarm. (0.5) REFERENCE' i GGNS'OP-D21-501. L.O. #3 and NRC Exam Bank 272OOOK303 -272OOOK206 ...(KA'S) i ANSWER 3.09 (2.00)
- a. Causes r eactor level to INCREASE EO.5] due to the Level Coritr ol Svutem having a LEVEL ERROR. with NO compensating FLOW ERROR resulting 'n a SIGNAL. to INCREASE the SPEED OF THE REACTOR FEED PUMPS-LO.5]. i
- b. Reactor l evel should REMAIN CONSTANT E0.53 because the "B" FEED PUt'o '
Turbine Control Unit will lock the pumps at the speed at'the time of the fa21ure LO.5]. 1 REFERENCE l GGNS OP-C34-501, L.O. #*s 6.a and 6.c. I 2590C2K507 259002K105 259002K103 ...(KA*S) I i i l l
. - - - - - _ -- - - - - - _ - _ - - ___ _-- .. J
l' l l' 5s_.JNSIBLIEgNIS OUI.) GOUIRO( S PAGE 4 21 j , . l ANSWERS - GRAND OLLF 1 -G7/05/18-CLARK, - F. ! l i 1 1 ! l 1' l ANSWER 3.10 (1,25) 1
- a. Cannot be teset (0.5)
- b. The contr ol s fer the RCIC systen on the P150 panel are- j
- 1. RC.?C Flow Control valve E. ad
- 2. Handrwitch to #' shi{t N ,contrgl and ' dof RCIC RCIC tc panel P150 (C ";)
- 3. Handswitch to .
"SY be.thh. u-419t (CSI CC6L)
- 9. Hard wnlA Tor V.t\C. y ecisen u \/d ug.h w._ (65 I - FO' s (Es t- FCAts)'5d .
REFERENCE 5 O p /-S k a ~t t\C.*M e.=.n 599) c , Ett C. G G N 5 O P --C 6 1 - O O ,, L.O. 43.d and 6.b (+o m atqvenes % L 7, c.c.it w A m 295000K007 ...(KA'9) b e. m C D E . M E \ C v m n m u.n D ua u a.t u v < 9 , ILO.\ t L A e_ O.n Ccx1=v- ucd ue io. ECLC l anese ; 6\a.ncl 4ha.L values c. cCs 5 T~mp( ESI- FOM 5,00ST)u .,'Te_ d l ANSWER ".11
. (3.00) (, g g o,75 cac, j
- a. Reactor Scram (0. 5 ) end G.y' 1salation
'I signal (0.5)
- b. Reactor scram (0.5) and Group I isolation (0.5)
- c. 1. All inboard and outboard MGIV shut (0.25)
- 2. The Recirculation system sample valve shuts (0.25)
- 3. The Mechanical Vacuum Pump (MVP) trips (If operating) (0.25)-
- 4. The MVP suction valve shuts .lf open) (0.25) hEFERENCE GGN5 DP D17-03, L. D . # 5.2.a. 5.T.a. 6.r, - 9.a and 7 272OOOK601 272 GOOK 402 272OOOK304 272OOOK101 ...(KA'S) j Ad5WER 3.12 (2.50)
- a. U-234 i s added to the coeting. (0.5)
- b. 1. B y p a t. s light on panel 608 J. LPRM bypass indication or the APRh front panel
- 3. LPRM bypass 11 cih t on the .ul1 Cor e D2sp1ay
- 4. % r eading on APRM w/ function nwitch in count (Any 3 first correct response 0.34, othero 0.33 each)
- c. 1. LPRM iipscale trip circult
- 2. LPRn downstal e trip circuit
- 3. The associated APRM channel
- 4. Plant Pr ocesn Computer
- 5. RCIS for i ndi c a ti on
- 6. Hr-e a m l e t i cc. E l m ,- ~. . m
, ._ M de + %bdibn bNc1 (4 h) 0.25 each)
REFER?NCE GSNS UP- C51 02, L.O. # ? 3.e, and 4 GONS OP-C51-4-02, L.O. # 7 215005K104 21DOOSE109 215005K406 215005K604 ...(EA'S)
Th._ _ld s!I[jyhE NI S,_ AU D,, C OUI BO L S PAGE 43 e ... ; ANSWERS - GRAND L,ULF 1 -87/05/18-CLARK, F. ANSWER 3.13- (1.50)
% . Starts in auto.on low flow througli the associated demin urn t. (0.5)
Stopu wnen condensate ficw is re-established (0.5)
- b. Precoat on thu filter dcmin may be lost (0.5)
REFERENCE GEN 5 OP C34-301, L . C; . # 3, 5.e.1 c c .1 6. 5 256000K404 ...(K4"G) fiNSWER 3.14 (1.DO)
- a. No (O.25) The wide range level indication is calibrated with no Jet' pump flow ( 0. 2b)- and at 100 % power the level indication will be 1ower due-to the draw down e f -f ec t (O.25)
- b. No (0.23)-TAF is .16'~ inchen (0.25) and tha low end of the wide range is -160 inches'.(0.25)
REFERENCE GGNS OP-E(21~O3, L.O. # 3.b and 4 l 216000K501 ...(KA'G)
' ANSWER 3.15 (2.00)
- a. 1. Diesel speed (frequencv) (0.S)
- 2. Di esel load contrcl (0. 5)
- b. 1. _Voltege control (0.5)
- 2. VAR contr ol (0. 5)
REFERENCE GGNS OF- P7"i-02. L.D. # 5 l 264000K305 264000N101 ...TA'S', l' L l l l l
. _ _ _ _ _ .__. _ _ _ _ _ . _ _ . _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ - . _ _ ._ _-a
. _3. __lUS.TBut1EUJE.,. QUDJONIBOL.S -' PAGE- 44 ANSWERS -- GRAND GULF 1 -87/05/18-CLARK, F. 1
]
1 l 1
)
1
\ .~15 AiGWER 3.16. (M) l I
- a. 1 ., Reactor or st.eam header pres.sure (0.25) 2 ., Generator load (MWe) (0.23)
- 3. Turbine speed (0.25) {
- b. Tl' e preLsure control circuit (0.5)
- c. Spe44-s 4wh e ! .ctt ~ .5' 'h ; , ,2 L c c cu . -m the c m2 ; g.m . __ , . im
" t i odt.- tM4mwp44) d I" , d f- ::p ;c j . ' - - 'al. t5c sp :i "
t h e-. l ce-sw-M<,; ' . G) R o n e_ (o ,5 ) l l REFERENCE ; 3GNS OP-N32--2, L.D. # 4 and 5 { 245000K409 245000K108 ...(KA'5) l 1 I l 1 l l
4 2. ,JtBOCEp6!BES_ ,_t40B(18L3._,,6BNQB!%_ElgBGENCL6dD f PAGE- 45'
?> SGDIOLQgICOL.JOMlBOL
~
. ANSWERS GRAND GULF 1 -87/05/18-CLARK, F.
l ANSWER 4.01 (1. 00)
- 1. The individual as in trainang to qualify for an operating 1icense-(0.5)
- 2. - 11u st be under the direction of a licensed RO or GRO and the RO/
SRO muut. be at the control s at all times ( 0. 5 ) REFERENCE GONS 01-S-06-26.4.5 2c?4001K105 ...(KW S)
) ,50 ANSWER- 4.02 (6M)
(.O .3 'J ) LO I b g' v a, Verify turbine. speed steady between 1800 and 1C50 RPM '(:: . 5)
- b. 1. Manually trip the turbine (with the MN TURB TRIP pushbutton on l the P680 panel) (0.5)
- 2. Run the MHC START DVC on P680 to zero (to bian the. turbine valves closed) (0.5)
~+ re iI. 6i -5. ,; .t - st ;! . .a : i,; g,g, ;,
cp. '
. tu + ' '- Cr -'
- 14: ;p '- " '
A-t b !_ , m , . ; u m. s l
.p , ti; , v .1 REFERENCE ONEP-05 l-02-I-2, Steps 4.4 and 4.5 246000A201 2450000014 ...(KA"S)
ANSWER 4.03 (2.00)
-c. 14.5 ieet (0.51
- b. 140 degrees F (0.5) c 21'2 degrees F (0.5) e , z.3
- d. The limitt ensure adequate NP5H ava2]able to the ECCS pumps. (+:-G )
TLM e .w-sums e2 de. ud e_ crs, Lug be-\(_ XQ1( Lh,_0.L (O,2_%) e REFERENCE GGNS EP-3, Steps L-16. L-17 and L-20 GGNS Ol'-SPD5-OU4-01 295029K201 295029k203 295029h209 29503Oh102 2950301:.203
... (LW S)
, O___[80CEDUBES_. _pDBd8Lt _GONOBOO!a._Et1[RGEUGL OUD PACE 46
* .. . BODIOLOGICOL._COUJOOL
\
y . ANGWERG - GRAND GULF 1 -87/05/10-CLARK. F. , J l
; i t
j l
. 2.00 j ANSWEP 4.04 (. . $)
'3 a. Containment sprav will be initiated incediately (0.5) (whether
.d er riot edequale core cool i ng car, be maintained ) l b b. No assurance can be alven that edecntet O.J e or e coc11ng can be l 1
l ma i sita) ied 1f th: containment fails (M> he failure mechanism l will probably result in loos of source water to the ECCS componente l
. :z i and a loss of adequate core cool i ng) s . E l
- c. Can be initiated (0.5) l
- d. (When the pressure and t.emp er a t ur e rel at i onehi p are in the thaded i portion af the spray initiation curve. the combi nati on o-f l evaporative c ool i ng and convective cooling), results i n depressurization rates which exceeds the negativt dess.gn pressure of the containment (O. 5) w o3 t , i R ca.ndic\1kt M dt'5 N.-
nErERENCE 9 Q qamh T>t 3 > n.25:t/j GONS OP-SPDS-LP-GO4-01, L.D. tt 13 " '4 . , Q em(;t "N5024K303 .. (KA'D) " Vcd b p e qcm ., <> h e tu s b'z. d - C " '('t ' 2.00 (' AN3WER 4.05 (T ?)
- a. 1. 1151 V low level isolation A---FM-meer y ,<1 , t em ,s . r, '.s ,
., . i s- . 2- ...;.2 ca
---4 !-wtt-dM*;-{U~S'*?*.".Yf ?
A b. 1. Util::stian of the main condenter as a heat sink to prevent e :ceedi ng the suppression pool heat capacity temperature limit i r, cf sufilcsent impor t ance to war r alit bypassing the i n t e.- l oc k (1.0) 2.--__J1vpm1_na_t12c_pniam:2_ c c:4t; inmout irnet-<.a.r--low Ami- 4w+4A%-on H ,e ustw+M.4maeM-wppby- A.+44*+--bM,(*d -H&I-V -bo--#Hw crcrurt i M 7)
', N OT C . G Q.ADES W CL 1E%L C (DN 100 % (WY\ i t0 A7 i C IO '7. C:DD REFERENCE ( Ei h Wit NM ,oc j D E W=w b 1 )
00NS OP-SPDD-LP-010-01, L.D. 4t a ., 4 and S 295025K205 275025h301 ...(LA"S> Se N 5 _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ O. _
# 1
,!41._lBOCliDL1EE~S_:_tJOB00L.c_0E209BOL _EMESGEUCLAUI) PAGE 47,
..y.. 80DIOLOGIC A .coVIBg6
. JANSWERS '- GRAND GULFS 1 ' -87/05/1G-CL. ARK, F. -
L i ANSWER- 4;06 'i2.00)
- 1. Secr>ndary containment ar na or 'HVAC rad l evel s above - max norm l evel s
~
'2. Secondary containment fl oor or sump- leve] s above the man nor'm leveln
- 3. . Secondary. containment-temperature or'HVAC differential temperature alsove _ man norm l evel s
- 4. Secondary containment differential pressure at or above zero inches water. . ;
(4 G) O.5 each) REFERENCE GGNS OP-SPDS-LP-006-01, L 0. # 2 295000G011' . .(KA'S) ANSWER .4.07 ( .50)
- 1. . (0.25) C 'De.w n w e dt. - sncb cdce Sta N 12% (0.25) 2.-
REFERENCE GGNS 03-1-01"1, Precautions and Limitations 2.1.9 and 2.1.8 215.COSK401 ... O(A' S) 2 00' ANSWER 4.08 (T _1)
- 1. Scram condition and reactor power greater than 4 % or undetermined
- 2. Drywell pressure ebove .1.23 psig.
- 3. Reattor pressur e above 1064.7 psig
- 4. Reactor vessel water below M 1.6 ihches or undeter mi ned (4 G) 0.5 each)
REFERENCE GGNS OP-SPDS-LP-OO3-0A, L.O. # ~ and 5 273000B011 ... (K4" 5) 1 l l l l l 1-
~
- 45. .PBQGEDUBES,..- _UOBd@L3,,_6pNQBd8tzt,_EUEBGENgy_AND. PAGE- '48
~ * . . . =. - RADIOLOGICAL CONTROL'
+- . ANSWERS -- GRAND GULF.'1'; -- 87 / 05 /18-CL ARK , F.
2.50 ANSWER 4.09 (1 ~ ? L
- a. Water ~1evel is allowed to decrease to ;
- 1. 2, u g i , mse 1 .. .
m.,.. _ a ,i , m.m ..m
, , - bu ( 9*7e (0,G )
m.w) e<~. um. , u , i m u, a
- 2. R! ' ' !'
c1-2r'sl1 exed to d e r r ._ _:- tc *" ')u, .. t , t.,, .
~
-1 e u i s v. Of Dt y ucd %%vye_ < \ 2.3 gr-Ag ef n o 5?N s r,.g c.h. ( O.G )
TAF w;,, c ;, 4
- 3. , , . _ . , ..1- , . 25 ) (o,s)
- b. 1 ., .CRD (_.15) (o.5)
- 2. SDLC E.5) (o, "J) because they are_being used to control reacter power (0.25)
REFERENCE GGNS.OP-SPDS-LP-010, L.O. # 3 and 4 i 295000G011 ...(KA S) ANSWER 4.10 ( .50) The precaution prevents hydrrang the recirculatico system to CRD pressure. (0.5) REFERENCE GGNS S O I--04 01- 01 -C 11 - 1,, Page 5 20?OO11t404 202OO1K405 . .,(KA"S) ANSWEh 4.11 (1.00) The pump must be star ted to prevent a 1 ass of hydrogen seal oil (0,5) snd p r .e v en t s the possible ignition of the hydrogen f rom the gas er-cape vel oc i ty (O.D) REFERENCE GGNS ONEP-05-1-02--I-4 294OO1K115 295003K206 ...(KA"S) l l l I l l i l l l i _ _ _ _ _ _ _ .-__ I
% . PBOCED(JBES_7 NOBUGL3_GB[40SL10L t, EUEOGEUCL GUD PAGE 49
* - ** . - BfMDTl:OGICOL._CDUTPOL
. ANSWERS - GRAND GULF 1 -87/OS/10-CLARK, F. l l
l ANSWER '+ . 1 2 (2.00)
\
- a. 2 1
- b. 4
- c. 1
- d. 3 REFERENCE GGNS: Procedur es 10-S-01--2. 7. 4 & 5: OP-EM-502, 503, 504, b 505 i 29E0006002 ...(KA'5)
ANDWER 4.13 (2.00)
- a. J. DNE (1) hour (GG admin quarterly limit w2thout form 4 on fale s 100:: mrem) (0.5)
- 7. ELEVEN (11) hours (O. L)
- b. IMMEDIATELY ex 2 L the area a 1d IMMEDIATELY report ta HEALTH PHYSICS (1.0)
REFERENCE CFR 20 UGNG Adm2n procedure 01 - B -- 0 8 - 2 ., PagP 16 and Attachment III. Page 2 294001K103 ... (KA'S) ANSWEP 4.14 (1.50)
- a. To prevent wetting of the c harcoal adsorber bed. i0.5)
- b. 1. To allow ts y s t e;n warmup . . . . , CO. G )
- 2. To r emove- any hydrogen from the system (~ h Co.di~)
REFERENCE GDNG GDI - 1- 01 -N64-1
;11000).;4 N , .(KW 5)
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ = _ _ _ ____ -
, ih _ _EBOCEDVBE[$_;,,,UDBOO( . _O[jyOBdOL1_EMEE[l[NCy._OhlD PAGE 50
= .. , - CODIOLOGICOL_ CON.IBOL
- ANSWLRS - GRAND OULF 1 -87/05/10-CLARK. F.
I 1 ANSWER 4.15 (2.00) I 1
- a. 1. Preclude injury to percor nel
- 2. Pr at et t the Public health and safety
- 3. Prevent vi t al plant equipment damage
- 4. Immediate actions mus.t be taken to mitigate the consequences of the emergency (No procedur e addresses the problem)
(3 G 0.5 each)
- b. As soon as prattItal advise t tw a r appropriate ammediate supervisor (0.5)
REFERENCE GGNS Procedure 01-5-02-1. Page 4 2070000015 223OOOG015 2630000015 ...(KA'S) ANSWER 4.16 (1.00) a ., Voltage gradient capactors will overheat (0.5)
- b. Open the disconnects (O. S)
REF EREl' ICE GONS ONEP 05-1-O?-I-02 262OO1K103 ...(KA"S)
TEST CROSS REFERENCE PAGE 1 v , '. e 6 . O,UESTION 'VALUEl REFERENCE 01.01 1.25 FECOOOO184 01.02 1.50 FECOOOO188 01.03 1.50 FiiCOOOO109
-01.04 .50 FECOOOO190 01.05 1.00 FECOOOO191 01.06 1.00 FECOOOO192 01.07 1.00 FECOOOO193
- 01. 08 .50 FECOOOO194 01.09 2,50 FECOOOO197 01.10 1.00 FECOOOO237 il 01.11 1.00 FECOOOO254 01.12 1.00 FECOOOO236 01.15 1.00 FECOOOO239 01.14- 1.00 FECOOOO242 01.15 1.0c FECOOOO243 ,
01.16 -1,00 FECOOOO241 l 01,17 2.00 FECOOOO246 01.18 2;50 FECOOOO251 01.19 1.50 FECOOOO245 01.20 1.'50 FECOOOO253 Q1.21 2.00 FECOOOO248 27.25 02.01 1,00 FECOOOO255 02.02 2.50 FECOOOO270 , 02.03 2.00 FECOOOO263 02.04 2.00 FECOOOO256 02.05 1.50 FECOOOO261 02.06 1.00 FECOOOO262 02.07 1.00 FECOOOO264 , 02.00 ?.00 FECOOOO265 ~! 02.09 1.50 FECOOOO267 02.10 1.50 FECOOOO268 02.11 1.00 FECOOOO272 02,12 2.50 FECOOOO258 02.13 1.00 FECOOOO259 02.14 1.00 FECOOOO260 02.15 2.00 FEC.OOOO266 02.16 1.75 FECOOOO269 02.17 1.00 FECOOOO271 02.10 1.50 FECOOOO257 27.75 03.01 1.00 FECOOOO198 03.02 1.00 FECOOOO199 03.03 1.00 FECOOOO2OO 03.04 2.00 FECOOOO201 03.05 3.00 F EC OOt:.>0202
TEST CROSS REFEFiENCE PAGE 2
- s. . . . .
OLfESTION VALUE REFERENCE 03.06 1.00 FECOnOO203 03.07 1.50 FECOOOO204 0.7.00 1.00 FECOOOO205 03.09 2.00 FECOOOO206 03.10 1. .'5 FECOOOO207 03,11 3.00 FECOOOO200 03.12 2.50 FECOOOO209 03,13 1.50 FECOOO>210 03,,14 1.50 FECOOOO213 03.15 2.00 FECOOOO214 03.16 2.25 FECOOOO215 27.50 04.01 1.00 FECOOOO216 i
- 04. O'? ' 00 FECOOOO217 04.03 2.00 FECOOOO219 04.04 z.25 FECOOOO220 04.05 '.00
. FECOOOO221 04.06 2.00 FECOOOO222 04.07 .50 FECOOOO223 04.00 3.00 FECOOOO224 04.09 2.00 FECOOOO225 04.10 .50 FECOOOO226 04.11 1.00 FECOOOO227 04,19 '? . 00 FECOOOO230 04.13 2.00 FECOOOO231 04.14 1.50 FECOOOO232 04.15 2.00 FECOOOO234 04.16 1.00 FECOOOO235 27.'75 110.25
, Endom z g QM(-
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _GRANQ_@ULF_1____________ REACTOR TYPE: _gWB-gE6_________________ DATE ADMINISTERED: _@ZZ99f1@________________ EXAMINER: _PAYNEz_p.Cz _____________ CANDIDATE: _________________________ . INSIBUGIl999_I9_G8BDID8IE1 Use separate paper for the answers. Write answers on one side only. Steple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after tna examination starts.
% OF CATEGOR( % OF CANDIDATE'S CATEGORY
-_Y86UE_ _19186 ___SCQBE___ _y@6UE__ ______________C@lEGQBY_____________ _2@t@9__ _2bz21 ___________ ___. ___ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 26z2E__ _23169 ___________ ________
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_26z99__ _2E121 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 22xE9__ _2Ez22 ___________ ________ B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1991ZE-_ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature l'V\ ASTER
^
'u NRC' RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration' of 'thic examination the f ollowing rules apply:
'11: Cheating. on the examination means an automatic denial of your application I and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to f acilitate legiblo reproductions.
- 4. Print your name in the blank provided on the cover LP of the ex ami nati on . 1 ti. ' Fill in the date on the cover sheet of the examination (if necessary) .
- 6. Use only the paper provided for answers.
- 7. ' Print your name in the upper right-hand corner of the first page of gach !
section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write gnly gn gne side of the paper,.and write "Last Page" on the last answer sheet.
l .9. Number each answer as to category and number, for example, 1.4, 6.3. 'j 110 . Skip at least three lines between each answer. 1 (11. Separate answer sheets from pad and place finished answer sheets face j down on your desk or table. i
- 12. Use abbreviations only if they are commonly used in facility litetatute. j L 13. The point value f or each question is indicated in parentheses after the question and can be used as a guide f or the depth of answer required.
- 14. Show all c al cul at i on s, methods, or assumptions used to obtain an answer to mathematical' problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examinet only.
(17. You must sign the statement on the cover sheet that indicates that the l work is your own and you have not received or been given assistance in
. compl eti ng the examinati on. This must be done after the examination has been completed. I l_ _ ________________________i
= ,
i
- 18. When you complete your examinat;on, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exar aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as d efined by the examiner. If af ter' leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
'5:__IUE96Y_9E_ NUCLE 98_EQWE3_g69NI_gPEB91]QN2 _ELUJpg1_$Np PAGE .2
, .T H_ _E _R _M _O _D _Y _N _A _M _I C_ _S -
QUESTION 5.01 (1.00)- WHICH ONE of the f ollowing neutron sources is NOT considered an INTRINSIC neutron source?
- a. Cm-242
- b. Sb-Be
- c. gamma-deuterium (photo ineutron)
- d. alpha-oxygen QUESTION 5.02 (1.25)
INDICATE if the below listed parameters will INCREASE, DECREASE or REMAIN THE SAME, if the facility experienced a " JET PUMP RAMS HEAD HOLDDOWN
. FAILURE".(ASSUME THE FAILED JET PUMP IS NOT FULLY INSTRUMENTED AND RECIRC CONTROL IS IN " FLUX AUTO")
- n. Flow for the Failed Jet Pump
- b. Core Differential Pressure
- c. Reactor (APRM) Power
- d. Indicated Core Flow
- e. Actual Core F1os
- i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) _ _ ________ _ a
PAGE 3 '5___IHggBy_gE_NUCL@@B_EgyE3_fL9NJ_gEgg@IJgNy_ FLU]Qgz_9ND IHEBMOQVNAMICS o QUESTION 5.03 (2.00) With regard to PCIOMR, IDENTIFY EACH of the f ollowing statements as TRUE or FALSE.
- a. PCI failures are dependent on absolute pcwer, increase in power, duration of power in:rease, previous power history and fuel exposure,
- b. Regardless of the stress 1 state, strain rate, temperature, presence of i dine and state of irradiation, zircalloy f uel tubes will be u
J' ductile.
- c. If power l evel is reduced prior tc completing the 12 hour soak, pre-conditioning is resumed at either the new power level or the power l evel 12 hours before the power decrease, whichever is higher.
- d. PCIOMR limits are given in Section 3/4.2 of the Technical Specifications.
QUESTION 5.04 (2.00)
- a. Define Net Positive Suction Head (NPSH).
- b. Opening the Recirc System Flow Control Valve (FCV) will cause the available NPSH for the Retirc Pumps to (INCREASE, DECREASE, or REMAIN THE SAME)?
- c. Will the required NPSH for the Retirc Pumps (INCREASES, DECREASES, or REMAINS THE SAME) when the Recirc Pumps arc shifted from slow speed (15 Hz) to fast speed (60 Hz)?
QUESTION 5.05 (1.00) A temperature instrument with an out-of-date sticker on it is reading 400 degrees F. A recently calibrated pressure gage, sensing the same area, indicates 350 psig. CALCULATE the temperature which should be indicated on the temperature instrument? (SHOW ALL WORK AND ASSUME SATURATED CONDITIONS) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'5 __IHgggy_gE_NggLgg3_fgygB_EL99I_gggB9))QN2 _g(glggz_gNg PAGE 4 THERMODYNAMICS QUESTION 5.06 (1.00) STATE the mode (s) of heat transfer for the following situatiens
- a. Center of fuel pellet out to the pellet edge.
- b. Across the Helium gap in the fuel rod.
- c. Clad surfe:e to the center of the coolant channel.
d .. Clad surf ace to coolant under film boiling conditions. QUES TION 5.07 (1.25) MATCH one of the -items listed below with each of the five statements. A lotter number sequence is suf ficient.
- 1. GEXL 6. PCIDMR
- 2. APLHGR 7. MCPR
- 3. CPR B. Total PF
- 4. FLPD 9. LHGR
- 5. Axial' PF 10. Radial PF
- o. Parameter by which plastic strain and deformation are limited to less than 1 */. .
- b. The ratio of bundle power required to pradace OTB in the bundle to actual bundle power.
- c. Parameter by which peak clad temperature is maintained less than 2200 degrees F during a design basis accident.
- d. The ratio of individual bundle power to core average bundle power.
- e. The rctio of nodal power to a/erage nodal power.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'5:__ISEQ8Y_gE_NQCLggg_EgWEg_f68NI_QEEBSJJgN2 ,.,[69Jgg3_6Ng PAGE 5 l
THERMODYNAMICS i QUESTION 5.08 (2.00)
. 1
~
The attachud Figure #1 111ustrates a transient that could occur at a BWR, l
.Given: (1) No operate r actions are taken ]
(2) Normal functioning of instrumentation, control and protection systems (3) Normal f unctioning oi the Reactor Protection System a, . IDENTIFY the initial cause of this event.
- b. EXPLAIN the causo, and actuation signal (s) if appropriate, of the following recorder indications:
(1) Reactor level increase from time t =
- 4.5 see to time
, t. = 7.5 sec (Graph 1) (2) Initial decrease in neutron flux from tine t = 0 to time t =
- 3 sec (Graph 2)
(3) Vessel pressure increase f rom time t =
- 3 sec to time t =
- 5 sec (Graph 5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
' 5 __Ibs95Y_9E. 09CLE95_E9Wg6_E68EI_gEEBSIJgN 2_E(91933,@NQ PAGE 6
, 'I!jsB099YN6d]CS QUESTION 5.09 (2.50)
The attached Figure #2 illustrates a transient that could occur at a E WR. 1 Given: (1) No operator actions are taken 1 (2) Normal functioning of instrumentation, control and protection systems (3) Normal functioning of the Reactor Protection System
- o. IDENTIFY the i ni ti al cause of this event.
- b. EXPLAIN the cause, and actuation si gnal (s) if appropriate, of the f ollowing recorder indications:
(1) Reactor pressure increase from time t =
- 1.8 see to time t =
- 3.5 sec (Graph 2)
(2) Initial decrease in neutron flux from time t = 0 to time t =
- 2.0 sec (Graph 1)
(3) Decrease in Vessel Steam flow f rom time t =* 1 sec to time t =' 1.6 sec (Graph 3) (4) Increase in Vessel Steam flow f rom time t =
- 3.4 see to time t =
- 3.75 sec (Graph 3)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) s
'5:__IHggdy_gE_NgC6ggg_fgWEB_fkggI_QEgggIJgN 2_Ekg}Qg1_gNp PAGE 7
_T_H_E_R_MO_D_Y_NA__M_IC_S , _ QUESTION 5.10 (1.00)
- a. Approximately WHAT percentage of neutrons from U-235 are born delayed? (0.5)
- 1. O.26%
- 2. 0.65%
- 3. 0.88%
- 4. 1.48%
- b. The power generated by the reactor at the beginning of core life comes from U-235 thermal fission and U-238 fast fission. Later in core life, larger and larger fractions of power generation are produced by fission of WHAT isotope? (0.5)
- 1. Cm-244
- 2. Am-241
- 3. Pu-238
- 4. Pu-239 QUESTION 5.11 (1.00)
WHICH ONE of the f ollowing most accurately describes the Point of Adding Heat (POAH) and its effect on reactor power?
- a. Due to the moderator temperature coefficient, positive reactivity is inserted which raises power and temperature.
- b. The Heating Range starts occurring at about 10% power.
- c. Heat production in the reactor at the POAH is suf ficient to offset ambient losses,
- d. Once PDAH i s attained, additional rod manipuistions are not required until normal operating pressure is reached.
1 l J (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) i i l J
'51- IHEggy_gE_NUC(g65_EgWEB_E(9NI_QfEBBIlgy2_E(UlDSg_6ND PAGE B THERMODYNAMICS I
QUESTION 5.12 (1.00) While operating at power, a RWCU filter demineralized ruptures causing ; resin intrusion into the reactor vessel. State the effects (INCREASE, DECREASE, REMAIN THE SAME) for EACH of the f ollowing:
- a. Reactor water pH.
- b. Reactor water conductivity
- c. Steam line nitrogen-16 activity
- d. Reactor Water ac ti vi ty.
QUESTION 5.13 (1.50) With a recirculation flow INCREASE ct power, characterize (INCREASE, DECREASE, REMAIN THE SAME) the effect on the final equilibrium values of the f ollowings
- a. Void fraction
- b. Doppl er reacti vi ty coef ficient
- c. Feedwater enthalpy QUESTION 5.14 (2.00)
Answer EACH of the f ollowing TRUE or FALSE.
- c. Void Coefficient is directly proportional to core size.
- b. The larger the core, the lower the relative neutron leakage.
- c. The two isotopes that make the largest contri buti on to the Doppler effect are U-235 and Pu-240.
- d. During start-up phases, the effect of the Temperature Coefficient is insignificant in controlling power wher, compared to the effects of the Doppler and Void Coefficients.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'5 __IHEg6X_QE_ NUCLE @B_CQWE6_E6@NI_g6E6611gN t _ELUlgSt_@Ng PAGE 9 It!E6tjQQR!@MICS
' QUESTION 5.15 (2.50)
Answar EACH of the f ollowing TRUE or FALSE.
- a. Xenon and Samarium concentrations increase f ollowing a scram f rom high power operation (within the first five hours).
b.. Samarium has a higher microscopic absorption cross section than Xenon.
- c. A reactor start-up several days after a scram from extended high power operati on is considered to be. Xenon and Samarium free,
- d. The equilibrium concentration of Samarium at 50% power is approximately the same as at 100% power.
- e. The equilibrium concentration of Xenon at 50% power is approximately one-half the equilibrium concentration at 100% power.
QUESTION 5.16 (1.00) FILL IN THE BLANKS The stable reactor period following the prompt drop in reactor power from control rod insertion after a scram will be ___________ (time) because of _______________. QUESTION 5.17 (1.50) LIST three (3) 3 actors upon which a reactor's decay heat generation l rate is dependent. QUESTION 5.18 (1.00) EXPLAIN why the Fuel Temperature Coefficient (Doppler Coefficient) changes as Void Fraction INCREASES. (INCLUDE in your EXPLANATION whether it becomes MORE or LESS negative). I (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ h
'5:__IHEgBY_gE_ NUCLE @B_EQWE8_E6@NI_QEEB8IlgNt _E6U1DS i_@Ng PAGE 10
, TH_E_R_M_O_D_YN_A_M__IC_S_
QUESTION 5.19 (2.00) Use the 1/M plot and predict the number of control' rods required to be withdrawn to achieve criticality. NOTES: .1. CR = C'ount Rate
- 2. USE THE FIGURE BELOW TO SKETCH YOUR SOLUTION CRO = 40 cps CR4 = 191 cps CR1 = 50 cps CR5 = 333 cps CR2 = 89 cps CR6 = 800 cps CR3 = 129 cps Each CR reading is recorded following a 5 rod withdrawal with CRO representing 100% rod density.
5 10 15 20 25 30 35 40 45 50 55 1.01----!----l----l ---l---- ----
----l----!----l----l---- --- 1.0 O.9- -0.9 O.8- -0.8 O.7- ~0.7 1/M O.6- -0.6 O.5- -0.5 O.4- -0.4 O 3- -0.3 O.2- -0.2 O.1- -0.1 O201----:----l----!----l---- ----l----l----l----l----l----l---!
O 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn (***** END OF CATEGORY 05 *****)
'6___PL9NI_pYgIEMg_DEg]gN 2 _CgNIggL 2 _96D_JygIBUMgyI9IJgy PAGE 11 QUESTION 6.01 (1.00)
LIST the four interlocks that inhibit recirc FCV motion. (SETPOINTS NOT REQUIRED)
-QUESTION 6.02 (2.00)
EXPLAIN what condition will generate EACH of the following indications on the Operator Control Module (P6BO):
- a. Data Fault
- b. Scram Valves
- c. Channel Disagree
- d. Insert Required QUESTION 6.03 (1.00)
List FOUR (4) systems that have components that can be operated or controlled 'f rom BOTH Remote Shutdown Panels P150 and P151. QUESTION 6.04 (1.50) With regard to the Nuclear Instrumentation Systems
- a. During a reactor shutdown, with the mode switch in STARTUP, the IRM's are rerding 13 on range 5. You downscale all IRM's to range 4.
STATE the expected level reading which would occur and LIST any automatic action (s) - JUSTIFY your answer.
- b. While operating at 100% power, you bypass APRM Channel A. STATE the effect, if any, that this has on the reactor recirculation system?
QUESTION 6.05 (1.00) l The Area Radiation Menitors have installed check sources which, when l activated or deactivated, will provide an indication of an ARM's l operability. BRIEFLY describe HOW operability is demonstrated with the check source in BOTH the activated and deactivated condition. i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) L I i _ _ _ _ _ _ _ _ _ J
- '6 3__PL@NI_pYSIggp_DEp]GN 1_CgDIggL _@Ng_INDIBUDENI@IJgN 1 PAGE 12 1
QUESTION 6.06 (2.00) Assume the FEEDWATER LEVEL CONTROL SYSTEM is being operated in 3-ELEMENT control using reactor LEVEL DETECTOR CHANNEL 'A'. Reactor powar is at 85%, STEADY STATE. For each of the instrument or control signal f ailures listed below. STATE HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase, decrease, or remain constant) and BRIEFLY EXPLAIN WHY in terms of WHAT is happening in the Foedwater Control System IMMEDIATELY AFTER THE FAILURE. (FOR EXAMPLE, your answers should include the ic11owiag detail, "Causes reactor level to decrease due to a steam flow /f eed flow error signal, steam flow < feed flow, resulting in a signal to increase the speed of the reactor feedpump(s)," IF APPLICABLE.)
- a. . Channel A REACTOR LEVEL detector signal fails LOW
- b. LOSS OF CONTROL SIGNAL to B Reactor Feed Pump Speed Controller QUESTION 6.07 (2.00)
Concerning the Division I Di esel Generator PARALLEL CONTROL HANDSWITCH.
- a. EXPLAIN the TWO (2) functions accomplished by placing the parallel control handswitch in tne PARALLEL positon.
- b. STATE what the GOVERNOR CONTROL is used for:
- 1. BEFORE the output breaker is closed.
- 2. AFTER the output breaker is closed.
QUESTION 6.08 (1.50) When op9 rating at 100% power, the wide range level instr uments read lower than the narrow range indication.
- a. STATE if this mismatch is a concern to the operator? EXPLAIN the answer.
- b. STALE if the wide range level instruments can be us9d to accurately indicate level at the TOP cf active fuel? EXPLAIN the answer.
(***** CATEGORY 06 CONTINUED ON NEXT PACE *****)
"62L.P[9NI_pypIEdg_pgg]gN_ggNIggL_BNp_]NgIBUdgyI91]gd z 3 PAGE 13 QUESTION 6.09 (1.75)
Concerning the Electrohydraulic Control Cystem (EHC):
- a. LIST the THREE (3) parameters sensed and evaluated by the coitrol system. (0.75)
- b. STATE what controlling circuit is actually positioning the control valves. ASSUME 100% power and normal svstem operation. (0.5)
- c. STATE what control circuit is out of the, control scheme when the Main Generator is synchronized to the grid. (0.5)
QUESTION 6.10 (2.50) Consider en Off-Normal Event in which Instrument Air System pressure is lost.
- a. How will the following valves FAIL? (CLOSED, OPEN, or AS IS)
- 1. CRD Flow Control Valve
- 7. RFP Minimum Flow Valve
- 3. Stram Valves
- 4. CCW Surge Tank Make-up Valve (2.0)
- b. EXPLAIN the cause of the potential Hi gh Radiation l evels in the Off-Gas Building. (0.5)
QUESTION 6.11 (2.00) The Standby Li quid Control (SLC) System is designed to provi de enough reactivity compensation to shut down the reactor from rated to zero power including shutdown margin and to allow cooling the nuclear system to 70 deg F with control rods remaining withdrawn in rated power pattern. LIST four (4) additional reactivity gains that must be compensated f or by SLC upon actuation as specified by the system's design bases. ) l 1 l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
'6.__PLgNI_gYgJgMg_DEg]QB1_Cg6J69L _gND_JNg16UMgNJgIJgN 2
PAGE 14 QUESTION 6.12 (1.00) The plant is operating at 90% power when the operator notices an INCREASE in the #2 seal pressure on Recirc Pump "A". The #1 seal pressure for the pump has NOT increased. These observations are f oll owed shortl y by a PUMP "A" EEAL STAGING FLOW HIGH/ LOW alarm. LIST the TWO malf unctions which give this alarm f or the given conditions. QUESTION 6.13 (2.00)
- a. LIST four (4) trip functions which ARE bypassed when the DIVISION III Emergency Diesel Generator is operating on a LOCA initiation signal. (1 O)
- b. LIST three (3) signals which wil? AUTOMATICALLY start the Division III EDG and bring it up to speed and voltage.
QUESTION 6.14 (1.50) With regard to the diesel driven FIRE PUMPS:
- a. LIST the discharge pressure at which the "A" and "B" di esel driven fire pumps will AUTOMATICALLY start.
- b. LIST the number of AUTOMATIC start attempts which EACH diesel will make.
- c. If the diesel goes throuch all of its attempts without starting, LIST the action (s) which must be taken in order to try to start the diesel. BE SPECIFIC AS TO LOCATION (S) AND CONTROL (S) TO BE OPERATED!
QUESTION 6.15 (1.00) Dur' ; reactor start-up under Cold Conditions, the operator adjusts the ontros Rod Drive pressure control valve to maintain a +260 psid between CRD and reactor pressure. EXPLAIN HOW this pressure differential is AUTOMATICALLY maintained as reactor pressure increases during the ensuing start-up. (IGNORE OPERATOR REQUIRED MINOR ADJUSTMENTS) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
'6___PL9NJ_gYglgM3_Qg3Jg31_CQNIBQL 2_9NQ_JN316UMENJgIIgN PAGE 15
= .
QUESTION 6.16 (1.25) The Standby Gas Treatment System (SBGTS) is in operation with an auto initiation signal present. EXPLAIN the effect(s) on SBGTS Train "A" for the f ollowing conditions, a, ' The SBGT9 Mode Sel ect Switch is in AUTO. The operator places SBGTS Div I " MAN INIT RESET SW" to RESET and than returns to NORMAL, then he places the " ENCL BLDG RECIRC FAN 'A'" and "SBGTS FILTER TRAIN 'A'" switches to STOP.
- b. The SBGTS Mode Select Switch is in STANDBY cnd the operator perf orms the same switch manipulations as in part "a" above.
QUESTION 6.17 (1.75) Answer the f ollowing with respect to the Liquid Radwaste System:
- a. LIST the three primary divisions of the system. (0.75)
- b. You are making a normal radwaste discharge to the basiri via the Li quid Radwaste Discharge Isolation Valve (F0355) when suddenly it closes. Give four (4) signals that will cause this valve to AUTOMATICALLY shut. (1.0) l
) (***** END OF CATEGORY 06 *****)
'Zz__PBgggguggg_;_BggdB61_BgNgBD862_EDggggNgy_BNg- PAGE 16 .. 6.891969919eb_99NIggb l
l QUESTION 7.01 (1.50) According to ONEP-05-1-02-I-2; " Turbine and Generator Trips" IMMEDIATE OPERATOR ACTIONS:
- a. STATE how to determine if the EHC speed demand circuit is controlling turbine speed. (SETPOINT REQUIRED for full credit)
- 6. STATE the IMMEDIATE ACTIONS if the Generator has tripped and Turbine has not tripped.
QUESTION 7.02 (1.50) Accor di ng to ONEP-05-1-02-I-3, "Si ngl e MSIV Cl osure", STATE what actions are' required when:
- a. A reactor SCRAM should have occured but did not. l
- b. A GROUP I isolation should have occurred but did not.
QUESTION 7.03 (2.00) Concerning EP-3, Containment Control:
- a. STATE the MINIMUM Suppression Pool Level that will be maintained when operating PCIC, LPCS, HPCS, and RHR in the LFCI mode,
- b. STATE the MAXIMUM Suppression Pool Temperature while operating RCIC.
- c. STATE the MAXIMUM Suppression Pool Temperature while operating HPCS, RHR in LPCI mode, and LPCS.
- d. STATE the BASIS for the temperature and level limitations, (i n a., b., and c. above), for the Suppression Pool.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE ******) __ ________-_-__-_A
'Z:__fegCgouggg_ _Ng808L _BgNg8 1 DBL 2_gdggggNgy_9ND PAGE 17 RADIOLOGICAL CONTROL QUESTION 7.04 (2.00)
Concerning Containment Pressure Control per EP-3:
- a. STATE what must be accomplished when Containment pressure reaches 17.25 psig.
- b. STATE the BASIS for part a. above.
- c. STATE if Containment Spray CAN BE INITIATED or CANNOT BE INITIATED when containment temperature is 150 degrees F and Containment Pressur e is 9 psig. (REFER to attached figure CN-T-1.)
- d. EXPLAIN the BASIS for precluding containment spray actuation tshaded area) per CN-T-1. i l
l QUESTION 7.05 (2.00) EP-14, " Level / Power Control" permits bypassing two low l evel i sol ati ons:
- a. STATE the TWO (2) low level isolation that may be bypassed.
- b. STATE the BASIS for allowing the bypassing of these isolations.
QUESTION 7.06 (2.00) LIST the entry conditions for EP-5, " Secondary Containment Control". (INCLUDE SETPOINTS) (TABLES NOT REQUIRED - JUST THEIR TOPICS) DUESTION 7.07 (2.00) l LIST the entry conditions for EP-2. (INCLUDE SETPOINTS) ) 1 I I i i L l I i I I (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) l L . _ _ _ _ _ - - - _ _ _ _ _
'Z__,PDggEQUBEp_;_NgBd861_@gygBMBL3_EdEggEygY_BNQ PAGE 10
.BSP1960eIG66_99NIBQ6 QUESTION 7.08 (2.50)
In step Li'-13 of EP-14, " Level / Power Control ", lowering of reactor vessel level is directed:
- a. LIST the THREE (3) condi ti on (s) /l evel (s) at which the lowering of level is terminated,
- b. LIST ALL inj ection systems which are not terminated by Step LP-13.
QUESTION 7.C9 (1.00) IMMEDIATE OPERATOR ACTION step 4.5 of DNEP-05-1-02-I-4, " Loss of Offsite POWER" requires the operator to check that three specific DC oil pumps auto start or to manually start them. STATE why the DC Generator Seal Oil Pump is needed and WHAT dangerous condition is prevented by operating the pump. QUESTION 7.10 (3.00) A condition arises that requires entry into a HIGH RADIATION AREA. } The operator will recieve an estimated whole body dose of 50 mrem. J You have the following data available: } Candidate 1 2 3 Age 27 38 24 Exposure: Week 15 mrem 35 mrem 5 mrem Guarter 2954 mrem 2460 mrem 16 mrem Life 18000 mrem 45720 mrem 29995 mrem Remarks: NRC FORM 4 on file YES YES NO Each candidate i s technically competent and physically capable of performing the task. Emergency limi ts do not apply and time constraints do not permit finding another candidate. EXPLAIN the reason f or accepting or rej ecting each candidhte. (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) i
'Z1__EBQgEDUBES_;_NOBd@bt_@SNOBd@(t_EDEBggggy_AND PAGE 19
, B,9D1Q(Q@lgAL_QONIBOL QUESTION 7.11 (1.S0) 'a. STATE why the OFFGAS syster CHARCOAL ADSORBER beds are bypsssed during i ni ti al system startup,
- b. STATE why a dry air purge must be established for the RECOMBINER TRAIN:
- 1. PRIOR to STARTUP?
- 2. IMMEDIATELY after SHUTDOWN?
QUESTION 7.12 (2.00)
- a. GGNS P-ocedure 01-S-02-1, " Description and Use of the GGNS Operations Manual ", li sts f our requirements, one of which must be met prior to deviating from approved procedures. LIST three of these requirements.
- b. WHO and WHEN must plant personnel notify after deviating from approved procedures.
QUESTION 7.13 (1.00) Procedure ONEP 05-1-C2-I-02, " Turbine and Generator Tripu," provides the operator with a CAUTION not to allow the 500 KV Breaker to be OPEN with VOLTAGE applied to i t f or greater than THIRTY (30) minutes.
- a. STATE the BASIS for this precaution (limiting component) . ]
- b. LIST the action required if the 500 KV Breaker will be open for more than THIRTY minutes.
QUESTION 7.14 02.00) Concerr.ing the GGNS Emergency Plant
- a. STATE the locations of the Technical Support Center (TSC) and the Operations Support Center (OSC). (1.0)
- b. STATE the two (2) responsibilities which the Site Emergency Director (ED) CAN NOT delegate. (1.0)
(***** END OF CATEGORY 07 *****)
'9 __e901NJgIggIlyE_P89CggUBEB2_CgUgIIJgNB,_gNg_LJUJIgIJgNB PAGE 20 QUESTION B.01 (1.00) i Ragarding the Protective Tagging System (AP 01-S-06-1), WHICH ONE of the following is NOT a purpose of the system?
- a. Provides administrative controls necessary to prevent the release of radioactive materials to the environment.
- b. Establishes the division of tagging authoritv between the Operations and Maintenance Sections.
- c. Provides detailed instructions for the use of Red Equipment Clearances and Information Tags.
- d. Provides administrative controls necessary to control equipment used in order to prevent personnel injury or equipment damage.
QUESTIOM B.02 (1.00) According to AP 01-S-06-2, "ConLuct of Operations", WHICH ONE of the following is NOT a responsibi?.ity of the Shi f t Supervi sa-?
- a. functioning as Shift Fire Chief
- b. approving Drywell access
- c. enkuring a licensed operator is present at-t he-c on t r ol s
- d. ensuring the operators on his shift conduct a proper shift ,
turnover QUESTION 8.03 (1.50) With the exception of breaker position, LIST the THREE (3) items an cperator should check on a breaker, if applicable, during the performance of a system lineup checksheet per 02-S-01-2, " Control and Use of Operations Section Directives"? Cor. sider LOCAL checks pal y and a 4.16 KV I.T.E. Circuit Breaker as an example. (***** CATEGORY OB CONTINUED ON NEXT PAGE *****)
l
'B- __0901Ni@I6@llyE_PBgCEQg8E@t_CQNgillgN@t_@@Q_LIN11@IlgNS PAGE 21 a .
l QUESTICN B.04 (1.00) The symbols (*) and (#) may proceed step (s) in the Pre-Startup Checksheet of an 01. STATE the meaning of each of these symbols. NOTE: FIGURE # 533 IS PROVIDED AS AN EXAMPLE. QUESTION 8.05 (1.00) FILL IN THE BLANK with one of the f ollowing 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 f ailure trips. "
- a. Channel Calibration
- b. Channel Check
- c. Channel Functional Test
- d. Logic System Functinnal Test l
l l l (***** CATEGORY 08 CONTINUED ON NEXT PAG 6 *****)
l 'R:__0DMINig16@llyE_P6gCEDURE5t_CQNDlIlgNS t_@ND_(lMlI@IlgN@ PAGE 22 OUESTION 8.06 (1.00) Technical Speci f i cati on 3/ 4.1. 4, " Control Hod Program Controls - Control Rod Withdrawal" states that control rods shall nct be withdrawn in OPERATIONAL CONDITIONS 1 and 2 when the main turbine bypass valves are not fully closed and when THERMAL POWER 1s greater than the low power setpoint of the RC&IS. WHICH ONE of the f ollowing best expresses the basis for this requirement ?
- a. Control rod withdrawal sequences are established at less than 20% RATED THERMAL POWER to ensure peak fuel enthalpy is maintained less than 200 cal /gme
- b. To prevent two methods of reactivity addition (rod manipulations and steam flow during main turbine start-up) from acting in an additive manner when Reactor Protection System instrument response is slow (< 10% RATFO THERMAL POWER) .
- c. At the end of core life with reactor power greater than 20%
and main turbine bypass valves open, there may insufficient capacity to ted steam flow during an Anticipated Transient Without Scram (ATWG) event.
- d. The rod withdrawal limiter system input power signal will indicate a core power level which is less than true core power.
(***** CATEGORY OS CONTINUED ON NEXT PAGE *****)
'9___B9dINJgIBg))yg_BBgggguggp1_ggNg}}}gNg2_@gg_(Jd]Igf]QNp PAGE 23 QUESTION B.07 (1.00)
Unit 1 is in Operational Condition 1, at 75 perr.ent of Rated Thermal
~
Power and. Tech Spec 3/4.5.1 ACTION step d.1 is being applied to unit operation. WHICH ONE of t.he below sets of circumstances is the cause for implementing this ACTION statemer.t? NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE
- a. LPCI pump " A" st arting Time Del ay (TD) relay is inoperative; two channels of HPCS low CST level logic are incapable of being tripped; ADS timer trip system "B" is inoperative.
- b. LPCI pump "A" motor has an overcurrent lockout condition; one i LPCI pump "B" ADS high pressure permissive is. inoperative; DG l 12 i s inoperative.
- c. ADS trip system "A" is inoperable; RCIC Trip & Throttle valve 3 has failed closed; LPCI pump "B" has an overcurrent lockout !
condi t f or.. -
- d. ADS timer trip system "A" is inoperative; LPCI pump "B" inot or has an overcurrent lockout condation; one channel of the HPCS CST low 1cvel logic has been tripped.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
" 9___8DdlNi@lB@llyE_PBgCEDUBE@t_CghpillgN@t_@yD_(ldll@llgN@ PAGE 24
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QUESTION 8.08 (1.00) The plant is at 60% power with only one outstanding LCO: l
- The LPCS pump is INOP due to an in-progress (one day) repair. I There is no estimate of repair time.
Ten minutes into the shift, DG 12 f ails to start twice during the performance of a scheduled surveillance and is declared INOP. There is no estimate of repair time. WHICH of the following ACTIONS most correctly detail the allowance and/or l i mi t ati ons imposed by the Technical Specifications in this instance? NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE.
- a. Be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- b. Power operation may continue for 12 hours; and then, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the f ollowing L4 hours.
- c. Power operation may continue for 72 hours; and then, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the f ollowing 24 hours.
- d. Power operation may continue for 6 days; and then, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the f ollowing 24 hours.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
' 9 __'8pMJNJg]BgIJyg_PB9CgpURgg1_CgNp))]QNpg_gNp_LJMJIgIJgNp PAGE 25
-QUESTION 8.09 (1.00)
Unit 1 is operating at 75% RATED THERMAL POWER. Channel Functional Tests are performed on all the Main Steam Line (MSL) Radiation Monitoring System channels. Channels A and D test UNSAT; Channels B and C test SAT. Maintenance has no estimate of repair time and will not be able to commence troubleshooting and repair for at least 16-20 l hours. WHICH of the f ollowing ACTIONS most correctly detail the all owances and/or limitations imposed by the Technical Specifications in this instance? NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE
- a. Be in at least HOT SHUTDOWN within 12 hours.
- b. Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- c. Place MSL Rad Monitor Channel "A" in the tripped condition within one hour -AND- be in at least HOT SHUTDOWN within 12 hours.
- d. Place MSL Rad Monitor Channel "A" in the tripped condition within one hnur -AND- be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOP'N within the next 24 hours.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
^9t__B90lN1@l@@llyE_PBQCEQUBE@t_CQNQlligN@t_@NQ_Lidll@llgN@ PAGE 26
~ QUESTION B.10 (3.00)
Indicate whether the f ollowing statements are TRUE or FALSE.
- a. STAS are required to be licensed Senior Operators.
- b. A Shift Supervisor may concurrently fill the position of the STA while on shift.
- c. All core alterations must be directly supervised by a licensed Senior Operator (or Senior Operator limited to fuel handling).
- d. An Operator license is required for an operator to perform a core alteration.
- o. During Operatiora1 Condition 4 or 5, an individual with a valid Operator license may be designated to assume the control room command function during the absence of the Shift Supervisor from the control room.
- f. The Fire Brigade must include at least one of the following:
SS, STA or COF. QUESTION 8.11 (2.00) FILL IN THE BLANKS Regarding the allowed working hours and overtime per the Grand Gulf TECHNICfL SPECIFICATIONS:
- a. An individual should not work more than ____________ hours straight, excluding shift turnover.
- b. There should be an least a(n) _____________ hour break between all wor k periods,
- c. An individual should not work more than _________ hours in any seven day period.
- d. Any deviation from the above guidelines shall be authorized by
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
'gz__BQMJN}p]ggIJyg_PggggpUggp1_gpNp]IlgNpz_gNp_LIMJIgIJgNg PAGE 27 a .
QUESTION 8.12 (3.00) FILL IN THE BLANKS regarding the Grand Gulf Safety Limits: The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed _____(a) _____. THERMAL POWER shall not exceed _____ (b ) _____ of RATED THERMAL POWER with the reactor vessel steam dome pressure l ess than _____ (c ) _____ or core flow l ess than _____ (d ) _____ of rated flow. The reactor vessel water level shall be above _____(e) _____. The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than _____ ( f) _____ during two loop operation with the reactor vessel steam dome pressure greater than _____ (c) _____ and core flow greater than _____( d) _____ of rated flow. QUESTION 8.13 (1.50) Administrative Procedure Ol-S-06-2, " Conduct of Operations" discusses the succession of Responsibility and Authority for Operation of GGNS. PLACE the following position titles in their proper sequence of authority from HIGHER to LOWER.
- a. Shift Superintendent
- b. Manager, Plant Maintenance
- c. Operations Superintendent
- d. GGNS General Manager
- e. Shift Supervisor
- f. Manager, Plant Operations QUESTION B.14 (3.00)
- a. Per 10 CFR 20, LIST thw three (3) Federal l i mi tc (whole body, ex tremi ti es, skin) f or QUARTERLY personnel exposure in restr.icted areas. (1.5)
- b. STATE the non-emergency CONDITIONS which must be met for a worker to exceed the limits given in (a) above. (1.5)
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l l 'at__'BDMINISIg@IlVE_PBgCEDUBES t_CgNDillgNgt_@Ng_ LIM 11@IlgN@ PAGE 28
< e l
l l -QUESTION 8.15 (1.00) l l The reactor is operating at 85% of Rated Thermal Power with the following plant conditions:
- two subsystems of Standby Service Water are OPERABLE,
- the LPCS pump has just been taken out of service for maintenance,
- Suppression Pool level is id' 3-1/2",
- average control room temperature for the last shift was 95 deg F,
- Drywell Airlock leakage rate is 1 scf/ hour & Pa = 11.5 psig,
- only one channel of the Turbine Control Valve - Fast Closure trip system A is OPERABLE.
For the conditions given above, STATE the applicable LCOs (i f any) requiring ACTION within ONE HOUR. QUESTION B.16 (2.00) The reactor is operating at 50% of RATED THERMAL POWER with one channel of Reactor Vessel Water Level - Low Low, Level 2 (non ECCS - Division 3) operable per trip system. LIST ALL SPECIFIC ACTIONS which is(are) required in this condition. (Include all ACTIONS required by any referenced graphs or tables) NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE QUESTION 8.17 (2.50) The plant is oeprating at 100% load when generator f requency suddenly drops to 50 Hz.
- a. EXPLAIN HOW operating in this condition for an extended period will cause problems with the Main Turbine. (1.0)
- 6. What SPECIFIC component of the Main Turbine is at risk in this situation? (0.5)
- c. What LIMIT is associated with this condition? (1.0)
(***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************)
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8 E 6
' INTEGRATED OPERATING INSTRUCTI;N J GRAND GCLF NCCLEAR STATION 03-1-01-1 l Revision: 35 lPage: 12 l
Title:
Cold Shutdown To Genetator INo.: l l Carrying Minimum Load I l l l 3.3.2 (Continued)
- d. HPCS System - In Standby per SOI 04-1-01 E22-1. /
- e. RCIC Systes - Ready for operation .-
or in Standby per SOI 04-1-01-E51-1. (Isolation valves cannot be opened until the low pressure isolation clears.) /
- f. Remote Shutdown System - In Standby per SOI 04-1-01-C61-Temp-1 (To be performed af ter B21, B33. E12, G33 P41, T46, Z77 lineups complete) /
- g. MSIV LCS - In Standby per SOI l 04-1-01-E32-1. /
- h. Feedwater LCS - In Standby per i SOI 04-1-01-E38-1. / .
- i. Process Sampling System - In operation per SOI 04-1-01-P33-1. / l 3.3.4 Containmen . Buildings and Systems
# a. Primary Containment integrity verified per Technical Specification 3.6.1. /
# b. Drywell integrity verified per Technical Specification 3.6.2. /
# c. Secondary Containment integrity verified per Tech Spec 3.6.6. /
- d. PRM System - In operation per 501 04-1-01-D17-1. /
- e. ARM System - In operation per SOI 04-1-01-D21-1. /
Figure #533
)
03-1-01-1 TEXT
i 3/4.0 APPLICABILITY LIMITING ~ CONDITION FOR OPERATION
- 3. 0.1 Compliance with the Limiting Conditions for Operation contained in the succeedi conditio,ng Specifications ns specified is required therein; exceptduring the OPERATIONAL that upon failure to meetCONDITIONS the Limiting or oth Conditions for Operation, the associated ACTION requirements shall be 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 reautrements, within one hour action shall be initiated te place the unit in an OPERATIONAL CONDITION in which the Specification does nt, apply by placing it, as applicable, in:
- 1. At least STARTUP within the next 6 hours,
- 2. At least HOT SHUTDOWN within the following 6 hours, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures arb completed that per?it operation under the ACTION requirements, the ACTION may be taken in accordcnce with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. . Exceptions to these requirements are stated in the individual Speci-fications. This specification is not applicable in OPERATIONAL CONDITION 4 or 5. 3.0.4 Entry into an OPERATIONAL CONDITION or other specificd condition shall not be made unless the conditiens for the Limiting Condition for Operation are met without reliance on provisions contained in the tr. TION requirements. This provision shall not prevent passage tnrough or to OffsATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications. GRAND GULF-UNIT 1 3/4 0-1
8 e g 3/4.3 INSTRUMENTATION l 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels
, shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME'as shown in Table 3.3.1-2.
APPLICABILITY: As shown in Table 3.3.1-1. ACTION:
- a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel and/or that trip system in the tripped condition" within one hour. The provisions of Specification 3.0.4 are not applicable.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS _ 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1. 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip I functional unit shown in Table 3.3.1-2 shall be demonstrated to be within is t limit at least once per 18 months. Each test shall include at least one cnan-nel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip syste.m. i
*An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall bs restored to OPERABLE status within 2 hours or the ACTION required by Tabie 3.3.1-1 for that Trip Function shall be taken.
**The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be p! aced in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip s _. system in the tripped condition.
GRAND GULF-UNIT 1 3/4 3-1
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O e6 INSTRUMENTATION TABLE 3.3.1-1 (Continued)
- . REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 -
Be in at least HOT SHUTDOWN within 12 hours. ACTION 2 - Verify all insertable control rods to be inserted in the core and Icek the reactor mode switch in the SNUTDOWN position within one hour. ACTION 3 - Suspend all operations involving CORE ALTERATION 5*, and insert all insertable control rods within one hour.
. ACTION 4 -
Be in at least STARTUP within 6 hours. ACTION 5 - Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least NOT SHUTDOWN within 12 hours. ACi10N 6 -
]nitiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours.
ACTION 7 - Verify all insertable control rods to be inserted within one hou-ACTION ! - Lock the reactor mode switch in the SHUT 00VN positior within ene hour. ACTION 9 - Suspend all coerations involving CORE ALTERATIONS *, and insert all insertable cont-c! rods and lock the reactor mode switch in the SHUTDOWN position within one hour.
"Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings pruvided SRM instrumentation is OPERABLE per Specifica* ion 3.9.2.
GRAND GULF-UNIT 1 3/4 3-4
l TABLE 3.3.1-1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION i TABLE NOTATIONS (4) A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped condi-tion provided at least one OPERA 8LE channel in the same trip system is monitoring that parameter. . (b) Tne " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- per Specification 3.9.2 .
and shutdown margin demonstrations performed per Specification 3.10.3. (c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel. (d) This function is not required to be OPERA 8LE when the reactor pressure f vessel head is unbolted or removed per Specification 3.30.1. (e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position. (f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required. (g) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. ' (h) This function shall be automatically bypassed when operating below the appropriate turbine first stage pressure setpoint of: (1) 126.9%** of the value of turbine first-stage pressure at valves wide open (WO) steam flow when operating with rated feedwater temperature of greater than er equal to 420*F, or l
)
(2) < 22.5%** of the value of turbine first-stage pressure at WO steam Ilow when operating with rated feedwater temperature between 370*F and 420*F.
"Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
** Allowable setpoint values of turbine first-stage pressure equivalent to THERMAL PQWER less than 40% of RATED THERMAL POWER. I 1
GP.AND GULF-UNIT 1 3/4 3-5 Amendment No. I6 l
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I INSTRUMENTATION l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION i
'IMITING
. CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERA 8LE with their trip setpoints set consistoit with the values shown in the Trip setpoint coliaan of Table 3.3.2-2 and with ISOLt. TION SYSTEM RESPONSE TIME as shown in Table 3.3.24.
APPLICA$ILITY: As shown ir. Table 3.3.2-La ACTION: {
- a. With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERA 8LE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels less than required by the Ministni 0PERA8LE Channels per Trip System requirement for one trip system, piece the inoperable channel (s) and/or that trip system in the tripped condition" within one hour. The provisions of Specification 3.0.4 are not applicable.
- c. With the number of OPERABLE channels less than required cy the Minime OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Tacle 3.3.2-1.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL 1EST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the f requencies shown in Table 4.3.2.1-1.
- 4.3.2.2 LOGIC SYSTEM FUNCTIO *iAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of ecch isolatinn trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system such that all channels are tested at least once overy N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
'An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these casas, the inoperable chtnnel shall be restored to CPERABLE status within 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
**The trip systes need not be placed in the tripped condition if this weald cause the Trip Function to occur. When a trip systes can be placed in the tripped condition without csusing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip systee in the tripped condition.
GRAND GULF-UNIT 1 3/4 3-9 Amendment No 22 l
INSTRUMENTATION l 1 i f! solation actuttien instrumentation is not required to be OPERABLE during control rod removal, reinstallation and movement within defueled core cells j for the period from October 3,1986, through October 10, 1986. GRAND GULF-UNIT 1 3/4 3-Sa Amerdnent No. 22 _ _ _ _ _ _ _ _ _ . ___________w
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i INSTRUMENTATION pgggg 3,3,p.1 (Continued) fQ( [ h h $ ' ISOLATION ACTUATION INSTRUMENTATION E g ppg g y - ACTION ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUT 00WN within the next 24 hours. ACTION 21 - Close the affee.ted system isolation valve (s) within one hour or: a., , In OPERATIONAL CONDITION 1. 2, or 3,.be in at least HOT SHUTDOWN within the next 12 hours and in COLO SMUT 00VN within the following 24 hours. , b. In OPERATIONAL CONDITION *, suspend CORE ALTERATIONS, l handling of irradiated fuel in the primary containment and operations with a potential for draining the reactor vessel. ACTION 22 - Restort the manual initiation function to OPERABLF. status within 48 hour's of tse in at least HOT SHUTDOWN within the next 12 hours and in COLD SHU700WN within the following 24 hours. ACTION 23 - 8e in at least STARTUD with the associated isolation valwes closed within 6 hours or be in at lear: HOT SHUT 00WN within 12 hcurs and in COL SHUTDOWN within the next 24 hours. ACTION 24 - Be in at least STARTUP within 6 houes. ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with th6 standby gas treatment system operating within one hour. ACTION 26 - Restore the manual initiation function to OPERABLE status within 8 hours or close the affected system isolation valves within the next hour and declare the affected system inoperable. ACTION 27 - Close the affected system isolation valves within one hour and declare the affected system inoperable. - ACTION 28 - Within one hour lock the affected system isolation valves closed, or verify, by remote indication, that the valve is closed and electrically disarmed, or isolate the penetration (s) and declare the af facted system inoperable. ACTION 29 - Close the affeci.ed system isolation valves within one hour and declare the affected system or component inoperable or:
- a. In OPERATIONAL CONDITION 1, 2 or 3 be in at least HOT SHUT C.N within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- b. In OPERATIONAL CONDITION # suspend CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel.
ACTION 30 - Declare the affected SLCS pump inoperable. NOTES When handlir.g irradiated fuel in Lt.e primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The low condenser vacuum MS!V closure may be manually bypassed during reac w ) SHUTDOWN or for reactor STARTUP when condenser vacuum is below the trip set-i 1 point to allow opening of the MSIVs. The manual bypass sna11 be removed .%, condenser vacuum exceeds the trip setpoint. During CORE ALTERATIONS and operations with a potential for draining tne reactof vessel.
## With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(a) See Specification 3.6.4, Table 3.6. c l for valves in each valve group. GRAND GULF-'JNIT 1 3/4 3-3 f
-9
< i INSTRUMENTATf0N TABLE 3.3.2-1 (Continued)
-ISOLATION ACTUATT5R INSTRUMENTATION tt0TES (Continued)
(b) A channel may be placed in an inoperable status for~ up to 2 hours for required surveillance without placing the trip :.ystem in the tripped con-dition provided at least one other OPERABLE channel in the same trip system ! is monitoring that parameter. (c) Also actuates the standby gas treatment system. (d) Also actuates the control room emergency filtration system in the isolation mode of operation. (e) Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system and initiate isolation of the associated containment and drywell isolation valves. (f) Also trips and isolates the mechanical vacuem pumps. l (g) Celeted. (h) Also actuates secondary containment ventilation isolation dahipers and valves per Table 3.6.6.2-1. (f) Closes only RWCU systee isolation valves G33-F001, G33-F004, and G33-F351. (j) Actuates the Standby Gas Treatment System and isolates Auxiliary Building = penetration of the ventilation systems within the Auxiliary Building. (k) Closes only RCIC outboard valves. A concurrent RCIC initiation signal is required for isolation to occer. (1) Valves E12-F037A and E12-F0378 are closed by high drywell pressure. All other Group 3 valves are closed by high reactor pressure. j (m) Valve Group 9 requires concurrent drywell high pressure and RCIC Steam - l Supply Pressure-Loa signals to isolate.
. (n) Valves E12-F042A and E12-F042B are closed by Containment. Spray System initiation signals.
(o) Also isolates valves E61-F009, E61-F010, E61-F056, and E61-F057 from Valve' ' Group 7. i I GRAND GULF-UNIT 1 3/4 3-15 p
INSTRUMENTATION 3 /4. 3. 3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setooints ! set consistent with the values shown in the Trip Setpoint column of 'aole ' 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3. nor.lCAEILIT5 As shown in Table 3. 3. 3-1. ACTION:
- d. With an ECC5 actuation instrumentation channel trir setpoint less conservative tf.an the value shown in the Allowable .alues column of Table 3.3.3-2 declare the channel inoperab a until tne channel is restored to OPEF'E status with its trip setpoint adjusted consistent with Trip (*tpoint value. '
O. With one or more . . actuation instr'umentation channels incperable, take the ACTION required by Table 3.3.3-1.
- c. With either ADS trip system "A" or "B" inoperable, t estore the inoperable trip system to OPEkABLE status within:
- 1. 7 days, provided that the HPC5 and RCIC systems are OPERABLE.
- 2. 72 nours.
Otherwise, be in at least HOT SHUiDOWN within the.next 12 hours and reduce reactor steam dome pressure to less than or equal to 135 psig within th( following 24 hours. SURVEILLANCE REQUIREMENTS 4.3.3.1 Eacn ECCS actuatiora instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNE. CHECK CHANNEL FUNCTI0t.AL TEST and CHANNEL CALIBR;~ ION operations for the OPERATIONAL CONDITIONS anc at the f recuenc'es sn. n in Table 4.3.3.1-1. 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.3.3 The ECCS RESPONSE TIME of each ECC5 trip functie. snown in Table 1 3.3.3-3 shall be demonstrated to be within the limit at least once per i 18 months. Each test shall include at least one channel per trip system
- such that all channels are tested at least once every N times 18 months l where N is the total number of redundant channels in a specific ECCS trip system.
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EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - With the number of OPERABLE channels less than recu d i by the Minimum OPERA 8LE Channels per Trip Function requirec.. :
- a. With one channel inoperable, place the inoperatie channel in the tripped condition within one hour" or eeclare the associated system (s) inoperable.
- b. With more than one channel inoperable, declare the associated sys'.em(s) inoperable.
ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per i ip Function requirement, ceclare the associated ADS tr . system or ECCS inoperable. ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to CPERABLE status within 6 hours or ceclare the associated ADS trir system or ECCS inoperable. ACTION 33 - With the number of OPERABLE channels less than reevired by the Minimum OPERABLE Channels per Trip Function requirement, p' ace the inoperable channel (s) in the tripped condition within one hour
- cr declare the HPCS systerr inoperable.
ACT10h 34 - With the number of OPERABLE channals less than required by the Minimum 0FERABLE Channels per Trip Function requirement, place at least one i'sperable channel in the tripped condition within one hour" or ceclare the NPCS system incoerable. ACT!'JN 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement. place the inoperable channel (s) in the tripped condition within one hour" or declare the associated system (s) inoperable.
"Tne provisions of Specification 3.0.4 are nc applicable.
1 GRAND GULF-UNIT 1 3/4 3-30 r andment he. 20 l Enective Date: gti 1 3 1566
^
',, 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION for Tech Spec 35.1 3.5.1 ECCS divisions 1, 2 and 3 shall be OPERABLE with:
See TSPS y OBA -
- a. ECCS division 1 consisting of:
- 1. The OPERA 8LE 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.
- 2. The OPERA 8'E low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.
- 3. Eight OPERABLE A05 valves.
- b. ECCS division 2 consisting of:
- 1. The OPERA 8LE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.
l For Tech Spec 35.I. 2 . c si,ht opea,,Lt Aos ,ai,e,. See TSPS f _0'79- TCCS division 3 consisting of the OPERA 8LE high pressure core spray (HPC5) 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. AP/ BILITY: OPERATIONAL CONDITION 1, 2a # and 3* I bE
- 4. For ECCS division 1, provided that ECCS divisions 2 and 3 are OPEUB.i
- 1. With the LPCS system inoperable, restore the inoperable LPC5 system to OPERABLE status within 7 days.
- 2. With LPCI subsystem "A" inoperable, restore the inoperaDie ;;.
subsystem "A" to OPERABLE status witnin 7 days.
- 3. With the LPCS system inoperable and LPCI subsystem "A" t. 9 :cer n -
restore at least the inoperable LPCI subsystem "A" or tre inoperable LPCS system to OPERABLE status within 72 hours. 4 Otherwise, be in at least HOT SHUT 00WN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.~ *
*The A05 is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.
#5ee Specjal Test Exception 3.10.5.
**Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. .
GRAND GULF-UNIT 1 3/4 5-1 i
.i i
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'S 8 EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued).
- b. For ECCS division 2, provided that ECCS divisions 1 and 3 are OPERABLE:
- 1. With either LPCI subsysten "B" or "C" inoperable, restore the inoperable LPCI subsysten "B" or "C" to OPERABLE status within 7 days.
- 2. With both LPCI subsystems "B" and "C" inoperable, restore at least the inoperable LPCI subsysten "B" or "C" to OPERA 8LE status within 72 hours.
- 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours".
- c. For ECCS division 3, provided that ECCS divisions 1 and 2 and the RCIC system are OPERABLE:
- 1. With ECCS division 3 inoperable, restore the inoperable division to OPERABLE status within 14 days.
- 2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.
- d. For ECCS divisions I and 2, provided that ECCS division 3 is I OPERABLE:
1, With LPCI subsystem "A" and either LPCI subsystem "B" or "C" inoperable, restore at least the inoperable LPCI subsystem " A" or the inoperable LPCI subsystem "B" or "C" to OPERABLE states within 72 hours.
- 2. With the LPCS system inoperable and either LDCI subsystems 1F or "C" inoperable, restore at least the incoerable ' PC5 sy stem or the inoperable LPCI subsystem "B" or 'C" to CPERABLE 5 tat.s within 72 hours.
- 3. Otherwise, be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours' "Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTOOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
GRAND GULF-UNIT 1 3/4 5-2 1 L - .. .
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ENERGENCY CORE C0OLING SYSTENS LINITING CONDITION FOR OPERATION (Continced) ACTION: (Continued) e. For ECCS divisions 1 and 2, provided that ECCS division 3 is i OPERA 8LE and divisions 1 and 2 are otherwise OPERABLE: 1. With one of the above required ADS valves inoperable, restore i the inoperable ADS valve to 0PERA8LE status within 14 days or i be in at least NDT SNUTDOWN within the next 12 hours and reduce reactor steam dose pressure to 5,135 psig within the next 24 hours. ; 2. With two or more of the above required ADS valves inoperable, be in at least NOT SHUTDOWN within 12 hours and reduce reactor steam done pressure to 1135 psig within the next 24 hours. f. With an ECC5 discharge line " keep filled" pressure alan instrumentation channel inoperable, perfom Survef11ance Requirement 4.5.1.a.1 at least once per 24 hours. g. With an ECCS header delta P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status with 72 hours or determine ECCS header delta P locally at least once per 12 hours; otherwise declare the associated ECCS inoperable.
- h. In the event an ECCS systee is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the useage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
- 1. With an ADS accumulator low pressure alars system instrumentation channel (s) inoperable, detemine the associated ADS accumulator pressure locally at least once per 12 hours; restore the inoperable channel (s) to OPERABLE status within 7 days; otherwise declare the associated ADS valves inoperable.
1 "Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. GRAND GULF-UNIT 1 3/4 5-3 Amendment No. 21 l Effective Date: OCT 2 0 lHi
~ .
3/4.8 ELEC7RICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. SOURCES - CPERATING LI_NITING COM0! TION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERA 8LE:
- a. Two physically independent circuits between the offsite transmission network and the ensite Class IE di6tribution systes, and
- b. Three separate and independent diesel generators, each with:
- 1. Separate day fuel tanks containing a sinimum of 220 gallons of fuel.
- 2. A separate fuel storage systes containing a sinicus of:
a) 57,200 gallons of fuel each for diesel generators 11 and l 12, and b) 39,000 gallons of fuel for diesel generator 13.
- 3. A separate fuel transfer pump.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With either one offsite circuit or diesel generator 11 or 12 of t5e above required A.C. electrical power sources inoperable, demonstrate the OPERA 81LITY cf the remaining A.C. sources by performing Surveil-Iance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4,"
for one diesel generator at a time, within two hours and at least once per 8 hours thereaf ter; restore at least two offsite circuits and diesel generators 11 and 12 to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- b. With one offsite circuit and diesel generator 11 or 12 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4,* for one diesel generator at a time, within two hours and at least once per 8 hours thereaf ter; restore at least one of the inoperable A.C.
sources to CPERA8LE status within 12 hours or be in at least HOT SHU DOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Restore at least two offsite circuits and diesel genera-tors 11 and 12 to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
" Specification 4.8.1.1.2.a.4 sust be performed for diesel generator 13 only when the HPCS systes is OPERA 8LE.
Amendment No. 5 l GRAND GULF-UNIT 1 3/4 3-1
. i ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
For Tech Spec 3.8.l.l c. ACTION (Continued) l See TSPS f OSO
- c. With either diesel generator 11 or 12 of the above required A.C.
electrical power sources inoperable, in addition to ACTION a or b above, as appitcable, verify within 2 hours that all required systems, subsystaas, trains, coeponents and devices that depenil en the remaining diesel generater 11 or 12 as a source of emergency power are aise OPERA 8LE; otherwise, be in at leact NOT SHUT 00W within the next 12 hours and in COLD SHUTDOW within the following 24 hours,
- d. With two of the above required offsite circuits inoperable, demonstrate the OPERA 8ILITY of three diesel generators by performing Survet11ance Requirement 4.8.1.1.2.a.4*, for one diesel generator at a time, within two hours and at least once per 8 hours thereaf ter, unless the diesel generators are alreadf operating; restore at least one of the inoperable offsite circuits to 0PERABLE status within 24 hours or be in at 16ast NOT SHUTDOWN within the next 12 hours.
With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to CPERA8LE status within 72 hours from time of initial loss or be in at least HDT SHUT 00W within the next 12 hours'and in COLD SHUTDOWN within the following 24 hours.
- e. With diesel generators 11 and 12 of the above required A.C.
electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.48, within two hours and at least once per 8 hours thereafter; restore at least one of the in-operable ufesel generators 11 and 12 to OPERA 8LE status within 2 hours or be in at least NOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Restore both diesel genera-tors 11 and 12 to 0PERA8LE status within 72 hours froe time of initial loss or be in at least NOT SHUTDOWN within the next 12 hours and in C010 SHUTDOWN within the following 24 hours.
- f. With dia:e1 generator 13 of the above required A.C. electrical power sources inoperable, demonstrate the OPERA 81LITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4, for one diesel generator at a ties, within two hours and at least once per 8 hours thereafter; restore the inoperable diesel generator 13 to OPERA 8LE status within 72 hours or declare the NPCS system inoperable and take the ACTION required by Specification 3.5.1.
" Specification 4.8.1.1.2.a.4 sust be performed for diesel generator 13 only when the HPCS system is OPERA 8LE.
GRAND GULF-UNIT 1 3/4 8-2
p.
. GRAND OULT MJCLEAR STATION PLANT ADMIN 157RAI1YE PMUCt.UUNE 01-5-15-2 Rev. 1 Attachment II Page 1 of 1 f
TECHNICAL SPECIFICATION POSITION STATEMENT Statement No [ $ 0) Part 1 Originator: GBu Date : I ~2. - 1 " K7 Technical Specification ~a +. 3.1. t -i ,en. Q ha +. 3. 2. . I - i
,i ch c c. ) > T.M u. +. 3. 3. i - 4
References:
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- Manager Approval Date l
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+0ne of the following is required: Plant Manager Assistant Manager, Operations O
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%. - QRAND GJLF MJCLEAR STAT 20N PLANV ADMINISTRAl1VL WUt;tDUMt.
x .,
- 01-5-15-2 i Rev. 1 I
Attachment II l Pace 1 of f i t TECHNICAL SPECIFICATION POSITION STATEMENT StatementNorfqpj.62//[ Part 1 Originator: Jerry L. Parker Date : 3-13-84 Technical Specification 4.8.1.1.3
References:
Regulatory Guide 1.108 Position: T.S. section 4.8.1.1.3 requires that all diesel generator f ailures be reported to the commission. It is our Position that the meaning of " diesel generator failures" in section 4.8.1.1.3 is the same as the definition given in Regulatory Guide 1.108 position B and expanded upon in position C.2.e. See attached Technical Justification. 1 Position Af fect/ponsnents: This position clarifies the deportability of f ailures' or events which are related to one diesel generators but do not meet the re-l This is needed because Tect.ni a; ' porting criteria of Regulatory Guide 1.105. Specifications do not define wnat constitutes a diesel generator failure vn le Regulatory Guide 1.108 does. +
~
Part 2 Licensing Engineer M/ / J/N/# Date Signature 7
' art 3 Review and Approvals Approve Disaccreve
'(1) ___ _ _ r bl a / /rh Date (X) ' ;
Tbennicai,'5ucyr int encent (2) . 2-/08b ([) i a> 01 f%?' J~
" r SRC/ Meeting Date 3 - n ,rv pA
- l. (4) +v V-5-61 ( ^^) (
' Manager Approval Date
*0ne of the following is recuired: Plant Manager Assistant Manager, Operations
A
; P, a q T6 4 g Technical Specification Position Statement on the Defini'. ion of
" Diesel Generator Failure" Technical Specification section 4.8.1.1.3 requires that "all diesel generato'r failures, valid or non-valid" be reported to the Commission.
It then references Regulatory Guide 1.108 and further states that the report shall include the information recommended in Regulatory Guide 1.108. Technical Specification section 4.8.1.1.2 and Table 4.8.1.1.2-1 show how the diesel surveillance frequency shall be determined by the number of f ailures in the last 100 valid test, identically to Regulatory. Guide 1.108. The table also states that the criteria for determining the number of failures shall be in accordance with position C.2.e of Regulatory Guide 1.108. Due to the f act that the Technical Specification sections closely parallel, ref erence, and have the same requirements as Regulatory Guide 1.108, it is our position that the meaning of " diesel generator f ailures" in section 4.8.1.1.3 is the same'as the definition'of diesel generator failure in Regulatory Guide 1.108 position B and expounded upon in position C.2.e.
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Part 2 Compliance Superintendent
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1 Part 3 Review and Ac reval: Accreve Disaccreve 1 (1) .5/ G /8 i (X) ( Tecnn'rdal Superigent / Date (:) bf r - ee4 //AP/8f (A) ' (3) L . F5a;/ Me ing a hff ' l {m (
/ Date (4) v 2/ (M ( )
f vanager Appreva1 / CTate l
*0ne of the following is recuired: CONS Ceneral Manager Manager, P1 ant :cerati:ns 9
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TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION ITDI INPLANT INSTRUMENT 1.g D17RITSK609A,B,C D 3.c D17RITSK617A,B.C.D 3.d D17RITSK618A,B,C.D l i h
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. GRAND GUIJ NUCIX.AR STATION . ADHDEISTRATIVE FloCEDURE
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, 01-S-15-2 Rev. 4 Attachment III Pese 1 of 1 l
TECHNICAL $ SPECIFICATION POSITION STATEMENT Part 1- Originator: A VA f vised 6//%/87 i Name / Date l l Technical Specification 3.f. / neferences: _ lifk 1 .. Position L- C. O 3.3.1.C Trowans 7 war HA*S se "capas:a er T 4 64ee,rC, A $GC 710M Fdees ng $effAfsyou AWL * * ' ' . pss j$ psiff4*/ % ~72rp 7D J)fMAJ Ff/J' A//t'$ A9os)* Be CA/A44&' 0F AV7 Din 47/CAlly A .blA 77eA,/ FJg.et /265~ fpf,//r.nsso u )@d& . - . ' T A fs4 <o . iI Poeition Af tecElCaamenta: TP 7pr ja' c7MU #9e NAS #4s Aursersncxs e
~7/445 F / Pfe**O 73 7hr J~wMi%av Akt 74er 50c n ou _5Neos.b Bf 44 cowcD in Pme.~ Antif/ob n 7Ws' Sv A.3iUSsw AWL lik/14 7MG" tt+Ds715. LuNsFM M SEs) yWF ks%AI7 C 7P & tai *1t' #43
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MD RE ALIGNd1) Escr ?D fru f.5*7' M2ff 7k/ Cl*Jn?> s RC 12-R**'A7AC 7X'#4/S F f t >V45 2 M'y [ k'A'A'i tV E*~D 7nr N/d'.$ M W 2r 2%enCLodd2) /A/Of 247~4 7MT $NM/2=/ /5 MEdC./% Part 2 Compliance Superintendent / / T'.5
" ' S4 nature # / Date Part 3 Review and Approval: Approve Disapprov e (1) (8 "
( ) TecB6ic u dent / Date (2) 2,5 ' VWsf vr ( ) S / ate
/ . . . a) l$14 A0) _l l Ol'1k5 '/bW (d ( )
(4) - *-/[ '" (M ~~ ( ) ~
, Manager, Approv al / 1459
*0ne of the follotring in required: CGNS General Manager Manager, Plant Operatione
." ." GRAND GUI.F NUCLEAR STATION ADMINISTRATIVE PROCEDURE 01-S-15-2 m.v. 6 #
Attachment 111 Page 1 of J3i
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TECHNICAL SPECIFICATION POSITION STATEMENT Part 1 Originator: but d s.41/ /0/3/87 Nhme / Date , Technical Specification 3.6.f.I, denes SN M e rC , noterencee: 7 Fed stat 843/s uc- s/+. r s , rMR se c. r, tr. i. 3 l 4tC 6tuctic um usa mtco sidov , try G..uf /.fs,3c 1999 Position ,$EE .4WMMD List of Aw 'Ren waro swsrrms S A sys;>w yens ~ s , awsv wa rs a o w,ess stor smo k, rus rema.+,+c sa.-we teksen W u o g i > es a so.ece en .am u r e n s. M ae r. 7,rsors w nern Ic AR.g d Ac so c%st yyy pfmnsv&6 b es9x d gr cupelyg yy ,,gy ,,,<sep, 40c kor w siY b arcause Ak-neu surswivr e c.vvos.b rveu .wh esr- _ A'enn s rww ar e. ' Poeieion Atiect/Consenta: The lis r ut n. /E60/Df & tssrou c, s.v ryp-Abf/ttstrev t>" rNa PEcusato Acrot.r n o w et der f.s .4 cAzex Gar To '2ebuce- 7?*r 22ptswer m .vec cuess .sver M pa .Misseo ira'my
/) M Part 2 Compliance Superintendent ~p / /8 C/' %
' Date '
Signatp j 1 l Part 3 Review and prev al.e Approve Di s appr ov e l
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( ) g Te ical Supbr4p_tJndent / Date f (2) h I. LLM to/1/ff (X) ( ) S LO ,
/ Date (3) 1 Nd[ ( )
[I / Da te 5 et' (4 ) ~
/- } } =']G ~hb ( ( )
,, *Mandter, Approval / Date
*0ne of the following is required: CCNS Ceneral Manager Manager , Plant Operations i
l l
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* . . TSPS No.: 080 Page 2 of 3 When Div I is declared INOP the following items shall be verified operable per LCO 3.8.1.1, ACTION c.
- 1. LPCI mode of RHR-B System
- 2. LPCI mode of RHR-C System
- 3. C*lHT Spray mode of RHR-B System 4 Suppression Pool Cocling mode of RHR-B System
- 5. Shutdown Cooling mode of RHR-B System (required only in Mode 3)
- 6. Suppression Pool Makeup System, Div II
- 7. CTMT and Dryvell Isolation Valves as listed in Tech. Spec. Table 3.6.4-1 section 1 Div II
- 8. Post LOCA Vacuum Breaker, Div II
- 9. Secondary CTMT Isolation valves as listed in Tech. Spee.
Table 3.6.6.2-1, Div II i
- 10. Standby Gas Treatment B System I
- 11. Hydrogen Recombiner fystem, Div II
- 12. CTMT and Dryvell Hydrogen Ignition System, Div II
- 13. Control Room Emergency Filtratirn System. Div II 14 MSIV Leakage Control System, Div II
- 15. FW Leakage Control System, Div II
- 16. RHR-B Room HVAC
- 17. RHR-C Roon HVAC
- 18. EST SWGR Room HVAC, Div II
- 19. Standby Liquid Control System, Div II
- 20. Dryvell Purge Compressor, Div II
- 21. Battery Charger, Div II (either IB4 or 1B5) i l
l
- - - - -- ----- --- --- A
, TSPS E0.: 080' l-Pcg2 3 of 3 i
When Div II D/G is declared INOP the following items shall be verified OPERA 3LE per LCO 3.8.1.1. ACTION c.
- 1. LPCS Systes
- 2. LPCI mode of RER-A System l 3. CTMT Sprav mode of RHR-A System 4 Suppression Pool Cooling mode of RER-A Systea
)
- 5. Shutdown Cooling modo of RHR-A System (required only in Mode 3) I
- 6. Suppression Pool Makeup System. Div I 1
- 7. CTMT and Drywell Isolation Valves as listed in Tech. Spec.
Table 3.6.4-1, section 1. Div I
- 8. Post LOCA Vacuus Breaker, Div I
- 9. Secondary CTMT Isolation Valves as listed in Tech. Spec. Table 3.6.6.2-1. Div I
- 10. Standby Cas Treatment - A System
- 11. Hydrogen Recombiner System, Div I
- 12. CTMT and Drywell Hydrogen Ignition System, Div I
- 13. Control Room Emergency Filtration System, Div I 14 MSIV Leakage Control System. Div I
- 15. FW Leakage Control System, Div I
- 16. LPCS Roos EVAC
- 17. RER-A Room HVAC
- 18. EST SWCR Room KVAC, Div I
- 19. Standby Liquid Control System. Div I
- 20. Dryvell Purge Comprestor, Div I
- 21. Battery Charger Div I (either IA4 or IAS)
. , j GRAND GU1J .WCI.IiJ. STATION ADMINISTRATIVE PRCC~;;.72 l:01-5 13-2 ' Revision 3l l At tac.went III !?are 1 er 1 i 1
TICXNICAL SPECIFICATION POSITION STATEMENT j i TSPS No.: 082 Part 1 Originator: Moulder 1/15/86 Name / Date Technical Specification 3.5.1 and 3.5.2 Referene.es: IE 85-94 _ Technical Specification Change Required (30 NO YES ( ) If Yes CR0 Pesition If the min flow valve on any ECCS system vill not perform its intended function then the associated ECCS function is inop. l Position Affect /Conusents: Example: If E12-T064A vill not function automatically then the LPCI and CTMI spray mode of RHR ' A' will be declared inop. However, S. pool cooling and SDC will not be declared ineo if the min flow valve can be closed and adequate flow est. in these modes for Dunre protection. The sin fitv valve will be deacti-vared einmed and the eume breaker racked ouc until the min flow valve
<= renstred uniene ninne are made to eince the system in SP cooling
- m. c:nt- and specific attention is placAd on systent performance until adequtte flow is established (lZ, Part 2 Coop 11ance Superintendent p; . f L., / / [/ 5$,, ?[
Sigiaturf ' Date Part 3 Review and Approval: Approve Disapprove (1) _
/ /t 3 [8 (., X) ( )
Techalcal knig, dent / Date l a, w w n; (g) ( , I ( f' l l o>/ 9lbloMtlllLAhan/ RC/ et 'r //l' S /' Date Dateist-iksk_ p ( ) (4) . . ,
# '.b' /-/ 7 -8 [ {><Q ( )
" *Managir Approval / Date
*0ce of the following is required: GGNS General Macager Manager, Plant Operstions f
01-S-15 2 ATT III
3 m 3 GUII NUCLEAR STATIOM ADMINISTRATIVE PROCEDURE 1 l 01-S-15-2 (Revision 7 l
- l. Attachment III lPage 1 of 3 l 1 -
I \ TECHNICAL SPECIFICATION POSITION STATEMENT l TSPS No.: O97 l Part 1 Originato'r: A OAu /A TGW /I/% ; Name / Date
. Technical Specification AS//S7f.) of1 0 W 3 Ofl
References:
d A A 2 00 / Technical Specification Change Required (,4 NO YES ( ) If Yes CR# A//d Position C f e AA fr A 2-Position Affact/Cg' aments: 7"I'rf A O $ s PtCN G 7M7f MLA/T ((AttletL*$ TA G N rL M O h4A A tonAb TEA 7" bA VW Y$$ 9 0 ($N?fA $ s-c DA n w Th f A A
- l re d. t/C S S4 L ', +-h i $~ fos e riow ln Tek an/ T C u ff A C e D A _C 9 D C 17 tea,/ fin rfhen~T TsPS-c8V.
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/ l /)At Part 2 Compliance Superintendent
/ /I Id Sign Wure Date Part 3 Revi Appro 1: Approve Disapprove (1) .
TectNical Supe (rintendent
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/ Date k) ( ) -
(2) F M ll / Ulk ( K) ( )
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(3) -
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PSRC/. , nting f 206/9G / Date (4) /'!1 (.) ( )
' Mahager
- Appr dal / Date'
'One of the follcwing is required: 3GNS Ceneral Manager Manager, Plant Operations 01-5-15-2 ATI III
I
\
l TECHNICAL SPECIFICATION POSITION STATEMENT STATEMENT NO. 097 ] 1 PAGE 2 of 3 I Any of the following, associated with a reactor vessel pressure boundary penetration greater than 1 inch in diameter located below the top of active i fuel, is considered to be an operation with the potential for draining the ' reactor vessel: (1) Any operation (S'hutdown Cooling, RWCU, etc.) not protected by at least one operable automatic isolation system (Note 1) or (2) Any operation (CRD removal, vessel draining through the bottom head drain, ! recirculation pump changeout, etc.) that is not isolated from the RPV by at ' least one closed manual valve, valve backseat, blind flange, or de-activated automatic (non-manual) valve secured in the closed position. Other temporary plugs (freeze seals, plumbers plugs, inflatable bladders, etc.) are not adequate to meet this requirement. 4 (NOTE 1) If electrical power is required to accomplish isolation, one of the following is required:
.. A functional on-site emergency power system. a l
- 2. Two physically independent offsite power sources.
- 3. Operator (in the vicinity of the valve) dedicated to manually close the valve, or if power is avaliable per 1 or 2, the the operator can be stationed in the vicinity of the control switch.
TSPS No. 097 Page 3 of 3 Tech Spec Tech Spec Page Section Subject 3/4 3-14 3.3.2-1, Action 21 Isolation Actuation Instrumentation. Action 29. Footnotes Primary / Secondary Ctat Isolations
- and i 3/4 3-26 4.3.2.1-1. Footnotes Isolation Actuation Instrumentation
- and i 3/4 5-6 3.5.2 Action a and b ECCS - Shutdown 3/4 5-8 3.5.3 Action b Suppression Pool Level 3/4 6-48 3.6.6.1. Action b Secondary Ctat Integrity and Footnote
- 3/4 6-49 3.6.6.2 Footnote
- Secondary Ctat Isolation Valves 3/4 6-55 3.6.6.3 Footnote
- Standby Cas Treatment System and Action a.2.
3/4 7-5 3.7.2 Action b.2 Control Room Filtration System 3/4 8-9 3.8.1.2 Action a AC Sources (D/G's) - Shutdown 3/4 8-14 3.8.2.2 Action a DC Sources - Shutdown 3/4 8-18 3.8.3.2 Action a.1 Electrical Distribution - Shutdown and b.1 PCOM TECH SPEC POS STATE - 2
'5 . -_IHEQ6Y_gE_NUC6E@6_PQWE6_PL@NI_QPE6@Ilgyt_6LylpS t_@NQ PAGE 30
- IME6dgQYN@dlCS ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C.
I l ANSWER 5.04 (2.00)
- a. NPSH, the difference between the total pressure at the eye of the pump and the saturation pressure of the liquid. (1.0)
- OR-NPSH = P actual -
P sat.
- b. Decrease (or Increase) (0.5)
- c. Increases (0.5)
REFERENCE GGNS HT&FF, Chapter 6, L.O. #10 293OO6K103 293OO6K110 ...(KA'S) ANSWER 5.05 (1.00) 350 psig + 14.7 psia = 364.7 psia (0.5) Saturation temperature for 364.7 psia: (444.6 degrees F - 431.73 degrees F) (14.7/50) + 431.73 degrees F = 435.5 degrees F (0.5) [ Answer should be within + 2 degrees FJ REFERENCE Steam Tables and GGNS HTLFF, Chapter 3, L.O. #'s 1 and 3 293OO3K123 ...(KA'S) ANSWER 5.06 (1.00)
- a. Conduction (0.25)
- b. Conduction (0.25)
- c. Conduc ti on (0.15) and convection (0.1)
- d. Radiation (0.25)
REFERENCE GGNS OP-HF-507/Rev. 1, L.O. #1 GGNS OP-HF-508/Rev. 1, L.D. #7 293OO7K101 293OO8K112 ...(KA'S)
m - q b 29-
'5 . THEORY OF NUCLEAR POWER PLANT OPERATION 1 FLUIDS1_AND PAGE l 4
IOEBD9PXN9d]CS ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 5.01 (1.00) REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 2, LOM 1.2, 2.1 292OO3K110 292OO3K111 ...(KA'S) ANSWER 5.02 (1.25)
- a. Decrease
- b. Decrease
- c. Decrease
- d. Increase
- e. Decrease (5 0 0.25 each)
REFERENCE GGNS OP-B33-1-03., L.O. # 3.g.6 202OO1K301 202OO1K303 202OO1K601 ...(KA'S) ANSWER 5.03 (2.00)
- a. True
- b. False
- c. True
- d. False (4 & O.5 each)
REFERENCE GGNS Heat Transfer and Fluid Flow, Chapter 9, L.D. # 8 3 293OO9K136 ...(KA*S) )
'7 1
1
'.5 ' THEORY OF-NUCLEAR FOWER PLANT OPERATION 1 _FLUIpSy_ANp PAGE 31
'- IMEBDgDyN9dJ.CS 1.
L -87/05/18-PAYNE, ANSWERS -- GRAND GULF 1 D.C.
.c
' ANSWER 5.07 (1.25)
- a. 9'
- b. 3
- c. 2
- d. 10
- e. 5 (5 D O.25 pts each)
REFERENCE GGNS'OP-HF-509/Rev. 1, L.O. #'s 1, 3, 4, &5
'293OO9K101 293OO9K102 293OO9K106 293OO9K111 293OO9K117 g ...(KA'S)
ANSWER 5.08 (2.00) l a. Trip of both Recirc Pumps (0.5).
- b. (1) As the SRV's open, level swells due to the reactor depressurizing and causing void formation. (0.5)
(2) Neutron flux drop due to flow coastdown causing accelerated voiding effects [0.53. (3) Vessel Pressure increase due to 53.5" level tripping the main and RFP turbine which causes a rapid reduction in steam flow CO.53. REFERENCE GGNS OP-DT-503/Rev. 1, L.O. #'s 1 and 2 202OO1K102 202OO1K105 202OO1K122 ...(KA'S)
5 t__IdEggy_QE_NygLE98_EgWE6_EL@NI_gEEB@IlgNt_EgglDS _@ND t PAGE 32
< IdE6dQDyN@ digs ANSWERS -- GRAND GULF 1 -87/05/1S-PAYNE, D.C.
ANSWER 5.09 (2.50)
- a. Closure of all MSIV's (0.5).
- b. (1) Pressure increase due to closure of all MSIV's (0.5)
(2) Neutron flux drop due to reactor scram CO.253 caused by MSIV closure [0.25] (stram when three MSIV's less . t h an 94 */. op en ) (3) Vessel Steam flow decrease due closure of the MSIV's (0.5) (4) Vessel Steam flow increase due to SRV's opening (0.5) REFERENCE GGNS OP-DT-507/Rev. 1, L.O. #'s 1 and 2 { 202OO1K102 202OO1K105 202OO1K122 ...(KA'S) { ANSWER 5.10 (1.00)
- a. 2
- b. 4 REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 3, pp. 25-30, LO# 4.1 292OO1K102 292OO3K103 ...(KA'S)
ANSWER 5.11 (1.00) C REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 7, pg. 10, LO# 4.3 292OOBK111 ...(KA'S)
"5:__IHEQBy_gE_NgCLE68_EQWE8_E6@NI_QEE6@IlgN _ELQ1QS t _@NQ t PAGE 33 IME80QQ1NBdlCS ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C.
ANSWER 5.12- (1.00) ,
.a. decrease .
*b. increase
- c. increase
- d. increase l
REFERENCE j GGNS: OP-LO-SYS-LP-G33/36, LO# 11.b 204000A109 204000K301 204000K306 ...(KA'S) i ANSWER 5.13 (1.50)
- a. decrease b .- increase
- c. increase REFERENCE GE BWR Academi c Series, Reactor Theory, Chap. 4, pp. 43-46, LO# 4.3, 6.3 GE BWR Academic Series, Heat Transfer and Fluid Flow, Chap. 3, LOM 4 292OO4K111 293OO2K104 ...(KA'S)
. ANSWER 5.14 (2.00)
- a. False
- b. True
- c. False
- d. False (0.5 each)
REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 4, LO# 7.4 292OO4K101 292OO4K103 292OO4K113 ...(KA'S) ANSWER 5.15 (2.50)
- a. True (0,5)
- b. False (0.5)
- c. False (0.5)
- d. True (0.51.
- e. False (0.5) l
1:__IdEQBl_QE_NyCLE@B_EQWEB_E(@N1_QEE6@llgNt _ELylDS _@ND t PAGE 34
+ IdE@dQDYN@ digs ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ;
i l REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 6, LO# 2, 3-292OO6K110 292OO6K114 292OO6K121 ...(KA'S) ANSWER '5.16 (1.00) a..-80 sec (0.5)
- b. long lived neutron' precursors (0.5)
(1/2 credit f or delayed neutron ef f ect) REFERENCE GE BWR Academic Seri es, Reactor Theory, Chap. 7, pg. 22, LO# 7.1, 7.2 292OO3K105 292OOOK125 ...(KA'S) ANSWER 5.17 (1.50)
- a. Power level
- b. Time at power
- c. Time since shutdown REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 7, pp. 23-24, LOM 8.3 292OOOK129 292OO8K130 ...(KA*S)
ANSWER 5.18 (1.00) As voids increase, the density of the moderator decreases so the neutron slowing down time and length become longer, thus resonant absorption increases due to neutrons spending more time in the resonant energy spectrum (0.5) and thus, the fuel temperature coefficient becomes more negative (0.5). I REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 4, p. 39, LO# 6.3 292OO4K107 ...(KA'S)
~ 5 1__18Eggy_QE_Nyg6E@B_EgWE6_E6@NI_QEEg@IlgN t _E(ylg h_@ND' PAGE 35 IdE659Dyd@ digs ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. i i l ANSWER 5.19 (2.00) i i Drawing a straight line between the last two *'s predicts 34-36 control rods must be withdrawn. (0.25 for each point plotted, 0.50 for line and prediction) 5 10 15 20 25 30 35 40- 45 50 55 1.O*----l----l----l----l----l----l----l----l----l----!---- ---l1.0 O.9- -0.9 O.8- * -0.8 O.7- -0.7 1/M 0,6- -0.6 O.5- -0.5 O.4- -0.4 l O.3- * -0.3 : O.2- * -0.2 O.1- * -O,1 _ g _ O.01----l----!----l----l----!----l----X---- ----l----l----l---! O 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 3, pp. 13-15, LOM 2.3, 2.4 292OOBK104 ...(KA'S) l l i
I PAGE- 36 1 [6__E60NI_SYSIEdg_QESIGN_CQN16g(t_@NQ_lNSIBUdEUI@llgN t t i ANSWERS -- GRAND GULF 1 -87/05/10-PAYNE, D.C. l l 'li 1 ANSWER 6.01 (1.00)
- 1. High Drywell Pressure -(1.23 psig) i 2.- Analog Control Circuit failure (or analog circuit demand high; velocity feedback high; position feedback rate of change high; velocity controller oscillations)
- 3. Loss of control power
- 4. Hydraulic Power unit failure (or hydraulic fluid tank low-low l evel; hydraulic fluid high-high oil temperature)
(4 O O.25 pts each) REFERENCE GGNS OP B33-2-501, L.O. #'s 3.c and 5.a.1 202OO2K112 ...(KA'S) ANSWER 6.02 (2.00)
- a. Indicates that at least-one rod has a defective position probe CO.5].
- b. Indicates that all scram valves are not in the same position EO.5].
- c. Indicates that the RGDS finds disagreement between the signals received from the RACS EO.5].
- d. Indicates that the selected rod must be fully inserted before any other control rod can be moved CO.53.
REFERENCE GGNS OP-C11-2-03, L.D. #'s 6.a and 6.b 201005K102 201005K103 201005K601 ...(KA'S) ANSWER 6.03 (1.00) SRVs or Relief Valves RHR SSW CRD Pump (0.25 each)
56t__BLQN1_SYSIEDS_DEgl@Nt_CQNIBQ(t_@ND_lN@l@UDENI@llgN PAGE 37 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. REFERENCE 2950.15K202 ...(KA'S)
. ANSWER 6.04 (1.50)
- c. 130 E0.33; Rx scram (IRM's > 120/125 of scale and mode switch not in RUN) [0.52.
- b. Rx recirc. flux. controller auto switches to APRM channel E. (0.5)
' REFERENCE GGNS OP-C51-4, L.D. #3 GGNS OP-B33-2, L.O. #4 GGNS OP-C51-2, L.O. #5 202OO2K607 215003K101- 215005K109 ...(KA'S)
ANSWER 6.05 (1.00) When activated - The source is positioned to irradiate the detector causing an upscale meter deflection. (0.5) When deactivated - There is sufficient leakage to cause a background level reading, so that a channel failure would be indicated by a downscale alarm. (0.5) REFERENCE GGNS OP-D21-5C?, L.O. #3 and NRC Exam Bank 272OOOK206 272OOOK303 ...(KA'S) ANSWER 6.06 (2.00)
- a. Causes reactor l evel to INCREASE [0.53 due to the Level Control System having a LEVEL ERROR, with NO compensating FLOW ERROR resulting in a SIGNAL to INCREASE the SPEED OF THE REACTOR FEED PUMPS
[0.53.
- b. Reactor level should REMAIN CONSTANT CO.53 because the "B" FEED PUMP Turbine Control Unit will lock the pumps at the speed at the time of the failure CO.53.
REFERENCE GGNS OP-C34-501, L.O. #'s 6.a and 6.c. 259002K103 259002K105 259002K507 ...(KA'S) { l
o '6g__PL@N1_@y@ led @_DE@l@Ut_GQN16QL t _QND_ lng 16UDENI@llgN PAGE 38 l . ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 6.07 (2.00)
- a. 1. Allows the output breaker to be shut (and parallel the diesel with normal source) (0.5)
- 2. Resets the diesel governor to the droop mode (0.5)
'b . 1. Diesel speed (frequency) (0.5)
- 2. Diesel load control (0,5)
REFERENCE GGNS OP-P75-02, L.D. #5 GGNS SOI-P75-1
'264000K101 264000K403 264000K406 264000K505 264000K507
...(KA'S)
ANSWER 6.08 (1.50)
- a. No (0.25). The wide range l evel indication is calibrated with no j et pump f l ow (0.25) and at 1 0 0*/. p o w e r the level indication will be lower due to the draw down effect (0.25).
- b. No (0.25). TAF i s -167 inches (0.25) and the low end of the wide range is -160 inches (0.25).
REFERENCE GGNS OP-B21-03, L.O. # 3.b and 4 216000K501 ...(KA'S) ANSWER 6.09 (1.75)
- a. 1. Reactor or steam header pressure (0.25) l
- 2. Generator load (MWe) (0.25)
- 3. Turbine speed (0.25)
- b. The pressure control circuit (0.5)
- c. None (0.5)
REFERENCE
'GGNS OP-N32-2, L.O. # 4 and 5 245000K108 245000K409 ...(KA'S)
I L [,61__P.L@N1_@ySIEUS_ DESIGN t _CQNIBQL_@ND_INSIBUDENI@llgy t PAGE 39 ANSWERS - GRAND GULF 1 -87/05/18-PAYNE, D.C. I
- 1 ANSWER- 6.10 (2.50) .
- a. 1. AI
- 2. FO
- 3. FO
- 4. FC l b. Valve stem air to the Off-Gas System is lost providing a potential flow path for airborne contamination.
REFERENCE l GGNS: DNEP 05-1-02-V-9, OP-LO-SYS-LP-P53-03, OP-N21-501, LO#6 OP-LO-SYS-LP-P42-02,= LOMBd, OP-C71-501, LO# 7b OP-LO-SYS-LP-C11-1A, LO #Ba 1 201001K109 223OO1K110 259001K106 295019AK2O ...(KA'S) l l ANSWER 6.11 (2.00) l 1. Decay of rated power Xenon inventory l 2. Positive reactivity ef f ects f rom eliminating steam voids - j (or reduced neutron leakage from boiling to cold) ! 3. Reduced doppler effect l 4. Decreasing rod worth as moderator cool s 1: 5. Imperfect mixing (Any 4 0 0.50 each) l REFERENCE-l GGNS: OP-LO-SYS-LP-C41, LO #2 211000K301 ...(KA'S) ANSWER 6.12 (1.00)
- a. Failure of Recirc Pump "A" #1 seal
-b. Plugging of Recirc Pump "A" #2 seal internal restriction / breakdown orifice
-(0.5 each) i REFERENCE l GGNS: OP-LO-SYS-LP-B33-1-03, LO #3c 202OO1A109 202OO1A110 202OO1A210 202OO1A411 202OO1K605
...(KA'S) l i
t
'6 t __ P(@NI _ @1SIEd @ _DE @ l@y t _ ggNI696 t _QND _lNS169dE gl@llgy PAGE 40 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 6.13 (2.00)
- a. 1. Low lobe oil pressure
- 2. High jacket water temperature
- 3. Loss of excitation (generator lockout)
- 4. Reverse power (generator lockout)
- 5. Generator overturrent with voltage restraint
- 6. High crankcase pressure (any 4 & O.25 each)
- b. 1. Reactor vessel level low -41.6"
- 2. High drywell pressure +1.39 psig
- 3. Manual HPCS System initiation pushbutton
- 4. Loss of normal power to ESF Bus 17AC f or 2.3 seconds (any 3 0 0.33 each)
REFERENCE GGNS: OP-LD-SYS-LP-PB1, LO #4a 264000K402 264000K408 ...(KA'S) ANSWER 6.14 (1.50)
- a. 125 psig (after preset time delay) (0.5)
- b. Six (6) (0.5)
- c. Reset the starting circuit on the local panel (P134) (0.25) by placing the Deisel Driven Fire Pump A/B Selector Switch to the OFF position (0.25).
REFERENCE GGNS: OP-P64-501, LO #3b,c 286000A301 286000A304 286000A406 286000K402 ...(KA'S) ANSWER 6.15 (1.00) The FCV opens up as reactor pressure increases maintaining a constant flow and therefore, a constant pressure to the PCV. REFERENCE GGNS: OP-LO-SYS-LP-C11-1A-03, LO #3e 201001A101 201001K408 ...(KA'S)
Ac__eLeNI_@y@lEMS_DESl@N_CQN16Q(t_6ND_lN@l69MENI@llON t PAGE 41 l .
- l. . ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C.
l 1 l ANSWER 6.16 (1.25)
- a. SBGTS Train "A" will continue to run (1.0)
- b. SBGTS Train "A" will stop (0.25)
- j. fiFERENCE
- GGNS 'OP-LO-SYS-LP-T48, LO P5 l 261000A301 261000A302 261000K401 ...(KA'S)
ANSWER 6.17 (1.75)
- a. The three primary divisions of the Li quid Radwaste System are:
- 1. Equipment Drains
- 2. Floor Drains
- 3. Chemical Wastes (0.25 each)
- 6. 1. Inoperative
- 2. High, high radiation .
1
- 3. Downstale
- 4. High discha:ge flow (through F0355 valve >100 gpm)
- 5. Low PSW or cire, water flow (or low dilution flow) (<2500 gpm) i (any 4 0 0.25 each)
REFERENCE GGNS: OP-LO-SYS-LP-G17, LO #1,5 26BOOOGOO4 268000K105 268000K106 ...(KA'S) l 1
t t PAGE 42 lZt__es91969G1ceL_c9
.- PROCEDURES-NORMAL NIB 96 _ABNgRMAL_ EMERGENCY _AND l-L ANSWERS -- GRAND GULF'l -87/05/18-PAYNE, D.C.
l l l , ANSWER 7.01 (1.50)
- s. Verify turbine speed steady between 1800 and 1850 RPM (0.5)
- b. 1. Manually trip the turbine (with the MN TURB TRIP pushbutton on the P680 panel) (0.5)
- 2. Run the MHC START DVC on P6BO to zero (to bias the turbine l valves closed) (0.5)
REFL,tENCE ONEP-05-1-02-I-2, Steps 4.4 and 4.5 245000A201 245000G014 ...(KA'S)
-ANSWER 7.02 (1.50)
- a. Enter procedure EP-2, RPV Control (0.5)
- b. Place the mode switch in shutdown (0.5) and close all group I isolation valves (0.5)
REFERENCE i GGNS ONEP-05-1-02-I-3, Steps 4.2 and 4.3 239001K401 239001K606 ...(KA*S) ANSWER 7.03 (2.00) 1
- a. 14.5 feet (0.5) {
- b. 140 degrees F (0.5) i e 212 degrees F (0,5)
- d. The limits ensure adequate NPSH available to the ECCS pumps (0.25) and. adequate cooling of the RCIC lube oil (0.25)
REFERENCE GGNS EP-3, Steps L-16, L-17 and L-20
.GGNS OP-SPDS-OO4-01 295029K201 295029K203 295029K209 29503OK102 29503OK203
...(KA'S)
I J l ________ N
Z___EBgCEDUBES_;_Uggd@bt_@BygBd@6t_EDEBGENC1_6MD PAGE 43 bed 1969G1G86_GQNIB96 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C.
ANSWER 7.04 (2.00)
- a. Containment spray will be initiated immediately (0.5) (whether or not adequate core cooling can be maintained) -OR-Vent containment (per step PC/P-10) (0.25) if >17.25 psig (0.25)
- b. No assurance can'be given that adequate core cooling can be maintained if the containment f ails (0.5). (The failure mecnanism will probably result in loss of source water to the ECCS components EO.253 and a loss of adequate core cooling CO.253).
- c. Can be initiated (0.5)
- d. (When the pressure and temperature rel ationship are in the shaded portion of the spray initiation curve, the combination of evaporative cooling and convective cooling), results in depressurization rates which exceeds the negative design pressure of the containment (0.5)
REFERENCE GGNS OP-SPDS-LP-OO4-01, L.D. # 11 295024K303 ...(KA'S) ANSWER 7.05 (2.00)
- a. MSIV low level i sol ati on (1.0)
- b. Utilization of the main condenser as a heat sink to prevent exceeding the suppression pool heat capacity temperature limit is of sufficient importance to warrant bypassing the interlock (1.0)
REFERENCE GGNS OP-SPOS-LP-010-01, L.O. # 3, 4 and 5 295025K203 295025K301 ...(KA'S) I
ZL _E@OgEQQ6ES_ _UQBd@(t,6@NQBd@(t_EUEBGENgy_@ND PAGE 44 B091969 Gig @b_ggNI@Q6 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 7.06 (2.00)
- 1. Secondary containment area or HVAC rad level above max norm levels
- 2. Secondary containment floor or sump l evel s above the max norm l e v el'4
- 3. Secondary containment temperature or HVAC differential temperature above max norm levels
- 4. Secondary etntainment dif f erential pressure at or above zero inches water.
(4 @ O.5 each) REFERENCE GGNS OP-SPDS-LP-OO6-01, L.O. # 2 295000G011 ...(KA'S) ANSWER 7-.07 (2,00)
- 1. Scram condition and reactor power greater t h an 4 */. or undetermined
- 2. Drywell pressure above 1.23 psig.
- 3. Reactor pressure above 1064.7 psig
- 4. Reactor vessel wate- below -41.6 inches or undetermined (4 @ O.5 each)
REFEFtENCE GGNS OP-SPDS-LP-OO3-01, L.D. # 3 and 5 295000G011 ...(KA'S) ANSWER 7.08 (2.50)
- a. ' Water level is allowed to decrease til;
- 1. power < 4 */. o r
- 2. reactor level decreases to TAF or
- 3. D/W pressure < 1.23 psig and no SRVs cycling
- b. 1. CRD
- 2. SBLC (0.5 each)
REFERENCE GGNS OP-SPDS-LP-010, L.O. # 3 and 4 295000G011 ...(KA'S)
lt__P6DCEDU6ES_;_NQBd@(t_@@NgBd@(t_EdEEGENCy_@ND PAGE 45 B8919600lG86_GQNIBQL ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 7.09 (1.00) The pump must be started to prevent a loss of hydrogen seal oil (0.5) and prevents the possible ignition of the hydrogen (0.5) (f rom the gas ascape velocity) . REFERENCE GGNS ONEP-05-1-02-I-4 294001K115 295003K206 ...(KA'S) ANSWER 7.10 (3.00) Cendidate #1 - Reject, would exceed the 10 CFR 20 quarterly limit
-DR-Reject, has exceeded the adiin quarterly limit Candidate #2 - Reject, would exceed the administrative quarterly limit
-OP.-
Accept, with proper higher management authority (Radiation Protection Manager & General Manager) , Candidate #3 - Accept, Even though the candidate would exceed the 1 5(N-18) limit, the limit only applies if the person is going to 3000 mrem / quarter. (3 0 1.0 each) REFERENCE 10 CFR 20 and GGNS Admin procedure 01-5-08-2 294001K103 ...(KA'S) ANSWER 7.11 (1.50)
- a. To prevent wetting of the charcoal adsorber bed. (0.5)
- b. 1. To allow system warmup (0.5)
- 2. To remove any hydrogen from the system (0.5)
REFERENCE GGNS SOI-04-1-01-N64-1 271000K404 ...(KA'S) 1
- _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ i
'Zi__PBQCEQUBES_ _NQBd@bt_8BNQBd@bt_EdE6GENCy_@NQ- PAGE 46 ' 8691969GIC86_CgN16Q(
ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER 7.12 (2.00)
- a. 1. Preclude injury to personnel
- 2. Protect the Public health and safety
- 3. Prevent vital pla . squipment damage
- 4. Immediate actions must be taken to mitigate the consequences of the emergency (No procedure addresses the problem)
(3 & O.50 each)
- b. As soon as practical advise their immediate supervisor (0.5)
REFERENCE GGNS Procedure 01-S-02-1, Page 4 207000G015 223OOOG015 263OOOG015 ...(KA'S) ANSWER 7.13 (1.00)
- a. Voltage gradient capactors will overheat (0.5)
- b. Open the disconnects (0. 5's REFERENCE GGNS ONEP 05-1-02-I-02 262OO1K103 ...(KA*S)
ANSWER 7.14 (2.00)
- a. TSC - (177' Elevation of ) the Control Building (0.5)
OSC - Maintenance Shop (of the Admini stration Building) (0.5)
- b. 1) Assess emergency situations (0.5)
- 2) PAG Recommendations (to state and local of fici als) (0.5)
REFERENCE GGNS: OP-DT-533, LOM B.1, E.1, K.2 EPP 10-S-01-7, 10-S-01-25 294001A116 ...(KA'S) i
}
'B. ADMINISTRATIVE PROCEDURES t _CgNDITIQNgt_AND_LIMITATIgN@
PAGE 47 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. l l ANSWER B.01 (1.00) l a REFERENCE GGNS: OP-AD-501, LO#1; AP 01-5-06-1 294001K102 ...(KA'S) b ANSWER 8.02 (1.00) a or c REFERENCE GGNS: OP-AD-501, LO#2; AP 01-S-06-2 2940014103 294001A110 294001A111 294001K116 ...(KA*S) ANSWER 8.03 (1.50) The f ollowing checks should be made:
- Breaker racked in
- Breat<er charging springs charged.
- Charging motor disconnect switch on.
-- Control power on or breaker fusesinsta: fled. '
(any 3 D O.5 each) REFERENCE GGNS: Procedure 02-S-01-2; OP-AD-501, LC# 3 294001A116 ...(KA'S) a e e
l l 'B __B901NISIB811yE_PBQCEQQBES t _CQNQlligNSt _@NQ_LidlI@IlgNQ PAGE 48 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE,.D.C. ANSWER 8.04 (1.00) (*) - Preceeding step requires the checksheet to be perf ormed if the plant has been shutdown for a given period of time (> l2 weeks). (0) - Indicates that a significant period of time may be required for the completion of this step of the checksheet. (0.5 each) REFERENCE GGNS: Procedure 02-S-01-2; 101-03-1-01-1 OP-AD-501, LOM 3,5 294001A101 294001A102 ...(KA'S) ANSWER 8.05 (1.00) C REFERENCE GGNS: TS DEFINITION 1.6 294001A111 294001A112 ...(KA'S) ANSWER 8.06 (1.00) d REFERENCE GGNS: TS 3/4.1.4 and Bases 201005 GOO 6 ...(KA'S) ANSWER 8.07 (1.00) d REFERENCE GGNS: .OP-PB-601, LOH4; TS 3/4.5.1 203OOOGOO5 209001 GOO 5 209002 GOO 5 218000 GOO 5 ...(KA'S) 1
i 8. ADMINISTRATIVE PROCEDURES t _CgNDITIQNSt ,,AND_LIMITATIgNg PAGE 49 1
- j .; . ANSWERS -- GRAND GULF 1. -87/05/18-PAYNE, D.C.
l'
. ANSWER .8.08 (1.00)
D REFERENCE GGNSt.OP-PB-601, LO#4; TS 3.8.1.1
'2620010005 264000 GOO 5 ...(KA'S)
ANSWER 8.09 (1.00)
'd REFERENCE GGNS: OP-PB-601, LO#4; TS 3,3.1 and 3.3.2 272OOOGOO5 ...(KA'S)
ANSWER 8.10 (3.00)
- a. - Fal se
- b. False L c. True
- d. False
- e. True f ,. False (0.5 each)
REFERENCE GGNS: OP-PB-601, LO#2; TS, secti on 6. 2. 2 294001A103 294001A109 294001A111 ...(KA'S) l ANSWER 8.11 (2.00)
- a. 16
- b. 8 l c. 72
- d. Accept any of the following: GGNS General Manager, or his designee; l or higher lovels of management, in accordance with established procedures.
l (0.5 each) 1 I l l ._ _ _ _ _ _ _ _ _ _ . . . . _ _ . _ _ _ _ _ _ _ _ _ _
's.__eDUJNJpIggIlyg_PgggggUgggg_ggNQJIlgNg1_gNQ_L}D]I@IlgNg PAGE 50 ANSWERS -- GRAND GULF 1 -87/05/10-PAYNE, D.C. I REFERENCE GGNS: OP-AD-501, LOM4; TS, pg. 6-2 294001A103 294001A109 ...(KA'S) ANSWER 8.12 (3.00)
- a. 1325 psig
- b. 25'/.
- c. 785 psig
- d. 10*/.
- e. the top of the active irradi ated f uel (-167" H2O or 366.3" from l inside vessel bottom head)
- f. 1.06 (0.5 each)
REFERENCE GGNS: TS, section 2.1, pg. 2-1, 2-2 202OO2 GOO 5 295025 GOO 3 295031 GOO 3 ...(KA'S) ANSWER 8.13 (1.50) d,f,b,c,a,e (0.25 each) REFERENCE GGNS: OP-AD-501, LO#2; AP 01-S-06-2 294001A116 . . . (K A' S)
~@i__bDylN1 SIB @llVE_P8QGEQUBEgt_GQUQillgd@t_@NQ_Gidll@llgN@ PAGE 51 ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. . ANSWER B.14 (3.00)
- a. Whole body - 1.25 Rem Extremities -
18.75 Rem Skin - 7.5 Rem (0.5 each)
- b. 1. Total WB dose per quarter less than or equal to 3 Rem, AND
- 2. WB dose + accumulated occupational dose to the WB less than or equal _to 5(N-19) Cwhere N = age in years at last birthday], AND
- 3. Licensee has determined the individual's accumulated occupational dose to the WB on NRC Form 4. (0.5 each)
REFERENCE GGNS: OP-PB-601, LO#5; 10 CFR 20 294001K103 ...(KA'S) ANSWER 8.15 (1.00)
- a. Suppr essi on Pool Operability (lS 3/4.6.3)
- 6. End-of-Cycle Recirculation Pump Trip System Instrumentation (TS 3.3.4.2)
- c. Reactor Protection System Instrumentation (TS 3/4.3.1) for TCV-fast closure (0.33 each)
REFERENCE GGNS: OP-PB-601, LO#1; TS 3/4.6.3, TS 3/4.3.4.2 & TS 3/4.3.1 202OO2 GOO 5 218000 GOO 5 223OO1 GOO 5 ...(KA'S) l
1
'es__@DdlN1QIB@IlVE_PBQCEQUBE@t_CQNDillgN@t_@NQ_LidlI@IlgN@ PAGE 52 l
ANSWERS -- GRAND GULF 1 -87/05/18-PAYNE, D.C. ANSWER B.16 (2.00) L(This situation places the plant in ACTION section "c" of TS 3/4.3.2 , which requiress) 1
- a. Place at least one trip system in the tripped condition within one hour, AND (take ACTION required by Table 3.3.2-1) (see below)
(f or answers b-d below either the ACTION number or actual ACTION statement is a suf ficient response) L b. ACTION 20 - be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. 1
- c. ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the Standby Gas Treatment System. operating within one hour,
- d. ACTION 27 - Close the af f ected system isolation valves within one hour and declare the af f ected system inoperable.
(0.5 each) REFERENCE GGNS: OP-PB-601, LO#4; TS 3/4.3.2 223OO1 GOO 5 ...(KA'S) ANSWER B.17 (2.50)
- a. Operating under heavy load at < or = 59.5 Hz may cause severe i vibration of the turbine blading CO.53 (as a result of the steam l excitation frequency approaching blade resonance) which produces fatigue stress CO.53.
- b. (Last rows of the long) low pressure turbine blades CO.53.
- c. Turbine blades should not be subjected to significant periods ,
(more than 10 minutes) CO.53 of severe vibration CUMULATIVE over I their lifespan EO.53. REFERENCE GGNS: OP-AD-549, LOM6 245000A209 245000K405 245000K502 294001K106 295005 GOO 7
...(KA'S) i
_-----_-_------_j
TEST CROSS REFERENCE PAGE 1 QUESTION- VALUE REFERENCE 05.01 1.00 DCPOOO1143 05.01 1.25 DCPOOO1183 05.03 2.00 DCPOOO1184
.05.04 2.00 DCPOOO1187 05.05 1.00 DCPOOO1189 05.06 1.00 DCPOOO1191 05.07 1.25 DCPOOO1193 05.08 2.00 DCPOOO1194 05.09 2.50 DCPOOO1195 05.10 1.00 DCPOOO1138 05.11 1.00 DCPOOO1146 05.12 1.00 DCPOOO1134 05.13 1.50 DCPOOO1141 05.14 2.00 DCPOOO1140 05.15 2.50 DCPOOO1145 j 05.16 1.00 DCPOOO1144 05.17 1.50 DCPOOO1147 05.18 1.00 DCPOOO1132 05.19 2.00 DCPOOO1142 28.50 ;
06.01 1.00 DCPOOO1197 06.02 2.00 DCPOOO12OO 06.03 1.00 DCPOOO1201 06.04 1.50 DCPOOO1202 06.05 1.00 DCPOOO1203 06.06 2.00 DCPOOO1204 06.07 2.00 DCPOOO1209 06.08 1.50 DCPOOO1211 06.09 1.75 DCPOOO1213 06.10 2.50 DCPOOO1165 06.11 2.00 DCPOOO1151 06.12 1.00 DCPOOO1157 06.13 2.00 DCPOOO1160 06.14 1.50 DCPOOO1163 06.15 1.00 DCPOOO1155 06.16 1.25 DCPOOO1161 06.17 1.75 DCPOOO1164 26.75 07.01 1.50 DCPOOO1215 07.02 1.50 DCPOOO1216 07.03 2.00 DCPOOO1217 07.04 2.00 DCPOOO1218 07.05 2.00 DCPOOO1219 07.06 2.00 DCPOOO1220 07.07 2.00 DCPOOO1222 07.08 2.50 DCPOOO1223
TEST CRDSS REFERENCE PAGE- 2
- v. ,
QUESTION VALUE REFERENCE 07.09 1.00 DCPOOO1225 1 07.10 3.00 DCPOOO1226 07.11 1.50 DCPOOO1229 07.12 2.00 DCPOOO1231 07-.13 1.00 DCPOOO1232 07.14 2.00 DCPOOO1233 26.00 08.01 1.dO DCPOOO1168 08.02 1.00 DCPOOO1172 08.03 1.50 DCPOOO1234 08.04 1.00 DCPOOO1235 08.05 1.00 DCPOOO1236 08.06 1.00 DCPOOO1175 08.07 1.00 DCPOOO1177 08.08 1.00 DCPOOO1181 08.09 1.00 DCPOOO1182 08.10 3.00 DCPOOO1173 08.11 2.00 DCPOOO1170 08.12 3.00 DCPOOO1171 08.13 1.50 DCPOOO1169 08.14 3.00 DCPOOO1178 08.15 1.00 DCPOOO1179 08.16 2.00 DCPOOO1180 08.17 2.50 DCPOOO1176 27.50 108.75 l l l I I _ - - - - - _ - _ - _ _ _ - _ _ _ _ _A}}