ML20211A197
| ML20211A197 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/10/1987 |
| From: | Cooley R, Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20211A114 | List: |
| References | |
| 50-285-OL-87-01, 50-285-OL-87-1, NUDOCS 8702190112 | |
| Download: ML20211A197 (118) | |
Text
{{#Wiki_filter:- . .- . . =. .. 6 Fort Calhoun Station Examination Report 4 50-285/0L-87-01 i ~ Docket No: 50-285 License No.: DPR-40 Licensee: Omaha Public Power District 1623 Harney Omaha, Ne 68012 Examinations administered at Fort Calhoun Station Chief Examiner: - - S. L. McCrory, Le$ ExMiner Date Approved by: k C1- h//0/ h7 t,A. Cooley,SectynChief Ddte ' C Summary Requalification examirations were conducted on November 18, 1986. Written and operating (oral) examinations were administered to three Reactor Operators and five Senior Reactor Operators. Three Senior Reactor Operators failed the written examination and one of these three also failed the operating examination. All others passed the examinations. 8702190112 070211 PDR V ADOCK 05000298 PDR
2 Report Details
- 1. Examination Results SRO Requalification R0 Requalification Total Pass Fail % Total Pass Fail %
5 2 3 40 3 3 0 100
- 2. Examiners S. L. McCrory, Chief Examiner.
R. G. Clark, PNL J. L. Pellet, NRC
- 3. Examination Report' This Examination Report is composed of the sections listed below.
, A. Examination Review Comment Resolution B. Exit Meeting Minutes C. General Comments D. Requalification Program Evaluation Report E. FCS Examination Key (SR0/R0 Questions and Answers) Performance results for individual examinees are not included in this report because examination reports are placed in'NRC's Public Document Room as a matter of course. Individual results may be retained in the NRC , Region office during the period that the facility is evaluated as unsatisfactory but are not subject to public disclosure. A. Examination Review Comment Resolution In general, editorial comments or changes made during the examination, the examination review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made during the examination review. The modifications discussed below are included in the master examination key which is provided elsewhere in this report as are all other changes mentioned above but not' discussed herein. Attachments I and 2 are the facility comments on the examination. Only NRC responses to facility comments are provided below. Refer to 4 i _...y_ . _ _ .
,.,___...,-_y , , _ . , _ _ , - _ . , , , _ _ , . - , , , . . - _ - . . _ _ , _ , _ , , .. . . . , .
3 attachment 1 or 2 as indicated for the facility comments. Unless otherwise indicated in this section, the facility comments were incorporated into the answer key. RESPONSES (1) 5.03 NRC endeavors to ensure that system parameter values used in its questions reflect' those that operators expect to find at their plant. In this particular question the magnitude of Tave is irrelevant since it is only used as a reference point and not as a potential indication of abnormal plant conditions. Att 1 (2) 5.04 This is not double-jeopardy-because wrong values or calculations were only penalized once. If an incorrect values was computed in part A but used correctly in part B, then part B is given full credit. In part C, credit was given for final values of RCS temperature between 532 and 540 ONLY if it was clear that the value was a result of operator control via the steam dumps and not a result of MTC. Att 1 & 2 (3) 5.06 This question was deleted from the examination. Att 1 & 2 (4) 6.05 Since the examinee ic given a drawing which limits the range of possible answers to what can be found on the drawing, it is not necessary to identify tne number of responses sought. This question was given in a previous examination at FCS and the drawing was evaluated by training staff personnel as superior for training purposes and much easy to follow than the P&ID. Att 1 (5) 6.06a The examinees were verbally told how to interpret this question during the written examination. Att 1 (6) 7.02 This question is consistent with item development techniques for matching type questions. The question is not open-ended because all possible combinations of answers are provided. The addition of distractors (answers not used), repeated use of an answer, and multiple answers to one question item are considered appropriate methods by item development experts. This reveals a greater depth of understanding of the subject covered by the question. Att 1 (7) 8.03b FCS memo FC-T-033-86 dated February 13, 1986 stipulates the administrative procedures (standing orders)for which operators are held responsible. Neither 50 G-33 nor NPDPP G-12 are contained in that memo. The question
4 refers to SO G-10, found in the memo, which does not identify anyone other than NRC for approval of T.S. changes. The comment was rejected. Att 2 (8) 8.05 The procedure states that NRC requires imediate reports of information covered by the six subject areas. NRC interprets this to imediately upon receipt of the information or recognition of the condition NOT immediately upon discovering the report requirement. Therefore, it is appropriate to expect senior operators to be conversant on these subject areas without reference to procedure. Att 1 (9) 8.09b The question asked if NRC approval was REQUIRED. The fact that NRC concurrence was sought is irrelevant. Att 1 (10)8.09c This coment encourages tunnel vision toward Tech Specs. When a senior operator is reviewing Tech Specs to determine compliance no one will always be looking over his/her shoulder to direct his/her attention to all the parts of Tech Specs which may apply. Att 1 (11) Gen. The Reactor Operator examinations were graded prior to receipt of the facility review coments which were not submitted in a timely manner. Since all Reactor Operators passed the examination on th initial grading, the facility review coments were not incorporated into the answer key nor a regrade performed. These coments will be reviewed for revision of questions in the examination bank for future use. (12) Gen. Several facility review coments interpreted NUREG 1021. Interpretation and application of NUREG 1021 is an NRC function. Facility coments should focus on factual technical matters. B. Exit Meeting Sumary At the conclusion of the examination period, the examiner met with the members of the facility staff to discuss the results of the examinations. The following personnel were present for the exit meeting:
5 NRC UTILITY T"hcCrory J. Fluehr P. Harrell J. Gasper R. Clark, PNL C. Brunnert M. Core D. Munderloh L. Kusek NRC informed the facility staff that all examinations must be reviewed by Region IV staff prior to release of any examination performance results. NRC examiners reported that Senior Operators demonstrated significant difficulty discussing or using the Emergency Plan Implementing Procedures (EPIPs). Discussions on information collection, dissemination, and personnel utilization revealed a general lack of
- understanding and ability to use the EPIPs beyond basic event 1 classification.
The chief examiner expressed a safety concern which was identified during the course of oral examinations. There are potential iradequacies in the procedures and comunication systems used to shutdown and control the plant from the Safe Shutdown Panel. When required to evacuate the Control Room as a result of a fire, the operators proceeded to shutdown the plant using A0Ps-687. No operator indicated that the-E0Ps would be consulted as a matter of course for this situation. Further, it could not be demonstrated l that the A0Ps used satisfied the requirements of E0P-1 or cross referenced any E0Ps based on observed indications. 4 C. General Coments i During the week of November 18, 1986, NRC was conducting Emergency Preparedness evaluations concurrent with the requalification program
- evaluation. Some senior. operators who were not participating in the l requalification examinations that week were interviewed to determine their ability to execute the EPIP. Those senior operators
, demonstrated weaknesses similar to the weaknesses noted in the senior l operators who took the NRC requalification operational examination. l It is apparent that a significant weakness exists in the area of l emergency plan implementation by senior licensed operators. l During the week of December 15, 1986, additional senior operators i were given modified operating examinations to determine if the l facility could continue to operate. Those senior operators'were evaluated as satisfactory to continue perfonning their current licensed activities subject to FCS administrative requirements for specific additional training for certain supervisory functions.
l 6 I l On December 22, 1986, NRC met with OPPD staff in the Region IV offices to discuss the weaknesses of the FCS requalification training ; program and review the facility proposed corrective actions. At this meeting 0 PPD staff gave additional examination review comments, Attachment 2, tc the NRC license examiners and asked for a second regrade of specific senior operator examinations. The resolution of those comments is contained in section 3.B of this report. The facility's concerns about the failed operating examination were discussed and the facility was told that the evaluation would not be reversed. The presentation given by OPPD, Attachment 3, addressed both the specifics of the recent requalification examination, and facility immediate and Icng term corrective actions. NRC generally concurred with OPPD findings concerning program deficiencies and proposed corrective actions. NRC informed OPPD that licensed operators who had not been evaluated by NRC within the past two years must be removed from licensed. duties until they could be examined by NRC. NRC provided the names of those operators who were affected by this requirement. Further examinations were tentatively scheduled for March and June, 1987. These examinations are not intended as further evaluations of the FCS requalification training program, but are required for those operators restricted from licensed activities as a result of failure of the November 18, 1986 examination or not having received NRC evaluation in the last two years. The program will be reevaluated during the next requalification cycle examination period. In an OPPD letter dated January 23, 1987, Attachment 4, OPPD presented a schedule of corrective actions for upgrading the FCS requalification training program and to confirm dates for _ reexaminations and program reevaluation. NRC ccncurs with the OPPD proposed corrective actions and the schedule of implementation. NRC also confirms the requested examination schedule. Except that examinations will be given the week of March 2, 1987 vice on March 5.
7 D. Requalification Program Evaluation Report Facility: Fort Calhoun Station D Examiner: S. L. McCrory Dates of Evaluation: 11/18-20/86 12/15-18/86 Areas Evaluated: X Written X Oral Simulator l R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M, or U) Written Examination 3/0 2/3 5/3 U Operating Examination Oral 3/0 10/1 13/1 S Simulator na Evaluation of facility written examination grading na Overall Program Evaluation Satisfactory: Marginal: Unsatisfactory: X (List major deficiency areas with brief descriptive comments.) The FCS requalification training program is unsatisfactory for certifying the proficiency of licensed operators for license renewal. Submitted: Forwarded: Approved: Examiner Section Chief Branch Chief
----------------------___ . I
8 t E. FCS EXAMINATION KEY Date Administered: 11/18/86 Exam Type: Senior Reactor Operator Requalification Reactor Operator Requalification l
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _EIz_QaLUQUN REACTOR TYPE: _PW8_QE__________________ DATE ADMINISTERED: _ghtillig________________ EXAMINER: _MQQBQRyz_Sz_____________ CANDIDATE: _________________________ IN1189QI1981_IQ_Q6NQ1Q81El Reed the attached instruction page carefully. This examination replaces the current cycle 'acility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of-a requalification examination prepared and administered.by your training staff. 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 four (4) hours after the examination starts.
% OF
- CATEGORY % OF CANDIDATE'S . CATEGORY
__V8LUE_ _IDI6L ___EQQBE___ _V8LUE__ ______________QeIEn0BY___________.__ l _16tHQ__ _2512Q ___________ ________
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
. THERMODYNAMICS f _1kzQQ__ _25z20 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _15z50__ _24z41 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, i EMERGENCY AND RADIOLOGICAL CONTROL l _1kzQQ__ _2512Q ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS _h2tEQ__ ___________ Totals , Final Orade l-l All work done on this examination is my own. I have neither given
- nor received aid.
l I Candidate's Signature t
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. . Cheating on the examination means an automatic denial of your application L 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 galy to facilitate legible reproductions. l 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
, 6. Use only_the paper provided for answers. j 7.. Print your name in the upper right-hand corner of the first page of ggeb ? section of the answer sheet. 8.. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a Dgw page, write GDly 20 90e gidt of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
10.' Skip at least thtgg lines between each answer. 11.. Separate answer sheets from pad and place finished' answer sheets face down on your desk or table. , 12. Use abbreviations only if they are commonly used in facility 111gtgtung.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the gggmingt only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
18.'When you complete your examination, you shall: j a.- Assemble your examination as follows: (1) Exam questions on top. (2) . Exam sids - 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 defined by the examiner. If after leaving, you are found in this area while the examination is still in-progress, your license may be denied or revoked.
I l l' 5t__IBEQBl_QE_UUQLE8B_EQWEB_EL8UI_QEEBall0Nt_ELVIQSt_6NQ PAGE 2
.IBEBdQQ188dIQS l l
QUESTION 5.01 (2.50)
- c. What are the bases (3) for the Transient Rod Insertion Limit? (1.5)
- b. Explain how and why boron concentration does or does not affect the Long Term Steady State Insertion Limit while operating at power.(1.0)
QUESTION 5.02 (3.00)
- a. What TWO parameters can the operator control to prevent pressurized thermal shoc<? (1.0)
- b. What is the sequence of generic events that could lead to pressurized thermal shock conditions? (1.0)
- c. Why does the concern about brittle fracture of the reactor pressure vessel i n c r e. as e as the FCS plant ages? Include in your answer the specific material property that is affected. (1.0)
QUESTION 5.03 (1.00) Pressurizer pressure is 2200 psia; RCS Tavg is 580 degrees F and core j delta T is 40 degrees F. The RCS is ___ degrees subcooled. Choose the correct answer below. Show calculations for full credit. (1.0) l o. 29.45
- b. 49.45
- c. 69.45 i
- d. 89.45 l
l l i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) ,
3 3
- St__IBEQBl_QE_UVQLEaB_EQWEB_ELBUI_QEEB8110Nt_ELVIDSt_aND 'PAGE 3.
IBEBdQQ1Ned10S t _ QUESTION 5.04 (3.00)
;The RCS is at 2100 psis, 532 deg. F (Mode 3), .and Keff = 0.997 rods at 75" on group 3. Core depletion is 8000 MWD /MTU anc boron concentration is 750 ppm. Ignore-xenon and samarium concentrations unless otherwise stated.
Graphs have been provided to assist with calculations but it may not be necessary to use each graph.
- a. How much rod motion will be required to bring the reactor just critical? Show all calculations. (1.0)
- b. How much; additional rod motion will be required to establish a 1 DPM startup rate after the reactor is made critical? Show all calculations. (1,0) c.- If the 1 OPM startup rate is allowed to continue until power enters
. the power range, what will be-the final values of RCS temperature, pressure, and reactor power? Assume N0 further rod motion (manual or automatic) and all other systems are-operated as required for a normal reactor startup. (1.0) 4 QUESTION 5.05 (2.00)
- a. If a trip from 100% power occurs _with xenon at equilibrium, what is the approximate time interval after trip thar xenon will again pass through the 100% equilibrium value? (0.5)
- b. How would_this approximate time interval. compare if the trip occurred from 50% equilibrium conditions? (0,5)
- c. When a reactor is returned to 100% power from peak xenon conditions, 4 why does xenon reactivity undershoot the 100% equilibrium value?(1.0)
. QUESTION 5.06 (1.50)
Figure 1 contains recorder traces of specific primary plant parameters observed during recovery from a loss of off site power with a steam generator tube rupture. Explain WHAT happened at point A including WHY the i parameter behaved as indicated. (1.5)
, (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
7 St__IBEQBl_QE_UUCLE88 EQWEB_EL8UI_DEEB81100t_ELUIDSt_8UR PAGE 4 IBEBdQQ18851CS QUESTION 5.07 (3.00) The plant has been operating at 60% power for several days with all rods out. A single CEA drops into the core but does not cause a reactor trip.
- o. Explain HOW and WHY the dropped CEA will affect the MAXIMUM (peak) linear power in the core? (1.5)
- b. HOW and WHY will the maximum linear power in the core change over the next three hours if the CEA is not recovered? (1.5)
I t l l i
- (***** END OF CATEGORY 05 *****)
ht__ELeNI_11SIEUS_QE1109t_QQUIBQLt_eUQ_INSIBudENIeIIQU PAGE 5 QUESTION 6.01 (3.00) Assume a large LOCA has occurred.
- c. Can an operator determine if the break is in the hot or cold leg from indications in the control room only? (0.5)
- b. Describe how core cooling is achieved for each break location prior to operator intervention. (1.0)
- c. HOW and WHAT does the operator do to compensate for inadequacies in the cooling regimes dependent on break location? Include a brief description of flow path alterations if any. (1.5)
QUESTION 6.02 (3.00) Describe the primary protective function (basis) for the following reactor trips: (3.0)
- c. The-mal margin / low pressure
- b. Axial power distribution
- c. High reactor power
- d. High rate of change of reactor power
- e. High pressurizer pressure
- f. Containment high pressure 9 Low steam generator water level
- h. Low steam generator pressure
- 1. Low reactor coolant flow
- j. Manual QUESTION 6.03 (3.00)
Sketch a one line diagram of ONE TRAIN of safety inj ect ion. Identify all potential water sources and all intrasystem cross connects. Include all
. pumps, remote / automatically operated valves (valve numbers or names are NOT required), tanks, and heat exchangers. Pressure, temperature, and flow detectors are NOT required.
(3.0) (***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)
K W L - ht__ELeNI_SYSIEdS_DESIGNt_QQNIBQLt_aND_INSIBVdENI8I1QN PAGE 6 L
' QUESTION 6.04 .C3.00)
Figure 3.2 is a simplified drawing of the CVCS. On figure 3.2 indicate all Ecomponents which receive Engineered Safeguard Actuation Signals. For each-
' component, indicate'the type of ESAS (CSAS, VIAS, etc.):and-the final
- condition of the component (open-closed, on-off). ,
(3.0)
- QUESTION 6.05 ~C2.00)
- s. What provides indication of RCS level / volume in the control room when the-system is drained below the_ pressurizer level taps? (1.0)
- b. Aside from indication, what are two specific uses of the signals generated by the low range pressurizer pressure detector? (1. 0 3 -
QUESTION 6.06 (2.00) ,
- n. What is the design fuentiongot the Containment Spray System? .(1.0)
b., How does the CSS continue to perform-its function after~ receipt of a RAS? Include the automatic responses of support / interface systems essential to the proper function of the CSS. (1.0)
(***** END OF CATEGORY 06 *****)
Zz__EBQQEQUBEH_:_NQBdoLt_8QNQBd6Lt_EdE8QEUQ1_3NQ PAGE 7 B8010LQQIQ8L_QQUIBQL QUESTION 7.01 (2.00)
- c. True or false. A person discovering a fire should immediately attempt to extinguish the fire if it is small and then notify the control room. (0.5)
- b. What determines whether S.O. G-28 (Station Fire Protection Plan) or AOP-6 (Emergency Fire Procedure) is the governing procedure for fighting a fire? (0.5)
- c. What two fire locations REQUIRE that the Blair Fire Department be called and asked to provide assistar.ce? (1.0)
QUESTION 7.02 (2.50). Match each of the T.S. limiting conditions for operation (LCOs) with the system or function to which it applies as found in the Technical Specifications. The systems / functions, numbered items, may match more than once or not et all. (2.5) LCO SYSTEM / FUNCTION
- a. unidentified leakage >lgpm 1. CORE
- b. at least 1 boron injection 2. ECCS path with core fueled 3. CCS
- c. at least one mean of temp.- 4. ELEC SYS indication of the SIRWT operable 5. ESFS
- d. DG base day tank full 6. RCS
- e. reactor coolant temp. <515 deg.F 7. CVCS
- 8. RPS
-QUESTION 7.03 (3.00) What are three ALTERNATE means of cooling the core if Shutdown Cooling is lost and cannot be rustored while the reactor vessel head is removed. Include a brief description of the flow path for each case. (3.0) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) __ _ _ _ _ , _ . - -~ _ ._
Zz__EBQQEQUBES_:_UQBdakt_6BUQBdakt_EUEBQEUQ1_aNQ PAGE 8 BaQ19LQQ10aL_QQUIBQL QUESTION 7.04 (3.00) List the means used to verify each of the ~ following conditions as per E0P-1 (Reactor Trip). The number in parentheses indicates the required number of items per condition. Setpoints and parameter values are required for full
-credit if they are stipulated in the EOP. (3.0)
- c. Reactor tripped / shutdown (3)
- b. RCS inventory control (2)
- c. Forced circulation heat removal (3)
- d. Steum dump and bypass valves functioning (2) o Normal containment conditions (5)
QUESTION 7.05 (2.50)- TRUE OR FALSE
- o. The "Employe'e ALARA Concern Form" is signed by personnel who have been given RWP pre-planning to satisfy ALARA requirements. (0.5)
- b. A tbsk with a collective dose of less than one man-rem is exempt from documented ALARA pre-planning. (0,5)
- c. The permissible accumulated dose (PAD) to the head, eyes, liver, or large bones is the same as the PAD to the whole body. (0.5)
- d. The protective clothing prescribed on a RWP is intended to reduce the whole body accumulated dose from exposure to external raciation.(0.5)
- e. The shift HP technician may authorize increased RWP exposure limits for back shift work. (0.5) a
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
i LZi__EBQQEQMBES_=_NQBd6Lt_8BNQBd8Lt_EDEBGEUQ1_6NQ PAGE 9
-88Q10LQQ196L_QQNIBQL . QUESTION 7.06 (2.50)
Fill.in the blank for each of the following statements derived from the precautionsLfor Reactor Startup (OP-7). The blanks may represent single words or pnrases. The answer need NOT be the exact wording found in the procedure. (2.5)
.o. ' Criticality must be anticipated whenever CEAs are being moved or ___.
- b. If core geometry has been altered from that which existed at the beginning of the cycle, perform a ___.
- c. When increasing power, monitor RCS temperature, pressure,'pzr level,.
and charging rate ~ changes to determine if the reactor is at ___.
- d. Do not exceed a sustained startup rate of ___.
.e. The RCS boron concentration MUST NEVER be at a level which would allow criticality ___.
l (***** END OF CATEGORY 07 *****)
st__ogdINISIB8IIVE_EBQggguBgst_gggg111 gust _800_L1511811QU3 PAGE 10 QUESTION 8.01 (1.00) Observing what two limits assures that minimum DNBR is not exceeded? (1.0) QUESTION 8.02 (3.00)
- 8. What is the basis for the limits on radioactivity on the RCS' coolant?
(1.0)
- b. Why are there provisions for allowing operation to continue when a particular sample is significantly above one of the radioactivity limits? (1.0)
- c. Why is there a RCS temperature limit imposed when a radioactivity limit is exceeded? (1.3)
QUESTION 8.03 (1.50) Fill in the blanks for the statements related to SO G-10, Technical Specifications. Blanks may represent single words or phrases. (1.5)
- c. Technical Specifications take priority over any ___.
- b. Exceptions and changes to Technical Specifications for use on an immediate and interim basis are approved by ___.
- c. Standing Order R-4 requires that all Technical Specification violations be reported IMMEDIATELY to ___.
QUESTION 8.04 (3.00)
- c. What are the main fission product barriers in the context of emergency classification under the EPIP? (1.0)
- b. Define the emergency classifications using the barrier criteria.(1.0)
- c. What are the three basic considerations for classification when initiating conditions do not clearly define the event as an ALERT or SITE AREA EMERGENCY? (1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Az__eQd1NISIBoIIVE_EBQQEQUBESt_QQNQ1Il0Nat_aNQ_Lldll611QN1 PAGE 11 QUESTION 8.05 (2.00) After the initial report of an on-going reportable event (one hour notification), the NRC requires IMMEDIATE reports on six information items which.may change after the initial report orfmey not be available at the time of..the initial report. List four of these six general items / topics which must be' reported immediately when the information is available. (2.0) QUESTION 8.06 (1.00) What is the basis for requiring that 3 of 4 high power level channels in the Reactor. Protection System be operable to continue full power operation for any length of time? (1.0)
-QUESTION G.07 (1.00)
!- What is the basis for the minimum EDG fuel oil availability (day tank and
' underground storage combined)? (1.0) t i
QUESTION 8.08 (1.00) l
.With a minimum water level in the. emergency feed water storage. tank, how i long can the S/Gs be fed with all steam being dumped to atmosphere before
! having to use river water as feed? (1.0) i-QUESTION 8.09 (2.50) Special Order 35 revised the limit on the primary-to-secondary leak rate at which action in accordance Technical Specification 2.1.4(3) (allowable [ primary-to-secondary leakage) must be taken.
- o. What is the new limit? (0.5)
- b. Did the new limit require NRC approval? Why or why not? (0.75)
- c. If this new limit is exceeded and action is not taken in accordance with T.S. 2.1.4, explain HOW and WHY Technical S p ec ', f ic at io n compliance is or is not maintained. (1.25)
'.***** END OF CATEGORY 08 *****)
R00000000*000*_ ***************3
NRC LICENSE EXAMINATION HANDOUT EQUATIONS, CONSTANTS, AND CONVERSIONS 6 = m*Cp *deltaT 6=U*A*deltaT P = Po*10sur*(t) p , po.et/T SUR = 26/T T=1*/p+(p-p)/Ip T=1/(p-5) T = ($-p)/X p p = (Keff-1)/Keff = deltaKeff/Keff p=1*/TKeff+hff/(1+1T)
- A = In2/tg = 0.693/tg K = 0.1 seconds-I I = Io*e "*
CR = S/(1-Keff) 2 R/hr = 6*CE/d feet l Water Parameters I gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 ibm /ft3 = 1 gm/cm3 Heat of vaporization = 970 Stu/lbm Heat of fusion = 144 Btu /1% 1 atmosphere = 14.7 psia = 29.9 inches Hg. Miscellaneous Conversions I curie = 3.7 x 10iu disintegrations per second 1 kilogram = 2.21 lbm 1 horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 105 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C 4 32 degrees C = 5/9 (degrees F - 32) 1 Btu = 778 ft-1bf
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Ez__IBEQBl_QE_UUGLEaB_EQWEB_ELeUI_DEEBoIIQut_ELUIRSt_6UQ PAGE 12
'IBEBdQQ1U8510S ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S.
ANSWER 5.01 (2.50)
- a. The Transient Insertion Limits are provided to ensure that;
- 1. acceptable power psaking factors are maintained, (0.5)
- 2. the minimum available shutdown margin is maintained, (0.5)
- 3. the potential consequences of a CEA ejection accident are limited to acceptable levels. (0.5)
- b. Boron concentration has no effect on Insertion Limits. (0.5) The Steady State Insertion Limit is power dependent and requires a boron concentration sufficient to achieve the required minimum CEA height.
(0.5) (Critical boron concentration varies both as a function of core power and core age.) REFERENCE FCS TS 2.10.2 BASES AND FIG 2-4 K/A 001-000-KS.08 I.R. 3.9/4.4 ANSWER 5.02 (3.00)
- n. Coolant temperature (0.5)
Coolant prassure (0.5)
- b. Rapid cooldown and depressurization (0.5) followed by rapid represrurization (0.5).
- c. Neutron exposure (integrated) (0.5) makes the material more brittle (raises NDTT) (0.5).
REFERENCE CE EQB K/A 000-040-EK1.01 I.R. 4.1/4.4 e ANSWER 5.03 (1.00) 8.(0.5) Tsat = 649.45, Th = 580 + (1/2 x 40) = 600 Tsat - Th = 649.45 - 600 = 49.45 (0.5) REFERENCE DASIC HT&FF K/A 001-000-K5.56 1.R. 4.2/4.6
St__IHEQBl_QE_NUQLEeB_EQWEB_ELaNI_QEEBel10Nt_ELU10$t_SUQ PAGE 13 IHEBdQQ1U6dIQS ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. I ANSWER 5.04 (3.00)
- c. rho = (Keff -1)/Keff (0.25)
= (0.997-1)/0.937
= -0.003/0.997
= -0.003009 (or -0.003)
= -0.3% (0.25)
From figure II.B.2, rod insertion worth at 75" on group 3~ 3.0% dp. Therefore, rod insertion worth when j ust critical = 3.0% dp - C-0.3% dp) = 3.3. From figure II.B.2, this is about 42" on group 4 or about 42" (accept 40 to 45) of rod motion. (0.5)
- b. rho = beta /(1+1amda T), T =26/SUR =26/1dpm =26, beta =0.5-0.65% C0.25)
= 0.006/(1+(0.1x263)
= 0.006/3.6
= 0.00167
= 0.167% (accept 0.139 - 0.18%) (0.25)
From figure II.B.2, rod position based on a final integrated worth of 3.467 (3.439 - 3.48) dp, rod height will be about 63" (60 - 65) on group 4. Therefore, the additional rod motion will be 21" (18 - 23). (0.5)
- c. (At this point in mode 3, the SOBCS would be set to control S/G pressure at about 900 psia (532 deg. F). As reactor power passes the point of adding heat (1%), the RCS will begin to heatup which will raise S/G pressure causing the bypass valves to modulate open thereby reducing temperature and pressure. Equilibrium will be achieved when l reactivity added by the power defect (figure II-15) equals the excess inserted by the rods.) Final conditions will be:
t RCS temperature - 532 deg. F (controlled by SDBCS) (0.4) i RCS pressure - 2100 psia (pzr press control system in auto) (0.2) j Reactor Power - 12% (10 - 15%) (based on power defect) (0.4) I REFERENCE l FCS TOB, OP-7, 01-RC-3 i K/A 015-000-K5.05 I.R. 4.1/4.4 I K/A 001-000-KS.05 I.R. 3.5/3.9 K/A 001-010-KS.36 I.R. 3.4/3.7 i i i
I-l t , St__IBEQBl_QE_NUGLEeB_EQWEB_ELeNI_QEEBoI198t_ELUIDSt_oND PAGE 14 I IBEBdQQldadICS ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 5.05 (2.00)
- c. ~ 24 Hrs. (accept 20 - 30 hrs.) (0.5)
- b. Time would be shorter. (< 20 hrs.) (0.5)
- c. Due to the time delay in Xenon production from the decay of Iodine.
(1.0) REFERENCE CE EQB K/A 001-000-K5.38 I.R. 3.5/4.1 ANSWER 5.06 (1.50) QUESTION DELETED from examination as result of facility comments. REFERENCE l BASIC 1HERM0 DYNAMICS 1 K/A 011-000-K5.10 I.R. 3.7/4.0 ANSWER 5.07 (3.00)
- a. The core maximum linear power will increase (0.5). Total core power does not change. The linear power near the dropped CFA will decrease (0.5), but the rest of the core will be producing a higher linear power (0.5).
- b. Xenon will build in near the inserted C~A due to burnout decreasing (0.5). This will further supp*ess power generation in this area (0.5 and cause the maximum linear power to increase elsewhere in the core (0.5).
REFERENCE Basic Reactor Theory K/A 001-000-K5.04 1.R. 4.3/4.7 K/A 001-000-K5.38 I.R. 3.5/4.1
. . . . . . . _________________1
Ez__PL6MI_gISIEdg_QE11QUt_QQUIBQLt_66Q_lUgIBQUEUISIlQU PAGE 15 ANSWERS -- FT. CALHOUN -86/ll/18-MCCRORY, S. ANSWER 6.01 (3.00)
- a. NO (0.5)
- b. Cold leg break - core cooling is maintained by boiling off inj ected fluids, excess inj ection spills out the break and to the sump. (0.5)
Hot leg break - ECC water enters the cold leg, flows down the annulus 1 through the core (cooling it) and out the break and to the sump.(0.5)
- c. He realigns the discharge of the HPSI pumps to be split equally between hot and cold injection paths.(0.2) The hot leg inj ect ion path is established by cross connecting the discharge of the HPSI pumps with the normal charging header.(0.5) Flow is then directed through the auxiliary spray line into the pressurizer causing flow out the su'rge line into the vessel resulting in reverse flow through the core.(0.8)
REFERENCE FCS SD I-8, I-g K/A 002-000-Kl.08 1.R. 4.5/4.6 ANSWER 6.02 (3.00)
- c. Prevents reactor operations when DNBR is LT minimum design.
- b. Prevents peak local power frcm damaging core (KW/ft fuel centerline melt).
- c. Protect fuel cladding against reactivity excursions too rapid to be protected by high pressure or TM/LP (CEA ej ection),
- d. Uncontrolled CEA withdrawal or boron dilution incidert during startup or very low power levels.
- e. Prevent excessive blowdown of RCS by a PORV or safety valve opening by pr eventing the reactor from generating more heat than can be removed by their steam generators. OR RCS overpressuriz ation.
- f. Ensures the reactor is tripped on conditions which require safety inj ec t ie n.
- g. Lor.s of feedwater accident, assures RCS pressure does not exceed design.
- h. Protects against excessively high steam flow caused by a maj or steam leak. (Protects against effects of any steam line break.)
- 1. DNB core protection on a sudden flow decrease.
J. Permits their operator to trip when the reactor should be tripped prior to forcing automatic action. (0.3 each)
ht__ELaNI_SYSIEUS_DESIGUt_QQUIBQLt_8UQ_lUSIBudEUI611QU PAGE 16 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. REFERENCE FCS SD II-5 K/A 012-000-K4.02 I.R. 3.9/4.0 ANSWER 6.03 (3.00) SEE DRAWING REFERENCE FCS SD I-9, P&ID E-2S866-210-130 K/A 002-000-K1.08 I.R. 4.5/4.6 ANSWER 6.04 (3.00) SEE FIGURE 3.2 (0.1 pt for ESAS, 0.1pt for condition - 15 components) REFERENCE FCS SD II-7, P& ids E-23866-210-120,121 K/A 013-000-K1.11 1.R. 3.3/3.8 ANSWER 6.05 (2.00)
- a. Control room indication is provided by differential pressure transmitter (LI 197) between a low pressurizer tap and a hot leg.
(1.0)
- b. 1. Shutdown cooling pressure interlock. (0.5)
- 2. Low temperature overpressure protection. (0.5)
REFERENCE FCS SD I-4 K/A 002-000-K4.10 I.R. 4.2/4.4
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w, Ez__ELeNI_SYSIEUS_DESIQut_CQUIBQLt_6NQ_INSIBubENIol1QN PAGE 17 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 6.06 (2.00) i
- a. The function of the CSS is to prevent exceeding the design pressure of the containment (0.7S) following a large break LOCA or main steam line break in containment (0.25).
- b. Upon receipt,of a RAS, the CSS pump suction valves shift to take a suction on the containment sump (0.5). At the same time the CCW valves to the Shutdown Cooling Heat Exchangers open to allow CCW to cool the recirculated sump water (0.5). (The LPSI pumps automatically stop on RAS.)
REFERENCE FCS SD I-10-4 K/A 026-000-K1.01 I.R. 4.1/4.1 K/A 026-000-K4.01 1.R. 4.2/4.3 4 i 1
- , , ._,_.....m __ . - - , _ . _ , _ - . . . . _ , _ - _ , _____--.,_-.m_,.._ __--....,____..,.m. _ _ . _ _ _ _ . . _ _ _ , . . . , _ _ , _ _ _ - _ , , , _ . _ _ . . -, -
Zz__EBQQEQVBES_:_NQBdaL&_aHUQBueLt_EUEBQENQ1_8UQ PAGE 18 B6DIQLQQ1CeL_CQUIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 7.01 (2.00) ,
- c. false
- b. AOP-6 governs if plant safety is at risk, otherwise, S.G. 0-28 governs.
- c. 1. Aux building controlled area (0.5) (" controlled" must be included for full credit)
- 2. Containment (0.5)
REFERENCE FCS A0P-6, S.O. 0-28 s K/A 000-067-EPEG10 I.R. 3.9/4.3 K/A PWG19 I.R. 3.4/4.2 ANSWER 7.02 (2.50)
- a. 6. RCS
- b. 7. CVCS
- c. 2. ECCS
. d. 4. ELEC SYS l e. 1. CORE, & 6. RCS 1 (0.5 EACH) REFERENCE FCS T.S. 2.1.4, 2.2, 2.3, 2.7, 2.10 K/A SWG8 I.R. 3.5/4.5 1 l l
Zt__EBQCEDUBES_:_UQBd8Lt_aBUQBdekt_EUEBGENC1_ouD PAGE 19 88010LQQ1 cal _CQUIBQL ANSWERS -- FT. CALHOUN -86/ll/18-MCCRORY, S. ANSWER 7.03 (3.00) (ANY 3, VERBATIM RESPONSE NOT REQUIRED)
- 1. Establish charging and letdown.
- 2. Establish a flow path through the SDC suction line - containment spray pumps - HPSI pumps - loop inj ect ion valves.
- 3. If fuel transfer canal is open and refueling pit is flooded, align fuel transfer canal drain pumps to take suction from the SIRWT to SFP and align storage pool pumps to take suction from the SFP to the SIRWT. (Ensure suction is not taken from the transfer canal drain.)
- 4. As a last resort, establish a flow path from the SIRWT, through the HPSI pumps, into the primary system, and out of the loop drains to the RCDT. Align the RCDT pumps to the SIRWT.
(Give full credit for any path which can be verified using P& ids and which does not use a LPSI pump as the prime mover.) (1.0 EACH) REFERENCE FCS A0P-19 K/A 005-000-Kl.06 I.R. 3.5/3.6 K/A 005-000-Kl.ll I.R. 3.5/3.6 t i r i l i
- , _ - - - - . _-- . ._- . _.._- - __.-,,- ---_- , --__,.~.- _ ._ _ -
Zz__EBQQEQUBER_:_UQB06Lt_eBNQBuett_EdEBGENC1_oNR PAGE 20 88Q10LQQ108L_CQUIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 7.04 (3.00)
- n. 1. reactor power decreasing
- 2. negative SUR
- 3. no more than 1 regulating or shutdown CEA NOT inserted
- b. 1. Pzr level 30-70% and trending to 45-60%
- 2. RCS subcooling >20 deg. F
- c. 1. at least one RCP running
- 2. core delta T <10 de. F
- 3. RCS subcooling >20 deg. F
- d. 1. RCS Tc 525-535 deg. F
- 2. S/G pressure 850-925 psia
- e. 1. no unexplained increase in containment sump level
- 2. no containment area radiation monitor alarms
- 3. no atmospheric or stack radiation monitor alarms
- 4. no S/G blowdown or condenser off-gas radiation monitor alarms
- 5. containment pressure <3.0 psig (Each lettered response is 0.6 pt, this value is divided equally between the subparts for the lettered response.)
REFERENCE FCS E0P-1 K/A 000-007-EPEG11 I.R. 4.4/4.5 K/A 000-029-EK3.01 I.R. 4.2/4.5 ANSWER 7.05 (2.50)
- a. false
- b. true
- c. true
- d. false
- e. false (0.5 EA)
REFERENCE FCS RPM VII-2, VII-7 + K/A PWG15 I.R. 3.4/3.9 K/A PWG16 1.R. 3.4/3.7
Zt__EBQQEQUBES_:_NQBdakt_aHUQBuaLt_EMEBGENQ1_3NQ PAGE 21 BaQ10LQQ1 cal _CQNIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 7.06 (2.50)
- a. boron is being diluted
- b. 1/M plot or low power physics testing
- c. the point of adding heat (1%) (normal temp., press., etc. 1/2 credit)
- d. 1 DPM
- e. below the rod insertion limits (POIL)
(0.5 each) REFERENCE FCS OP-7 . K/A 001-000-K5.18 I.R. 4.2/4.3 K/A 001-010-K5.16 I.R. 2.9/3.5 K/A SWG7 I.R. 3.5/4.0 i 1 i i t 1 i !i
81__aQd1NISIB611YE_EBQQEQuBESt_QQUQII1QUSt_600_LidII611QUS PAGE 22 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 8.01 (1.00)
- 1. CEA insertion limits (PDIL)
- 2. Axial power distribution limits (APD)
- 3. Tc
; 4. Pzr Press
- 5. RCS flow
- 6. ASI
- 7. Peaking factors
- 8. Quadrant tilt
- 9. TM/LP trip (Pvar)
- 10. Asymmetric S/G transient trip (any 2 at 0.5 ea)
REFERENCE FCS TS pgs 1-1,2 K/A SWG5 I.R. 3.9 K/A 001-000-K5.08 I.R. 4.4 X/A 001-000-K5.06 I.R. 4.1 4 ANSWER 8.02 (3.00)
- a. To limit the (2 hour) dose at the site boundary (0.5) following a S/G tube rupture (0.5).
- b. To accommodate the possible iodine spiking which may accompany changes in thermal power. (1.0)
- c. Prevents a release by keeping saturation temperature below setpoint for S/G atmospheric relief valves. (1.0) i REFERENCE j FCS TS 2.1.3 Bases K/A 000-076-EPEG5 I.R. 3.6*
l K/A 000-076-EPEG7 I.R. 3.4 l ! ANSWER 8.03 (1.50) l
- a. procedure (verbal or written) i b. NRC 1 c. Manager-FCS i (0.5 each) l REFERENCE l FCS SO G-10 X/A SWG2 1.R. 3.0
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az__6Dd1N1118611VE_EBQQEQVBEft_QQUDIIIQU$t_6NQ_LidlI611QU$ PAGE 23 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 8.04 (3.00)
- a. 1. fuel clad
- 2. RCS pressure boundary
- 3. Containment
- b. ALERT - 1/3 barriers failed / challenged SAE - 2/3 barriers failed / challenged GE - 1/3 barriers failed / challenged
- c. 1. long term radioactive release
- 2. plant safety (affected by contamination)
- 3. public safety (affected by plant degradation)
(ALL ans, 0.333 each) REFERENCE FCS EPIP-OSC-1 REV 9 K/A PWG36 I.R. 4.7 ANSWER 8.05 (2.00)
- 1. further plant degradation (worsening of plant conditions)
- 2. change in Emergency Classif ication (including initial classification)
- 3. termination of Emergency Classification
- 4. evaluation / assessment results
- 5. protective / mitigator actions taken
- 6. information related to plant behavior that is not understood (any 4 at 0.5 each) 4 REFERENCE FCS SO R-11 2.6 l K/A SWG3 1.R. 3.8 I
ANSWER 8.06 (1.00)
- 1. To provide adequate power tilt detection. (1.0)
- 2. Longterm operation in 2/3 logic configuration should be avoided to maintain separation criteria. (alternate answer) 3
! REFERCNCE f FCS TS DASES'2.15 l K/A 012-000-SG5 I.R. 4.0 l
1 1 St__8DblNISIB611YE_EBQQEQVBESt_GQ001I10US&_800_L1d116I1003 PAGE 24 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. t ANSWER 8.07 (1.00) Ensure a minimum of 7 days of operation (0.5) if only one EOG is operable l I 1 (0.5). REFERENCE FCS TS 2.7 BASES X/A 062-000-SGS I.R. 4.0 l
- l
! ANSWER 8.08 (1.00) 1 I 8 hours (1.0) REFERENCE
- FCS TS 2.5 BASES K/A 000-040-EPEG5 I.R. 4.1 l
ANSWER 8.09 (2.50) ! a. 0.3 gpm (0.75)
- b. No (0.25), because it is more conservative than the approved i
limit (0.25).
- c. T.S. compliance is not maintained (0.5) because T.S. require compliance with facility established procedures (through the 1
commitment to ANSI 18.7-1972 section 5.11 (0.75). REFERENCE '
- FCS SPEC. ORD. NO. 35, TS 5.8, ANSI 18.7-1972 K/A 000-037-EPEGS 1.R. 4.0 X/A 000-037-EPEG8 I.R. 4.3
)
h U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _EIz_CoLBQUN_____________ REACTOR TYPE: _EWB-CE__________________ DATE ADMINISTERED:_Q6411/1B________________ EXAMINER: _dQQBQBXt_St_____________ CANDIDATE: _________________________ a INSIBUCIIQUS_IQ_CoNQ1DoIE1 Rood the attached instruction page co efully. This examination replaces 4 the current cycle facility administered requalification examination. 4 Rotraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. Points for each question are indicated in
, perentheses 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 four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__VoLUE. _IQIoL ___SQQBE___ _VoLUE__ ______________CaIEQQBl_____________ 15t50__ 2dt22 ___________ ________ l. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _16t00__ _252Q0 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _16t50__ _25t20 ___________ ________ 3. INSTRUMENTS AND CONTROLS _16A00__ _25100 ___________ ________ 4. PROCEDURES _ NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _64tDQ__ ___________ Totale Final Grade All work done on this examination is my own. I have neither given
! nor received aid, i ___________________________________
i Candidate's Signature 1
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Ouring the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil SD1y to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of gggb section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __ " as sporopriate, start each category on a Dgg page, write SDAY SD 202 Eldt of the paper, and write "Lart Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least ibtag lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Une abbreviations only if they are commonly used in facility 111gtg1gtg.
- 13. The point value for each question is indic ated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER DLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the g59T10gt only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done ofter the examination has been completed.
- 18. When you complete your examination, you shall
- c. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam 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 defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
t w y+m~-.- -
---n ~-g-yg-g -- ,~vyw--w w-,.n- --
it__EBINCIELER_QE_NUCLEeB_EQWEB_EL8NI_QEEBoIIQNi PAGE 2 IUEBdQQIN80101t_UE6I_IB8NSEEB_8NQ_ELu1D_ELQW QUESTION 1.01 (1.50) The reactor is at power when a single RCP fails. Describe HOW and WHY the following parameters change. Assume NO reactor trip and constant steam d:mond.
- c. Margin to DNB (DNBR) (0.75)
- b. Core delta T (0.75)
QUESTION 1.02 (1.50) The heat output of the reactor con continue to cause the fuel to heatup following a reactor shutdown. WHAT is(ore) the source (s) of this thermal power, and WHY does it NOT indicate on the nuclear instrumentation? (1.5) QUESTION 1.03 (1.00) y Why is the required shutdown margin for modes 1-4 greater than that for
; mode 57 (1.0)
QUESTIONa 1.04 (1.00) Pressurizer pressure is 2200 psia: RCS Tavg is 500 degrees F; and core dolta T is 40 degrees F. The RCS is ___ degrees subcooled. Choose the correct answer below. Show calculations for full credit. (1.0)
- c. 29.45
- b. 40.45
- c. 60.45
- d. 89.45 i
I i (***** CATEGORY 01 CON 11NUE0 ON NEXT PAGF *****) ,.
It__EBINQIELES_QE_NVQLE68_EQWEB_EL8NI_QEEB8IIQNt PAGE 3 IBEBdQQ1Ned10St_BE8I_IBoNSEEB_600_ELV10_ELQW QUESTION 1.05 (3.00) The RCS is at 2100 psia, 532 deg. F (Mode 3), and Keff = 0.997 rods at 75" on group 3. Core depletion is 8000 MWD /MTV and boron concentration is 750 ppm. Ignore xenon and samarium concentrations unless otherwise stated. Graphs have been provided to assist with calculations but it may not be nocessary to use each graph.
- c. How much rod motion will be required to bring the reactor j ust critical? Show all calculations. (1.0)
- b. How much additional rod motion will be required to establish a 1 DPM startup rate after the reactor is made critical? Show all calculations. (1.0)
- c. If the 1 DPM startup rate is allowed to continue until power enters the power range, what will be the final values of RCS temperature, pressure, and reactor power? Assume N0 further rod motion (manual or automatic) and all other systems are operated as required for a normal reactor startup. (1.0)
QUESTION 1.06 (1.00) Consider the operating characteristics of a centrifugal pump and explain c benefit of having the HPSI pumps arranged in parallel rather than in cortes. (1.0) QUESTION 1.07 (2.00)
- c. Curve A on figure 2-10 is a plot of RCS temperature and pressure following a reactor trip. What transient in indicated by curve A and what is the principle concern if the RCS is subjected to thin transtant? (1.0)
- b. As the plant operates through several core cycles, curve O shifts.
Why does curve 0 shift and which direction does it shift? (1.0) L (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
It__EBINQ1ELES_QE_NUQLE8B_EQWEB_EL6HI_QEEBoIl0Hz PAGE 4 IBEBdQDYNed1GSt_BE8I_IB6NSEEB_8ND_ELu1D_ELQW QUESTION 1.08 (2.50)
- o. Where in the reactor core is nucleate boiling most likely to begin?
'Why does it start there rather than elsewhere? (Assume ideal fuel distribution and no clad surface defects.) (1.5)
- b. What is the impact of going to nucleate boiling from the non boiling condition? (1.0)
QUESTION 1.09 (2.00)
- c. If a trip from 100% power occurs with xenon at equilibrium, what in the approximate time interval after trip that xenon will again pass through the 100% equilibrium value? (0.5)
- b. How would this approximate time interval compare if the trip occurred from 50% equilibrium conditiona? (0.5)
- c. When a reactor is returned to 100% power from peak xenon conditions, why does xenon reactivity undershoot the 100% equilibrium value?(1.0)
I i j (***** END OF CATC00RY 01 *****)
2t__EL8NI_ DESIGN _INCLUDINQ_S8EEI1_8ND_EMEBGENQ1_SYSIEd3 PAGE 5 QUESTION 2.01 (2.00) Fill in the blanks in the following precautions taken from 01-DG-1. Blanks mcy contain one or more words or numbers. (2.0)
- o. In an EXTREME EMERGENCY the engine can be shutdown by pulling _____.
- b. Damage to the turbocharger may result if a diesel is started or shutdown without _____.
- c. Placing the Master Emergency Switch 183/MES at AI-133A/B out of the
" Normal" position will _____.
- d. Do not load the diesel until _____ temperature is greater than 120 degrees F.
- c. After unloading and prior to shutdown, allow the diesel generator to run at _____.
QUESTION 2.02 (3.00) 4 Technical Specification 2.2 on the Chemical and Volume Control System requires that there be at least one flow path to the core for boric acid injection whenever fuel is in the reactor. Describe the four possible flow poths which meet this requirement per Tech Specs. (3.0) i QUESTION 2.03 (3.00) I Sketch a one line diagram of ONE TRAIN of safety inj ect ion. Identify all potential water sources and all introsystem cross connects. Include all , pumps, remote /outomatically operated valves (volve numbers or names are NOT required), tanks, and heat exchangers. Pressure, temperature, and flow dotectors are NOT required. (3.0) QUESTION 2.04 (2.50) What five parameters are used by the Engineered Safeguards Control System to generate actuation signals? (2.5) 1 (***** CATE00RY 02 CONTINUED ON NEXT PAGE *****)
2t__EL6NI_QESIGN_INCLUQ1NQ 18EEI1_6ND_EMEBQENCY_11SIEd3 PAGE 6 QUESTION 2.05 (2.50)
- o. WHAT and HOW MANY major components of the CCW system are required to handle a Design Base Accident? 00 NOT consider valves, tanks, instruments or detectors. (1.0)
- b. What loads continue to be cooled by the CCWS following a steam line break inside containment prior to RAS? (1.5)
! QUESTION 2.06 (1.00) Why does the " Compressed Air System - Normal Operation Operating Instructions," 01-CA-1, require that cooling water outlet valves be throttled down? (1.0) 1 QUESTION 2.07 (2.00) ! o. What is the design function of the Containment Spray System? (1.0)
- b. How does the CSS continue to perform its function after receipt of a RAS? Include the automatic responses of support / interface systems essential to the proper functioning of the CSS. (1.0)
J 1 i I (***** END Or CATEGORY 02 *****)
2x__INSIBudENIS_6NQ_QQUIBQLS PAGE 7 QUESTION 3.01 (3.00)
- c. What two fluid systems (not necessarily liquid) are essential to operation of the Turbine Driven Auxiliary Feed Pump? (1.0)
- b. How may the Turbine Driven AFW pump be operated if all air and electrical power, including DC, are lost? (1.0)
- c. What four parameters can be monitored in the Control Room to provide indication of adeounte suction for a Motor Driven AFW pump? (1.0)
QUESTION 3.02 (2.50)
- c. L is t three reactor trips which will NOT initiate a trip on a loss of analog signal? (1.5)
- b. WHAT are the effect(s), if any, on the protection system if a logic matrix trip relay fails to trip? WHY? (1.0)
QUESTION 3.03 (2.00)
- o. What provides indication of RCS level / volume in the control room when the system is drained below the pressurizer level taps? (1.0)
- b. Aside from indication, what are two specific uses of the signals generated by the low range pressurizer pressure detector? (1.0)
QUESTION 3.04 (3.00) List four conditions which automatically reduce letdown flow. For each condition, give the component (usually a control valve) that actually operakas to reduce flow. (3.0) QUESTION 3.05 (3.00) Figure 3.2 tu a simpitfied drawing of the CVCS. On figure 3.2 indicate all components which receive Engineered Safeguard Actuation Signals. For each component, indicate the type of ESAS (CSAS, V!AS, etc.) and the final condition of the component (open-closed, on-off). (3.0) (***** CATEGORY 03 CONTINUED ON NEXT PAGF *****)
2t__INSIBudENIS_6NQ_GQNIBQL3 PAGE 8 QUESTION 3.06 (3.00) List all the automatic control functions generated by the process radiation conitoring system. Include the type of radioactivity (particulate, gtseous, etc.), the system monitored, and the intended system responses to the control signal (valves ___ and ___ open/close, CIAS, etc.). For cignals which generate an engineered safeguards actuation signal, it is nocessary only to indicate the type of ESAS. (3.0) (***** END OF CATEGORY 03 *****)
l l di__EBQGEQUBEH_:_NQBdeL&_8ENQBd8Lt_EUEBQENGl_80Q PAGE 9 l 88Q10LQQ1G6L_GQUIBQL l i i 1 QUESTION 4.01 (1.00)
- c. True or false. A person discovering a fire should immediately attempt to extinguish the fire if it is small and then notify the control room. (0.5)
- b. What determines whether S.O. G-28 (Station Fire Protection Plan) or AOP-6 (Emergency Fire Procedure) is the governing procedure for fighting a fire? (0.5)
QUESTION 4.02 (1.00) A single channel of instrument bus power is lost and no reactor trip occurs. WHAT operator action is required prior to attempting to ro-energize the bun and HOW does this action ef f ect reactor protection? (1.0) QUESTION 4.03 (3.00) What are three ALTERNATE means of cooling the core if Shutdown Cooling is lost and cannot be restored while the reactor vessel head is removed. Include a brief description of the flow path for each case. (3.0) QUESTION 4.04 (3.00) List the means used to verify each of the following conditions as per E0P-1 (Reactor Trip). The number in parentheses indicates the required number of items per condition. Setpoints and parameter values are required for full credit if they are stipulated in the E0P. (3.0)
- c. Reactor tripped / shutdown (3)
- b. RCS inventory control (2)
- c. Forced circulation heat removal (3)
- d. Steam dump and bypass valves functioning (2) o Normal containment conditions (5)
(***** CATEGORY 04 CONTINVED ON NEXT PAGE *****)
di__EBQQEQUBES_:_UQBd8Lt_8BNQBd8Lt_EdEBGENQ1_86Q PAGE 10 88Q10LQQ1Q8L_QQUIBQL e QUESTION 4.05 (2.50) TRUE OR FALSE
- o. The " Employee ALARA Concern Form" is signed by personnel who have been given RWP pre-planning to satisfy ALARA requirements. (0.5)
- b. A task with a collective dose of less than one man-rem is exempt from documented ALARA pre-planning. (0,5)
- c. The permissible accumulated dose (PAD) to the head, eyes, liver, or large bones is the same as the PAD to the whole body. (0.5)
- d. The protective clothing prescribed on a RWP is intended to reduce the whole body accumulated dose from exposure to external radiation.(0.5)
- o. The shift HP technician may authorize increased RWP exposure limits for back shift work. (0.5)
QUESTION 4.06 (3.00)
- o. List eight (8) direct ALARMS that would immediately indicate a loss o Component Cooling Water. (00 NOT list temperature alarms from systems or components serviced by CCW. These are secondary symptoms.
Redundant components will be counted as one.) (2.0)
- b. If CCW cannot be restored in five minutes, what two specific actions must be performed immediately? (1.0)
(***** CATEGORY 04 CONTINUCD ON NEXT PAGE *****)
di__EBQGEQUBES_:_UQBU6Lt_8000BuaLt_EUEBQCUQ1_6HD PAGE 11 88010LQQ1Q8L_CQUIBQL s QUESTION 4.07 (2.50) Fill in the blank for each of the'following statements derived from the precautions for Reactor Startup (OP-7). The blanks may represent single words or phrases. The answer need NOT be the exact wording found in the procedure. (2.5)
- c. Criticality must be anticipated whenever CEAs are being moved or ___.
- b. If core geometry has been altered from that which existed at the beginning of the cycle, perform a ___.
- c. When increasing power, monitor RCS temperature, pressure, pzr level, and charging rate changes to determine if the reactor is at ___.
- d. Do not exceed a sustained startup rate of ___.
- o. The RCS boron concentration MUST NEVER be at a level which would allow criticality ___.
1 s (***** END OF CATEGORY 04 *****) ( ______ ___ _____?? '?_ ' ' ??? ' '??___ "0 _ AM1NAT1ON **n ***n*n ***)
MtC LICENSE EXMIMTION HANDOUT EQUATIONS. CONSTMTS, AND CONVERSIONS 6 = rii*Cp'*deltaT 6=U*A*deltaT P = Po*10sur*(t) p , po et/T SUR = 26/T T=1*/p+(p-p)/Kp T=1/(g-$) T=($-p)/Xy p = (Keff-1)/Keff = deltakeff/Keff p=1*/TKeff+hfr/(1+XT)
- A = In2/tg = 0.693/tg X = 0.1 seconds-I I = Io'e-ux CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1gajlon=8.345lbm=3.87 1 ft = 7.48 gallons liters Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 8tu/lbm fleat of fusion = 144 8tu/lba 1 atmosphere = 14.7 psia = 29.9 inches Hg.
Miscellaneous Conversions I curie = 3.7 x 10 w disintegrations per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 100 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf
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it__EBIN91ELES_9E_NV9LEeB_E9 WEB _EL6NI_9EE86119dt PAGE 12 IBEBd9u1 nom 19St_UgeI_IgeNSEgg_6NQ_ELVIQ_EL9W ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 1.01 (1.50)
- c. The margin to ONB will decrease (0.25) because Th will increase.(0.5)
- b. Core dT will increase (0.25) because of the following relationship:
Heat transfer rate = mass flow rate X (Th - Tc) (0.25) Since the steam demand is constant (heat transfer rate) and mass flow rate has decreased, core delta T must increase (0.25). REFERENCE HT&FF THEORY, CE CORE CHARACTERISTICS K/A 003-000-K5.01 I.R. 3.3/3.9 ANSWER 1.02 (1.50) This thermal power is mainly from the decay of fission products (0.5). It does not indicate on the nuclear instrumentation because it consists mcinly of alpha, beta, and gamma radiation from the decay of the fission products, which are not detected on the nuclear instrumentation (1.0). REFERENCE Standard Reactor Theory K/A 001-000-K5.37 I.R. 3.6/4.1 ANSWER 1.03 (1.00) In mode 5 the plant is completely cooled down (below 200 deg. F) so that there is no need to have reserve shutdown margin to offset the positive reactivity addition resulting from a rapid cooldown accident (steam line break) while in modes 1-4. (1.0) REFERENCE CE SAFETY ANALYSIS FCS TS 2.10.2 BASES K/A 000-040-EK1.05 I.R. 4.1/4.4 f
11__EBINQ1ELEH_QE_NUQLE68_EQWEB_EL6NI_QEE8611982 PAGE 13 IBEBdQQ186dIQHt_BE6I_IB6NSEEB_6ND_ELUID_ELQW ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 1.04 (1.00) 8.(0.5) Tnat = 649.45, Th 580 + (1/2 x 40) = 600 Tsat - Th = 649.45 - 600 = 49.45 (0.5) REFERENCE BASIC HT8FF K/A 001-000-K5.56 I.R. 4.2/4.6 l l l l t l l l 3
Iz__EBINQ1 ELE 1_QE_NUQLEeB_EQWEB_ELeNI_QEEBoI1QNt PAGE 14 IBEBdQQ1Ned191t_BE6I IBeNSEEB_6NQ_ELMIQ_ELQW ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 1.05 (3.00)
- a. rho = (Keff -1)/Keff (0.25)
= (0.997-1)/0.997
= -0.003/0.997
= -0.003000 (or -0.003)
= -0.3% (0.25)
From figure II.B.2, rod insertion worth at 75" on group 3 ~ 3.0% dp. Therefore, rod insertion worth when j ust critical = 3.0% dp ,'(-0.3% dp) = 3.3. From figure II.B.2, this is about 42" on group 4 or about 42" (accept 40 to 45) of rod motion. (0,5)
- b. rho = beta /(1+1amda T), T =26/SUR =26/1dpm =26, beta =0.5-0.65% (0.25)
= 0.006/(1+(0.1x263)
{
= 0.006/3.6
= 0.00167
= 0.167% (accept 0.139 - 0.18%) (0.25)
From figure 11.8.2, rod position based on a final integrated worth of 3.467 (3.439 - 3.48) dp, rod height will.be about 63" (60 - 65) on group 4. Therefore, the additional rod motion will be 21" (18 - 23). l (0,5)
- c. (At this point in mode 3, the SDBCS would be set to control S/G pressure at about 900 psia (532 deg. F). As reactor power passes the point of adding heat (1%), the RCS will begin to heatup which will raise S/G pressure causing the bypass valves to modulate open thereby reducing temperature and pressure. Equilibrium will be achieved when
, reactivity added by the power defect (figure II-15) equals the excess
- inserted by the rods.) Final conditions will be
RCS temperature - 532 deg. F (controlled by SDBCS) (0.4)
~RCS pressure - 2100 psia (pzr press control system in auto) (0.2)
Reactor Power - 12% (10 - 15%) (based on power defect) (0.4) REFERENCE I FCS TDB, OP-7, OI-RC-3 l
-K/A 015-000-K5.05 1.R. 4.1/4.4 l -K/A 001-000-KS.05 I.R. 3.5/3.9 l K/A 001-010-KS.36 I.R. 3.4/3.7 4
l l \ -. _ - - . - . - - _ _ _ -._-_ - - - -. - .- - - - .-
it__EBINCIELER_QE_NVQLE68_EQWEB_EL6NI_QEEBeIIQNi PAGE 15 IBEBdQQ1Ned1CS&_BE6I_IB6NSEEB_68Q_ELU10_ELQW ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 1.06 (1.00) When centrifugal pumps are arranged in parallel, you get an increase in the volumetric flowrate in the system for each pump that is started. (This ability to increase flowrate may be critical in an emergency situation. The parallel arrangement helps ensure reliability such that the seizure of ! a pump.will not impede flow from the other pumps but this conside' ration is not based on centrifugal pump characteristics.) REFERENCE PUMP LAWS and SAR considerations K/A 006-000-K4.18 I.R. 3.3/3.8
. ANSWER 1.07 (2.00)
- a. The reactor vessel has been subjected to pressurized thermal shock i (0,5) as a result of the rapid cooldown with only a small change in pressure. If the vessel is flawed and sufficiently overstressed, brittle failure could occur. (0.5)
- b. As the vessel is subjected to continuous fast (0.1) neutron flux over repeated core cycles (0.4), the curve shifts to the right. (0,5)
REFERENCE FCS TS 2.1.2 BASES, BASIC FRACTURE MECHANICS K/A 004-000-KS.09 I.R. 3.7/4.2 K/A 000-040-EK1.01 I.R. 4.1/4.4 ANSWER 1.08 (2.50)
- n. Nucleate boiling is most likely to start at the top of the core or fuel assembly (0.5) because this is the hottest part of the core and therefore nearer saturation conditions. (1.0)
- b. At the onset of nucleate boiling the heat transfer rate increases (1.0) (due to the removal of the heat of vaporization and the break up of the laminar layer at the clad surface).
REFERENCE basic thermo and core characteristics K/A 002-020-K5.01 I.R. 3.2/3.6 L
Iz__EBINGIELEH_QE_UUGLE68_EQWEB_EL8HI_QEEBoI1QNt PAGE 16 IBEBdQQYNedIGSt_BE6I_IB8NSEEB_eNQ_ELVIQ_ELQW ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 1.09 (2.00)
- c. ~ 24 Hrs. (accept 20 - 30 hrs.) (0.5)
- b. Time would be shorter. (< 20 hrs.) (0,5)
- c. Due to the time delay in Xenon production from the decay of Iodine.
(1.0) REFERENCE CE EQB K/A 001-000-K5.38 I.R. 3.5/4.1 ' l i I
.21__EL8NI_QE11GN_INGLVQ1NQ 18EEI1_6NQ_EMEBGENQ1_11SIEd3 PAGE 17 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S.
ANSWER 2.01 (2.00)
- e. the manual inj ector control lever out (away-from engine).
- b. the oil circulating pump in operation. Clube oil)
- c. de-energize the 4160 volt busses. Cload shed emergency bus, isolate control circuits from control room and shift DG control to local)
- d. Jacket water. (engine coolant) e.. 900 rpm Cidle) for at least 5 minutes.
(0.4 PT EACH) REFERENCE FCS 0I-DG-1, p. 1 K/A 064-000-K1.03 I.R. 3.6/4.0 i K/A 062/000-K1.02 I.R. 4.1/4.4 ANSWER 2.02 (3.00)
- 1. The boric acid pumps (0.3) can deliver boric acid tank (0.15) contents to the charging pumps (0.3).
- 2. The safety inj ection pumps (0.3) can take suction from the SIRW j tank (0.15) and inj ect into the core (0.3).
! 3. The charging pumps (0.3) can take suction by gravity from the , SIRW tank (0.45). l 4. The charging pumps (0.3) can take suction by gravity from the
- boric acid tanks (0.45).
! REFERENCE l FCS TS 2.2 Basis, p. 2-18, 19 l K/A 004-010-K6.09 I.R. 4.4/4.6 l K/A 004-000-K4.05 I.R. 3.3/3.2 I j ANSWER 2.03 (3.00) i i SEE DRAWING l [ I
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2x__INSIBudEUIS_680_QQUIBQLS PAGE 22 ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S.
ANSWER 3.06 (3.00) e SYSTEM ACTIVITY FORM CONTROL FCN
- 1. Containment Vent part./ gas initiate VIAS discharge duct (RMs 61, 62)
- 2. Steam Generator liquid shuts blowdown blowdown wonitor isolation valves (RMs 54A & B) (HCV-1387A,8 &
HCV-1388A,8)
- 3. Plant stack gas /part/ iodine initiate VIAS (RMs 50, 51, 60)
- 4. Condenser off-gas gas secure turbine (RM 57) extract steam to aux steam sys
- 5. Waste disposal liquid close ROWS liquid (RM 55) effluent overboard control valve
- 6. Weste disposal liquid close RCV-059 aux steam cond (2 phase) return (RM 59)
(0.5 per numbered response or 0.15 per individual item) REFERENCE FCS SO II-9 K/A 073-000-K1.01 I.R. 3.6/3.9 L
di__EBQQEQUBES_:_NQBd8Lt_8BUQBd8Lt_EDEBQENGl_eUQ PAGE 23 86DIQLQQIG6L_GQUIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 4.01 (1.00)
- o. false (0.5)
- b. AOP-6 governs if plant safety is at risk, otherwise, S.G. 0-28 governs. (0.5)
REFERENCE FCS AOP-6, S.O. 0-28 K/A 000-067-EPEG10 1.R. 3.9/4.3 K/A PWG19 I.R. 3.4/4.2 ANSWER 4.02 (1.00) Place the affected channel in bypass (0.5) which shifts the RPS logic to 2/3 (0.5). REFERENCE FCS AOP-16 K/A 012-000-K1.01 I.R. 3.4/3.7 K/A 012-000-K4.01 I.R. 3.7/4.0 ANSWER 4.03 (3.00) (ANY 3, VER8ATIM RESPONSE NOT REQUIRED)
- 1. Establish charging and letdown.
- 2. Establish a flow path through the SOC suction line - containment spray pumps - HPSI pumps - loop injection valves.
- 3. If fuel transfer canal is open and refueling pit is flooded, align fuel transfer canal drain pumps to take suction from the SIRWT to SFP and align storage pool pumps to take suction from the SFP to the SIRWT. (Ensure suction is not taken from the transfer canal drain.)
4 As a last resort, establish a flow path from the SIRWT, through the HPSI pumps, into the primary system, and out of the loop drains to the RCOT. Align the RCDT pumps to the SIRWT. (1.0 EACH) REFERENCE FCS AOP-19 K/A 005-000-K1.06 1.R. 3.5/3.6 K/A 005-000-K1.11 1.R. 3.5/3.6
Az__EBQQEQUBES_:_UQ8d6Lt_6BUQBdakt_EDEBQEUQ1_8UQ PAGE 24 86019LQQ196L_QQUIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 4.04 (3.00)
- c. 1. reactor power decreasing
- 2. negative SUR
- 3. no more than 1 regulating or shutdown CEA NOT inserted
- b. 1. Pzr level 30-70% and trending to 45-60%
- 2. RCS subcooling >20 deg. F
- c. 1. at least one RCP running
- 2. core delta T <10 de. F
- 3. RCS subcooling >20 deg. F
- d. 1. RCS Tc 525-535 deg. F
- 2. S/G pressure 850-925 psia
- o. 1. no unexplained increase in containment sump level
- 2. no containment area radiation monitor alarms
- 3. no atmospheric or stack radiation monitor alarms
- 4. no S/G blowdown or condenser off-gas radiation monitor alarms
- 5. containment pressure <3.0 psig (Each lettered response is 0.5 pt, this value is divided equally between the subparts for the lettered response.)
REFERENCE FCS E0P-1 K/A 000-007-EPEG11 1.R. 4.4/4.5 K/A 000-029-EK3.01 1.R. 4.2/4.5 ANSWER 4.05 (2.50)
- c. false
- b. true
- c. true
- d. falso
- o. false (0.5 EA)
REFERENCE FCS RPM VII-2, VII-7 K/A PWG15 I.R. 3.4/3.9 K/A PWG16 I.R. 3.4/3.7
e at__EBQQEQUBES_:_UQBd8Lt_8BUQBd8Lt_EDEBQENQ1_86D PAGE 25 88Q10LQQ1Q8L_QQUIBQL ANSWERS -- FT. CALHOUN -86/11/18-MCCRORY, S. ANSWER 4.06 (3.00)
- e. 1. CCW pump discharge pressure low
- 2. CCW discharge header low flow
- 3. CCW surge tank level high-low
- 4. CCW pump trip
- 5. AC/RW interface valve secondary solenoid off normal
- 6. CCW from (containment air cooler) coils no flow
- 7. HPSI pump bearing cooler no flow
- 8. LPSI pump bearing cooler no flow
- 9. Spray pump bearing cooler no flow
- 10. CCW from (SIT) leakage cooler low flow
- 11. CCW pump AC-3A/8/C standby start (any 8 at 0.25 each)
- b. 1. Trip the reactor (proceed to E0P-1) (0.5)
- 2. Trip all RCPs (0.5)
REFERENCE FCS AOP-11, OP-10, SD I-7 K/A 008-030-SG10 1.R. 3.9/4.2 K/A 008-030-SG11 1.R. 3.9/4.0 ANSWER 4.07 (2.50)
- a. boron is being diluted
- b. 1/M plot
- c. the point of adding heat (1%)
- d. 1 DPM
- o. below the rod insertion limits (0.5 each)
REFERENCE FCS OP-7 K/A 001-000-K5.18 I.R. 4.2/4.3 K/A 001-010-K5.16 I.R. 2.9/3.5 K/A SWG7 I.R. 3.5/4.0
r Attachment 1 ATTACHMENT Ouestion 5.01 b. It is agreed that boron concentration does not affect the insertion limit because the limit is independent of boron concentration. The fact that the limit is power dependent, and that sufficient boron must be present to operate above the limit, does not explain the relationship between boron concentration and CEA insertion limits. Any reasonable discussion which addresses the independence of boron concentration from the long term insertion limit should be accepted for full credit. Question 5.02 a. Any discussion which identifies Reactor Coolant System temperature and Reactor Coolant System pressure should be accepted for full credit. Question 5.02 b. An uncontrolled heat extraction event is characterized by rapid cooldown and depressurization. Therefore, " uncontrolled heat extraction" should also be accepted as a correct answer. (
Reference:
USAR Sect. 14) Question 5.02 c. Any discussion of neutron exposure, such as fluence and neutron bombardment, should be accepted for full credit, in addition to
" neutron exposure".
Question 5.03. This question could have been misleading since a Tave of 580 0 F is far above Fort Calhoun Station's Tave, and in fact, is much closer to Fort Calhoun's Thot. Had an examinee mistakenly identified 580 F as Thot, he would have identified (c) as the correct answer. Substantial partial credit (e.g., .75 points) should be awarded in the event that the examinee properly obtained a subcooled margin using 5800 F as Thot'
A correct answer to this question requires that the examinee know the relationship between Tave, TH , T, C and Tsat, how to obtain T sat from the steam tables, and temperature which subcooling is based on (TH )* If the examinee's method is correct, but he calculates the wrong valve of subcooling, partial credit should be awarded. Question 5.04. This is a double-jeopardy question contrary to NUREG-1021, ES-202E.13. Part (a) must be answered correctly in order for part (b) to be answered correctly. Part (b) must be answered correctly in order to answer part (c) to be answered correctly. Full credit should be given on part (b) for proper use of calculations regardless of the input from part (a). It would be difficult for an individual to perform any calculations in part (c) if he were unable to work part (b). Therefore, an individual failing to correctly solve part (b) should be given full credit for an accurate discussion of the approach to part (c). Based upon the acceptable reactivity range in part (b) of the answer key, and the power defect curve supplied with the examination, a final reactor power of 8% to 15% for part (c) should be accepted. The examination key prosently allows a reactor power range of 10% to 15%. Question 5.05 c. The answer key only addresses the production side of the xenon equation. An equally acceptable answer is that xenon removal due to the return to 100% power exceeds xenon production due to the decay of iodine. Any discussion that indicates an understanding of xenon production and removal following a power increase should be given full credit. Question 5.06 The curves depicted in Figure 1 are extremely difficult to use to the point of being misleading. No time domain in specified. It is extremoly difficult to determine whether the arrows on the axes indicato increasing or decreasing temperature and pressure. Those omissions are contrary to NUREG-1021, ES-202E.15. which states, in part ".....all necessary information is provided .....". An additional source of confusion is the fact that the curves supplied with the examination purport to show formation of a head bubble with_ no correspondina decrease in_ pressurizer crossure. This is unrealistic. (
Reference:
IE Circular No. 81-10, attached)
r Lack of scales on the curves, and the fact that the initial reactor power level is not specified, allow interpretations to be made which yield correct answers other than those in the key. For example, an explanation that additional charging pumps were started and refilled the pressurizer, subsequently, letdown was adjusted to maintain a constant pressurizer level. Another acceptable explanation is that pressure decreased to a point where HPSI flow started and the HPSI flow just matched the tube rupture flow. These alternate acceptable answers demonstrate that the lack of information about the transient and lack of scales allow a variety of interpretations. Therefore, alternative answers which indicate refill of the pressurizer by charging or safety injection should also be accepted for full credit. Question 6.01 The phrase ".....following a DBA ......" in the answer key should be optional and not required for full credit. The Containment Spray System is designed to control containment pressure regardless of the source of steam that caused the pressure increase. The terms loss of coolant accident, steamline break, or " accident" are equally as valid as the term DBA. Question 6.01 b. Examinees may state that recirculation valves close upon receipt of RAS. Points should not be deducted for this statement. (
Reference:
P & ID E-23866-210-130) Qugstion 6.02 b. The term "....and to the sump...." in the answer key should not be required for full credit. The fact that coolant eventually goes to the sump has no effect on the way core cooling is achieved. Question 6.02 c. The ambiguity of the question could lead to a number of answers since nothing in the Fort Calhoun Station Emergency Operating Procedures require the operator to take action based upon identification of break location. The only action currently required in the Emergency Operating Procedures (EOP-03) is to realign the safety injection in a long term core cooling mode, in accordance with the size of the break an determined by RCS pressure.
Because of these ambiguities, three correct answers are possible: l a) The operator does nothing in accordance with EOP-3 since no determination of break location is possible, nor is any action required even if break location could be determined.; b) The examinee may have interpreted this question as requesting information from System Descriptions which have not been upgraded to reflect the new Emergency Operating Procedures. In this case, the examinee may have responded that the Safety Injection flow to each loop should be balanced in accordance with System Description I-9, paragraph 3.2.2; c) The examinee may have indicated that long-term core cooling should have been established to compensate for boron precipitation in the core. This would have been done per Step 3.40 of EOP-03. Question 6.03 The correct answer for all sections except (d) and (j) are found in the basis for Technical Specification 1.3. The following are the suggested correct answers for these questions:
- a. The thermal margin low pressure trip'is provided to prevent operation when the DNBR 'is less than the minimum design (1.18).
- b. The axial power distribution trip is provided to ensure that excess axial peaking will not cause fuel damage.
- c. The high power level trip is provided to prevent damage to the fuel cladding resulting from reactivity excursions too rapid to be detected by pressure and temperature measurements.
- d. The high rate of change reactor powerrtrip is provided as protection against power excursions below the variable overpower trip setpoint as an anticipatory trip to minimize the effects of any high startup rate event. This trip is not required by the Technical Specifications and, as such, is currently in the RPS for equipment protection purposes only.
s
- e. The high pressurizer pressure trip is provided to prevent Reactor Coolant System overpressurization.
- f. The containment high pressure trip is provided to assure that the reactor is shut down simultaneously with the initiation of the Safety Injection System. ,
- g. Low steam generator water level trip is provided to mitigate the consequences of loss of feedwater acciden't or loss of secondary side heat sink event.
- h. In addition to the answer given, the low steam generator pressure trip setpoint protects against any steam line break event.
- i. The low reactor coolant flow trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly.
Question 6.04 This question is very difficult to interpret. The definition of "one train of safety injection" is unclear and will probably vary from person to person. This is inconsistent with NUREG-1021, ES-202E.15., which states "... ensure that the stipulations and requirements posed in the question are complete and unambiguous...". ES-202E.10.a. states "There should be only one correct answer". The correct answer will vary depending on how the individual interpreted "one train of safety injection." Requiring the examinee to draw the entire sketch is inconsistent with ES-202,E20. Any drawing consistent with P & ID E-23866-210-130, which demonstrates knowledge of the safety injection system, should be accepted. Provided that an understanding is demonstrated, alternate terms should be accepted in place of terms such as, RC sump, redundant HPSI path, and RC loop cold leg. Question 6.05 This question does not state the number of answers required for full credit. Thus, it is an "open-ended" question. NUREG-1021, ES-202E.16. states " Vague, "open-ended" questions should be avoided." It is accurate to say that charging pumps are started by the sequencers. (
Reference:
System Description II-7, page 2-7-7 thru 9) The schematic provided is confusing. Valves are not numbered and the arrangement is much different from P & ID's. The examinee must therefore waste valuable time deciphering this unfamiliar sketch before he can even begin responding to the question. It is much more appropriate to supply marked-up P & ID drawings for this type of question. Question 6.06 a. The question as written does not ask what type of instrument LI-197 is, nor does the question ask where it taps off the system. Therefore, neither of these pieces of information should be required for full credit. Either LI-197, or some suitable description of the instrument, should be accepted for full credit.
r- , i
'N t
.)
1 Question 6.06'b. In addition to low temperature overpressure protection, additional correct answers should be; low temperature PORV circuitry, LTOP, or low temperature PORV setpoints. QuestiEn 7.02 s The question' states "the system / functions, numbered items, may match more than once or~not at all". This is confusing. For example, that statement'could mean that there,are multiple answers for part (a). On the cther hand, that statement'could mean that system / function number one (1) matches more than one of the LCO's. Confusion created by that statementtin the question'kesults in the question being open-ended, inconsistent with NUREG-1021, ES-202G.16. An. additional acceptable answer for part (e) is "6 RCS" (Tech. Spec. 2.1.1) . ' (
, lOuestion 7.03 It is inappropriate for an operator to have to memorize flowpath precisely as stated in an AOP. These procedures are,available for reference should they ever be needed. This being the case, any three independent flowpaths which would provide alternate shutdown cooling
'should be accepted for full credit. The validity of an answer can be verified using procedure OI-SC-3, and P & ID's ll405-M-ll, and E-23866-210-130, sheet 1 of 2, in addition to AOP-19.
Quest [on 7.04 Additional correct answers are "RCS subcooling greater than or equal to 200F" and " Core delta T less than or equal to 10 F". Question 7.06 NUREG-lO21,ES-402A.3. states: "The, candidate should be able to describe generally the objectives and methods used in the normal, off-normal, and Emergency Operating. procedures...". This does not say that the examinee must memorize normal operating procedures. Since memorization is not required, additional correct answers should be accepted. For example, on part (b) low power physics testing (SP-PRCPT-1) which requires 1/M plotting, should lua accepted in lieu of 1/M plot (
Reference:
Cycle 10 Reload Submittal). In part (c), answers such as " Normal temperature, pressure and level" should be accepted. For part (e), PDIL should be accepted.
\
I
Question 8.01 There are far more than two (2) limits which assure that minimum DNBR is not violated. Any two of the following answers should also be accepted for full credit. Per Technical Specification 2.10.4: (1) cold leg temperature (2) pressurizer pressure (3) reactor coolant flow (4) axial shape index (5) peaking factors (F T, y T R, Fxy, FR) (6) azimuthal (quandrant) tilt Per Technical Soecification 1.3 (1) TM/LP trip (Pvar) (2) asymmetric steam generator transient trip Question 8.02 a. It is suggested that "2 hour" not be required for full credit. Question 8.04 c. This question is inappropriate for memorization. The operator would always have EPIP-OSC-1 open before him when classifying an event. This question _is definitely not performance-based. Question 8.05 This question is also inappropriate for memorization. Once again, the operator would always have the procedures open before him whenever a situation occurred which would require notifications of this type.
i t Question 8.06 L
.In addition to the answer given in the answer key, a general basis l for operability limits in Tech. Spec. 2.15 can be found in the SER l for Tech. Spec. Amendment No. 88. That general basis gives i
i separation criteria between the different channels as a reason for preventing long-term operation in a 2-out-of-3 logic configuration. 2-out-of-3 logic configuration would exist if one channel were bypassed. ) ' Question 8.09 The weighting of each part is inappropriate. The most important ' performance-based portion of this question is part (a). A weight of ' l.25 is appropriate. The least important is part (b) and should have
- a weight of .50 point.
a e
- Question 8.09 b.
} The limit was submitted to the NRC by OPPD as part of the commitments I made by OPPD to justify restart of' Fort Calhoun following the tube rupture event. The NRC wrote an SER to allow OPPD to restart. i Therefore, it also is accurate to say that this limit was approved by the NRC depending upon the explanation given by the examinee. The l fact that this limit was specified during NRC negotiations may not be well known by all operators. i Question 8.09 c. The introduction to Question 8.09, and the question in part (c), direct the examinee's attention to Tech. Spec. 2.1.4. Therefore, the ! examinee may assume that the question is concerning compliance.with Tech. Spec. 2.1.4. Thus, if the examinee answered the question of compliance relative to Tech. Spec. 2.1.4, credit should be given for j a "no violation" answer since neither the limit nor the corrective action for exceeding the limit is described in' Tech. Spec. 2.1.4. - Since the key answer is from section 5 of the Tech. Spec.s, there i should have been an effort to redirect the examinee's attention to Section 5. 5 t i Question 1.01 l The governing equation for this question is Q = mc o delta t. Failure of a single reactor coolant pump means that a decreases. ! Therefore, delta T must increase. An increase in delta T implies i both a decrease in T c and an increase in T h because heat extraction remains constant and T will remain constant. Therefore, it should also be cons $h! red a correct answer if the ! examinee stated-that margin to DNBR decreases since delta T ! increases. i. 4
,yv . - = , - - - - _n---n,-.-n,wn _ . - _ _ ___, _.--,--._m-,.- --v- , , - , , , - - - - - - - -
9 Question 1.02 Alpha decay is a very small contributor to decay heat and should not be required for full credit. (
Reference:
Glasstone & Sesonske, Nuclear Reactor Encineering) Ouestion 1.03 This question is not applicable to Fort Calhoun Station and should be deleted. Review of Technical Specification 2.10.2 and Technical Specification definitions will reveal that the required shutdown margin for modes one (1) thru three (3) is greater than or equal to 4.0% delta k/k. The shutdown margin required for mode four (4) (cold j shutdown) is greater than or equal to 3.0% delta k/k. The shutdown margin required for mode five (5) (refueling shutdown) is at least 5% delta k/k with all CEA's withdrawn. The question states that the required shutdown margin for modes (1) thru (4) is greater than that for mode (5). Clearly, this statement is incorrect. Due to the confusion that the question would have caused, the question should be deleted. l Question 1.07 a. Curve A depicts a physically impossible event. On any rapid cooldown, system pressure will be driven by the saturation temperature until repressurization occurs due principally to high pressure safety injection. Therefore, curve A does not depict a PTS event since PTS requires cooldown, depressurization, and repressurization. This is supported by the NRC answer for question 5.02 b. What curve A does depict is a cooldown with minimal depressurization, followed by depressurization. Because of the physical impossibility of the curve, the examinees may have provided i answers which discuss cooldown events such as steam line break or { overcooling. Answers which discuss any overcooling transient are acceptable. The principle concern if the RCS is subjected to this transient is pressurized thermal shock and/or brittle fracture. Question 1.07 b. Curve B is not labeled on Figure 2-1B. It is requested that the points awarded in the answer key be adjusted as follows: Fast neutron flux (or fluence). (.5 points) It should not be necessary to repeat in the answer the statement "through several core cycles" contained in the question for full credit.
r-Question 2.01 This is a procedural question appropriate to Category 4. NUREG-1021, ES-202B.4. states: "The candidate is not expected to have normal procedures committed to memory but should be able to explain reasons, preacutions, and limitations of normal operating procedures". This question requires memorization of a normal procedure. It is ngt an explanation of a caution or limitation. Any reasonable, correct response should be accepted for full credit. Question 2.02 Any flowpath that would provide boric acid injection to the core is acceptable for meeting Technical Specification 2.2 (1). The flow-paths described in the Technical Specification basis are not the only three possible boration flowpaths. They are merely examples. Detailed knowledge of Technical Specification bases is not required of an RO, but of an SRO (
Reference:
NUREG-1021, ES-402 A.4.). In addition, ES-202E.1. states: " Technical Specification questions for Reactor Operators should be conceptual in nature". Memorization of bases goes well beyond conceptual. Since the intent of Category (2) is to examine the operator's knowledge of " plant design", any three independent flowpaths should be accepted for full credit (
Reference:
P & ID-E-23866-210-120, 121, and 130). i
- Ouestion 2.04 l
I Three additional parameters are also ured by the Engineered Safeguards Control system: (1) low voltage on 22/4.16kv or 161/4.16kv . transformers, (2) low voltage on buses lA3 or lA4; or (3) low steam generator water level. The first two parameters generate an off-site power low signal (OPLS) and the third parameter generates an auxiliary feedwater actuation signal (AFWAS). (
Reference:
USAR Section 7.3.2.8 and 7.3.2.9) Full credit should be given for any five (5) answers. Question 2.05 a. Credit should not be deducted if the examinee, in addition to the answers stated in the key, accurately discusses components cooled by the component cooling water system. Examples could include items such as safety injection pump coolers, containment spray pump coolers, and containment cooling / filtering units. , t I
Question 2.05 b. This question could easily be misleading due to the lack of proper punctuation. The examinee could easily be led to believe that the only loads inside of containment are required to be specified. In addition, the examinee should not be penalized for listing component cooling water loads in addition to those listed in the key, provided they are not automatically isolated by a safeguards signal. This is an example of an "open-ended" question, contrary to the guidelines in NUREG-1021, ES -202E.16. Question 2.06 ' Operating instruction OI-CA-1 does not state that "the operating compressor may trip if the valves are not properly throttled". Therefore, this portion of the answer should not be required for point credit. Question 3.01 a. i The term " main steam" or " steam" should be acceptable for full credit in answer (1). The question does not ask for a flowpath but only asks for a fluid system. In addition, " water", "feedwater", or
" auxiliary feedwater" should be accepted as one of the two fluid systems, since without this fluid system the pump would have no water to pump. For answer (2), " lube oil" or " oil" should be accepted for full credit.
Question 3.01 b. The second sentence in the answer key should not be required for full credit. The question asks "How may the .... pump be operated." It does not ask "How is the lube oil pump driven". Examinees are aware that the lube oil pump is driven by the auxiliary feedwater pump. However, the question does not elicit that response. Therefore,
" steam may be manually controlled locally at the pump" is sufficient for full credit.
Question 3.01 c. Additional correct answers exist. The answer key lists pump flow as a single parameter. However, pump flow is actually composed of two separate paramaters: condensate, or suction, flow via flow elements FE-1369 and FE-1368 (see P & ID ll45-M-254) and feedwater, or discharge, flow via flow elements FE-1109 and FE-ll10 (see P & ID 1145-M-253). An examinee who names both of these flows should get credit for two parameters. In addition, steam generator level stable or increasing should be considered correct answers. Steam generator level is monitored in the control room. If auxiliary feedwater pump suction were inadequate, steam generator level would decrease. While this is not a direct indication of adequate suction, it is unquestionably an indication nonetheless. A portion of the answer key is incorrect. Discharge pressure cannot be monitored in the control room (see P & ID 11405-M-253). This being the case, only three parameters should be required for full credit. Question 3.02 b. Full credit should be given for " additional logic matrices" as opposed to "six logic matrices". It should not be necessary to specify the number of logic matrices. Question 3.04 The question could be misleading since it asks for conditions which automatically reduce letdown flow, not necessarily isolate letdown flow. The question should be revised to state "... automatically reduce or isolate letdown flow". The condition column in the answer key lists: " conditions, and the instruments which give those conditions". Identification of the instruments should not be required for full credit since the question asks only for conditions. Two additional correct answers exist for this question: 1) HCV-204 l 1s closed by high letdown pressure (greater than 550 psig) ; 2) HCV-204 is closed by high letdown pressure (greater than 550 psig);
- 3) HCV-204 is closed by high letdown pressure (greater than 190 psig). The proper reference for these additional answers is drawing i
ll405-EM-242/244 (attached). l l
r ',l . ~ Question 3.06 i-Either descriptions, or tag numbers, of valves should be accepted for full credit. For example, RCV-059 is the isolation valve for conden-sate return to the drip and drain tank.
- The answer key needs some correction. Answer (1), rather than j stating " containment vent discharge duct" should state " containment i
atmosphere" and is monitored by RM-050 and RM-051. (
Reference:
Technical Specification Table 2-1) The plant stack is monitored by RM-060 and may be monitored by either RM-050/51 or RM-061/62. l (
Reference:
Technical Specification Table 2-1) Waste disposal may be
- monitored by either RM-055 or RM-055A. (
Reference:
USAR Section i 11.2.3.6) l 1 This is an example of an "open-ended" question, inconsistent with } NUREG-lO21, ES-202E.16. l i
;- Question 4.02 i
- Operators are not required to acmorize AOP or EOP actions other than the immediate actions for EOP-01. This question requires j memorization of an AOP. It is therefore inappropriate.
l Question 4.07
- NUREG-1021, ES-202B.4. states: "The candidate is not expected to i have normal procedures committed to memory ...". This question
- requires memorization of OP-7, a normal operating procedure. The
- operator would always have the procedure open before him when
- performing a reactor startup. Since memorization is required, contrary to NUREG-1021, any correct response should be accepted i regardless of whether it occurs in OP-7. See also comments on SRO j question 7.06.
i The third line of the equation sheet contains at least one variation
- of the Inhour equation which is incorrect. This needs to be
. corrected to avoid confusing the examinees.
l I I i i
. Attachment 2 4 ,- = =m - - _ _ . _ _ - _ _
l Omaha Pubilt Power District 1623 Harney Omaha, Nebraska 68102 402/536 4000 l l December 18, 1986 Docket tio. 50-285 I Mr. Ralph A. Cooley Operating Licensing U.S. fluclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, TX 76011
Dear Mr. Cooley:
The graded licensed operator requalification examinations from flovember 18, 1986, were delivered to Omaha Public Power District on December 15, 1986. The Company has completed review of the graded examinations. The review revealed the presence of addi- ! tional correct answers which are not presently included in the l answer keys. It is requested that the answer key should be l revised to include the correct answers discussed below. Question 5.04a: , The photocopy of the grid for the sequential rod worth curve (Fig. II.B.2) is of poor quality. Consequently, it is difficult or impossible to discern the smallest divisions on the graphs. Due to the virtual disappearance of the smallest divisions, and also due to the broadness of the curve itself, it is unreason-able to expect an examinee to read the curve to the nearest sinall di vi s i on . It is much more reasonable to allow a tolerance of plus or minus two (2) small divisions. Plus or minus two (2) small divisions corresponds to plus or minus five (5) inches. Reading the graph with the assistance of straight edge indicates that 117 inches is an accurate answer for final group three (3) rod position. Therefore, an examinee who responded with an answer within a band of 117 plus or minus five (5) inches (group 3), or 42 plus or minus five (5) inches (group 4), should re-ceive full credit. Question 5.04c: The question states: " Assume . . . all other systems are oper-ated as required for a normal reactor startup." The examinee may have assumed that reactor coolant system temperature would be controlled by the steam dump and bypass system in which case T c would be controlled at 532*F. However, it is equally valid to assume that T c w uld be controlled at it's programmed value by the operator 4sse4 tmpmym ngpggua;opponunay
Mr. Ralph A. Cooley December 18, 1986 Page Two (FC-T-038-86) Question 5.04c (cont.): Using that assumption, a final value of T c between 532*F and 540*F, and following its programmed value, should be accepted for full credit. It is also correct to state that T c would be maintained on program to 536*F, then maintained below 537'F administratively (Ref. 01-RC-1, attached). Question 5.06: The curves supplied with the examination were a great source of confusion for the following reasons: o None of the Y-axes have a time scale. It is impossible to determine if the changes in parameters are rapid, slow, or somewhere in between. o The X-axes, with the exception of pressurizer level, have no scale. It is impossible to determine the amount by which RCS pressure, T h , and T c change during the course of the event. o It is not clear whether RCS pressure, T h , and T g are increasing or decreasing from left to right. Discussions with examinees, instructors, and thermal hydraulics engi-neers revealed tremendous confusion in this area. Indeed, the engineer who ran the attached computer simulation first modeled the transient with Th and T c increasing from left to right. o The apparently rapid and large increase in pressurizer level at point A is not realistic for formation of a head bubble following a steam generator tube rupture. This is particu-larly true in view of the fact that no mention was made of pressurizer spray initiation. I.E. Circular No. 81-10, attached, indicates that increase in pressurizer level fol t lowing bubble formation is induced by commencement of pres-surizer spray. In addition, with loss of offsite power a steam generator tube rupture case was modeled for Fort Calhoun using a bench-marked simulation code. The results, t which are attached, indicate that a large and sudden in-crease in pressurizer level would not be experienced l following head bubble formation. The confusion and inconsistency discussed above were of suffi-cient magnitude to cause great difficuty on the part of the examinees. It is therefore requested that question 5.06 be deleted from the examination. l t
r 1 1 Mr. Ralph L. Cooley December 18, 1986 Page Three (FC-T-038-86) Question 6.01a: The question asks for "the design function" of the containment i spray system, not "the design basis". The design function of j the containment spray system is to limit the increase in con- i tainment pressure following either a main steam line break in containment or a loss of coolant accident. The peak containment pressure is almost identical in both cases. Either steam line break or LOCA should be accepted for full credit. (Ref., USAR l Containment Pressure Analysis, attached) Question 6.030: ! Section 7.2 of the USAR indicates that the low steam generator water level trip exists to trip the reactor in the event that secondary side water inventory is lost or decreases to the trip An answer which reflects an understanding of this l setpoint. l concept, for example, loss of feedwater or loss of feed, should be accepted. A ccpy of the USAR section is attached. Question 7.03: The comments in the November 25, 1986, letter suggested that any three (3) independent flow paths which would provide automatic shutdown cooling should be accepted for full credit. The question did not specify that the only alternate means permis-sible were those stated in A0P-19. All of the SR0 examinations were reviewed to determine which of the answers in fact indi-cated true alternate means of cooling the core if shutdown cooling were lost. P & ID's were then marked up to indicate ; what these flowpaths are. (These marked up P & ID's were provided to Mr. McCrory during our meeting of December 22, ' 1986.) Once again, it is requested that any three (3) alternate flowpaths which would provide cooling of the core with the reactor vessel head removed should be accepted for full credit, provided the flowpaths are described in the amount of detail consistent with a three-part question worth 3.0 points. Question 8.02a: 10 CFR 100 states that the dose limit in question is a two-hour l dose limit. An examinee who responds that the basis for the Technical Specification is to maintain exposure below 10 CFR 100 limits should not also have to describe a two-hour limit since this is redundant information. Mention of two hours should not i be required for full credit.
Mr. Ralph A. Cooley December 18, 1986 Page Four (FC-T-038-86) 1 Ouestion 8.03b: Many individuals and groups are involved in the approval of Technical Specification changes. These include the plant manager, the Plant Review Committee, and the Safety Audit and Review Committee. The question does not ask for the final approval authority, i.e., the NRC. Therefore, any answer which specifies a person or group in the approval chain should be accepted for full credit. (Ref., Standing Order G-33 and Nuclear Production Division Policy / Procedure G-12), attached) 0uestion 8.06: It appears that the comment in the November 25, 1986, letter was not accepted. It is again requested that this comment be , accepted. A general basis for operability limits, and Technical Specification 2.15, is described in the SER for Technical Specification Ammendment No. 88, copy attached. That general
- basis gives separation criteria between the different channels 4
as a reason for preventing long-term operation in a two-out-of-three logic configuration. A two-out-of-three configuration would exist if one channel were bypassed. A discussion of this
- basis should be accepted for full credit.
These comments were provided to you during our meeting of December 22, 1986. We request that you review these comments and regrade at least two of the examinations based on this additional information.
/
, , /' l 's e R. L. Andrews Division Manager , Nuclear Production RLA/JJF:ket Attach. i c: File (FC-T-038-86) i t
Attachnent 3 OPERATOR RE00ALIFICATION TRAINING ACTIONS SINCE DECEMBER 11. 1985
-Training personnel have reviewed Yankee Rowe, Oyster Creek, Maine Yankee, Connecticut Yankee and Indian Point operator
. programs as participants on INP0 Accreditation Team Visits.
No significant weaknesses identified in OPPD program. j
- Training organization changed so that all training l departments report to a single manager.
Imoroved Attendance 9 Special provisions were made to assure Shift Super-j visors attended training. ) - Policy issued which requires all personnel to take all quizzes before annual requal is administered. Provide Performance-Based Trainino Materials Objectives provided to NRC in March, 1986. i - Questions provided to NRC in September, 1986. ) - All lesson plans initially developed but review showed
- upgrade necessary.
Operators heavily involved in review of task analysis i and objectives. Upgrade of theory and system lesson plans will be i completed by April 1987, i Upgrade of procedure and Technical Specification lesson plans will be completed by March 1987. i Address the Need for Imoroved OJT Trainina Program completed in September, 1986. i Implementation to start January, 1987. ?
i 1 Imoroved 1986 Reaualification Trainina Weekly quizzes given. I
- Ass'ured 10 CFR55 and NUREG-1021 subject matter covered , in requal training although scope of coverage was not i adequate to pass NRC exams, 1
l Dedicated preparation time provided prior to annual requali-fication exam. l 4 Inhouse requalifiestion exams given to NRC requalification j participants. i l i i i I i i i I
F IMMEDIATE ACTIONS
- 1) Removed operators who had below acceptable scores from licensed duties.
- 2) Notified affected individuals by letter that they were not to perform licensed duties.
- 3) Constant monitoring of operator morale.
- 4) Ensure adequate staffing without excessive overtime.
~
J 1 CONSULTANT FEEDBACK ON
! NRC EXAMINATIONS i
FINDINGS ON SCOPE AND CONTENT The scope and content of the written examination are consistent with NRC guidelines, with a few exceptions. There were a variety of sub-jects covered, appropriate point values assigned, and no one concept, ! system, or administrative area was over-emphasized so that a specific weakness would cause failure of the examination. However, several questions were close to 20% of the category total value. There was one question that we considered inappropriate. The examination was lengthy, allowing no opportunity to review answers to questions. However, this did not appear to be a major factor in the results. 1 FINDINGS ON NRC GRADING The grading was, for the most part, accurate. However, there was one i question where we believe credit should be given for some of the j answers supplied by the licensees. NRC also factored Fort Calhoun comments into their grading. Nevertheless, additional credit for j answers given should be granted. i FINDINGS ON NRC WALKTHROUGH Documentation of failure is not adequate to justify failure. i RECOMMENDATIONS OPPD should request the NRC to delete one question from the exami-
- nation.
l - OPPD should present evidence to the NRC that the answers supplied to l one question are correct. j OPPD should present evidence to the NRC to give more credit for
- several other questions.
j CONCLUSION 4 f If the NRC agrees to these items, two additional senior operators will pass the examination. 'o
i j CONSULTANT FEEDBACK ON OPPD 4 i REQUAllFICATION EXAMINATIONS ' i
; FINDINGS i
The examination required only 51 responses and 6 hours were allowed for completion. The NRC examination required 75 responses and 4 hours
! were allowed for completion.
i The examination contained an unusually large number of questions that ! required only a definition or listing of items. j - Several examination questions did not require answers that would indi-j cate a depth of knowledge normally associated with a senior operator. 4 RECOMMENDATIONS The scope and content of the Fort Calhoun examinations should be ex- < panded. !
- The Fort Calhoun examinations should be more demanding than the NRC 1 j examination.
The time allowed for completing the examinations should be reduced to j 4 hours. i I i i 4 l I i i i l t l
r CONSULTANT FEE 0BACK ON OPPD TRAINING PROCESS FINDINGS
- It is essential that NRC areas be factored into training program.
Training personnel are concerned about considerable amount of memori-zation during written exainination, especially involving procedures and Technical Specifications. These documents are readily available during plant operations; therefore, they should not be required to memorize them. RECOMMENDATIONS
- OPPD should request the NRC to consider administering "open book" examinations for questions that address procedures and Technical Specifications. - Several CE facilities have had similar experience on the NRC Region IV .
examinations in the past. Both Waterford and Arkansas Nuclear One Unit #2 apparently have made the necessary corrections and recently have achieved a satisfactory record of success. We suggest that one of these facilities be contacted for discussions regarding their ex- ! i periences.
OPPD FINDINGS Role of Reactor Operator (SRO) as Shift Supervisor needs Strengthening Old vs. new licensed R0 and SR0 liigher than normal anxiety based on the perception that no boundaries existed on the NRC range or depth in testing Lack of appropriate performance indicators to monitor training performance and offectiveness by management All candidates exhibited their best effort in preparing for and taking the last NRC administered requalification examination f i 1
OPPD FINDINGS
- OPPD does well on NRC R0 exams, but not on NRC SRO exams.
NRC Exam Results Since June, 1985 (Exclusive of Requal retake) Written SR0 5 Passes 11 Failures RO 9 Passes 1 Failure QLill. SR0 12 Passes 4 Failures R0 10 Passes 0 Failures l
i . i l OPPD FINDINGS i l 1 - OPPD exam focus not consistent with supporting preparation I for NRC exams. OPPO exams need to be consistent with NRC exams per consultant feedback. 1 i OPPD and NRC differ on definition of "Immediate Action". ] OPPD training objectives need to incorporate NRC definition !
; for E0P, A0P, OP and Tech. Spec training.
j - OPPD organization does not have a single department respon- I j sible for Operator Training. OPPD has not sought assistance of other utilities particu- ; larly in Region IV, who have had similar difficulties. I i I t OPPD has not provided*NRC " good", easy to use, examination i " preparation material. Material developed to date was ; developed without adequate OPPD Licensed Operator input. l i This caused our review of material to show the need for j upgrades. l l > 1 j - OPPD weekly requal quizzes are not sufficiently challenging. l
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Attachmert 4 i l _ - .__ . - _ _ _ _._ - - . _ . . . . . . - _ u- 2 = - - - i Omaha Public Power District ; i 1623 Harney Omaha. Nebraska 68102 2247 ' 402/536 4000 ! [ i , January 23, 1987 l ] 4 i i Mr Eric l1. Johnson, Director Division of Reactor Safety and Projects i U. S. fluclear Regulatory Commission Region IV 611 Ryan Plaza Drive. Suite 1000 Arlington, Texas 76011 l I
Reference:
Docket flo. 50-285 i
Dear Mr. Johnson:
I l
SUBJECT:
Fort Calhoun Station Licensed Operator Requalification Program i i On December 11, 1986, the Omaha Public Power District was notified, by tele- i phone, of the results of the flRC license requalification examinations conducted i the week of flovember 8,1986. Of the five Corrpany personnel who took the flRC examination and who hold SRO licenses, four failed the flRC examination. The three Company personnel who took the flRC examination and who hold R0 licenses passed the flRC examination, following this notification, the following actions were taken: (1) The Corrpany removed those who failed the flRC examination from their assigned licensed duties, in accordance with the Fort Calhoun Station licensed operator requalification prooram. (2) The Corrpany employed two consultants (f tr. Paul Collins and f1r. Jerry llolman) to review the Company's requalification examination and the t'RC : requalificat109 examinatlon. (3) fir. John Pellet of your staf f was dispatched to the f ort Calhoun Station . to give a limited scope walk-through to six Company SRO licensed per- ' sonnel who normally perform operator dutles and who had not recently taken an f4RC examinat ion, ffr. Pellet and ftr. Ray llall provided feedhack for the examinations discussed in i item (3) above. Of the six Company personnel examined, five were clear passes i and one was a marginal pass. This individual will be given additional training ; and will not be assigned Shift Supervisor duties untti weaknesses identified are corrected, t p, s a y g I nip iqq wed a dh l qii.91 tipp utuemy Y 4 4 t' h $*f fT,, . m,. v . . , ~ - - - - . - - - - -
I Mr. Eric H. Johnson January 23, 1987 Page Two Company representatives met with you and members of your staff on December 22, 1986. The Company presented a preliminary assessment of the cause of the exami-nation failures, an outline of a corrective action program, and justification for passing two additional Company personnel who took the flRC examination. Sub- ' sequent to this meeting, a member of your staff contacted the Company, indicat-ing that one additional person did pass the flRC examination. Since our meeting on December 22, 1986, Company personnel have visited both Arkansas Power & Light and Louisiana Power & Light to investigate the operator training programs conducted at those utilities. Also, an evaluation of the Fort Calhoun Station licensed operator requalification training program has I been conducted by an independent consulting group. Based on this additional information, the Company now believes the primary causes of the examination failures were the following: (1) Licensed operator requalification training failed to place adequate emphasis on Procedures and Technical Specifications. (2) The Company failed to provide adequate training material to the NRC from which the requalification examination could have boon prepared. (3) The Company's requalification examinations and weekly quizzes are not of sufficient depth and difficulty to adequately prepare Company personnel for the NRC requalification examinations. Because these failures and their primary causes indicate a need for substantial changesinourlicensedoperatorrequalificationtrainingprogram,theCom-pany s corrective action plan has several elements. The major elements of this plan are the following: Traininn Department Ornanization (1) The Manager - Administrative & Training Services assumed responsibility for all Nuclear Production Division training in September, 1986. He will continue to increase his involvement in the daily operation of the Training Departments to provide improved management oversight of train-ing. (2) The Supervisor - Station Training has been assigned responsibility for all operator training, including both development and implementation. (3) The Reactor Engineer has been temporarily assigned to the Station Train-ing Department and is responsible for requalification and initial SR0 licensed operator training. The responsibility for initial R0 training has been assigned to another qualified individual. The Operations Train-ing Supervisor and the Licensed Operator Training Coordinator have been removed from line responsibility for licensed operator training.
Mr. Eric H. Johnson January 23, 1987 Page Three Trainino Materials (1) Revised SR0 training program material will be provided to the fiRC before March 30, 1987. (2) This material will be reviewed with members of your staff during the weeks of March 2,1987, in Omaha, and April 6,1987, in Arlington. (3) Revised R0 training program material will be provided to the f4RC at least six weeks before the next initial or requalification R0 exami-nation. Trainino Instruction _an<l Examination (1) The Company's accelerated and normal licensed operator requalification programs will be revised to increase the emphasis on Procedures and Technical Specifications. (2) The Company's requalification examinations will be prepared and graded by an independent consultant until our licensed operator requalification program is found satisfactory. The weekly quizzes given during requall-fication training will also be made more demanding with respect to both scope and depth. In addition to those actions to upgrade our overall operator training program, actions will be taken to prepare individuals for the June,1987 requalification examination to be administered by the f4RC. This program will include the fol-lowing elements: (1) Administration of an audit examination to identify weak areas. (2) Accelerated training in identified weak areas. (3) Frequent quizzes during the accelerated training to assure mastery of learning objectives. (4) Training on written and walk through examination skills. (5) An independently prepared examination to check overall readiness for re-certification. Licensed operatcc s equalification examinations are currently given in flovember and December of each year. The Fort Calhoun Station is on eighteen month fuel cycles, with alternating spring and fall refuelings. This results in the re-qualification examination falling during a refueling outage every third year. To allow the Company to remedy this situation, implement the improved requall-fication traininij program, and better coordinate upcoming initial R0 and SRO examinations, we request permission to delay administration of the 1987 requall-fication examinations until January and February of 1988. Subsequent annual examinations would also be given in January and February of each year.
Mr. Eric H. Johnson January 23, 1987 Page Four Based on the various changes to our requalification program and conversations with members of your staff, the following is our understanding of the NRC exami-nation schedule: (1) March 5. 1912 (a) Oral and written examinations given to SR0 upgrade candidates. (b) Oral examinations given to two Company personnel who are currently restricted from using their valid SR0 licenses. (2) Merk_p LJune 1. 1987 (a) Oral and written examinations given to the three Company personnel who failed the requalification examination and two additional per-sonnel identified by Mr. Cooley, (b) Written examinations given to the two Company individuals who were given walk-throughs in March, 1987. (3) WacA_9LJa_nnatyJQ,J938 (a) Oral and written licensed operator requalification examinations. (b) Oral and written initial R0 examinations. (c) Oral and written initial SR0 examinations. I appreciated the efforts of your staff in helping the Company understand the nature of the problem with our requalification program and their volunteering to meet with us in the future to discuss our proposed solutions. We will con-tinue to interact with your staff in the future to keep them abreast of our pro-gress. I can assure you the Company is making every effort to develop a compre-hensive recovery program and will provide the necessary resources to implement this program. Sincerely, s' R. l. Andrews Olvision Manager Nuclear production RLA/Jmm _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _}}