ML20214U831
| ML20214U831 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/22/1987 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20214U768 | List: |
| References | |
| NUDOCS 8706110357 | |
| Download: ML20214U831 (7) | |
Text
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y BALTIMORE GAS AND ELECTRIC CHARLES CENTER. P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ObAUTY ASSURANCE & STAFF SERVICES DEPARTMENT fuS."'uIn"EE"mYr" "" **'
April 22, 1987 Mr. Robert M. Keller, Projects Chief l Region 1 U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa.
19406
Dear Mr. Keller:
The attached comments and answer key corrections are in regards to the operator license exam administered on April 21, 1987, at our Baltimore Gas and Electric Company, Calvert Cliffs facility.
They are being formally submitted-to you at the request of your lead examiner, Mr. D. H. Coe.
Also included is the necessary documentation to support each suggested answer key correction.
Should you have any comments or questions concerning our comments or corrections, please contact Mr. J. R. Hill at (301) 260-4955.
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nG Manager-Quality Assurance & Staff Services RMD/JRH/tjg 8706110357 870608 PDR ADOCK 05000317 V
ATTACHMENT 1 RECOMMENDED ANSWER KEY CORRECTIONS
-2.0 PLANT DESIGN,-INCLUDING SAFETY AND EMERGENCY SYSTEMS ~
2.06.a Both units have now been modified so that the response required for unit-2 now applies to-unit-1.
Reference:
FCR-1094, attached 2.08.a The answer key should also reflect the Tech. Spec.
value of 4.75 as a correct response.
Reference:
Table 3.3-4 pg. 3/4-17 of the Tech.
Specs.
l 3.0 INSTRUMENTATION AND CONTROL 3.1.a The question asks the effect of a failure on control rod withdrawal. The answer key requires both' withdrawal and insertion for full credit.
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~We therefore recommend the insertion requirement be deleted from the key for full credit.
3.5.c.The answer key identifies the component as the 4
feedwater LEAD / LAG unit. The candidate may refer to this as the flow error signal which compensates for shrink and swell.
Reference:
RO-103-1-0 pg.10 attached 3.11.a The three inputs to the APD calculator as stated in the answer key are incorrect. The correct inputs are:
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- 1) Linear range NI upper and lower detector
- outputs, i
- 2) positive and negative limits (Yp and Yn),
- 3) CEA function.
Reference:
Same as that listed in key b The wording in the answer key needs to be changed from " power dependent insertion limit" to power dependent limits. These terms are not the same.
Reference:
Same as that listed in key i
3.12.a An alternative answer to this question is:
- 1) Control valves shut,
- 2) stop valves open, and l
- 3) bleeder trip valves open.
Reference:
OI-43A pg.9 step 3
1 4.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.07.b In' addition to those RMS indicators listed in the answer key, the candidate may include containment RMS. The question did not specify inside or outside the containment.
Reference:
EO?-0 pg.8 E.3 4.08.a Condenser vacuum needs to be deleted since this is a special case of EOP-1 and is not reflected in the other EOPs. Additionally EOP-0 is the only EOP-required to be committed to memory.
Reference:
EOP-0 pg.11 b Two safety functions identified in the answer key are not the same terminology as used at CCNPP. The correct terminology is:
- 1) Containment Environment vs. Isolation
- 2) Radiation Levels External to the Containment vs. Radioactivity-Control
Reference:
EOP Safety Function Headings 4.11.c OP-2 ntates that shutdown rods shall be fully withdrawn, this value is 135" not 132" as stated in the answer key.
Reference:
Sys. Description 60, Table 5-7
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ATTACHMENT 2 EXAMINATION COMMENTS
' 1. 0 - PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER,-AND FLUID FLOW l.06.a We believe the answer is technically correct.;A review of our existing reference material does not-identify this specific case, however.~Therefore,'
the candidate would have to make a judgement-without being able to support his conclusion.
This question also requires the RO candidate to recall the diameter of a safety valve, calculate its area, and then compare this to the area that defined-as a small break LOCA. Neither the EOP for LOCA nor the reference document (CEN-152)_
specifically address this case.
Additionally, the reference document used, CEN-152 Rev.3 is a draft document and not a final publication available to the training department.
4.O PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.02 This type of question is not at the RO level.
i Administrative Procedures are. covered in section l.
8.0 of the SRO exam and are not the responsibility of the RO. This type of question is more j
appropriate for the operating _ exam where the candidate has access to the information needed.
Recall of this type of procedure is not necessary i
for plant' safety.
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These examination -concerns are being presented because the training ~
program at CCNPP does not address these responsibilities'as those of a Reactor Operator. Additionally the reference to the KSA catalog in the answer key was not correct for the revision available to the utilities at the time of the exam. It is therefore our contention that these questions be reevaluated for fairness in this and future exams.
During the simulator portion of the exam, a member of the training staff supervision was not allowed to observe. In the future de request that a member of training staff supervision I
be allowed to observe from a remote location.
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RESOLUTION OF COMMENTS ON.R0 EXAMINATION FOR CALVERT. CLIFFS APRIL 21, 1987 2.06a
. Agree, answer key changed to reflect current plant conditions.
-Ref.:
FCR-1094 2.08a Agree, answer key changed from "4 PSIG i.25 PSIG" to "4.75 PSIG".
Ref.: Tech. Specs. Table 3.3-4 3.01.a~
Agree, question was misworded. Any reference to insertion will be deleted and not required for full credit.
Point value changed accordingly.
3.05c
' Agree, will-accept " flow error signal" as correct response.
In addition the following will also be accepted:
flow unit, rate unit, downcomer lead / lag unit (level). Additional reference added to key.
Ref.: As listed in answer key and R0-103-1-0 Page 10 3.11a Partially accepted.
Part a.1 changed to Positive and Negative limits (Yp and Yn).
Part a.2 not changed because reference as listed on answer key lists ASI from NIS as the input to the APD calculator.
3.11b Agree, wording changed as requested.
In addition will also accept "if ASI exceeds the -limiting value on tent curve ' for full credit.
3.12a Not accepted. The question asks for events occuring in the EHC system.
~The alternative answer concerns equipment affected by the EHC system.
No change to answer key made.
4.07.b Agree, Question did not specify inside or outside of containment.
" Containment Monitors" is added as possible correct response.
Point value changed to require any 3 responses 0 0.33 each.
4.08.a Agree, Condenser vacuum deleted from answer key.
Ref..
E0P-0 Page 11 4.08.b Agree, Answer key changed to reflect names of safety functions as stated in body of procedure.
Ref.:
E0P-0 I
O t
2 4.11.c Agree, Answer key changed to reflect proper values.
REF.:
SD-60 Table 5-7 1.06a The facility comment contends that the second part of the question is not required knowledge for a Reactor Operator. The NRC disagrees, believing that the Reactor Operator should be able to demonstrate an understanding of the importance of maintaining natural circulation during SBLOCA conditions by identifying the relatively likely case of a stuck open pressurizer safety valve as a condition during which once-through cooling is NOT sufficient to cool the core.
This objec-tive is operationally oriented and backed by the following NUREG 1122 KA's.
000009EK3.22 Knowledge of the bases or reasons for maintenance of heat sink (during a SBLOCA).
Importance 4.4 cut of 5.0.
000009EA2.39 Ability to determine or interpret adequate core cooling (during a SBLOCA).
Importance 4.3 out of 5.0.
The use of CEN-152 Rev. 3 as a reference was to further support the facility's preface to E0P-5 which documents that certain SBLOCA's require natural circulation to ensure core cooling.
4.02 Not accepted. Part (a) requires the R0, who may be a safety tagger, to know the conditions under which he should not perform independent verification (for personal safety and health) as well as when he may not.
Parts (b) and (c) require the R0 to understand the facility hierarchy of persons authorized to perform safety tagging functions. These knowledges are covered under NUREG 1122 KA 194001K1.02 Knowledge of Tagging and Clearance Procedures (Importance 3.7 out of 5.0).
In response to the facility's request to allow a member of their training staff supervision to observe the simulator evaluation, the NRC will follow the requirements of NUREG 1021 ES-301 paragraph G.
The NRC policy is that only those persons necessary to operate the simulator or provide appropriate simulation role playing will be allowed to be present during simulator evaluations.
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"a Changes Made to Answer Key Made by Grader i
1.01.b Also accept " Fuel Temperature Defect stays the same", if discussion about off setting affects of FTC is made.
1.03.a Deleted reference to FTC having negative affect.
The FTC will more likely have a positive affect if any at all, based on same references used originally.
4.02.c Also accept " Control Room Supervisor" as correct response, since the Control Room Supervisor is always a Senior Safety Tagger.
4.04 Added point breakdown to answer key.
4.10.c Change answer to Read 1st - Steam Generator Level 2nd - CET Temperature or Steam Generator Level To clarify the intended meaning of the answer.
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