ML17264A424
| ML17264A424 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/01/1996 |
| From: | Widay J ROCHESTER GAS & ELECTRIC CORP. |
| To: | Dantonio J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML17264A422 | List: |
| References | |
| NUDOCS 9604020157 | |
| Download: ML17264A424 (9) | |
Text
ROCHESTER GAS ANDELECTRIC CORPORA TIOtd ~
'OSEPt! A. WlDAY P'os; Monogamy C ~m> N'wcs~e< Pgt 89EASTAVEIIUE, ROCHESTER, IV. Y.
Y4649-OOOY 1ELEVHONE
/"iC4CCDC Y>C 5'5 '700 March 1, 1996 Mr. Joseph D'Antonio Lead Examiner U.S. Nuclear Regulatory Commission Region 1
475 Allendale Road King of Prussia, PA 19406
Dear Mr. D'Antonio:
Please consider this correspondence as superseding our previous letter dated February 21, 1996'egarding the'ame subject matter.
As required by ROI 95-25, the final grades for the License Candidates that participated in the exam administered at Ginna Station during the week of February 12th are as follows:
Candidate A
Candidate B
- 81. 54
- 77. 174 Comments on exam questions are enclosed with this letter along with material that justifies any changes or deletions made to our grading.
Should you have any concerns, we would be more than willing to discuss the changes via a conference call or meeting at your facility.
As discussed during our telephone conference call, a root cause is being conducted to determine what factors contributed to the failures.
Respectfully, oseph A. Wid Plant Manager Ginna Nuclear Power Plant 9604020157 9h0325 PDR ADOCK 05000244 8
Exam Key Corrections:
Q4 (E000.0004):
The question stem and scenario information set up a
situation that does not meet the definition of an ATWS condition as defined by the Westinghouse Emergency Response Guidelines (ERGs).
Regarding the definition of an
- ATWS, procedure FRS.1 background document states:
"eventually the definition of ATWS evolved into an unspecified common-cause failure (either electrical or mechanical) which precluded control rods from being inserted into the core in response to an anticipated transient (Condition IIevent) which requires reactor trip.
ATWS events are postulated to be initiated from Condition Il transients.
Commonly termed "Anticipated Transients" to distinguish them from the more
- severe, lower probability Condition III and IV transients, they include:
- 1) Uncontrolled RCCA Bank Withdrawal
- 2) Uncontrolled RCCA Bank Misalignment
- 3) Partial Loss of Forced Reactor Coolant Plow
- 4) Loss of Load and /or Turbine Trip
- 5) Loss of Normal Feedwater 6)
Station Blackout (Loss of Offsite Power)
- 7) Accidental RCS Depressurization" In addition, the statement "or conduct a
controlled shutdown" in choice "A" is not consistent with the Ginna Management Philosophy regarding Operator action during an ATWS situation. This philosophy is captured in step 3.6.1.6 of Procedure A-503.1 (Emergency and Abnormal Procedures Users Guide) which states:
"If at any time during a plant transient, a parameter related to reactor or turbine protection is approaching a trip setpoint and cannot be controlled, manual action should be initiated to trip the reactor or turbine as appropriate."
Considering that the question does not set up a
defined ATWS condition and all choices have an incorrect condition or option, this
question does not meet any model outlined in section 5.3.2 of NUREG/BR-0122 (Examiners'andbook).
Question is deleted.
QS (E000.0005)
This question was developed using the generic ERG's rather than the site specific procedures.
The CAUTION/NOTE referenced in the question stem does not apply to Ginna's EOP's and is specifically shown as being deleted per the Documented Step Deviation book.
In it's place Ginna has added a number of steps to sequence the Operators through the process.
This question should not have been asked as Ginna does not test the specific notes and cautions of the ERGs.
Question has been deleted.
Q9 (E000.0009):
Answer "a" is too close to the correct value to be a valid distractor.
Using a calculator, superheated temperature for 775 psia 514.52'F.
Since the crew cannot accurately read tenths of a degree on the Main Control
- Board, and since FIRST indication was stressed within the
- stem, the candidate's choice is logical. Accept "a" and "b" as correct.
Q10 (EOOO. 0010)
The question stem refers to a
specific situation and the answer choices are general in nature.
The correct answer "c"
refers specifically to Step 19, however the Response Not Obtained (RNO) for step 41 of the same procedure also offers the guidance of choice "d". Accept answer "c" or "d" as correct.
Q18 (E000.0021)
Recent procedure change to procedure 0-9 (Shift Relief Turnover-Control Room) added step 5.3.1 (choice "b"). This change was added due to an industry event where this policy was not practiced. It is Ginna Operations Management policy and expectation that this be accomplished with the off going watch. Accept "a" and "b" as correct.
Q28 (E000.0034):
For the power level stated in the question stem there is no correct answer.
Per drawing 33013-1353 both the output of Turbine Power network and the Intermediate Range/Power Range Low Setpoint Trip Block (P-10) feed into a
"1/2" OR gate in the P-7 logic. For Answer "C" to be fully correct, power would have to be less than 8%. Question is deleted.
Q29 (E000.0035)
This question was pointed out by the candidates during the exam as having more than one correct answer.
The question is looking for the the difference between the Single Loop Loss of Flow and the 2
Loop Loss of Flow.
The initial conditions set up a scenario where power is
)Permissive P7 (8%) and ( Permissive P8 (49%).
Looking strictly at the Single Loop Loss of Flow circuit, the correct choice per the key was choice "b". If, however, one were to consider the Two Loop Loss of Flow, the failed RCP would cause a loss of flow signal from the flow transmitters and the RCP breaker.
The trip signal, in this case, is inhibited by power being greater than P7, thus choice "a" is correct.
The question stem does not reference any particular
- trip, therefore the second argument answers the question in that is correct.
Accept "a" and "b" as correct.
Q40 (E000.0046)
The conditions.
in the question stem give limited information which does not rule out choice "b'".
To arrive at the correct choice "c",
the candidate has to discount other indications on the board which would have occurred such as Annunciator H30 (Condensate Bypass Valve OPEN),
the start of the third condensate pump, and fluctuations in condensate header and feed pump suction pressure.
Lacking this information, a feed line leak is also a
possible choice especially if it is located close to the Steam Generator.
The indications in the question would occur if the feed ring were uncovered and steam started to escape from the Steam Generator.
Given that the question asks which of the following COULD result in the given indications and there are no other conditions
- given, choice "b"
is also acceptable.
Accept "b" and, "c" as correct.
Q45 (E000.0051):
Answering this question correctly requires the candidate to memorize the basis for step 6 of ECA-1.1 (Loss of Emergency Coolant Recirculation).
- Further, the basis of step 7
uses the 28%
swap over value as a reference point at which time the SRO makes a
determination for alignment of the safety
- systems, using the procedural guidance.
This level of procedure familiarization is beyond the scope of the RO program and as such is inappropriate for an RO candidate.
This information would be more appropriately tested by asking the basis of Caution 2 of Step 1.
NUREG/BR-0122 states that the examiner should
make sure "that knowledge is not tested for knowledge's sake, nor is an important K/A tested in such a
way that it is far removed from the candidate's use of the R/A in the actual performance of the job."
r Reactor Operators are expected to have the fold out page setpoints memorized, choice "A" is one of the setpoints that they use and is the one that they chose.
Reactor Operators are not expected to have "Monitor" steps memorized, let alone the
- basis, since the SRO would review the step with them during the performance of the procedure.
Question is deleted.
Q48 (EOOO.OOS4):
Given the SI condition in the stem, the EDG's are alread runnin The distractors in "b" and "c" do not match the stated condition and "c" is not true. Question is deleted.
Q68 (E000.0074):
Answer requires a
"judgement call" which would probably not be required of an RO and, in fact, is not clearly spelled out in the procedures.
The given answer, "d", is arguably incorrect because the ability to override the P10 circuit malfunction does exist, therefore the answer which both candidates selected
("b") is the least incorrect answer.
- Again, answer "b"
also reflects the more conservative approach which has been emphasized by Management since the Salem event.
To penalize the candidates for correctly selecting the strategy that Management has felt is in the best interest of safety is counter productive and sends the wrong message.
The correct answer has been changed to "b".
Q74 (E000.0080)
As per the E-2 background documents and the Transient and Accident Analysis text, there is a difference between a feed line and steam line
- break, especially on the primary
- response, until the feed ring is uncovered.
Initial primary response for a feed break is primary temperature increasing until the feed ring is uncovered.
Once the feed ring is uncovered, plant response to the feed and steam line breaks is the same.
Question is deleted.
Q77 (E000.0083)
Per the lesson plan and the Mitigating Core
Q81 (E000.0087):
Damage Student text, either choice "a" or "d" should be considered correct.
The text states that the clad damage is caused initiallyby the Eire-steam reaction.
The word "immediately" in
'choice "a" agrees with the material in the student text which outlines when the reaction will begin
- and, thus, Hydrogen would be generated.
The other condition in choice "a"
regarding the buildup of hydrogen does not indicate how much hydrogen has built up but it is reasonable to assume that if the Zirc steam reaction is producing hydrogen and there is a LOCA in
- progress, hydrogen will begin to buildup in the containment.
Accept "a" or "d".
There is no objective which requires an RO to memorize 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> tech specs, instead at the RO level they are required to recognize Tech Spec components (see objective 1.1 in the Q
84 (E000.0090 package).
Given that all choices are correct tech spec
- choices, the differentiation required to determine the 1
hour tech spec is beyond what would expected of the ROs on the job.
The reasoning for deletion is similar to that of Q45 above.
Question is deleted.
Q84 (E000.0090)
Although choice "a" offers one definition of inoperable per Tech
- Specs, choice "b" also defines a condition that would place the rod inoperable per procedure A-52.4 (Control of Limiting Conditions for Operating Equipment).
Per the procedure items in Section 3.10 of Ginna Custom Tech Specs are to be documented on Attachment 1 and declared inoperable.
Since the question is vague, both answers are acceptable.
Accept "a" or "b".
Q95 (E000.0102):
There is no clear, objective answer to this question based on the background document for Step 1 of E-0. The choice listed as correct is alluded to in the Supplemental Information of the background document but this material is beyond the scope of knowledge for an RO candidate.
The procedural transition and the basis for that transition is the responsibility of the
Question is deleted.
ATTACHMENT 3 NRC RESOLUTION OF FACILITY CONNENTS ON WRITTEN EXANINATION Q4 Not Accepted.
Q5 Accepted.
Q9 Not Accepted.
Q10 Accepted.
Q18 Accepted.
Q28 Accepted.
Q29 Not Accepted.
Q40 Accepted.
Q45 Not Accepted.
This question was discussed with the facility author during exam development.
The training center taught that a trip was required in the given circumstances, although the policy submitted by the facility did not explicitly require this; from a strictly technical specification standpoint the given situation would be interpreted as a failure of both reactor trip logic channels requiring shutdown, but not trip.
Distractor "a" was thus worded to give the candidate both options, either a technical specification shutdown or a manual reactor trip.
Since distractor "a" contained the correct acceptable actions for the conditions in the question stem and the other distractors do not, NRC saw no reason to delete this question.
Delete question due to technical inaccuracy.
Question did not ask candidate to read a meter, "b"
was the only correct answer.
Both "c" and "d" correct depending on where you are in E technical inaccuracy.
Both "a" and "b" correct Delete question due to technical inaccuracy.
Comment withdrawn by facility per telephone conversation, "b" was the only correct answer.
"b" and "c" correct technical imprecision.
Question was technically correct.
Post-exam evaluation that the question was too difficult for an RO was not acceptable by the NRC as a reason to delete.
Q48 Accepted.
Q68 Accepted.
Q74 Accepted Q77 Accepted.
Q81 Not Accepted.
Question deleted due to technical inaccuracy.
Answer key error.
Delete technical inaccuracy.
"a" and "d" correct Question was technically correct, one hour tech specs were identified by facility during exam development as an expected level of knowledge.
Q84 Accept partially. Question deleted, NRC review indicated that there were actually three correct answers.
Q95 Accept partially Question
- remains, two correct answers "a" and "b".
ATTACHMENT 4 SINULATION FACILITY REPORT Facility Docket No:
50-244 Operating Test Administration:
2/13-15, 1996 This form is to be used only to report observations.
These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations.
No licensee action is required in response to these observations.
ITEN DESCRIPTION None
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