IR 05000454/2014301
| ML14212A099 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/30/2014 |
| From: | Hironori Peterson Operations Branch III |
| To: | Pacilio M Exelon Generation Co, Exelon Nuclear |
| R. K. Walton | |
| References | |
| 50-454/OL-14, 50-455/OL-14 | |
| Download: ML14212A099 (17) | |
Text
UNITED STATES uly 30, 2014
SUBJECT:
BYRON STATION, UNITS 1 AND 2 NRC INITIAL LICENSE EXAMINATION REPORT 05000454/2014301; 05000455/2014301
Dear Mr. Pacilio:
On June 19, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed at your Byron Station, Units 1 and 2. The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on June 11, 2014, with Mr. B. Youman and other members of your staff. An exit meeting was conducted by telephone on June 27, 2014, between Mr. T. Chalmers, Plant Manager; other members of your staff; and Mr. R. K. Walton, Chief Operator Licensing Examiner, to review the proposed final grading of the written examination for the license applicants. During the telephone conversation, NRC resolutions of the stations post-examination comments, initially received by the NRC on June 19, 2014, were discussed.
The NRC examiners administered an initial license examination operating test during the weeks of June 2 and June 9, 2014. The written examination was administered by Byron Station Training Department personnel on June 13, 2014. Eight Senior Reactor Operator and four Reactor Operator applicants were administered license examinations. The results of the examinations were finalized on July 11, 2014. Twelve applicants passed all sections of their respective examinations. Eight applicants were issued senior operator licenses and three applicants were issued operator licenses. One Reactor Operator license was withheld pending completion of a medical review by the NRCs medical review officer. Upon completion of the review, the withheld license will be issued if the operators health meets licensing requirements.
During the debrief meeting on June 11, 2014, you requested the written examination documentation be withheld from public disclosure for 24 months. The written examination will be withheld per your request. In accordance with Title 10 of the Code of Federal Regulations, Section 2.390 of the NRC's
"Rules of Practice," a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Hironori Peterson, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66
Enclosures:
1. Initial License Exam Report (ER)
05000454/2014301; 05000455/2014301 w/Attachment: Supplemental Information 2. Simulation Facility Report 3. Written Examination Post-Examination Comment Resolution
REGION III==
Docket Nos: 50-454; 50-455 License Nos: NPF-37; NPF-66 Report No: 05000454/2014301; 05000455/2014301 Licensee: Exelon Generation Company, LLC Facility: Byron Station, Units 1 and 2 Location: Byron, Illinois Dates: May 12, 2014 - June 19, 2014 Inspectors: R. K. Walton, Chief Examiner D. Reeser, Examiner S. Garchow, Examiner Approved by: H. Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000454/2014301; 05000455/2014301; 5/12/2014 - 6/19/2014; Exelon
Generating Co, LLC, Byron Station, Units 1 and 2; Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional U.S. Nuclear Regulatory Commission (NRC) examiners in accordance with the guidance of NUREG-1021,
Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1.
Examination Summary All 12 applicants passed all sections of their respective examinations. Eight applicants were issued senior operator licenses and three applicants were issued operator licenses. One Reactor Operator license was withheld pending completion of a medical review by the NRCs medical review officer.
Upon completion of the review, the withheld license will be issued if the operators health meets licensing requirements. (Section 4OA5.1)
REPORT DETAILS
4OA5 Other Activities
.1 Initial Licensing Examinations
a. Examination Scope
The U.S. Nuclear Regulatory Commission (NRC) examiners and members of the facility licensees staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1, to develop, validate, administer, and grade the written examination and operating test. Members of the facility licensees staff prepared the outline and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of May 12, 2014, with the assistance of members of the facility licensees staff.
During the onsite validation week, the examiners audited three license applications for accuracy. The NRC examiners, with the assistance of members of the facility licensees staff, administered the operating test, consisting of job performance measures (JPMs)and dynamic simulator scenarios, during the period of June 2 - 10, 2014. The facility licensee administered the written examination on June 13, 2014.
b. Findings
- (1) Written Examination The NRC examiners determined that the written examination, as proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Less than 20 percent of the proposed examination questions were determined to be unsatisfactory and required modification or replacement.
On June 19, 2014, the licensee submitted documentation noting that there were two post-examination comments for consideration by the NRC examiners when grading the written examination. The post-examination comments and the NRC resolution for the post-examination comments are included in Enclosure 3 of this report.
The final as-administered examination and answer key (ADAMS Accession Numbers ML14203A290) will be available in 24 months, electronically in the NRC Public Document Room or from the Agencywide Documents Access and Management System (ADAMS). All changes made to the proposed written examination were made in accordance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors, and documented on Form ES-401-9, Written Examination Review Worksheet.
The NRC examiners graded the written examination on June 27, 2014, and conducted a review of each missed question to determine the accuracy and validity of the examination questions.
- (2) Operating Test The NRC examiners determined that the operating test, as originally proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Changes made to the operating test, documented in a document titled, Operating Test Comments, as well as the final as-administered dynamic simulator scenarios and JPMs are available electronically in the NRC Public Document Room or from ADAMS.
The NRC examiners completed operating test grading on June 23, 2014.
- (3) Examination Results Eight applicants at the Senior Reactor Operator (SRO) level and four applicants at the Reactor Operator (RO) level were administered written examinations and operating tests. Twelve applicants passed all portions of their examinations. Eight SRO applicants were issued senior reactor operator licenses and three RO applicants were issued reactor operator licenses. One RO applicants license was withheld pending a medical review of the applicants health conditions.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's implementation of examination security requirements during the examination validation and administration to assure compliance with Title 10 of the Code of Federal Regulations, Section 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG-1021, "Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
b. Findings
No findings were identified.
4OA6 Management Meetings
.1 Debrief
The chief examiner presented the examination team's preliminary observations and findings on June 11, 2014, to Mr. B. Youman, Plant Manager, and other members of the Operations and Training Department staff.
.2 Exit Meeting Summary
The chief examiner conducted an exit meeting on June 27, 2014, with Mr. T. Chalmers, Plant Manager, and other members of your staff by telephone. The NRCs final disposition of the Byron Stations post-examination comments were disclosed and discussed with licensee staff during the telephone exit meeting. The examiners asked the licensee whether any of the material used to develop or administer the examination should be considered proprietary. No proprietary or sensitive information was identified during the examination or debrief/exit meetings.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- B. Youman, Plant Manager
- E. Hernandez, Director Site Operations
- S. Kerr, Director Site Training
- S. Gackstetter, Regulatory Assurance Manager
- J. Ruth, Operations Training Program Specialist
- R. Lawlor, Operating Director (Acting)
- T. Cain, Nuclear Oversight Manager (Acting)
- M. McCue, Operations Training Manager
- M. Coffman, ILT Lead Instructor
- R. Peterson, Exam Author
NRC
- J. McGhee, Senior Resident Inspector, Byron Station
- R. K. Walton, Chief Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
ADAMS Agencywide Document Access and Management System
CV Charging System
ECCS Emergency Core Cooling System
ER Examination Report
LCO Limiting Condition of Operation
LOCA Loss of Coolant Accident
MCR Main Control Room
NRC U.S. Nuclear Regulatory Commission
PARS Publicly Available Records System
RCFC Reactor Containment Fan Coolers
RO Reactor Operator
RWST Refueling Water Storage Tank
SI Safety Injection
SRO Senior Reactor Operator
SX Essential Cooling Water System
TRM Technical Requirements Manual
TS Technical Specification
SIMULATION FACILITY REPORT
Facility Licensee: Byron Station, Units 1 and 2
Facility Docket Nos: 50-454; 50-455;
Operating Tests Administered: June 2 - 10, 2014
The following documents observations made by the NRC examination team during the initial
operator license examination. These observations do not constitute audit or inspection
findings and are not, without further verification and review, indicative of non-compliance with
CFR 55.45(b). These observations do not affect NRC certification or approval of the
simulation facility other than to provide information which may be used in future evaluations. No
licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM DESCRIPTION
Simulator Lock-up During performance of Control Room JPM f, SX Flooding Requiring
RCFC Isolation, the simulator stopped operating. Subsequent trouble
shooting determined that incorrect operator actions and a simulated leak
in the SX system generated a fault condition on the simulator. (See
simulator work request 15414).
WRITTEN EXAMINATION POST-EXAMINATION COMMENT RESOLUTION
SRO Question 91:
When moving irradiated fuel assemblies in the Refueling Cavity, the minimum water level above
the vessel flange can be monitored __(1)__.
The Tech Spec basis for the minimum water level is to __(2)__.
________(1)________ _______(2)_______
A. locally and in the MCR limit iodine fission product release
B. locally and in the MCR provide longer time to core boil
C. ONLY in the MCR limit iodine fission product release
D. ONLY in the MCR provide longer time to core boil
Answer: A
Answer Explanation:
A is CORRECT: TS 3.9.7 requires at least 23 feet above the vessel flange when moving
irradiated fuel. The top of the cavity is 26 feet above the vessel, and the TS basis is to lower
iodine activity.
The local and MCR indicators are both individually used in some surveillances and procedures,
making it plausible that only one exists. Time to core boil is calculated and monitored, but is not
a basis for TS required minimum level.
B is incorrect: Time to core boil is calculated and monitored, but is not a basis for TS required
minimum level.
C is incorrect: The local and MCR indicators are both individually used in some surveillances
and procedures, making it plausible that only one exists.
D is incorrect: The local and MCR indicators are both individually used in some surveillances
and procedures, making it plausible that only one exists. Time to core boil is calculated and
monitored, but is not a basis for TS required minimum level.
Applicants comment:
I selected Answer (B) due to the fact that I had the basis of LCO 3.9.5 - RH and Coolant
Circulation High Water Level in mind. Specifically, Tech Spec Basis (B 3.9.5-3) includes:
With no forced circulation cooling, decay heat removal from the core occurs by
natural convection to the heat sink provided by the water above the core.
A minimum refueling water level of 23 feet above the reactor vessel flange provides an
adequate available heat sink. Suspending any operation that would increase decay heat load,
such as loading a fuel assembly, is a prudent action under this condition. Therefore, actions
shall be taken immediately to suspend loading of irradiated fuel assemblies in the core.
It continues to say on page (B 3.9.5-4):
With the RHR loop requirements not met, the potential exists for the coolant to boil and release
radioactive gas to the containment atmosphere.
Applicants Conclusion:
I believe without asking for the basis of the specific technical specification 3.9.7, the question
left an opportunity to consider a different LCO. Limiting condition for operation 3.9.5 includes
language to capture both iodine fission product release as well as time to core boil, while
LCO 3.9.7 (that the answer key relied on) contains only language specific to iodine fission
product release. Consequently, I believe there are two correct answers, both (A) and (B).
Facility recommendation:
The question was written and validated with the basis of TS 3.9.7, Refueling Cavity Water Level
in mind. However, that TS was not specified in the question, leaving the question open to the
interpretation of the examinee. Technical Specification 3.9.5 does apply with > 23 feet in the
refueling cavity, and the basis for it does discuss adequate level to maintain time to core boil
and decay heat removal. The facility agrees with the applicants comment. The question
should be changed to allow both choices A and B as correct answers.
References:
Technical Specification 3.9.5 and 3.9.7.
NRC Resolution:
The applicant noted in TS B 3.9.5, with one train of RHR system operable in Mode 6, water level
was to be maintained > 23 feet above the top of the reactor vessel flange to prevent the onset of
boiling and resultant reduction of boron concentration that would eventually challenge the
integrity of the fuel cladding. Technical Specification 3.9.5, (RHR and Coolant Circulation -
High Water Level) specified that reactor water level was to be maintained >23 feet to ensure
that an adequate volume of water was available for backup decay heat removal capability.
Additionally, TS 3.9.5 applicability bases stated, The 23 foot water level was selected because
it corresponds to the 23 foot requirement established for fuel movement in LCO 3.9.7.
Technical Specification 3.9.7, (Refueling Cavity Level) states the reason for maintaining reactor
water level > 23 feet above the top of the reactor vessel flange in Mode 6 was to ensure a
sufficient level of water was maintained in the refueling cavity to retain iodine fission product
activity resulting from a fuel handling accident in containment. The basis for maintaining a 23
foot water level in the refueling cavity was both to provide adequate available heat sink during
times of no forced circulation cooling and to aid in retaining iodine from a leaking fuel pin.
The intent of Question 91 was to evaluate the applicants knowledge of parameters to prevent
exceeding design limits associated with the refueling cavity water level. The applicant applied
knowledge from TS 3.9.5 in lieu of knowledge from TS 3.9.7 to answer this question.
Question 91 did not specify RHR system status. As such, the requirements of TS 3.9.5 could
not be ruled out. The question only specified that irradiated fuel was being moved. The
conditions of TS 3.9.7 were applicable for moving irradiated fuel, independent of RHR system
status. Since both TSs 3.9.5 and 3.9.7 specified the same level requirement for different
reasons, the examiners concluded that both A and B distractors be correct answers for the
conditions provided in Question 91.
SRO Question 97:
Unit 1 is at 100% power.
Which valve failure(s) will result in the Emergency Core Cooling System UNABLE to meet its
safety function?
- N.
B. 1SI8812A, RWST to RH pump suction valve, and 1SI8811A, RH pump CNMT sump
suction valve, BOTH failed OPE
- N.
C. 1SI8807A and 1SI8807B, CV and SI pump crosstie valves, failed CLOSE
- D.
D. 1CV8804A and 1SI8804B, RH to CV and SI pump suction valves, failed CLOSED.
Answer: D
Answer Explanation:
A is incorrect: Plausible, in that if the seal injection throttle valves are too far open, injection flow
could be lower than design. There is a surveillance to verify throttle valve position.
B is incorrect: Plausible in that a failure to close this valve would result in RWST draining to the
CNMT sump on recirc actuation. The water is fully available to the RH pumps when in the
sump.
C is incorrect: These are parallel suction header crosstie valves to allow all CV and SI pumps to
be supplied from 1 RH pump in the event of a pump trip. However, other valve alignments will
supply water to the pumps even if they failed. This is evidenced by the existence of 1SI8924, a
single valve in series with these two paralleled valves.
D is CORRECT: The listed valves must be able to be opened for the RH pumps to supply the
CV and SI pumps for cold leg recirculation.
Applicants comment:
I selected Answer (B) due to the fact that I envisioned the RWST draining/drained to the
containment sump based on the valve lineup in this answer choice and believed this
challenged/extinguished the safety function of ECCS.
According to Byron Operating Department Policy Statement (No. 600-04, Rev 25):
The RWST will gravity drain to the containment sump [via an open 1SI8811A/B] at a rate of
7000 gpm/train in 30 minutesto achieve 40 percent RWST level. [as time progresses, the
RWST will eventually empty at the same rate.]
According to the Byron UFSAR (B/B 6.3-31, 32):
Following a small break LOCA, the RWST will deplete much slower than a large break LOCA
since high RCS pressure prevents RHR pump injection and containment spray pump actuation
on HI-3 containment pressure is not expected. Under these conditions, switchover to
recirculation is not time-critical since a relatively large amount of time is available. However, if
the containment spray pumps actuate during the injection mode, the RWST level LO-2 alarm
could be reached in a relatively short period of time, necessitating switchover to the recirculation
mode. If RCS pressure remains above RHR pump shutoff head pressure under these
conditions, only the SI and CV pumps are capable of providing core cooling flow. [but
the RWST volume is inadequate or non-existent]. Therefore, completing the manual actions to
align RHR pump discharge to the suction of the SI and CV pumps becomes time critical since
the SI and CV pumps will lose their suction source unless manual switchover actions are
completed prior to reaching the RWST empty level. [the suction source is already lost
from the RWST.]
According to TS Basis (B 3.5.4-2), with respect to the notion that the 1SI8812A is failed open in
the answer choice I selected, it says:
When the suction for the ECCS and Containment Spray System pumps is transferred to the
containment sump, the RWST flow paths must be isolated [i.e., 1SI8812A closed] to
prevent a release of the containment sump contents to the RWST, which could result in a
release of contaminants to the atmosphere and the eventual loss of suction head for the
ECCS pumps.
It continues to say on page (B 3.5.4-7):
With the RWST inoperable for reasons other than Condition A (e.g., water volume), it must be
restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, neither the ECCS nor the
Containment Spray System can perform its design function. Therefore, prompt action must
be taken to restore the tank to OPERABLE status or to place the unit in a MODE in which the
RWST is not required.
Lastly, the available NPSH for the CV and SI pumps is also challenged. According to the
UFSAR (B/B 6.3-8):
The net positive suction head for the safety injection pumps and the centrifugal charging pumps
is evaluated for both the injection and recirculation modes of operation for the design-basis
accident. The end of the injection mode of operation gives the limiting net positive suction head
available (minimum static head). The net positive suction head available is determined from the
elevation head and vapor pressure of the water in the refueling water storage tank, the tank air
space pressure, and the pressure drop in the suction piping from the tank to the pumps.
First Conclusion:
The stem of the question only said that Unit 1 is at 100% power. It did not say that an SI had
just been actuated or was in progress. I had assumed that the RWST drained itself at least
adequately enough to the containment sump such that the available NPSH for the CV and SI
pumps would be challenged and potentially not have an adequate suction source to deliver flow
for the injection phase of a LOC
if RCS pressure stays high enough, the RH pumps will never deliver water to the RCS until cold
leg recirc is manually aligned, and thus the safety function of ECCS to deliver core cooling at
this time does not exist. The time necessary to manually align for cold leg recirc before the
RWST is empty is further challenged due to the low/lowering/empty volume of the RWS
- T.
Furthermore, shutdown margin and the ability to maintain the reactor subcritical during a
secondary LOCA would be challenged due to the lack of borated water getting to the reactor
during the injection phase.
To the extent that there may be doubt regarding to the validity of answer (B) despite the
referenced material above, the UFSAR (B/B 6.3-31) offers this information to challenge answer
(D) and suggests that 1CV8804A and 1SI8804B failing closed is of no consequence with
respect to ECCS meeting its ability to deliver core cooling during the cold leg recirculation
phase, upon which the answer key relies:
following a large break LOCA. In the event the manual actions to align RHR pump discharge
to the suction of the SI and CV pumps are not completed prior to reaching the RWST empty
level, required core cooling capability will be maintained by the RHR pumps.
Second Conclusion:
While the failed closed positions of 1CV8804A and 1SI8804B will prevent cold recirculation via
the CV and SI pumps, the UFSAR says that ECCS will still be able to maintain core cooling
capability via the RH pumps with the water in the containment sump alone. In which case, I
would argue that the safety function to maintain core cooling by ECCS is not totally extinguished
with respect to answer (D).
Third Conclusion:
Lastly, there is one more argument to support the notion that a loss of safety function for
ECCS exists. According to TRM, Appendix O, 3.5.4 RWST is a support system for 3.5.2
ECCS (Operating). The continued draining of the RWST lower than 89 percent
TS impacts not just one train of ECCS, but both. All ECCS pumps take suction off a
common discharge header from the RWST. If the available volume isnt adequate (or in
existence) for one pump, then its not available for any of the other ECCS pumps either. It
appears this would be outside the design basis for Byron, which is probably also why
CFR 50.54X is referenced in the Byron Operating Department Policy Statement (No. 600-04)
for gravity draining the RWST during a security threat.
Facility recommendation:
The distractor was written with the idea that the RWST would completely drain to the
containment sump, as the applicant correctly surmised. While the RWST level would be less
than TS 3.5.4 required level, the inventory would be available to the ECCS system via the
Containment Recirculation Sumps.
An automatic signal caused by a Safety Injection signal in conjunction with Low-2 RWST level
(below 47 percent) will automatically align both RH pumps suctions to the containment sump.
Further manual actions are necessary to close the RWST to RH pump suction valves, and to
align the RH pump discharge to the high head injection pumps suction.
With RWST level below 9 percent, the operators would stop the high head injection pumps that
are aligned to the RWST based on the Operator Action Summary page of 1BEP ES-1.3,
Transfer to Cold Leg Recirculation procedure. While this would render the high head injection
pumps incapable of immediately providing flow on a Safety Injection actuation, the Byron
emergency procedure network would direct operators to use 1BEP ES-1.3 to align the RH pump
discharge to the high head injection pumps suction, restoring high head injection flow.
The facility recommends making no change to the grading of the question.
References:
Operations Policy 600-04
UFSAR 6.3 Emergency Core Cooling System
TRM Appendix O, Safety Function Determination Program
1BEP ES-1.3 Transfer to Cold Leg Recirculation
NRC Resolution:
The intent of Question 97 was to evaluate the applicants knowledge of operability/availability of
ECCS equipment. The applicant was to evaluate a series of choices of valve positions and
determine which condition would prohibit the ECCS system from performing its safety function.
Distractor D resulted in a condition that would prohibit RH pump discharge from being re-aligned
to both CV and SI pump suctions during a LOC
- A. During a LOCA, the RWST would drain
requiring both auto and manual actions to realign ECCS pumps to draw suction from the
containment sump. For a small break LOCA, (where RCS pressure is greater than RHR pump
discharge pressure), realigning RHR pump discharge to the SI and CV pumps would be
necessary to ensure adequate core cooling. Hence distractor D was considered the only
correct answer.
However, the applicant argues in his first and second conclusions that for a large break LOCA,
realigning RHR pump discharge through valves CV8804A and SI8804B would be unnecessary
to ensure adequate core cooling since the flow from the RHR pumps alone would be sufficient
to keep the core cooled. This scenario is supported by UFSAR page 6.3-31. This implies that
distractor D would not always be the correct answer.
The applicant believed that the B distractor was correct since it resulted in draining the RWST
to the containment sump; a condition that he believed would challenge the ECCS safety
function. This condition does not challenge the ECCS safety function since ECCS is designed
to draw suction from the containment sump and inject into the reactor vessel. In his first
conclusion, the applicant assumed the RWST was drained to such an extent that NPSH for the
high head injection pumps would be challenged. Additionally, if RCS pressure remained high,
RH pumps also would not be able to inject. There would be minimal injection until the ECCS
system was realigned from injection mode to recirculation mode. This scenario represents a
challenged ECCS system, but not an ECCS system that was unable to meet its safety
function, as stipulated in the question.
In his third conclusion, the applicant argues that both ECCS systems are impacted by a loss of
RWST level as distractor D hypothesizes. Additionally, the TRM and TS have limiting
conditions that would prohibit this condition. The examiners agree that the condition in
distractor D warrants entry into the TRM and TS and continued operation in this condition
would warrant entry into 10 CFR 50.54x (permission to operate in a condition prohibited by
Technical Specifications). But this condition is not outside of the design basis of the facility
since recirculation phase of ECCS is recognized by the UFSAR and is controlled by licensees
emergency procedures.
None of the arguments provided by the applicant can result in distracter B being correct; but
the arguments do partially discredit D distractor being correct always. Specifically, during the
recirculation phase of a LOCA, the condition described in distractor D would not result in
inadequate core cooling for a large break LOCA but would result in an inadequately cooled core
for a small break LOCA. Since Question 97 asked for conditions that will result in the ECCS
system [being] unable to meet its safety function, and since there were no conditions as to what
type of accident existed in the stem of the question, the condition presented in distracter D
meets the condition in the stem of the question for a small break LOC
- A. Therefore, the answer
to Question 97 remains D.
M. Pacilio -2-
In accordance with Title 10 of the Code of Federal Regulations, Section 2.390 of the NRC's
"Rules of Practice," a copy of this letter and its enclosures will be available electronically for
public inspection in the NRC Public Document Room or from the Publicly Available Records
System (PARS) component of NRC's Agencywide Documents Access and Management
System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Hironori Peterson, Chief
Operations Branch
Division of Reactor Safety
Docket Nos. 50-454; 50-455
Enclosures:
1. ER 05000454/2014301; 05000455/2014301
w/Attachment: Supplemental Information
2. Simulation Facility Report
3. Written Examination Post-Examination Comment
Resolution
cc w/ encls: Distribution via LISTSERV