ML080100408
| ML080100408 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/10/2008 |
| From: | Widmann M NRC/RGN-II/DRS/OLB |
| To: | Madison D Southern Nuclear Operating Co |
| References | |
| 50-321/07-301, 50-366/07-301 | |
| Download: ML080100408 (13) | |
See also: IR 05000321/2007301
Text
January 10, 2008
Southern Nuclear Operating Company, Inc.
ATTN: Mr. Dennis R. Madison
Vice President - Hatch
Edwin I. Hatch Nuclear Plant
11028 Hatch Parkway North
Baxley, GA 31513
SUBJECT: EDWIN I. HATCH NUCLEAR PLANT- NRC EXAMINATION REPORT
05000321/2007301 AND 05000366/2007301
Dear Mr. Madison:
During the period December 3 - 6, 2007, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to
operate the Edwin I. Hatch Nuclear Plant Units 1 and 2. At the conclusion of the tests, the
examiners discussed the tests and preliminary findings with those members of your staff
identified in the enclosed report. The written examination was administered by your staff on
December 10, 2007.
Four Senior Reactor Operator (SRO) applicants passed both the written examination and
operating test. One Reactor Operator (RO) and one SRO applicant failed the written
examination. One RO applicant failed both the written examination and the operating test.
There were five post examination comments. The NRC resolutions to these comments are
summarized in Enclosure 2.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter
and its enclosure will be available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARS) component of NRC's
document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
SNC 2
Should you have any questions concerning this letter, please contact me at (404) 562-4550.
Sincerely,
Malcolm T. Widmann, Chief
Operations Branch
Division of Reactor Safety
Docket Nos.: 50-321, 50-366
Enclosures: 1. Report Details
2. NRC Resolution to the Facility Comments
cc: See Page 3
SNC 3
J. T. Gasser Resident Manager
Executive Vice President Oglethorpe Power Corporation
Southern Nuclear Operating Company, Inc. Edwin I. Hatch Nuclear Plant
Electronic Mail Distribution Electronic Mail Distribution
David H. Jones Senior Engineer - Power Supply
Vice President - Engineering Municipal Electric Authority
Southern Nuclear Operating Company, Inc. of Georgia
P.O. Box 1295 Electronic Mail Distribution
Birmingham, AL 35201-1295
Reece McAlister
L. M. Stinson, Vice President, Executive Secretary
Fleet Operations Support Georgia Public Service Commission
Southern Nuclear Operating Company, Inc. 244 Washington Street, SW
11028 Hatch Parkway North Atlanta, GA 30334
Baxley, GA 31513
John C. Lewis,Training Manager
Raymond D. Baker c/o Edwin C. Hatch Nuclear
Manager Licensing - Hatch Generating Plant
Southern Nuclear Operating Company, Inc. Southern Nuclear Operating Company
Electronic Mail Distribution 11028 Hatch Parkway North
Baxley, GA 31513
Arthur H. Domby, Esq.
Troutman Sanders
Electronic Mail Distribution
Laurence Bergen
Oglethorpe Power Corporation
Electronic Mail Distribution
Moanica Caston
Southern Nuclear Operating Company, Inc.
Bin B-022
P. O. Box 1295
Birmingham, AL 35201-1295
Director
Department of Natural Resources
205 Butler Street, SE, Suite 1252
Atlanta, GA 30334
Manager, Radioactive Materials Program
Department of Natural Resources
Electronic Mail Distribution
Chairman
Appling County Commissioners
69 Tippins St., Suite 201
Baxley, GA 31513
OFFICE RII:DRS RII:DRS RII:DRS RIV:DRS
SIGNATUR * / RA / * /RA / * /RA/ * /RA /
NAME RAiello:pmd RBaldwin BCaballero MWidmann
DATE 1/4/08 1/4/08 1/4/08 1/10/08
E-MAIL YES NO YES NO YES NO YES NO
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-321, 50-366
Report No.: 05000321/2007301 and 05000366/2007301
Licensee: Southern Nuclear Power Company (SNPCO)
Facility: Edwin I. Hatch Nuclear Plant, Units 1 & 2
Location: 11030 Hatch Parkway N.
Baxley, GA, 31513
Dates: Operating Tests - December 3 - 6, 2007
Written Examination - December 10, 2007
Examiners: R. Aiello, Chief, Senior Operations Engineer
R. Baldwin, Senior Operations Engineer
B. Caballero, Operations Engineer
Approved by: Malcolm T. Widmann, Chief
Operations Branch
Division of Reactor Safety
Enclosure 1
SUMMARY OF FINDINGS
ER05000321/2007301 and ER05000366/2007301; 12/03 - 06, 2007 & 12/10, 2007; Edwin I.
Hatch Nuclear Plant, Units 1& 2, Licensed Operator Examinations.
The NRC examiners conducted operator licensing initial examinations in accordance with the
guidance in NUREG-1021, Revision 9, Operator Licensing Examination Standards for Power
Reactors. This examination implemented the operator licensing requirements of 10 CFR
§55.41, §55.43, and §55.45.
The NRC administered the operating tests during the period of December 3 - 6, 2007.
Members of the Edwin I. Hatch training staff administered the written examination on December
10, 2007. The written examination was developed by the NRC and the operating test was
developed by the Edwin I. Hatch Training Department.
Four Senior Reactor Operator (SRO) applicants passed both the written examination and
operating test. One Reactor Operator (RO) and one SRO applicant failed the written
examination. One RO applicant failed both the written examination and the operating test.
There were five post examination comments. The NRC resolutions to these comments are
summarized in Enclosure 2.
No findings of significance were identified.
Enclosure 1
Report Details
4. OTHER ACTIVITIES
4OA5 Operator Licensing Initial Examinations
a. Inspection Scope
The facility developed the operating test and the NRC developed the written
examination in accordance with NUREG-1021, Operator Licensing Examination
Standards for Power Reactors, Revision 9. The NRC reviewed the proposed operating
test. Examination changes agreed upon between the NRC and the licensee were made
according to NUREG-1021 and incorporated into the final version of the test materials.
The examiners reviewed the licensees examination and test security measures while
preparing and administering the examinations and tests to ensure examination and test
security and integrity complied with 10 CFR 55.49, Integrity of examinations and tests.
The examiners evaluated two RO and five SRO applicants who were being assessed
under the guidelines specified in NUREG-1021. The examiners administered the
operating tests during the period of December 3 - 6, 2007. The written examination was
administered by the Edwin I. Hatch training staff on December 10, 2007. The
evaluations of the applicants and review of documentation were performed to determine
if the applicants, who applied for licenses to operate the Edwin I. Hatch Nuclear Plant,
met requirements specified in 10 CFR 55, Operators Licenses.
b. Findings
No findings of significance were identified.
The licensee and the NRC reviewed the final version of the written examination and
operating test, and indicated that these exams were within the range of acceptability
expected for the proposed examination and test respectively. Four SRO applicants
passed both the written examination and operating test. One RO and one SRO
applicant failed the written examination. One RO applicant failed both the written
examination and the operating test. Each applicant who passed the operating test and
written examination with an overall score greater than 82% and SRO-only score greater
than 74%, as applicable, was issued an operator license commensurate with the level of
examination administered.
The combined RO and SRO written examinations with knowledge and abilities (K/As)
question references/answers and examination references may be accessed in the
ADAMS system (ADAMS Accession Numbers, ML 080040289 and ML080040296).
Enclosure 1
4
4OA6 Meetings
Exit Meeting Summary
On December 6, 2007, the examination team discussed generic issues with Mr. Dennis
Madison, Site Vice President, and members of his staff. The examiners asked the
licensee whether any materials examined during the examination should be considered
proprietary. No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED
Licensee personnel
S. Bargeron, Plant Manager
C. Edmund, Nuclear Operations Instructor
S. Grantham, Operation Training Supervisor
J. Lewis, Training Manager
D. Madison, Plant Hatch Vice President
R. Musgrove, Operations Superintendent
K. Wainwright, Initial License Senior Instructor
NRC personnel
J. Hickey, Senior Resident inspector
P. Niebaum, Resident Inspector
S. Shaeffer, NRC Branch Chief
Enclosure 1
NRC Resolution to the Facility Comment
A complete Text of the licensee's post examination comments can be found in ADAMS under
Accession Number ML080040300.
RO QUESTION # 18
LICENSEE COMMENT:
This question deals with a loss of condenser vacuum with the unit at 20% and asks the
applicant for the status of the main steam isolation valves (MSIVs) and the appropriate
procedure for depressurizing the reactor to cold shutdown.
The licensee contends that the stem lacks sufficient focus to ensure that only choice D is
correct. The licensee contends that choice C is also correct based on the fact that at 20%
rated thermal power (RTP), when the MSIVs close on low vacuum, reactor pressure will
increase above the scram setpoint, requiring entry into the EOP RC flowchart. Once the EOP
RC flow chart is entered, guidance in the RC/P path at location F2 states:
If desired use one or more of the following: Low Low Set (LLS) or Alternate Reactor
pressure control system(s) in Table 1 per 31EO-EOP-107-2.
The licensee also contends that the systems used to cooldown in both procedures are basically
the same. The licensee recommended that both choices, C and D be accepted as correct
answers.
NRC DISCUSSION:
The NRC disagrees with the licensees comments. When the EOP RC flow chart is entered
because of reactor pressure rising above the scram setpoint, the guidance in the RC/P path at
location F2 fully states:
Stabilize reactor pressure below 1074 psig with main turbine bypass valves. If desired
used one or more of the following: Low Low Set (LLS) or Alternate Reactor pressure
control system(s) in Table 1 per 31EO-EOP-107-2.
Further down the RC/P path at location G2, the guidance in the RC/P path also states:
Begin reactor pressure reduction per 34GO-OPS-013-1 AND maintain the cooldown
rate below 100 °F/hr
The question stem provided sufficient focus to ensure only D is correct because it specifically
asked for the procedure to depressurize the reactor to cold shutdown.
Furthermore, a procedure note in section 7.5 (Reactor Depressurization) on page 25 of 34GO-
OPS-013-1, version 26.13 states:
31EO-EOP-107-2, Alternate RPV Pressure Control, will be used IF this section is being
performed per the EOPs AND the following systems are NOT available.
Enclosure 2
2
1) Bypass Valves (~22% total steam flow)
2) HPCI per 34SO-E41-001-2, (~10% total steam flow)
3) RCIC per 34SO-E51-001-2 (~2% total steam flow)
4) Main steam line drains (<1% total steam flow)
5) RHR per 34SO-E11-010-2
6) Manually lifting safety relief valves per 34SO-B21-001-2"
Since the question did not specify that the relief valves, HPCI, and RCIC were unavailable, the
applicant must assume that these systems were available. Therefore, the Alternate RPV
Pressure Control procedure, 31EO-EOP-107-2, was not the correct procedure to cooldown and
depressurize the plant in accordance with the note on page 25 of 34GO-OPS-013-1, version
26.13.
NRC RESOLUTION:
Based on the above discussion, the licensee's recommendation is not accepted and answer
choice "D" will be considered as the only correct answer.
RO QUESTION # 31
LICENSEE COMMENT:
This question deals with unloading Diesel Generator 2A when offsite power has been restored
following a loss of offsite power on Unit 2. The question asks the applicant to identify the
synchroscope direction as the control room operator transfers the Bus 2E load (300 kw) from
the Diesel Generator 2A to the offsite power supply.
The licensee contends that the stem lacks sufficient focus to ensure that only choice "C"
(counterclockwise) is correct because the stem failed to identify the choices as being "in
accordance with procedure." The licensee also contends that unstated assumptions can be
made that also support choice "A" (clockwise) because the procedure states that
counterclockwise at less than 500 KW is "desirable"; i.e., not the required direction.
Additionally, the licensee contends that paralleling at 300 KW with the rotation in the clockwise
direction would not be enough to result in reverse power.
Based on the above discussion, the licensee requests that answer choices "A" and "C" both be
accepted as correct.
NRC DISCUSSION:
The NRC disagrees with the licensee's comment because the question stem asked for the
"required" synchroscope direction. The word "requirement" implies a procedural requirement.
The procedure step in 34SO-R43-001-2, Diesel Generator Standby AC System, Section 7.3.1,
Transferring Power From Diesel Generator 2A To Normal Or Alternate Power states:
Enclosure 2
3
7.3.1.8 Using Diesel Gen 2A (2C) Speed Adjust switch, adjust diesel speed to attain a
slow synchroscope rotation in the desired direction (1 to 3 RPM)
The caution preceding this step states:
IF THE DIESEL GENERATOR LOAD IS LESS THAN 500 KW, IT IS DESIRABLE TO
HAVE THE SYNCHROSCOPE ROTATING IN THE COUNTERCLOCKWISE
DIRECTION TO AVOID OPERATING THE DIESEL AT LOW LOADS WHEN
PARALLELED TO THE GRID.
Because of the procedure step 7.3.1.8 requirement, then counterclockwise is the required
synchroscope direction.
NRC RESOLUTION:
Based on the above discussion, the licensee's recommendation is not accepted and choice "C"
will be considered as the only correct answer.
RO QUESTION # 59
LICENSEE COMMENT:
This question deals with a plant event which causes conditions that require an emergency
depressurization when torus level is very low (55") and the 4160 V busses 2A and 2B are
de-energized. The question asks the applicant to choose one system to emergency
depressurize the reactor given these circumstances.
The licensee contends that the stem lacks sufficient focus because it contained neither an
events timeline nor the current value of condenser vacuum. Consequently, an applicant may
reasonably assume that torus water level was approaching the heat capacity temperature limit
(HCTL) before the 4160 VAC busses became de-energized. In this case, the bypass valves
may still be available until low vacuum conditions eventually prevented their use. This
reasonable assumption makes choice B correct.
Additionally, the licensee also contends that the stem did not specify the initial power level.
Consequently, if the initial power level was low, and the 4160 VAC busses became
de-energized after a scram, then it could take some time for the condenser vacuum to diminish.
In this case the bypass valves would continue to be used until low vacuum conditions prevented
their use. Per 31EO-EOP-108-1, Alternate RPV Depressurization, the NOTE before step 3.1.1
states that if >10 inches of vacuum exists, then the bypass valves could be used as long as a
steam line break did not exist outside secondary containment (which was not the case). This
procedure (31EO-EOP-108-1), implies that the circulating water system should be in service IF
possible. Therefore, the licensee contends that a low initial power level assumption also makes
choice B correct.
Based on the above discussion, the licensee requests that both choices B and C be
accepted as correct.
Enclosure 2
4
NRC DISCUSSION:
The NRC disagrees with the licensees comment because the question stem required the
applicant to choose ONE system to accomplish the emergency depressurization. Additionally,
the applicants must not assume any unstated conditions in the stem, i.e., NUREG 1021, Rev 9,
Appendix E, Policies and Guidelines for Taking NRC Examinations, Section B.7 states:
When answering a question, do not make assumptions regarding conditions that are
not specified in the question unless they occur as a consequence of other conditions
that are stated in the question.
The question did not state that the 4160 VAC busses de-energized subsequent to the low torus
water level condition. The question is technically correct since the condenser is not available
following the loss of 4160 VAC busses 2A and 2B and the SRVs and HPCI are unavailable
due to the very low torus level (55). Furthermore, none of the applicants asked the exam
proctors to clarify the sequence of events or the initial power level when the exam was being
administered.
NRC RESOLUTION:
Based on the above discussion, the licensee's recommendation is not accepted and choice "C"
will be considered as the only correct answer.
SRO QUESTION # 85
LICENSEE COMMENT:
This question deals with a rod that was inadvertently withdrawn from position 16 to 22, versus
an intended position 18 during a startup when reactor power was 14%. The question asked the
SRO applicants to determine whether Tech Spec 3.1.6, Rod Pattern Control, contained a
required action statement applicable to these plant conditions.
The licensee contends that the stem lacked sufficient focus to ensure that only choice D is
correct because it did not exclude the initiation of a "tracking" required action statement. As an
example, the licensee referenced a previous question (#79) which did include a statement in
the stem that instructed the applicants to not consider a "tracking" required action statement
when choosing their answer. Because this question (#85) did not include a similar statement,
the licensee contends that the applicants thought the question was asking if an actual or a
"tracking" required action statement existed for this rod movement error.
The licensee contends that this rod movement error constituted a failure to meet banked
position withdrawal sequence (BPWS) requirements and required the implementation of a
"tracking" required action statement. The licensee contends that if power decreased below
10% (when the TS limiting condition of operation was exceeded) then an actual required action
statement would be required.
Enclosure 2
5
Based on the above discussion, the licensee requests that both choices C and D be
accepted as correct.
NRC DISCUSSION:
The NRC disagrees with the licensees comment because the stem asked specifically whether
Tech Spec 3.1.6, Rod Pattern Control, contained a required action statement when reactor
power was greater than 10%. There are no required action statements for Tech Spec 3.1.6,
Rod Pattern Control, when reactor power is 14%. Additionally, at the time the exam was
being administered, none of the applicants asked the exam proctors to clarify an actual or
tracking required action statement.
NRC RESOLUTION:
Based on the above discussion, the licensee's recommendation is not accepted and choice "D"
will be considered as the only correct answer.
SRO QUESTION # 86
LICENSEE COMMENT:
This question deals with an inadvertent HPCI low steam supply pressure isolation that occurs
as the MSIVs are being opened during a Unit 2 plant heatup and pressurization (during the
pressure equalization process across the MSIVs, reactor pressure drifts from 170 psig to 125
psig).
The licensee contends that the question is not valid because HPCI was no longer required to be
operable when reactor pressure is lowered to 125 psig. The licensee contends that the
applicants stated that there is no correct answer because at 125 psig, there is no required
action statement for HPCI.
The licensee requests that this question be deleted because there is no correct answer.
NRC DISCUSSION:
The NRC agrees with the licensee. The stem for the original question (submitted by the NRC)
included a sequence of events where reactor pressure lowered to 125 psig and then was
allowed to return to 160 psig without having HPCI re-aligned. The original question was
subsequently modified during the review processes by the exam team such that reactor
pressure remained at 125 psig (versus returning to a point greater than 150 psig). HPCI is not
required operable less than 150 psig, therefore the question, as presented to the applicants,
had no correct answer.
NRC RESOLUTION:
Based on the above discussion, the licensee's recommendation is accepted and this question is
deleted from the SRO exam.
Enclosure 2