ML051570095
| ML051570095 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/03/2005 |
| From: | Gody A Operations Branch IV |
| To: | Ray H Southern California Edison Co |
| References | |
| 50-361/05-301, 50-362/05-301 | |
| Download: ML051570095 (17) | |
See also: IR 05000361/2005301
Text
June 3, 2005
Harold B. Ray, Executive Vice President
San Onofre, Units 2 and 3
Southern California Edison Co.
P.O. Box 128, Mail Stop D-3-F
San Clemente, CA 92674-0128
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 - NRC
EXAMINATION REPORT 05000361/2005301; 05000362/2005301
Dear Mr. Ray:
On April 25 to 29, 2005, the NRC completed an inspection at your San Onofre Nuclear
Generating Station, Units 2 and 3. The enclosed examination report documents the results of
the NRC examination and the inspection findings which were discussed on April 28, May 25,
and June 2, 2005 with you and other members of your staff.
The examinations included an evaluation of four applicants for reactor operator licenses and
two applicants for senior operator licenses. The written and operating examinations were
developed using NUREG-1021, "Operator Licensing Examination Standards for Power
Reactors," Revision 9. We determined that five applicants satisfied the requirements of
10 CFR Part 55, and the licenses have been issued or deferred as appropriate.
Based on the results of this examination/inspection, the NRC has determined that one violation
of NRC requirements occurred. Based on the results of this inspection, the NRC has identified
an issue that was evaluated under the risk significance determination process as having very
low safety significance (green). The NRC has also determined that a violation was associated
with this issue. This violation is being treated as a noncited violation (NCV), consistent with
Section VI.A of the Enforcement Policy. The NCV is described in the subject inspection report.
If you contest the violation or significance of the NCV, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with
copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV,
611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident
Inspector at the San Onofre Nuclear Generating Station, Units 2 and 3 facility.
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
Southern California Edison Co.
-2-
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).
Sincerely,
/RA/
Anthony T. Gody, Chief
Operations Branch
Division of Reactor Safety
Dockets: 50-361; 50-362
Enclosure:
Examination Report 05000361; 05000362/2005301
w/Attachment Supplemental Information
cc w/enclosure:
Chairman, Board of Supervisors
County of San Diego
1600 Pacific Highway, Room 335
San Diego, CA 92101
Gary L. Nolff
Power Projects/Contracts Manager
Riverside Public Utilities
2911 Adams Street
Riverside, CA 92504
Eileen M. Teichert, Esq.
Supervising Deputy City Attorney
City of Riverside
3900 Main Street
Riverside, CA 92522
Raymond Waldo, Vice President,
Nuclear Generation
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Southern California Edison Co.
-3-
David Spath, Chief
Division of Drinking Water and
Environmental Management
California Department of Health Services
P.O. Box 942732
Sacramento, CA 94234-7320
Michael R. Olson
San Onofre Liaison
San Diego Gas & Electric Company
P.O. Box 1831
San Diego, CA 92112-4150
Ed Bailey, Chief
Radiologic Health Branch
State Department of Health Services
P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414
Mayor
City of San Clemente
100 Avenida Presidio
San Clemente, CA 92672
James D. Boyd, Commissioner
California Energy Commission
1516 Ninth Street (MS 34)
Sacramento, CA 95814
Douglas K. Porter, Esq.
Southern California Edison Company
2244 Walnut Grove Avenue
Rosemead, CA 91770
Dwight E. Nunn, Vice President
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Daniel P. Breig, Station Manager
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Southern California Edison Co.
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A. Edward Scherer
Southern California Edison
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Brian Katz, Vice President, Nuclear
Oversight and Regulatory Affairs
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Adolfo Bailon
Field Representative
United States Senator Barbara Boxer
312 N. Spring Street, Suite 1748
Los Angeles, CA 90012
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Enclosure
ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Dockets:
50-361; 50-361
Licenses:
Report No.:
05000361/2005-301; 05000362/2005-301
Licensee:
Southern California Edison Co.
Facility:
San Onofre Nuclear Generating Station, Units 2 and 3
Location:
5000 S. Pacific Coast Hwy.
San Clemente, California
Dates:
April 25 to 28, 2005 and May 6 to June 3, 2005
Inspectors:
Gary W. Johnston, Sr. Operations Engineer
Mark Haire, Operations Engineer
James F. Drake, Operations Engineer
Approved By:
Anthony T. Gody, Chief
Operations Branch
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
ER 05000361/2005301; 05000-362/2005301; 4/25-29/2005; San Onofre Nuclear Generating
Station, Units 2 and 3: Initial Operator Licensing Examination Report. Equipment Operation.
NRC examiners evaluated the competency of four applicants for reactor operator licenses
and two applicants for senior operator licenses. The licensee developed the operating portions
of the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power
Reactors," Revision 9. The NRC staff developed the written portions of the examinations.
Licensee proctors administered the written examinations to all applicants on April 29, 2005, in
accordance with instructions provided by the chief examiner. The NRC examiners administered
the operating tests on April 25 through 28, 2005. Five of the applicants passed all portion of the
examinations, one failed the written portion of the examinations. One Green noncited violation
was identified. The significance of this finding is indicated by its color (Green, White, Yellow,
Red) using Inspection Manual Chapter 0609, Significance Determination Process. The NRCs
program for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
Cornerstone: Mitigating Systems
Green. The examiners identified a noncited violation of Technical Specification 5.4.1
associated with an inadequate operating procedure. Specifically, the examiners determined
that Procedure SO23-3-2.6, Shutdown Cooling System Operation, Revision 21, was not
adequate, in that, attachments to the procedure had conflicting values for the maximum
temperature allowed for the shutdown cooling system and interconnecting piping. The licensee
is correcting the procedure to resolve the discrepancy and has documented this issue in Action
Request number 0505014401 Action 1.
The finding is a performance deficiency in that the licensee failed to identify that the procedure
did not correctly identify the limiting temperature for the shutdown cooling system and
interconnecting piping. The finding is more than minor because it affects the Mitigating
Systems Cornerstone of procedural quality and equipment performance, in that, it could result
in exceeding the maximum operating temperature for the shutdown cooling system and
interconnecting piping. Using the Phase 1 worksheet in Manual Chapter 0609, Significance
Determination Process, this finding is determined to be of very low safety significance because
there was no actual system temperature limits exceeded (Section 1R04).
Enclosure
Report Details
1.
REACTOR SAFETY
Cornerstone: Mitigating Systems
1R04
Equipment Operation
a.
Examination Scope
The examiners reviewed plant operating procedures and references for the evaluation of
post examination comments from the licensee to ensure that the answers to all
challenged questions were relevant and accurate in accordance with the plant
procedures and references.
b.
Findings
Introduction. A Green noncited violation of Technical Specification 5.4.1 was identified
for an inadequate procedure associated with operation of the Shutdown Cooling
System.
Description. On May 25 , 2005, while evaluating licensee post examination comments
on the written examination portion of an initial license examination (Section 4OA5.1.1.b)
the NRC staff noted that the procedure had conflicting values for the maximum
allowable temperature for the shutdown cooling system and interconnecting piping. The
licensee placed an entry into their corrective action program for a procedure change to
correct Procedure SO23-3-2.6 Shutdown Cooling System Operation, Attachment 17,
Action Request (AR) number 0505014401, Action 1.
Analysis. The finding is a performance deficiency in that the licensee failed to identify
that the procedure did not correctly/consistently identify the maximum allowable
temperature for the shutdown cooling system and interconnecting piping. The finding is
more than minor because it affects the Mitigating Systems Cornerstone for procedure
quality in that it could result in a failure to properly operate the shutdown cooling system.
The finding is of very low safety significance (Green) because the finding did not result
in temperature limits of the shutdown cooling system and interconnecting piping being
exceeded. Using the Phase 1 worksheet in Manual Chapter 0609, Significance
Determination Process, this finding is determined to be of very low safety significance
because there was no actual system temperature limits exceeded (Section 1R04).
Enforcement. Technical Specification 5.4.1 requires, in part, that written procedures be
established, implemented, and maintained as recommended in Appendix A of
Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation),"
Revision 2, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, provides
the typical activities that should be covered by written procedures. Section 2 requires
general plant operating procedures.
Enclosure
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Contrary to the above, Procedure SO23-3-2.6, Attachment 17, Revision 21, failed to
implement proper/consistent maximum allowable temperature limits for systems
operated using this procedure. Because this example of an inadequate procedure is of
very low safety significance and has been entered into the licensees corrective action
program (AR #0505014401, Action 1), this violation is being treated as a noncited
violation, consistent with Section VI.A of the NRC Enforcement Policy: Noncited
Violation 05000361;05000362/2005301-01, Inadequate Procedure SO23-3-2.6,
Revision 21.
4.
OTHER ACTIVITIES
4OA5 Other Activities
.1
Initial License Examination Administration
.1.1
Operator Knowledge and Performance
a.
Examination Scope
The NRC examination team administered the operating examinations to the
six applicants on April 25-28, 2005. The applicants participated in two dynamic
simulator scenarios, a control room and facilities walkthrough test consisting of 14 to
15 system and administrative tasks.
On April 29, 2005, the licensee proctored the administration of the written examinations
to all six applicants and forwarded the proposed grades together with the performance
analysis to the NRC staff for approval.
b.
Findings
Five applicants passed all parts of the examinations. One applicant failed the written
portion of the examination. For the written examinations, the average score for reactor
operator applicants was 82.34 percent, and the average score for senior operator
applicants was 90.00 percent. The reactor operator applicant scores ranged from 78.6
to 88.00 percent, and the senior operator applicant scores ranged from 88 to 92
percent.
The licensee conducted a performance analysis for the written examinations with
emphasis on questions missed by half or more of the applicants. After reviewing the
licensees analysis, the examiners concluded that the analysis adequately addressed
potential training deficiencies, and that the licensee provided adequate remediation for
these apparent deficiencies. The text of the licensees examination question
performance analysis may be accessed in the ADAMS system as noted in the
attachment.
Enclosure
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.2
Initial Licensing Examination Development
The licensee developed the operating portion of the examinations in accordance with
NUREG-1021, Revision 9. Licensee facility training and operations staff involved in
examination development were on a security agreement. The written examination was
developed by the NRC staff, also in accordance with NUREG-1021, Revision 9.
.2.1
Examination Quality
a.
Examination Scope
The facility licensee submitted post examination comments and challenges on May 6,
2005. Examiners reviewed the submittal against the plant operating procedures and
reference materials.
b.
Findings
The chief examiner determined that, although the overall discrimination validity of
examination was reduced, the written examination overall as modified by the post
examination changes was within the range of acceptability expected for an initial
licensing examination and was satisfactory. Documentation of the licensees post
examination comments and the NRC staff review are located in Section .2.5 below
.2.2
Operating Examination Outline and Examination Package
a.
Examination Scope
The facility licensee submitted the operating examination outlines and draft operating
examinations on March 14, 2005. Examiners reviewed the submittal against the
requirements of NUREG-1021, Revision 9, and forwarded minor comments to the
licensee on Mach 15, 2005. The chief examiner conducted an onsite validation of the
examinations and provided further comments during the week of April 4-7, 2005. The
licensee satisfactorily completed comment resolution on April 11, 2005.
b.
Findings
Examiners approved the initial examination outline with minor comments and advised
the licensee to proceed with the operating examination development.
The chief examiner determined that the operating examinations initially submitted by the
licensee were within the range of acceptability expected for a proposed examination and
were satisfactory.
No findings of significance were identified.
Enclosure
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.2.3
Simulation Facility Performance
a.
Scope
The examination team observed simulator performance with regard to plant fidelity
during the examination validation and administration. The chief examiner also reviewed
the outstanding simulator work orders to determined if there were any conflicts with
examinations administered on the simulator.
b.
Findings
No simulator deficiency was noted during validation and no findings of significance were
identified.
.2.4
Examination Security
a.
Scope
The examiners reviewed examination security both during the onsite preparation and
examination administration weeks with respect to NUREG-1021 requirements. Written
plans for simulator security and applicant control were reviewed and discussed with
licensee personnel.
b.
Findings
No findings of significance were identified.
.2.5
Post Examination Comments
Question
32.
SO23-3-2.6, Shutdown Cooling System, cautions the operator to maintain the
SDCS and interconnecting piping within the following limit:
A. < 340°F (normal operations)
B. < 350°F (normal operations)
C. < 370°F (post-accident)
D. < 380°F (post-accident)
Proposed Answer: B
Licensee Challenge:
The licensee contended that both A and B are correct answers based on several
references provided. Procedure SO23-3-2.6 Shutdown Cooling System Operation,
Step 6.8.3, requires the operator to ensure reactor coolant system cold-leg temperature
is between 260EF and 340EF as a condition for establishing operation of the system.
Enclosure
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Further, the licensee points out that Attachment 1 of the procedure requires that reactor
coolant system hot-leg temperature be below 340EF prior to placing low temperature
overpressure protection in service. Therefore, the licensee contends that both A and B
are correct based on these assertions.
NRC Review
The NRC staff review of the question took note of the graphs and notes in Attachment 1
of Procedure SO23-3-2.6 which require the operator to maintain the shutdown cooling
system and interconnecting piping within the operating limit of # 340EF. This is in
conflict with the limitations for the shutdown cooling system described in Attachment 17
of Procedure SO23-3-2.6 which state the limit for the shutdown cooling system as
< 350EF. Additional discussions with the licensee identified that the 350EF limit was in
error and a change to the procedure had been submitted to correct this error.
Therefore, the NRC staff considers that only A is the correct answer.
Question
47.
Minimum allowable starting air pressure for Emergency Diesel Generator
OPERABILITY is:
A.
150 psig
B.
175 psig
C.
185 psig
D.
210 psig
Proposed Answer: B
Licensee Challenge:
The licensee contended that both A and B are correct based on the Technical Specification 3.8.3. Action F, which allowed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to re-establish starting air pressure
to $175 psig. If A was chosen, the reasoning assumes that the emergency diesel
generator would remain operable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as 150 psig is above the lower bound for
Action F of 136 psig.
NRC Review:
The NRC staff review noted that the stem could reasonably be interpreted to be
expanded to the OPERABLE condition when air pressure is between 175 and 136 psig.
The Technical Specification definition of Operable- Operability states in part: A system,
subsystem, train, component, or device shall be operable when it is capable of
performing its specified safety function(s) and when all necessary attendant
instrumentation, controls, seal water, lubrication, and other auxiliary equipment that are
required for the system, subsystem, train, component, or device to preform its specified
safety function(s) are also capable of performing their related support functions.
The starting air system is required to have a minimum capacity for five successive start
attempts on the diesel generator without recharging the air start receivers. The starting
Enclosure
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air system is not capable of performing its specified safety function when pressure is
<175 psig. The minimum pressure to maintain operabilty of the starting air system is
175 psig. However, the NRC Inspection Manual, Part 9900, Standard Technical Specifications, Section 1.0 OPERABILITY states in part that some plant Technical
Specifications are designed to allow for different modes of operation and/or situations in
which a system, structure, or component remains operable for a limted period of time
even though the situation does not meet either GDC, single-failure, or design bases
requirements. The staff has permitted licensee's to use risk arguments to develop
plant Technical Specifications that allow periods of operation when systems, structures,
or components do not meet GDC, single-failure, or design bases requirements while still
being considered operable for either a period of time or plant condition. The design of
this particular Technical Specification becomes apparent when you read Limting
Condition for Operation 3.8.3.G which states in part, if...."Required Action and
associated Completion Time of condition A, B, C, D, E, or F are not met...." then
...."Declare the associated DG inoperable." This point is further supported by the fact
that the staff has approved this particular risk or probabilistic argument and it is
discussed in the bases for TS 3.8.3 Action F.1. which states in part, "A period of
48-hours is considered sufficient ..... based on the remaining air start capacity, the fact
that most DG starts are accomplished on the first attempt, and the low probability of an
event during this brief period." The NRC staff concluded that since this particular
Technical Specification (TS 3.8.3) represents one of these requirements, in that the
EDG would remain operable for a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period with starting air pressure between 136
and 175 psig, both answers A and B are correct.
Question
54.
Which of the following conditions would prevent an emergency containment
entry in accordance with Procedure SO23-3-2.34, Containment Access Control,
Inspections and Airlocks Operation?
A.
Containment Pressure is 3.4 psig.
B.
Containment Humidity is 100%.
C.
Containment Temperature is 158EF.
D.
Containment atmosphere Oxygen level is 14%.
Proposed Answer: A
Licensee Challenge:
The licensee recommended that both A and C be accepted as correct answers. Their
basis for C being correct is that Procedure SO23-3-2.34, Containment Access Control,
Inspections and Airlocks Operation, required that stay times with regard to containment
entry be done in accordance with Procedure SO123-XVI-22, Safety Procedure,
Attachment 3, Recommended Work Time Limits, of which provides recommended
work time limits with regard to temperature. The chart stops at a maximum temperature
of 140EF. The contention was that the 158EF would preclude entry per the chart and,
therefore, C would also be correct.
Enclosure
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NRC Review:
While the temperature of 158EF is not a condition that would prevent an emergency
containment entry in accordance with Procedure SO23-3-2.34, Containment Access
Control, Inspections and Airlocks Operation, it is beyond the limit of the chart in
Attachment 3 of Procedure SO123-XVI-22 which was referenced by
Procedure SO23-3-2.34. Therefore, since there is no guidance provided procedurally
to a reactor operator as to a case assessment or the capabilities of cooling gear for
entry above 140EF, the NRC staff will accept A and C as correct answers.
Question
61.
Unit 2 is at 45% power after coming out of an outage. A grid disturbance has
resulted in grid frequency and turbine speed increasing. With turbine speed now
at 1950 rpm, the response of the turbine and generator control systems should
be to:
A.
Throttle the governor valves in the closed direction.
B.
Trip close the turbine stop and governor valves.
C.
Trip the generator output breaker on a high volts to hertz signal.
D.
Throttle open the steam bypass control valves.
Proposed Answer: B
Licensee Challenge:
The licensee contended that because there are three separate signal paths (the turbine
control system, the turbine supervisory system, and the turbine protection system), there
may have been confusion among the applicants as to what the question stem was
asking with regard to the 1950 rpm speed of the turbine. The licensee noted that A was
the correct answer if the turbine had not tripped. The licensee further stated that B is
not correct because the question stem asks what the response of the turbine and
governor control systems would be for the conditions presented. The licensee stated
that the training for the turbine controls and instrumentation systems discusses the
turbine control systems as the turbine control system, the turbine protection system, and
the turbine supervisory system. The licensee stated that this may also have contributed
to confusion among the applicants, since they would not consider the control, protection,
and supervisory systems as subsystems of the turbine control and instrumentation
systems. Therefore, the licensee requested that A be accepted as the correct answer.
NRC Review:
The question stem stated ...the response of the turbine and generator control
systems... The stem of this question was developed in this manner to intentionally
include all the turbine and generator control sub-systems. The condition presented in
the stem of 1950 rpm was clearly above the setpoint for the turbine control system to trip
close the turbine stop and governor valves and should have resulted in a trip close
signal being sent to the valves. The turbine and generator controls systems are
Enclosure
-8-
designed to allow only the lowest control or trip signal to pass through an analog module
low value gate which in this case would mean the 1950 rpm trip signal, the lowest value,
should pass to the valves. While the NRC staff acknowledged the confusion operator
candidates may have experienced due to the inconsistent use of terminology in training,
weak training is not considered a valid justification for accepting a clearly technically
wrong distractor as a correct answer. The NRC staff determined that only B was the
correct answer.
4OA6 Meetings, Including Exit
On April 28, 2005, the examination team presented the examination results to Mr. Ray
Waldo, Vice President Nuclear Generation, and other members of the licensee's
management staff at the conclusion of the operating examinations. The licensee
acknowledged the findings presented. The licensee did not identify as proprietary any
information or materials examined during the examination.
On May 25 and June 3, 2005 the results of the NRC staff review of the post examination
comments from the licensee and subsequent examination regrading were discussed
with Kurt Rauch, Operations Training Manager, and other members of the licensee staff.
The final resolution of post examination comments were acknowledged by the licensee.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
D. Brieg, Station Manager
A. Hagenmeyer, Supervisor, Nuclear Training Division
C. McAndrews, Manager, Nuclear Oversight and Assessment
M. McBrearty, Engineer, Nuclear Regulatory Affairs
M. Ney, Operator Initial Training Supervisor
D. Nunn, VP Engineering and Technical Services
J. Osborne, Corrective Action Program Lead
K. Rauch, Manager, Nuclear Training Division
R. Sandstrom, Manager, Training
A. Scherer, Manager, Nuclear Regulatory Affairs
R. Waldo, VP Nuclear Generation
NRC personnel
C. Osterholtz, Senior Resident Inspector
M. Sitek, Resident Inspector
ADAMS DOCUMENTS REFERENCED
Accession No.: ML051310380 Written examination for reactor and senior operators
Accession No.: ML051310408 Written examination performance analysis
Documents Reviewed
Open Simulator Work Orders as of April 27, 2005