ML20198S797
| ML20198S797 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/04/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20198S790 | List: |
| References | |
| 50-361-98-302, 50-362-98-302, NUDOCS 9901120007 | |
| Download: ML20198S797 (107) | |
See also: IR 05000361/1998302
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ENCLOSURE
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kl.S. NUCLEAR REGULATORY COMMISSION
.
REGION IV '
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- Docket Nos.:
.50 361,50 362.
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L'icense Nos.:)
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- Report No.:
- 50-361/98-302, 50-362/98-302
Licensee:
Southern California Edison Co.
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Facility: .
San Onofre Nuclear Generating Station, Units 2 and 3
< Location:
5000 S. Pacific Coast Hwy.
San Clemente, Califomia .
- Dates
November 15 through December 3,1998
Inspectors:
S. L. McCrory, Senior Reactor Engineer, Examiner / Inspector, Chief Examiner
T. O. McKernon, Senior Reactor Engineer, Examiner / Inspector'
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M. E. Murphy, Senior Reactor Engineer, Examiner / Inspector -
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T. R.' Meadows, Senior Reactor Engineer, Examiner / Inspector'
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R. E Lantz, Reactor Engineer, Examiner / Inspector
s Approved By:
J. L. Pellet, Chief, Operations Branch
- Division of Reactor Safety -
' ATTACHMENTS:
Attachment 1:
SupplementalInformation
~ Attachment 2:
Final Written Examinations and Answer Keys
Attachment 3:
Post Examination Comments
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9901120007 990104
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PDR - ADOCK 05000361
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EXECUTIVE SUMMARY
San Onofre Nuclear Generating Station, Units 2 and 3
NRC Inspection Report 50 361/98-302; 50-362/98-302
NRC examiners evaluated the competency of nine senior operator applicants and five reactor
operator applicants for issuance of operating licenses at the San Onofre Nuclear Generating
Station facility. The licensee developed the initiallicense examinations using NUREG-1021,
" Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The
NRC examiners administered the operating tests on November 1519,1998. The facility
licensee administered the initial written examinations to all applicants on November 20,1998.
Operations
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The large number of questions missed and the high number of common error responses
by most applicants indicated training weaknesses. This conclusion was further
supported by performance weaknesses observed during the operating examination
(Sections 04.1 and 04.2).
The facility licensee devele> ped an adequate written and operating examination.
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However, the post-examination review of the written examination identified a large
number of technicalinaccuracies. The large number of technicalinaccuracies indicated
a significant weakness in the facility licensee's initial technical review (Section O5.1.2).
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An inadvertent breach of examination security did not result in an examination
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compromise (Section 05.3).
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Reoort Details
Summary of Plant Status
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The units operated at essentially 100 percent power for the duration of this inspection.
I. Operations
04
Operator Knowledge and Performance
04.1 Initial Written Examination
a.
Insoection Scoce -
On November 20,1998, the facility licensee proctored the administration of the written
examination to nine senior operator license applicants and five reactor operator license
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applicants. The facility licensee provided post-examination comments (Attachment 3)
following the administration of the written examination. The chief examiner reviewed
the comments for technical adequacy. The chief examiner reviewed the written
examination grading on December 2,1998.
b.
Observations and Findinos
Three of five reactor operators and six of eight senior reactor opt . tor applicants
passed the written examination. The written examination was waived for one senior
reactor operator applicant who had passed the written examination on a prior licensing
examination. Reactor operator applicant scores ranged from 72.6 to 85.3 percent with
an average of 78.7 percent. Senior reactor operator applicant scores ranged from
63.8 to 85.1 percent with an average of 79.8 percent. The overall written examination
average was 79.4 percent.
- The following questions were missed by at least one half of the applicants. Questions
common to both examinations are shown with the number from the reactor operator
examination first.
Common questions: 1/1,6/7,11/13, 14/18*,16/21*,24/27*,28/29*,42/39*,58/52*,
65/58*, 78/74*, 81/79*, 85/84*, 86/85*, 94/94*, 98/99*
Reactor Operator only: 45*,49*,57,63*,64,88*,
Senior Operator only: 22*,32*,57*,78*,87
Most applicants gave the same incorrect answers to the above questions marked with
an asterisk (*) plus common question 59/53, and senior operator question 15. The
knowledge deficiencies fell roughly equally into two broad categories - systems and
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procedures. Of the system knowledge based errors, about two thirds related to logic or
control circuit performance. During the pre-examination review, the chief examiner
expressed concern to the facility licensee about the number of control logic questions
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and whether references should be provided for some of them. Licensee staff responded
that the tested areas were required knowledge.
c.
Conclusions
The large number of questions missed and the high number of common error responses
by most applicants indicated training weaknesses.
O4.2 Initial Ooeratina Test
a.
Inspection Scoce
The examination team administered the various portions of the operating test to the
14 applicants on November 15-19,1998. Each applicant participated in at least two
dynamic simulator scenarios and received a walkthrough test, which consisted of ten
system tasks together with followup questions for each system. Additionally, each
applicant was tasted on five subjects in four administrative areas with a combination of
administrative tasks and questions.
b.
Observations and Findinas
All applicants passed the operating examination.
The examiners observed consistently good three-way communications and supervision
of control panel activities during the dynamic simulator and dynamic walkthrough
portions of the operating test.
During Simulator Scenario 2, simultaneous steam generator tube rupture and failed
open steam generator safety valve malfunctions occurred on the same steam generator.
In one crew, no applicants observed the abnormal cooldown caused by the failed open
safety valve and, therefore, did not diagnose and respond to the bomonitored
radioactive release. During the same scenario, only one of five crews communicated to
management or support personnel any precautions or concerns regarding the
radiological conditions impacting recovery efforts.
There were three instances in which applicants read or operated the wrong radiation
monitors in response to system tasks or scenario events. The nature of the errors was
similar, and the examiners concluded that instrument label placement contributed to the
errors. The instrument labels were positioned below the instruments for a small number
of radiation monitors. Virtually all other instruments and controls in the control room had
the labels positioned above the instrument.
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c.
Conclusions
All applicants passed the operating examinations but exhibited some knowledge and
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ability weaknesses. This performance further supported that training weaknesses
existed.
05
Operator Training and Qualification
05.1
Initial Licensina Examination Develooment
The licensee developed the initial licensing examination in accordance with guidance
provided in NUREG-1021," Operator Licensing Examination Standards for Power
Reactors," Interim Revision 8, and additional guidance provided by the chief examiner.
05.1.1 Examination Outline
The facility licensee submitted the initial examination outline on September 2,1998.
The chief examiner reviewed the submittal against the requirements of NUREG-1021,
Interim Revision 8. The examination outlines satisfied the requirements of the
examination standards with regard to breadth, depth, and scope.
O5.1.2 Examination Packaae
a.
Inspection Scope
The facility licensee submitted the completed draft examination package by
October 5,1998. The chief examiner and peer reviewers reviewed the formal submittal
against the requirements of NUREG-1021, interim Revision 8. An onsite validation of
the operating examination was conducted during the period November 4-6,1998.
b.
Observations and Findinas
The reviewer directed that 18 of 125 written examination questions be revised or
replaced as a result of being assessed as discriminating at too high or too low a level.
The reviewer provided enhancement comments on an additional 25 questions. The
reviewer commented on several questions related to control systems logic as possibly
being too difficult to answer without a reference; however, the reviewer left the decision
with the facility licensee to propose the use of specific references.
Approximately 50 percent of the prescripted questions developed for Parts A and B of
the operating test had to be revised or replaced for various deficiencies including low
discrimination, direct look-up, and wrong focus. Overall, the walkthrough portion was
assessed as marginally adequate because there was at least one acceptable
prescripted question per task.
The reviewer identified two system tasks in one of the walkthrough test that tested the
same operator ability and directed that one be replaced. Both tasks required the
operator to parallel electrical generating sources (one for the main turbine generator and
one for an emergency diesel generator). During the onsite validation of the operating
examination, the chief examiner identified that two of the simulator malfunctions were
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also included as system tasks in the walkthrough part of the operating examination and
directed that the scenario malfunctions be replaced. Apart from this minor task
duplication, the reviewer determined that the simulator scenarios were of good quality.
The facility licensee provided a total of 22 post-examination comments (see
Attachment 3) on the written examination recommending question deletion and
acceptance of additional answers. Nearly all of the comments addressed technical
inaccuracies. The chief examiner accepted all the facility licensee post-examination
comments except the following:
Senior Operator Comment 5 - The facility licensee recommended deleting the
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question on the basis that the allowed maximum value was a pressurizer level of
57 percent, which was not one of the choices. The chief examiner rejected this
recommendation because the reference cited required that pressurizer level
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must be less than 900 ft , which equated to 57 percent. Therefore, Choice C
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(53 percent) remained as the only correct answer since it was the highest value
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below 57 percent.
Senior / Reactor Operator Comment 16/11 - The facility licensee recommended
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accepting Choice B as an additional correct answer based on the possible
assumption that condenser backpressure did not continue to increase above the
last reported value. The chief examiner rejected this recommendation based on
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the procedural requirement to take action if the condenser backpressure
increased to equal to or greater than 3.5 inches of mercury. Therefore, taking no
action until condenser backpressure increased to 4.5 inches of mercury
(Choice B) was not acceptable.
c.
Conclusions
The facility licensee developed an adequate written and operating examination that had
job performance measure prescripted questions of marginal quality. However, the
post-examination review of the written examination identified a large number of technical
inaccuracies. The large number of technicalinaccuracies indicated a significant
weakness in the facility licensee's initial technical review.
O5.2 Simulation Facility Performance
The examiners observed simulator performance with regard to fidelity during the
examination validation and administration. The simulation facility supported the
examination administration well. The examiners observed no problems.
05.3 Examination Security
During examination administration, the examination material was maintained in a locked
room to which only the examiners and limited members of the training staff, in the
security agreement, had keys.
On Tuesday, November 17,1998, between 5:30 a.m. and 6 a.m., a site security guard
opened the examination material room with a master key to permit the cleaning staff to
remove trash. The cleaning staff left the door to the room slightly ajar, but not open
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enough to permit observing the contents of the room. The examiners arrived at about
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6:30 a.m. and found the door ajar. The examination material did not appear to have
been disturbed. Only one license applicant had arrived onsite by that time. The
applicant was interviewed, and he stated that he had gone directly to the applicant's
sequestering room upon arrival. Members of the training staff had noted the applicant's
arrival and had not seen him anywhere 'near the examination material room. The
security guard and cleaning individual were added to the security agreement. There
were no discernable improvement in the performance of any applicant, nor other
indication of any applicant having obtained knowledge of the examination content
following the incident. The chief examiner determined that examination material security
had been inadvertently breached but that no examination compromise had occurred.
V. Management Meetings
- X1
Exit Meeting Summary
The examiners presented partial inspection results to members of the licensee
management at the conclusion of the onsite inspection on November 19,1998. After
the graoing of the written examinations and analysis of the results, the chief examiner
held a final exit with the licensee telephonically on December 18,1998. The licensee
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acknowledged the findings presented.
The licensee did not identify as proprietary any information or materials examined during
the inspection.
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ATTACHMENT 1
SUPPLEMENTAL INFORMATION
PARTIAL LIST OF PERSONS CONTACTED
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Licensee .
- M. Jones, Manager, Operations
R. Sandstrom, Manager, Training
K. Rauch, Supervisor, Operations Training
T. Frey, Compliance
T.Vogt Operations
. D. Axline, Licensing
L. Germann, Training
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ATTACHMENT 2
FACILITY LICENSEE POST-EXAMINATION COMMENTS
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RO Exam Comments
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COMMENT #1
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RO Examination Q esuon 7
(SRO9)
The question stem referenas SO23-3 3.27,3 as do the possible answers. The actual procedure that should have
been referenced is SO23-3-3.23, Emergency Diesel Generator Monthly Surveillance. The given procedure, SO23-
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3-3.27.3 is not used to perform the AC Sources check that is required by the given scenario. The correct answer
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was not prended. Southern California Edison believes there are no correct answers to this question.
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- Delete the question.
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NUCLEAR ORGANIZATION
SURVEILLANCE OPERATING INSTRUCTION 5023-3-3.23
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UNITS 2 AND 3
REVISION 14 TCN
14-2
PAGE 72 0F 88
ATTACHMENT 7
A. C SOURCES VERIFICATION (MODES 141
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OBJECTIVE
To provide verification that sufficient AC Sources are available to
the IE 4.16kV Busses when any combination of Offsite Circuits,
Onsite Circuits, and Diesel Generators are Inoperable. This
attachment satisfies Surveillance requirement of Tech. Spec.
LC0 3.8.1 AC Sources Verification.
UNIT
MODE
(1-4)
DATE
TIME
PERF. SY
1.0
PREREOUISITES
INITIALS
1.1
Verify this document is current by checking a controlled
copy or by using the method described in 50123-VI-0.9.
1.2
List the reason for performing this attachment (e.g., Diesel
Generator 2G002 Inoperability).
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2.0
AC SOURCES VERIFICATION
2.1
If this attachment is being performed prior to declaring
a piece of equipment Inoperable, then assume the
equipment is Inoperable when performing the attachment.
2.2
If the specific equipment Inoperability has placed both
Units in action statements, then a separate attachment
will have to be performed for each Unit.
2.3
If a Diesel is Inoperable, then determine if the cause of
the Diesel Generator Inoperability may exist on the other
Diesel Generator (s).
2.351
If the cause of the Diesel Generator
Inoperability exists on the other Ofesel
Generator (s),thendeclaretheaffected
Diesel (s) Inoperable,
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2.4
If desired use the last page of this Attachment to assist
in performance of this Attachment.
ATTACHMENT 7
PAGE 1 0F 7
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NUCLEAR ORGANf2ATION
SURVEILLANCE OPERATING INSTRUCTION
S023-3-3.27.2
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UNITS 2 AND 3
REVISION 10
PAGE 4 0F 26
--ATTACHMENT 1-
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WEEKLY ELECTRICAL BUS SURVEILLANCE - Both Units in Modes 1 thru 4
OBJECTIVE
To verify Operability of the offsite transmission network, onsite Class 1E
distribution system (except the diesel generators), and the onsite DC systems
as required by-the Technical Specification Surveillance requirements:
SR 3.8.1.1, SR 3.8.4.1, SR 3.8.7.1, SR 3.8.9.1.
To verify the functionality of the Spent Fuel Pool Cooling System power
availability as required by the Administrative Technical Specification.
UNIT 2 MODE
UNIT 3 MCDE
DATE
PERF. BY
1.0
PREREOUISITES
INITIALS
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VERIFY this document is current by checking a controlled
copy or by using the method described in 50123-VI-0.9.
1.2
DETERMINE the performance requirements of this attachment,
as follows:
SRO Ops.
O This Attachment is being performed for a scheduled
surveillance.
O This Attachment is being performed for operability
verification. LIS) the Components and Sections Steps
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to be performed. After approval, then CIRCLE N A for
the remaining unused steps.
COMPONENTS
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SECTIONS / STEPS
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OPERABILITY VERIFICATION
PREPARED BY:
Control Room Operator
OPERABILITY VERIFICATION
APPROVED BY:
SR0 Ops. Supv.
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ATTACHMENT 1
PAGE 1 0F 7
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1R0 EXAMINATION QUESTION #9'
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S023-5-1.8 is thelreference for "A" to be a correct answer.-
"C" is
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Jalso correct based'on. Technical-Specification 3.4.6 and 3.4.7, which-
. requires'the-RCS LOOP to be. operable.- S6uthern California Edison
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believes" there-are two correct answers to this question,
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iAccept answers A & C:
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NUCLEAR ORGANIZATION
INTEGRATED OPERATING INSTRUCTION
S023-5-1.8
UNITS 2 AND 3
REVISION 9
PAGE 86 0F 91
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ATTACHMENT 13
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9.o kcP OPERATION
9.1
With at least one RCP operating, reverse flow will be present in the
idle loop.
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9.2
The loss of RCP heat will affect cooldown rate. Consequently, Tcold
should be maintained 2125'F to prevent entering the restrictive heatup
and cooldown limitations that apply when s120*F.
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9.3
When securing RCPs, it may be necessary to reduce PZR heater output due
to the reduction of PZR Spray Valve bypass flow.
9.4
Due to insufficient Pretsurizer heater capacity, it may be necessary to
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secure all RCPs and main spray prior to initiating Auxiliary Spray.
Otherwise, loss of NPSH for the RCPs could occur.
(Ref. 2.3.17)
9.5
Pressurizer insurge reay occur when securing the last RCP. This is
caused due to the lower RCS flow across the core. As Core Exit
Temperature rises, the RCS will swell into the Pressurizer. Adjusting
letdown flow will help minimize this insurge.
9.6
Indicated Tcold will initially rapidly lower in any loop where 500 is
injecting, if the RCP operating in that loop is stopped or when the
last RCP ts stopped. This is due to cooler SDCS injection water
flowing over the loop Tcold temperature element.
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9.7
If any RCPs are operating, then the Tcold associated with an operating
RCP should be used for RCS temperature monitoring.
9.8
WhentherearenoRCPsoperating,thenTR-0351A(T351X),SDCCombined
Outlet Temperature, should be used for Tcold temperature monitoring.
9.9
I.E RCPs are running, IllE!( one RCP shall remain in service until
completing RCS boration to Mode 5, or refueling concentration and other
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forced circulation dependent parameters are met (e.g., hydrogen,
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peroxide,etc.).
9.10 When the last RCP is stopped, then RCS Thot indications (T351Y and
CETs) will begin to rise due to the increased time coolant is in the
Core region (i.e., no RCP forced circulation). Consequently, SDCS
flowrate should be adjusted to maintain RCS Tcold TR-0351A (T351X) at
the desired teniperature.
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ATTACHMENT 13
PAGE 6 0F 11
T0 *d
CP:li
86. Of ^0N
9122-891-6v6:xe3
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RCS Loops--MODE 4
3.4.6
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3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.6 RCS Loops'--MODE 4
Two loops or trains consisting of any combination of RCS loops
and shutdown cooling (SDC) trains shall be OPERABLE and at least
one loop or train shall be in operation.
NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and SDC pumps may be
de-energized for s I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
a.
No operations are permitted that would cause
reduction of the RCS boron concentration; and
b.
Core outlet temperature is maintained at least 10*F
below saturation temperature.
2.
No RCP shall be started with any RCS cold leg
temperature 5 256*F unless:
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a.
Pressurizer water volume is < 900 ft , or
b.
Secondary side water temperature in each steam
generator (SG) is < 100'F above each of the RCS cold
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leg temperatures.
____________________________________________________________
APPLICABILITY:
MODE 4.
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SAN ONOFRE--UNIT 2
3.4-18
Amendment No. 127
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RCS Loops--MODE 4
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3.4.6
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ACTIONS
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CONDI' TION
REQUIRED ACTION
COMPLETION TIME
A.
One required RCS loop
A.1
Initiate action to
Immediately
restore a second loop
or train to OPERABLE
AND
status.
Two SDC trains
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B.
One required SDC train
B.1
Be in MODE 5.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
AND
Two required RCS loops
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C.
Required RCS loop (s)
C.1
Suspend all
Immediately
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or SDC train (s)
operations involving
reduction r~ e'.S
boron conce ; ration.
AND
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train in operation.
C.2
Initiate action to
Immediately
restore one loop or
train to OPERABLE
status and operation.
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SAN ONOFRE -UNIT 2
3.4-19
Amendment No. 127
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RCS Loops-Mn0E 4
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3.4.6
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' SURVEILLANCE REQUIREMENTS
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SURVEILLANCE
FREQUENCY
Verify at least one RCS loop.or SDC train
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
is in operation.
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Verify' secondary side water level in
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
required SG(s) is it 50% (wide range).
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Verify the second required RCS Loop or SDC
7 days
train is OPERABLE.
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SAN ONOFRE--UIIIT 2
3.4-20
Amendment No. 127
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RCS Loops--MODE 5. Loops Filled
3.4.7
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3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.7 RCSLoops3 MODE 5, Loops Filled
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At least one of the following loop (s)/ trains listed below
shall be OPERABLE and in operation:
Reactor Coolant Loop 1 and its associated steam
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a.
generator and at leas'. one associated Reactor Coolant
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Pump;
b.
Reactor Coolant loop 2 and its associated steam
generator and at least one associated Reactor Coolant
Pump;
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c.
Shutdown Cooling Train A; or
d.
Shutdown Cooling Train B
One additional Reactor Coolant Loop / shutdown cooling train
'shall be OPERABLE, or
The secondary side water level of each steam generator shall
be greater than 50% (wide range).
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NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and pumps providing
-
shutdown cooling may be de-energized for 51 hour5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> per
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
i
a.
No operations are permitted that' would cause
reduction of the RCS boron concentration; and
b.
Core outlet temperature is maintained a' least 10*F
t
below saturation temperature.
2.
One required SDC train may be inoperable for up to
'
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other
SDC train or RCS loop is OPERABLE and in operation.
3.
One required RCS loop may be inoperable for up to 2
hours for surveillance testing provided that the other
RCS loop or SDC train is OPERABLE and in operation.
(continued)
)
SANONOFRE-bHIT2
3.4-21
Amendment No. 127
-
- . - . . ~ . .
.
. ~ . - _ - - _ - . . .
. . - - . . . .
. ~ - - . . . .
. - . . .
-
,
.
.
RCS Loops--MODE 5,. Loops Filled
.
3.4.7
i
.
_,
9%
)
!
NOTES
(continued)---------------------
,
t
4.
No reactor coolant pump (RCP) shall be started with one
or more of the RCS cold leg temperatures s 256*F unless:
1
a.
The pressurizer water volume is < 900 ft3 or
b.
The secondary side water temperature in each steam
generator (SG) is < 100*F' above each of the RCS cold
leg temperatures.
'
E
5.
- A containment spray pump may be used in place:of a icw
pressure safety injection pump in either or both.
shutdown cooling trains to provide shutdown cooling flow
i
-
provided the reactor has been suberitical for a period
> 24. hours and the RCS is fully depressurized and vented
in accordance with LCO 3.4.12.1.
i
6.
All SDC trains may be removed from operation during
,
!
planned heatup to MODE 4 vnen at least one RCS loop is
,
in operation.
,
l
1
APPLICABILITY:
MODE 5 with RCS loops filled.
{
,,
]
..
,
,
ACTIONS
~
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.
Less than the required
A.1
Initiate action to
Immediately
SDC trains /RCS loops
restore the' required
SDC trains /RCS loops
-
to OPERABLE status.
.A#.Q
08
'
Any.SG with secondary
side water level not .
A.2
Initiate actior to
Immediately
within limit.
restore SG secondary
side water levels to
within limits.
1
(continued)
i
i
SAN ONOFRE--UNIT 2'
3.4-22
Amendment No. 127
I
iL
!
, - _ .
,
_
-
-
. ...
.
. . .
--
_ , - _
..
. . _ _
.
_ . . . .
._m.
_ . . _ _ _ _ . . . _ _ _ . _ . . _ . . _ . . - - _ _ _ _ _ . . . _ _ . _ . _ _ . _ . . _ . _ _ _ _ _ .
..
.
.*
'RCS Loops--MODE 5, Loops Filled
-
.
3.4.7
i
_ _ ,
"}
. ACTIONS (continued)
~
CONDITION
REQUIRED ACTION
COMPLETION TIME
,
.
,
B.
No SDC train /RCS loop.
B.1
Suspend all
Immediately
'
in operation.
operations involving
"
,
reduction in RCS
boron concentration.
.
'
.
AND
B.2
Initiate action to
Immediately
restore required SDC
-
train /RCS loop to
-
operation.
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
)
Verify at least one RCS loop or SDC train
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
~
.is in operation.
_
'
Verify required SG secondary side water
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
'
level is 2 50% (wide range).
!
i
Verify the second required RCS loop, SDC
7 days
tFain or steam generator secondary is
'
.
l
J
.
,
"
p.
SAN ON0FRE--UNIT 2
3.4-23
Amendment No. 127
..
.
-
-
_ ..
_ ,._
.
_. _.
. . _
_ . . . _ _ _ _ _ . - _ _ . . .
-_____..-m.
__
. . _ _ . _ _ . ~_
.
.
,
.
COMMENT #3
' RO Examination Mr.14
(SROI8)
De generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. The Reed
.,
Switch Position Transenitters, RSPT's, actually sense the CEA's position. The Control Element Assembly
Calculator, CEAC, uses the input Dom the RSFT and sends a signal to the alarm. Both components are =ri~l to
generate a deviation alarm. Southern California Edison believes there are two correct answers to this question.
Accept anu a B & C
l
i
!
,
.
NUCLEAR ORGAtl!ZATION
ALARM RESPONSE INSTRUCTION 5023-15-50.Al
.,
UNZTS 2 AND 3
REVIS!0N 2
PAGE 710F 76
-
ATTACHMENT 2
.
APPLICABILITY
PRIORITY
REFLASH
ASSOCIATED WINDOWS
Modes 1-3
AMBER
NO
NONE
IhlTIATING
NOUN NAME
SETPOINT
VALIDATION
PMS 10
LINK #
DEVICE
INSTRUMENT
U2/U3
2(3)LO91,- CEAC 1
Control Element
5 Inches
NONE
DEVIAR56
641/663
j!L
or CEAC 2
Assembly Deviation
1.0
REOUIRED ACTIONS:
1.1 Position the CEOMCS Mode Selector Switch on 2(3)CR50 to 0FF.
1.2 Verify which CEA is misaligned and the amount of misalignment, by
observatipnofthefollowing:
CEAC display CRT
CEAC remote operators modules
e PMS alarms
e PMS readout
2.0
CORRECTIVE ACTIONS:
SPECIFIC CAUSES
SPECIFIC CORRECTIVE ACTIONS
2.1 Misaligned CEA
2.1 After the misaligned CEA has been
determined, lhmt:
2.1.1
Notify the SR0 Ops. Supv.
2.1.2
Realign the CEA per 5023-3-2.19,
Section for Manual individual
R
Operation.
2.2 Slipped or Dropped CEA
2.2 GO TO S023-13-13, Misaligned t.ontrol Element
Assembly.
3.0
ASSOCIATED RESPONSES:.
3.1 NotifytheCRS/SSandtheSTAtoreviewTech. Specs.LCO3.1.5and
LCS3.1.105,andinitiateanEDMR/LC0AR,asrequired.
~g.,
~
,
,
.
.
. . .
. . .
.___.
._
..
.
. - _ . .
_. .
m._
, _ _ _ - _.
_ _ . -
-,
. _ _.
. _ _ _ . . _ _ . . _ - . . .
.
- *
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NttCLEAR ORGANIZATION
.
SYSTEM DESCRIPTION $0.$023 710
UNITS 2 AND 3
REVIS!CN 3
PAGE 72 0F 76
.._
fl$.iL8 E !! - CONTROL ELEMENT ASSEuBLv sultGm00R REED TWITM
P95fTf0N TRAN!NITTER TIGNAL AtsfGNWENTS
.
l
h EX4CRE CHANNEL
h
RSP7
at SPT \\
'VA
23CEAS
h
/ RSP7
22CEAS
_
2
i
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/
g
22 CEAS
CEAS
22 CEAS
"f
45 CE,a5
g
d$ C(A$
_MQEh
22 CE.
'
N
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y
2
/
23 Co.1
_
_
_
ISCLATION
_
,
/
"
"
CALCULATOR
CALCULATOR
\\
CEA POSIT:Ch
NO 1
MO. 2
CD #CSITCN
N[ikif,k
ctUatc=
a=S{
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A COR E
8 CORE
CCCRE
OCORE
PROTECT CN
PROTECTION
PROTECTCh
PROTECT:ON
.
'.
CALCULATCR
CALCut/ TOR
CALCULATOR
CALCULATCR
I
I
c---
.
$
I
OPERATOR $
OPERATCR'S
OPERATOR 4
CPERATCR'S
MODULE
MOCULE
MOCULE
WOOULE
CRTCISPLAY
NOTES.1. SIGNAL FRCM CEA 213 CCNPECTED TO CPC's A AND C, BUT IT 13 NOT USED AS A TARGET CEA.
2. SIGNAL l'MCM CEA 313 CONNECTED TO CPC's 8 AND C,8tf7 ITIS NOT USE3 A5 A TUtGET CEA.
3 SIGNALS FRCM 23 CENs ARE CONNECTED TO EACH CPC.
-
CNLY 22 CF THE 23 SICNALS ARE USEQ AS TARGET CEAs.
t
.
.
.v
5
,
_ -
._
m
__ _
. . . _ . _ _ . ._ _. _ _ . .
_ . . . - - _ . _ . _ _ _ _
_
_ . . .
?
!- - .,
COMMENT # 4
.
- -'
- RO EXAMINATION QUESTION 26.
i
Electncal drawing 30718 Rev. 9 shows the automatic makeup circuitry has been deleted. Southern California Edison believes there
. are no correct answers for this quesuon and the question should be deleted from the examination.
.
Delete'4ueshon.
1
-
l
l
- -
1
- - - .
.
.
...
. ._
_.. .-... _.__ . . _ _ . . . _ .
, _ . . . . . . _ . . _ _ . . . . _ _ _ _...
.
.
.
' COMMENT #5-
'
c-R0 EXAMINATION QUESTION'#27
-
-(SR0 28)
'
i
-Answer "B" is correct because-pressurizer spray valves are open at-
l
-
- 2300 psia.. Answer "D" is also correct because the backup heaters turn
i
off.at 2225 psia and a backup signal to turn.off the backup heaters at
2275 psia._' Southern California Edison. believes there are two correct
l
answers _to this question.'
)
.
L
1
Accept answers _B 8 0.:
,
1
a
'
,
-
1
1
'
<
d
4
4
s-
.-
._
_
_ _
_ _
._
. . _
_ . _
NUCLEAR ORGANIZATION
UNITS 2 AND 3:
SYSTEM DESCRIPTION SD.SO23-360
.
REVISION S
Page 168 of 205
,
q
FIGURE 111-5 PRESSURIZER PRESSURE CONTROL SYSTEM EILdCK
.s
,
.
100X
100Y
1? HEATER
'
~ PMS/CFMS
PMS/CFMS -,1E HEATER
<
"
~ STEAM EYPASS SYSTEM
STEAM EYPASS SYSTEM -
<
i
E/S
'
RED
2500 -
GREEN
E!S
'
-
Lp00 -l
P lA
HS-100A
2200/2225
RECORDER
220C/2225
IX
PR0100-A/c
Hl/LO
E/S
i
RED
$y* PMS
AuRM
50A14
- -
(2275/2175)
E/S _
c
.~
S
ONIOFF
,
(2IOb25)
E/S
i
-
c
Hi HI
'
'
0
=
'!
8/U HEATER
TRIP (2340)
-
c
1
j
e
,p
0100
>
3ROPORTIONAL
l
1
DE-ENERGlZE
jEATERS
~
~ E/UHEATERS
(2275)
l
,
-
%HCC
4LDM
J
o
9,
y
a o
V4 Mi
M V4
s trM%iiT' "
--
l
VALV PCSmON
Hl/LORM g
7chg33
w=;grg!;
tsme-
hoQ14
(2275/2175)
'#' MEH9f#
1
l
PV 100A
PV 1008
A . e ren
V4 .gsvr
.
.
p e
I
.
.-
_ . _ . _ . . _ .
.. . _ _ _ . . _ _ _ _ . . . _ _
. _ . . . _ . _ . _ . _ _ . . _ . _ _ . _ . _ . . _ . _ _
. . _ - . . _ , _
.
.
.
4
.-
COMMENT #6
RO Examination Question 37
(SRO37)
Procedure, SO23 12-7. Ims of Offhite Power / Loss of Forced Circulation, floating step 2 states that Thot and
CET's are compared as are Thot and Tc are not rising to verify natural circulation is occurring. The S/G pressure
can be used to correlate to Tc there fore answer B is also etnTect. Soutirr's California Edison believes there are two
correct answers to this question.
Accept answers A & B
,
.
_ _ _ _ _ _ _ _
,
NUCLEAR CEGANIZA~ ION
UNITS 2 AND 3
EMERGENCY CPERATING INSTR'JCTICN
i
REVISION 15
5023-12-7
.
(*
ATTACHMENT 2
PAGE 30 CF 122
'j
LC'S OF FORCED CIRCULATICN/ LOSS CF 0FFSITE PCW
s
-
FLOATING STEPS
ACTION /EXPECTE0' RESPONSE
i
RESPONSE NOT OBTAINED
NOTE:
Lcw flow during Natural Circulation slows RCS
response to temperature changes.
Lec: transit
time rises to between: S minutes and 10 minutes.
FS-2
HONITOR Natural Circulation
_
Established:
CHECX all RCPs - stopped.
a.
GO TO FS-4, MONITCM RCP Operating
a.
Limits.
b.
CHECXatleastoneS/G
b.
GO TO S023-12-9, FUNCTIONAL RECOVERY
operating:
AND
1) SBCS - operating
,,
INITIATE S023-12-9 Attach ent 8
RECOVERY . HEAT REMOVAL.
ADV . operating.
AND
.
2)
Feedwater - available.
CHECX operating icop AT - less
c.
than SS*F.
IF any criteria c through g NOT
o
satisfied.
d.
CHECX Tc and Th - NOT-rising.
THEN
-
CHECK Reactor Vessel Level
!
e.
(Plenum) - greater than or
MAXIMIZE S/G 1evel - less than
.
'
equal to 100%t.
80% NR.
QSPDS page 622
RAISE available S/G steaming
l
.
CFMS page 312
rate.
I
Attachment 4.
}
i
RAISE Core Exit Saturation Margin
.
CHECX operating loop Ts and REP
- greater than 20*F:
j
f.
i
CET - within 16*F:
QSPDS page 611
!
QSPDS page 611
i
CFMS page 311.
!
-
/
,
.
3
P
ATTACHMENT 2
PAGE 4 0F 29
. . . .... - .....
.
-
. . - . , . . _.
.. . ..
..
2As- /
- NUCLEAR CRGANIZAilCN
UNITS 2 AND 3l
EMERGENCY CFERATING INSTRUCTICN S023-12
.
-
REVISION 15
i_g.. -
ATTACHMENT 2
PAGE 31 0F17)
.
-
"
'
LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWE
.
FLOATING STEPS
ACTION / EXPECTED RESPONSE,
_ RESPONSE NOT OBTAINED
!'
'FS-2
MONITOR Natural Circulation
Established:
(Continued)
g.
CHECK Core Exit Saturation
Margin - greater than 20*F:
IF any criteria c through g NOT
o
satisfied,
i
QSPOS page 611
THEN
CFMS page 311.
,
MAXIMIZE S/G level - less th'an
.-
Sok NR.
\\
.
RAISE avcilable S/G steaming
.
rate.
l
RAISE Core Exit Saturation Margin
.
.
- greater than 20*F:
'
'
QSPOS page 611
'
i
CFMS page 311.
.
.
I
}'
L
.
-s==
ATTACHMENT 2.
PAGE 5 0F 29
- N>.'41,'- }..' Wif # !y'l!--{R'[,jiiI,51 y,3ITEI.0,69,7s,(S$359,M975UYN'
N
,
-
.
. . . - - .
.
...
.
.
.-
. _ - . - . . .- -
. - . .
_ .
. - - - -
.
..
.
.
. COMMENT #7
RO Examination Question 46
(SRO43)
Answer B is correct based on the strictest interpretation ofinunedsately before and aAer a trip. Immediately before
the trip, S/O level exmeded 85% causing a High Level Override, HLO, signal to be applied to the main and bypass
foM regulating valves causing both to close. The valves both stay closed until level decreases below 85% at which
time the HLO signal clears and the valves go to either the Reactor Tripped override, RTO, position or to the
position set by the demand from the feed water argulating control system. With a reactor trip, an RTO signal is
sent which as soon as the HLO condition clears seconds aAer the trip due to normal shrink of water levels, the
RTO signal is applied and the bypass valve opens to 50%, answer C. Southern California Edison believes there are
two correct answers to this question.
'
Amept answers B & C
l
i
t
I
- _ - ___________ _ _ _ _ _
.
COMMENT #8
.
R0 EXAMINATION QUESTION #54
.
5023-6.2.9 states " Select the Bus Transfer Control AUT0/ MANUAL
iWitch to AUTO after completion of the breaker manipulations
that return the bus to its " Normal" configuration." Having a
bus on the tie-brk is not a normal configuration.
So the
AUT0/ MANUAL switch would not be placed in the AUTO position.
Southern California Edison believes there are no correct answers
to this question.
Delete question
I
_ _ _ _ . _ _
.
-
.
-
--
,
-
.
- * T.?MT.".C*T ~. .? ^
,
NUCLEAR ORGANIZATION
OPERATING INSTRUCTION
S023-6-2
.
UNITS 2 AND 3
REVISION 5
PAGE 7 0F 26
I
9..
)
6.0
PROCEDURE (Continued)
'
l
6.2'.8
If the INCOMING 4160V source is a bus tie or diesel
l
generator output breaker, then open the RUNNING breaker,
'
if desired.
u
6.2.9
Select the Bus Transfer Controls AUT0/ MANUAL switch to
.
'
AUTO after completion of the breaker manipulations that
return the bus to its " normal" corfiguration.
6.2.10
Remove the synchronizing circuit from service by
depressing SYNC pushbutton for the INCOMING breaker.
6.2.11
Remove the synchroscope from service by selecting the
respective key-operated Master Control switch to 0FF.
6.2.12
Clear any annunciator alarms resulting from the transfer
operation.
'
NOTE:
For Bus Transfers using the Bus Tie Breakers,
the synchroscope and synchronizing circuits can
.
only be in service on one Unit at a time. After
the first Bus Tie Breaker (regardless of Unit)
is operated, its associated synchronizing
i
circuit must be de-energized and its
'
syr.chroscope removed from service prior to
.-
)
starting the evolution on the remaining Bus Tie
--'
Breaker.
6.2.13
For IE 4160V Bus Tie transfer schemes, perform the
'
following:
.
.1
Starting from the Unit which is to SUPPLY power:
NOTE:
The INCOMING Voltage and Frequency are sensed
directly from the Tie Bus. The " BUSES
PARALLELED" alarm logic is satisfied on a Unit
'
when BOTH Units Bus Tie Breakers are closed AND
either a Reserve Aux Transformer or a Unit Aux
Transformer Power Source breaker is closed.
When BOTH Bus Tie Breakers are closed A_ND a
Transformer breaker on each Unit is Closed, BOTH
Units will have BUSES PARALLELED alarms
annunciated.
.1.1
Place the synchroscope in service per
Step 6.2.3.
!
l
',
I
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w_-
(
+-
1
.
.
..
.
..
COMMENT #9
RO Exanunation Question 74
(SRO69)
The question setup has the VCr pressure at 40 psig. The head of the Refueling Water Storage tank, RWST, is 70'
>
of H2O or 30.3 psig(70' x 0.433 psi /A) In the scenario provided the head of the VCT with the over pressure, will
keep the check valves from the RWST closed. Southern California Edison believes there are no correct answers to
this question and it should be deleted -
- Delete this Twiaa.
j
)
)
l
i
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-,
m
m
. .
--
.
- .
-
-
..
_ _ _ - . . . - . - , - . - . .
. .
-.
.
. -
. .
.
..
l
'
l
COMMENT #10
l
' RO Exanunation Quesuon 79
!
(SRO 75)
\\
!
!
Question was based on old Tavg program. New program has normal pressurizer level of 48% at 100% power.
j
This is based on the reduced Tc program of $48 deg F @ 100% power. The lower Tc at full power equates to a
Tave of $74 deg F. Per the attached reference. the expected level would be 48% and no additional charging pumps
- would be operating. Southern California Edison believes there are no correct answers for this question and the
'
question should be deleted from the examination.
Delete Question.
'
,
h
,
I
l
l
4
t
!
!
l
I
l
.
_
.
. . .
-
- .- .
. - -
- -
.
.
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4
NUCLEAR ORGANIZATION
OPERATING INSTRUCTION
'
,
REVISION 8
S 2 -3 1.10
l
UNITS 2 AND 3
.
l
ATTACHMENT 5
AuE 30 0F 34
.
PRESSURIZER LEVEL CONTROL PROGRAM
.
e
l
.
I
l
I
l
70
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. . . . . ;.
..
. .
. . . . .. . . . .
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.
.
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.
.
.
.
.
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. . . . .. . . . .
... ... ... ... . ........... . .. . . . . . . . . . . . . . . . . . ...
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. . .
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,
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,
30
.
s s c s s
s s s s
s s s . i
s .
,
s . s
s
s
544
550
560
570
58Q
590
RCS AVERAGE TEMPERATURE ( F)
010 -1.C H T
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D10 8.wS1
ATTACHMENT 5
PAGE 1 0F 1
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. . _ _ - _
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COMMENT #11
RO EXAMINATION QUESTION 85
' (SRO 84)
The trends given are inconclusive as to whether vacuum has stabilized at 3.3", if assumed vacuum is stable at 3.3 "
- no further action will be required and answer B would be correct, ifit is assumed vacuum will continue the current
- irend, the listed action of"D" could be taken to return the plant to a more stable condition. Southern California
Edison believes "B" & *D" are correct answers.
Accept B & D.
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NUCLEAR ORGANIZATION
ABNORMAL OPERATING INSTRUCTION S023-13-10
.
UNITS 2 AND 3
REVISION 2
PAGE 7 0F 13
.
ACTION /EXPiCTED RESPONSE
RESPONSE NOT OBTAINED
4 Actions for loss of vacuum due to
Condenser fouling.
,
i
CAUTION
During periods of heavy influx, rapid and aggressive action may be required in :
@
order to avoid a Unit trip. ' Power may need to be reduced in order to,
,
t
Bumpand/orStopCirculatingWaterPumpsontheCondenserquadrantswith l
e
the highest differential pressures
Maintain Condenser backpressure < 3.5" Hg
j
l
a. REDUCE Regctor power to 75% TO
854.
b. BUMP Circulating Water Pump (s)
per direction of Shift
Superintendent,
c. VERIFY backpressure < 3.5" Hg _
c.
1) REDUCE Reactor power
and stable.
to s 65%.
'
2) STOP two Circulating Water
Pumps on opposite ends of
the Condenser.
j
,
3) INITIATE isolating stopped
pumps per 5023-2-5,
Attachment for Stopping a
Circulating Water Pump Due
to Fouled Condenser
.
Tubesheet/High AP/ Debris
f
'
Removal .
4) IF not < 3.5" Hg and stable,
THEN REDUCE Reactor Power
as necessary to establish
backpressure < 3.5" Hg and
stable.
5) GO TO Step 5.
d. EVALUATE stopping pumps based on
Waterbox differential pressure
and pump vibration,
e. GO TO Step 6.
- 2
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,
.
COMMEW # 12
RO Examination Questson 93
(SRO93)
Answer C is conect based on HV9217 and HV9218 being open and providing a direct path from inside
containment to the outside, in this case to the VCT. Answer D is correct bcause given this event, Controlled Bleed -
,
O\\ff flow is routed to the quench tank via a relief valve when HV9217 and HV9218 are closed. With HV0514 and
HV0515 being failed open, a dLect path for RCS water exists from the quench tank to the chemistry sample sink.
Therefore, acapt both answers C and D
Accept answers C & D
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COMMENT # 13
RO EXAMINATION QUESTION 95
(SRO.%)
Both answers A & B will cause a FTS event to occur if the operator fails to initiate release of steam from E088
S/G. "A"is correct based on failing to steam the good S/G to establish a heat sink. "B"is also correct in that HPSI
throttle /stop criteria will NOT be met because of the unavailability of the S/G as stated in the stem (step FS4 a.1
requires operating S/g with an ADV operating), and continuing to inject water into the RCS will increase pressure
also leading to FTS event. Southern California Edison believes there are two correct answers to this question.
Accept answers A & B.
.
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NUCLEAR CRGANIZATION
EMERGENCY ~0PERATING INSTRUCTION S023-12-5
UNITS 2 AND 3
REVISION 15
PAGE 43 0F 143
ATTACHMENT 2
.
i
EXCESS STEAM DEMAND EVENT
'
-
FLOATING STEPS
' ACTIONS / EXPECTED RESPONSE'
RESPONSE NOT CBTAINED
FS-6
CHECK HPSI Throttle /Stop
,
Criteria:
a.
CHECKatleastone.S/G
a. GO TO S023-12-9, FUNCTIONAL RECOVERY..
operating:
AND
'1)
SBCS -' operating.
INITIATE 5023-12-9, Attachment 8,
n
RECOVERY - HEAT REMOVAL.
,
ADV'- operating.
-
AND
2)
Feedwapr - availa.ble,
b.
CHECK PZR level
o =IF any criteria of steps b through d
NOT met,
-- greater.than'30%
THEN
'AND
OPERATE Charging and HPSI systems
,
.
- NOT lowering.-
as necessary to maintain
(
Throttle /Stop criteria
c.
CHECK Core Exit Saturation
- satisfied.
'
Margin . greater than 20*F:
THROTTLE Loop Injection Valves.
.
.QSPDS page'611
CFMS page 311.
ENSURE auxiliaries to SI pumps:
.
a)
Electrical power to pumps and
j
valves.
)
b)
Proper system alignment.
,
c) CCW flow.
d) HVAC.
. ATTACHMENT 2
PAGE 13 0F 43
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NUCLEAR GRGANIZATICN
EMERGENCY OPERATING INSTRUCTION S023-12-5
- .-
UNITS 2'AND 3
REVISION 15-
PAGE a2 CF 123
"
ATTACHMENT.2
...
EXCESS STEAM DEMAND EVENT
,
s
.
FLOATING STEPS-
-
. ACTIONS / EXPECTED RESPONSE
RESPONSE NOT OBTAINED
FS-6-
CHECK-HPSI. Throttle /Stop
Criteria:
(Continued)
d.
CHECK Reactor Vessel Level'
.o-
IF any criteria of steos b thrcugn. d
~
(Plenum)-- greater than or
NOT met,
,
. equal to 1004:
-THEN
'
.0SPDS page 622
CFMS page 312
OPERATE Charging and HPSI. systems
.
Attachment 5.
as necessary to. maintain
{
Throttle / Step criteria
,
- satisfied.
THROTTLE Loop Injection Valves.
.
ENSURE auxiliaries to Si pumps.:
- .
.
a)
Electrical power to pumps and
valves.
b)
Proper system alignment.
c)
CCW flow,
d)
HVAC.
e. ; VERIFY RCS borated - greater-
e. MAINTAIN Emergency Boration
.
than ~ Technical Specification ~
- at least 40 GPM.
Shutdown Margin for T vt > 200*F
A
per Operations Physics Summary
-Figure 2.3-1,
j
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'-
RCS Cooldown - NOT in progress.
f,
THROTTLE ~0R STOP HPSI as
required one train at a time.
4
%
g.
STOP charging pumps as required
one at a time,
,
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ATTACHMENT 2
PAGE 14 0F 43
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NUCLEAR ORG'ANIZATION
,
EMERGENCY OPERATING INSTRUCTION 5023-12-5
UNITS 2 AND'3
REVISION 15
PAGE 45 0F I:3 -
,
-
ATTACHMENT 2
. . -
EXCESS STEAM DEMAND EVENT
'
FLOATING STEPS
ACTIONS / EXPECTED-RESPONSE
RESPONSE NOT OSTA!NEO
'FS-6
CHECX HPSI Throttle /Stop
Criteria:
(Continued)
h.
MAINTAIN Criteria of steps a
~ through e - satisfied.
'.
CHECK Containment' pressure-
.i. 1)- ENSURE SIAS - actuated.
i
.less than 3.4-'PSIG.
2)
GO'TO FS-7, CHECX LPSI
Termination Criteria.
~
J.
CHECK PZR Level
J. INITIATE FS-22, ESTABLISH CVCS
- less than 80%.
Letdown Flow,
k.
RESET'SIAS per 5023-3-2.22,
.ESFAS OPERA-TION.
.
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9
ATTACHMENT 2
PAGE IS OF 43
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d
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COMMEW # 14
i
RO EXAMINATION QUESTION %
(SRO 97)
i
Answer "B" is correct based on the information given. However answer "A" is also correct based on procedure
I
SO23 12-7, Safety Function Status Checks, which requires subcooling > 20F or you are directed to the Functional
Recovery for not meeting natural circulation. Southern California Edison believes there are two correct answers to
this question.
- Accept answers A & B.
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NUCLEAR ORGANIZATION
UNITS 2lAND 3
EMERGENCY OPERATING INSTRUCTION 5023-12-7
f
- .
REVISION 15
-
ATTACHMENT 2
PAGE 30 CF 122
.
LOSS. OF FORCED CIRCULATION / LOSS OF 0FFSITE POWER
,
FLOATING STEPS
.
ACTION / EXPECTED RESPONSE
RESPONSE NOT OBTAINED
i
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NOTE:
Low flow during Natural Circulation slows RCS
'
response to temperature changes.
Loop transit
time rises.to between 5 minutes and 10 minutes.
.
'
.
FS-2
MONITOR Natural Circulation
Established:
)
a.
CHECK'all RCPs - stopped.
a. GO TO FS t, MONITOR RCP Operating
Limits.
l
b.
CHECK at least one S/G
b. GO TO S023-12-9, FUNCTIONAL RECOVERY
operating:
AND
..
.
1) SBCS - operating
'
INITIATE S023-12-9, Attachment 8,
.
RECOVERY - HEAT REMOVAL.
,
ADV - operating.
i
AND
2)
Feedwater - available.
I
c.
CHECK operating loop AT - less
o
IF any criteria c through g NOT
than'58'F.
satisfied,
j
-d
CHECK Tc and TH - NOT rising.
THEN
e.
CHECK Reactor Vessel Level
MAXIMIZES /Glevel-lessthan
.
,
(Plenum) - greater than or
80% NR.
t
equal to.1004:
RAISE available S/G steaming
.
QSPDS page 622
rate.
CFMS page 312
Attachment 4
RAISE Core Exit Saturation Margin
.
- greater than 20'F-
-
-f. 'CHECN operating loop Ts and REP
i
CET - within 16*F:
QSPOS page 611
CFMS page 311.
QSPDS page 611
.
CFMS page 311.
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ATTACHMENT 2
PAGE 4 0F 29
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NUCLEAR ORGANIZATION
t-
UNITS 2 AND 3
EMERGENCY OPERATING INSTRUCTION S023-12-7
REVISION 15
PAGE 31 0F 122
ATTACHMENT 2
.
LO.SOFFORCEDCIRCULATION/LOSSOF0FFSITEPOWER
FLOATING STEPS
i
ACTION / EXPECTED RESPONSE
RESPONSE NOT OBTAINED
FS-2
MONITOR Natural Circulation
Established:
(Continued)
g.
CHECK' Core Exit Saturation
o
IF any criteria-c through g NOT
Margin - greater than 20*F:
_ satisfied,
-
OSPOS page 611
THEN
CFMS page 311.
MAXIMIZE S/G 1evel - less than
.
804 NR.
RAISE available S/G steaming
i
.
rate.
- '
.
RAISE Core Exit Saturation Margin
.
- greater than 20*F-
,
QSPOS page 611
CFMS page 311.
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ATTACHMENT 2
PAGE 5 0F 29
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SRO Exam Comments
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COMMENT #1
i
SRO Examination Question 9
(RO7) .
The question stem references SO23 3-3.27.3 as do the possible answers. The actual procedure that should have
been referenad is SO23-3-3.23, Emergency Diesel Generator Monthlv Surveillance. The given procedure, SO23-
3-3.27.3 is not used to perform the AC Sources check that is required by the given scenario. The correct answer
was not provuled. Southern California Edison believes there are no correct answers to this question.
Delete the que. tion.
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COMMENT # 18
SRO Examination Question 93
(RO93)
Answer C is correct based on HV9217 and HV9218 being open and providing a direct path from inside
containment to the outside, in this case to the VC7 Answer D is correct because given this event, Controlled Bleed
O\\ff flow is routed to the quench tank via a relief valve when HV9217 and HV9218 are closed. With HV0514 and
HV0515 being failed open, a direct path for RCS water exists from the quench tank to the chemistry sample sink.
Therefore, accept both answers C and D
Accept answers C & D
,
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NUCLEAR ORGANIZATION
SURVEILLANCE OPERATING INSTRUCTION S023-3-3.23
UNITS 2 AND 3
REVISION 14 TCH
14-2
PAGE 72 0F 88
.
ATTACHMENT 7
A. C. SOURCES VERIFICATION (MODES 1-41
.,
OBJECTIVE
To provide verification that sufficient AC Sources are available to
the 1E 4.16kV Busses when any combination of Offsite Circuits,
Onsite Circuits, and Diesel Generators are Inoperable. This
,
attachment satisfies Surveillance. requirement of Tech. Spec.
LC0 3.8.1'AC Sources Verification.
'
UNIT
MODE-
(1-4)
DATE
TIME
PERF. BY
1.0
PREREOUISITES
INITIALS
1.1
Verify this document is current by checking a controlled
copy or by using the method described in 50123-VI-0.9.
1.2
List the reason for performing this attachment (e.g., Diesel
'
Generator 2G002 Inoperability).
2.0
AC SOURCES VERIFICATION
2.1
If this attachment is being performed prior to declaring
a piece of equipment Inoperable, then assume the
equipment is Inoperable when performing the attachment.
2.2
If the specific equipment Inoperability has placed both
Units in action statements, then a separate attachment
will have to be performed for each Unit.
2.3
If a Diesel is Inoperable, then determine if the cause of
the Diesel Generator Inoperability may exist on the other
Diesel Generator (s).
2.3;1
If the cause of the Diesel Generator
Inoperability exists on the other Diesel
- Geneiator(s), then declare the affected
Diasel(s).'noperable.
01
2.4
If desired u',e the last page of this Attachment to assist
!
in performan.:e of this Attach;aent.
'
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<-
ATTACHMENT 7
PAGE 1 0F 7
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NUCLEAR ORGANIZATION
SURVEILLANCE OPERATING INSTRUCTION
S023-3-3.27.2
UNITS 2 AND 3
REVISION 10
PAGE 4 0F 26
ATTACHMENT 1
WEEKLY ELECTRICAL BUS SURVEILLANCE - Both Units in Modes 1 thru 4
OBJECTIVE
To verify Operability of the offsite transmission network, onsite Class 1E
distribution system (except the diesel generators), and the onsite DC systems
as required by the Technical Specification Surveillance requirements:
SR 3.8.1.1, SR 3.8.4.1, SR 3.8.7.1, SR 3.8.9.1.
To verify the functionality of the Spent Fuel Pool Cooling System power
availability as required by the Administrative Technical Specification.
UNIT 2 MODE
UNIT 3 MODE
DATE
_
PERF BY
1.0
PRERE0VISITES
181114LS
1.1
VERIFY this document is current by checking a controlled
copy or by using the method described in 50123-VI-0.9.
1.2
DETERMINE the performance requirements of this attachment,
,
as f~ollows:
)
SR0 Ops.
'
[] This Attachment is being performed for a scheduled
surveillance.
j
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[] This Attachment is being performed for operability
verification. LIST the Components and Sections Steps
lR
to be performed. After approval, then CIRCLE N A for
the remaining unused steps.
COMP 0NENTS
SECTIONS / STEPS
OPERABILITY VERIFICATION
PREPARED BY:
Control Room Operator
OPERABILITY VERIFICATION
APPROVED BY:
=='
ATTACHMENT 1
PAGE 1 0F 7
.
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' COMMENT #2
SRO Examination Question 10
- Infrequently performed test can also be interpreted to be special tests. SO123-IT-1, Infrequently Performed Tests,
states that infrequently pedormed tests can also be performed under the special test procedure. Since 5023-0-23 is
also used to conduct short term valve lineup changes, it too is a correct choice. Southern California Edison believes
that there are two answers to this question.
Accept answers A & D
D
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NUCLEAR ORGANIZATION
GENERAL ORDER
50123-IT-1
Uti!TS 1, 2 AND 3
REVISION 4
PAGE 3 0F 16
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III. RESPONSIBILITIES (Continued)
F.
The Manaaement Desianee (see Definitions, Attachment A) exercises
continuous responsibility for Management Oversight. With the
approval of the Vice President, Nuclear Generation and/or the Senior
Vice President, Power Generation, may exercise Management Oversight
on a " spot-check" basis.
G.
The Test Soecialist (see Definitions, Attachment A) is a technical
resource to the supervisor who has operational responsibility for
conduct of the test or evolution.
H.
Licensed Goerattr_1 and Plant Manacement Staff (see Definitions,
Attachment A) have the responsibility to recognize tests and
evolutions which are (or should be) included in the IPTE List.
IV.
REQUIREMENTS
A.
Infrequently Performed Tests and Evolutions (IPTE) which take plant
personnel or equipment beyond the bounds of normal procedures,
traininqroperaTiTig-ban 6, ur exper3ence. ed (ich repr"eem
~~-
i
. siggificant safety or economic risk, require controlling documents
with enhanced development and review as outlined by this order.y
B.
Execution of IPTE activities require Management O M
Definitions, Attachment A) with clear direction, clear communication
of management expectations with respect to margins of safety,
,
expected plant response, termination criteria, and actions to be
l
taken in the event of unexpected results.
C.
Direction to licensed or non-licensed personnel with regard to the
operation of the plant shall be given only by personnel who possess a
SR0/R0 license a.nd are designated with responsibility for the safe
conduct of the evolution or test.
D.
The highest margin of safety shall be maintained throughout the test
or evolution exercising caution and conservatism, particulariy when
uncertainties or unexpected plant behavior is encountered.
E.
Req 61rements for test or evolution termination shall be clearly
defined, conhunicated, and understood by all persons involved with
the conduct of the test or evolution.
F.
IPTEs shall be conducted and documented using the IPTE Checklist
(Attachment D) per the Keypoints guidance (Attachment E).
G.
If it is necessary to change the IPTE controlling document prior to
or during use, ItiEH the associated affect on IPTE fntent (see
,
Definitions, Attachment A) and plent safety shall be considered.
'
'
lE the intent is affected, THER prior to starting or continuing with
the IPTE, the same approval level as the original controlling
l
document is required. Prior to any IPTE controlling document
'
changes, extreme caation and consideration should be exercised.
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NUCLEAR ORGANIZATION'
OPERATIONS DIVISION PROCEDURE S0123-0-23
UNITS 1, 2 AND 3
REVISION 5
PAGE 47 0F 62
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ATTACHMENT 4
1
XEYPOINTS FOR ABNORMAL ALIGNMENT / EVOLUTION
PAGE 10F .D]
]
COMPONENT: [1]
LOG NUM3ER
- [2]
-
PURPOSE OF ALIGNMENT:
Idl
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Procedure Change Required
O No
O Yes
!
Verify this~ document is current by checking a controlled copy or by using the
!
method described in 50123-VI-0.9. [5]
l
EFFECT OF ABNORMAL ALIGNMENT / EVOLUTION
NO
YES
l
Has it been addressed in a completed:
// O INDICATE document Type and
'
Number:
1
,
50.59 Safety Evaluation. E
O ATTACH PF(123)109-1, Unreviewed
e
Safety Question Screening
1
PF(123)109-1, Unreviewed Safety
/
Criteria
e
Question Screening Criteria? [6]
fj
1
Was SCE PF(123)l09-1 checked YES in
O
O 00 NOT PERFORM until Part II is
PART I? (Check N0 if form not used.)[7]
completed.
'
Does it:
O
O OBTAIN approval from Manager,
f
.
e Change the intent of the Operating
Operations prior to
Instruction, E -
impiementation.
.
Constitute an Evolution, E
e
Require a new or additional 50.59 [8]
e
Could it pose adverse environmental
O
O DO NOT PERFORM until a review
effects of any] type directly or
from Environmental Protection
j
indirectly? [9
is attached.
Is it involved with multiple evolutions
O
O ATTACH Marked-up P& ids, and
on the same system, an interconnected
OBTAIN approval from the Shif t
system, E will rasult in_ theJDov0 ment
Superintendent as the SR0/CFH.
'
of gases or1Tquids? [10]
~
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,
_
Is t'a complex alignment which:
O
O OBTAIN approval from the Plant
'a '
e
Is requested by another division, E /
Superintendent or designee as
.
Requires non-rautine interdivisional /
the Plant Management Staff-
/ coordination, E
/
Operations.
- Installs temporary plant equipmept'
l
that could alter the function
path of existing plan
com
nts? [11]
PREPARED / REVIEWED & APPROVED
DATE
TIME
PREPARER
[12]
MANAGER, REQUESTING ORGANIZATION
O N/A
(13]
PLANT MANAGEMENT STAFF - OPERATIONS
[13]
l
(13]
MANAGER, CPERATIONS
[14]
.
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ATTACMMENT 4
PAGE 1 0F 14
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NUCLEAR ORGANIZATION
OPERATIONS DIVISION PROCEDURE 50123-0-23
UNITS 1, 2 AND 3
REVISION 5
PAGE 57 0F 62
.
ATTACHMENT 4
KEYPOINTS FOR ABNORMAL ALIGNMENT / EVOLUTION (Continued)
9.
H this Abnormal Alignment / Evolution could pose any type of adverse
L
environmental effects, than Environmental Protection must review this
'
permit before implementation, and provide documentation to be attacned.
!
10.
If the Abnormal Alignment / Evolution is involved in multiple evolutions on
the same system, an interconnected system, E the evolution will result in
I
the movement or gases or liquids, ihan check YES, and attach marked-up
!R
,
'
P&lDs.
In November 1993, failure to properly evaluate system
i
irterconnection flowpaths resulted in HPSI Pump run-out and caused
'
extensive pump damage. Drawings are required to assist in Abnormal
l
Alignment / Evolution review and tailboard, and therefore are not recuired to
)
be attached to the ccmpleted Abnormal Alignment / Evolution.
(Ref. 2.4.7)
)
,
11.
If this a complex alignment requested by another division, E requires non-
routine interdivisional coordination, E installs temporary plant equipment
,
that could alter the function or flowpath of existing plant components,
'
including in-service or hydrostatic testing, then check YES.
12.
After preparing the document for use (including Return-to-Service
instructions) the Preparer will enter name, date, and time in the space
provided. -The individual preparing the document SHALL NOT sign any of the
Reviewed and Approved By lines.
l
13.
Approval is normally required prior to using)the document. (Refertomain
body, Steps 6.1.4 and 6.8.5.1 for exception.
H the Abnormal
Alignment / Evolution (AA) was requested by an organization other than
Operations, then the Manager of that division is required to review and
approve the activity.
If the Operations Division initiated this AA, then
ChecktheN/Aboxinthe" Manager,RequestingOrganization" space.
14.
Approval is required by the Manager, Operations prior to implementation if
the Abnormal Alignment / Evolution changes the intent of the Operating
Instruction, E constitutes an Evolution. H not, then implementation may
proceed prior to the Manager, Operations final approval, provided approval
isobtainedwithin14daysofSR0/CFHApproval.
alignment .(pecific document number that will allow closure of this
15.
Enter the s
e.g., Closure of a WAR, TFM, or NCR}. H closure is " completion
of this al'ignment", and no other documents will be involved, than state so.
E a procedure change is required, then check YES. OPG should also be
g
notified (e.g., E-Mail).
Editorial information may be included by USER (S) in the form of numbered
notes in the Comments section (e.g., add su
WAR Number to the Closure Document section)pporting information such as a
Such information does not
,
change the intent, method, or outcome of the Abnormal Alignment / Evolution.
l
16.
Insert the number and name of the associated System Operating
i
Instruction (s).
I
17.
Enter any pertinent additional references (e.g., Technical Manual, UFSAR,
l-
. Site Procedure, etc.). If none, then check NO.
f
ATTACHMENT 4
PAGE 11 0F 14
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NUCLEAR ORGANIZATION
OPERATIONS DIVISION PROCEDURE S0123-0-23
.
UNITS 1, 2 AND 3
REVISION 5
PAGE 10 0F 62
.
6.0
PROCEDURE (Continued)
,
-6.4
' Cont'rol of System Alionments Affected By System Modifications
'
6.4.1
Permanent facility modifications will be accounted for by
TCNs or revisions to the system Operating Instructions.
NOTE:
S0123-0-22 provides specific direction regarding
control of system alignrrents due to temporary
,
facility modifications,
6.4.2
Temporary modifications lasting greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
should be accounted for by TCNs or Revisions to the system
Operating Instruction (s).
.1
When the expected duration of the temporary modification
is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then Section 6.8 should be used to
document the change.
.2
Whita the. temporary modification is expected to 'last for
1
'
greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, And it is not yet covered by a
-
1
procedure TCH or revision, then Section 6.8 should be used
to document the change. This.is allowed provided the
1
Operations Procedures Group is actively preparing a TCN or
revision for' issuance.
END OF SECTION 6.4
=
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~* COMMENT * O
.
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SRO bXAMINATION QUESTION 12
- ')iR O 9)f
-S023 5-1.8 is the reference for "A" to be a correct answer. "C" is also correct based on Technical Specification 3.4.6 and 3.4.7. which
requires the RCS LOOP !b he operable. Southern California Edison believe there are two correct answers to this question.
' Accept answers A & C
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NUCLEAR ORGANIZATION
INTEGRATED OPERATING INSTRUCTION
5023-5-1.8
UNITS 2 AND 3
REVISION 9
PAGE 86 0F 91
.
ATTACHMENT 13
9.o kCP OPERATION
9.1
With at least one RCP operating, reverse flow will be present in the
idle loop.
jD
9.2
The loss of RCP heat will affect cooldown rate. Consequently, Tcold
should be maintained 2125'F to prevent entering the restrictive heatup
and cooldown limitations that apply when s120'F.
9.3
When securing RCPs, it may be necessary to reduce PZR heater output due
to.the reduction of PZR Spray Valve bypass flow.
9.4
Due to insufficient Pressurizer heater capacity, it may be necessary to
secure all RCPs and main spray prior to initiating Auxiliary Sp ay.
Otherwise, loss of MPSH for the RCPs could occur.
(Ref. 2.3.17
9.5
Pressurizer insurge may occur when securing the last RCP. This is
caused due to the lower RCS flow across the co-e. As Core Exit
Temperature rises, the RCS will swell into the Pressurizer. Adjusting
letdown flow will help minimize this insurge.
9.6
Indicated Tcold will initially rapidly lower in any loop where SDC is
injecting, if the RCP operating in that loop is stopped or when the
last RCP is stopped. This is due to cooler SDCS injection water
flowing over the loop Tcold temperature element.
6
9.7
If any RCPs are operating, then the Tcold associated with an operating
RCP should be used for RCS temperature monitoring.
9.8
WhentherearenoRCPsoperating,thenTR-0351A(T351X),SDCCombined
outlet Temperature, should be used for Tcold ternperature monitoring.
9.9
IE RCPs are running. IllEli one RCP shall remcin in service until
completing RCS boration to Mode 5, or refueling concentration and other
forced circulation dependent parameters are met (e.g., hydrogen,
peroxide, etc.).
9.10 When the last RCP is stopped, then RCS Thot indications (T351Y and
CETs) will begin to rise due to the increased time coolant is in the
j
Core region (i.e., no RCP forced circulation). Consequently, SDCS
flowrate should be adjusted to maintain RCS Tcold TR-0351A (T351X) at
the desired tentperature.
1
O
f
ATTACHMENT 13
PAGE 6 0F 11
!
10 'd
20:2T
86. Of ACN
9122-891-606:W;l
Rd TdDO 1/2 0 $9N0s
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RCS Loops--MODE 4
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3.4.6
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3.4 REACTOR COOLANT SYSTEM (RCS)-
3.4.6. RCS Loopi--MODE 4 -
Two loops or trains consisting of ray combination of RCS loops
and shutdown cooling (SDC) trains shall be OPERABLE and at least
one loop or train shall be in operation.
NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and SDC pumps may be
de-energized for s 1 bour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
a.
No operatiens are permitted that would cause
reduction of the RCS baron concentration; and
b.
Core outlet temperature is maintained at least 10*F
below saturation temperature.
2.
No RCP shall be started with any RCS cold leg
temperature s 256*F unless:
.
a.
Pres:urizer water volume is < 900 ft', or
b.
Secondary side water temperature in each steam
generator (SG) is < 100*F above each of the RCS cold
..
'
leg temperatures.
-
____________________________________________________________
APPLICABILITY:
MODE 4.
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[
SANONOFRE-bHIT2
3.4-18
Amendment No. 127
e
_ _ _ _ _ _ _ _ _ - _ - - _ - - _ - - - - _ _ - - - - - _ - - _ - - - - - - - _ - - - - - - - - - -
,
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RCS Loops--MODE 4
-
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3.4.6
'
.
M
ACTIONS
,
?
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.
One required RCS loop
A.1
Initiate action to
Immediately
restore a second loop
or train to OPERABLE
AND
status.
Two SDC trains
.
B.
One required SDC train
B.I
Be in MODE 5.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
AND
Two required RCS loops
C.
Required RCS loop (s)
C.1
Suspend all
Immediately
-
_i
or SDC train (s)
operations involving
reduction of RCS
boron concentration.
EE
AND
train in operation.
C.2
Initiate action to
Immediately
re. store one loop or
train to OPERABLE
status and operation.
.
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~
SAN ON0FRE--UNIT 2
3.4-19
Amendment No. 127
_ - _ _ _
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RCS Loops-MODE 4
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3.4.6
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A
SURVEILLANCE REQUIREMENTS.
1-
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SURVEILLANCE
FREQUENCY
Verify at least' one RCS loop or SDC train
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
is in operation.
'5R 3.4.6.2
Verify secondary side water level in
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
required SG(s) is 2 50% (wide range).
.
Verify the second required RCS Loop or SDC
7 days
train is OPERABLE.
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SAN ONOFRE--UNIT 2
3.4-20
Amendment No. 127
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RCS Loops-MODE 5, Loops Filled
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3.4.7
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3.4 REACTOR COOLANT SYSTEM (RCS)
~
3.4.7 RCS Loops-MODE 5, Loops Filled
At least one of the.following loop (s)/ trains listed below
shall be OPERABLE and in operation:
,
a.
Reactor Coolant Loop 1 and its associated steam
generator and at least one associated Reactor Coolant
Pump;
b.
Reactor Coolant Loop 2 and its associated steam
generator and at least one associated Reactor Coolant
i
Pump;
-
l
c.
Shutdown Cooling Train A; or
d.
Shutdown Cooling Train B
One additional Reactor Coolant Loop / shutdown cooling train
!
' hall be OPERABLE, or
s
The secondary side water level of each steam generator shall
be greater than 50% (wide range).
l
y
,
)
NOTES---------------------------
o
1.
All reactor coolant pumps (RCPs) and pumps providing
-
shutdown cooling may be de-energized for i I hour per
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
a.
No operations are permitted that would cause
reduction of the RCS boron concentration; and
b.
Core outlet temperature is maintained ai least 10'F
below saturation temperature.
2.
One required SDC train may be inoperable for up to
i
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other
'
SDC train or RCS loop is OPERABLE and in operation.
3.
One required RCS loop may be inoperable for up to 2
hours for surveillance testing provided that the other
RCS loop or SDC train is OPERABLE and in operation.
(continued)
L
)
!
SAN ON0FRE--UNIT 2
3.4-21
Amendment No. 127
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RCS Loops--MODE 5, Loops Filled
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3.4.7
,
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NOTES
(continusd)---------------------
4.
No reactor' coolant pump (RCP) shall be started with one
or more of the RCS cold leg temperatures s 256*F unless:
,
l
a. -The pressurizer water volume is < 900 ft3 or
b.
The secondary side water temperature in each steam
generator (SG) is < 100*F above each of the RCS cold
leg temperatures.
_
5.
A containment spray pump may be used in placs of.a low
!
pressure safety injection pump in either or both
shutdown cooling trains to provide shutdown cooling flow
provided the reactor has been suberitical for a period
> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the RCS is fully depressurized and vented
in accordance with LCO 3.4.12.1.
,
'
6.
All SDC trains may be removed from operation during
planned heatup to MODE 4 when at least one RCS loop is
,
in operation.
____________________________________________________________
,
APPLICABILITY:
MODE 5 with RCS loops filled.
, . ,
_'..l
ACTIONS
~
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.
Less than the required
A.1
Initiate action to
Immediately
SDC trains /RCS loops
restore the required
SDC trains /RCS loops
,
to OPERABLE status.
AND
Any SG with secondary
side water level not
A.2
Initiate action to
Immedi ately
within limit.
restore SG secondary
side water levels to
within limits.
(continued)
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SAN'ON0FRE--UNIT 2
3.4-22
Amendment No. 127
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RCS Loops-MODE 5, Loops Filled
.
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3.4.7
"
.
'}
ACTIONS- (continued)
. - .
l
CONDITIO'N
REQUIRED ACTION
COMPLETION TIME
~B.
No SDC train /RCS loop _
B.1
Suspend all
Immediately
in operation.
operations involving
- -
!
reduction-in RCS
i
baron concentration.
.
.E.N.Q .
B.2
Initiate action to
Immediately
i
restore required SDC
L
train /RCS loop to
-
l
operation.
' SURVEILLANCE REQUIREMENTS
!
' SURVEILLANCE
FREQUENCY
.. . .)
SR 3.4./.1
Verify at least one RCS loop or SDC train
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
,
~~,
is in operation.
L
,
'
Verify required SG secondary side water'
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
level is 2 50Y. (wide range).
'
L
Verify the second required RCS icop, SDC
7 days
tFain or steam generator secondary is
t.
!
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SAN ONOCRE--UNIT 2
3.4-23
Amendment No. 127
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COMMENT H
SRO Examination Question 18
(RO14)
The generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. The Reed
Switch Position Transmitters, RSM's, actually sense the CEA's position. The Control Element Assembly
Calculator, CEAC, uses the input from the RSPT and sends a signal to the alarm. Both components are needed to
,
generate a deviation alarm. Southern California Edison believes there are two correct answers to this question.
- Accept answers B & C
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. NUCLEAR ORGANIZATION
ALARM RESPO!ISE. INSTRUCTION 5023-15-50.Al
. UNITS 2-AND 3
REVISION 2
PAGE 71 0F 76
ATTACHMENT 2
..
50A28
CEA DEyIATION
APPLICABILITY
PRIORITY.
REFLASH
ASSOCIATED WINDOWS
i
-
Modes 1-3
AMBER
NO
NONE
.
,
INITIATING
NOUN NAME
SETPOINT
VALIDATION
PMS 10
LINK t
DEVICE
INSTRUMENT
U2/U3
,
i
l
2(3)LO91, CEAC 1
Control Element
5 Inches
NONE-
DEV1AR56
641/663
lR
or CEAC 2
Assembly Deviation
1.0
RE0VIRED ACTIONS:
1.1 Position the CEDMCS Mode Selector Switch on 2(3)CR50 to 0FF.
1.2 Verify which'CEA is misaligned and the amount of misalignment, by
observation.of the following:
CEAC display CRT
CEAC remote operators modules
- PMS alarms
- PMS readout
'
l.
2.0
CORRECTIVE ACTIONS:
SPECIFIC.CAUSES
SPECIFIC CORRECTIVE ACTIONS
'
.
2.1' Misaligned CEA
2.1 After the misaligned CEA has been
determined, _thful:
,
,
2.1.1
Notify the SR0 Ops. Supv.
2.1.2
Realign.the CEA per 5023-3-2.19,
Section for Manual Individual
(
l
Operation.
2.2 Slipped or Dropped' CEA
2.2 GO TO S023-13-13, Misaligned Control Element
Assembly.
l-
3.0-
ASSOCIATED RESPONSES:
,
3.1 Notify the CRS/SS and the STA to review lech. Specs. LC0 3.1.5 and
LCS3.1.105,andinitiateanEDMR/LC0AR,asrequired.
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NUCLEAR ORGANIZAfl04
SYSTEM DESCRIPTION SD.5023-710
.
UNITS 2 AND 3
ls
REVISION 3
PAGE 72 0F 76
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FIGURE 15? CONTROL ELEMENT ASSEMBLY SUBGROUP WEED SWITCH
POSITION TRANSMITTER SIGNAL AS$1GNMENTS
l
. h EX-CORE CHANNEL
RSPTA RSPT\\
'
_2
23 CEAS
l
,,
/
h
RSPT
t
v
2
'
i
i
22 CEAS
Chs
i
1
23 CEAS
22 CgAs
RSPT
i
U
2
MQU\\
h
22 CE
N
1
g
2
l
l
23 CEAS
23 CEA3
h
C
-
-
CALCULATOR
CALCULATOR
-
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CEA POSITCN
NO.1
NO. 2
CEA POSITION
l
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3 ?:1ra
Eni;?
CA? A tmG
CAT A UMci
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4
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A CORE
8 CORE
C CORE
DCORE
' ,*
PROTECTON
PROTECTION
PROTECTION
PROTECTON
CALCULATOR
CALCULATOR
CALCULATOR
CALCULATOR
.
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OPERATOR'S
OPERATOR'S
OPERATORM
OPERATOR'S
MODULE
MCOULE
MODULE i
MODULE
CRT OtSPLAY
I
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NOTES.1. SIGNAL FROM CEA 2 IS CCNNECTED TO CPC's A AND C, BUT IT IS NOT USED AS A TARGET CEA.
2. SIGNAL FROM CEA 3 CS CONNECTED TO CPC's 5 AND D. BUT IT is NOT USEO AS A TARGET CEA.
l
3. SIGNALS FRCN 23 CEA's ARE CONNECTED TO EACH CPC.
CNLY 22 CF THE 23 SIGNALS ARE USED AS TARGET CEAs.
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,
COMMENT # 3
.
SRO EXAMINATION QUESTION 22
. .
There is no correct answer. Actual allowable maximum level is $7% per SO23 3 1,7. L&S 2.2. Southern California Edison believes
there is no correct answer to this question.
' Delete Question
s
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NUCLEAR ORGANIZATION
OPERATING INSTRUCTION
S023-3-1.7
UNITS 2 AND 3
REVISION 20 TCN
20-2
PAGE 56 0F 56
ATTACHMENT 16
1.0
REACTOR COOLANT PUMPS (Continued)
1.15 2 (3)'P-002 : For the ABB RCP Motors, the Lif t Oil Pumps normal
discharge pressure is 1400 psig (allowable range: 1377 to 1450 psig),
1.16 Bleed-off flow normally is proportional to RCS pressure.
At 2250 psia, bleed-off flow should be between 1.25 gpm and 1.75 gpm.
If at a low pressure, and CB0 flow is < 0.25 gpm, .th23 one of the
following is required:
CB0 line temperature (at the local flow indicator) is warm
(i.e., cold line indicates no flow)
93
RCP Seal Cavity pressures are properly staged.
2.0
2.1
If Boron concentration in an idle loop is suspected of being lower
than Reactor Core boron concentration, IHf3 DO NOT attempt to Start a
RCP in that idle loop. This will prevent a possible reactivity
transient upon restart of forced flow.
(Ref. 2.1.5)
With'any RCS cold leg temperature s 260'F, DO NOT Start a RCP unless
2.2
the following conditions for PZR level and RCS temperature are met.
Use the most conservative values available in order to maximize
delta T (Tsat-Tc).
(Ref. 2.3.1 and Tech Spec. LCO 3.4.6, LCO 3.4.7)
PZR LEVEL
.
RCS TEMP (ADMIN LIMIT)
30%"
T,,, (S/G) <T, + 20
~)
57% (<900 f t')
T,,, (S/G) <T, + 10
T,,, (S/G) <T, + 100
2.3
If the RCS has just been initially filled (air trapped in S/G 'U'
tubes), then RCS pressure may drop rapidly below the minimum pressure
for RCP operation.
2.4
If the RCS is solid, then RCS pressure may rapidly rise above the
maximum pressure for SDC loop operation due to heat transfer from the
S/GtotheRCS.
2.5
If RCS pressure is being maintained by the Letdown Backpressure
controller, then automatic operation may tend to raise RCS pressure
by the amount of RCP differential pressure since letdown comes from
the pump suction cold leg.
2.6
Failure of the seals to stage on an operating RCP with RCS pressure
greater than 700 psia is an indication of a failed seal (s).
3.0
STOPPING AN RCP
3.1
When in Mode 3, then failure to bypass the associated SG Low Flow
Trip before stopping an RCP will rr tult in a Reactor trip signal.
52
ATTACHMENT 16
PAGE 4 0F 4
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COMMENT #6
SRO Examination Question 28
(RO27)
Answer "B" is correct because 656 deg F corresponds to 2300 psia. 'Ibe pressurizer spray valves open at 2275 psia.
Answer "D" is also correct because at 2225 psia and a backup signal at 2275 psia, the heaters get a signal to turn
off. Southern California Edison believes that there are two correct answers to this question.
'
Accept answers B & D -
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$UCLEAR ORGANIZATION
~ ~
SYSTEM DESCRIPTION SD-SO23-360.
-
UNITS 2 AND 3
.. -
REVISION 5
Page 168 of 205
^--
FIGURE 111-5 ' PRESSURIZER PRESSURE CONTROL SYSTEM B' LOCK DIAGRAM
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COMMENT #7
-
SRO Examination Question 30
The design basis for adding the steam generator delta p trip was based on steam line or feed line break (harsh
environment) inside containment accompanied by a loss of offsite power The steam generator delta p signal used
for a reactor trip due to a sheared RCP shaft is not related to the loss of offsite power. This makes the answer A
incorrect.
Southern California Edison believes there are no correct answers to the question.
Delete this question.-
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SOUTHERN CALIF 0RNIA EDISON
NUCLEAR ENGINEERING, SAFETY AND LICENSING
PLANT PROTECTION SYSTEM
'
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- DESIGN BASES DOCUMENT.
0B0-5023-710. REV. 4
PAGE 92 0F 569
ADDRESSABLE CONSTANTS
.
,
Symbol
Definition
Range
' BUFTRP -
Snapshot Buffer Control Flag
0 or 1
TCBP
Maximum time tha' the RPC Flag can remain set
0 to 40.0
(seconds)
TCOUNT-
CRT Display Rewrite Control Flag
0 or 1
2.1.1.15
LowReactorCoolantFlow(LRCF)
Eachsteamgenerator'2(3)E088and2
measurement of differential pressure (3)E089 has an RCS four channe
measured across the primary
- side, which is indicative of RCS Flow. This ' function was originally
-added to the RPS to provide a qualified means of tripping the reactor
for a SLB or FWLB inside containment accompanied by a loss of offsite
power, since not all of the LDNBR signal inputs (i.e., RCPSSSS) were
qualified to function in the harsh environment created by those
accidents."**""'"""
Another substantive reason for adding this
function was to provide batter protection against the sheared shaft
event, whose impor.tance in the safety analysis of design basis events
had elevated since the original plant analyses was performed.
The
shearing of a RCP shaft was not considered a design basis event in
the initial 3410 MWt reactor design, since it was not required by
Revision 1 of R.G.1.70.
However, Revision 2 of R.G.1.70 requires
that this event be considered in the preparation of the FSAR.
Additionally, analyses performed demonstrated that acceptable
'
consequences cannot be demonstrated without pn. iding some sort of
protective action, and that this event has about the same probability
of occurrence as the seized shaft event. The protection offered by
the CPCs for this event could be compromised if the RCP shaft were to
shear above the RCPSS sensors.- The low flow trip function utilizing
a variable setpoint based on steam generator primary differential
-pressure was selected as the optimum design to mitigate this event,
since it does not depend sn the CPCs, and could be developed,
installed, and meet licensing schedules.""
Th'e PPS provides a channel trip when the ACS flow-produced
differential pressure falls below the setpoirit. . A reactor trip then
follows on a 2-out-of-4 basis. This trip function is presently
credited to help mitigate the consequences of a sheared RCP shaft
accident, or a two-pump or four-pump coast down event, and is
therefore classified as a safety function per 2.1 (iii). See Table-
B-12. Applicable Modes are 1 and 2.
This trip function has a variable setpoint feature that causes the
differential pressure setpoint to track below the measured
differential pressure by a pre-determined increment. The tracking
rate of the setpoint is rate-limited, in that it can decrease only at
a pre-selected maximum rate, and only to a pre-selected minimum value
(" floor"). Should the signal level fall below the setpoint level
m
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. COMMENT #8
SRO Examination Question 37
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(RO37)
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Procedure, SO23 12-7. Ims of Offsite Power / Loss of Forced Circulation, floating step 2 states that Thot and
CET's are compared as are Thot and Tc are not rising to verify natural circulation is occurring. The S/G pressure
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can be used to correlate to Tc there fore ans'ver B is also correct. Southern California Edison believes there are two
correct answers to this question.
Accept answers A & B
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NUCLEAR ORGAN!2ATICN
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UNITS 2 AND 3
EMERGENCY CPERATING INSTRUCTICN
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REVISION 15
5023-12-7
[
ATTACHMENT 2
PAGE 30 0F 122
LOSS OF FORCED CIRCULATION / LOSS OF OFFSIT
-
FLOATING STEPS
ACTION / EXPECTED RES_PONSE
RESPONSE NOT OBTAINED
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NOTE:
Low flow during Natural Circulation slows RCS
response to temperature changes.
L0cc transit
time rises to between 5 minutes and 10 minutes.
FS-2
MONITOR Natural Circulation
Established:
CHECK all RCPs - stopped.
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a.
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GO TO FS-4, MONITOR RCP Operating
a.
Limits.
b.
CHECKatleastoneS/G
b.
GO TO S023-12-9, FUNCTIONAL RECOVERY
operating:
AND
1) SBCS - operating
,
INITIATE S023-12-9, Attachment 8
-
REC 0VERY - HEAT REMOVAL.
ADV - operating.
\\
AND
2)
Feedwater - available.
CHECK operating loop AT - less
c.
than 58'F.
IF any criteria c through g NOT
o
satisfied,
,
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CHECK Tc and Ts - NOT rising.
THEN
CHECK Reactor Vessel level
e.
(Plenum) - greater than or
MAXIMIZE S/G level - less than
equal to 100M
80% NR.
QSPOS page 622
RAISEavailableS/Gsteaming
.
i,
rate.
ii
CFMS page 312
Attachment 4.
RAISE Core Exit Saturation Margin
f.
CHECK operating loop Ta and REP
- greater than 20'F:
'
CET - within 16*F:
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QSPDS page 611
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QSPDS page 611
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CFMS page 311.
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ATTACHMENT 2
PAGE 4 0F 29
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- NUCLEAR CRGANIZATICN
1
UNITS 2 AND 3
EMERGENCY OPERATING INSTRUCTICN 5023-12-
-' *
REVISION 15
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ATTACHMENT 2
PAGE 31 0F 122
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LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWE
FLOATING STEPS
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ACTION / EXPECTED RESPONSE
RESPONSE NOT O8TAINED
FS-2
MONITOR: Natural Circulation
Established:
(Continued)
9
CHECK Core Exit Saturation-
Margin - greater than 20*F:
IF any criteria c through g NOT
o
'
satisfied;
J
QSPOS page 611
THEN
CFMS page 311.
MAXIMIZE S/G level - less'than
.
80% NR.
,
RAISE available S/G steaming-
rate.
RAISE Core' Exit Saturation Margin
.
- greater than 20*F:
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,QSPOS page 611
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CFMS page 311.
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ATTACHMENT 2
PAGE 5 0F 29
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COMMENT #9
SRO Examination Question 43
(RO46)
Answer B is correct based on the strictest interpretation ofimmediately before and afict a trip. Immediately before
the trip, S/G level exceeded 85% causing a High Level Override, HLO, signal to be applied to the main and bypass
food regulating valves causing both to close. The valves both st9y closed until level decreases below 85% at which
time the HLO signal clears and the valves go to either the Reactor Tripped override, RTO, position or to the
position set by the demand from the feed water regulating control system. With a reactor trip, an RTO signal is
sent which as soon as the HLO condition clean seconds after the trip due to normal shrink of water 1svels, the
RTO signal is applied and the bypass valve opens to 50%, answer C. Southern California Edison believes there are
two correct answers to this question.
Accept answers B & C
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- SRO EXAMINATION QUESTION 51
c
'All four answers contain the statement "over current reset".. There is no over current relay in the circuitry for the 50.54X cross-tie for
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the EDGs.1 Southern California Edison believes there is no correct answer for this question and the question should be deleted from
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the examination.
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COMMENT #11
SRO Examination Question 60
(RO7)
Procedure SO23 13-13, Misaligned CEA, has a note after step 1 stating " Initial and stabilized reactor power levels
are required for the subsequent shutdown margin calculation." This is the basis for answer A being correct.
In attachment 3 of the same procedure, there is anotl'er caution that states: "Within 15 minutes of misalignment
'
discovery, a power reduction may be required.. " laitial and final stabilized power levels are used to determine the
further power reduction requirements within the first hour to actintain compliance with the acceptable operating
region in technical specification LCO 5.1.5 and LCS 3.1.105. This is why answer C is correct.
Southern California Edison beheves the there are two correct answers to this question.
Accept A & C
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NUCLEAR ORGANIZATION
ABNORMAL OPERATING INSTRUCTION
S023-13-13
UNITS 2 AND 3
REVISION 4 TCN 4-1
PAGE 5 0F 24
.
MISALIGNED CONTROL ELEMENT ASSEMBLY
<
OPERATOR ACTIONS
ACTION / EXPECTED RESPONSE
RESPONSE NOT OBTAINED
3 COMMENCE plant load reduction:
a,
'If Reactor power is < 50%,
THEN GO TO Step 3c:
-
fddl.UEff
Within fifteen minutes of misalignment discovery a power reduction
may be required. The negative reactivity of the misaligned CEA is
censidered part of the required power reduction.
Failure to
maintain Reactor Power in the Region of Acceptable Operation is a
violation of Tech. Spec. LCO 3.1.5~and LCS 3.1.105.
NOTE: The power reduction shall be in accordance with
the applicable LCS 3.1.105 Figure.
The boration flowrate
shall be sufficient to achieve the target power level
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 45 minutes of the rod drop time.
b.
INITIATE required RX power
reduction to maintain RX
power in the Region of
Acceptable Operation per the
applicable LCS 3.1.105
figure.
1) LOWER Turbine Generator
load using CVOL wlfle
maintaining Tcold within
the Operating Band per
Attachment 1.
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COMMENT # 12
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, .SRO EXAMIN tT'O , G " *1 ""> "
Both answcrs "A" & "B" will increau i ' ine.ts for operation of the reactor coolant pumps. Southern California Edison believes there
are two correct answers to this question.
Accept answers A & B
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h0 CLEAR OAGAi!IAT10N
EMERGENCY OPERATING lh5tRUCTION 5023-12-3
LN(T5 2 AND 3
21:15101 15
PAGE 141 0F 163
ATTACHMENT 14
LCSS Or COOLANT ACCIDEMI
PO$T ACCIDENT PRE 55URI/TD4PERATORE LIMIT 5
2500
(2380 PSIA) MA UMUM OPERATI NAL PRESSUR
,
100*F/1;R
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200*F
, TEMPERATURE
SATURATION
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MARGIN TC
2000
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SDC ENTRY CONblTIONS
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0
130
200
300
400
500
600
700
800
RCS TEMPERATURE (*F)
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NOTE 1
THis CLRVE 15 IM EFFECT ANY TIME AN OMCONTROLLED COOLDOWN TO RC5 Tc LESS T4AN 500'F
HAS OCCURRED,
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ATTACHMENT 14
PAGE 1 CF 1
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COMMENT #13
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SRO Examination Question 69
(RO74)
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The question setup has the VCT pressure at 40 psig. The head of the Refueling Water Storage tank, RWST, is 70'
of H2O or 30.3 psig(70' x 0.433 psi /ft) In the scenario provided the head of the VCT with the over pressure, will
keep the check valves from the RWST closed. Southern California Edison believes there are no correct answers to
this question and it should be deleted.
Delete this question.
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COMMENT #14
SRO Examination Question 14
(RO78)
All pressurizer heaters receive a backup signal to turn off at a pressure of 2275 psia. This is true for the heaters
that are in auto. The stem states that the heaters are in auto. With PT0100X failing high, the signal to the heaters
j
will exceed the 2275 psia shutoff set point for the heaters in auto. The correct answer should be D.
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Change correct answer to D.
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NUCLEAR ORGANIZATION
SYSTEM DESCRIPTION SD-SO23-360
UNITS 2 AND 3
REVISION 5
Page 168 of 205
FIGURE lil-5 PRESSURIZER PRESSURE CONTROL SYSTEM BLdCK DIAGRAM
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PT
100X
100Y
1E HEATER r-- PMS/CFMS
PMS/CFMS -,1E HEATER
<
ETEAM BYPASS SYSTEM
STEAM BYPASS SYSTEM -
4-
i
B/S
'
B/S
RED
2500 -
GREEN
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HS-100A
1 Oh-
2200G225
0
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10
Hl/LO
B/S
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50Y14M
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PV 1008
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CUMMENT# 15
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SR0 EXAMINATION QUESTION #75
(R0 79)
Question was based on old Tavg program. New program has normal pressurizer level cf 48% at 100% power. This is based on the
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reduced Tc program of 548 deg F @ 100% power. The lower Tc at full power equates to a Tave of 574 deg F. Per the attached
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reference, the expected level would be 48% and no additional charging pumps would be operating. Southern California Edison
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believes there are no correct answers for this question and the question should be deleted from the examination.
,
Delete Question.
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OPERAT!NG 1NSTRUCTION
5023-3-1.10
UNITS 2 AND 3
REVISION 8
PAGE 30 0F 34
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ATTACHMENT 5
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ATTACHMENT 5
PAGE 1 0F 1
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COMMENT # 16
SRO Examination Question 77
DSS is not covered by Technical Specifications. There is no correct answer to the question as stated.
Delete the question.
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COMMENT #17
SRO EXAMINATION QUESTION 84
(RO 85)
The trends given are inconclusive as to whether vacuum has stabilized at 3.3", if assumed vacuum is stable at 3.3 "
no further action will be required and answer B would be correct. If it is assumed vacuum will continue the current
trend, the listed action of"D" could be taken to return the plant to a more stab!c condition. Southern California
Edison believes "B" & "D" are correct answers.
Accept B & D.
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NUCLEAR ORGANIZATION
ABNORMAL OPERATING INSTRUCTION 5023-13-10
UNITS 2 AND 3
REVISION 2
PAGE 7 0F 13
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ACTION / EXPECTED RESPONSE
RESPONSE NOT OB(AINED
4 Actions for loss of vacuum due to
Condenser fouling:
. CAUTION
During periods of heavy influx, rapid and aggressive action may be required in
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, order to avoid a Unit trip. Power may need to be reduced in order to:
Bump and/or Stop Circulating Water Pumps on the Condenser quadrants with
e
the highest differential pressures
Maintain Condenser backpressure < 3.5" Hg
a. REDUCE Reactor power to 75% TO
85%.
b. BUMP Circulating Water Pump (s)
per direction of Shift
Superintendent.
c. VERIFY backpressure < 3.5" Hg
c.
1) REDUCE Reactor power
and stable.
to s 65%.
2) STOP two Circulating Water
Pumps on opposite ends of
the Condenser.
3) INITIATE isolating stopped
pumps per 5023-2-5,
Attachment for Stopping a
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Circulating Nater Pump Due
to Fouled Condenser
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Tubesheet/High 6P/ Debris
K.
Removal.
4) IF not < 3.5" Hg and stable,
THEN REDUCE Reactor Power
as necessary to establish
backpressure < 3.5" Hg and
stable.
5) GO TO Step 5.
d.
EVALUATE stopping pumps based on
Waterbox differential pressure
and pump vibration.
e. GO TO Step 6.
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C0tWENT #19
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SR0 EXAMINATION QUESTION #76~
(R0 95)
Both answerss A & B will cause a PTS event to occur if the operator
fails to initiate release of steam from E088 S/G.
"A" is correct
based on failing to steam the good S/G to establish a heat sink.
"B" is also correct in that HPSI throttle /stop criteria will NOT
be met because of the unavailability of the S/G as stated in the
stem (step FS-6 a.1 requires operating S/G with an ADV operating),
and continuing to inject water into the RCS will increase
pressure also leading to PTS event.
56uthern California Edison
believes there are two correct answers to this question.
Accept answers A & B
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NUCLEAR ORGANIZAT10N
EMERGENCY OPERATING INSTRUCTICN S023-12-5
.
UNfTS 2 AND 3
REVIS10N 15
PAGE 45 0F 143
ATTACHMENT 2
EXCESS STEAM DEMAND EVENT
.
FLOATING STEPS
ACTIONS / EXPECTED RESPONSE
REgp0NSE NOT 08TAINE0
FS-6
CHECK HPSI Throttle /Stop
,
Criteria:
(Continued)
h.
MAINTAIN Criteria of steps a
through e - satisfied.
i.
CHECK Containment pressure
i. 1)
ENSURE SIAS - actuated.
- less than 3.4 PSIG.
2)
GO TO FS-7, CHECK LPSI
6
Terniination Criteria.
J.
CHECK PZR Level
J. INITIATE FS-22 ESTABLISH CVCS
- less than 80%.
Letdown Flow.
k.
RESET SIAS per 5023-3-2.22,
ESFAS OPERATION.
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ATTACHMENT 2
PAGE 15 0F 43
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NUCLEAR ORGANIZATICN
EMERGENCY CPERATING INSTRUCTION S023-12-5
UtilTS 2 AND 3
REVIS10N 15
PAGE la CF 113
ATTACHMENT 2
EXCESS STEAM DEMAND EVENT
.
FLOATING STEPS
ACTIONS / EXPECTED RESPONSE
RESPONSE NOT CBTAINED
FS-6
CHECK HPSI Throttle /Stop
Criteria:
(Continued)
d.
CHECK Reactor Vessel Level
c
IF any criteria of ste:s b tnrcugh c
(Plenum) - greater than or
NOT met,
equal to 100%:
THEN
QSPDS page 622
CFMS page 312
OPERATE Charging and HPSI systems
.
Attachment S.
as necessary to maintain
{
Throttle / Step criteria
- satisfied.
THROTTLE Loop Injection Valves.
.
2
ENSURE auxiliaries to SI pumps.:
a)
Electrical power to pumps and
valves.
b)
Proper system alignment.
c)
CCW flew.
d)
HVAC.
e.
VERIFY RCS borated - greater
e. MAINTAIN Emergency Boration
than Technical Specification
- at least 40 GPM.
Shutdown Margin for T4E > 200"F
per Operations Physics Summary
Figure 2.3-1,
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RCS Cooldown - NOT in progress.
f.
required one train at a time.
g.
STOP charging pumps as required
one at a time.
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ATTACHMENT 2
PAGE 14 0F 43
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NUCLEAR ORGAN!ZATION
EMERGENCY OPERATING INSTRUCTION S023-12-5
UNITS 2 AND 3
REVIS10N 15
PAGE 43 0F 143
ATTACHMENT 2
EXCESS STEAM DEMAND EVENT
FLOATING STEPS
ACTIONS / EXPECTED RESPONSE
RESPONSE NOT OBTAINED
FS-6
CHECK HPSI Throttle /Stop
. Criteria:
a.
CHECKatleastoneS/G
a.
GO TO S023-12-9, FUNCTIONAL RECOVERY
operating:
AND
1)
SBCS - operating
INITIATE S023-12-9, Attachment 8,
RECOVERY - HEAT REMOVAL.
-ADV - operating.
.
AND
2)
Feedwater - available.
-
b.
CHECK PZR level
o
IF any criteria of steps b through d
NOT met.
- greater than 30%
THEN
AND
.
OPERATE Charging and HPSI systems
.
- NOT lowering.
as necessary to maintain
g
Throttle /Stop criteria
c.
CHECK Core Exit Saturation
- satisfied.
Margin - greater than 20*F:
THROTTLE Loop Injection Valves.
.
CFMS page 311.
ENSURE auxiliaries to SI pumps:
.
a)
Electrical power to pumps and
valves,
b)
Proper system alignment.
c)
CCW flow.
d)
HVAC.
Q.
ATTACHMENT 2
PAGE 13 0F 43
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COMMLYT # 20
SRO EXAMINATION QUEST 10N 97
(RO 96)
Answer "B" is correct based on the information given. Howner answer "A" is also correct based on procedure
5023-12-7, Safety Function Status Checks, which requires rabcooling > 20F or you are directed to the Functional
Recovery for not meeting natural circulation. Southern California Edison believes there are two correct answers to
this question.
Accept answers A & B.
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NUCLEAR ORGANIZATICN
EMERGENCY OPERAT8NG INSTRUCT!CN S023-12-7
,
UNITS 2 AND 3
REVISION 15
PAGE 30 CF 122
ATTACHMENT 2
LOSS. OF FORCED CIRCULATION / LOSS OF 0FFSITE PCWER
,
,
FLOATING STEPS
ACTION /EXPCCTED RESPONSE
RESPONSE NOT OBTAINED
NOTE:
Low flow during Natural Circulation slows RCS
response to temperature changes.
Loop transit
time rises to between 5 minutes and 10 minutes.
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FS-2
MONITOR Natural Circulation
Established:
a.
CHECK all RCPs - stopped.
a.
GO TO FS-4, MONITOR RCP Operating
Limits.
,
b.
CHECKatleastoneS/G
b.
GO TO S023-12-9, FUNCTIONAL RECOVERY
operating:
.
AND
1)
SBCS - operating
INITIATE 5023-12-9, Attachment 8,
i
RECOVERY - HEAT REMOVAL.
ADV - operating.
AND
,
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2)
Feedwater - available.
t
c.
CHECK operating loop AT
.less
o
IF any criteria c through g NOT
{
than 58'F.
satisfied,
d.
CHECK Tc and Tu - NOT rising.
THEN
e.
CHECK Reactor Vessel Level
MAXIMIZES /Glevel-lessthan
.
.
(Plenum) - greater than or
80% NR.
equal to 1004:
1
RAISEavailableS/Gsteaming
.
4
QSPOS page 622
rate.
CFMS page 312
i
Attachment 4.
RAISE Core Exit Saturation Margin
.
f.
CHECK operating loop Ts and REP.
CET - within 16'F:
QSPOS page 611
CFMS page 311.
QSPOS page 611
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CFMS page 311.
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ATTACHMENT 2
PAGE 4 0F 29
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NUCLEAR ORGANIZATION
EMERGENCY CPERAT!NG 1NSTRUCT10N S023-12-7
l*
UN!TS 2'AND 3
REVISION 15
ATTACHMENT 2
PAGE 31 0F 122
LO 5 0F FORCED CIRCULATION / LOSS OF 0FFSITE POWER
FLOATING STEPS
ACTION / EXPECTED RESPONSE
. RESPONSE NOT OBTAINED
FS-2
MONITOR Natural Circulation
Established:
(Continued)
g.
CHECK Core' Exit Saturation
o
IF any ' criteria c through g NOT-
Margin - greater than 20'F:
satisfied,
OSPOS page 611-
THEN
CFMS page 311.
MAXIM 1ZE S/G level - less than
.
804 NR.
i
RAISE available S/G steaming
a
rate.
RAISE Core Exit Saturation Margin
- greater than 20*F:
QSPDS page 611
CFMS page'311.
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ATTACHMENT 2
PAGE 5 0F 29
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