ML20202C393
| ML20202C393 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/22/1999 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9902010009 | |
| Download: ML20202C393 (1) | |
Text
.._. _
j#"'%,k UNITED STATES 3
. NUCLEAR RESULATORY COMMISSION B
a j.
REGION IV
[
611 RYAN PLAZA DRIVE, SUITE 400 g,,
ARLINGTON, TEXAS 76011-4064 January 22,1999 NOTE TO:
NRC Document Control Desk '
Mail Stop O-6-D-24. ~ '
FROM:
Laura Hurley, Licensing Assistant Operations Brancii, Region IV
SUBJECT:
OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON NOVEMBER 15-20,1998, AT SAN ONOFRE NUCLEAR GENERATING STATION.
DOCKETS #50-361/362 On November 15-20,1998, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and the initial exam submittal for distribution under RIDS Code A070.
b) As given operating examination, designated for distribution under RIDS Code A070.
Item #2 -
Examination Report with the as given written examination attached, designated for distributt on under RIDS Code IE42.
If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860-8253.
2C0117 9902010009 990122 PDR ADOCK 05000361 V
._m._.
___ ~
k e u...
8 UNITED STATES
~
NUCLEAR REG Y COMMISSf0N s
. S.fhb
- -, m
- v... e w, t 4 j
. s,, m t, n. w -w 1
g JAN 1999 i
1 i
Harold B. Ray, Executive Vice President Southern California Edison Co.
-San Onofre Nuclear Ger.erating Station P.O. Box 128 San Clemente, California 92674-0128
SUBJECT:
NRC INSPECTION REPORT 50-361/98-302; 50-362/98-302
Dear Mr. Ray:
From November 15-20,1998, an operator licensing certification inspection was conducted at your San Onofre Nuclear Generating Station, Units 2 and 3, reactor facilities. The enclosed report presents the scope and results of that inspection.
The inspection included an evaluation of five applicants for reactor operator licenses and nine applicants for senior operator licenses. We determined that ten of the applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
Your facility training and operations staff reviewed the written examinations after administration and recommended changes to 22 of the questions submitted for the reactor operator and.
senior operator examinations combined. Nearly all of the comments cited technical inaccuracies in the questions that made more than one choice correct or resulted in no correct choices. This represents a significant weakness in the technical review conducted prior to examination submittal. Additionally, in several cases, applicants responded with the same wrong answer to a number of questions. We concluded that this represented significant training deficiencies. This conclusion was further supported by some of the operational errors made by applicants.
Based on these technical review and training deficiencies, we request that you respond in writing to this report within 90 days of receipt of this letter. Your response should address the root causes for the lack of adequate technical review and the training deficiencies represented by the applicant performance and your proposed corrective actions.
. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its
/
, erdlosure will be placed in the NRC Public Document Room (PDR).
/
f ryk
//
^
17puw W p,'
Southern California Edison Co. Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, John L. Pellet, Chief Operations Branch Division of Reactor Safety Docket Nos.: 50-361, 50-362 License Nos.: NPF-10, NPF-15
Enclosure:
NRC Inspection Report 50-361/98-302; 50-362/98-302 cc w/ Enclosure and Attachments 1-3:
Rob Sandstrom, Training Manager Southern California Edison Co.
San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 cc w/ Enclosure and Attachments 1 and 2:
Chairman, Board of Supervisors County of San Diego 1600 Pacific Highway, Room 335 San Diego, California 92101 Alan R. Watts, Esq.
Woodruff, Spradlin & Smart 701 S. Parker St. Suite 7000 Orange, California 92868-4720 Sherwin Harris, Resource Project Manager Public Utilities Department City of Riverside 3900 Main Street l
Riverside, California 92522 I
d
Southern California Edison Co. R. W. Krieger, Vice President Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, California 92674-0128 Stephen A. Woods, Senior Health Physicist Division of Drinking Water and Environmental Management l
Nuclear Emergency Response Program California Department of Health Seivices P.O. Box 942732, M/S 396 Sacramento, California 94334-7320 Mr. Gary D. Cotton, Sr. Vice President Energy Supply San Diego Gas & Electric Company P. O. Box 1831 San Diego, California 92112-4150 Mr. Steve Hsu Radiological Health Branch i
State Department of Health Services P.O. Box 942732 Sacramento, California 94234 i
l Mayor City of San Clemente 100 Avenida Presidio San Clemente, California 92672 Mr. Truman Burns \\Mr. Robert Kinosian Califomia Public Utilities Commission 505 Van Ness, Rm. 4102 San Francisco, California 94102 l
i
Southern California Edison Co. E-Mail report to T. Frye (TJF) i E-Mail report to D. Lange (DJL)
E-Mail report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK) i bec to DCD (IE01) l bcc distrib. by RIV w/ Enclosure and Attachments 1 and 2:
Regional Administrator JONGS Resident inspector DRS Director DRS Deputy Director DRP Director DRS-PSB Branch Chief (DRP/E)
MIS System Senior Project inspector (DRP/E)
RIV File Branch Chief (DRP/TSS) bec w/ Enclosure and Attachments 1-3:
Laura Hurley R. Gallo, HOLB/DRCH DOCUMENT NAME: R:\\_SO23\\SO8302RP.SLM To receive copy of document, indicate in box: "C" = Copy without enclosures *E" = Copy with enclosures "N" = No copy RIV:SRE:OB E
SRE:OB E
SRE:OB
,, E, RE:OB E
RE:OB l
- SLMcCrory/nh MOMcKernon WEMurptiv/+M TRMeadows MElantz 12/
/98 12/
/98 12/Jl /98' 12/
/98 12/ /98 j
C:OB E
CORS\\E E
C:OB tl l
JLPellet-R/
/ GPick L&oh JLPellet /d 12/L e /98 l 12/2.'l/98 1ff ( /90l
~
'Previously concurred.(
OFFICIAL RECORD COPY p;, - ti ~
i 1
ENCLOSURE 1
i U.S. NUCLEAR REGULATORY COMMISSION REGION IV i
f' t
Docket Nos.:
50-361,50-362 l
License Nos.:
l 50-361/98-302,50-362/98-302 l
Report No.:
Licensee:
Southern California Edison Co.
Facility:
San Onofre Nuclear Generating Station, Units 2 and 3 Location:
5000 S. Pacific Coast Hwy, i
San Clemente, California Dates:
November 15 through December 3,1998 l
Inspectors:
S. L. McCrory, Senior Reactor Engineer, Examiner / Inspector, Chief Examiner I
T. O. McKernon, Senior Reactor Engineer, Examiner / Inspector M. E. Murphy, Senior Reactor Engineer, Examiner / Inspector l
T. R. Meadows, Senior Reactor Engineer, Examiner / Inspector R. E. Lantz, Reactor Engineer, Examiner / Inspector Approved By:
J. L. Pellet, Chief, Operations Branch Division of Reactor Safety i
ATTACHMENTS:
' Attachment 1:
SupplementalInformation :
Final Written Examinations and Answer Keys :
Post Examination Comments l
l
$80020ccz9,.
i t
i EXECUTIVE
SUMMARY
San Onofre Nuclear Generating Station, Units 2 and 3 NRC Inspection Report 50-361/98-302; 50-362/98-302 i
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 l
Stauon facility. The licensee developed the initial license examinations using NUREG-1021,
" Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The NRC examiners administered the operating tes' ts on November 15-19,1998. The facility licensee administered the initial written examinations to all applicants on November 20,1998.
Ooerations The large number of questions missed and the high number of common error responses by most applicants indicated training weaknesses. This conclusion was further i
supported by performance weaknesses observed during the operating examination (Sections 04.1 and O4.2).
The f acility licensee developed an adequate written and operating examination, f
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).
An inadvertent breach of examination security did not result in an examination compromise (Section 05.3).
_~
I i
[
Report Details l
Summarv of Plant Status The units operated at essentially 100 percent power for the duration of this inspection.
- l. Operations 04 Operator Knowledge and Performance 04.1 initial Written Examination a.
Insoection Scope On November 20,1998, the facility licensee proctored the administration of the written i
examination to nine senior operator license applicants and five reactor operator license 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 Findinas Three of five reactor operators and six of eight senior reactor operator 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 app!; cants 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 t
l
=
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 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 Operatina Test a.
Insoection Scope 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 tested on five subjects in four administrative areas with a combination of I
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 unmonitored 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 label: ps5;iioned above the instrument.
l I
l l
l l
)
c.
Conclusions All applicants passed the operating examinations but exhibited some knowledge and 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, ar.d scope.
O5.1.2 Ednination Packaae a.
Inspection Scoce 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 Findinos 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 questiens related to control systems logic as possibly being too difficult to answer without a refvence; 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 l
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 1
l
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 ) 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 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 must be less than 900 ft, which equated to 57 percent. Therefore, Choice C (53 percent) remained as the only correct answer since it was the highest value below 57 percent.
Senior / Reactor Operator Comment 16/11 - The facility licensee recommended accepting Choice B as an additional correct enswer 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 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 i
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
i enough to permit observing the contents of the room. The examiners arrived at about f
6:30 a.m. and found the door ajar. The examination material did not appear to have been disturbed. Oni 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 i
arrival and had not seen him anywhere near the examination material roorn. 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 i
management at the conclusion of the onsite inspection on November 19,1998. After the grading 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 acknowledged the findings presented.
The licensee did not identify as proprietary any information or materials examined during the inspection.
i i
i
l
^
ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED 1
Licensee M. Jones, Manager, Operations R. Sandstrom, Manager, Training
' K. Rauch, Supervisor, Operations Tra!ning T. Frey, Compliance T. Vogt, Operations D. Axline, Licensing L. Gennann, Training i
j l
?
f f
i
I ATTACHMENT 2 FACILITY LICENSEE POST EXAMINATION COMMENTS 1
1 i
i i
RO Exam Conunents 1
4 y.
__m..
l COMMENT #1 RO Examination Quesuon 7 (SRO9) i, The question stern references SO23-3-3.27.3 as do the pc aible answers. The actual procedure that should have been referenced is SO23-3-3.23, Etnergency Diesel Generator Monthly 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 prended Southern California Edison believes there are no correct answers to this question.
i Delete the quesuon J
l i
I l
i i
l l
I l
i l
F I
f k
d i
a
j NUCLEAR ORGANIZATION SURVEILLANCE OPERATING INSTRUCTION S023 3-3.23 UNITS 2 AND 3 REVISION 14 TCH 14-2 PAGE 72 0F 88
{
ATTACHMENT 7 s
A. C. SOURCES VERIFICATION (MODES 1-41 l
.0BJECTIVE
-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 r
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 Ino)erability has placed both Units in action statements, tien 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 Diesel Generator (s), then declare the affected Diesel (s) Inoperable, el 2.4 If desired use the last page of this Attachment to assist in performance of this Attachment.
l^
l ATTACHMENT 7 PAGE 1 0F 7 f
NUCLEAR ORGAfdZATION SURVE1LLANCE OPERATING INSTRUCTION S023-3-3.27.2 UNITS 2 AND 3 REVIS!0N 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 pcwer availability as required by the Administrative Technical Specification.
UNIT 2 MODE UNIT 3 MODE DATE PERF. BY 1.0 PREFEOUISITES 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 DETERMINE the performa7ce 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. LIST the Components and Sections Steps lR i
to be performed. After approval, then CIRCLE N A for i
the remaining unused steps.
COMPONENTS i
SECTIONS / STEPS i
OPERABILITY VERIFICATION PREPARED BY:
Control Room Operator OPERABILITY YERIFICATION APPROVED BY:
SR0 Ops. Supv.
~
ATTACHMENT 1 PAGE 1 0F 7
~
I
j COMMENT 02 R0 EXAMINATION QUESTION #9 (SR0'12) 5023-5-1.8 is the reference for "A" to be a correct answer.
"C" is also correct based on Technical Specification 3.4.6 cnd 3.4.7, which requires the RCS LOOP to be operable.
56uthern California Edison believes there are two correct answers to this question.
Accept answers A & C h
i f
i J
u
NUCLEAR ORGANIZATION INTEGRATED OPERAT3NG INSTRUCTf 3 S023-5-L8 UNITS 2 AND 3 REVISION 9 PAGE 86 0F 91 ATTACHMENT 13 1
I 9.0 kcP OPERATION I
9.1 With at least one RCP operating, reverse flow will be present in the idle loop.
lD 9.2 The loss of RCP heat will affect cooldown rate. Consequently, Tcold i
should be maintained 2125'F to prevent entering the restrictive heatup and cooldown limitations that apply when s120*F.
1 9.3 When securing RCPs, it may be necessary to reduce PIR heater output due i
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 Spray.
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 core. As Core Exit Temperature rises, the RCS will swell into the Pressurizer. Adjusting letiown 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.
r 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 Teold temperature monitoring.
9.9
,ff RCPs are running, Ili1H one RCP shall remain in service until completing RCS boration to Hode 5, or refueling concentration and other forced circulation dependent parameters are met (e.g., hydrogen, i
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 temperature.
\\
F ATTACHMENT 13 PAGE 6 0F 11 10 *d Ct': LI
- 86. Of AoN 91CL-891-6t'6:W3 Wd id1NS t/2 h 5905 J
RCS Loops--MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (kCS) 3.4.6 RCS Loops'--MODE 4 LCO 3.4.6 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 termitted that would cause reduction of the RCi 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 s 256*F unless:
3 a.
Pressurizer water volume 's < 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.
~
SANONOFRE-bNIT2 3.4-18 Amendment No. 127 L-
RCS Loops-MODE 4 3.4.6
^
ACTIONS CONDI' TION REQUIRED ACTION COMPLETION TIME A.
One required RCS loop A.1 Initiate action to Imediately inoperable.
restore a second loop or train to OPERABLE A_ND status.
Two SDC trains inoperable.
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 /> inoperable.
AND Two required RCS loops inoperable.
C.
Required RCS loop (s)
C.1 Suspend all Imediately or SDC train (s) operations involving inoperable.
reduction or' RCS boron concentration.
03 MQ
~
No RCS loop or SDC train in operation.
C.2 Initiate action to Imediately restore one loop or train to OPERABLE status and operation.
l l
=
SAN ONOFRE--UNIT 2 3.4-19 Amendment No. 127
._....._m i
RCS Loops-MODE 4 i
-3.4.6 h-SURVEILLANCE REQUIREMENTS i
SURVEILLANCE L
FREQUENCY SR 3.4.6.1-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.
1 i
h SR 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 1 50% (wide range).
p l
1 i
SR 3.4.6.3 Verify the second required RCS Loop or SDC 7 days train is OPERABLE.
j l
l 1
4 ls -
SAN ONOFRE--UNIT 2 3.4-20 Amendment No. 127 i
4 I
l
RCS Loops-MODE 5 Loops Filled i
3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-MODE 5, Loops Filled LCO 3.4.7 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; i
b.
Reactor Coolant loop 2 and its associated steam generator and at least one associated Reactor Coolant Pump; c.
Shutdown Cooling Train A; or r
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 I
.be greater than 50% (wide range).
1
NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and pumps providing shutdown cooling may be de-energized for s I hour per i
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 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 i
hours for' surveillance testing provided that the other RCS loop or SDC train is OPERABLE and in operation.
____________________.--------------------------- (continued)
SAN ONOFRE--UNIT 2 3.4-21 Amendment No. 127
_ ~. _.._.
RCS Loops-MODE 5, Locps Filled i'
3.4.7
NOTES (continued)---------------------
4~.
No reactor coolant pump (RCP) shall be started with one or more of the RCS cold leg temperatures s 256*F unless:
a.
The pressurizer water volume is < 900 ft3 or b.
The secondary side water temperature in each steam i
generator (SG) is < 100*F above each of the RCS cold leg temperatures.
5.
A containment spray pump may be used in place 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 i
> 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.I.
6.
All SDC trains may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.
tj l
I 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 OPERABLE.
SDC trains /RCS loops to OPERABLE status.
AND E
Any SG with secondary side water level not A.2 Initiate action to Imediately within limit.
restore SG secondary _
)
side water levels to within limits.
l (continued)
- =
SAN ONOFRE--UNIT 2 3.4-22 Amendment No. 127
RCS Loops--MODE 5, Loops Filled 3.4.7 ACTIONS (continued)
CONDITIdN REQUIRED ACTION COMPLETION TIME B.
No SDC train /RCS loop B.1 Suspend all Immediately in operation.
operations involving reduction in RCS boron concentration, b.ND B.2 Initiate action to Immediately restore required SDC train /RCS loop to operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 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.
SR 3.4.7.2 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).
SR 3.4.7.3 Verify the second required RCS loop, SDC 7 days tFain or steam generator secondary is OPERABLE.
.s==-
l SAN ONOFRE--UNIT 2 3.4-23 Amendment No. 127
l t-COMMENT #3 RO Examination Quesuon 14 l
(SROl8) l I
The generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. The Reed Switch Position Trnaamitters, RSPT's, actually sense the CEA's position. The Control Element Assembly C+ %r, CEAC, uses the input from the RSPT and sends a signal to the alarm. Both --+;-wnts are n=w to generate a deviatwo alarm. Southern California Edison believes there are two correct answers to this question.
Acapt answers B & C l
I l
[
l t
i
. ~
NUCLEAR ORGANIZATION ALARM RESPONSE INSTRUCTION 5023-15-50.Al UNITS 2 AND 3 REVISION 2 PAGE 710F 76 ATTACHMENT 2 50A28 CEA DE.VIATION APPLICABILITY PRIORITY REFLASH ASSOCIATED WINDOWS Modes 1-3 AMBER NO NONE INITIATING NOUN NAME SETPOINT VALIDATION PMS ID LINK #
DEVICE INSTRUMENT U2/U3
- (3)LO91, CEAC 1 Control Element 5 Inches NONE DEVIAR56 641/663 (R
or CEAC 2 Assembly Deviation 1.0 REOUIRED ACTIONS:
1.1 Position the CE0MCS Mode Selector Switch on 2(3)CR50 to 0FF.
1.2 Verify which CEA is misaligned and the amount of misalignment, by observatynofthefollowing:
CEAC display CRT CEAC remote operators modules
- PMS alarms
- PMS readout 2.0 CORRECTIVE ACTIONS:
SPECIFIC CAUSES SPECIFIC CORRECTIVE ACTIONS 2.1 Misaligned CEA 2.1 After the misaligned CEA has been determined,.tAgn:
i 2.1.1 Notify the SRO 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 Control Element Assembly.
3.0 ASSOCIATED RESPONSES:
3.1 Notify the CRS/SS and the STA to review Tech. Specs. LCO 3.1.5 and LCS 3.1.105, and initiate an EDMR/LC0AR, as required.
=J=.
i
f l
NUCLEAR ORGANIZATION SYSTEM CESCRIPTION $0 5023-710 UNITS 2 AND 3 l
REVISION 3 PAGE 72 CF 75 FIGUDE !!*
CONTROL ELEMENT A$$EMBLY SUBGROUP REED SWITCH DSITION TDANIMITTER SIGNAL A$$1GNME4TS I
l
. h E.X CORE CHANNEL RSPTO RSPT\\
l,,,
23 CEAS i,
i f
V l
I 22 CEAS CEAS 23 CE AS 22 CEAS RS 2
45CDS 45CEAS TEMQM\\
22 CE O
g 1
y 2
23 CEA1 ISOLATON _
y
/
CALCULATOR CALCULATOR
\\
CEA POSITch NO 1 NC.2 ggA pCSf7CN ISOLATCN
^ f]{
],08gfy ISO (ATON j
AMPLIFIER y
AMPLftER CA?A uMG CATA uhms
-H-i i
i PROTECTckl A CORE 8 CORE C CCRE QCORE PROTECTCN PROTECTCN PROTECTCN CALCutATOsd CALCULATOR CALCULATCR CALCULATCR 1
1 CPERATOR1 OPERATOR $
OPERATOR 1 CPERATCR S MOOwtE McCULE ko0VLE MooutE CRT DISPLAY NOTES.1. SIGNAL PMCM CEA 2 IS CCNPECTED TO CPC's A AND C, BUTIT is NCT USED As A TARGET CEA.
CNLY 2: CF THE 23 SGNALS ARE USED AS TARGET CEAs.
i s'*;ET
- t
.~
~
.. =..
I l
COMMENT # 4 j
RO EXAMINATION QUESTION 26
. cilectrical 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 queshon.
I 1
i i
l e
i 9
l i
.i i
t i
I i
T l
i'
1 e
1
~ COMMENT 05 R0 EXAMINATION QUESTION #27 (SR0 28)
Answer "B" is correct because pressurizer spray valves are open at 2300 psia. Answer "D" is also correct because the backup heaters turn 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 answers to this question.
Accept answers B & D
)
j l
l 1
l
i
. NUCLEAR ORGANIZATION
\\
SYSTEM DESCRIPTION SD-SO23-260
. UNITS 2 AND 3 REVISION 5 Page 168 of 205 l
I;-
FIGURE 1115 PRESSURIZER PRESSURE CONTROL SYSTEM 8LdCK DIAGR I
l
[
1EHEATER
. FMS/CFMS FMS/CFMS ~>'EHE#IER 1
o
~ STEAM EYFASS SYSTEM STEAM EYPASS SYSTEM ~,
E/S EIS RED 0-GREEN A-l lA HS-100A 1
2200/2225 R DRDER 220C12225 X
P 0100-Al Hl/L E/S i
RED 7'
PMS Au M 0
50A14 (2275/2175)
O' c
~.~3
~>
FF UH
~R j
(2200 j
c gf3 Hl HI i
=
"b"MI&R Q
C
=
o PIC 0100 PROPORTIONAL I
I DE-ENERGlZE HEATERS E/U HEATERS l
(2275) o mom o
o o o V/4 Mi 4W/4 e iRJgggf{lON I
Hl/L g A
VALVC PCymON mre 50 1 o
o i-(2275/2175)
"'[MMM" o-PV 100A PV 1005
.. ren c3: ge s:: :
l I
+
- +
M e*-
- py
COMMENTM RO Examination Question 37
- (SRO37)
Procedure, SO23 12-7. Loss of Offsite Power / Loss of Forced Circulation, floating step 2 states that That and CET's are ar,,a. red as are Thot and Tc are not rising to verify natural circulation is occurnng. The S/G pressure can be used to correlate to Tc there fore answer B is also correct. Southern California Edison believes there are two conect answers to this quesuon.
Accept answers A & B t
..--.. - _ - -. - -. ~ _ -. _ _ _. - -. ~ _ -,
NUCLEARCfGAN'IAT*C's UNITS 2 m 0 3 EMER3ENCY CFERATING INSTRUCTICN R: VISION 15 SC23-12 7
- TACHMENT 2 FAGE 30 CF 122 1
LOSS OF FORCED CIRCULATICN/ LOSS CF 0FFSITE FLOATING STEPS 1
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
\\
NOTE:
Lcw flew during Natural Circula icn slows RCS i
response to temperature changes.
Lee: t rar.s i t time rises to between 5 minutes and 10 minutes. <
l FS-2 MONITOR Natural Circulation Established:
i CHECX all RCPS - stopped.
a.
i GO TO FS-4, MONITOR RCP Operatir.g a.
Limits.
b.
CHECKatleastoneS/G b.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1)
SBCS - operating OR INITIATE 5023-12-9, Atta:hment 8 I
RECOVERY - HEAT REMOVAL.
ADV - operating.
AND
\\
i 2)
Feedwater - available.
i t
CHECK operating loop AT - less c.
than 58'F.
IF any criteria c thrcugh g NOT o
satisfied.
d.
CHECK Tc and Ts - NOT rising.
THEN CHECK Reactor Vessel Level i
e.
t (Plenum) - greater than or MAXIMIZE S/G level - less than C*i NR.
equal to 100%t.
QSPDS page 622 P';SEavailableS/Gsteaming rate.
CFMS page 312 t
1 RAISE Core Exit Saturation Margin f.
CHECX operating loop Ta and REP
- greater than 20'F j
CET - within 16*F:
QSPDS page 611 i
CFMS page 311.
e 311.
i
)
c=r
(
ATTACHMENT 2 PAGE 4 0F 29 4
l'
~* *
- l'
. NUCLEAR CRGANIZATICN UNITS 2 AND 3 EMERGENty cpgR; TING INSTRUCTICN SC23-12-7 REVISION 15 fg. -
ATTACHMENT 2 PAGE 31 c': 393 LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWE FLOATING STEPS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-2 MONITOR Natural Circulation f'
Established:
(Continued) 9 CHECK Core Exit Saturation
. Margin - greater than 20*F:
IF any criteria c through g NOT e
satisfied, OSPOS page 611 THEN CFMS page 311.
i MAXIMIZE S/G 1evel - less than 80% NR.
RAISE available S/G steaming rate.
RAISE Core Exit Saturation Margin
- greater than 20'F:
~
i QSPOS page 611 CFMS page 311.
j 3
.s N:
ATTACHMENT 2 PAGE 5 0F 29
, ; c :.Tn t.1:r :.O;F.: Ye;'!:'.V:.'F.T'.M5??fg'.iBW' pvl~FJ7Tce'm?$q!<,77.?F,~,.q;O 7:9~ yAb d t.
3
. :. ~...
- ..'~~-
.9~"
+
v
- -. ~. _ - -.
-~.
-- -...- ~. -.
t i
COMMENT #7 i
l
. RO Examination Question 46 (SRO43) i Answer B is correct based on the strictest interpretation ofimmediately before and after a trip. Immediately before the trip, S/G level cy* 85% causing a High Ise! Override, HLO, signal to be applied to the main and b pass 3
feed 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 f:ed water regulating control system. With a reactor trip, an RTO signal is j
sent which as soon as the HLO condition clears seconds aAer the trip due to normal shrink of water levels, the i
RTO signal is apphed stad the bypass valve opens to 50%, answer C. Southern California Edison believes there are two correct answers to this question.
Accept answers B & C I
l t
l
\\
I l
1
COMMENT 08 R0 EXAMINATION QUESTION #54 S023-6.2.9 states " Select the Bus Transfer Control AUT0/ MANUAL switch 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.
1 Southern California Edison believes there are no correct answers to this question.
Delete question
M e,
NUCLEAR ORGANIZATION OPERATING INSTRUCTION S023-6-2 UNITS 2 AND 3 REVISION 5 PAGE 7 0F 26 m.I 6.0 PROCEDURE (Continued) 6.2'.B If the INCOMING 4160V source is a bus tie or diesel generator output breaker, then open the RUNNING breaker, if desired.
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" configuration.
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 synchrcscope 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 circuit must be de-energized and its I
synchroscope removed from service prior to starting the evolution on the remaining Bus Tie Breaker.
6.2.13 For 1E 4160V Bus Tie transfer schemes, perform the foilowing:.
.1
. Starting from the Unit which is to SUPPLY power:
i 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 AN) a Transformer breaker on each Unit is Closed, BD.T!!
Units will have BUSES PARALLELED alarms annunciated.
.1.1 Place the synchroscope in service per Step 6.2.3.
$ e-9
COMMENT #9 RO Examination Question 74 (SRO69)
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 prmided 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.
f
._. ~
i l
t
(
COMMENT #10 l
RO Exanunation Question 79 (SRO 75) l l
Question was based on old Tavg program New program has normal pressurizer level of 48% at 100% power.
This is based on the reduced Tc program of 548 deg F @ 100% power. The lower Tc at full power equates to a
.l Tave of $74 deg F. Per the attached reference, the expected level would be 48% and no additional charging pumps l
would be operating. Southern California Edison believes there are no correct answers for this question and the quesuon should be deleted from the txamination Delete Question.
1 i
.1 l
NUCLEAR ORGANIZATION OPEPATING INSTRUCTION 3N
,0 op 34 UNITS 2 AND 3
- U REVISION 8 PmE 3 ATTACHMENT 5 PRESSURIZER LEVEL CONTROL PROGRAM 70 l
l i
t t
.i...i.
.i.
.i.
..e a
6 0 -"
n
,3R J
i W
...o..<.
3.. 3.
3.. 3...,....,.
..s..
w J
t t
.n.
.t.
.p...g...<.
8.
,.1 h.
W N
.lf
..... l
.........,...Q C
3
...........:.................e...e...<...
VJ
...........e...t.....-.
.(.....
6..
..s...s....
..{...{...
W 40 e
n
...i...f.. 4....e...;...........
.,...)...)...;...,...f.....
4.. 4....s.
..g.........'....i...'.,..
. ;... I... ;.. l.... ;.. s....
....y.
.....a.y...y....y.
...<...g..
30
.. r... i i.........
4 544 550 560 570 58Q 590 RCS AVERAGE TEMPERATURE { F) 010 - 1.C H T i
I i
I
~
D10-8.wS1 ATTACHMENT 5 PAGE 1 0F 1
i COMME.NT #11 i
RO EXAMINATION QUESTION 85 l"
(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 schon will be required and answer B would be correct. Ifit is assumed vacuum will continue the current trend, the listed acuon of"D* could be taken to return the plant to a more stable condition. Southern California Edison believes "B" & *D" are correct answers l
\\
Accept B & D.
i h
t i
P i
i I
i i
1 l
NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTICN S023-13-10 UNITS 2 AND 3 REVISION 2 PAGE 7 0F 13 LOSS OF CONDENSER VACUUM
~
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 4 Actions for loss of vacuum due to Condenser fouling:
CAUTION i
During periods of heavy influx, rapid and aggressive action may be required in :
{ g,
{
order to avoid a Unit trip. Power may need to be reduced in order to:
i 4
Bumpand/orStopCirculatingWaterPumpsontheCondenserquadrantswith j e
the highest. differential pressures Maintain Condenser backpressure < 3.5" Hg l
- a. REDUCE Regetor power to 75% TO 854.
l
- b. BUMP Circulating Water Pump (s) per direction of Shift Superintendent.
j
- c. VERIFY backpressure < 3.5" Hg c.
- 1) REDUCE Reactor power j
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 l
Tubesheet/High AP/ Debris f-Removal.
l
- 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.
)
l l
i l
l t
l COMMENT # 12 RO Examination Question 93 (SRO93)
Answer C is correct based on HV9217 and HV9218 being open and providing a direct path from inside i
containment to the outside, in this case to the VCT Answer D is correct because given this event, Controlled Bleed j
O'E 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 3
l Accept answers C & D 1
1 l
t
COMMENT # 13 RO EXAMINATION QUESTION 95 (SRO %)
l Both answers 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. Southern California Edison believes there are two mrrect answers to this question.
Accept answers A & B.
)
i j
I i
i v,
-c w
+
,__.___m.
. _ _ _ ~ _.. _ _. _... _
_~
NUCLEAR CRGANIZATION EMERGENCY OPERATING INSTRUCTION 5023-12
- JUNITS 2 AND 3 REVISION 15 PAGE 43 0F 113 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT i
FLOATING STEPS ACTIONS / EXPECTED RESPONSE PESPONSE NOT CBTAINED FS-6 CHECK HPSI Throttle /Stop Criteria:
l
.a.
CHECKatleastoneS/G a.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1
1)
SBCS - operating i
INITIATE S023-12-9. Attachment 8, OR RECOVERY - HEAT REMOVAL.
ADV - operating.
' AND -
.2)
Feedwajar - available.
b..
CHECK PZR level o
IF any criteria of steps b through d
-l NOT met,
- greater than 30%
l THEN t
AND l
OPERATE Charging and HPSI systems'
- NOT lowering.
as necessary to maintain
{
t Throttle /Stop criteria
- c. : CHECK Core Exit Saturation
- satisfied.
l 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.
i i
i ATTACHMENT 2 PAGE 13 0F 43 3
x n
e g
w
- NUCLEAR.CRGANIZAT!;R EMERGENCY CPERATIN3 INSTRUCTICN SC23-12-5 UNITS 2 AND 3 REVISION.15 PAGE :: CF 1 3 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT FLOATING STEPS ACTIONS / EXPECTED RESPONSE RESPONSE NOT CBTAINE3 FS-6 CHECK HPSI. Throttle /Stop Criteria:.(Continued)-
d.
CHECK Reactor Vessel Level
.o IF any criteria cf stems t thrcuen d (Plenum) - greater than or NOT met,
~
equal to 100%:
THEN QSPDS page 622 CFMS page 312 OPERATE Charging and HPSI systsis as necessary to maintain
{
Throttle / Step criteria
- satisfied.
THROTTLE Locp Injecticn Valves.
ENSURE auxiliaries to $! pumps:
a)
Electrical power to pumps and valves.
bl 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 l
Figure 2.3-1, OR RCS Cooldown NOT in progress.
.. f. THROTTLE OR STOP HPSI as required one train at a time.
g.
STOP charging pumps as required one at a' time.
)
L-7,;.
ATTACHMENT 2 PAGE 14 0F 43 m
i
- NUCLEAR ORGANIZATICN EMERGENCY GPERATING INSTRUCTICN SC23.2-5
- UNITS 2'AND-3 REv!5ICN 15 PAGE 45 CF 1:3 ATTACHMENT 2 i
l EXCESS STEAM DEMAND EVENT l
FL0ATING STEPS ACTIONS / EXPECTED RESPONSE.
RESPONSE NOT OSTAINED i
- FS-6' CHECK HPSI Throttle /Stop Criteria:
(Continued) h.
MAINTAIN Criteria of steps a through e - satisfied.
j
- i. CHECK Containment pressure
- i. 1)
ENSURE SIAS~- actuated.
- less than 3.4 PSIG.
2)
GO TO FS-7, CHECX LPSI Termination Criteria, j
- j. CHECK PZR Level _
- j. INITIATE FS-22 ESTABLISH CVCS I
--less than 80%.
Letdown Flow.
k.
RESET SIAS per 5023-3-2.22,.
ESFAS OPERATION.
1
)
t 4
q.
ATTACHMENT 2 PAGE 15 0F 43
t I
l i
COMMENT # 14 RO EXAMINATION QUESTION %
(SRO 97) j~
Answer "B" is correct based on the information given. However answer "A" is also correct based on procedure j
SO23 12-7 Safety Function Status Checks, which requires ahling > 20F or you are directed to the Functional l
i Recovery for not meeting natural circulation. Southern California Edison believes there are two correct answers to this questson.
)
f i
Accept answers A & B.
l 1
i i
l l
l' l
1 ~.
l i
t i
NUCLEAR CRGAN!ZATICN UN!?S 2 AND 3 EMERGENCY OPERATING INSTRUCTION 5023-12-7 REV! SIGN 15 ATTACHMENT 2 PAGE 30 CF 122 LOSS. OF FORCEO CIRCULATION / LOSS OF OFFSITE POWER FLOATING STEPS ACTION / EXPECTED RESPONSE R,.FSPONSE 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.
-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, OR RECOVERY - HEAT REMOVAL.
ADV - operating.
AND
'2)
Feedwater - available.
c.
CHECK operating loop AT - less o
IF any criteria c through g NOT than 58'F.
sati sfied, I
d.
CHECK Tc and TH - NOT rising.
THEN e.
CHECK Reactor Vessel Level MAXIMIZES /G1evel-lessthan (Plenum) - greater than or 80% NR.
equal to 100%:
RAISE available S/G steaming OSPDS page 622 rate.
CFMS page 312.
RAISE Core Exit Saturation Margin J
- greater than 20'F:
f.
CHECK operating loop Ts and REP CET - within 16*F:
OSPDS page 611 CFMS page 311.
~
- =-
ATTACHPENT 2 PAGE 4 0F 29
-., y
.i NUCLEAR CRGANIZATICN i
UNITS 2 AND 3 EMERGENCY OPERATING INSTRUCTION 5023-12-7
~
REVISION 15 ATTACHMENT 2 PAGE 310F 122 LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWER 5
FLOATING STEPS ACTION / EXPECTED RESPONSE RESPONSE NOT O8TAINED FS-2
. MONITOR Natural Circulation Established:
(Continued) g.
CHECK Core Exit Saturation IF any criteria c through g NOT l
o Margin - greater than 20*F:
satisfied, OSPOS page 611 THEN l
CFMS page 311.
MAXIMIZE S/G 1evel - less than i
80% NR.
RAISE available S/G steaming rate.
RAISE Core Exit Saturation Margin
- greater than 20*F i
QSPDS page 611 l
CFMS page 311.
i l
t i
i
/
po ATTACHMENT 2 PAGE 5 0F 29
\\
I l
SRO Exam Comments 4
$=*
COMMENT #1 SRO Exanunaten Quesuon 9 (RO7) l The quesuon stem references 5023-3 3.27.3 as do the possible answers. The actual procedure that should have been referenced is 5023-3-3.23, Emergency Diesel Generator Monthly Surveillance. The given procedure, S023-3 3.27.3 is not used to perform the AC Sources check that is required by the given scenario. The correct answer was not provided. Southern California Edison believes there are no correct answers to this question.
Delete the question.
1 I
i
l I
NUCLEAR ORGANIZATION SURVEZLLANCE OPERATING INSTRUCTION 5023-3-3.23 UNITS 2 AND 3 REVISION 14 TCN 14-2 PAGE 72 0F 88 ATTACHMENT 7 A. C. SOURCES VERIFICA110N (MODES 1-4)
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).
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;l If the cause of the Diesel Generator Inoperability exists on the other Diesel Generator (s), then declare the affected Diesel (s) Inoperable, of 2.4 If desired use the last page of this Attachment to assist in performance of this Attachment.
ATTACHMENT 7 PAGE 1 0F 7
I i
i NUCLEAR ORGANIZATION SURVE!LLANCE OPERATING INSTRUCTION 5023-3-3.27.2 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 IE 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 i
availability as required by the Administrative Technical Specification.
i UNIT 2 MODE UNIT 3 MODE DATE 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.
l 1.2 DETERMINE the performance requirements of this attachment, as follows:
)
SR0 Ops.
Supv.
O This Attachment is being performed for a scheduled surveillance.
O This Attachment is being performed for operability verification. LISTtheComponentsandSections/ Steps lR to be performed. After approval, then CIRCLE N/A for the remaining unused steps.
COMPONENTS SECTIONS / STEPS OPERABILITY VERIFICATION PREPARED BY:
Control Room Operator OPERABILITY VERIFICATION 2
APPROVED BY:
SR0 Ops. Supv.
=
ATTACHMENT 1 PAGE 10F 7
h i
f COMMENT #2 SRO Examination Question 10 i
infrequently performed test can also be interpreted to be special tests. SO123-IT 1, Infrequently Performed Tests, states that infrequently performed tests can also be performed under the special test procedure. Since 5O23-0 23 is l
also used to conduct short term valve lineup changes, it too is a correct choice. Southern California Edison beheves that there are two answers to this question.
j l
Accept answers A & D l
i l
l 1
k 1
i.
NUCLEAR ORGANIZATION GENERAL ORDER 50123-1T-1 UNITS 1, 2 AND 3 REVISION 4 PAGE 3 0F 16 III. RESPONSIBILITIES (Continued)
F.
The Manaaement Desionee (see Definitions, Attachment A) exe"cises continuous responsibility for Management Oversight. With the tsproval of the Vice President, Nuclear Generation and/or the Senior Vice President, Power Generation, may exercise Management Oversight on a " spot-check" basis.
't G.
The Test Scecialist (see Definitions, Attachment A) is a technical resource to the supervisor who has operational responsibility for conduct of the test or evolution.
H.
Licansed Goerators and Plant Manacement Staff (see Definiticns, I
Attachment A) have the responsibility to recognize tests and evolutions which are (or should be) included in the IPTE Lis.
IV.
REQUIREMENTS 4
A.
Infrequently Performed Tests and Evolutions (IPTE) which take plant personnel or le uf pment beyond the bounds of normal procedures, trajningroperatTng-bans, vi spe-ience. W which rene"en*e
.significant safety or economic risk, require controlling documents with enhanced development and review as outlined by this order, j B.
Execution of IPTE activities require Management Oversigh Definitions, Attachment A) with clear direction, clear cor:munication of management expectations with respect to margins of safety, expected plant response, termination criteria, and actions to be 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 and 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, particularly when uncertainties or unexpected plant behavior is encountered.
E.
Req 61rements for test or evolution termination shall be clearly defined, communicated, 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.
1E it is necessary to change the IPTE controlling document prior to or during use, M the associated affect on IPTE fntent (see Definitions, Attachment A) and plant safety shall be considered.
1E the intent is affected, M prior to starting or continuin3; with the IPTE, the same approval level as the original controlling document is required. Prior to any IPTE controlling document changes, extreme caution and consideration should be exercised.
- =
NULLEAR ORGANIZATION OPERATIONS DIVISION PROCEDURE 50123-0-23 UNITS 1, 2 AND 3 REVISION 5 PAGE 47 0F 62 ATTACHMENT 4 KEYPOINTS FOR ABNORMAL ALIGNMENT / EVOLUTION PAGE 1 0F.13.]
COMPONENT: [1]
LOG NUMBER
- [2]
PURPOSE OF ALIGNMENT:
I41 Procedure Change Required O No O Yes i
Verify this document is current by checking a centrolled copy or by using the method described in 50123-VI-0.9. [5]
i EFFECT OF ABNORMAL ALIGNMENT / EVOLUTION NO YES l
Has it been addressed in a completed:
// O INDICATE document Type and Number:
50.59 Safety Evaluation. E O ATTACH PF(123)l09-1, Unreviewed Safety Question Screening PF(123)109-1, Unreviewed Safety Criteria Question Screenir.g Criteria? [6]
/j 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, Change the intent of the Operating Operations prior to Instruction, E implementation.
Constitute an Evolution, E Require a new or additional 50.59 [8]
Could it pose adverse environmental O
O DO NOT PERFORM until a review effects of any] type directly or from Environmental Protection indirectly? [9 is attached.
Is it involved with multiple evolutions O
O ATTACH Marked-up P& ids, A.nji on the same system, an interconnected OBTAIN approval from the Shif t system, E wilJ result in the mov.ement Superintendent as the SR0/CFH.
of gases. ort 1guids? [10]
'~
Is 't'a complex alignment which:
O O OBTAIN approval from the Plant Is requested by another division, E /
Superintendent or designee as i
Requires non-routine interdivisional /
the Plant Management Staff-
/ coordination, E
/
Operations.
Installs temporary plant equipment that could alter the functio of owpath of existing pla com nents? [11]
PREPARED / REVIEWED & APPROVED DATE TIME l
PREPARER (12]
l MANAGER, REQUESTING CRGANIZATION O N/A (13]
PLANT MANAGEMENT STAFF - OPERATIONS (13]
[13)
MANAGER. CPERATIONs (14]
ATTACHMENT 4 PAGE 1 0F 14 4
NUCLEAR ORGANIZATION CPERATIONS O!UISION PROCEDURE 50123-0-23 UNITS 1, 2 AND 3 REVISION 5 PAGE 57 0F 62 ATTACHMENT 4 KEYPOINTS FOR ABNOPMAL ALIGNMENT /EVOLUTtqH (Continued) 9.
H this Abnormal Alignment / Evolution could pose any type of adverse environmental ef fects, Lhan Environmental Protection must review this permit before implementation, inji 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 j
the movement of gases or liquids, than check YES, and attach marked-up IR P& ids.
In November 1993, failure to properly evaluate system interconnection flowpaths resulted in HPSI Pump run-out and caused extensive pump damage. Drawings are required to assist in Abnormal Alignment / Evolution review and tailboard, and therefore are not recui-ed 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 l
provided. The individual preparing the document SHALL NOT sign any of the Reviewed and Approved By lines.
Approval is normally required prior to using)the document.
13.
(Refer to main body, Steps 6.1.4 and 6.8.5.1 for exception.
H the Abncmal 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 Check the N/A box in the " Manager, Requesting Organization" 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 is obtained within 14 days of SR0/CFH Approval.
15.
Enter the s alignment.(pecific document number that will allow closure of this e.g., Closure of a WAR, TFM, or NCR). H closure is " completion of this al'ignment", and no other documents will be involved, lhen state so.
H a procedure change is required, then check YES. OPG should also be notified (e.g., E-Mail).
g 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 1
Such information does not change the intent, method, or outcome of the Abnormal Alignment / Evolution.
16.
Insert the number and name of the associated System Operating Instruction (s).
17.
Enter any pertinent additional references (e.g., Technical Manual, UFSAR, Site Procedure, etc.).
If none, then check NO.
ATTACHMENT 4 PAGE 11 0F 14
-. - ~. - _ -
-. -. ~ - -
?
NUCLEAR ORGANIZATION
- UNITS 1. 2 AND 3 OPERATIONS DIVISION PROCEDURE S0123-0-23 REVISION 5 PAGE 10 0F 62
(
6.0 PROCEDURE,(Continued) 6.4 Control of System Al f onments Af fected By Svstem Modifications 6.4.1 Permanent facility modifications will be accounted for by I
TCHs or revisions to the system Operating Instructions.
NOTE:
50123-0-22 provides specific direction regarding control of system alignments due to temporary l
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 /> i
should be accounted for by TCNs or Revisions to the system t
Operating Instruction (s).
.1 Whan 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 /> 1han Section 6.8 should be used to document the change.
.2 Whan the temporary modification is expected to last for I
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 procedure TCN or revision, ihan Section 6.8 should be used i
to document the change. This is allowed provided the j
Operations Procedures Group is actively preparing a.TCN or revision for issuance.
END OF SECTION 6.4 I
l l-4
)
p===.
J
COMMENT s.1 SRO EX AMINATION QUESTION 12 (RO 9) 3 5-18 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. w hich
.. quires the RCS LOOP to be operable. Southern California Edison believe there are two correct answers to this question.
Accept answers A & C l
I i
1 i
1 i
I i
l i
i 1
I J
(
NUCLEAR ORGAN!ZATION INTEGRATED OPERATING INSTRUCTION S023-5-1.8 UNITS 2 AND 3 REv!SION 9 PAGE 86 0F 91 ATTACHMENT 13 9.o keP OPteATI6N 9.1 With at least one RCP operating, reverse flow will be present in the idle loop.
lD 9.2 The loss of RCP heat will affect cooldown rate. Consequently, Tcold should be maintained zl25'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 m&y be necessary to secure all RCP5 and main spray prior to initiating Auxiliary spray.
l Otherwise. loss of NPSH 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 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 SDC is injecting, if the RCP operating in that loop is stopped or when the 1ast RCP is stopped. This is due to cooler SDCS injection water flowing over the loop Tcold temperature element.
r 9.7 If any RCPs are operating, then the Tcold associated with an operating RCP should be used for RCS temperature monitoring.
9.8 When there are no RCPs operating, then TR-0351A (T351X), SDC Combined outlet Temperature, should be used for Teold temperature monitoring.
9.9 TE RCPs are running, IER one RCP shall remain in service until completing RCS boration to Mode 5, or refueling concentration and other j
i forced circulation dependent parameters are met (e.g., hydrogen, f\\
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 temperature.
i t
ATTACHMENT 13 PAGE 6 0F 11 T0 *d Cv:21
- 86. 01 ^0N 9tC2-891-606:n.d WB TdDO t/C n s90s
- -. ~.
RCS Loops--MODE 4 3.4.6 f
3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops-MODE 4 LCO 3.4.6 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 l
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.
Pressurizer water volume is < 900 ft, or 3
b.
Secondary side water temperature in each steam generator (SG) is < 100*F above each of the RCS cold l
leg temperatures.
APPLICABILITY:
MODE 4.
L I
l l
i SAN ONOFRE-_ UNIT 2 3.4-18 Amendment No. 127
l...
RCS Loops-MODE 4 3.4.6 l^-
ACTIONS CONDI' TION REQUIRED ACTION COMPLETION TIME A.
One required RCS loop A.I Initiate action to Imediately inoperable.
restore a second loop or train to OPERABLE M
status.
Two SDC trains inoperable.
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 /> inoperable.
E Two required RCS loops inoperable.
-C.
Required RCS loop (s)
C.1 Suspend all Imediately or SDC train (s) operations involving inoperable.
reduction of RCS boron concentration.
9.3 AND No'RCS loop or SDC train in operation.
C.2 Initiate action to Imediately restore one loop or train to OPERABLE status and operation.
l I
i
'SANONOFPbUNIT2 3.4-19 Amendment No. 127
.. -. ~.
- ---.... --..=-.__._ _.-= _.-.~.~.
- 8..
l RCS Loops-MODE 4.
3.4.6 I
i
^
SURVEILLANCE REQUIREMENTS' i
l SURVEILLANCE FREQUENCY f
SR 3.4.6.1 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 /> i
is in operation.
I 4
i
'SR 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 it 50% (wide range).
j
!j SR 3.4.6.3 Verify the second required RCS Loop or SDC 7 days train is OPERABLE.
J 1
~.
i 7
. SAN ONOFRE--UNIT 2 3.4-20 Amendment No. 127 1
j
RCS Loops--MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops 2 MODE 5, Loops Filled LCO 3.4.7 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 Pump; 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).
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:
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 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 RCS loop or SDC train is OPERABLE and in operation.
(continued)
SAN ONOFRE--UNIT 2 3.4-21 Amendment No. 127
-4 l
RCS Loops-MODE 5, Loops Filled 1
3.4.7
NOTES (continusd)---------------------
4 No reactor coolant pump (RCP) shall be started with one i
or more of the RCS cold leg temperatures s 256*F unless:
c i
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 place of a low pressure safety injection pump in either or both i
shutdown cooling trains to provide shutdown cooling flow provided the reactor has been subcritical 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.
1 APPLICABILITY:
MODE 5 with RCS loops filled.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1
A.
Less than the required A.1 Initiate action to Imediately
-SDC trains /RCS loops restore the required OPERABLE.
SDC trains /RCS loops i
to OPERABLE status.
AND 0.8 Any SG with secondary side water level not A.2 Initiate action to Immediately within limit.
restore SG secondary i
side water levels to within limits.
(continued)
SAN ONOFRE--UNIT 2 3.4-22 Amendment No. 127
RCS Loops--MODE 5, Loops Filled 3.4.7 ACTIONS (continued)
CONDITIC'N 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 Inmediately restore required SDC train /RCS loop to operation.
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
\\
SR 3.4.7.1 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.
SR 3.4.7.2 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).
SR 3.4.7.3 Verify the second required RCS loop, SDC 7 days tfain or steam generator secondary is OPERABLE.
s==-
SAN ONOFRE--UNIT 2 3.4-23 Amendment No. 127
- l. -
t COMMENTN SRO Examination Quesuon 18 (RO14)
The generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. De Reed Switch Position Transminers, RSPT's, actually sense the CEA's position. The Control Element Assembly Calculator, CEAC, uses the input from the RSFT and sends a signal to the alarm. Both componcats are needed to generate a devution alarm. Southern California Edison believes there are two correct answers to this question.
Accept answers B & C i
t 1
l I
I 1
_ _ _ _ - - - ~ ~ - -
i NUCLEAR ORGANIZATION ALARM RESPONSE INSTRUCTICN 5023-15-5C.Al l
UNITS 2 AND 3 REVISION 2 PAGE 710F 76 i
ATTACHMENT 2 j
i 50A28 CEA DEVIATION l
APPLICABILITY PRIORITY REFLASH ASSOCIATED WINDOWS l
i Modes 1-3 AMBER NO NONE l
INITIATING NOUN NAME SETPOINT VALIDATION PMS ID LINK e OEVICE INSTRUMENT U2/U3 l
l 2(3)LO91, CEAC 1 Control Element 5 Inches NONE DEVIAR56 641/663 lR j
or.CEAC 2 Assembly Deviation 1.0 REOUIRED ACTIONS:
1.1 Position the CEDMCS Mode Selector Switch on 2(3)CR50 to 0FF.
l 1.2 Verify which CEA is misaligned and the amount of misalignment, by observati,on of the following:
CEAC display CRT CEAC remote operators modules PMS alarms PMS readout 2.0 CORRECTIVE ACTIONS:
SPECIFIC CAUSES SPECIFIC CORRECTIVE ACTIONS 2.1 Misaligned CEA 2.1 Af.11t the misaligned CEA has been determined,.thsta:
2.1.1 Notify the SR0 Ops. Supv.
2.1.2 Realign the CEA per 5023-3-2.19, Section for Manual Individual
(
Operation.
2.2 Slipped or Dropped CEA 2.2 GO TO S023-13-13, Misaligned Control Element Assembly.
3.0- ASSOCIATED RESPONSES:
3.1 NotifytheCRS/SSandtheSTAtoreviewTech. Specs.LCO3.1.5and LCS3.1.105,andinitiateanEDMR/LC0AR,asrequired.
NUCLEAR ORGANIZA710N
- UNITS 2 AND 3 SYSTEM DESCRIPTION $D-5023 710 REVISION 3 PAGE 72 0F 76 FIGURE th CONTROL ELEMENT A$$EMBLY $UGGROUP #EED $ WITCH POSITION TRANSMITTER TIGNAL AS$1GhMENTS 1
h Ex CORE CHANNEL P$7TO RSPT\\
d 23 CEAS N
h RSPT
- RSPT 27CFAS t
v21
/
t i
45 CEAS h
W5F?QM\\
22 CE 1
y 2
g 23 CEAa C
CEA k
CALCULATOR CALCULATOR N
CEA POSITCh NO t NO. 2 CEA POSITCN
[O g,g,,c, ogd,jc, au a==
ame i
i A CORE SCORE C CORE D CORE PROTECTCN PROTECTION PROTECTON PROTECTON CALCULATOR CALCULATOR CALCULATOR CALCULATOR I
1 I
1 CPERATORS OPERATORS OPERATOR'S WODULE MODULE OPERATOR's MODULE CRT DiSPtAY MODULE NOTES.1. SiONAL PROM CEA 2 IS CCf#ecTED TO CPC's A AND C. BUT (T IS NOT USED AS A TARGET CEA.
2 SIGNAL PROM CEA 3 IS CONFECTED TO CPC's B AND D. StfTITIS NOT USED AS A TARGET CEA.
- 3. SIGNALS PROM 23 CEA's ARE CommECTED TO EACHCPC.
ONLY 22 CP THE 23 SIGNALS ARE USED AS TARGET CEAs.
l l
COMME.%T d 9
. SRO EXAMINATION QUESTION 22 There is no correct answer. Actual allowable maximum les el is 57% per SO23 3 1.7. L&S 2.2. Southern California Edison believes i
e is no correct answer to this question.
Delete Question
}
f 4
h l
i l
f l-i I
NUCLEAR ORGANIZATION OPERATING INSTRUCTION 5023-3-1.7 UNITS 2 AND 3 REVISION 20 TCH 20-2 ATTACHMENT 16 PAGE 56 0F 56 1.0 REACTOR COOLANT PUMPS (Continued) 1.15 2 (3)'P-002 : For the ABB RCP Motors, the Lift 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.
Lt at a low pressure, And CB0 flow is < 0.25 gpm,.then one of the following is required:
CB0 line temperature (at the local flow indicator) is warm (i.e., cold line indicates no flow)
DE RCP Seal Cavity pressures are properly staged.
2.0 REACTOR COOLANT SYSTEM 2.1 1E Boron concentration in an idle loop is suspected of being lower than Reactor Core boron concentration, IHEN 00 NOT attempt to Start a RCP in that idle loop. This will prevent a possible reactivity 1
transient upon restart of forced flow.
(Ref. 2.1.5)
With'any RCS cold leg temperature < 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)
T,.. (S/G) <T, + 20
-9 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 I
for RCP operation.
2.4 If the RCS is solid, then RCS pressure may rapidly rise above the maxi um pressure for SDC loop operation due to heat transfer from the
?
S/G to the RCS.
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 i
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 oypass the associated SG Low Flow Trip before stopping an RCP will re: ult in a Reactor trip signal.
Or ATTACHMENT 16 PAGE 4 0F 4
l 1
COMMENT #6 SRO Examination Question 28 r
(RO27)
Answer "B" is correct because 656 deg F corresponds to 2300 psia. The pressurizer spray valves open at 2275 psia.
Ansur "D" is also correct because at 2225 psia and a backur Jgnal at 2275 psia, the heaters get a signal to turn
- off. Southern California Edison believes that tirre are two conect answers to this question.
Accept answers B & D j
i
)
l I
l I
l I
f
7 l
l NUCLEAR ORGANIZATION -
UNITS 2 AND 3 SYSTEM DESCRIPTION SD-SO23-360 I-REVISION 5 Page 168 of 205 i
l FIGURE 111-5 PRESSURIZER PRESSURF_ CONTROL SYSTEM BLOC
[
t PT 100X PT 100Y 1E HEATER PMS/CFMS PMS/CFMS -,1E HEATER
<,-. STEAM BYPASS SYSTEM STEAM BYPASS SYSTEM - i E/S RED 7 0 GREEN B/S h-HS-100A 1
2200/2225 RECORDER 2200/2225 IX
~
PRO 100-A/B Hl/LO B/S RED
%y* PMS foY3PM (2275/2175)
~
S E/S -
e ON/OFF 0
B/U HEATER c
- (2200/222d)
- a 8/S i
Hi H' B/U IiEATER c
- TRIP (2340) 3
=
u PIC 0100 Hg{gONA) l l
- DE-ENERGlZE R
~ B/UHEATERS (2275) b o
o o
o o V/4 E-
- EV/4 e MMRef'o" H
A I
$b]/L VALVC PopmON
%gpg mm MV 44 s.ggg u
u (2275/2175)
-# =[y&@
PV 100A
,, V 1008
,,,,rg
.cysv l
l J
i l
l COMMENT #7 i
l 1
l SRO Examination Question 30 l
The design basis for adding the steam generator delta p trip was based on steam line or feed line break (harsh l
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 shaA is not related to the loss of offsite power. This makes the answer A incormet.
Southern California Edison believes there are no correct answers to the question.
Delete this question.
1 1
l L
i 1
i.
i i
SOUTHERN CALIFORNIA EDISON PLANT PROTECTION SYSTEM l
NUCLEAR ENGINEERING, SAFETY AND LICENSING 0B0-5023 710. REV. 4 i
DESIGN BASES DOCUMENT PAGE 92 0F 569 A00RESSABLE CONSTANTS Symbol Definition Range BUFTRP Snapshot Buffer Control Flag 0 or 1 TCBP Maximum time that the RPC Flag can remain set 0 to 40.0 (seconds) d TCOUNT CRT Display Rewrite Control Flag 0 or 1 2.1.1.15 Low Reactor Coolant Flow (LRCF)
Each steam generator 2(3)E088 and 2(3)E089 has an RCS four channel measurement of differential pressure 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.2se.nas uro.zar Another suhtantive reason for adding this function was to provide better protection against the sheared shaft event, whose importance 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 providing 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 cogromised 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 on the CPCs, and could be developed, installed, and meet licensing schedules.""
Th'e PPS provides a channel trip when the RCS flow-produced differential pressure falls below the setpoint. A reactor trip then follows on a 2-out-of-4 basis. This trip function is presently l
credited to help mitigate the consequences of a sheared RCP shaft i
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 4
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
- =
l 1
COMMENT #8 SRO Examination Question 37 (RO37)
Procedure, 5023-12-7. Imss 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 can be used to correlate to Tc there fore answer B is also correct. Southern California Edison believes there are two correct answers to this questson.
l l
Accept answers A & B l
\\
l l
l l
1 l
1
)
I i
l i
1 I'
)
I i
l l
l.
I
l NUCLEAR CRGANIZATICN UNITS 2 AND 3 EMERGENCY CPERATING INSTRUCTICN REVISION 15 5023-12-7 ATTACHMENT 2 PAGE 30 CF 122
(:
LOSS OF FORCED CIRCULATIQN/ LOSS OF OFFSITE F0WER FLOATING STEPS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:
Low flow during Natural Circulation slows RCS response to temperature changes.
Leco transit time rises to between 5 minutes and 10 minutes. 1 Y
FS-2 MONITOR Natural Circulation Established:
CHECK all RCPs - stopped.
a.
GO TO FS-4, MONITOR RCP Operating a.
Limits.
b.
CHECK at least one S/G
- b. GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND
- 1) SBCS - operating OR INITIATE S023-12-9 Attachment 8, RECOVERY - 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, d.
CHECK Tc and TH - NOT rising.
THEN e.
CHECK Reactor vessel Level (Plenum) - greater than or MAXIMIZE S/G level - less tiian equal to 100%1 80% NR.
QSPOS page 622 RAISE available S/G steaming I
CFMS page 312 rate.
}
i RAISE Core Exit Saturation Margin f.
CHECK operating loop Tu and REP
- greater than 20'F:
CET - within 16*F:
QSPDS page 611 l
e CFMS page 311.
i
~.
m=
m-ATTACHMENT 2 PAGE 4 0F 29
NUCLEAR CRGANI~'TICN UNITS 2 AND 3 EMERGENCY CPERATING INSTRUCTICN REVISION 15 5023-12-7 ATTACHMENT 2 PAGE 31 0F 122 LOSS OF FORCED CIRCULATION / LOSS OF 0FFSITE PO FLOATING STEPS i
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 N0i o
satisfied, QSPDS 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:
9 l
\\
1 f
e.
C 1
> >h ATTACHMENT 2 PAGE 5 0F 29
,;:.e,*.v.,::.*;:::.9_] T n if: '[P5M9f.fQ9Qf}f'~'QTW;T,Ti?"TMTS?}WYS#'
T l
l COMMENT #9 SRO Examination Question 43 (RO46) 6 Answer B is correct based on the strictest interpretation ofimmediately before and after 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 i
feed 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 merride, 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 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.
l Accept answers B & C i
l l
l l
l I
t r
l t
L
{
l, l
t
. ~...
COMAIENT
- 10 i
SRO EXAMINATION QUESTION 51 -
l:
l l All four answers contain the statement *oser current reset" There is no over current relay in the circuitry for the 50.54X cross-tic for I
EDGs. Southern California Edison believes there is no correct answer for this question and the question should be deleted from I'
- examination.
Delete Question 9
Y W
l 1
t 6
f i
?
l i
i l
l r.
l J
l i
L 1
1
1 1
l 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.
l In attachment 3 of the same procedure, there is another caution that states: "Within 15 minutes of misalignment discovery, a power reduction snay be required..." laitial and final stabilized power levels are used to determine the l
further power reduction requirements within the first hour to maintain compliance with the acceptable operating
'i
- region in technical specificasson 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.
I Accept A & C i
l 1
l l
l l
i J
i 5
4 I
J NUCLEAR ORGAN!ZATION ABNORMAL 0PERATING 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 OBTAINER 1
3 COMMENCE plant load reduction:
a.
IF Reactor power is < 50%,
THEN GO TO Step 3c:
GallIlafi Within fifteen minutes of misalignment-discovery a power reduction may.be required. The negative reactivity of the misaligned CEA is considered 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 while maintaining Tcold within the Operating Band per.
r
....... ~.
t I
l CQMMENT # 12 l.
l SRO EXAMIN ATION QUESTION 65
th answers " A" & -B" will increase PT limits for operation of the reactor coolant pumps. Southern California Edison believes there two correct answers to this question.
Accept answers A & B l
b l
i P
1 L
l 6
h t
I i
l t
s 1
1 t
4 1
I i
l l-I
kUCL[AR ORGAN!ZAfl0N p[RG[qay Op[ PAT!4 IMTRUCTKN 5023-12-3 L4175 2 AND 3 8Ev!S104 15 PAGE 147 0F 163 ATTACHPENT 14 LC55 0F COOLANT ACCIDENT POST ACCIDENT PRE 55URE/TDtPERAT5RE LIMIT 5 2500 (2380 PSIA) MA OMUM OPERAT1 NAL PRESSUR
+
9 8
100*F/HR
/
- LOWEST SERVICE 200*F TEMPERATUPE SATURATION.
MARGIN TC
. (209'F) TC I
I g
l l
NOTE,1 l
l f
I 1500 i
i 0
l l
1 9
i I
i 5
i i
MARGIN TH, i
i I
i l
1000 I l
e i
i l
l 0*Fi SATURATION i
4 l
MARGIN Tg e
i e
l 6
s-l l
l i
i i
i i
1 i
l f
500 i
l l
8 i
340 '
291 SDC ENTRY CONDITIONS l
2%
0
- 00 200 300 400 500 60C 700 80 RCS TEMPERATURE (T)
NOTE 1: THIS CLRVE 15 IN EFFECT ANT TIME AN UNCONTROLLED C00LDOWN TO RC5 Tc LESS T O N 500*F MAS OCCURRED.
$3 *5.w61 ATTACWENT 14 PAGE 1 0F 1
i i
l
~ COMMENT #13 i
l SRO Exanunation Question 69 (RO74) l The question setup has the VCT pressure at 40 psig. The head of the Refueling Water Storage tank, RWST, is 70' i
t 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 I
keep the check valves from the RWST closed. Southern California Edison believes there are no correct answers to l
this queshon and it should be deleted.
l I
Delete this question.
i I
I i
l I
l t
1 f
l 1
l l
1
f I
COMMENT #14 SRO Ev==ination Queshon 74 (RO 78)
All pressunzer 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 will exceed the 2275 psia shutoff set point for the heaters in auto. The correct answer should be D.
Change correct answer to D.
1 r
t h
l r
I 1
l I
1 a
j
NUCLEAR ORGANIZATION SYSTEM DESCRIPTION SD-SO23-360 UNITS 2 AND 3 REVISION 5 Page 168 of 205 FIGURE lil-5 PRESSURIZER PRESSURE CONTROL SYSTEM BLOCK DIAGRAM l.
PT PT 100X 100 1E HEATER PMS/CFMS PMS/CFMS -,1E HEATER i
,~ STEAM BYPASS SYSTEM STEAM BYPASS SYSTEM - i B/S RED g500- GREEN i
B/S ifN-HS-100A 2200/2225 1
Hl/LO 8/S E
kx y PMS foY34RM O
(2275/2175) c B/S -
S ON/OFF
>--* 8/U HEATER (2200/2225) e B/S i
_ HI HI
~
P( 3 )
g c
2 o
PIC 0100
'RgypglONA1.
l l
DE-ENERGIZE q
~
B/U HEATERS (2275) ll 1I 1f lI 1I 10 1 OB
/
h hhhAO l
~
"~*
l Hl/LORM g VALVg POSmON mm h014 P?^
"'IAbTIN o
o (2275/2175)
-3 ggg%~"
PV 100A
,V 1006 g.
10, r
g.c3o oy.n 1
- ~~
. - -. - -.... ~. -
i COMMENT # 15 -
SR0 EXAMINATION QUESTION L#75 l
(R0 79)-
Question was based on old Tavg program. New program has normal pressurizer level of 48% at 100% power. This is based on the 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 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. -
e t
P
NUCLEAR ORGANIZAT80N OPERATING INSTRUCTION REVISION 8 h0 UNITS 2 AND 3 34 ATTACHMENT 5 PRESSURIZER LEVEL CONTROL PROGRAM 70 y.........
_.+........
...>...).
.t...}.
4
.s.
.g.
.i.
.q.
.{.
.>. 9.;.
.g.,
.J...*....'.....
g 8
......4..
- 4....;.. 3...
w
.J
...;...L...L........
UJ
.y...y...)..
3...g.
i
...(*.<...<...
m
.>...*...h..f..i..
..j.
.J
....g v..
E E
1 N
...........:53 C
o (f)
(f)
.p.....
...z...g........
y W
g
.. j....:..
\\
...i...;.
4..
4...
...<.. 4...
...1
....g........k....>...',..
..;...;.......{..<.
4........
... s....;.... 3.. 3...3 30
. i r..........
,.. i i...
550 560' 570 14 58Q 590 RCS AVERAGE TEMPE A(ATURE ( F) 544 D 10 -1.C H T I
w.
D10 8.wS1 ATTACHMENT 5 PAGE 1 0F 1 4
l L
COMMENT # 16 SRO Frammahan Queshon 77 DSS is not covered by Tochacal Specifications. There is no correct answer to the question as stated.
Delete the question l
L-
l I
COMMENT #17 SRO EXAMINATION QUESTION 84 (RO 85) i I'
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 i
trend, the listed action of *D" could be taken to return the plant to a more stable condition. Southern California Ediam believes "B" & "D" are correct answers.
Accept B & D.
i l
I i
l I
l i
l I
i l
l l
!~
i l
r l
o l
l l('
5 I
i
NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTTON S023-13-10 UNITS 2 AND 3 REVISION 2 PAGE 7 0F 13 LOSS OF CONDENSER VACUUM
' ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 4 Actions for loss of vacuum due to Condenser fouling:
CAUTION During periods of heavy influx, rapid and aggressive action may be required in (f.
order to avoid a Unit trip. Power may need to be reduced in orcer to:
Bump and/or Stop Circulating Water Pumps on the Condenser que.drants 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 S023-2-5, 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 stacle.
- 5) GO TO Step 5.
- d. EVALUATE stopping pumps based on Waterbox differential pressure and pump vibration.
- e. GO TO Step 6.
s=
k
l l
I COMMENT # 18 SRO Examination Quesuon 93
' (RO93)
Answer C is correct based on HV9217 and HV9218 being open and presiding a direct path from inside l
containment to the outside, in this case to the VCT. Answer D is correct because given this event, Controlled Bleed O\\ffflow 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.
j Therefore, accept both answers C and D l
Accept answers C & D i
l i
COMMENT-#19
.fg-i SRO EXAMINATION QUESTION #76-
+
(R0 95) l
~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 l
+-
stem (step FS-6 a.1 requires operating S/G with an ADV operating).
I and continuing to inject water into the RCS will increase i
pressure also leading to PTS event. S6uthern California Edison believes.there are two correct answers to this question.
Accept answers A & B I
e i
c e
i
- - =
NUCLEAR ORGANIZATION EMERGENCY OPERATING INSTRUCTICN S023-12-5
. UNITS.2 AND 3 REVISION 15 PAGE 45 0F 143 ATTACHMENT 2
(
EXCESS STEAM DEMAND EVENT FLOATING STEPS
~ ACTIONS / EXPECTED RESPONSE RESPONSE NOT OBTAINED j
FS-6 CHECK HPS?. 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 Termination Criteria.
j.. CHECK PZR Level
- j. INITIATE FS-22, ESTABLISH CVCS i
- less than 80%.
Letdown Flow.
t k.
RESET SIAS per 5023-3-2.22, ESFAS OPERATION.
i i
i i
l.
ATTACHPENT 2 PAGE 15 0F 43 i
i
1
%UCLEAR CRGANIZATICN EMERGENCY CPERATING INSTRUCTION 5023-12-5 UNITS 2 AND 3-REVISION 15 PAGE : OF 123 x-ATTACHMENT 2 EXCESS STEAM DEMAND EVENT-FLOATING STEPS i-ACTIONS / EXPECTED RESPONSE RESPONSE NOT CBTAINED
.FS-6 CHECK HPSI Throttle /Stop Criteria:
(Continued) d.
CHECK Reactor Vessel Level e
IF any criteria of ste:s o tnrcugh c (Plenum) - greater than or NOT met, equal to 100%:
j i
THEN QSPDS page 622 CFMS page 312 OPERATE Charging and HPSI systems as necessary to maintain
{
Throttle / Step criteria
- satisfied.
j THROTTLE Loop Injection Valves.
ENSURE auxiliaries to SI pumps:
i i
a)
Electrical power to pumps and i
valves.
b)
Proper system alignment.
c)
CCW flow.
d)
HVAC.
l e.
VERIFY RCS borated - greater
- e. MAINTAIN Emergency Boration than Technical Specification
- at least 40 GPM.
Shutdown Margin for T vE > 200*F A
per Operations Physics Summary Figure 2.3-1,
~~
OR RCS Cooldown - NOT in progress.
.g.
STOP charging pumps as required i
one at a time.
L i
c=-
~
l ATTACHMENT 2 PAGE 14 0F 43 y
NUCLEAR CRGANIZATION EMERGENCY OPERATING INSTRUCTION 5023-12-5 UNITS 2 AND 3 REVISION 15 PAGE 43 0F 143 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT FLOATING STEPS ACTIONS / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-6 CHECK HTSI Throttle /Stop Criteria:
a.
CHECKatleastoneS/G a.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1)
SBCS - operating INITIATE 5023-12-9, Attachment 8, OR RECOVERY - HEAT REMOVAL.
ADV - operating.
AND 2)
Feedwatfr - 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
{
Throttle /Stop criteria c.
CHECK Core Exit Saturation
- satisfied.
Margin - greater than 20*F:
THROTTLE Loop Injection Valves.
ENSURE auxiliaries to SI pumps:
1 a)
Electrical power to pumps and valves.
b)
Proper system alignment, c) CCW flow.
d) HVAC.
a.
ATTACHMENT 2 PAGE 13 0F 43 J
I COMMENT # 20 SRO EXAMINATION QUESTION 97 (RO96)
Answer "B" is correct based on the information given. However answer "A" is also correct based on procedure 5023-12-7, 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 j
this question.
Accept answers A & B.
1 k
NUCLEAR CRGANIZATION EMERGENCY OPERATI*d INSTRUCTICN 5023-12-7
/
UNITS 2 AND 3 REVlSION 15 ATTACHMENT 2 PAGE 30 CF 12'9 LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE PCWER FLOATING STEPS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:
Low flow during Natural Circulation slcws RCS response to temperature changes.
Loop transi.
time rises to between 5 minutes and 10 minutes.
FS-2 MONITOR Natural Circulation Established:
a.
CHECK all RCPs - stopped.
- a. GO TO FS-4, MONITOR RCP C.perating Limi ts.
b.
CHECKatleastoneS/G
- b. GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1)
SBCS - operating INITIATE 5023-12-9, Attachment 8 OR RECOVERY - HEAT REMOVAL.
ADV - operating.
AND 2)
Feedwater - available.
c.
CHECK operating loop AT'- less o
IF any criteria c through g NOT than 58'F.
satisfied, d.
CHECK Tc and Ta - NOT rising.
THEN e.
CHECK Reactor Vessel Level MAXIMIZE S/G 1evel - less than (Plenum) - greater than or 80% NR.
equal to 100%:
RAISE available S/G steaming QSPDS page 622 rate.
CFMS page 312.
RAISE Core Exit Saturation Margin
- greater than 20'F:
f.
CHECK operating loop Tw and REP CET - within 16*F:
i-QSPOS page 611 l
CFMS page 311.
l
~
7 s=
ATTACHMENT 2 PAGE 4 0F 29 E
)
i'
' NUCLEAR ORGAN 12ATION UNfTS 2 AND 3 EMERGENCY CPERATING INSTRUCTf0N 5023-12-7 REV!SION 15 ATTACHMENT 2 PAGE 31 0F 12'7 i
LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWER FLOATING STEPS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-2 MONITOR Natural Circulation l
Established:
(Continued)
/
g.
CHECK Core Exit Saturation IF any criteria c through g NOT o
Margin - greater than 20*F:
satisfied, OSPDS page 611 Tags CFMS page 311.
MAXIMIZE S/G level - less than i
'80% NR.
RAISE available S/G steaming rate.
RAISE Core Exit Saturation Margin
- greater than 20*F:
l l
ATTACHMENT-2 PAGE 5 0F 29 4