IR 05000361/1998302
| ML20198S797 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/04/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20198S790 | List: |
| References | |
| 50-361-98-302, 50-362-98-302, NUDOCS 9901120007 | |
| Download: ML20198S797 (107) | |
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ENCLOSURE
kl.S. NUCLEAR REGULATORY COMMISSION
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REGION IV '
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- Docket Nos.:
.50 361,50 362.
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L'icense Nos.:)
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- Report No.:
- 50-361/98-302, 50-362/98-302 Licensee:
Southern California Edison Co.
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Facility:.
San Onofre Nuclear Generating Station, Units 2 and 3
< Location:
5000 S. Pacific Coast Hwy.
San Clemente, Califomia.
- Dates
November 15 through December 3,1998 Inspectors:
S. L. McCrory, Senior Reactor Engineer, Examiner / Inspector, Chief Examiner T. O. McKernon, Senior Reactor Engineer, Examiner / Inspector'
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M. E. Murphy, Senior Reactor Engineer, Examiner / Inspector -
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T. R.' Meadows, Senior Reactor Engineer, Examiner / Inspector'
R. E Lantz, Reactor Engineer, Examiner / Inspector s Approved By:
J. L. Pellet, Chief, Operations Branch
- Division of Reactor Safety -
' ATTACHMENTS:
Attachment 1:
SupplementalInformation
~ Attachment 2:
Final Written Examinations and Answer Keys Attachment 3:
Post Examination Comments l
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9901120007 990104 L
PDR - ADOCK 05000361 e
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EXECUTIVE SUMMARY San Onofre Nuclear Generating Station, Units 2 and 3 NRC Inspection Report 50 361/98-302; 50-362/98-302 NRC examiners evaluated the competency of nine senior operator applicants and five reactor operator applicants for issuance of operating licenses at the San Onofre Nuclear Generating Station facility. The licensee developed the initiallicense examinations using NUREG-1021,
" Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The NRC examiners administered the operating tests on November 1519,1998. The facility licensee administered the initial written examinations to all applicants on November 20,1998.
Operations
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The large number of questions missed and the high number of common error responses by most applicants indicated training weaknesses. This conclusion was further supported by performance weaknesses observed during the operating examination (Sections 04.1 and 04.2).
The facility licensee devele> ped an adequate written and operating examination.
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However, the post-examination review of the written examination identified a large number of technicalinaccuracies. The large number of technicalinaccuracies indicated a significant weakness in the facility licensee's initial technical review (Section O5.1.2).
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An inadvertent breach of examination security did not result in an examination
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i-compromise (Section 05.3).
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Reoort Details Summary of Plant Status l
The units operated at essentially 100 percent power for the duration of this inspection.
I. Operations
Operator Knowledge and Performance 04.1 Initial Written Examination a.
Insoection Scoce -
On November 20,1998, the facility licensee proctored the administration of the written examination to nine senior operator license applicants and five reactor operator license
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applicants. The facility licensee provided post-examination comments (Attachment 3)
following the administration of the written examination. The chief examiner reviewed the comments for technical adequacy. The chief examiner reviewed the written examination grading on December 2,1998.
b.
Observations and Findinos Three of five reactor operators and six of eight senior reactor opt. tor applicants passed the written examination. The written examination was waived for one senior reactor operator applicant who had passed the written examination on a prior licensing examination. Reactor operator applicant scores ranged from 72.6 to 85.3 percent with an average of 78.7 percent. Senior reactor operator applicant scores ranged from 63.8 to 85.1 percent with an average of 79.8 percent. The overall written examination average was 79.4 percent.
- The following questions were missed by at least one half of the applicants. Questions common to both examinations are shown with the number from the reactor operator examination first.
Common questions: 1/1,6/7,11/13, 14/18*,16/21*,24/27*,28/29*,42/39*,58/52*,
65/58*, 78/74*, 81/79*, 85/84*, 86/85*, 94/94*, 98/99*
Reactor Operator only: 45*,49*,57,63*,64,88*,
Senior Operator only: 22*,32*,57*,78*,87 Most applicants gave the same incorrect answers to the above questions marked with an asterisk (*) plus common question 59/53, and senior operator question 15. The knowledge deficiencies fell roughly equally into two broad categories - systems and l
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procedures. Of the system knowledge based errors, about two thirds related to logic or control circuit performance. During the pre-examination review, the chief examiner expressed concern to the facility licensee about the number of control logic questions
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and whether references should be provided for some of them. Licensee staff responded that the tested areas were required knowledge.
c.
Conclusions The large number of questions missed and the high number of common error responses by most applicants indicated training weaknesses.
O4.2 Initial Ooeratina Test a.
Inspection Scoce The examination team administered the various portions of the operating test to the 14 applicants on November 15-19,1998. Each applicant participated in at least two dynamic simulator scenarios and received a walkthrough test, which consisted of ten system tasks together with followup questions for each system. Additionally, each applicant was tasted on five subjects in four administrative areas with a combination of administrative tasks and questions.
b.
Observations and Findinas All applicants passed the operating examination.
The examiners observed consistently good three-way communications and supervision of control panel activities during the dynamic simulator and dynamic walkthrough portions of the operating test.
During Simulator Scenario 2, simultaneous steam generator tube rupture and failed open steam generator safety valve malfunctions occurred on the same steam generator.
In one crew, no applicants observed the abnormal cooldown caused by the failed open safety valve and, therefore, did not diagnose and respond to the bomonitored radioactive release. During the same scenario, only one of five crews communicated to management or support personnel any precautions or concerns regarding the radiological conditions impacting recovery efforts.
There were three instances in which applicants read or operated the wrong radiation monitors in response to system tasks or scenario events. The nature of the errors was similar, and the examiners concluded that instrument label placement contributed to the errors. The instrument labels were positioned below the instruments for a small number of radiation monitors. Virtually all other instruments and controls in the control room had the labels positioned above the instrument.
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Conclusions All applicants passed the operating examinations but exhibited some knowledge and
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ability weaknesses. This performance further supported that training weaknesses existed.
Operator Training and Qualification 05.1 Initial Licensina Examination Develooment The licensee developed the initial licensing examination in accordance with guidance provided in NUREG-1021," Operator Licensing Examination Standards for Power Reactors," Interim Revision 8, and additional guidance provided by the chief examiner.
05.1.1 Examination Outline The facility licensee submitted the initial examination outline on September 2,1998.
The chief examiner reviewed the submittal against the requirements of NUREG-1021, Interim Revision 8. The examination outlines satisfied the requirements of the examination standards with regard to breadth, depth, and scope.
O5.1.2 Examination Packaae a.
Inspection Scope The facility licensee submitted the completed draft examination package by October 5,1998. The chief examiner and peer reviewers reviewed the formal submittal against the requirements of NUREG-1021, interim Revision 8. An onsite validation of the operating examination was conducted during the period November 4-6,1998.
b.
Observations and Findinas The reviewer directed that 18 of 125 written examination questions be revised or replaced as a result of being assessed as discriminating at too high or too low a level.
The reviewer provided enhancement comments on an additional 25 questions. The reviewer commented on several questions related to control systems logic as possibly being too difficult to answer without a reference; however, the reviewer left the decision with the facility licensee to propose the use of specific references.
Approximately 50 percent of the prescripted questions developed for Parts A and B of the operating test had to be revised or replaced for various deficiencies including low discrimination, direct look-up, and wrong focus. Overall, the walkthrough portion was assessed as marginally adequate because there was at least one acceptable prescripted question per task.
The reviewer identified two system tasks in one of the walkthrough test that tested the same operator ability and directed that one be replaced. Both tasks required the operator to parallel electrical generating sources (one for the main turbine generator and one for an emergency diesel generator). During the onsite validation of the operating examination, the chief examiner identified that two of the simulator malfunctions were
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also included as system tasks in the walkthrough part of the operating examination and directed that the scenario malfunctions be replaced. Apart from this minor task duplication, the reviewer determined that the simulator scenarios were of good quality.
The facility licensee provided a total of 22 post-examination comments (see Attachment 3) on the written examination recommending question deletion and acceptance of additional answers. Nearly all of the comments addressed technical inaccuracies. The chief examiner accepted all the facility licensee post-examination comments except the following:
Senior Operator Comment 5 - The facility licensee recommended deleting the
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question on the basis that the allowed maximum value was a pressurizer level of 57 percent, which was not one of the choices. The chief examiner rejected this recommendation because the reference cited required that pressurizer level
must be less than 900 ft, which equated to 57 percent. Therefore, Choice C i
(53 percent) remained as the only correct answer since it was the highest value
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below 57 percent.
Senior / Reactor Operator Comment 16/11 - The facility licensee recommended
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accepting Choice B as an additional correct answer based on the possible assumption that condenser backpressure did not continue to increase above the last reported value. The chief examiner rejected this recommendation based on l
the procedural requirement to take action if the condenser backpressure increased to equal to or greater than 3.5 inches of mercury. Therefore, taking no action until condenser backpressure increased to 4.5 inches of mercury (Choice B) was not acceptable.
c.
Conclusions The facility licensee developed an adequate written and operating examination that had job performance measure prescripted questions of marginal quality. However, the post-examination review of the written examination identified a large number of technical inaccuracies. The large number of technicalinaccuracies indicated a significant weakness in the facility licensee's initial technical review.
O5.2 Simulation Facility Performance The examiners observed simulator performance with regard to fidelity during the examination validation and administration. The simulation facility supported the examination administration well. The examiners observed no problems.
05.3 Examination Security During examination administration, the examination material was maintained in a locked room to which only the examiners and limited members of the training staff, in the security agreement, had keys.
On Tuesday, November 17,1998, between 5:30 a.m. and 6 a.m., a site security guard opened the examination material room with a master key to permit the cleaning staff to remove trash. The cleaning staff left the door to the room slightly ajar, but not open
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enough to permit observing the contents of the room. The examiners arrived at about
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6:30 a.m. and found the door ajar. The examination material did not appear to have been disturbed. Only one license applicant had arrived onsite by that time. The applicant was interviewed, and he stated that he had gone directly to the applicant's sequestering room upon arrival. Members of the training staff had noted the applicant's arrival and had not seen him anywhere 'near the examination material room. The security guard and cleaning individual were added to the security agreement. There were no discernable improvement in the performance of any applicant, nor other indication of any applicant having obtained knowledge of the examination content following the incident. The chief examiner determined that examination material security had been inadvertently breached but that no examination compromise had occurred.
V. Management Meetings
- X1 Exit Meeting Summary The examiners presented partial inspection results to members of the licensee management at the conclusion of the onsite inspection on November 19,1998. After the graoing of the written examinations and analysis of the results, the chief examiner held a final exit with the licensee telephonically on December 18,1998. The licensee j
acknowledged the findings presented.
The licensee did not identify as proprietary any information or materials examined during the inspection.
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ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED
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Licensee.
- M. Jones, Manager, Operations R. Sandstrom, Manager, Training K. Rauch, Supervisor, Operations Training T. Frey, Compliance T.Vogt Operations
. D. Axline, Licensing L. Germann, Training i
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e ATTACHMENT 2 FACILITY LICENSEE POST-EXAMINATION COMMENTS
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RO Exam Comments
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COMMENT #1
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RO Examination Q esuon 7 (SRO9)
The question stem referenas SO23-3 3.27,3 as do the possible answers. The actual procedure that should have
been referenced is SO23-3-3.23, Emergency Diesel Generator Monthly Surveillance. The given procedure, SO23-
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3-3.27.3 is not used to perform the AC Sources check that is required by the given scenario. The correct answer
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was not prended. Southern California Edison believes there are no correct answers to this question.
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- Delete the question.
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NUCLEAR ORGANIZATION SURVEILLANCE OPERATING INSTRUCTION 5023-3-3.23
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UNITS 2 AND 3 REVISION 14 TCN 14-2 PAGE 72 0F 88
ATTACHMENT 7 A. C SOURCES VERIFICATION (MODES 141
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OBJECTIVE To provide verification that sufficient AC Sources are available to the IE 4.16kV Busses when any combination of Offsite Circuits, Onsite Circuits, and Diesel Generators are Inoperable. This attachment satisfies Surveillance requirement of Tech. Spec.
LC0 3.8.1 AC Sources Verification.
UNIT MODE (1-4)
DATE TIME PERF. SY 1.0 PREREOUISITES INITIALS 1.1 Verify this document is current by checking a controlled copy or by using the method described in 50123-VI-0.9.
1.2 List the reason for performing this attachment (e.g., Diesel Generator 2G002 Inoperability).
I 2.0 AC SOURCES VERIFICATION 2.1 If this attachment is being performed prior to declaring a piece of equipment Inoperable, then assume the equipment is Inoperable when performing the attachment.
2.2 If the specific equipment Inoperability has placed both Units in action statements, then a separate attachment will have to be performed for each Unit.
2.3 If a Diesel is Inoperable, then determine if the cause of the Diesel Generator Inoperability may exist on the other Diesel Generator (s).
2.351 If the cause of the Diesel Generator Inoperability exists on the other Ofesel Generator (s),thendeclaretheaffected Diesel (s) Inoperable, j
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2.4 If desired use the last page of this Attachment to assist in performance of this Attachment.
ATTACHMENT 7 PAGE 1 0F 7
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.o NUCLEAR ORGANf2ATION SURVEILLANCE OPERATING INSTRUCTION S023-3-3.27.2
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UNITS 2 AND 3 REVISION 10 PAGE 4 0F 26
--ATTACHMENT 1-
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WEEKLY ELECTRICAL BUS SURVEILLANCE - Both Units in Modes 1 thru 4 OBJECTIVE To verify Operability of the offsite transmission network, onsite Class 1E distribution system (except the diesel generators), and the onsite DC systems as required by-the Technical Specification Surveillance requirements:
SR 3.8.1.1, SR 3.8.4.1, SR 3.8.7.1, SR 3.8.9.1.
To verify the functionality of the Spent Fuel Pool Cooling System power availability as required by the Administrative Technical Specification.
UNIT 2 MODE UNIT 3 MCDE DATE PERF. BY 1.0 PREREOUISITES INITIALS
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.1.1 VERIFY this document is current by checking a controlled copy or by using the method described in 50123-VI-0.9.
1.2 DETERMINE the performance requirements of this attachment, as follows:
SRO Ops.
O This Attachment is being performed for a scheduled surveillance.
O This Attachment is being performed for operability verification. LIS) the Components and Sections Steps lR to be performed. After approval, then CIRCLE N A for the remaining unused steps.
COMPONENTS
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SECTIONS / STEPS
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I t-OPERABILITY VERIFICATION PREPARED BY:
Control Room Operator OPERABILITY VERIFICATION APPROVED BY:
SR0 Ops. Supv.
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C ATTACHMENT 1 PAGE 1 0F 7 i
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i 1R0 EXAMINATION QUESTION #9'
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S023-5-1.8 is thelreference for "A" to be a correct answer.-
"C" is i
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Jalso correct based'on. Technical-Specification 3.4.6 and 3.4.7, which-
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believes" there-are two correct answers to this question, i
iAccept answers A & C:
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NUCLEAR ORGANIZATION INTEGRATED OPERATING INSTRUCTION S023-5-1.8
UNITS 2 AND 3 REVISION 9 PAGE 86 0F 91
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ATTACHMENT 13 i
9.o kcP OPERATION 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 2125'F to prevent entering the restrictive heatup and cooldown limitations that apply when s120*F.
j 9.3 When securing RCPs, it may be necessary to reduce PZR heater output due to the reduction of PZR Spray Valve bypass flow.
9.4 Due to insufficient Pretsurizer heater capacity, it may be necessary to
secure all RCPs and main spray prior to initiating Auxiliary Spray.
Otherwise, loss of NPSH for the RCPs could occur.
(Ref. 2.3.17)
9.5 Pressurizer insurge reay occur when securing the last RCP. This is caused due to the lower RCS flow across the core. As Core Exit Temperature rises, the RCS will swell into the Pressurizer. Adjusting letdown flow will help minimize this insurge.
9.6 Indicated Tcold will initially rapidly lower in any loop where 500 is injecting, if the RCP operating in that loop is stopped or when the last RCP ts stopped. This is due to cooler SDCS injection water flowing over the loop Tcold temperature element.
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 Tcold temperature monitoring.
9.9 I.E RCPs are running, IllE!( one RCP shall remain in service until completing RCS boration to Mode 5, or refueling concentration and other o
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forced circulation dependent parameters are met (e.g., hydrogen,
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peroxide,etc.).
9.10 When the last RCP is stopped, then RCS Thot indications (T351Y and CETs) will begin to rise due to the increased time coolant is in the Core region (i.e., no RCP forced circulation). Consequently, SDCS flowrate should be adjusted to maintain RCS Tcold TR-0351A (T351X) at the desired teniperature.
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ATTACHMENT 13 PAGE 6 0F 11
T0 *d CP:li 86. Of ^0N 9122-891-6v6:xe3 gd wito seg n sgos l
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RCS Loops--MODE 4
3.4.6
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3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.6 RCS Loops'--MODE 4 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---------------------------
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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:
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No operations are permitted that would cause reduction of the RCS boron concentration; and b.
Core outlet temperature is maintained at least 10*F below saturation temperature.
2.
No RCP shall be started with any RCS cold leg temperature 5 256*F unless:
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Pressurizer water volume is < 900 ft, or b.
Secondary side water temperature in each steam generator (SG) is < 100'F above each of the RCS cold
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leg temperatures.
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APPLICABILITY:
MODE 4.
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SAN ONOFRE--UNIT 2 3.4-18 Amendment No. 127
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RCS Loops--MODE 4
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ACTIONS e
CONDI' TION REQUIRED ACTION COMPLETION TIME A.
One required RCS loop A.1 Initiate action to Immediately inoperable.
restore a second loop or train to OPERABLE AND status.
Two SDC trains inoperable.
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B.
One required SDC train B.1 Be in MODE 5.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable.
AND Two required RCS loops inoperable.
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Required RCS loop (s)
C.1 Suspend all Immediately
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or SDC train (s)
operations involving inoperable.
reduction r~ e'.S boron conce ; ration.
QB AND
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No RCS loop or SDC train in operation.
C.2 Initiate action to Immediately restore one loop or train to OPERABLE status and operation.
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SAN ONOFRE -UNIT 2 3.4-19 Amendment No. 127 t
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RCS Loops-Mn0E 4
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' SURVEILLANCE REQUIREMENTS I
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SURVEILLANCE 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.
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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).
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SR 3.4.6.3 Verify the second required RCS Loop or SDC 7 days train is OPERABLE.
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SAN ONOFRE--UIIIT 2 3.4-20 Amendment No. 127
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RCS Loops--MODE 5. Loops Filled 3.4.7
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3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.7 RCSLoops3 MODE 5, Loops Filled
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LCO 3.4.7 At least one of the following loop (s)/ trains listed below shall be OPERABLE and in operation:
Reactor Coolant Loop 1 and its associated steam i
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generator and at leas'. one associated Reactor Coolant
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Pump; b.
Reactor Coolant loop 2 and its associated steam generator and at least one associated Reactor Coolant Pump;
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c.
Shutdown Cooling Train A; or d.
Shutdown Cooling Train B One additional Reactor Coolant Loop / shutdown cooling train
'shall be OPERABLE, or The secondary side water level of each steam generator shall be greater than 50% (wide range).
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NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and pumps providing
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shutdown cooling may be de-energized for 51 hour5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
i a.
No operations are permitted that' would cause reduction of the RCS boron concentration; and b.
Core outlet temperature is maintained a' least 10*F t
below saturation temperature.
2.
One required SDC train may be inoperable for up to
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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)
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SANONOFRE-bHIT2 3.4-21 Amendment No. 127
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RCS Loops--MODE 5,. Loops Filled
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3.4.7 i
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NOTES (continued)---------------------
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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 generator (SG) is < 100*F' above each of the RCS cold leg temperatures.
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- A containment spray pump may be used in place:of a icw pressure safety injection pump in either or both.
shutdown cooling trains to provide shutdown cooling flow
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provided the reactor has been suberitical for a period
> 24. hours and the RCS is fully depressurized and vented
in accordance with LCO 3.4.12.1.
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All SDC trains may be removed from operation during
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planned heatup to MODE 4 vnen at least one RCS loop is
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in operation.
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APPLICABILITY:
MODE 5 with RCS loops filled.
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ACTIONS
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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
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to OPERABLE status.
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Any.SG with secondary side water level not.
A.2 Initiate actior to Immediately within limit.
restore SG secondary side water levels to within limits.
(continued)
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i SAN ONOFRE--UNIT 2'
3.4-22 Amendment No. 127
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_.. _ _ _ _... _ _ _. _.. _.. _.. - - _ _ _ _ _... _ _. _. _ _. _.. _. _ _ _ _ _.
..
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'RCS Loops--MODE 5, Loops Filled
-
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3.4.7 i
_ _,
"}
. ACTIONS (continued)
~
CONDITION REQUIRED ACTION COMPLETION TIME
,
.
,
B.
No SDC train /RCS loop.
B.1 Suspend all Immediately
'
in operation.
operations involving
"
,
reduction in RCS boron concentration.
.
'
.
AND B.2 Initiate action to Immediately restore required SDC
-
train /RCS loop to
-
operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
)
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).
!
i SR 3.4.7.3 Verify the second required RCS loop, SDC 7 days tFain or steam generator secondary is
'
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p.
SAN ON0FRE--UNIT 2 3.4-23 Amendment No. 127
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COMMENT #3
' RO Examination Mr.14 (SROI8)
De generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. The Reed
.,
Switch Position Transenitters, RSPT's, actually sense the CEA's position. The Control Element Assembly Calculator, CEAC, uses the input Dom the RSFT and sends a signal to the alarm. Both components are =ri~l to generate a deviation alarm. Southern California Edison believes there are two correct answers to this question.
Accept anu a B & C l
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NUCLEAR ORGAtl!ZATION ALARM RESPONSE INSTRUCTION 5023-15-50.Al
.,
UNZTS 2 AND 3 REVIS!0N 2 PAGE 710F 76
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ATTACHMENT 2
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50A28.CEA DE.VIATION APPLICABILITY PRIORITY REFLASH ASSOCIATED WINDOWS Modes 1-3 AMBER NO NONE IhlTIATING NOUN NAME SETPOINT VALIDATION PMS 10 LINK #
DEVICE INSTRUMENT U2/U3 2(3)LO91,- CEAC 1 Control Element 5 Inches NONE DEVIAR56 641/663 j!L or CEAC 2 Assembly Deviation 1.0 REOUIRED ACTIONS:
1.1 Position the CEOMCS Mode Selector Switch on 2(3)CR50 to 0FF.
1.2 Verify which CEA is misaligned and the amount of misalignment, by observatipnofthefollowing:
CEAC display CRT
CEAC remote operators modules
e PMS alarms e PMS readout 2.0 CORRECTIVE ACTIONS:
SPECIFIC CAUSES SPECIFIC CORRECTIVE ACTIONS 2.1 Misaligned CEA 2.1 After the misaligned CEA has been determined, lhmt:
2.1.1 Notify the SR0 Ops. Supv.
2.1.2 Realign the CEA per 5023-3-2.19, Section for Manual individual R
Operation.
2.2 Slipped or Dropped CEA 2.2 GO TO S023-13-13, Misaligned t.ontrol Element Assembly.
3.0 ASSOCIATED RESPONSES:.
3.1 NotifytheCRS/SSandtheSTAtoreviewTech. Specs.LCO3.1.5and LCS3.1.105,andinitiateanEDMR/LC0AR,asrequired.
~g.,
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NttCLEAR ORGANIZATION
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SYSTEM DESCRIPTION $0.$023 710 UNITS 2 AND 3 REVIS!CN 3 PAGE 72 0F 76
.._
fl$.iL8 E !! - CONTROL ELEMENT ASSEuBLv sultGm00R REED TWITM P95fTf0N TRAN!NITTER TIGNAL AtsfGNWENTS
.
l h EX4CRE CHANNEL h
RSP7 at SPT \\
'VA 23CEAS h
/ RSP7 22CEAS
_
i V
/
g 22 CEAS CEAS
"f 45 CE,a5 g
d$ C(A$
_MQEh 22 CE.
'
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23 Co.1
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ISCLATION
_
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/
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CALCULATOR CALCULATOR
\\
CEA POSIT:Ch NO 1 MO. 2 CD #CSITCN N[ikif,k ctUatc=
a=S{
] a=EwAfCN=
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A COR E 8 CORE CCCRE OCORE PROTECT CN PROTECTION PROTECTCh PROTECT:ON
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CALCULATCR CALCut/ TOR CALCULATOR CALCULATCR I
I
c---
.
$
I OPERATOR $
OPERATCR'S OPERATOR 4 CPERATCR'S MODULE MOCULE MOCULE WOOULE CRTCISPLAY NOTES.1. SIGNAL FRCM CEA 213 CCNPECTED TO CPC's A AND C, BUT IT 13 NOT USED AS A TARGET CEA.
2. SIGNAL l'MCM CEA 313 CONNECTED TO CPC's 8 AND C,8tf7 ITIS NOT USE3 A5 A TUtGET CEA.
3 SIGNALS FRCM 23 CENs ARE CONNECTED TO EACH CPC.
-
CNLY 22 CF THE 23 SICNALS ARE USEQ AS TARGET CEAs.
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COMMENT # 4
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- -'
- RO EXAMINATION QUESTION 26.
i Electncal drawing 30718 Rev. 9 shows the automatic makeup circuitry has been deleted. Southern California Edison believes there
. are no correct answers for this quesuon and the question should be deleted from the examination.
.
Delete'4ueshon.
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' COMMENT #5-
'
c-R0 EXAMINATION QUESTION'#27
-
-(SR0 28)
'
i-Answer "B" is correct because-pressurizer spray valves are open at-l
-
- 2300 psia.. Answer "D" is also correct because the backup heaters turn i
off.at 2225 psia and a backup signal to turn.off the backup heaters at 2275 psia._' Southern California Edison. believes there are two correct l
answers _to this question.'
)
.
L
Accept answers _B 8 0.:
,
a
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d
4 s-
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NUCLEAR ORGANIZATION UNITS 2 AND 3:
SYSTEM DESCRIPTION SD.SO23-360
.
REVISION S Page 168 of 205
,
q FIGURE 111-5 PRESSURIZER PRESSURE CONTROL SYSTEM EILdCK
.s
,
.
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~ PMS/CFMS PMS/CFMS -,1E HEATER
<
"
~ STEAM EYPASS SYSTEM STEAM EYPASS SYSTEM -
<
i E/S
'
RED 2500 -
GREEN E!S
'
-
Lp00 -l P lA HS-100A 2200/2225 RECORDER 220C/2225 IX PR0100-A/c Hl/LO E/S i
RED
$y* PMS AuRM 50A14
- -
(2275/2175)
E/S _
c
.~
S ONIOFF
,
(2IOb25)
E/S i
-
c
Hi HI
'
'
=
'!
8/U HEATER TRIP (2340)
-
c
j e
,p PIC 0100
>
3ROPORTIONAL l
DE-ENERGlZE jEATERS
~
~ E/UHEATERS (2275)
l
,
-
%HCC 4LDM J
o 9,
y a o V4 Mi M V4 s trM%iiT' "
--
l VALV PCSmON Hl/LORM g 7chg33 w=;grg!;
tsme-hoQ14 (2275/2175)
'#' MEH9f#
l PV 100A PV 1008 A. e ren V4.gsvr
.
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p e I
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. _... _. _. _. _ _.. _. _ _. _. _.. _. _ _
.. _ -.. _, _
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COMMENT #6 RO Examination Question 37 (SRO37)
Procedure, SO23 12-7. Ims of Offhite Power / Loss of Forced Circulation, floating step 2 states that Thot and CET's are compared as are Thot and Tc are not rising to verify natural circulation is occurring. The S/G pressure can be used to correlate to Tc there fore answer B is also etnTect. Soutirr's California Edison believes there are two correct answers to this question.
Accept answers A & B
,
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_ _ _ _ _ _ _ _
,
NUCLEAR CEGANIZA~ ION
UNITS 2 AND 3 EMERGENCY CPERATING INSTR'JCTICN i
REVISION 15 5023-12-7
.
(*
ATTACHMENT 2 PAGE 30 CF 122
'j LC'S OF FORCED CIRCULATICN/ LOSS CF 0FFSITE PCW s
-
FLOATING STEPS ACTION /EXPECTE0' RESPONSE i
RESPONSE NOT OBTAINED NOTE:
Lcw flow during Natural Circulation slows RCS response to temperature changes.
Lec: transit time rises to between: S minutes and 10 minutes.
FS-2 HONITOR Natural Circulation
_
Established:
CHECX all RCPs - stopped.
a.
GO TO FS-4, MONITCM RCP Operating a.
Limits.
b.
CHECXatleastoneS/G b.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1) SBCS - operating
,,
OR INITIATE S023-12-9 Attach ent 8 RECOVERY. HEAT REMOVAL.
ADV. operating.
AND
.
2)
Feedwater - available.
CHECX operating icop AT - less c.
than SS*F.
IF any criteria c through g NOT o
satisfied.
d.
CHECX Tc and Th - NOT-rising.
THEN
-
CHECK Reactor Vessel Level
!
e.
(Plenum) - greater than or MAXIMIZE S/G 1evel - less than
.
'
equal to 100%t.
80% NR.
QSPDS page 622 RAISE available S/G steaming l
.
CFMS page 312 rate.
I Attachment 4.
}
i RAISE Core Exit Saturation Margin
.
CHECX operating loop Ts and REP
- greater than 20*F:
j f.
i CET - within 16*F:
QSPDS page 611
!
QSPDS page 611 i
CFMS page 311.
!
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/
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P ATTACHMENT 2 PAGE 4 0F 29
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- NUCLEAR CRGANIZAilCN UNITS 2 AND 3l EMERGENCY CFERATING INSTRUCTICN S023-12
.
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REVISION 15 i_g.. -
ATTACHMENT 2 PAGE 31 0F17)
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LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWE
.
FLOATING STEPS
ACTION / EXPECTED RESPONSE,
_ RESPONSE NOT OBTAINED
!'
'FS-2 MONITOR Natural Circulation Established:
(Continued)
g.
CHECK Core Exit Saturation Margin - greater than 20*F:
IF any criteria c through g NOT o
satisfied, i
QSPOS page 611 THEN CFMS page 311.
,
MAXIMIZE S/G level - less th'an
.-
Sok NR.
\\
.
RAISE avcilable S/G steaming
.
rate.
l RAISE Core Exit Saturation Margin
.
.
- greater than 20*F:
'
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QSPOS page 611
'
i CFMS page 311.
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ATTACHMENT 2.
PAGE 5 0F 29
- N>.'41,'- }..' Wif # !y'l!--{R'[,jiiI,51 y,3ITEI.0,69,7s,(S$359,M975UYN'
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. COMMENT #7 RO Examination Question 46 (SRO43)
Answer B is correct based on the strictest interpretation ofinunedsately before and aAer a trip. Immediately before the trip, S/O level exmeded 85% causing a High Level Override, HLO, signal to be applied to the main and bypass foM regulating valves causing both to close. The valves both stay closed until level decreases below 85% at which time the HLO signal clears and the valves go to either the Reactor Tripped override, RTO, position or to the position set by the demand from the feed water argulating control system. With a reactor trip, an RTO signal is sent which as soon as the HLO condition clears seconds aAer the trip due to normal shrink of water levels, the RTO signal is applied and the bypass valve opens to 50%, answer C. Southern California Edison believes there are two correct answers to this question.
'
Amept answers B & C l
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COMMENT #8
.
R0 EXAMINATION QUESTION #54
.
5023-6.2.9 states " Select the Bus Transfer Control AUT0/ MANUAL iWitch to AUTO after completion of the breaker manipulations that return the bus to its " Normal" configuration." Having a bus on the tie-brk is not a normal configuration.
So the AUT0/ MANUAL switch would not be placed in the AUTO position.
Southern California Edison believes there are no correct answers to this question.
Delete question I
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- * T.?MT.".C*T ~..? ^
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NUCLEAR ORGANIZATION OPERATING INSTRUCTION S023-6-2
.
UNITS 2 AND 3 REVISION 5 PAGE 7 0F 26 I
9..
)
6.0 PROCEDURE (Continued)
'
l 6.2'.8 If the INCOMING 4160V source is a bus tie or diesel l
generator output breaker, then open the RUNNING breaker,
'
if desired.
u 6.2.9 Select the Bus Transfer Controls AUT0/ MANUAL switch to
.
'
AUTO after completion of the breaker manipulations that return the bus to its " normal" corfiguration.
6.2.10 Remove the synchronizing circuit from service by
depressing SYNC pushbutton for the INCOMING breaker.
6.2.11 Remove the synchroscope from service by selecting the respective key-operated Master Control switch to 0FF.
6.2.12 Clear any annunciator alarms resulting from the transfer operation.
'
NOTE:
For Bus Transfers using the Bus Tie Breakers, the synchroscope and synchronizing circuits can
.
only be in service on one Unit at a time. After the first Bus Tie Breaker (regardless of Unit)
is operated, its associated synchronizing i
circuit must be de-energized and its
'
syr.chroscope removed from service prior to
.-
)
starting the evolution on the remaining Bus Tie
--'
Breaker.
6.2.13 For IE 4160V Bus Tie transfer schemes, perform the
'
following:
.
.1 Starting from the Unit which is to SUPPLY power:
NOTE:
The INCOMING Voltage and Frequency are sensed directly from the Tie Bus. The " BUSES PARALLELED" alarm logic is satisfied on a Unit
'
when BOTH Units Bus Tie Breakers are closed AND either a Reserve Aux Transformer or a Unit Aux Transformer Power Source breaker is closed.
When BOTH Bus Tie Breakers are closed A_ND a Transformer breaker on each Unit is Closed, BOTH Units will have BUSES PARALLELED alarms annunciated.
.1.1 Place the synchroscope in service per Step 6.2.3.
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COMMENT #9 RO Exanunation Question 74 (SRO69)
The question setup has the VCr pressure at 40 psig. The head of the Refueling Water Storage tank, RWST, is 70'
>
of H2O or 30.3 psig(70' x 0.433 psi /A) In the scenario provided the head of the VCT with the over pressure, will keep the check valves from the RWST closed. Southern California Edison believes there are no correct answers to this question and it should be deleted -
- Delete this Twiaa.
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l COMMENT #10 l
' RO Exanunation Quesuon 79
!
(SRO 75)
\\
!
!
Question was based on old Tavg program. New program has normal pressurizer level of 48% at 100% power.
j This is based on the reduced Tc program of $48 deg F @ 100% power. The lower Tc at full power equates to a Tave of $74 deg F. Per the attached reference. the expected level would be 48% and no additional charging pumps
- would be operating. Southern California Edison believes there are no correct answers for this question and the
'
question should be deleted from the examination.
Delete Question.
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NUCLEAR ORGANIZATION OPERATING INSTRUCTION
'
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REVISION 8 S 2 -3 1.10 l
UNITS 2 AND 3
.
l ATTACHMENT 5 AuE 30 0F 34
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PRESSURIZER LEVEL CONTROL PROGRAM
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s s c s s s s s s s s s. i s.
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s. s s
s 544 550 560 570 58Q 590 RCS AVERAGE TEMPERATURE ( F)
010 -1.C H T
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D10 8.wS1 ATTACHMENT 5 PAGE 1 0F 1 r
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COMMENT #11 RO EXAMINATION QUESTION 85
' (SRO 84)
The trends given are inconclusive as to whether vacuum has stabilized at 3.3", if assumed vacuum is stable at 3.3 "
- no further action will be required and answer B would be correct, ifit is assumed vacuum will continue the current
- irend, the listed action of"D" could be taken to return the plant to a more stable condition. Southern California Edison believes "B" & *D" are correct answers.
Accept B & D.
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NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTION S023-13-10
.
UNITS 2 AND 3 REVISION 2 PAGE 7 0F 13
.
LOSS OF CONDENSER VACUUM ACTION /EXPiCTED RESPONSE RESPONSE NOT OBTAINED 4 Actions for loss of vacuum due to Condenser fouling.
,
i CAUTION During periods of heavy influx, rapid and aggressive action may be required in :
@
order to avoid a Unit trip. ' Power may need to be reduced in order to,
,
t Bumpand/orStopCirculatingWaterPumpsontheCondenserquadrantswith l e
the highest differential pressures Maintain Condenser backpressure < 3.5" Hg j
l a. REDUCE Regctor power to 75% TO 854.
b. BUMP Circulating Water Pump (s)
per direction of Shift Superintendent, c. VERIFY backpressure < 3.5" Hg _
c.
1) REDUCE Reactor power and stable.
to s 65%.
'
2) STOP two Circulating Water Pumps on opposite ends of the Condenser.
j
,
3) INITIATE isolating stopped pumps per 5023-2-5, Attachment for Stopping a Circulating Water Pump Due to Fouled Condenser
.
Tubesheet/High AP/ Debris f
'
Removal.
4) IF not < 3.5" Hg and stable, THEN REDUCE Reactor Power as necessary to establish backpressure < 3.5" Hg and stable.
5) GO TO Step 5.
d. EVALUATE stopping pumps based on Waterbox differential pressure and pump vibration, e. GO TO Step 6.
- 2
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COMMEW # 12 RO Examination Questson 93 (SRO93)
Answer C is conect based on HV9217 and HV9218 being open and providing a direct path from inside containment to the outside, in this case to the VCT. Answer D is correct bcause given this event, Controlled Bleed -
,
O\\ff flow is routed to the quench tank via a relief valve when HV9217 and HV9218 are closed. With HV0514 and HV0515 being failed open, a dLect path for RCS water exists from the quench tank to the chemistry sample sink.
Therefore, acapt both answers C and D Accept answers C & D l
.
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l COMMENT # 13 RO EXAMINATION QUESTION 95 (SRO.%)
Both answers A & B will cause a FTS event to occur if the operator fails to initiate release of steam from E088 S/G. "A"is correct based on failing to steam the good S/G to establish a heat sink. "B"is also correct in that HPSI throttle /stop criteria will NOT be met because of the unavailability of the S/G as stated in the stem (step FS4 a.1 requires operating S/g with an ADV operating), and continuing to inject water into the RCS will increase pressure also leading to FTS event. Southern California Edison believes there are two correct answers to this question.
Accept answers A & B.
.
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NUCLEAR CRGANIZATION EMERGENCY ~0PERATING INSTRUCTION S023-12-5
UNITS 2 AND 3 REVISION 15 PAGE 43 0F 143 ATTACHMENT 2
.
i EXCESS STEAM DEMAND EVENT
'
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FLOATING STEPS
' ACTIONS / EXPECTED RESPONSE'
RESPONSE NOT CBTAINED FS-6 CHECK HPSI Throttle /Stop
,
Criteria:
a.
CHECKatleastone.S/G a. GO TO S023-12-9, FUNCTIONAL RECOVERY..
operating:
AND
'1)
SBCS -' operating.
INITIATE 5023-12-9, Attachment 8, n
OR RECOVERY - HEAT REMOVAL.
,
ADV'- operating.
-
AND 2)
Feedwapr - availa.ble, b.
CHECK PZR level o =IF any criteria of steps b through d NOT met,
-- greater.than'30%
THEN
'AND OPERATE Charging and HPSI systems
,
.
- NOT lowering.-
as necessary to maintain (
Throttle /Stop criteria c.
CHECK Core Exit Saturation
- satisfied.
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Margin. greater than 20*F:
THROTTLE Loop Injection Valves.
.
.QSPDS page'611 CFMS page 311.
ENSURE auxiliaries to SI pumps:
.
a)
Electrical power to pumps and j
valves.
)
b)
Proper system alignment.
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c) CCW flow.
d) HVAC.
. ATTACHMENT 2 PAGE 13 0F 43
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NUCLEAR GRGANIZATICN EMERGENCY OPERATING INSTRUCTION S023-12-5
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UNITS 2'AND 3 REVISION 15-PAGE a2 CF 123
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ATTACHMENT.2
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EXCESS STEAM DEMAND EVENT
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FLOATING STEPS-
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. ACTIONS / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-6-CHECK-HPSI. Throttle /Stop Criteria:
(Continued)
d.
CHECK Reactor Vessel Level'
.o-IF any criteria of steos b thrcugn. d
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(Plenum)-- greater than or NOT met,
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. equal to 1004:
-THEN
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.0SPDS page 622 CFMS page 312 OPERATE Charging and HPSI. systems
.
Attachment 5.
as necessary to. maintain
{
Throttle / Step criteria
,
- satisfied.
THROTTLE Loop Injection Valves.
.
ENSURE auxiliaries to Si pumps.:
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a)
Electrical power to pumps and valves.
b)
Proper system alignment.
c)
CCW flow, d)
HVAC.
e. ; VERIFY RCS borated - greater-e. MAINTAIN Emergency Boration
.
than ~ Technical Specification ~
- at least 40 GPM.
Shutdown Margin for T vt > 200*F A
per Operations Physics Summary-Figure 2.3-1, j
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RCS Cooldown - NOT in progress.
f, THROTTLE ~0R STOP HPSI as required one train at a time.
%
g.
STOP charging pumps as required one at a time,
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ATTACHMENT 2 PAGE 14 0F 43
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NUCLEAR ORG'ANIZATION
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EMERGENCY OPERATING INSTRUCTION 5023-12-5 UNITS 2 AND'3 REVISION 15 PAGE 45 0F I:3 -
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ATTACHMENT 2
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EXCESS STEAM DEMAND EVENT
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FLOATING STEPS ACTIONS / EXPECTED-RESPONSE RESPONSE NOT OSTA!NEO
'FS-6 CHECX HPSI Throttle /Stop Criteria:
(Continued)
h.
MAINTAIN Criteria of steps a
~ through e - satisfied.
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CHECK Containment' pressure-
.i. 1)- ENSURE SIAS - actuated.
i
.less than 3.4-'PSIG.
2)
GO'TO FS-7, CHECX LPSI Termination Criteria.
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CHECK PZR Level J. INITIATE FS-22, ESTABLISH CVCS
- less than 80%.
Letdown Flow, k.
RESET'SIAS per 5023-3-2.22,
.ESFAS OPERA-TION.
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ATTACHMENT 2 PAGE IS OF 43
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COMMEW # 14 i
RO EXAMINATION QUESTION %
(SRO 97)
i Answer "B" is correct based on the information given. However answer "A" is also correct based on procedure I
SO23 12-7, Safety Function Status Checks, which requires subcooling > 20F or you are directed to the Functional Recovery for not meeting natural circulation. Southern California Edison believes there are two correct answers to this question.
- Accept answers A & B.
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NUCLEAR ORGANIZATION UNITS 2lAND 3 EMERGENCY OPERATING INSTRUCTION 5023-12-7 f
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REVISION 15
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ATTACHMENT 2 PAGE 30 CF 122
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LOSS. OF FORCED CIRCULATION / LOSS OF 0FFSITE POWER
,
FLOATING STEPS
.
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED i
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NOTE:
Low flow during Natural Circulation slows RCS
'
response to temperature changes.
Loop transit time rises.to between 5 minutes and 10 minutes.
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FS-2 MONITOR Natural Circulation Established:
)
a.
CHECK'all RCPs - stopped.
a. GO TO FS t, MONITOR RCP Operating Limits.
l b.
CHECK at least one S/G b. GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND
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1) SBCS - operating
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INITIATE S023-12-9, Attachment 8,
.
OR RECOVERY - HEAT REMOVAL.
,
ADV - operating.
i AND 2)
Feedwater - available.
I c.
CHECK operating loop AT - less o
IF any criteria c through g NOT than'58'F.
satisfied, j-d CHECK Tc and TH - NOT rising.
THEN e.
CHECK Reactor Vessel Level MAXIMIZES /Glevel-lessthan
.
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(Plenum) - greater than or 80% NR.
t equal to.1004:
RAISE available S/G steaming
.
QSPDS page 622 rate.
CFMS page 312 Attachment 4 RAISE Core Exit Saturation Margin
.
- greater than 20'F-
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-f. 'CHECN operating loop Ts and REP i
CET - within 16*F:
QSPOS page 611 CFMS page 311.
QSPDS page 611
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CFMS page 311.
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NUCLEAR ORGANIZATION t-UNITS 2 AND 3 EMERGENCY OPERATING INSTRUCTION S023-12-7 REVISION 15 PAGE 31 0F 122 ATTACHMENT 2
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LO.SOFFORCEDCIRCULATION/LOSSOF0FFSITEPOWER FLOATING STEPS i
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-2 MONITOR Natural Circulation Established:
(Continued)
g.
CHECK' Core Exit Saturation o
IF any criteria-c through g NOT Margin - greater than 20*F:
_ satisfied,
-
OSPOS page 611 THEN CFMS page 311.
MAXIMIZE S/G 1evel - less than
.
804 NR.
RAISE available S/G steaming i
.
rate.
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RAISE Core Exit Saturation Margin
.
- greater than 20*F-
,
QSPOS page 611 CFMS page 311.
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COMMENT #1 i
SRO Examination Question 9 (RO7).
The question stem references SO23 3-3.27.3 as do the possible answers. The actual procedure that should have been referenad is SO23-3-3.23, Emergency Diesel Generator Monthlv Surveillance. The given procedure, SO23-3-3.27.3 is not used to perform the AC Sources check that is required by the given scenario. The correct answer was not provuled. Southern California Edison believes there are no correct answers to this question.
Delete the que. tion.
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COMMENT # 18 SRO Examination Question 93 (RO93)
Answer C is correct based on HV9217 and HV9218 being open and providing a direct path from inside containment to the outside, in this case to the VC7 Answer D is correct because given this event, Controlled Bleed O\\ff flow is routed to the quench tank via a relief valve when HV9217 and HV9218 are closed. With HV0514 and HV0515 being failed open, a direct path for RCS water exists from the quench tank to the chemistry sample sink.
Therefore, accept both answers C and D Accept answers C & D
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NUCLEAR ORGANIZATION SURVEILLANCE OPERATING INSTRUCTION S023-3-3.23 UNITS 2 AND 3 REVISION 14 TCH 14-2 PAGE 72 0F 88
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ATTACHMENT 7 A. C. SOURCES VERIFICATION (MODES 1-41
.,
OBJECTIVE To provide verification that sufficient AC Sources are available to the 1E 4.16kV Busses when any combination of Offsite Circuits, Onsite Circuits, and Diesel Generators are Inoperable. This
,
attachment satisfies Surveillance. requirement of Tech. Spec.
LC0 3.8.1'AC Sources Verification.
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UNIT MODE-(1-4)
DATE TIME PERF. BY 1.0 PREREOUISITES INITIALS 1.1 Verify this document is current by checking a controlled copy or by using the method described in 50123-VI-0.9.
1.2 List the reason for performing this attachment (e.g., Diesel
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Generator 2G002 Inoperability).
2.0 AC SOURCES VERIFICATION 2.1 If this attachment is being performed prior to declaring a piece of equipment Inoperable, then assume the equipment is Inoperable when performing the attachment.
2.2 If the specific equipment Inoperability has placed both Units in action statements, then a separate attachment will have to be performed for each Unit.
2.3 If a Diesel is Inoperable, then determine if the cause of the Diesel Generator Inoperability may exist on the other Diesel Generator (s).
2.3;1 If the cause of the Diesel Generator Inoperability exists on the other Diesel
- Geneiator(s), then declare the affected Diasel(s).'noperable.
2.4 If desired u',e the last page of this Attachment to assist
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ATTACHMENT 7 PAGE 1 0F 7
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NUCLEAR ORGANIZATION SURVEILLANCE OPERATING INSTRUCTION S023-3-3.27.2
UNITS 2 AND 3 REVISION 10 PAGE 4 0F 26 ATTACHMENT 1 WEEKLY ELECTRICAL BUS SURVEILLANCE - Both Units in Modes 1 thru 4 OBJECTIVE To verify Operability of the offsite transmission network, onsite Class 1E distribution system (except the diesel generators), and the onsite DC systems as required by the Technical Specification Surveillance requirements:
SR 3.8.1.1, SR 3.8.4.1, SR 3.8.7.1, SR 3.8.9.1.
To verify the functionality of the Spent Fuel Pool Cooling System power availability as required by the Administrative Technical Specification.
UNIT 2 MODE UNIT 3 MODE DATE
_
PERF BY 1.0 PRERE0VISITES 181114LS 1.1 VERIFY this document is current by checking a controlled copy or by using the method described in 50123-VI-0.9.
1.2 DETERMINE the performance requirements of this attachment,
,
as f~ollows:
)
SR0 Ops.
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[] This Attachment is being performed for a scheduled surveillance.
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[] This Attachment is being performed for operability verification. LIST the Components and Sections Steps lR to be performed. After approval, then CIRCLE N A for the remaining unused steps.
COMP 0NENTS SECTIONS / STEPS
OPERABILITY VERIFICATION PREPARED BY:
Control Room Operator OPERABILITY VERIFICATION APPROVED BY:
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ATTACHMENT 1 PAGE 1 0F 7
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' COMMENT #2 SRO Examination Question 10
- Infrequently performed test can also be interpreted to be special tests. SO123-IT-1, Infrequently Performed Tests, states that infrequently pedormed tests can also be performed under the special test procedure. Since 5023-0-23 is also used to conduct short term valve lineup changes, it too is a correct choice. Southern California Edison believes that there are two answers to this question.
Accept answers A & D D
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NUCLEAR ORGANIZATION GENERAL ORDER 50123-IT-1
Uti!TS 1, 2 AND 3 REVISION 4 PAGE 3 0F 16
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l III. RESPONSIBILITIES (Continued)
F.
The Manaaement Desianee (see Definitions, Attachment A) exercises continuous responsibility for Management Oversight. With the approval of the Vice President, Nuclear Generation and/or the Senior Vice President, Power Generation, may exercise Management Oversight on a " spot-check" basis.
G.
The Test Soecialist (see Definitions, Attachment A) is a technical resource to the supervisor who has operational responsibility for conduct of the test or evolution.
H.
Licensed Goerattr_1 and Plant Manacement Staff (see Definitions, Attachment A) have the responsibility to recognize tests and evolutions which are (or should be) included in the IPTE List.
IV.
REQUIREMENTS A.
Infrequently Performed Tests and Evolutions (IPTE) which take plant personnel or equipment beyond the bounds of normal procedures, traininqroperaTiTig-ban 6, ur exper3ence. ed (ich repr"eem
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. siggificant safety or economic risk, require controlling documents with enhanced development and review as outlined by this order.y B.
Execution of IPTE activities require Management O M Definitions, Attachment A) with clear direction, clear communication of management expectations with respect to margins of safety,
,
expected plant response, termination criteria, and actions to be l
taken in the event of unexpected results.
C.
Direction to licensed or non-licensed personnel with regard to the operation of the plant shall be given only by personnel who possess a SR0/R0 license a.nd are designated with responsibility for the safe conduct of the evolution or test.
D.
The highest margin of safety shall be maintained throughout the test or evolution exercising caution and conservatism, particulariy when uncertainties or unexpected plant behavior is encountered.
E.
Req 61rements for test or evolution termination shall be clearly defined, conhunicated, and understood by all persons involved with the conduct of the test or evolution.
F.
IPTEs shall be conducted and documented using the IPTE Checklist (Attachment D) per the Keypoints guidance (Attachment E).
G.
If it is necessary to change the IPTE controlling document prior to or during use, ItiEH the associated affect on IPTE fntent (see
,
Definitions, Attachment A) and plent safety shall be considered.
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lE the intent is affected, THER prior to starting or continuing with the IPTE, the same approval level as the original controlling l
document is required. Prior to any IPTE controlling document
'
changes, extreme caation and consideration should be exercised.
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NUCLEAR ORGANIZATION'
OPERATIONS DIVISION PROCEDURE S0123-0-23
UNITS 1, 2 AND 3 REVISION 5 PAGE 47 0F 62
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L ATTACHMENT 4
XEYPOINTS FOR ABNORMAL ALIGNMENT / EVOLUTION PAGE 10F.D]
]
COMPONENT: [1]
LOG NUM3ER
- [2]
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PURPOSE OF ALIGNMENT:
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Procedure Change Required O No O Yes
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Verify this~ document is current by checking a controlled copy or by using the
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method described in 50123-VI-0.9. [5]
l EFFECT OF ABNORMAL ALIGNMENT / EVOLUTION NO YES l
Has it been addressed in a completed:
// O INDICATE document Type and
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Number:
,
50.59 Safety Evaluation. E O ATTACH PF(123)109-1, Unreviewed e
Safety Question Screening
PF(123)109-1, Unreviewed Safety
/
Criteria e
Question Screening Criteria? [6]
fj
Was SCE PF(123)l09-1 checked YES in O
O 00 NOT PERFORM until Part II is PART I? (Check N0 if form not used.)[7]
completed.
'
Does it:
O O OBTAIN approval from Manager, f
.
e Change the intent of the Operating Operations prior to Instruction, E -
impiementation.
.
Constitute an Evolution, E e
Require a new or additional 50.59 [8]
e
Could it pose adverse environmental O
O DO NOT PERFORM until a review effects of any] type directly or from Environmental Protection j
indirectly? [9 is attached.
Is it involved with multiple evolutions O
O ATTACH Marked-up P& ids, and on the same system, an interconnected OBTAIN approval from the Shif t system, E will rasult in_ theJDov0 ment Superintendent as the SR0/CFH.
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of gases or1Tquids? [10]
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Is t'a complex alignment which:
O O OBTAIN approval from the Plant
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e Is requested by another division, E /
Superintendent or designee as
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Requires non-rautine interdivisional /
the Plant Management Staff-
/ coordination, E
/
Operations.
- Installs temporary plant equipmept'
l that could alter the function path of existing plan com nts? [11]
PREPARED / REVIEWED & APPROVED DATE TIME PREPARER
[12]
MANAGER, REQUESTING ORGANIZATION O N/A (13]
PLANT MANAGEMENT STAFF - OPERATIONS
[13]
(13]
MANAGER, CPERATIONS
[14]
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ATTACMMENT 4 PAGE 1 0F 14
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i NUCLEAR ORGANIZATION OPERATIONS DIVISION PROCEDURE 50123-0-23
UNITS 1, 2 AND 3 REVISION 5 PAGE 57 0F 62
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ATTACHMENT 4 KEYPOINTS FOR ABNORMAL ALIGNMENT / EVOLUTION (Continued)
9.
H this Abnormal Alignment / Evolution could pose any type of adverse L
environmental effects, than Environmental Protection must review this
'
permit before implementation, and provide documentation to be attacned.
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10.
If the Abnormal Alignment / Evolution is involved in multiple evolutions on the same system, an interconnected system, E the evolution will result in I
the movement or gases or liquids, ihan check YES, and attach marked-up
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P&lDs.
In November 1993, failure to properly evaluate system i
irterconnection flowpaths resulted in HPSI Pump run-out and caused
'
extensive pump damage. Drawings are required to assist in Abnormal l
Alignment / Evolution review and tailboard, and therefore are not recuired to
)
be attached to the ccmpleted Abnormal Alignment / Evolution.
(Ref. 2.4.7)
)
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11.
If this a complex alignment requested by another division, E requires non-routine interdivisional coordination, E installs temporary plant equipment
,
that could alter the function or flowpath of existing plant components,
'
including in-service or hydrostatic testing, then check YES.
12.
After preparing the document for use (including Return-to-Service instructions) the Preparer will enter name, date, and time in the space provided. -The individual preparing the document SHALL NOT sign any of the Reviewed and Approved By lines.
l 13.
Approval is normally required prior to using)the document. (Refertomain body, Steps 6.1.4 and 6.8.5.1 for exception.
H the Abnormal Alignment / Evolution (AA) was requested by an organization other than Operations, then the Manager of that division is required to review and approve the activity.
If the Operations Division initiated this AA, then ChecktheN/Aboxinthe" Manager,RequestingOrganization" space.
14.
Approval is required by the Manager, Operations prior to implementation if the Abnormal Alignment / Evolution changes the intent of the Operating Instruction, E constitutes an Evolution. H not, then implementation may proceed prior to the Manager, Operations final approval, provided approval isobtainedwithin14daysofSR0/CFHApproval.
alignment.(pecific document number that will allow closure of this 15.
Enter the s e.g., Closure of a WAR, TFM, or NCR}. H closure is " completion of this al'ignment", and no other documents will be involved, than state so.
E a procedure change is required, then check YES. OPG should also be g
notified (e.g., E-Mail).
Editorial information may be included by USER (S) in the form of numbered notes in the Comments section (e.g., add su WAR Number to the Closure Document section)pporting information such as a Such information does not
,
change the intent, method, or outcome of the Abnormal Alignment / Evolution.
l 16.
Insert the number and name of the associated System Operating i
Instruction (s).
I 17.
Enter any pertinent additional references (e.g., Technical Manual, UFSAR, l-
. Site Procedure, etc.). If none, then check NO.
f ATTACHMENT 4 PAGE 11 0F 14
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NUCLEAR ORGANIZATION OPERATIONS DIVISION PROCEDURE S0123-0-23
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UNITS 1, 2 AND 3 REVISION 5 PAGE 10 0F 62
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6.0 PROCEDURE (Continued)
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-6.4
' Cont'rol of System Alionments Affected By System Modifications
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6.4.1 Permanent facility modifications will be accounted for by TCNs or revisions to the system Operating Instructions.
NOTE:
S0123-0-22 provides specific direction regarding control of system alignrrents due to temporary
,
facility modifications, 6.4.2 Temporary modifications lasting greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> should be accounted for by TCNs or Revisions to the system Operating Instruction (s).
.1 When the expected duration of the temporary modification is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then Section 6.8 should be used to document the change.
.2 Whita the. temporary modification is expected to 'last for
'
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 TCH or revision, then Section 6.8 should be used to document the change. This.is allowed provided the
Operations Procedures Group is actively preparing a TCN or revision for' issuance.
END OF SECTION 6.4
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SRO bXAMINATION QUESTION 12
- ')iR O 9)f-S023 5-1.8 is the reference for "A" to be a correct answer. "C" is also correct based on Technical Specification 3.4.6 and 3.4.7. which requires the RCS LOOP !b he operable. Southern California Edison believe there are two correct answers to this question.
' Accept answers A & C
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NUCLEAR ORGANIZATION INTEGRATED OPERATING INSTRUCTION 5023-5-1.8
UNITS 2 AND 3 REVISION 9 PAGE 86 0F 91
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ATTACHMENT 13 9.o kCP OPERATION 9.1 With at least one RCP operating, reverse flow will be present in the idle loop.
jD 9.2 The loss of RCP heat will affect cooldown rate. Consequently, Tcold should be maintained 2125'F to prevent entering the restrictive heatup and cooldown limitations that apply when s120'F.
9.3 When securing RCPs, it may be necessary to reduce PZR heater output due to.the reduction of PZR Spray Valve bypass flow.
9.4 Due to insufficient Pressurizer heater capacity, it may be necessary to secure all RCPs and main spray prior to initiating Auxiliary Sp ay.
Otherwise, loss of MPSH for the RCPs could occur.
(Ref. 2.3.17 9.5 Pressurizer insurge may occur when securing the last RCP. This is caused due to the lower RCS flow across the co-e. As Core Exit Temperature rises, the RCS will swell into the Pressurizer. Adjusting letdown flow will help minimize this insurge.
9.6 Indicated Tcold will initially rapidly lower in any loop where SDC is injecting, if the RCP operating in that loop is stopped or when the last RCP is stopped. This is due to cooler SDCS injection water flowing over the loop Tcold temperature element.
9.7 If any RCPs are operating, then the Tcold associated with an operating RCP should be used for RCS temperature monitoring.
9.8 WhentherearenoRCPsoperating,thenTR-0351A(T351X),SDCCombined outlet Temperature, should be used for Tcold ternperature monitoring.
9.9 IE RCPs are running. IllEli one RCP shall remcin in service until completing RCS boration to Mode 5, or refueling concentration and other forced circulation dependent parameters are met (e.g., hydrogen, peroxide, etc.).
9.10 When the last RCP is stopped, then RCS Thot indications (T351Y and CETs) will begin to rise due to the increased time coolant is in the j
Core region (i.e., no RCP forced circulation). Consequently, SDCS flowrate should be adjusted to maintain RCS Tcold TR-0351A (T351X) at the desired tentperature.
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ATTACHMENT 13 PAGE 6 0F 11
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RCS Loops--MODE 4
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3.4 REACTOR COOLANT SYSTEM (RCS)-
3.4.6. RCS Loopi--MODE 4 -
LCO 3.4.6 Two loops or trains consisting of ray combination of RCS loops and shutdown cooling (SDC) trains shall be OPERABLE and at least one loop or train shall be in operation.
NOTES---------------------------
1.
All reactor coolant pumps (RCPs) and SDC pumps may be de-energized for s 1 bour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
a.
No operatiens are permitted that would cause reduction of the RCS baron concentration; and b.
Core outlet temperature is maintained at least 10*F below saturation temperature.
2.
No RCP shall be started with any RCS cold leg temperature s 256*F unless:
.
a.
Pres:urizer water volume is < 900 ft', or b.
Secondary side water temperature in each steam generator (SG) is < 100*F above each of the RCS cold
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leg temperatures.
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APPLICABILITY:
MODE 4.
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SANONOFRE-bHIT2 3.4-18 Amendment No. 127 e
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3.4.6
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CONDITION REQUIRED ACTION COMPLETION TIME A.
One required RCS loop A.1 Initiate action to Immediately inoperable.
restore a second loop or train to OPERABLE AND 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 Immediately
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operations involving inoperable.
reduction of RCS boron concentration.
EE AND
No RCS loop or SDC train in operation.
C.2 Initiate action to Immediately re. store one loop or train to OPERABLE status and operation.
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A SURVEILLANCE REQUIREMENTS.
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SURVEILLANCE 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.
'5R 3.4.6.2 Verify secondary side water level in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> required SG(s) is 2 50% (wide range).
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SR 3.4.6.3 Verify the second required RCS Loop or SDC 7 days train is OPERABLE.
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RCS Loops-MODE 5, Loops Filled
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3.4.7
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3.4 REACTOR COOLANT SYSTEM (RCS)
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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:
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a.
Reactor Coolant Loop 1 and its associated steam generator and at least one associated Reactor Coolant Pump; b.
Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reactor Coolant i
Pump;
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Shutdown Cooling Train A; or d.
Shutdown Cooling Train B
One additional Reactor Coolant Loop / shutdown cooling train
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' hall be OPERABLE, or s
The secondary side water level of each steam generator shall be greater than 50% (wide range).
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NOTES---------------------------
o 1.
All reactor coolant pumps (RCPs) and pumps providing
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shutdown cooling may be de-energized for i I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided:
a.
No operations are permitted that would cause reduction of the RCS boron concentration; and b.
Core outlet temperature is maintained ai least 10'F below saturation temperature.
2.
One required SDC train may be inoperable for up to i
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other
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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)
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SAN ON0FRE--UNIT 2 3.4-21 Amendment No. 127
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RCS Loops--MODE 5, Loops Filled
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NOTES (continusd)---------------------
4.
No reactor' coolant pump (RCP) shall be started with one or more of the RCS cold leg temperatures s 256*F unless:
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l a. -The pressurizer water volume is < 900 ft3 or b.
The secondary side water temperature in each steam generator (SG) is < 100*F above each of the RCS cold leg temperatures.
_
5.
A containment spray pump may be used in placs of.a low
!
pressure safety injection pump in either or both shutdown cooling trains to provide shutdown cooling flow provided the reactor has been suberitical for a period
> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the RCS is fully depressurized and vented in accordance with LCO 3.4.12.1.
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All SDC trains may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is
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in operation.
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APPLICABILITY:
MODE 5 with RCS loops filled.
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_'..l ACTIONS
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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
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to OPERABLE status.
AND DE Any SG with secondary side water level not A.2 Initiate action to Immedi ately within limit.
restore SG secondary side water levels to within limits.
(continued)
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SAN'ON0FRE--UNIT 2 3.4-22 Amendment No. 127
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ACTIONS- (continued)
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~B.
No SDC train /RCS loop _
B.1 Suspend all Immediately in operation.
operations involving
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reduction-in RCS i
baron concentration.
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.E.N.Q.
B.2 Initiate action to Immediately i
restore required SDC L
train /RCS loop to
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l operation.
' SURVEILLANCE REQUIREMENTS
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' SURVEILLANCE FREQUENCY
....)
SR 3.4./.1 Verify at least one RCS loop or SDC train 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
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is in operation.
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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 50Y. (wide range).
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L SR 3.4.7.3 Verify the second required RCS icop, SDC 7 days tFain or steam generator secondary is t.
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SAN ONOCRE--UNIT 2 3.4-23 Amendment No. 127
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COMMENT H SRO Examination Question 18 (RO14)
The generation of the Control Element Assembly, CEA, deviation alarm is a multi component process. The Reed Switch Position Transmitters, RSM's, actually sense the CEA's position. The Control Element Assembly Calculator, CEAC, uses the input from the RSPT and sends a signal to the alarm. Both components are needed to
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generate a deviation alarm. Southern California Edison believes there are two correct answers to this question.
- Accept answers B & C
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. NUCLEAR ORGANIZATION ALARM RESPO!ISE. INSTRUCTION 5023-15-50.Al
. UNITS 2-AND 3 REVISION 2 PAGE 71 0F 76 ATTACHMENT 2
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50A28 CEA DEyIATION APPLICABILITY PRIORITY.
REFLASH ASSOCIATED WINDOWS i
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Modes 1-3 AMBER NO NONE
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INITIATING NOUN NAME SETPOINT VALIDATION PMS 10 LINK t DEVICE INSTRUMENT U2/U3
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2(3)LO91, CEAC 1 Control Element 5 Inches NONE-DEV1AR56 641/663 lR or CEAC 2 Assembly Deviation 1.0 RE0VIRED ACTIONS:
1.1 Position the CEDMCS Mode Selector Switch on 2(3)CR50 to 0FF.
1.2 Verify which'CEA is misaligned and the amount of misalignment, by observation.of the following:
CEAC display CRT
CEAC remote operators modules
- PMS alarms
- PMS readout
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2.0 CORRECTIVE ACTIONS:
SPECIFIC.CAUSES SPECIFIC CORRECTIVE ACTIONS
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2.1' Misaligned CEA 2.1 After the misaligned CEA has been determined, _thful:
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2.1.1 Notify the SR0 Ops. Supv.
2.1.2 Realign.the CEA per 5023-3-2.19, Section for Manual Individual (
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2.2 Slipped or Dropped' CEA 2.2 GO TO S023-13-13, Misaligned Control Element Assembly.
l-3.0-ASSOCIATED RESPONSES:
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3.1 Notify the CRS/SS and the STA to review lech. Specs. LC0 3.1.5 and LCS3.1.105,andinitiateanEDMR/LC0AR,asrequired.
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NUCLEAR ORGANIZAfl04 SYSTEM DESCRIPTION SD.5023-710
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UNITS 2 AND 3 ls REVISION 3 PAGE 72 0F 76
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FIGURE 15? CONTROL ELEMENT ASSEMBLY SUBGROUP WEED SWITCH POSITION TRANSMITTER SIGNAL AS$1GNMENTS l
. h EX-CORE CHANNEL RSPTA RSPT\\
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_2 23 CEAS l
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h RSP RSPT
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22 CEAS Chs i
23 CEAS 22 CgAs RSPT i
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22 CE N
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23 CEAS
23 CEA3 h
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CEA CALCULATOR CALCULATOR
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CEA POSITCN NO.1 NO. 2 CEA POSITION l
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CA? A tmG CAT A UMci l
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A CORE 8 CORE C CORE DCORE
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PROTECTON PROTECTION PROTECTION PROTECTON
CALCULATOR CALCULATOR CALCULATOR CALCULATOR
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OPERATOR'S OPERATOR'S OPERATORM OPERATOR'S MODULE MCOULE MODULE i MODULE CRT OtSPLAY I
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NOTES.1. SIGNAL FROM CEA 2 IS CCNNECTED TO CPC's A AND C, BUT IT IS NOT USED AS A TARGET CEA.
2. SIGNAL FROM CEA 3 CS CONNECTED TO CPC's 5 AND D. BUT IT is NOT USEO AS A TARGET CEA.
l 3. SIGNALS FRCN 23 CEA's ARE CONNECTED TO EACH CPC.
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COMMENT # 3
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SRO EXAMINATION QUESTION 22
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There is no correct answer. Actual allowable maximum level is $7% per SO23 3 1,7. L&S 2.2. Southern California Edison believes there is no correct answer to this question.
' Delete Question s
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NUCLEAR ORGANIZATION OPERATING INSTRUCTION S023-3-1.7 UNITS 2 AND 3 REVISION 20 TCN 20-2 PAGE 56 0F 56
ATTACHMENT 16 1.0 REACTOR COOLANT PUMPS (Continued)
1.15 2 (3)'P-002 : For the ABB RCP Motors, the Lif t Oil Pumps normal discharge pressure is 1400 psig (allowable range: 1377 to 1450 psig),
1.16 Bleed-off flow normally is proportional to RCS pressure.
At 2250 psia, bleed-off flow should be between 1.25 gpm and 1.75 gpm.
If at a low pressure, and CB0 flow is < 0.25 gpm,.th23 one of the following is required:
CB0 line temperature (at the local flow indicator) is warm
(i.e., cold line indicates no flow)
RCP Seal Cavity pressures are properly staged.
- 2.0 REACTOR COOLANT SYSTEM 2.1 If Boron concentration in an idle loop is suspected of being lower than Reactor Core boron concentration, IHf3 DO NOT attempt to Start a RCP in that idle loop. This will prevent a possible reactivity transient upon restart of forced flow.
(Ref. 2.1.5)
With'any RCS cold leg temperature s 260'F, DO NOT Start a RCP unless 2.2 the following conditions for PZR level and RCS temperature are met.
Use the most conservative values available in order to maximize delta T (Tsat-Tc).
(Ref. 2.3.1 and Tech Spec. LCO 3.4.6, LCO 3.4.7)
PZR LEVEL
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RCS TEMP (ADMIN LIMIT)
RCS TEMP (TECH SPEC)
30%"
T,,, (S/G) <T, + 20
~)
57% (<900 f t')
T,,, (S/G) <T, + 10 T,,, (S/G) <T, + 100 2.3 If the RCS has just been initially filled (air trapped in S/G 'U'
tubes), then RCS pressure may drop rapidly below the minimum pressure for RCP operation.
2.4 If the RCS is solid, then RCS pressure may rapidly rise above the maximum pressure for SDC loop operation due to heat transfer from the S/GtotheRCS.
2.5 If RCS pressure is being maintained by the Letdown Backpressure controller, then automatic operation may tend to raise RCS pressure by the amount of RCP differential pressure since letdown comes from the pump suction cold leg.
2.6 Failure of the seals to stage on an operating RCP with RCS pressure greater than 700 psia is an indication of a failed seal (s).
3.0 STOPPING AN RCP 3.1 When in Mode 3, then failure to bypass the associated SG Low Flow Trip before stopping an RCP will rr tult in a Reactor trip signal.
ATTACHMENT 16 PAGE 4 0F 4
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COMMENT #6 SRO Examination Question 28 (RO27)
Answer "B" is correct because 656 deg F corresponds to 2300 psia. 'Ibe pressurizer spray valves open at 2275 psia.
Answer "D" is also correct because at 2225 psia and a backup signal at 2275 psia, the heaters get a signal to turn off. Southern California Edison believes that there are two correct answers to this question.
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Accept answers B & D -
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$UCLEAR ORGANIZATION
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SYSTEM DESCRIPTION SD-SO23-360.
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UNITS 2 AND 3
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REVISION 5 Page 168 of 205
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FIGURE 111-5 ' PRESSURIZER PRESSURE CONTROL SYSTEM B' LOCK DIAGRAM
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100X PT 100Y
. 1E HEATER >-+ PMS/CFMS PMS/CFMS -,1E HEATER
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E/S
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8/S RED GREEN
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A HS-100A
0-2200/2225
2200/2225 g1gDM Hg B/S i
RED
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PMS 50A14
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(2275/2175)
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COMMENT #7
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SRO Examination Question 30 The design basis for adding the steam generator delta p trip was based on steam line or feed line break (harsh environment) inside containment accompanied by a loss of offsite power The steam generator delta p signal used for a reactor trip due to a sheared RCP shaft is not related to the loss of offsite power. This makes the answer A incorrect.
Southern California Edison believes there are no correct answers to the question.
Delete this question.-
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SOUTHERN CALIF 0RNIA EDISON NUCLEAR ENGINEERING, SAFETY AND LICENSING PLANT PROTECTION SYSTEM
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- DESIGN BASES DOCUMENT.
0B0-5023-710. REV. 4 PAGE 92 0F 569 ADDRESSABLE CONSTANTS
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Symbol Definition Range
' BUFTRP -
Snapshot Buffer Control Flag 0 or 1 TCBP Maximum time tha' the RPC Flag can remain set 0 to 40.0 (seconds)
TCOUNT-CRT Display Rewrite Control Flag 0 or 1 2.1.1.15 LowReactorCoolantFlow(LRCF)
Eachsteamgenerator'2(3)E088and2 measurement of differential pressure (3)E089 has an RCS four channe measured across the primary
- side, which is indicative of RCS Flow. This ' function was originally-added to the RPS to provide a qualified means of tripping the reactor for a SLB or FWLB inside containment accompanied by a loss of offsite power, since not all of the LDNBR signal inputs (i.e., RCPSSSS) were qualified to function in the harsh environment created by those accidents."**""'"""
Another substantive reason for adding this function was to provide batter protection against the sheared shaft event, whose impor.tance in the safety analysis of design basis events had elevated since the original plant analyses was performed.
The shearing of a RCP shaft was not considered a design basis event in the initial 3410 MWt reactor design, since it was not required by Revision 1 of R.G.1.70.
However, Revision 2 of R.G.1.70 requires that this event be considered in the preparation of the FSAR.
Additionally, analyses performed demonstrated that acceptable
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consequences cannot be demonstrated without pn. iding some sort of protective action, and that this event has about the same probability of occurrence as the seized shaft event. The protection offered by the CPCs for this event could be compromised if the RCP shaft were to shear above the RCPSS sensors.- The low flow trip function utilizing a variable setpoint based on steam generator primary differential-pressure was selected as the optimum design to mitigate this event, since it does not depend sn the CPCs, and could be developed, installed, and meet licensing schedules.""
Th'e PPS provides a channel trip when the ACS flow-produced differential pressure falls below the setpoirit.. A reactor trip then follows on a 2-out-of-4 basis. This trip function is presently credited to help mitigate the consequences of a sheared RCP shaft accident, or a two-pump or four-pump coast down event, and is therefore classified as a safety function per 2.1 (iii). See Table-B-12. Applicable Modes are 1 and 2.
This trip function has a variable setpoint feature that causes the differential pressure setpoint to track below the measured differential pressure by a pre-determined increment. The tracking rate of the setpoint is rate-limited, in that it can decrease only at a pre-selected maximum rate, and only to a pre-selected minimum value (" floor"). Should the signal level fall below the setpoint level m
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. COMMENT #8 SRO Examination Question 37 l
(RO37)
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Procedure, SO23 12-7. Ims of Offsite Power / Loss of Forced Circulation, floating step 2 states that Thot and CET's are compared as are Thot and Tc are not rising to verify natural circulation is occurring. The S/G pressure L
can be used to correlate to Tc there fore ans'ver B is also correct. Southern California Edison believes there are two correct answers to this question.
Accept answers A & B i
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- NUCLEAR ORGAN!2ATICN l
UNITS 2 AND 3 EMERGENCY CPERATING INSTRUCTICN
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REVISION 15 5023-12-7
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ATTACHMENT 2 PAGE 30 0F 122 LOSS OF FORCED CIRCULATION / LOSS OF OFFSIT
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FLOATING STEPS ACTION / EXPECTED RES_PONSE RESPONSE NOT OBTAINED l
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Low flow during Natural Circulation slows RCS response to temperature changes.
L0cc transit time rises to between 5 minutes and 10 minutes.
FS-2 MONITOR Natural Circulation Established:
CHECK all RCPs - stopped.
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GO TO FS-4, MONITOR RCP Operating a.
Limits.
b.
CHECKatleastoneS/G b.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1) SBCS - operating
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OR INITIATE S023-12-9, Attachment 8
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REC 0VERY - HEAT REMOVAL.
ADV - operating.
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AND 2)
Feedwater - available.
CHECK operating loop AT - less c.
than 58'F.
IF any criteria c through g NOT o
satisfied,
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CHECK Tc and Ts - NOT rising.
THEN CHECK Reactor Vessel level e.
(Plenum) - greater than or MAXIMIZE S/G level - less than
equal to 100M 80% NR.
QSPOS page 622 RAISEavailableS/Gsteaming
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i, rate.
ii CFMS page 312 Attachment 4.
RAISE Core Exit Saturation Margin
f.
CHECK operating loop Ta and REP
- greater than 20'F:
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CET - within 16*F:
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QSPDS page 611
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CFMS page 311.
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ATTACHMENT 2 PAGE 4 0F 29
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- NUCLEAR CRGANIZATICN
UNITS 2 AND 3 EMERGENCY OPERATING INSTRUCTICN 5023-12-
-' *
REVISION 15
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ATTACHMENT 2 PAGE 31 0F 122
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LOSS OF FORCED CIRCULATION / LOSS OF OFFSITE POWE FLOATING STEPS
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ACTION / EXPECTED RESPONSE RESPONSE NOT O8TAINED FS-2 MONITOR: Natural Circulation Established:
(Continued)
CHECK Core Exit Saturation-Margin - greater than 20*F:
IF any criteria c through g NOT o
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satisfied; J
QSPOS page 611 THEN CFMS page 311.
MAXIMIZE S/G level - less'than
.
80% NR.
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RAISE available S/G steaming-
rate.
RAISE Core' Exit Saturation Margin
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- greater than 20*F:
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CFMS page 311.
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COMMENT #9 SRO Examination Question 43 (RO46)
Answer B is correct based on the strictest interpretation ofimmediately before and afict a trip. Immediately before the trip, S/G level exceeded 85% causing a High Level Override, HLO, signal to be applied to the main and bypass food regulating valves causing both to close. The valves both st9y closed until level decreases below 85% at which time the HLO signal clears and the valves go to either the Reactor Tripped override, RTO, position or to the position set by the demand from the feed water regulating control system. With a reactor trip, an RTO signal is sent which as soon as the HLO condition clean seconds after the trip due to normal shrink of water 1svels, the RTO signal is applied and the bypass valve opens to 50%, answer C. Southern California Edison believes there are two correct answers to this question.
Accept answers B & C i
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- SRO EXAMINATION QUESTION 51 c
'All four answers contain the statement "over current reset".. There is no over current relay in the circuitry for the 50.54X cross-tie for
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the EDGs.1 Southern California Edison believes there is no correct answer for this question and the question should be deleted from
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I r
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COMMENT #11 SRO Examination Question 60 (RO7)
Procedure SO23 13-13, Misaligned CEA, has a note after step 1 stating " Initial and stabilized reactor power levels are required for the subsequent shutdown margin calculation." This is the basis for answer A being correct.
In attachment 3 of the same procedure, there is anotl'er caution that states: "Within 15 minutes of misalignment
'
discovery, a power reduction may be required.. " laitial and final stabilized power levels are used to determine the further power reduction requirements within the first hour to actintain compliance with the acceptable operating region in technical specification LCO 5.1.5 and LCS 3.1.105. This is why answer C is correct.
Southern California Edison beheves the there are two correct answers to this question.
Accept A & C i
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NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTION S023-13-13
UNITS 2 AND 3 REVISION 4 TCN 4-1 PAGE 5 0F 24
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MISALIGNED CONTROL ELEMENT ASSEMBLY
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OPERATOR ACTIONS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 3 COMMENCE plant load reduction:
a,
'If Reactor power is < 50%,
THEN GO TO Step 3c:
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fddl.UEff Within fifteen minutes of misalignment discovery a power reduction may be required. The negative reactivity of the misaligned CEA is censidered part of the required power reduction.
Failure to maintain Reactor Power in the Region of Acceptable Operation is a violation of Tech. Spec. LCO 3.1.5~and LCS 3.1.105.
NOTE: The power reduction shall be in accordance with the applicable LCS 3.1.105 Figure.
The boration flowrate shall be sufficient to achieve the target power level within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 45 minutes of the rod drop time.
b.
INITIATE required RX power reduction to maintain RX power in the Region of Acceptable Operation per the applicable LCS 3.1.105 figure.
1) LOWER Turbine Generator load using CVOL wlfle maintaining Tcold within the Operating Band per Attachment 1.
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COMMENT # 12
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,.SRO EXAMIN tT'O, G " *1 ""> "
Both answcrs "A" & "B" will increau i ' ine.ts for operation of the reactor coolant pumps. Southern California Edison believes there are two correct answers to this question.
Accept answers A & B (
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h0 CLEAR OAGAi!IAT10N EMERGENCY OPERATING lh5tRUCTION 5023-12-3 LN(T5 2 AND 3 21:15101 15 PAGE 141 0F 163 ATTACHMENT 14 LCSS Or COOLANT ACCIDEMI PO$T ACCIDENT PRE 55URI/TD4PERATORE LIMIT 5 2500 (2380 PSIA) MA UMUM OPERATI NAL PRESSUR
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100*F/1;R j
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i LOWEST SERVICE 200*F
, TEMPERATURE SATURATION
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1500
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SDC ENTRY CONblTIONS ie
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130 200 300 400 500 600 700 800 RCS TEMPERATURE (*F)
l NOTE 1 THis CLRVE 15 IM EFFECT ANY TIME AN OMCONTROLLED COOLDOWN TO RC5 Tc LESS T4AN 500'F HAS OCCURRED, i
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53-15.W61 ATTACHMENT 14 PAGE 1 CF 1
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l COMMENT #13
i SRO Examination 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'
of H2O or 30.3 psig(70' x 0.433 psi /ft) In the scenario provided the head of the VCT with the over pressure, will keep the check valves from the RWST closed. Southern California Edison believes there are no correct answers to this question and it should be deleted.
Delete this question.
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l COMMENT #14 SRO Examination Question 14 (RO78)
All pressurizer heaters receive a backup signal to turn off at a pressure of 2275 psia. This is true for the heaters that are in auto. The stem states that the heaters are in auto. With PT0100X failing high, the signal to the heaters j
will exceed the 2275 psia shutoff set point for the heaters in auto. The correct answer should be D.
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Change correct answer to D.
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NUCLEAR ORGANIZATION SYSTEM DESCRIPTION SD-SO23-360 UNITS 2 AND 3 REVISION 5 Page 168 of 205 FIGURE lil-5 PRESSURIZER PRESSURE CONTROL SYSTEM BLdCK DIAGRAM
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i PT PT 100X 100Y 1E HEATER r-- PMS/CFMS PMS/CFMS -,1E HEATER
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ETEAM BYPASS SYSTEM STEAM BYPASS SYSTEM -
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CUMMENT# 15
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SR0 EXAMINATION QUESTION #75 (R0 79)
Question was based on old Tavg program. New program has normal pressurizer level cf 48% at 100% power. This is based on the l
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 l
reference, the expected level would be 48% and no additional charging pumps would be operating. Southern California Edison l
believes there are no correct answers for this question and the question should be deleted from the examination.
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Delete Question.
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NUCLEAR ORGANIZAT!ON OPERAT!NG 1NSTRUCTION 5023-3-1.10 UNITS 2 AND 3 REVISION 8 PAGE 30 0F 34
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ATTACHMENT 5 l
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PRESSURIZER LEVEL CONTROL PROGRAM
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COMMENT # 16 SRO Examination Question 77 DSS is not covered by Technical Specifications. There is no correct answer to the question as stated.
Delete the question.
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l COMMENT #17 SRO EXAMINATION QUESTION 84 (RO 85)
The trends given are inconclusive as to whether vacuum has stabilized at 3.3", if assumed vacuum is stable at 3.3 "
no further action will be required and answer B would be correct. If it is assumed vacuum will continue the current trend, the listed action of"D" could be taken to return the plant to a more stab!c condition. Southern California Edison believes "B" & "D" are correct answers.
Accept B & D.
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NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTION 5023-13-10 UNITS 2 AND 3 REVISION 2 PAGE 7 0F 13 i
LOSS OF CONDENSER VACUUM ACTION / EXPECTED RESPONSE RESPONSE NOT OB(AINED 4 Actions for loss of vacuum due to Condenser fouling:
. CAUTION During periods of heavy influx, rapid and aggressive action may be required in l&
, order to avoid a Unit trip. Power may need to be reduced in order to:
Bump and/or Stop Circulating Water Pumps on the Condenser quadrants with e
the highest differential pressures Maintain Condenser backpressure < 3.5" Hg
a. REDUCE Reactor power to 75% TO 85%.
b. BUMP Circulating Water Pump (s)
per direction of Shift Superintendent.
c. VERIFY backpressure < 3.5" Hg c.
1) REDUCE Reactor power and stable.
to s 65%.
2) STOP two Circulating Water Pumps on opposite ends of the Condenser.
3) INITIATE isolating stopped pumps per 5023-2-5, Attachment for Stopping a l
Circulating Nater Pump Due to Fouled Condenser
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Tubesheet/High 6P/ Debris K.
Removal.
4) IF not < 3.5" Hg and stable, THEN REDUCE Reactor Power as necessary to establish backpressure < 3.5" Hg and stable.
5) GO TO Step 5.
d.
EVALUATE stopping pumps based on Waterbox differential pressure and pump vibration.
e. GO TO Step 6.
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C0tWENT #19 N
SR0 EXAMINATION QUESTION #76~
(R0 95)
Both answerss A & B will cause a PTS event to occur if the operator fails to initiate release of steam from E088 S/G.
"A" is correct based on failing to steam the good S/G to establish a heat sink.
"B" is also correct in that HPSI throttle /stop criteria will NOT be met because of the unavailability of the S/G as stated in the stem (step FS-6 a.1 requires operating S/G with an ADV operating),
and continuing to inject water into the RCS will increase pressure also leading to PTS event.
56uthern California Edison believes there are two correct answers to this question.
Accept answers A & B i
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NUCLEAR ORGANIZAT10N EMERGENCY OPERATING INSTRUCTICN S023-12-5
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UNfTS 2 AND 3 REVIS10N 15 PAGE 45 0F 143 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT
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FLOATING STEPS ACTIONS / EXPECTED RESPONSE REgp0NSE NOT 08TAINE0 FS-6 CHECK HPSI Throttle /Stop
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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
Terniination Criteria.
J.
CHECK PZR Level J. INITIATE FS-22 ESTABLISH CVCS
- less than 80%.
Letdown Flow.
k.
RESET SIAS per 5023-3-2.22, ESFAS OPERATION.
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NUCLEAR ORGANIZATICN EMERGENCY CPERATING INSTRUCTION S023-12-5 UtilTS 2 AND 3 REVIS10N 15 PAGE la CF 113 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT
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FLOATING STEPS
ACTIONS / EXPECTED RESPONSE RESPONSE NOT CBTAINED FS-6 CHECK HPSI Throttle /Stop Criteria:
(Continued)
d.
CHECK Reactor Vessel Level c
IF any criteria of ste:s b tnrcugh c (Plenum) - greater than or NOT met, equal to 100%:
THEN QSPDS page 622 CFMS page 312 OPERATE Charging and HPSI systems
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Attachment S.
as necessary to maintain
{
Throttle / Step criteria
- satisfied.
THROTTLE Loop Injection Valves.
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ENSURE auxiliaries to SI pumps.:
a)
Electrical power to pumps and valves.
b)
Proper system alignment.
c)
CCW flew.
d)
HVAC.
e.
VERIFY RCS borated - greater e. MAINTAIN Emergency Boration than Technical Specification
- at least 40 GPM.
Shutdown Margin for T4E > 200"F per Operations Physics Summary Figure 2.3-1,
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RCS Cooldown - NOT in progress.
f.
THROTTLE OR STOP HPSI as required one train at a time.
g.
STOP charging pumps as required one at a time.
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ATTACHMENT 2 PAGE 14 0F 43
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NUCLEAR ORGAN!ZATION EMERGENCY OPERATING INSTRUCTION S023-12-5 UNITS 2 AND 3 REVIS10N 15 PAGE 43 0F 143 ATTACHMENT 2 EXCESS STEAM DEMAND EVENT FLOATING STEPS ACTIONS / EXPECTED RESPONSE RESPONSE NOT OBTAINED FS-6 CHECK HPSI Throttle /Stop
. Criteria:
a.
CHECKatleastoneS/G a.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
AND 1)
SBCS - operating INITIATE S023-12-9, Attachment 8, OR RECOVERY - HEAT REMOVAL.
-ADV - operating.
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AND 2)
Feedwater - available.
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CHECK PZR level o
IF any criteria of steps b through d NOT met.
- greater than 30%
THEN AND
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OPERATE Charging and HPSI systems
.
- NOT lowering.
as necessary to maintain g
Throttle /Stop criteria c.
CHECK Core Exit Saturation
- satisfied.
Margin - greater than 20*F:
THROTTLE Loop Injection Valves.
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CFMS page 311.
ENSURE auxiliaries to SI pumps:
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a)
Electrical power to pumps and
valves, b)
Proper system alignment.
c)
CCW flow.
d)
HVAC.
Q.
ATTACHMENT 2 PAGE 13 0F 43
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COMMLYT # 20 SRO EXAMINATION QUEST 10N 97 (RO 96)
Answer "B" is correct based on the information given. Howner answer "A" is also correct based on procedure 5023-12-7, Safety Function Status Checks, which requires rabcooling > 20F or you are directed to the Functional Recovery for not meeting natural circulation. Southern California Edison believes there are two correct answers to this question.
Accept answers A & B.
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NUCLEAR ORGANIZATICN EMERGENCY OPERAT8NG INSTRUCT!CN S023-12-7
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UNITS 2 AND 3 REVISION 15 PAGE 30 CF 122 ATTACHMENT 2 LOSS. OF FORCED CIRCULATION / LOSS OF 0FFSITE PCWER
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FLOATING STEPS ACTION /EXPCCTED RESPONSE RESPONSE NOT OBTAINED NOTE:
Low flow during Natural Circulation slows RCS response to temperature changes.
Loop transit time rises to between 5 minutes and 10 minutes.
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l FS-2 MONITOR Natural Circulation Established:
a.
CHECK all RCPs - stopped.
a.
GO TO FS-4, MONITOR RCP Operating Limits.
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b.
CHECKatleastoneS/G b.
GO TO S023-12-9, FUNCTIONAL RECOVERY operating:
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AND 1)
SBCS - operating INITIATE 5023-12-9, Attachment 8, i
OR RECOVERY - HEAT REMOVAL.
ADV - operating.
AND
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Feedwater - available.
t c.
CHECK operating loop AT
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IF any criteria c through g NOT
{
than 58'F.
satisfied, d.
CHECK Tc and Tu - NOT rising.
THEN e.
CHECK Reactor Vessel Level MAXIMIZES /Glevel-lessthan
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(Plenum) - greater than or 80% NR.
equal to 1004:
RAISEavailableS/Gsteaming
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QSPOS page 622 rate.
CFMS page 312 i
Attachment 4.
RAISE Core Exit Saturation Margin
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f.
CHECK operating loop Ts and REP.
CET - within 16'F:
QSPOS page 611 CFMS page 311.
QSPOS page 611 L
CFMS page 311.
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NUCLEAR ORGANIZATION EMERGENCY CPERAT!NG 1NSTRUCT10N S023-12-7 l*
UN!TS 2'AND 3 REVISION 15 ATTACHMENT 2 PAGE 31 0F 122 LO 5 0F FORCED CIRCULATION / LOSS OF 0FFSITE POWER FLOATING STEPS ACTION / EXPECTED RESPONSE
. RESPONSE NOT OBTAINED FS-2 MONITOR Natural Circulation Established:
(Continued)
g.
CHECK Core' Exit Saturation o
IF any ' criteria c through g NOT-Margin - greater than 20'F:
satisfied, OSPOS page 611-THEN CFMS page 311.
MAXIM 1ZE S/G level - less than
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804 NR.
i RAISE available S/G steaming a
rate.
RAISE Core Exit Saturation Margin
- greater than 20*F:
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ATTACHMENT 2 PAGE 5 0F 29
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