ML16342E124

NRC Operator Licensing Exam Repts 50-275/98-301 & 50-323/98-301 for Tests Administered on 980511-15

ML16342E124
Person / Time
Site:	Diablo Canyon
Issue date:	06/10/1998
From:	Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML16342E125	List: IR 05000275/1998301 ML16342E124
References
50-275-98-301, 50-323-98-301, NUDOCS 9806150330
Download: ML16342E124 (60)

See also: IR 05000275/1998301

Text

{{#Wiki_filter:ENCLOSURE U.S. NUCLEAR REGULATORYCOMMISSION REGION IV Docket Nos.: License Nos.: Report No.: Licensee: Facility: Location: Dates: Inspector(s): Approved By: 50-275; 50-323 DPR-80; DPR-82 50-275/98-301; 50-323/98-301 Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant, Units 1 and 2 7Nmiles NWofAvilaBeach Avila Beach, California May 11-15, 1998 T. O. McKernon, Chief Examiner, Operations Branch S. L. McCrory, Examiner, Operations Branch H. F. Bundy, Examiner, Operations Branch M. E. Murphy, Examiner, Operations Branch T. R. Meadows, Examiner, Operations Branch R. E. Lantz, Examiner, Operations Branch John L. Pellet, Chief, Operations Branch Division of, Reactor Safety ATTACHMENTS: Attachment 1: Supplemental Information Attachment 2: Facility Initial License Written Examination Comments Attachment 3: Final Written Examinations and Answer Keys (RO and SRO)

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-2- XEC TIVE UMMA Y Diablo Canyon Nuclear Power Plant, Units 1 and 2 NRC Inspection Report 50-275/98-301; 50-323/98-301 NRC examiners evaluated the competency of 6 reactor operator and 9 senior operator applicants for issuance of operating licenses at the Diablo Canyon Nuclear Power Plant facility. The licensee developed the initial license examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. NRC examiners reviewed, approved, and administered the examinations. The initial written examinations were administered to all 15 applicants on May 8, 1998, by facility proctors in accordance with instructions provided by the chief examiner. The NRC examiners administered the operating tests on May 11-15, 1998. ~eye Allapplicants (9 senior operators and 6 operators) passed their license examinations (Sections 04.1). I Overall good licensed operator applicant performance was observed during the initial license examinations. Effective communications and good peer checks were observed in the dynamic simulator scenarios (Section 04.2). Housekeeping and condition of external panels observed coincident with plant walkthroughs was good (Section F8.1).

-3- Re D ails Sum fPI n Sau Both units operated at essentially 100 percent power for the duration of this inspection. 04 Operator Knowledge and Performance 04.1 I iiaIWri e E ami a ion a. Ins e 'on o b. The licensee developed the written examination with dedicated training instructors on the security agreement and used facilitytraining and operations staff on security agreement to validate the examination. The licensee proctored the administration of the written examination to the license applicants on May 8, 1998. The licensee staff proposed grading for the written examinations, analyzed the proposed results, and presented their evaluation and resultant draft comments for examination revision to the chief examiner on May 15, 1998. The licensee formally transmitted the examination comments to the NRC on May 19, 1998. bs rvao sa d n in The minimum passing score was 80 percent. Allapplicants (9 senior operators and 6 operators) passed with scores ranging from 80.6 to 92.8 percent for the reactor operators, with an average score of 86.5 percent, and scores ranging from 83.6 to 92.8 percent for the senior reactor operators, with an average of 88 percent. The NRC speciTically notified the licensee learning services representative of two individuals, who passed with scores of 80.6 and 81.6 percent, for consideration of additional enhancement or remedial training. The grades reflected the results after examination changes recommended by the licensee as a result of post-examination question analysis were incorporated. The licensee provided comments and the appropriate references for four questions. Two questions were recommended for deletion: RO Question 9 (SRO Question 12) because no correct answer was listed; RO Question 72 (SRO Question 75) because three of the choices were correct answers. One question was revised to accept two correct answers; RO Question 11(SRO Question 14.) The fourth question was revised to accept answer B in lieu of A as the correct answer; RO Question 47 (SRO Question 54.) The chief examiner reviewed and accepted these recommendations based on the technical merits of each recommendation and the material references provided by the licensee. The licensee's submitted examination comments are included as Attachment 2 to this inspection report.

In addition to the above, while proctoring the written examination the licensee revised a technical error in the stem of RO Question 26 (SRO Question 32) by changing the term "Mode 5" to "Mode 6." Allapplicants taking the examination were orally advised of the change. No other questions asked during the examination resulted in changes to the examination. The licensee's post-examination test analysis indicated that more than half of the applicants missed the same 13 questions. Two of the questions were procedures related; 5 questions were plant systems related (all different systems); and 6 questions were generic in subject. The chief examiner determined that there were no significant interrelationships to indicate generic weaknesses in knowledge or ability. The licensee stated that all missed questions would be reviewed with the individuals as part of the training department's remediation prior to assuming shift watch. c.. ~Cclu~io s Allapplicants passed the written examination. in coe The examination team administered the various portions of the operating test to the 15 applicants between May 11-15, 1998. Each applicant participated in two or three dynamic simulator scenarios and received a walkthrough test, which consisted of ten system job performance tasks (except for the 1 reactor operator upgrading his license to senior operator, who performed five tasks), together with two followup questions for each system. Additionally, each applicant was tested on five subjects in four administrative areas by answering two questions or performing one task for each subject. b. b rvations and Fi din The examiners observed effective communications and good peer checks of control board activities during the dynamic simulator scenarios. Good plant and component awareness was observed during the walkthrough portion of the operating tests. The crews utilized effective two-way communications. Applicants displayed good knowledge of the location and operation of local plant components. With one exception, the applicants responded accurately to the walkthrough followup questions, which indicated a depth of associated system knowledge. The one exception involved the system effect and failure position of Letdown System Valve PCV-135 upon loss of its air supply. More than 50 percent of the applicants answered this question incorrectly, which indicated an isolated knowledge weakness of failure positions for air operated valves. The question was required to be answered without the use of references. The licensee stated that all applicants would have this information reviewed with them prior to assuming shift duties.

-5- Allapplicants passed all sections of the operating test. Effective communications and good peer checks were observed during the dynamic simulator scenarios. Overall, good operator performance was observed during the examination. 05 Operator Training and Qualification 05.1 Inii Li n in Examina ' v I The licensee developed the initial licensing examination'n accordance with NUREG-1021. 0 .1.1 ~i n e The licensee submitted the initial examination outline on February 9, 1998. The examiners reviewed the submittal against the requirements of NUREG-1021. b. Io a d The chief examiner provided only minor enhancement suggestions related to a balanced mix of malfunctions and power maneuvers in the dynamic simulator scenarios. Some other minor enhancements were suggested to the scenarios to ensure that senior operator applicants were evaluated in exercising the facility's technical specification. c. ~CincniL ions The licensee's examination outline was acceptable. However, minor enhancements suggested by the chief examiner were incorporated. 05.1.2 Exa ' io k i n The licensee submitted the initial examination package on March 11, 1998. The chief examiner reviewed the submittal against the requirements of NUREG-1 021. b. Obse a n n Fin in The licensee submitted 126 draft written examination questions, of which 74 were designated to be common to both the reactor operator and senior operator examinations. The chief examiner provided comments or questions on 20 reactor operator examination items and 21 senior operator examination items; 4 questions were common to both examinations. In resolving these comments, the licensee revised or replaced 11 questions. The remaining questions were found to be satisfactory. The majority of

-6- the chief examiner's comments were enhancements and not considered substantive. The examinations were acceptable for administration as submitted. As discussed in Section 04.1, following post-examination review, credit was given for 2 questions on both the reactor operator and senior operator examinations. Further, credit for multiple answers on 1 question was allowed on both examinations and on another question the correct answer was changed. The failure to make these changes would not have invalidated the examinations or degrade their discriminatory value. The examinations were considered acceptable for administration as submitted. The licensee submitted three sets of operating tests, which included a total of 30 job performance measures, 3 administrative tests, 8 scenarios, and 2 backup scenarios. The submitted scenarios were considered acceptable for administration. However, some enhancement suggestions were incorporated during the validation week to add better balance to the scenarios and ensure evaluated senior operator applicants exercised the facility's technical specification'0 The submitted facilitywalkthrough subsection of the examination discriminated at the required level. Some enhancement suggestions were incorporated to better facilitate the test administration and focus the initiating cues on the desired tasks. Some of the job performance followup questions were revised or replaced to avoid @5'irect lookup type questions and to better define the acceptance criteria for questions requiring multiple answers. While some enhancements and revisions to the operating tests were made, the number of revisions were few and the changes did not impact administering the examination. r~ of Final revisions to the examination were completed during the preparation week and prior to the examination week. The licensee's training department and operations department provided excellent support during the development and administration of the examination. oncl sion The licensee submitted an acceptable examination for administration to operator license applicants. Only minor enhancements were made to the submittal before and during the onsite preparation week. The facility provided excellent support during the development and administration of the examination. im I ion Faci'erfor a ce The examiners observed simulator performance with regard to fidelity during the examination validation and administration. The simulation facility supported the examination administration well. No problems were observed.

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-7- IV. Plant Su o F8 Nliscellaneous Fire Protection Issues 88.1 ~G The examiners observed good plant housekeeping and condition of external panels coincident with the inplant walkthrough portion of the examination. The facilitywas reasonably clean, well lighted, and the floors were clear and free of debris. V a a e en Nice in s X1 Exit Nleeting Summary The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on May 15, 1998. The licensee acknowledged the findings presented. The licensee did not identify as proprietary any information or materials examined during the inspection.

A ACHMENT 1 SUPPLEMENTAL INFORMATION PARTIALLIST OF PERSONS CONTACTED S. Kettlesen, Supervisor, Licensing G. Goelzer, Acting Operations Director T. Blake, Learning Services Director D. Burns, Training Instructor J. Buckley, Training Instructor R. Jett, Training Leader J. Haynes, Training Leader J. Molden, Operations Manager B. Garrett, Operations Director D. Christensen, Engineer B. McRory, Instructor J. Becerra, Instructor R. Burnside, Engineer

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ATTA HMENT2 Facility Initial License Written Examination Comments Question ¹ Question Recommendation Justification RO SRO 12 Given the following: ~ Operators are performing EOP ECA-2.1, "Uncontrolled Depressurization ofAllSteam Generators." ~ Cooldown rate is 200'F per hour. ~ Steam Generator levels range from 1% to 3% NR level. No correct answer. Delete question from both exams. Per ECA 2.1 Background document AFW flowis reduced to 25 gpm per Steam Generator to prevent "Dryout." Keeping the tubes covered implies 4% water level in the Faulted Steam Generator (see attached reference). WHICH ONE (1) of the following is the reason for maintaining a MINIMUMof 25 gpm AFWflowto each steam generator in this condition? a. To provide AFW pump minimum recirc flow requirements b. To minimize the potential water hammer by maintaining a minimum flowthrough the feed ring J tubes. c. To keep the S/G tubes covered while ensuring the minimum detectable feed flowis maintained. d. To conserve CST inventory until the end of the blowdown phase. ANSWER: c. Page1

Written Examination Comments (cont'd) Question ¹ Question Recommendation Justification RO SRO 14 Given the following: At 0803, the following conditions exist: ~ Reactor power = 56% ~ Condenser pressure = 4.0 inches Hg. At 0806, it is reported that the following conditions exist: Accept A 8 B as correct answers for both exams. Per the RNO column ofAP -7 "Degraded Condenser" both answer A and distracter B can be done, Turbine load is reduced "as necessary" since Condenser pressure is still in the Acceptable Operating Range, per Foldout page ofAP-7 and the condenser Vacuum pump is placed inservice (see attached reference). ~ Reactor power = 50% ~ Condenser pressure = 5.0 inches Hg. WHICH ONE (1) of the following describes the MINIMUMaction(s) required to be taken at 0806? a. Continue to reduce turbine load. b. Stabilize turbine load and start the vacuum pump. c. Immediately trip the turbine. d. Immediately trip the reactor. ANSWER: a. Page 2

Written EXaminatiOn COmmentS (cont"d) Question ¹ Question Recommendation Justification RO 47 SRO 54 Given the following: ~ The unit is at 30% power. ~ RCP 1-2 trips. ~ NO operator action is taken. WHICH ONE (1) of the following describes the INITIALunit response to the RCP trip? a. A reactor trip willNOT occur and S/G 1-2 water level will INCREASE. b. A reactor trip willNOT occur and S/G 1-2 water level will DECREASE. Change correct answer to B instead ofAfor both exams. When a RCP trips, loop flowgoes down in that loop which affects heat transfer into the Steam Generator (S/G). Less heat into the S/G means lower boiling rate, so steam pressure drops, until the Main Steam Line Isolation Check Valve doses. This drop in boiling rate causes smaller steam bubble formation and S/G level drops until steam flowalso drops. The initial response is a decrease in S/G water level, not an increase. Additionally reverse flowwillnot occur until the RCP has coasted down (see attached reference). c. A reactor trip WILLoccur and S/G 1-2 water level will INCREASE. d. A reactor trip WILLoccur and S/G 1-2 water level will DECREASE. ANSWER: a. Page 3

Written EXaminatiOn COmmentS (cont'd) Question ff Question Recommendation JustiTication RO 72 SRO 75 Given the following: ~ Unit is at 100% power. ~ Reactor protection system (RPS) testing is in progress. ~ Train "8" reactor trip breaker is CLOSED. ~ Train "B"bypass breaker is CLOSED. ~ Train "A"reactor trip breaker is OPEN. ~ Train "A"bypass breaker is OPEN. Multiple correct answers (3). Delete question from both exams. Since Bypass breaker A does not open no P4 train A signal is generated. This makes distractors A and D also correct answers (see attached reference). WHICH ONE (1) of the following is the system response following a spurious reactor trip signal and Bypass Breaker Train "B"fails to open? a. A Feedwater Isolation Signal willNOT be initiated by Train "A." b. Ifan Sl occurs and the signal is RESET, an automatic reinitiation of Sl would NOT be prevented. c. The Turbine Generator remains on line and must be manually tripped. d. Condenser steam dumps receive an open signaI, but do NOT arm. ANSWER: b. Page 4

Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56 Avita Beach, CA 93424 805/5454000 Robert P. Powers Vice President-Diablo Canyon Operations and Plant Manager May 20, 1998 PGKE Letter DCL-98-074 Thomas O. McKernon, Chief Examiner U.S. Nuclear Regulatory Commission, Region IV 611 Ryan Plaza Dr., Suite 400 Arlington, TX 76011-8064 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant Units 1 and 2 NRC License Written Examinatidn Formal Comments Dear Mr. McKerno: In accordance with NUREG 1021, Interim Revision 8, PGRE is providing the enclosed formal comments on the written examination administered to Diablo Canyon Power Plant license candidates on May 8, 1998. PG8 E appreciates the NRC staff efforts during the entire examination and review cycle. lfyou have any questions, please contact Roger Jett, Operations Training Supervisor, at (805) 545-3439. Sincerely, Robert P. Powers Enclosures cc: Joseph M. Haynes Roger L. Jett David L. Proulx TLH/1753

e losure PG8 E Letter 98-074 May 1998 Diablo Canyon Written Examination Formal Comments Question 0 Question Recommendation Justification RO SRO 12 Given the following: ~ Operators are performing EOP ECA-2.1, 'Uncontrolled Depressurization ofAllSteam Generators.' Cooldown rate is 200'F per hour. ~ Steam Generator levels range from 1% to 3% NR level. WHICH ONE (1) of the following is the reason for maintaining a MINIMUMof 25 gpm AFW liowto each steam generator in this condition? No correct answer. Delete question fromboth exams. Per EGA 2,1 Background document AFW liowis reduced to 25 gpm per Steam Generator to prevent "Dryout." Keeping the tubes covered implies 4% water level in the Faulted Steam Generator (see attached reference). a. To provide AFW pump minimum recirc flow requirements. b. To minimize the potential water hammer by maintaining a minimum liowthrough the feed ring J tubes. c. To keep the S/G tubes covered while ensuring the minimum detectable feed flowis maintained. d. To conserve CST inventory until the end ofthe blowdown phase. ANSWER: c.

4 nclosure PG&E Letter L 98-074 May 1998 Diablo Canyon Written Examination Formal Comments Question ¹ Question Recommendation Justification RO SRO "4 Given the following: At0803, the following conditions exist: ~ Reactor power = 56% ~ Condenser pressure = 4.0 inches Hg. At0806, it is reported that the following conditions exist: ~ Reactor power = 50% ~ Condenser pressure = 5.0 inches Hg. Accept A6 S as correct answers for both exams. Per the RNO column ofAP -7'Degraded Condenser" both answer A and distracter S can be done, Turbine load is reduced 'as necessary'ince Condenser pressure is still in the Acceptable Operating Range, per Foldout page ofAP-7 and the condenser Vacuum pump is placed in service (see attached reference). WHICH ONE (1) of the following describes the MINIMUMaction(s) required to be taken at 0806'? a. Continue to reduce turbine load. b. Stabilize turbine load and start the vacuum pump. c. Immediately trip the turbine. d. Immediately trip the reactor. ANSWER: a.

closure PG8 E Letter 98-074 May 1998 Diablo Canyon Written Examination Formal Comments Question 8 Question Recommendation Justification RO 47 SRO Given the following: ~ The unit is at 30% power. ~ RCP 1-2 trips. ~ NO operator action is taken. WHICH ONE (1) of the followingdescribes the INITIALunit response to the RCP tripf a. Areactor trip willNOT occur and S/G 1-2 water level viill INCREASE. b. Areactor fripwjIINOT occur and S/G 1-2wa'ter level will DECREASE. Change correct answer to B instead ofAfor both exams. When a RCP trips, loop flowgoes down in that loop which affects heat transfer into the Steam Generator (S/G). Less heat into the S/G means lower boiling rate, so steam pressure drops, until the Main Steam Line Isolation Check Valve doses. This drop in boiling rate causes smaller steam bubble formation and S/G level drops until steam flowalso drops. The initial response is a decrease in S/G water level, not an increase. Additionally reverse flowwillnot occur until the RCP has coasted down (see attached reference). c. Areactor trip WILLoccur and S/G 1-2 water level viill INCREASE. d. Areactor trip WILLoccur and S/G 1-2 water level will DECREASE. ANSWER: a.

closure PG&E Letter 98-074 May 1998 Diablo Canyon Written Examination Formal Comments Question ¹ RO SRO 72 75 Question Given the following: ~ Unit is at 100% power. ~ Reactor protection system (RPS) testing is in progress. ~ Train 'B'eactor trip breaker is CLOSED. ~ Train 'B'ypass breaker is CLOSED. ~ Train 'A'eactor trip breaker is OPEN. ~ Train 'A'ypass breaker is OPEN. Recommendation Multiple correct answers (3). Delete question from both exams. Justification Since Bypass breaker Adoes not open no P-4 train Asignal is generated. This makes distracters Aand D also correct answers (see attached reference). WHICH ONE (1) ofthe followingis the system response following a spurious reactor trip signal and Bypass Breaker Train 'B'ails to open' a. AFeedwater Isolation Signal willNOT be initiated by Train "A." b. Ifan Sl occurs and the signal is RESET, an automatic reinitiation of Sl would NOT be prevented. c. The Turbine Generator remains on tine and must be manually tripped. d. Condenser steam dumps receive an open signal, but do NOT arm. ANSWER: b.

Attachment pG&E Letter DCL 98-074 Reference Nlaterial

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STEP 0 RIPTION TABLE FOR EOP ECA-2.1 STEP ' STEP: CONTROL Feedf low To Hinimize RCS Cooldown To control feedf low to minimize the effects of the cooldown due to the secondary depressurization and to subsequently control the transient. BASIS: i. Depending upon the size of the effective break areas for the S/Gs, the cooldown rate experienced after reactor trip could exceed 100 F/hr. A reduction of feedflow to the S/Gs three primary effects: 1) 2) 3) To minimize any additional cooldown resulting from the addition of feedwater, To prevent S/G tube dryout by maintaining a minimum feedflow to the S/Gs, and To minimize the water inventory in the S/Gs that eventually is the source of additional steam flow to containment or the environment. The 25 gpm value is representative of a minimum measurable feedflow to a S/G. Values depend upon flow instrumentation and the sensitivity of the controls on the feedflow. As steam flow rate drops, the feedflow will eventually increase the S/G inventory. Feedflow is controlled to maintain S/G NR level less than 44K to prevent overfeeding the S/Gs. ition, as- S/G pressur e and steam flow rate drop, RCS hot leg temperatures will ize and start increasing. The operator controls feedflow or dumps steam to st ilize the RCS hot leg temperature. The operator controls feedflow or dumps injection flow to establish conditions for SI termination and minimizes thermal stresses that may be generated. ACTIONS: o Determine if NR level in all S/G greater than 4%[2(5j o Determine if cooldown rate in RCS cold legs is less than 100'F/hr o Determine if NR level in all S/Gs is less than 44K o Determine if RCS hot leg temperatures are stable or decreasing o Decrease feedflow to 25 gpm to each S/G o Control feedflow to maintain NR level less than 44K in all S/Gs o Control feedflow or dump steam to stabilize RCS hot leg temperatures INSTRUHENTATION: o RCS cold leg temperature indication o RCS hot leg temperature indication o S/G NR level indication o Total feedflow indication o Feedflow control valve position indication o earn dump valve. position indication o /G 10K steam dump valve position indication EOP ECA-2.1 N00026.EOP 22 Rev. 4

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PACIHC GAS AND COMPANY . DIABLOCANYONPOWER PLANT TITLE: DEGRADED CONDENSER NUMBER OP AP-7 REVISION 16 PAGE 3 OF 16 UNITS 1 AND2 SECTION A: LOSS OF CONDENSER VACUUM{Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED thtlrthth dt thtittP thdr thththtPththeththdrthdrthtlrrh rhthththdrththrhththtbtlt ththrhrhdr thdt thrhdt thrhtP thdrAdtththtlt thrhrhtP tlt tPttPrhththtPtlt thtPdr ththththtltatty ththrhththththth CAUTION: Ifboth Cps are lost, Attachment 4.1 should be implemented after EOP E-0 is performed. tulle tlrrh IlrtlttltIllthillQ IP IP tlllitlittlrrlrillrlrQ IlltltIP th Q IllIO rh th tltdr th Q IP + Q rh IltrlltP lg th litIlllhth th tltllrtP + Q IltlhQ th IltIh+ Ih lg tlrIlt+ tP tP th th lltth dr llrtllrk tlrdr Ih ur th tP th Q ilrIh Ih th 1. CHECK Condenser Parameters- NORMAL: a. Condenser Pressure LESS a. Trip the Turbine and GO TO EOP E-0 THAN7" Hg ABS {ifapplicable) CAUTION: Ifturbine power is reduced to less than 30%, condenser pressure requirements are more restrictive. Refer to Foldout Page. th dr dt th th thrh tltth tltth tlrth tP tltth thrh tltthrh th tP tlttltth tltth thrh th tlttP tdt tltth th4 th th tP tP tlttP th tlttP4 tP th tlttlrth4 dr th tltth tlttPt4 tP th th tlttlttltth th th 4 th th tltth4 tltth tP th th tP Condenser Pressure LESS THAN3.5" Hg ABS c. Condenser Pressure LESS THAN4" Hg ABS 2. CHECK SJAE After Condenser REDUCE Unit load as necessary to remain within the operating limitations of the Foldout Page. IF Condenser Pressure'xceeds the Foldout Page limitations, THEN Trip the Turbine and GO TO EOP E-O (if applicable) Condenser Pressure continues to INCREASE, THEN Place the Vacuum pump on the affected Unit per OP CW:I, Attachment 9.1. IF Investigate cause ofair inleakage AND valve in additional AirEjectors per OP C-6:I. NORMAL{LESS THAN 1S SCFM at 100% Power) 00066216.DOA 02 3 0130.0824

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09/10/96 Page 1 of 1 Condenser Pressure, in Hg Abs 10 FOLDOUT PAGE FOR P AP-7 Turbine Operating Limitations Trip Turbine Immediately Trip Turbine within 5 minutes Acceptable Operating Region 0 10 20 30 40 50 60 70 80 90 100 Turbine Load, % 00066216. DOA 02 Ol30.0824

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E rcp trip power O CD T Tfttp Rci I-1 wrcsloop(1) goop I F~ O 0 CI LUI- CI O o> I I III I I III I I II I II II I II I I I II III I I II I CD CV II I I II-r 0 20 40 80 TIME (Seconds) 100 120 140 rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

E rcp trip a ower qrcssgn(1) ~p I J4~ T~++~ FRom QC S 40 60 I IIII II I I I 80 I I III I J I I II I II I II 1 I I I II I I III I I I 100 I II IIL I II I I I I II I I I I I I II III I I 120 140 1I I I I I I I I I J TIME {Seconds) rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

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E rcp trip power C) CI I II I II I I II I JI I I II I II I I II J BSGNWL{{1),) gq I Wa Sc Law~ 1I IIIIII I I I II I I II II I I I' I I I I I LLJ)Ill pe.cea~ I TRiP Rcp I.I II I I I I I1 I II I II I I I I I I I II 1 0 20 40 60 80 TIME (Seconds) 100 120 140 rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

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E rcp trip a power oO C4 I II I I I II I PSGNS((1),) Loop I srr I~5~A I I I II I r 1 I C) CO C) C) I III I I III I II I I I IIIII I I III I I I I I I I II I II IJ I II I I II I II II I I I I II I II P h I I I I III I I I I III I II I I II II I I III I I I I I II I II I I II 0 20 40 60 80 TIME (Seconds) 100 120 140 rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

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E rcp trip a ower TRCSHLR((1),) L~P I TwV I I I I I I I I I I I I 4, I I I I I I I I III I I I 1 I 1 III I II I 20 40 60 TIME (Seconds) 80 100 120 140 rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

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E rcp trip power TRCSCLR((1),) L+oP 1- coco r II II I I I I Ir I I I I'1 I I I '1 I I I II II I I II II I I II II I I I I II I I I I II I I I I r I I I I I I I I II III I I I I II I I I I I I I '1 II I I I 1 I I I I I I II I I I I I I 20 40 60 TIME (Seconds) 80 100 120 140 rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

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E rcp trip power qrcssgn(2) 20 I 40 I 60 80 CO LA I C) C) I C) LA I C) CD CLC o LA I I I I I I I I I I IrI II II I I I I I I II I II II II J III I II I I II 1 I II I I II J C) C9 (0 LA CO II I I IIIII TIME (Seconds) rcptrip Date of Origin - DATE: 12May98 TIME: 14:14:05

REA TGR TRIP HREAI ERG BREAK"R DC CCNTRQL FQWER (l28 VDC) AUTO TRIP5 AND AUTO 51's <<e VDC) TRIP HREAKER5 ONLY CC RT VB RT 5I CC VH '5I CC 7 TRIP HLQCK ! O UNDER- VQLTAGE AUXIL1ARY COIL TRIP

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TR1F COIL UNDER- VQLTAGc. COIL TRIP5 BREAKER Ir ENERCiZED TRIP5 BREAi ER IF DE-c.NERLIZED TO RQD CONTROL ( I'TA

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BYA (TRAIN H) w I 0 i (;RAIN H) 1 HYH (TRAiN A) P-O'.=,A>V A = RIP " AND BYFA55 "A" QPEiV . =.A:N = = i RiF "5" AND HYPA55 "H" OF iV 0 TRAi~ p p Q $)QP~ 5-5-4C 4 "7ela 5ECT 4 EL=CT PR:NT5 REV C

Ste Dump Control Signa s Associated Inter7ocks: C-7A: Arming Signal to Groups I 8 II (in Tavg Mode). C-78: Arming Signal to Groups III 8 IV (in Tavg Mode). P-4(Train A): Arming Signal to Groups I 8 II (in Tavg Mode). Blocks Groups III 8 IV (in Tavg Mode). P-4(Train 8): Selects the Reactor Trip Controller (in Tavg Mode). Blocks the Load Rejection Controller (in Tavg Mode). Blocks Group IV (in Tavg Mode). C-9: Arming Signal to Groups I 8 II (in all Modes). P-12: Blocks all valves closed (may be bypassed for Group I valves during cooldown). 80 70 60 u 50 o 40 20 10 IV III ~ ~ CL (3 ID CL0 CLI- CL (9 CV D C 6) CL0 CL II-, I ~ II III II III I I II + I I CL I (g I I ~ I I I Q ', c -LQ I CL CL )I I I I (g , C C I g) ID ~ CL IO 0 CL CL I CL0 CLI- 100 75 C) C) 50 ~ 25 0 II $0 ------------- f30 20 10 15 20 Load Rejection (Auct Tavg - Tref) CV ' ~e

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CL~ 25 50 C) C) 25 ~ CL~ 10 15 20 Reac',or Trip (Auct Tavg - Tnl) C-2-2 25 ReV Ii

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