ML18038A761

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Forwards Addl Info Re Simulator Certification for Facility, Per 890803 Request.Schedule Extension Verbally Granted Until 890930
ML18038A761
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 09/29/1989
From: Terry C
NIAGARA MOHAWK POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NMP1L-0440, NMP1L-440, NUDOCS 8910060064
Download: ML18038A761 (397)
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pg CP~ ~TED IIQCRIBUTION DEMONS'GATI ON SYSI'EM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8910060064 DOC.DATE: 89/09/29 NOTARIZED: NO DOCKET I FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 05000220 AUTH. NAME AUTHOR AFFILIATION TERRY,C.D. Niagara Mohawk Power Corp.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards addi info re simulator certification for facility, per 890803 request.

DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED:LTR Submittal: General Distribution g ENCL $ SIZE: QG NOTES RECIPIENT . COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 LA 1 &0 PD1-1 PD 1 1 A SLOSSON,M 5 INTERNAL: ACRS 6 ~Z NRR/DEST/ADS 7E 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/RSB 8E 1 1 NRR/DOEA/TSB 11 1 1 NUDOCS-ABSTRACT 1 1 EN 1 0 OGC/HDS1 1 0 REG FILE 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 '1 NRC PDR NSIC 1 1 ENCLOSURES DUE TO SIZE F

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"* '.iQ'lH 7 NIAGAII U MQHAWK NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474.1511 September 29, 1989 NMP1L 0440 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Re: Nine Mile Point Unit 1 Docket No. 50-220 DPR-63 Gentlemen:

This letter is in reply to your letter dated August 3, 1989 regarding Simulator Certification for Nine Mile Point Unit l. Enclosure A includes answers to your questions and additional documents that you requested. On September 6, 1989, Ms. M.H. Slosson of your staff verbally granted a schedule extension until September 30, 1989.

Very truly yours, NIAGARA MOHAWK P WER CORPORATION C. D. Terry'ice President Nuclear Engineering and Licensing HS/mjd 7830G xc: Regional Administrator, Region I Hr. R. A. Capra, Director Ms. M. H. Slosson, Project Manager Hr. W. A. Cook, Resident Inspector Records Management

>ol 89i0060064 S90929 /

PDR ADOCK 05000220 P PDC

ENCLOSURE A Simulator Certification Unit 1 Response to Questions September 1989

..8910060064,"

NIAGARA MOHAWK POWER CORPORATION Nine Mile.-Point Unit 1 Docket No. 50-220 DPR-63

. Pi"-">i;((7830G)

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NIAGARA MOHAWK POWER CORPORATION RESPONSE TO NRC QUESTIONS ON UNIT 1 SIMULATOR CERTIFICATION Question 1:

On Form 474 you indicate that "Simulation Facility Performance Test Abstracts (are) attached." However, Attachment 1, "Performance Testing", is only a list of tests performed. Please provide the abstracts for these tests. These abstracts should include the following:

1. Date test was conducted.
2. Name and description of test (including relationship to Section 3.1.2, "Plant Malfunctions", of the Standard, if applicable).
3. Available options (e.g., range of rates or severity of which the simulation facility is capable).
4. Tested options (,i.e., what was actually tested for certification).
5. Initial conditions (for each tested option).
6. Final conditions (for each tested option).
7. Descriptions of baseline data used to determine fidelity to the reference plant.
8. Deficiencies found as a result of the test, corrective action planned and dates by which corrections will be made.
9. Exceptions taken to ANSI/ANS-3.5-1985 as a result of the test, with justification.

If the baseline data used was the judgment of a panel of experts, then documentation of their review, sufficient for a third party to evaluate the adequacy of the test(s) and results, should be'ncluded. This documentation may include such items as the make-up and qualifications of the panel and any differing professional opinions as to the outcome of the test(s).

~Res onse 1:

Enclosed are copies of the ANSI/ANS 3.5 Annual Reports for 1986 (initial),

1987 and 1988. The baseline data did not involve the judgement of a panel of experts. Also enclosed is the 1989 ANSI/ANS 3.5 Test Procedures (Test Methodology and Performance Tests). The detailed response is as follows:

1.1 Factory acceptance test March 1984 to May 1984.

On-site final acceptance test August 1984 to September 1984.

Initial ANSI 3.5 1986 test March 1986.

ANSI 3.5 1987 test May 1987.

ANSI 3 ' 1988 test May 1988.

(7830G)

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1.2 Tests are described in the following sections:

ANSI 3.5 1986, 1987, and 1988 Annual Reports, Section III.

ANSI 3.5 1986 Attachment 3 (lists the malfunction names).

1.3 The available options/malfunctions are listed in ANSI 3.5 1986 Annual Report, Attachment 3. Each malfunction has a specific cause and effect (refer to sample malfunction cause and effect ED09 enclosed). A complete copy of the malfunction cause and effect book will be supplied upon request.

1.4 Tested options are described in the ANSI 3.5 1986 Annual Report. All options/malfunctions (listed in Attachment 3) and other tests (listed in Section III) were tested for certification.

1.5 Each tested option/malfunction has a specific test procedure similar to the test procedure for malfunction ED09 enclosed. The complete malfunction test procedures occupies three books (two inch binders) which are available for your review on-site.

1.6 This response is the same as 1.5.

1.7 The baseline data was the simulator design data base which contained the following:

a. Reference plant controlled drawings (PID, Electrical, etc.).
b. Reference plant FSAR.
c. Selected GEK's.
d. Reference plant panel photographs.
e. Referent plant BOP logs.

f.. Reference plant P-1 edits and other OD printouts.

g. -

Reference plant Operating Procedures.

1.8 All deficiencies identified duri'ng the applicable tests were recorded and corrected. No Discrepancy Reports remain open from the original 474 submittal test procedures.

1.9 Exceptions are described in the following sections:

a. ANSI 3.5 1986 Annual Report, Section I.B.
b. ANSI 3.5 .1987 Annual Report, Section I.B.
c. ANSI 3.5 1988 Annual Report, Section I.B.

(7830G)

Question 2:

You have also indicated on Form 474 that a Simulation Facility Performance Testing Schedule (is) attached. However, Attachment 2, "Malfunction Testing",

appears to be only a list of 4-character malfunction designators, which are intended to indicate which malfunctions will be tested in each year. Please provide a schedule which more fully describes these tests.

R~es ense 2:

ANSI/ANS 3.5 1986 Annual Report, Attachment 3, provides a word description for each malfunction. They are tested using the original factory acceptance test and the malfunction cause and effect for that malfunction. 'A sample copy of malfunction ED09 (AC Power Board Electrical Fault - PB13, Section A) factory acceptance test and malfunction cause and effect is enclosed. Additional malfunction test procedures will be supplied upon request. All malfunctions have been tested during the past four year period. Malfunction CU05 was missed in 1987 and subsequently tested in 1989.

Question 3:

In Attachment 1, Item III indicates that an initial ANSI 3.5 test report was prepared. if It would be helpful you could provide this report as may it provide some of the information requested in Item of this enclosure.

1

~Res ense 3:

Enclosed please find the ANSI 3.5 Annual Reports for 1986 (initial), 1987, and 1988. Included, also, is the 1989 ANSI 3.5 Test Procedures.

Question 4:

Items III.3 and III.4 of Attachment 1 state that simulator performance was compared against the FSAR. ANSI/ANS-3.5-1985 states in Appendix A, Section A.3.3(2) that FSAR transients may be inappropriate for real-time dynamic simulation comparisons. Please provide justification for making such comparisons.

R~es onse 4:

The use of Nine Mile Point Unit FSAR for transient response was the only 1

transient data available to the testers, during initial construction. This issue was recognized by Niagara Mohawk in January 1989. The 1989 ANSI transient test results are being compared to General Electric transient data NED024708A Rev. 1 (December 1980). The 1989 ANSI Test Report will include the results of this analysis. Preliminary results indicate that the simulator response complies with ANSI 3.5 requirements.

(7830G)

Question 5:

ANSI/ANS 3.5-1985, Section 5.4.1, "Simulator Performance Testing", requires testing within the requirements of Section 4 (Performance Criteria). One of the criteria in Section 4 is that "administrative controls or other means shall be provided to alert the instructor when certain parameters approach values. indicative of events beyond the implemented model or known plant behavior." It appears that no testing was performed to ensure that this criterion was met. Performance test abstracts for such testing, or justification for exception to this requirement, should be provided.

~Res onse 5:

Review of the Factory acceptance, on-site final acceptance, and annual ANSI 3.5 testing shows no evidence of such testing. The simulator does have a light indication to warn the instructor when the operating limits are exceeded in accordance with ANSI/ANS 3.5 Section 4.3. A special test procedure is being written, will be conducted, and the resulting data included with the 1989 ANSI 3.5 test data, which is presently scheduled for December 1989

'7830G)

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UNIT 1 SIMULATOR 1989 ANSI/ANS 3.5 TEST PROCEDURES Submitted

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Gary n Da e Superv r Simulator Technology Reviewed R. T. Seifried D e Assistant Superintendent Training Approved A. D. Rivers Da Superintendent Training - Nuclear

n. Secondary plant heat balance data NOTE: Performance Criteria-
a. Simulator instrument error shall be NO greater than that of the related instrument in the Reference plant
b. Principal mass and energy balances shall be satisfied:
1) Net NSSS thermal power to generated electrical power
2) Feedwater flow to reactor thermal power
c. Simulator computed values for steady state, full power reference plant configuration operation, shall be stable and not vary more than + 2% of the initial values over a 60 minute period d, Simulator computed values of critical parameters shall agree within + 2% of the reference plant parameter at the specified power level C. Appendix "8" Transient tests:

(Sect~on 81.2.1)

1. Manual scram
2. Simultaneous trip of all feedwater pumps
3. Simultaneous closure of all Main Steam Isolation Valves
4. Main turbine trip (from max power NITHOUT immediate Rx. Scram)
5. Maximum rate power ramp (100% 75% 100%)

NOTE: Recric Flow Control in Master Manual Test parameters recorded vs. time with a resolution ( 0.5 seconds

a. Reactor power ( / neutron flux)
b. Total steam flow c, Total feedwater flow
d. Hide range reactor pressure
e. Narrow range reactor pressure
f. Hide range reactor water level
g. Narrow range reactor water level (feedwater control)
h. Generator gross electrical power Turbine steam flow
j. Total core flow (total recirc flow)

ADMIN/1 63

6. Simultaneous trip of all recirculation pumps
7. Single recirculation pump trip Test parameters recorded vs. time with a resolution.< 0.5 seconds
a. Reactor power (% neutron flux)
b. Total steam flow
c. Total feedwater flow
d. Narrow range reactor pressure
e. Narrow range reactor water le'vel (feedwater control)
f. Total core flow (no accurate indication with less than 5 loops)
g. Individual recirculation loop flows
8. Maximum size reactor coolant system rupture combined wi th loss of all offsite power
9. Maximum size unisolable main steam line rupture
10. Simultaneous closure of all Main Steam Isolation Valves combined with single stuck open safety/relief valve Test parameters recorded vs. time with a resolution ( 0.5 seconds
a. Reactor power (/ neutron flux)

Hide range pressure C. Hide range water level

d. Fuel zone water level
e. Total steam flow total feedwater flow/high pressure coolant injection flow
g. Torus water temperature Torus air temperature Torus pressure j Drywell temperature

~

k. Drywell pressure Total low pressure core spray flow NOTE: Transient test Performance Criteria:
1) He the same as reference plant startup test criteria where applicable.

3 ADMIN/163

2) The observable change in the monitored parameters correspond in direction to those expected from a best estimate analysis for the simulated transient and do not violate the physical laws of nature. Reference Document NED024708A Rev. 1 8/79 for transient data.
3) Not fail to cause an alarm or automatic action if the reference plant would have and not cause an alarm or automatic action if the reference plant would not have.

D. Mal function Tests (me ets requirements of ANSI A3.5 and Regulatory Guide 1.149)

l. AD04 Relief Valve Leaks
2. CS01 Core Spray Pump Trip CT02 Containment Spray RAN Water Pump Trip CU03 RWCU Reject FCV Fails Open
5. CU07 RWCU Low Pressure Control Valve Fails Open
6. CUll Coolant Leak Outside Primary Containment CW04 RBCLC Pump Trip CW08 Circulating Water intake Structure Icing DG02 Diesel Generator Trip
10. EC04 Emergency Cooling System Return Valve Fails Open ED01 Loss of Offsite 115KV Power
12. ED05 PB12 Electrical Fault 13 ED09 PB13 Section "A" Electrical Fault
14. ED13 PB14 Section "B" Electrical Fault
15. ED17 PBI5 Section "C" Electrical Fault
16. ED21 PB17 Section "B" Electrical Fault
17. ED25 Loss of Power to Instrument Control Bus 130 Normal and Alt
18. EG04 Main Generator Core Internal Heating
19. EG08 Generator Hydrogen Emergency Seal Oil Pump Failure
20. EG12 Power Grid Network Load Transient - Decrease
21. FP02 Electric Fire Pump Failure
22. FP06 Control Room Fire Detection Various Panels
23. FP10 Reactor Building Fire Detection
24. FW04 Shaft Driven Feedwater Pump 13 Failure
25. FW08 Feedwater Control Valve 11 Controller Fails Low
26. FW12 Feedwater Control Valve 13 Controller Fails Low ADMI N /163

D. Mal function Tests (Cont'd)

27. FN16 Feedwater Master Controller Fails as is
28. FN20 Condensate Recirc Valve (FCV 50-24) Fails Closed
29. FN24 Feedwater Control Valve Fails Closed (13A/13B)
30. FN28 HPCI Mode Failure to Initiate
31. IAOl Loss of Instrument Air
32. MC03 Hotwell Level Controllers in Auto Fail - High
33. MSOl Steam Line Rupture Outside Primary Containment
34. MS05 Turbine Steam Seal Regulator Fails Closed
35. MS09 Second State Reheater 112 Drain Tk Level Control Fail Low
36. NM02 SRM Channel Failure - Downscale
37. NM10 IRM Channel Failure - Upscale
38. NM18 IRM Channel Detector Stuck
39. NM25 LPRH Failure Upscale
40. NM36 Recirc Flow Converter Channel Failure Upscale
41. NM40 Recirc Flow Converter Failure - Comparator
42. GQ04 Off Gas Discharge to Stack Isolation Valve Fails Closed
43. PP01 Failure of Plant Process Computer
44. RD04 Control Rod Failure - Stuck
45. RD08 Control Rod Failure - RPIS
46. RD36 CRD Flow Control Valve Failure Closed
47. RD40 Reactor Manual Control System Timer Malfunction Settle
48. RM03 Area Radiation Monitor Drawer Upscale
49. RP04 Reactor Protection System Failure to Scram Automatic
50. RP08 Anticipated Transient Ni thout Scram (ATHS)
51. RR04 Recirculation Pump ll Control Signal Failure
52. RR08 Recirculation Pump 12 Seizure
53. RR12 Recirc Pump 13 Field Breaker Trip RR15 Recirc pump 14 Drive Breaker Trip
55. RR20 Recirc Pump 14 Incomplete Start Sequence
56. RR24 Recirc Pump 15 Control Signal Failure
57. RR27 Master Recirc Flow Controller Failure Low RR31 Reactor Vessel Pressure Recorder Failure - Low
59. RR35 Reactor Vessel Pressure Indicator Failure Upscale
60. RR39 Reactor Vessel Level Recorder Failure - Downscale ADMIN/1 63
61. RR43 Rx Vessel Level Indication (Control Sys) Fail - as is
62. RR47 Recirc Pump Discharge Valve Stem Separates From Gate
63. RR51 Rx Vessel Level Transmitter (RPS Input) Fails High
64. RR55 Rx Vessel Level Transmitter (Control Input) Fails Low
65. RR59 Rx Vessel Pressure Transmitter (RPS Input) Fails as is
66. RR63 Reactor Recirc Pump 12 Inner Seal Failure
67. RR67 Reactor Recirc Pump 15 Tachometer Fails - Oscillates
68. RR71 Reactor Safety Valve Inadvertently Opens
69. RX02 Increased Rod North
70. TC02 Turbine Governor Fails High
71. TC06 Electrical Pressure Regulator Fails - Oscillates
72. TC10 First Bypass Valve Sticks Open
73. TUOl Exhaust Hood Spray Valve Fails Closed
74. TU05 Main Turbine Bearing High Temperature NOTE: Performance Criteria-
a. Be the same as reference plant startup test criteria where applicable
b. The observable 'hange in the monitored parameters correspond in direction to those expected from a best estimate analysis for the simulated transient and do not violate the phys, ical laws of nature.
c. Not fail to cause an alarm or automatic action if the reference plant would have and not cause an alarm or automatic action if the reference plant would not have.
d. Respond in accordance wi th the malfunction cause and effect Appendix "A" Section A3.2 Steady State and Normal Operations NOTE: Using controlled operating pr'ocedures conduct all evolutions listed in ANSI 3.5 Section 3.1.1.
1. P'lant start-up to Hot Standby, ,Operation at Hot Standby, and Start-up from Hot Standby to Rated Power including Turbine and Generator Start-up and Load Changes (meets requirements of Section 3.1.1 (1), (2), (3), (5) and (6)]
a. Master Start-up Checkoff Form III
b. Master Systems Pre-Start-up Checkoff List Form IV
c. Plant Start-up Nl-OP-43 Sections "E" and "H.2.1" ADMIN/163
2. Plant Shutdown to Hot Standby, and cooldown to Cold Shutdown Conditions (meets requirements of Section 3.1.1 (8)
a. Plant Shutdown Nl-OP-43, Section "G"
b. Reactor Hot Standby Nl-OP-43, Section "H.2.2.2"
3. Reactor trip followed by three (3) Recirc Loop Reactor startup to 90'L power, Start-up of idle Recirc Loops, continuation to Rated Power, Shutdown of Recirc Pump at power, and Shutdown and Cooldown to Cold Shutdown Condition (meets requirements of 3.1.1 (4), and (7)
a. Special Operating Procedure No. 1 "Reactor Scram"
b. Short Pre-Start-up Check-Off-Systems Form I
c. Plant Start-up Nl-OP-43, Section "E"
d. Plant Shutdown Nl-OP-43, Section "G"
e. Nuclear Steam Supply System Nl-OP-l, Section "H.l, H.2 and H.3"
4. Surveillance test procedures (using Control Room installed ir.strumentation) [meets requirements of Section 3.1.1 (9) and (10)]. As designated by the Simulator Configuration Control Board" F. Appendix "A" Section A3.3 Transients
l. LER's with reference plant data, as selected by the Simulator Configuration Control Board ADMIN/1 63

NIAGARA HONAHK SIHULATION FACILITY PERFORHANCE TEST Simulation facility: NINE MILE POINT UNIT I Reference Plant: NINE MILE POINT UNIT 1 Performance Test: ANSI 3.5 A endix "A" Section A3.2 Stead State and Normal 2!

Initial conditions: As specified in each individual test section Data Collection:

A. Method:

l. Data for the 100/ steady state 60 minute run will be gathered on the critical parameters using a Simulator program called "ANSI SS60"
a. Data is gathered at 30 second intervals, and collected for the 60 minute period
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Data for the Normal operating and 25%, 50%%d, 75%%d 100% steady state operation tests will be gathered on the critical parameters using the Process P1ant Computer (PPC) on demand BOP log, P-1 and OD-3 edits and procedure checkoff forms as specified.

B. Critical Parameters: Computer Point

1. Neutron Flux (%%d) H441 Z. Core Thermal Power C875 3 ~ Total Core Flow (Total Recirc Flow) A445
4. Reactor Dome Pressure 0372
5. Reactor Hater Level (Narrow Range) 0377
6. Total Steam Flow 0376
7. Total Feedwater Flow ~

A391

8. Feedwater temperature A390 (after last stage of feedwater heaters)
9. Control Rod Drive System
a. Flow C328
b. Temperature 0417 ANSI 3.5 A3.2 NORMAL OP TEST -1 ADMIN/1 63
10. Reactor Water Cleanup System
a. Inlet Flow F361
b. Return Flow
c. Return Temperature F360
11. Main Generator Gross Electrical Watts F414
12. Drywell
a. Temperature D322
b. Pressure D320
13. Secondary Plant Heat Balance Data (OD-3 and BOP Log)

C. Supplemental Data:

1. None Prerequisites A. All participants read applicable test sections and fi 11 out section VIII.a (performed by)

IV. Procedure: (individual test sections)

A. Plant start-u to Hot Standb 0 eration at Hot Standb and Start-u from Hot Standb to Rate Power includin Turbine and Generator Start-u and Load Chan es Using controlled procedures attached perform the following and initial procedures where applicable

1. Initial conditions Cold Shutdown with normal reference plant equipment lineup
2. Complete Master Start-up Checkoff Form III
3. Complete Master Systems per-start-up Checkoff List Form IV
4. Perform a Plant Start-up to Hot Standby and maintain Hot Standby conditions in accordance with (IAW) NI-OP-43 Sections "E" and "H.2.1"
5. Continue Start-up to rated power IAW Nl-OP-43, Section "E"
a. Hold power steady at 25%, 501., 751. and 100't; gather critical data IAW Section II.A.2 NOTE: Verify AGAF's within technical specification 1 imi ts ANSI 3 5~ A3. 2 NORMAL OP TEST -2 ADMIN/1 63
6. Test Section Acceptance
a. The simulator plant start-up accomplished IAW reference plant controlled operating procedures, any procedure exceptions documented IAW NTI-4.5.3' Sati sfactory / Unsatisfactory Initial/Date Initial/Date
b. The simulator steady state parameters meet performance criteria of Section V.A, V.B, and Y.C

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. Plant Shutdown to Hot Standb and Cooldown to Cold S/0 Conditions Using controlled procedures attached perform the following and initial procedures where applicable,

l. Initial conditions rated power steady state with normal reference plant equipment lineup
2. Perform a plant shutdown to Hot Standby IAW Nl-OP-43, Section "G" and "H.2.2.2"
3. Continue plant cooldown to cold shutdown IAW Nl-OP-43, Section "G"
4. Test section acceptance
a. The simulator plant shutdown accomplished in accordance with (IAW) reference plant controlled operating procedures, any procedures exceptions documented IAW NTI-4.5.3.

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date ANSI 3.5 'A3.2 NORMAL OP TEST -3 ADMIN/1 63

C. Reactor tri followed b recover to rated ower Startu Shutdown and ower o eration with less than full reactor coolant flow Using controlled procedures attached perform the following and i,ni tial procedures where applicable

l. Initial conditions Rated power steady state with normal reference plant equipment lineup
2. Manually scram the Reactor and perform Special Operating Procedure No. 1 "Reactor Scram"
3. Perform Short Pre-Start-up Check-Off-Systems Form I
4. Shutdown Recirc Pumps 12, and 14 IAW Nl-OP-l, leaving 3 recirc loops in operation for Reactor startup
5. Perform a Reactor start-up to maximum attainable with 3 loop configuration IAW Nl-OP-43, Section "E"
6. Startup idle Recirc Loops IAW Nl-OP-l, Section "H.1.0"
7. Continue power ascension to rated
9. Reduce power to the maximum attainable in 3 loop configuration (Step IV.C.5) and shutdown recirc Pumps ll and 15 IAW Nl-OP-1, Section "H.2.0"
11. Perform a Reactor shutdown to all control rods inserted to position 00 IAW Nl-OP-43, Section "G"
12. Test section acceptance
a. All manipulations in test Section "III.C" accomplished in accordance with (IAW) reference plant controlled operating procedures, any procedure exceptions documented IAW NTI-4.5.3

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date D. 100% stead state 60 minute run erformance test

l. Initial conditions: Full power steady state, Equilibrium Xenon at 100% + 1/., Middle of cycle, with normal reference plant equipment lineup NOTE Use fast time Xe and snap shot a test IC .if necessary to establish initial conditions ANSI 3.5 A3.2 NORMAL OP TEST -4 ADMIN/1 63
2. Reset to test condition IC, DO NOT TAKE OUT OF FREEZE
3. Activate data gathering program ANSI DATA COLLECTION option - 4 4, Un-freeze the Simulator and allow it to run
5. At problem time 61:00 minutes freeze the simulator
6. Download the collected data to a PC and graph the information
7. Evaluate the critical parameter graphs or raw data for acceptance in accordance with performance criteria section V.D.

Satisfactory Unsatisfactory Initial/Date Initial/Date Surveillance Tests Using controlled procedures attached perform the following

.listed surveillance tests. Initial and sign test procedures where applicable.

NOTE: Use only installed Control Room indications Nl-ST-C2, Manual Opening of Solenoid Actuated Pressure Relief Valves and Flow Verification

2. Nl-ST-Cl, Auto Startup of High Press Cool Inj ec Sys
3. Nl-ST-C5, Secondary Containment and Reactor Building Emergency Ventilation System Operability
4. Nl-ST-C7, Automatic Securing and Isolation of the Mechanical Vacuum Pumps
5. Nl-ST-C8, Automatic Initiation of Off-Gas Isolation Valve
6. Nl-ST-C14, Alternate Control Rod Insertion/Back-up Scram Valve/Scram
7. Nl-ST-DO, Daily Checks
8. Nl-ST-ICl, Liq POI Pump Inoper Comp Oper Test
9. Nl-ST-IC2, Emer Cool Surv With an Inoper Sys Test
10. Nl-ST-IC3, Core Spr Redundant Comp or Sys Oper Test ll. Nl-ST-IC4, Control Rod Dr Pump Surv W/Inoper Comp Test
12. Nl-ST-IC5, High Pressure Coolant Injection Surveillance with Inoperable Component Test
13. Nl-ST-IC7, Emer Vent Sys Surv with an Inoper Branch
14. Nl-ST-IC9, Emer Diesel Gen Inoper Comp Oper Test ANSI 3.5 A3.2 NORMAL OP TEST -5 ADMIN/163
15. Nl-ST-Ml, Liquid Poision System Pump 5 Valve Operability Test
16. Nl-ST-M2, Emers Cool Sys Makeup Tanks Lvl Control Vlvs Exer
17. Nl-ST-M3, Supp Pool Drywell Relief Vlvs Exer
18. Nl-ST-M4, Emergency Diesel Generators Manual Start 5 1 Hr Rated Load Test Power Board 102 5 103 Undervoltage Relay Test
19. Nl-ST-M6, Core Spray Keep Fill System
20. Nl-ST-M8, Emer Vent Sys Oper Test
21. Nl-ST-M10, Scram Discharge Volume Vent 5 Drain Valve Position Verification
22. Nl-ST-Ql, Core Spray Pumps Valves Operability Test
23. Nl-ST-Q2, Control Rod Drive Pumps Flow Rate Test
24. Nl-ST-Q3, High Press Cool In]ec Pump and Valve Oper Test
25. Nl-ST-Q4, Reactor Coolant System Isolation Valves Operability Test
26. Nl-ST-Q5, Primary Cont Iso Vlvs Exercising
27. Nl-ST-Q6, Containment Spray System Quarterly Operability Test
28. Nl-ST-Q7, Manual Scram Instrument Channel Test
29. Nl-ST-QB, Liquid Poison Pump and Check Valve Operability Test
30. Nl-ST-Q13, Emerg Service Water Pump 5 Check Valve Operability Test
31. Nl-ST-Q15, Condensate Transfer System Operability Test
32. Nl-ST-Q16, Emergency Diesel Generator Quarterly Test
33. Nl-ST-Q17, N2 Supply Systems Valves Operability Test
34. Nl-ST-Q21, Instrument Air Valves
35. Nl-ST-Q24, Drywell/Torus and Torus/Reactor Building Vacuum Breakers Test
36. Nl-ST-R1, Control Rod Scram Insertion Time Test
37. Nl-ST-R8, Hi Cool 5 Prim Cont Iso Vlvs Timing
38. Nl-ST-R16, Emergency Service Water Pump Header Test
39. Nl-ST-V3, Rod Worth Minimizer Operability Test
40. Nl-ST-V4, Feedwater & Main Steam Line Power Operatied Iso Vlvs Ex ANSI 3. 5 A3. 2 NORMAL OP TEST -6 ADMIN/163
41. Nl-ST-V5, Suppression Temperature Moni toring During Relief Valve Operation
42. Nl-ST-V7, Reactor Building Closed Loop Cooling System Pump 5 Valve Operability Test
43. Nl-ST-VB, Main Steam Iso Vlv Full Closure Test
44. Nl-ST-Wl, Control Rod Exercising
45. Nl-ST-W4, Main Steamline High Radiaion Instrument Channel Test Section Acceptance
a. All manipulations in test Section "III.D" accomplished in accordance with (IAW) reference plant controlled operating procedures, any procedure exceptions documented IAW NTI-4.5.3

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date V. Performance Criteria A. Simulator instrument error shall be NO greater than that of the related instrument in the reference plant B. Principal mass and energy balances shall be satisfied

1. Net NSSS thermal power to generated electrical power
2. Feedwater flow to reactor thermal power C. Simulator computed values of critical parameters shall agree within + 2/ of the reference plant parameter at the specified power level D. Simulator computed values for steady state, full power reference plant configuration operation, shall be stable and not vary more than + 2% of the initial values over a 60 minute period ANSI 3.5 A3.2 NORMAL OP TEST -7 ADMIN/1 63

VI. Remarks:

VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3. 5 A3. 2 NORMAL OP TEST -8 ADMIN/1 63

0 NIAGARA IIOHAHK SIMULATION FACILITY PERFORMANCE TEST Simulation faci1 ity: NINE MILE POINT UNIT 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI 3.5 A3.4 Plant Malfunctions Tests Initial condi tions: In accordance, with the Malfunction applicable Acceptance Test Procedure (ATP) Section Data Collection:

A. Method: Manual as specified by the applicable ATP section B. Parameters: In accordance with the applicable ATP section Prerequisi tes A. All participants read test procedure and fill out section VII.A (performed by)

IV. Procedure A. Perform each of the following listed Malfunction tests in accordance with the applicable ATP section and Malfunction Cause and Effect

1. AD04 Relief Valve Leaks,
2. CS01 Core Spray Pump Trip
3. CT02 Containment Spray RAN Nater Pump Trip
4. CU03 RNCU Reject FCV Fails Open
5. CU07 RNCU Low Pressure Control Valve Fails Open
6. CUll Coolant Leak Outside Primary Containment
7. CN04 RBCLC Pump Trip
8. CN08 Circulating Nater Intake Structure Icing
9. DG02 Diesel Generator Trip
10. EC04 Emergency Cooling System Return Valve Fails Open
11. ED01 Loss of Offsite 115 KV Power
12. ED05 PB12 Electrical Fault
13. ED09 PB13 Section "A" Electrical Fault
14. ED13 PB14 Section "B" Electrical Fault
15. ED17 PB15 Section "C" Electrical Fault
16. ED21 PB17 Section "B" Electrical Fault ANSI 3.5 A3.4 MALFUNCTION TESTS -1 ADMIN/163
17. ED25 Loss of Power to Instrument Control Bus 130 Normal and Al t
18. EG04 Main Generator Core Internal Heating
19. EG08 Generator Hydrogen Emergency Seal Oil Pump Failure
20. EG12 Power Grid Network Load Transient - Decrease
21. FP02 Electric Fire Pump Failure
22. FP06 Control Room Fire Detection Various Panels
23. FP10 Reactor Building Fire Detection
24. FN04 Shaft Driven Feedwater Pump 13 Failure
25. FN08 Feedwater Control Valve ll Controller Fails - Low
26. FN12 Feedwater Control Valve 13 13 Controller Fails Low
27. FN16 Feedwater Master Controll'er Fails as is
28. FN20 Condensate Reci rc Valve (FCV 50-24) Fails Closed
29. FN24 Feedwater Control Valve Fails Closed (13A/13B)
30. FH28 HPCI lode Failure to Initiate
31. IA01 Loss of Instrument Air
32. RC03 Hotwell Level, Controllers in Auto Fail High
33. MS01 Steam Line Rupture Outside Primary Containment
34. MS05 Turbine Steam Seal Regulator Fails Closed
35. MS09 Second Stage Reheater 112 Drain Tk Level Control Fail Low
36. NM02 SRM Channel Failure Downscale
37. NM10 IRM Channel Failure Upscale
38. NM18 IRM Channel Detector Stuck
39. NM25 LPRM Failure Upscale
40. NM36 Recirc Flow Converter Channel Failure - Upscale
41. NM40 Recirc Flow Converter Failure - Comparator
42. OG04 Off Gas Discharge to Stack Isolation Valve Fails Closed
43. PP01 Failure of Plant Process Computer
44. RD04 Control Rod Failure Stuck
45. RD08 Control Rod Failure RPIS
46. RD36 CRD Flow Control Valve Failure - Closed
47. RD40 Reactor Manual Control System Timer Malfunction Settle ANSI 3.5 A3.4 MALFUNCTION TESTS -2 ADMIN/1 63
48. RM03 Area Radiation Monitor Drawer Upscale
49. RP04 Reactor Protection System Failure to Scram - Automatic 50.. RP08 Anticipated Transient Nithout Scram <ATNS)
51. RR04 Recirculation Pump ll Control Signal Failure
52. RR08 Recirculation Pump 12 Seizure
53. RR12 Recirc Pump 13 Field Breaker Trip
54. RR16 Recirc Pump 14 Drive Breaker Trip
55. RR20 Recirc Pump 14 Incomplete Start Sequence
56. RR24 Recirc Pump 15 Control Signal Failure
57. RR27 Master Recirc Flow Controller Failure Low

, 58. RR31 Reactor Vessel Pressure Recorder Failure Low

59. RR35 Reactor Vessel Pressure Indicator Failure - Upscale
60. RR39 Reactor Vessel Level Recorder Failure - Downscale
61. RR43 Rx Vessel Level Indication (Control Sys) Fail as is
62. RR47 Recirc Pump Discharge Valve Stem Separates From Gate
63. RR51 Rx Vessel Level Transmitter (RPS'nput) Fails High
64. RR55 Rx Vessel Level Transmitter <Control Input) Fails Low
65. RR59 Rx Vessel Pressure Transmitter (RPS Input) Fails as is
66. RR63 Reactor Recirc Pump 12 Inner Seal Failure
67. RR67 Reactor Recirc Pump 15 Tachometer Fails - Oscillates
68. RR71 Reactor Safety Valve Inadvertently Opens
69. RX02 Increased Rod North
70. TC02 Turbine Governor Fails High
71. TC06 Electrical Pressure Regulator Fails - Oscillates
72. TC10 First Bypass Valve Sticks Open
73. TU01 Exhaust Hood Spray Valve Fails Closed
74. TU05 Main Turbine Bearing High Temperature
75. CU05 RHCU High Pressure Control Valve Fails Open Acceptanc e Criteria A. All malfunction tests completed satisfactorily in accordance with applicable ATP sections and Malfunction Cause and Effect

/ Sati sfactory / Unsati,s factory Initial/Date Initial/Date ANSI 3.5 A3.4 MALFUNCTION TESTS -3 ADMIN/1 63

YI. Remarks:

YII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Perfo. med By: /

B. Approved by SCCB Signature Date ANSI 3.5 A3.4 MALFUNCTION TESTS -'4 ADMIN/163

NIAGARA NOHANK SINULATION FACILITY PERFORNANCE TEST Simulation facility: NINE MILE POINT UNIT Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "B" B1.2<1) MANUAL SCRAM TRANSIENT Ini tial conditions: Ful 1 power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

l. Data wi 11 be gathered on the critical parameters using a Simulator program called "ANSI B121"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a personal computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor Power <Neutron Flux /) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Hide Range Reactor Pressure D373
5. Narrow Range Reactor Pressure D372
6. Hide Range Reactor Hater Level J342
7. Narrow Range Reactor Hater Level (Feedwater Control) D377
8. Generator Gross Electrical Power F414
9. Turbine Steam Flow (By First Stage Shell Press) B465
10. Total Core Flow (Total Recirc Flow) A445 ANSI 3. 5 Bl . 2( 1 ) TRANSIENT TEST -1 ADMIN/1 63

0 C: Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following:

l. Insert the following malfunction with delay time start of problem time 1:00 minute
a. RP03 Reactor Scram
2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress 5.

eitk I operator act1ons At problem time ll:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date

6. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI 9, Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 Bl.2(l) TRANSIENT TEST -2 ADMIN/1 63

V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Ini t i al /Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

BE Approved by SCCB Signature Date ANSI 3.5 81.2(1) TRANSIENT TEST -3 ADMIN/1 63

0 NIAGARA NOHANK SINULATION FACILITY PERFORMANCE TEST Simulation facility: NINE MILE POINT UNIT- 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(2) SIMULTANEOUS TRIP OF ALL FEEDHATER PUMPS Initial conditions: Full power steady state, end of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

l. Data will be gathered on the critical parameters using a Simulator program called ANSI DATA COLLECTION option 1 seconds before the transient and lasting for 10 minutes of the transient
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power (% neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Hide Range Reactor Pressure D373
5. Narrow Range Reactor Pressure D372
6. Hide Range Reactor Hater Level J342
7. Narrow Range Reactor Hater Level (Feedwater Control) D377
8. Generator Gross Electrical Power F414
9. Turbine Steam Flow (By First Stage Shell Press) 8465
10. Total Core Flow (Total Recirc Flow) A445 ANSI 3.5 81.2(2) TRANSIENT TEST -1 ADMIN/163

0 C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of'vents Log Printout Prerequisi tes A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following:

l. Insert the following malfunction with delay time start of problem time 1:00 minute
a. FHOlA "A" feedwater pump trip
b.

FN018 "8" feedwater pump trip

c.

FNOlC "C" feedwater pump trip

2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un -freeze the Simulator and allow the transient to progress with NO operator actions
5. At problem time 11:30 freeze the simulator and gather the PPC printouts, and label wi th the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR',

and enter any appropriate remarks in Section VI,

7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
9. Simulator Configuration Control 8oard (SCCB) review and approve test results, Section VII.B ANSI 3.5 81,2(2) TRANSIENT TEST -2 ADMIN/163

V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 Bl.2(2) TRANSIENT TEST -3 F

ADMIN/1 63

NIAGARA llOHAIIK SIMULATION FACILITY PERFORMANCE TEST Simulation facility: NINE MILE POINT UNIT - 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "B" 81.2(3) SIMULTANEOUS CLOSURE OF ALL MAIN STEAM ISOLATION VALVE Initial conditions: Full power steady state, middle of cycle, wi th nor'mal reference plant equipment lineup Data Collection:

A. Method

1. Data will be gathered on the critical parameters using a Simulator program called "ANSI 8123"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer <PC) and graphed for evaluation
2. Process Plant Computer <PPC) alarm and Sequence of Events

<SOE) printouts will be gathered for alarm and automatic action-evaluation B. Cri ti cal Parameters: Computer Point Reactor Power (1. neutron flux) H441

2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Nide Range Reactor Pressure 0373 Narrow Range Reactor Pressure D372
6. Nide Range Reactor Nater Level J342
7. Narrow Range Reactor Hater Level (Feedwater Control) D377
8. Generator Gross Electrical Power F414
9. Turbine Steam Flow (By First Stage Shell Press) B465
10. Total Core Flow (Total Recirc Flow) A445 ANSI 3.5 81.2(3) TRANSIENT TEST ADMIN/163

C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following:,

1. Insert the following malfunction with delay time start of problem time 1:00 minute a ~ RP06 Vessel Isolation
2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress with I operator actions At problem time ll:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the information Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 81.2(3) TRANSIENT TEST -2 ADMIN/163

0 V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Un sat i sf ac tory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 81.2(3) TRANSIENT TEST -3 ADMIN/1 63

NIAGARA IIOHAIIK SIHIILATION FACILITY PERFONNANCE TEST Simulation faci i ty:

1 NINE MILE POINT UNIT 1 Reference Plant: NINE MILE POINT UNIT Performance Test: 1989 ANSI APPENDIX "8" 81.2(4) SIMULTANEOUS TRIP OF ALL RECIRCULATION PUMPS Enitial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

1. Data will be gathered on the critical parameters using a Simulator program called "ANSI 8124"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts wi 11 be gathered for alarm and automatic action evaluation
8. Critical Parameters: Computer Point
1. Reactor power ( / neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Narrow Range Reactor Pressure D372
5. Narrow Range Reactor Hater Level (Feedwater Control) 0377
6. Total Core Flow = (Total Recirc Flow) A445
7. Individual Recirculation Loop Flows
a. Loop ll A430
b. Loop 12 C A434
c. Loop 13 A438
d. Loop 14 A442
e. Loop 15 A446 ANSI 3.5 81.2(4) TRANSIENT TEST -1 ADMEN/163

C. Supplemental Data

1. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

l. Insert the following malfunction with delay time start of problem time 1:00 minute
a. RR01 Pump 11 Trip
b. RR06 Pump 12 Trip
c. RR11 Pump 13 Trip
d. RR16 Pump 14 Trip
e. RR21 Pump 15 Trip
2. Activate the data gathering program specified (II.A.l)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress with NO operator actions
5. At problem time ll:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data f'r acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.

Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 B1.2(4) TRANSIENT TEST -2 ADMIN/163

V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 81.2(4) TRANSIENT TEST -3 ADMIN/163

HIAGARA HOHAHK SIHULATIOH FACILITY PERFORHAHCE TEST Simulation facility: NINE MILE POINT UNIT 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(5) SINGLE RECIRCULATION PUMP TRIP Initial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

l. Data will be gathered on the critical parameters using a Simulator program called "ANSI 8125"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power (1. neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391 4 ~ Narrow Range Reactor Pressure D372
5. Narrow Range Reactor Hater Level (Feedwater Control) D377
6. Total Core Flow (accurate indication not available)
7. Individual Recirculation Loop Flows
a. Loop 11 A430
b. Loop 12 A434
c. Loop 13 A438
d. Loop 14 A442
e. Loop 15 A446 ANSI 3.5 81.2(5) TRANSIENT TEST -1 ADMI N/1 63

C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

l. Insert the following malfunction with delay time start of problem time 1:00 minute
a. RR21 Pump 15 Trip
2. Activate the data gathering program specified (II.A.l)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress with NO operator actions
5. At problem time 11:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
9. Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 81.2(5) TRANSIENT TEST -2 ADMIN/1 63

V. Acceptance Cri teria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsati sfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Ini ti al /Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 B1.2(5) TRANSIENT TEST -3 ADMIN/163

NIAGARA MONANK SIMULATION FACILITY PERFORMANCE TEST Simulation facility: NINE MILE POINT UNIT - 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(6) HAIN TURBINE TRIP TRANSIENT TEST I. Initial conditions: Maximum power level which does not result in immediate reactor scram, and normal reference plant equipment lineup Data Collection:

A. Method

1. Data wi 11 be gathered on the critical parameters using a Simulator program called "ANSI 8126"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power ('L neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Wide Range Reactor Pressure D373
5. Narrow Range Reactor Pressure D372 V
6. Wide Range Reactor Hater Level j342
7. Narrow Range Reactor Water Level (Feedwater Control) D377
8. Generator Gross Electrical Power F414
9. Turbine Steam Flow (By First Stage Shell Press) 8465
10. Total Core Flow (Total Recirc Flow) A445 ANSI 3.5 81.2(6) TRANSIENT TEST -1 ADMIN/1 63

C. Supplemental Data

1. Alarm Typer Printout
2. Sequence of Events Log Printout Prere quisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Proce dure A. Reset to full power steady state, end of cycle, with normal reference plant equipment lineup, and perform the following

1. Reduce reactor power until annunciator F3-4-6 comes in, in accordance with Nl-OP-43
2. Stabilize power and snap shot a test initial condition (IC) labeled as ANSI B1.2(6)
3. Reset to the test IC, DO NOT TAKE OUT OF FREEZE
4. Insert the following malfunction with delay time start of problem time 1:00 mi nute
a. TK1 Turbine Trip
5. Activate the data gathering program specified (II.A.l)
6. Roll ahead the printer paper for a clean start point
7. Un-freeze the Simulator and allow the transient to progress with NO operator actions
8. At problem time ll:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
9. Evaluate the printouts for acceptance criteria V.B and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
10. Download the collected data to a PC and graph the information
11. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
12. Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 B1.2(6) TRANSIENT TEST -2 ADMIN/163

V. Acceptance Cri teri a A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ . Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII, Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 B1.2(6) TRANSIENT TEST -3 ADMIN/163

0 NIAGARA HONAHK SIHULATION FACILITY PERFORHANCE TEST Simulation facility: NINE MILE POINT UNIT - 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "B" B1.2(7) MAXIMUM RATE POWER RAMP (100% - 75% 100%)

Initial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

l. Data will be gathered on the critical parameters using a Simulator program called "ANSI B127"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer <PC) and graphed for evaluation
2. Process Plant Computer <PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power <% neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater Flow A391
4. Wide Range Reactor Pressure D373
5. Narrow Range Reactor Pressure D372
6. Wide Range Reactor Water Level J342
7. Narrow Range Reactor Water Level.

(Feedwater Control) D377

8. Generator Gross Electrical Power ~ F414
9. Turbine Steam Flow

<By First Stage Shell Press) B465

10. Total Core Flow (Total Recirc Flow) A445 ANSI 3.5 B1.2(7) TRANSIENT TEST -1 ADMIN/163

C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events l.og Printout Prerequisites A. All participants read test procedure and fi 11 out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

l. Activate the data gathering program specified <II.A.1)
2. Roll ahead the printer paper for a clean start point
3. Un-freeze the Simulator and wait for minute problem time 1
4. At problem time = 1 minute, using the Master Controller, Ramp reactor power at a rate of 6 MNE/SEC by reducing recirculation flow until reactor power is 75/, then raise recirculation flow until reactor power is 97.5%
a. In accordance with Nl-OP-43
5. Hhen reactor power = 97.5%, raise reactor power to 100% at 5 MNE per hour using the Master recirc flow controller 4 V
6. At problem time 11:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
7. Evaluate the printouts for acceptance criteria V.B. and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
8. Download the collected data to a PC and graph the information
9. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
10. Simulator Configuration Control Board <SCCB) review and approve test results, Section VII.B ANSI 3.5 81.2(7) TRANSIENT TEST -2 ADMIN/1 63

V. Acceptance Cri teri a A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date Remarks YII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 B1.2(7) TRANSIENT TEST -3 ADMIN/163

NIAGARA NOHANK SINULATION FACILITY PERFORIIANCE TEST Simulation facility: NINE MILE POINT UNIT 1 Reference Plant: NINE MILE POINT UNIT - 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(8) MAXIMUM SIZE REACTOR COOLANT SYSTEM RUPTURE (LOCA) COMBINED WITH LOSS OF ALL OFF SITE POWER (LOOP)

Initial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

l. Data wi 11 be gathered on the critical parameters using a Simulator program called "ANSI B128"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power ('/ neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater/High Pressure Coolant In]ection Flow A391
4. Wide Range Reactor Pressure 0373
5. Hide Range Reactor Hater Level J342
6. Fuel Zone Hater Level ~ H447
7. Torus Hater Temperature H478
8. Torus Air Temperature H477
9. Torus Pressure D324
10. Drywell Temperature D322 ll. Drywell Pressure D320
12. Low Pressure Core Spray Flow (No Point Available See FI RV-35A Il B)

ANSI 3.5 B1.2(8) TRANSIENT TEST -1

C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

l. Insert the following malfunctions with delay time start of problem time 1:00 minute
a. RR29 - RRP Suction Break
b. ED01AB Loss of 115KV
2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress wi th NO cperator actions
5. At problem time ll:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the priniouts for acceptance criteria V.B. and V.C, initial the appropriate acceptance b1ocks, write any DR's, and enter'ny appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the informati on Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.

Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3. 5 81.2(8) TRANSIENT TEST -2 ADMIN/1 63

V. Acceptance Cri teri a A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date lk C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Sati sfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3.5 B1.2(8) TRANSIENT TEST -3 ADMIN/163

NIAGARA HOHAHK SIHULATION FACILITY PERFORHANCE TEST Simulation facility: NINE MILE POINT UNIT - 1 Reference Plant: NINE MILE POINT UNIT - 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(9) MAXIMUM SIZE UNISOLABLE MAIN STEAM LINE RUPTURE Initial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Collection:

A. Method

1. Data wi 1 1 be gathered on the critical parameters using a Simulator program called "ANSI B129"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: Computer Point
1. Reactor power (/. neutron flux) H441
2. Total Steam Flow D376
3. Total Feedwater/High Pressure Coolant Injection Flow A391
4. Hide Range Reactor Pressure D373
5. Hide Range Reactor Hater Level J342
6. Fuel Zone Hater Level H447
7. Torus Hater Temperature ~ H478
8. Torus Air Temperature H477
9. Torus Pressure D324 10 'rywell Temperature D322
11. Dryweli Pressure D320
12. Low Pressure Core Spray Flow (No Point Available See FI RV-35A 5 B)

ANSI 3.5 81.2(9) TRANSIENT TEST -1 ADMIN/1 63

C. Supplemental Data

l. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fi 11 out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

l. Insert the following malfunctions with delay time start of problem time 1:00 minute
a. NS04 Steam Line Rupture Inside Primary Containment
2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress with NO operator actions
5. At problem time 11:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B. and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.
7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
9. Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 B1.2(9) TRANSIENT TEST -2 ADMIN/163

V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Satisfactory / Unsatisfactory Initial/Date Ini ti al /Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Perrormed By: /

8. Approved by SCCB Signature Date ANSI 3.5 B1.2(9) TRANSIENT TEST -3 ADMIN/163

NIAGARA HOHAHK SIHULATION FACILITY PERFORHANCE TEST Simulation facility: NINE MILE POINT UNIT - 1 Reference Plant: NINE MILE POINT UNIT 1 Performance Test: 1989 ANSI APPENDIX "8" 81.2(10) SIMULATANEOUS CLOSURE OF ALL MAIN STEAM ISOLATION VALVES (MSIV) COMBINED NITH SINGLE STUCK OPEN ELECTROMATIC RELIEF VALVE TRANSIENT TEST Initial conditions: Full power steady state, middle of cycle, with normal reference plant equipment lineup Data Col 1 ection:

A. Method

1. Data wi 11 be gathered on the critical parameters using a Simulator program called "ANSI 81210"
a. Data is gathered at 0.5 second intervals, starting 15 seconds before the transient and lasting for 10 minutes of the transient
b. Data is then down loaded to a Personal Computer (PC) and graphed for evaluation
2. Process Plant Computer (PPC) alarm and Sequence of Events (SOE) printouts will be gathered for alarm and automatic action evaluation B. Critical Parameters: ~

Computer Point Reactor power (1. neutron flux), . . H441

2. Total Steam Flow D376
3. Total Feedwater/High Pressure Coolant Injection Flow A391
4. Hide Range Reactor Pressure D373
5. Hide Range Reactor Hater Level J342
6. Fuel Zone Hater Level H447
7. Torus Hater Temperature H478
8. Torus Air Temperature H477
9. Torus Pressure D324
10. Drywell Temperature D322
11. Drywell Pressure D320
12. Low Pressure Core Spray Flow (No Point Available See FI RV-3SA Il 8)

ANSI 3.5 81.2(10) TRANSIENT TEST -]

C. Supplemental Data

1. Alarm Typer Printout
2. Sequence of Events Log Printout Prerequisites A. All participants read test procedure and fill out Section VII.a (performed by)

IV. Procedure A. Reset to initial conditions specified in Section "I.", DO NOT TAKE OUT OF FREEZE, and perform the following

1. Insert the following malfunctions with delay time start of problem time 1:00 minute
a. RP06 - Vessel Isolation
b. AD06 - ERV ill Sticks Open
2. Activate the data gathering program specified (II.A.1)
3. Roll ahead the printer paper for a clean start point
4. Un-freeze the Simulator and allow the transient to progress wi th NO ope. ator actions
5. At problem time 11:30 freeze the simulator and gather the PPC printouts, and label with the performance test title and date
6. Evaluate the printouts for acceptance criteria V.B. and V.C, initial the appropriate acceptance blocks, write any DR's, and enter any appropriate remarks in Section VI.

r

7. Download the collected data to a PC and graph the information
8. Evaluate the critical parameter graphs or raw data for acceptance criteria V.A, initial the appropriate acceptance block, write any DR's, and enter any appropriate remarks in Section VI.
9. Simulator Configuration Control Board (SCCB) review and approve test results, Section VII.B ANSI 3.5 B1.2(10) TRANSIENT TEST -2 ADMIN/163

V. Acceptance Criteria A. The observable change in the Simulator monitored parameter corresponds in direction to those expected in the reference plant Re: Critical parameter raw data or PC graph

/ Sati sfactory / Unsatisfactory Initial/Date Initial/Date B. The Simulator caused alarms and automatic actions that would have happened in the reference plant Re: PPC alarm and SOE printouts

/ Satisfactory / Unsatisfactory Initial/Date Initial/Date C. The Simulator did not cause an alarm or automatic action that would NOT have happened in the reference plant Re: PPC alarm and SOE printouts

/ Sati sfactory / Unsatisfactory Initial/Date Initial/Date VI. Remarks VII. Acceptance INITIALS SIGNATURE PRINTED NAME /DEPARTMENT A. Performed By: /

B. Approved by SCCB Signature Date ANSI 3. 5 81. 2(10) TRANSIENT TEST -3 ADMIN/163

0 NINE MILE POINT UNIT 1 SIMULATOR TEST DATE: 05/17/89 SECTION: 3.051 REV: 2 MALF ED09 AC POWER BOARD ELECTRICAL FAULT (PB 13 SECTION A)

Ol INITIALIZE THE TRAINER TO A FULL POWER IC.

PCM:

TRAINER RESET 02 INSERT MALF ED09 PCM:

MALF ED09 ACTIVE THIS RESULTS IN THE TRIPPING OF THE MANUAL SUPPLY BREAKER FROM AUXILIARY FEEDER 11.

A4-4-3: "POWER BOARD 13-14-15 LOW BUS VOLTAGE" REF: C-19426-2 INITIAL DATE THE FOLLOWING LOADS (BUS TABLE 7-3A) WXLL EXPERIENCE A TOTAL LOSS OF POWER:

POWER BOARD 131A. REFER TO BUS'ABLE 7-11A FOR A LIST OF AFFECTED LOADS.

2, MECHANICAL VACUUM PUMP 11

3. STATOR WATER CIRC PUMP 11. REFER TO SECTION 3.076 FOR DETAXLS.

4, REACTOR TRIP MG SET. REFER TO BUS TABLE 7-29A. XNXTIAL DATE 03 ATTEMPT TO CLOSE THE MANUAL SUPPLY BREKAER FROM FEEDER 11 TO PB 13 SECTION A.

PCM:

REMOTE FUNCTION RED26 CLOSES BREAKER, BUT BREAKER IMMEDIATELY TRZPS OPEN. INITIAL DATE 04 ATTEMPT TO CLOSE THE MANUAL SUPPLY BREAKER PROM POWER BORAD 13 SECTION B.

PCM:

REMOTE FUNCTION RED03 CLOSES BREAKER, BUT BREAKER IMMEDIATELY TRXPS OPEN. INITIAL DATE

NINE MILE POINT UNIT 1 SIMULATOR TEST DATE: 05/17/89 SECTION: 3.051 REV: 2 PAGE 2 05 CLOSE THE MANUAL SUPPLY BREAKER FROM PB 131A TO PB 131B.

PCM:

REMOTE FUNCTION RED15 OPENS MANUAL SUPPLY TO PB 131A FROM PB 13A AND CLOSES SUPPLY BREAKER FROM PB 131B.

POWER IS RESTORED TO PB 131A.

INITIAL DATE 06 REMOVE MALF ED09 PCM:

MALF ED09 REMOVED RECLOSE THE MANUAL SUPPLY BREAKER FROM FEEDER 11 TO POWER BOARD 13 SECTION A PCM:

REMOTE FUNCTION RED26 CLOSES THE BREAKER A4-4-3 CLEARS: "POWER BOARD 13-14-15 LOW BUS VOLTAGE" INITIAL DATE POWER IS RESTORED TO PB 13A INITIAL DATE END OF MAXZ ED09

0

.,;. I.Ic'"l

>r,ii

~

~

-~t

~

.. ~ "w NIhE NIL'" FT ~) l'lALFUNCTION CAUSE AND EFFECTS MALF NO ~ MALFUNCTION TITLE/<ANGE/CAUSE AND EFFECT ED09 AC POWER BOARD ELECTRICAL FAULT (F913 SECTION A)

TYPP:

ED - DISCRETE CAUS:-:

SHORT TO 6"cOUND ON PB1 3A BUS BAR.

PLAN T S TATUS:

1007. POWER EFFECTS:

THIS MALFUNCTION MILL RESULT IN THE TRIPPING

( OF THE SUPPLY BREAKER TO F8134 FROM AUXILIARY FEEDER 11 ~ IF THE TIE BREAKER BETWEEN SECTIONS A AND B IS CLOSED IN (BY USE OF REMOTE FUNCTIO" ) IT WILL TRIP ~ THIS RESULTS IN A TOTAL LOSS OF POWER TO P91 3A.

THE FOLI OMI NG ANNUNCIATOR 'WILL BE P EC E I VED IMMEDIATELY:

PANEL MlNOOW ENGRAVING

~0 ~ W ~ WW W 0 r

- A4 A4-27 POWER BOARD 13-14 "15 LOM BUS YOLTAGF REF FR TO BUS TABLE 7 "3A FOR EQUIPMENT'HAT MILL BE AFFECTED EACH SYSTEll AFFECTED BY THIS MALFUNCTION MILL RESPOND APPROPRIATELY FOR THE LOSS OF THE VARIOUS COMPONFNTS ~

ALL A>PRVFRIATE ALARMS AND INDICATIONS MILL AC'TUATE ~

REMOVAL OF THIS MALFUNCTION WILL REMOVE THE SHORT FROM THE BUS BAR. THE MANUAL SUPPLY BREAKER. FROM AUXILIARY FEEDER 11 MILL CLOSE WHEN THE MALFUNCTION IS REMOVED ~

REFT C 19<26 C

ANSI/ANS 3.5 ANNUAL REPORT prepared for NIAGARA MOHAWK POWER CORPORATION GP-R-115007 March 1, 1986 GENERAL PHYSICS CORPORATION Oswego, New York

0 GENERAL PHYSICS CORPORATION G P-R-115007 March 1, 1986 I. SIMULATOR INFORMATION This section provides pertinent descriptive information on the Nine Mile Point Unit One simulator. This section summarizes the relevant information concerning the simulator and its suitability and applicability as an operator training device.

A. Genera 1

1. The Nine Hi le Point Unit One simul ator is owned by General Physi cs Ni a ga ra Corp orat i on, a whol ly owned subs i di a ry of General Physi cs Corporation. The simulator is operated jointly by Niagara Mohawk Power Corporation and General Physics Niagara Corporation instructors. All maintenance and modifications have been performed by lmneral Physics at the di rection of Niagara Mohawk Power Corporation. . The simulator was manufactured by Singer/Link.
2. The Nine Mi 1 e Point Unit One simulator is a full scope control room simulator which simulates the Nine Mile Point Unit One plant, owned and operated by Niagara Mohawk Power Corporation. Nine Mile Point Unit One is an 1850-MWt . BWR-2 manufactured by General Elect ri c. The rated electrical output is 620 MWe.
3. The Nine Mile Point Unit One simulator was declared Ready for Training on I September, 1984.
4. Thi s report is the ini t i a 1 report on the Uni t One simul ator performance. This report was prepared to document simulator modifications since the .Ready for Training date and to document the simulator's performance subsequent to the modifications.

As of I March, 1986, the simulator has been updated to ref 1 ect al 1 modifications made to the reference plant through I September, 1984. In addition, the simulator has been modi fied to reflect actual plant performance characteristics for several additional parameters.

GENERAL PHYSICS CORPORATION GP-R-115007 March I, 1986 B. Control Room

1. The physical configuration of the reference plant control room is shown in Figure 1. The physical configuration of the simulator control room is shown in Figure 2. A comparison of these figures reveals a high degree of simi 1 ar ity between the simul ator and the ref erence pl ant control room. The following di fferences exist in the physical arrangements:'imulator Reference Plant

~ Has the instructor console ~ Has a desk at this location.

next to the NSSS typer.

~ Does not have the ASSS ~ Has an addition on the SSS office addition. office for the ASSS.

~ No simulation of plant e A Meteorological Computer is Meteorological Computer. located in the control room.

2. Panels and Equipment The Nine Mile Point Unit One simualtor contains all the panels which are in the reference plant control room, and it fully simulates all front panel and back panel controls and indications . Some equipment whi ch has neither controls nor indications is cosmetically simulated. The following list summarizes the equipment differences between the simulator and the reference plant control room:

RPS relays are installed but are non-functional in the simulator.

Electr ical protective relays are cosmetically simulated utilizing photographs mounted inside the relay enclosures.

The simulator utilizes one operating TIP machine and a selector switch to functionally simulate the four plant TIP machines; the three remaining three TIP machines have all panel hardware installed but remai.n non-functional.

GENERAL PHYSICS CORPORATION GP-R-115007 March I, 1986

3. Systems All the, systems which are operable from the reference plant control room are fully simulated. For completeness, Attachment 1 lists those systems which are fully simulated.
4. Simulator Control Room Environment The simulator control room was specifically designed to duplicate as nearly as possible the reference plant control room environment. The floor tile and lighting systems are identical. The physical configuration of the panels is dimensionally identical, and false doors, and posts are placed in the simulator control room at the same locations in which they occur in the reference plant.

Al 1 inst rumentati on, control's, mimic, label s, and operator aids are identical within the limits of hardware availability. The paint color and shade used on the panels i s very nea rly the same. Al 1 communications equipment provided in the reference plant is provided in the simul ator. The ambi ent noi se level s in the simul ator are approximately the same as those in the reference plant control room.

C. Instructor Interface

1. There are twenty initial condition sets available for instructor use.

See Attachment 2 for a description of these. Thirty additional initial condition sets are available for use by the instructors.

2. The malfunctions are listed on At'tachment 3.
3. The remote functions, for controls located outside the control room are listed on Attachment 4.

3-

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986

4. Additional instructor aids include:

e Display of Monitored Parameters: DMP is available on an instructor station computer terminal; the instructor may select and display up to sixteen of the 100 available monitored plant parameters.

.Parameters are listed on Attachment 5.

~ Line Printer Plotter: The LPP function permits the instructor to print simultaneously up to 12 DMP variables on the computer room line printer. The time resolution of the LPP function is 0.1 seconds.

Record/Replay is an instructor-controlled feature by which the instructor may record trainee actions to magnetic tape in real time and later play this back through the simulator to enable a discussion of these actions.

D. Reference Plant Operating Procedures Trainees in the simulator utilize a controlled set of plant procedures whenever they are operating the simulator. When a plant procedure is inappropriate or cannot be utilized on the simulator, such as when a plant modification has not been incorporated, into the simulator at the time of a procedure revision, the difference is identified in the operating procedure in accordance with the Niagara Mohawk Power Corporation Nuclear Training Instruction for Simulator Instruction.

I I. SIMULATOR DESIGN DATA The simulator initial design data is listed in the original Database document on file at the simulator library. The initial simulator design data was frozen in April, 1982. Since the Ready for Training date (I September, 1984), a number of modifications have been made to the simulator, based on changes made to the reference plant.

As a result of these modifications, many new data items have been added to the simulator database, and several original database prints have been upgraded to more recent revision numbers. Simulator modifications and affected documents for the current reporting period are listed on Attachment 6.

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 III. SIMULATOR TESTS The simulator underwent a series of tests in the period between January 13, 1986 and February 28, 1986. Testing was done to verify real time operation, steady state and transient performance, and malfunction responses.

Documentation of the tests is available in the simulator database and records under the title "Simulator Performance Test Data - March 1986".

A. Computer Real Time Test Simulator real time testing was performed by measuring the indi vidual simulation model times during steady state and accident conditions. During this testing, no frame slippage or program overtimes accrued. In addition to the verification of timing, the simulator has designed safeguards which preclude operation outside of real time with two exceptions. The exceptions to real time operation are two instructor- controlled aids: Fast Time; and Slow Time. In slow time all simulator responses are in one-half of normal or real time. Fast time operation is specially handled and only affects reactor core xenon, condenser evacuation, and turbine warming.

B. Steady State and Normal Operations Steady state performance and simualtor stability have been verified at four power levels against reference plant Balance of Plant Logs. The testing power levels utilized were approximately 25, 50, 75 and 100 percent of rated power. Normal operation testing was performed by doing a plant startup, and shutdown in accordance with reference plant procedures. , Strip charts and process computer logs which document the acceptable stability of the simulator are contained in Volume 2 of the Simulator Performance Test Data books.

~

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 C. Transient Tests The following transient tests were performed on the Nine Mile Point Unit One simulator:

(1) Manual Scram (2) Simultaneous trip of all feedwater pumps (3) Simultaneous closure of all main steam isolation valves (4) Simultaneous trip of all recirculation pumps (5) Single recirculation pump trip (6) Main turbine tri p (maximum power level which does not result in immediate reactor scram)

(7) Maximum rate power ramp (master recirculation flow controller in "manual" ) down to approximately 75$ and back up to 100$

(8) Maximum size reactor coolant system rupture combined with loss of all offsite power (9) Maximum size uni solable main steam line rupture (10) Simultaneous closure of all main steam isolation valves combined with single stuck open safety/relief valve (inhibit activation of high pressure Emer gency Core Cooling Systems)

No actual plant response data is available for these specific transients.

However, a number of other plant transients (e.g., pressure control failure; turbine trip from rated power) have occurred. The simulator performance for these events compares very favorably with actual plant data.

The Final Safety Analysis Report (FSAR) analyzed six of the above transients. Four of the listed transients have neither specific plant data nor FSAR analysis. Those transients are the following: manual scram; main turbine trip at 355 power; maximum rate power ramp 100$ to 755 to 100$ ;

simultaneous closure of all MSIVs; with stuck open relief valve, and no high pressure ECCS. For these four events, the best-estimate method was used to evaluate simulator performance.

i

~

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 The transients were tested, and data was 'btained in, accordance with ANSI/ANS 3.5-1985 Appendix B. The fol lowing comments apply to the tested transients:

(1) Manual scram The simulator showed a 10 psi pressure spike when the turbine tripped, and a 6 psi pressure spike when the MSIVs closed on low pressure (no operator action). Opinion varies on the accuracy of these magnitudes. However, no plant data exists for this event (a procedural violation would be required), and the FSAR does not follow this event through MSIY closure.

(2) Simultaneous trip of all feedwater pumps The FSAR did not analyze a simultaneous trip of all feedwater pumps.

However, in the interest of comparing simulator performance to analyzed data, this event was compared to the Feedwater Control 1er (zero demand) Mal function. This changed the timing of the feedwater flow decrease, but otherwise the events are very similar.

The FSAR predicts a small, gradual increase in recirculation flow of approximately 5X over a period of about 5 seconds. The simulator shows no change in recirculation flow throughout the event. The reason for the discrepancy appears to be that recirculation flow increases due to the effects of reduced; two-phase . core flow. The simulator does not dynamically model two-phase pressure drop and flow effects . Since the discrepancy is slight and occurs after the low level scram, it is judged to be non-critical. This difference has no effect on training.

(3) Main steam isolation valve closure The simulator performance correlated very wel 1 with the FSAR prediction.

-e GENERAL PHYSICS CORPORATION G P-R-115007 March 1, 1986 (4) Simultaneous trip of all recirculation pumps The simulator performance correlated very wel 1 with the FSAR prediction.

(5) Single recirculation pump trip The simulator performance correlated very well with the FSAR prediction.

(6) Main turbine trip (maximum power level which does not result in immediate reactor scram)

The FSAR does not analyze this event, since it has no safety implications. There is no reference plant data available for this event. Therefore the simulator performance was compared to the best estimate. All measured parameters showed smooth, predictable responses. Reactor pressure rose very slightly (approximately 5 psi ). This was expected on the basis that the bypass valves present a slightly different flow resistance than the turbine does. A very mild feedwater flow transient occurred. This was also predicted based on the automatic trip of g13 feedwater pump and the automatic initiation of the HPCI mode of feedwater.

(7) Maximum rate power ramp (master recirculation flow controller in "manual" ) down to approximately 75$ and back up to 100$

This event was neither analyzed in'he FSAR nor performed at the plant as a test or normal evolution. The simulator performance was smooth and predictable on all parameters. No abnormalities were noted.

GENERAL PHYSICS CORPORATION GP-R-115007 March I, 1986 (8) Maximum size reactor coolant system rupture combined with loss of all offsite power The simulator response was very close to the FSAR prediction for all parameters except two: reactor vessel pressure response and drywell pressure response. The reactor vessel pressure decreased to zero in approximately 30 seconds on the simulator; the FSAR predicts that the pressure will reach zero in about 15 seconds. The simulator drywell pressure reached a maximum of approximately 21 psi; the FSAR predicts a peak of about 33 psi. The simulator drywell pressure peaks in approximately 20 seconds; the FSAR predicts that drywell pressure will peak in about 2 seconds. The discrepancy between the simulator performance and the FSAR prediction is traceable to the vessel blowdown rate, and lower plenum flashing. These problems are being addressed by the software support group. As a practical matter, this has no effect on training, since the operator doesn't have time to analyze this type of data until all immediate actions are taken, and the plant condition is stabilized. This discrepancy does not mask any valid indications, nor does it produce any false indications. No operator or automatic protective actions are affected.

(9) Maximum size unisolable main steam line rupture The simulator performance correl ated very wel 1 with the FSAR predi cti on.

(10) Simultaneous closure of all main steam isolation valves combined with single stuck open safety/relief valve (inhibit activation of high pressure Emergency Core Cooling Systems)

This event was not analyzed in the FSAR, and it has never occurred in the reference plant. The simulator performance correlated very closely with the FSAR prediction for MSIV Closure. When the simulator relief valve stuck open, the pressure response departed from the FSAR predicti on as expected. For the duration of the event, all parameters showed smooth, predictable response.

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 In addition, the following five events were tested:

Turbine trip without bypass 21 Recirculation pump stall (seizure)

3. Inadvertant actuation of an EMRV Safety valve actuation
5. EPR/MPR failure (reactor pressure decreases)

Although these transients are not part of the testing which is required for this report, they were performed because FSAR analysis data is available for them. In all five transients, the simulator performance correlated very closely with the FSAR prediction.

D. Malfunction Tests All malfunctions were tested during the Customer Acceptance Test at the manufacturer's facility during the first half of 1984. Approximately 30 percent of the malfunctions were re-tested for this report. See Attachment 7 for a list of all malfunctions verified during this testing period. The mal functions to be tested were chosen to include all generic mal functions required to be simul ted by ANSI/ANS 3.5 Regulatory Gui de 1.149. Minor discrepancies were noted during the 1986 testing. These discrepancies were documented and are scheduled for correction in accordance with the discrepancy reporting system.

IV. DISCREPANCY RESOLUTION AND UPDATING Discrepancies, modifications, and enhancements are all addressed by the Niagara Mohawk Power Corporation Nuclear Training Instruction 4.5.3, "Simulator Configuration Management." Discrepancy reports may be initiated by any individual utilizing the simulator for testing or training. Plant modifications are tracked by a computerized configuration management system.

10-

0 GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 SYSTEMS FULLY SIMULATED

1. Nuclear Boiler and Instrumentation
2. Reactor Recirculation System
a. Reactor Recirculation Loops
b. Boiler Process Instrumentation
c. Recirculation Flow Control
3. Control Rod Drive and Hydraulics System (CRDHS)
4. Reactor Manual Control System (RMCS)
5. Reactor Core (Physics and Thermodynamics)
a. Reactor Core Neutron Kinetics
b. Reactor Core Thermodynamics
6. Rod Worth Minimizer (RWM)
7. Main Steam Systems
a. Main Steam and Main Steam Bypass Systems
b. Moisture Separators-Reheaters
c. Extraction Steam System
d. Auxiliary Steam System
8. Reactor Water Cleanup System
9. Nuclear Instrumentation System
a. Source Range Monitor (SRH) System
b. Intermediate Range Monitor (IRM) System
c. Local Power Range Monitoring (LPRM) System
d. Average Power Range Monitoring (APRM) System
e. Rod Block Monitor (RBH) System
f. Traversing In-Core Probe (TIP) System
10. Reactor Protection System
11. Simulation of the Primary Containment and Isolation System
a. Primary Containment
b. Primary Containment Isolation System
12. Secondary Containment
13. Emergency Ventilation
a. Reactor Building Ventilation
b. Turbine Building Ventilation
c. Bui 1 ding Venti 1 ati on
14. Primary Containment Atmosphere Control and Sampling System

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986

15. Emergency Core Cooling Systems
a. Automatic Depressurization and Pressure Relief System
b. Core Spray
c. High Pressure Coolant Injection (HPCI) System
d. Containment Spray
e. Emergency Cooling System
16. Shutdown Cooling
17. Standby Liquid Control (SLC) System
18. Condensate and Feedwater System
a. Condensate System
b. Condensate Demineralizer System
c. Feedwater System
d. Condensate Storage and Transfer System
e. Reactor Vessel Level Control System
f. Feedwater Heaters, Vents and Drains t
19. Off-Gas Recombiner and Condenser Air Removal
20. Main Condenser
21. Circulating Water System
22. Reactor Building Closed Loop Cooling
23. Turbine Building Closed Loop Cooling
24. Service Water System
25. Instrument, Service, and Breathing Air
26. Area Radiation Monitoring System
27. Process Radiation Monitoring System
28. Venti 1 ati on Radi ati on Monitoring System
29. Main Turbine and Turbine Control
a. Turbine Oil System
b. Turbine Kinemat i cs
c. Turbine Mechanics
d. Turbine Supervisory and Safety System
e. Gland Seal System
f. Low Pressure Hood Spray System
g. Moisture Separator and Reheat System
h. Main Turbine Electro-Hydraulic Control System Al-2

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986

30. Plant Electrical System
a. Main Generator and Auxiliary Systems
1) Main Generator Synchronous Machine
2) Excitation and Voltage Regulator System
3) Synchroscope
4) Hydrogen Cooling System
5) Stator and Iso-Phase Duct Cooling System
6) Hydrogen Seal Oil System
b. Electrical Distribution System
1) Buses and Transformers
2) Breakers
3) Currents, Voltages, and Frequencies
4) DC Electrical Distribution and Control
5) Power System Electrical Grid
c. Di esel Generators
31. Containment Atmosphere Dilution, Vent and Purge System
32. Radiation Waste Disposal System Containment Equipment and Floor Drain Sump
33. Plant Carbon Dioxide System
34. Diesel Fire Pump and Pressurized Water Fire System
35. Fire Control Ventilation Systems
36. Control Room Heating, Ventilation, and Air Conditioning
37. Communi cat i on System
38. Plant Process Computer System
a. Appl i cabl e Experi ence
39. Meteorol ogi cal Experi ence
40. Plant Annunciators and Fire System Alarm Al-3

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 INITIAL CONDITIONS Ol COLD SHUTDOWN 02 COLD SHUTDOWN, AUXILIARY SYSTEMS RUNNING 03 STARTUP, 10 RODS SUBCRITICAL 04 9 RODS SUBCRITICAL SEQUENCE A 05 15 RODS SUBCRITICAL 06 HOT SCRAM RECOVERY 07 STARTUP, 75 PSIG 08 STARTUP, 375 PSIG 09 STARTUP, 800 PSIG 10 3.5 BYPASSES - COLD TURBINE 11 HOT TURBINE, 3.5 BYPASSES OPEN 12 NUMBER 13 FEEDWATER PUMP STARTUP 13 50~~ POWER 14 FULL POWER 15 FULL POWER ALL RODS OUT 16 POWER DECREASE FROM 100$ POWER 17 FOUR HOURS AFTER SCRAM 18 PRECOND ITIONING 19 20 A2-1

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 NINE MILE POINT UNIT ONE MALFUNCTIONS ADOl ADS FAILURE TO INITIATE - PRIMARY VALVES AD02 ADS FAILURE TO INITIATE - COMPLETE AD03 SOLENOID ACTUATED PRESSURE RELIEF VALVE (gill) FAILURE - SOLENOID AD04 SOLENOID ACTUATED PRESSURE RELIEF VALVE (gill) FAILURE - VALVE LEAKS AD05 SOLENOID ACTUATED PRESSURE RELIEF VALVE (gill) FAILURE - OPENS INADVERTENTLY AD06 SOLENOID ACTUATED PRESSURE RELIEF VALVE (gill) FAILURE - STUCK OPEN AN01 CONTROL ROOM ANNUNCIATOR SYSTEM FAILURE CS01 CORE SPRAY PUMP TRIP (111, 112, 121, 122, OR ANY)

CS02 CORE SPRAY TOPPING PUMP TRIP (ill, 112, 121, 122, OR ANY)

CS03 CORE SPRAY INBOARD INJECTION VALVE FAILURE TO OPEN (IV40-01, IV40-09, IV40-11, IV40-10, OR ANY)

CTOl CONTAINMENT SPRAY PUMP TRIP (111, 112, 121, 122, OR ANY)

CT02 CONTAINMENT SPRAY RAW WATER PUMP TRIP (111, 112, 121, 122, OR ANY)

CT03 CONTAINMENT SPRAY HEAT EXCHANGER (ill, 112, OR BOTH) TUBE LEAK CUOl COOLANT LEAKAGE INSIDE PRIMARY CONTAINMENT CU02 REACTOR WATER CLEANUP PUMP TRIP (11, 12, OR BOTH)

CU03 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-N022) FAILS OPEN CU04 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-ND22) FAILS CLOSED'EACTOR CU05 WATER CLEANUP HIGH PRESSURE CONTROL VALVE (PCV 33-39) FAILS OPEN CU06 REACTOR WATER CLEANUP HIGH PRESSURE CONTROL VALVE (PCV 33-39) FAILS CLOSED CU07 REACTOR WATER CLEANUP LOW PRESSURE CONTROL VALVE (PCV-ND37) FAILS OPEN CU08 REACTOR WATER CLEANUP LOW PRESSURE CONTROL VALVE (PCV-N037) FAILS CLOSED CU09 REACTOR WATER CLEANUP NON REGENERATIVE HEAT EXCHANGER TUBE LEAK CU10 REACTOR WATER CLEANUP DEMINERALIZER RESIN DEPLETION (ll, 12, OR BOTH)

A3-1

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 CU11 COOLANT LEAKAGE OUTSIDE PRIMARY CONTAINMENT CW01 HIGH RADIATION IN SERVICE WATER CW02 SERVICE WATER PUMP TRIP (11, 12, OR BOTH)

CW03 EMERGENCY SERVICE WATfR PUMP TRIP (11, 12, OR BOTH)

CW04 REACTOR BUILDING CLOSED LOOP COOLING (11, 12, 13, OR ANY) PUMP TRIP CW05 TURBINE BUILDING CLOSED LOOP COOLING PUMP TRIP (ll, 12, OR BOTH)

CW06 CIRCULATING WATER PUMP TRIP (11, 12, OR BOTH)

CW07 CIRCULATING WATER EXPANSION JOINT LEAKAGE CWOB CIRCULATING WATER INTAKE STRUCTURE ICING CW09 LOSS OF DRYWELL COOLING CW10 MAIN CONDENSER TUBE LEAK DG01 DIESEL GENERATOR FAILURE TO START (102, 103, OR BOTH)

DG02 DIESEL GENERATOR TRIP (102, 103, OR BOTH)

EC01 STfAM LEAKAGE INSIDE PRIMARY CONTAINMENT EC02 STEAM LEAKAGE OUTSIDE PRIMARY CONTAINMENT EC03 EMERGENCY COOLING SYSTEM RETURN VALVE FAILS OPEN (IV39-05, IV39-06, OR BOTH)

EC04 EMERGENCY COOLING SYSTEM RETURN VALVE FAILS TO OPEN (IV39-05,IV39-06, OR BOTH)

EC05 EMERGENCY COOLING SYSTEM EMERGENCY CONDENSER MAKEUP CONTROL VALVE FAILS CLOSED (LCV60-17, LCV60-18, OR BOTH)

EC06 fMERGENCY CONDENSER TUBE LEAK (111, 121, OR BOTH)

ED01 LOSS OF OFFSITE 115 KV POWER SOURCES (LIGHTHOUSE HILL-JAF, OSWEGO STEAM, OR BOTH)

ED02 BATTERY CHARGER AND EMERGENCY LIGHTING SUPPLY MOTOR GENERATOR TRIPS (161, 171, OR BOTH)

ED03 COMPUTfR POWER SUPPLY MOTOR GENERATOR TRIPS (167) 1 ED04 AC POWERBOARD ELECTRICAL FAULT (PB11)

ED05 AC POWERBOARD ELECTRICAL FAULT (PB12)

A3-2

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 E006 AC POWERBOARD ELECTRICAL FAULT (P8101)

ED07 AC POWERBOARD ELECTRICAL FAULT (PB102)

ED08 AC POWERBOARD ELfCTRICAL FAULT (P8103)

ED09 AC POWERBOARD ELECTR ICAL FAULT (P813 SECTION A)

ED10 AC POWERBOARD ELECTRICAL FAULT (PB13 SECTION 8)

ED11 AC POWERBOARD ELECTRICAL FAULT (P813 SECTION C)

ED12 AC POWERBOARD ELECTRICAL FAULT (PB14 SECTION A)

ED13 AC POWfRBOARD ELECTRICAL FAULT (PB14 SECTION 8)

F.D14 AC POWERBOARO ELECTRICAL FAULT (PB14 SECTION C)

ED15 AC POWERBOARD ELECTRICAL FAULT (P815 SECTION A)

ED16 AC POWERBOARD ELECTRICAL FAULT (PB15 SECTION 8)

ED17 AC POWERBOARD ELECTRICAL FAULT (PB15 SECTION C)

ED 18 AC POWERBOARD fLECTRICAL FAULT (PB16 SECTION A)

ED19 AC POWERBOARD fLECTRICAL FAULT (PB16 SECTION 8)

ED20 AC POWERBOARD ELECTRICAL FAULT (P817 SECTION A)

ED21 AC POWERBOARD ELECTRICAL FAULT (PB18 S'ECTION 8)

ED22 DC POWERBOARD ELECTRICAL FAULT (11, 12, OR BOTH)

ED23 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - NORHAL ED24 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - ALTERNATE ED25 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - NORMAL AND ALTERNATE EG01 HAIN GENERATOR TRIP - ELECTRICAL FAULT EG02 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS - INCREASE EG03 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS - DECREASE MAIN GENERATOR CORE INTERNAL HEATING EGOS HAIN TRANSFORMER LOSS OF COOLING EG06 GENERATOR HYDROGEN COOLING SYSTEH LEAKAGE A3-3

'0 GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 EG07 GENERATOR HYDROGfN MAIN SEAL OIL PUMP FAILURE EG08 GENERATOR HYDROGEN EMERGENCY SEAL OIL PUMP FAILURE EG09 STATOR COOLING PUMP TRIP. (11, 12, OR BOTH)

EG10 LOSS OF CONTROL AIR TO 345 KV BREAKER (R-915, R-925, OR BOTH)

EGll POWER GRID NETWORK LOAD TRANSIENT - INCREASE EG12 POWER GRID NETWORK I DAD TRANSIENT - DECREASE EG13 STATOR WATER COOLING DEMINERALIZER RESIN DEPLETION FP01 DIESEL FIRE PUMP FAILURE FP02 ELECTRIC FIRE PUMP FAILURE FP03 AC FOAM PUMP FAILURE FP04 DC FOAM PUMP FAILURE FP05 TURBINE ISLAND FIRE DETECTION (D-1195, 0-1155, 0-1165, 0>>1175, D-1061, DA-1114, DA-1131, OR AN Y)

FP06 CONTROL ROOM FIRE DETECTION (FIRE PANEL 2, CONTROL CONSOLE, "L" PANEL, "K" PANEL, "H" PANEL, "F" PANEL, "A" PANEL, OR ANY)

FP07 TURBINE BUILDING FIRE DfTECTION (DA-22092MG, DA-2083M, DA-2081S, DA2092E, 0-2102, OR ANY)

FP08 DIESEL ROOM FIRE DETECTION (DX-2113A, DX-2113B, DX-2141A, DA-2141, DX-2151B, DA-2151, 0-2151, OR ANY)

FP09 AUXILIARY CONTROL ROOM/CABLE SPREADING ROOM FIRE DETECTION (D-3031PL, DX-3031A, DX-30118, WD-8131, WD-8082, OR ANY)

FP10 REACTOR BUILDING FIRE DETECTION (DX-4217A, DA-4116W, DA-4076E, 0-4207, D-4156, SP-4126, 0-4086, OR ANY)

FW01 CONDENSATE PUMP TRIP (11, 12, 13, OR ANY)

FW02 FEEDWATfR BOOSTER PUMP TRIP (11, 12, 13, OR ANY)

FW03 FEEDWATER PUMP TRIP (11, 12, OR BOTH)

FW04 SHAFT DRIVEN FEEDWATER PUMP 13 FAILURE FW05 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAILURE TO fNGAGE FWn6 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAILURE TO DISfNGAGE FW07 FEEDWATER CONTROL VALVE 11 CONTROLLER FAILS HIGH A3-4

'0 GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 FEfDWATER CONTROL VALVE 11 CONTROLLER FAILS LOW FW09 FEEDWATfR CONTROL VALVE 12 CONTROLLER FAILS HIGH FW10 FEfDWATER CONTROL VALVE 12 CONTROLLER FAILS LOW FW11 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS HIGH FW12 FEEDWATER CONTROL VALVE 13'ONTROLLER FAILS LOW FW13 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS AS IS FW14 FEEDWATER MASTER CONTROLLER FAILS HIGH FW15 FEEDWATER MASTER CONTROLLER FAILS LOW FW16 FEEDWATER MASTER CONTROLLER FAILS AS IS FW17 CONDENSATE DEMINERALI ZER DEPLETION FW18 FEEDWATER CONDUCTIVITY INCREASE FW19 CONDENSATE RECIRCULATION VALVE (FCV 50-24) FAILS OPEN FW20 CONDENSATE RECIRCULATION VALVE (FCV 50-24) FAILS CLOSED FW21 FEEDWATER BOOSTER PUMP RfCIRCULATION VALVE FAILS OPEN (FCV 51-58, FCV 51-59, FCV 51-60, OR ANY)

FW22 FEEDWATER HEATER TUBE"LEAK FW23 FEEDWATER PUMP RECIRCULATION VALVES FAIL OPEN (11, 12, 13, OR ANY)

FW24 FEEDWATER CONTROL VALVE FAILS CLOSED (13A, 138, OR BOTH)

FW25 THREE MILE ISLAND ACCIDENT (BWR EQUIVALENT)

FW26 CONDENSATE BYPASS SPRAY TO MAIN CONDENSER FLOW CONTROL VALVE (FCV 50-22) FAILS CLOSED FW27 LOSS OF COMPENSATION TO FEEDWATER FLOW TRANSMITTER F W28 HPCI MODE FAILURE TO INITIATE (11, 12, OR'OTH)

FW29 HPCI MODE INADVERTANT INITIATION (ll, 12, OR BOTH)

HV01 REACTOR BUILDING EXHAUST FAN TRIP (11, 12, OR BOTH)

HV02 EMERGENCY VENTILATION FAN TRIP (ll, 12, OR BOTH)

IA01 LOSS OF INSTRUMENT AIR LP01 LIQUID POISON PUMP TRIP (A, B, OR BOTH)

A3-5

'0 G'ENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1 1986 Mcol HAIN CONDENSER AIR, INLEAKAGE HC02 STEAM JET AIR EJECTOR STEAM SUPPLY VALVE FAILS CLOSED MC03 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL HIGH MC04 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL LOW MC05 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL AS IS MC06 EXPLOSION IN AIR EJECTOR DISCHARGE PIPING HS01 STEAM LEAK RUPTURE OUTSIDE PRIMARY CONTAINMENT (DESIGN BASIS)

MS02 MSIV DISC SEPARATES FROM STEM HS03 ONE MSIV FAILS CLOSED (VALVE 122)

HS04 STEAM LINE RUPTURE INSIDE PRIHARY CONTAINMENT (DESIGN BASIS)

MSOS TURBINE STEAM SEAL REGULATOR FAILS CLOSED MS06 MOISTURE SEPARATOR DRAIN TANK LEVEL CONTROL FAIl S LOW HS07 FIRST STAGE REHEATER 111 STEAM SUPPLY VALVE CLOSES SECOND STAGE REHEATER 112 STEAM SUPPLY VALVE CLOSES HS09 SECOND STAGE REHEATER 112 DRAIN TANK LEVEL CONTROL FAILS LOW MS10 LOSS OF EXTRACTION STEAM TO HIGH PRESSURE FEEDWATER HEATER (115, 125, 135, OR ANY)

. MS11 LOSS OF COMPENSATION TO STEAM FLOW TRANSMITTER NM01 SRM CHANNEL (ll, 12, 13, 14, OR ANY) FAILURE - UPSCALE NM02 SRM CHANNEL (11, 12, 13, 14, OR ANY) FAILURE - DOWNSCALE NM03 SRH CHANNEL RECORDER FAILURE (RED, BLACK, OR BOTH PENS)

NM04 SRM CHANNEL (11, 12, 13, 14, OR ANY) FAILURE - INOPERATIVE NM05 SRM CHANNEL (11, 12, 13, 14, OR ANY) FAILURE - UPSCALE, RECORDER INOPERATIVE NM06 SRM CHANNEL (11, 12, 13, 14, OR ANY) FAILURE - DOWNSCALE NM07 SRM CHANNEL (11, 12, 13, 14, OR ANY) FAILURE - RECORDER NM08 SRM CHANNEl (11, 12, 13, 14, OR ANY) FAILURES- INOPERATIVE A3-6

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 NM09 SRM CHANNEL (11, 12, 13, 14, OR ANY) DETECTOR STUCK NM10 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-UPSCALE NMll IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-DOWHSCALE NM12 IRM/APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-RECORDER NM13 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-INOPERATIVE NM14 IRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18, OR ANY) fAILURE - UPSCALE NM15 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-DOWNSCALE IRM/APRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-RECORDER HM17 IRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18, OR ANY) INOPERATIVE NM18 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) DETECTOR STUCK NM19 APRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE - UPSCALE HM20 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-DOWHSCALE NM21 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE '-

INOPERATIVE NM22 APRM CHANNEL (11, 12', 13, 14, 15, 16, 17, 18, OR ANY) FAILURE - UPSCALE NM23 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-DOWHSCALE NM24 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18, OR ANY) FAILURE-INOPERATIVE ANY LPRM (X-Y-J) FAILURE - UPSCALE NM26 ANY LPRM ( X-Y-J) FAILURE - UPSCALE NM27 ANY LPRM (X-Y-J) FAILURE - UPSCALE NN28 ANY LPRM (X-Y-J) fAILURE - DOWNSCALE HM29 ANY LPRM (X-Y-J) FAILURE - DOWNSCALE NM30 ANY LPRM (X-Y-J) FAILURE - DOWNSCALE A3-7

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 NM31 ANY LPRM (X-Y-JO FAILURE - DOWNSCALE NM33 TIP DETECTOR STUCK IN CORE NM34 ANY LPRM (X-Y-J) DRIFT +/- 25$

NM35 ANY LPRM (X-Y-J) DRIFT +/- 25$

NM36 RECIRC FLOW CONVERTER CHANNEL (ll, 12, OR BOTH) FAILURE - UPSCALE NM37 RECIRC FLOW CONVERTER CHANNEL (11, 12, OR BOTH) FAILURE - DOWNSCALE NM38 RECIRC FLOW CONVERTER CHANNEL (11, 12, OR BOTH) FAILURE - AS IS NM39 RECIRC FLOW CONVERTER CHANNEL (11, 12, OR BOTH) FAILURE - INOPERATIVE NM40 RECIRC FLOW CONVERTER (11, 12, OR BOTH) FAILURE - COMPARATOR OG01 OFF GAS RECOMBINER PREHEATER STEAM SUPPLY FAILS CLOSED OG02 OFF GAS RECOMBINER MIXING JET STEAM SUPPLY FAILS OPEN OG03 OFf GAS RECOMB INER MIXING JET STEAM SUPPLY FAILS CLOSED OG04 OFF GAS DISCHARGE TO STACK ISOLATION VALVE FAILS CLOSED PC01 DRYWELL-TORUS DIFFERENTIAL PRESSURE CONTROL FAILURE - INCREASE PC02 DRYWELL-TORUS DIFFERENTIAL PRESSURE CONTROL FAILURE - DECREASE PC03 PRIMARY CONTAINMENT LEAKAGE PP01 FAILURE OF PLANT PROCESS COMPUTER RD01 CONTROL ROD XX-YY FAILURE - DRIFT IN RD02 CONTROL ROD XX-YY fAILURE - DRIFT OUT RD03 CONTROL ROD XX-YY FAILURE - ACCUMULATOR STUCK RD04 CONTROL ROD XX-YY FAILURE STUCK RD05 CONTROL ROD XX-YY FAILURE - UNCOUPLED RD06 CONTROL ROD XX-YY FAILURE - SCRAMMED RD07 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME R008 CONTROL ROD XX-YY FAILURE - RPIS RD09 CONTROL ROD XX-YY FAILURE - DRIFT IN RA10 CONTROL ROD XX-YY FAILURE - DRIFT OUT A3-8

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 R011 CONTROL ROD XX-YY FAILURE - ACCUMULATOR TROUBLE R012 CONTROL ROD XX-YY FAILURE - STUCK RD13 CONTROL ROD XX-YY FAILURE - UNCOUPLED R014 CONTROL ROD XX-YY FAILURE - SCRAMMED RD15 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME R016 CONTROL ROD XX-YY FAILURE - RPIS RD17 CONTROL ROD XX-YY FAILURE - DRIFT IN RD18 'CONTROL ROD XX-YY FAILURE - DRIFT OUT R019 CONTROL ROD XX-YY FAILURE - ACCUMULATOR TROUBLE RD20 CONTROL ROD XX-YY FAILURE - STUCK R021 CONTROL ROD XX-YY FAILURE - UNCOUPLED R022 CONTROL ROD XX-YY FAILURE - SCRAMMED RD23 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME R024 CONTROL ROD XX-YY FAILURE - RPIS R025 CONTROL ROD XX-YY FAILURE - DRIFT IN R026 CONTROL ROD XX-YY FAILURE - DRIFT OUT R027 CONTROL ROD XX-YY FAILURf - ACCUMULATOR TROUBLE RD28 CONTROL ROD XX-YY FAILURE - STUCK RD29 CONTROL ROD XX-YY FAILURE - UNCOUPLED RD30 CONTROL ROD XX-YY FAILURE - SCRAMMED Rll31 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME R032 CONTROL ROD XX-YY FAILURE - RPIS RD33 CONTROL ROD BANK FAILURf TO SCRAM (BANK I, II, III, IV, V, OR ANY)

RD34 LOSS OF CRD INSTRUMfNT AIR PRfSSURE R035 CRD HYDRAULIC PUMP TRIP (11, 12, OR BOTH)

R036 CRD FLOW CONTROL VALVE FAILURE - CLOSED (11, 12, OR BOTH)

A3-9

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 RD37 RPIS FAILURE - COMPLfTE SYSTf M FAILURE RD38 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION - WITHDRAWN RD39 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION - INSERT RD40 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION - SETTLE RD41 SCRAM DISCHARGE VOLUME RUPTURE RMOl DRAWER INOPERATIVE FOR ANY PROCESS RADIATION MONITOR SIMULATED

( INSTRUCTOR SELECT)

RM02 DRAWER DOWNSCALE FOR ANY ARfA RADIATION MONITOR SIMULATED (INSTRUCTOR SELECT)

RM03 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM04 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM05 AIR MONITOR FAILURE (TURBINE BUILDING, REACTOR BUILDING,

'I'ONTINUOUS WASTE BUILDING, DRYWELL)

RM06 ANY PROCESS RADIATION MONITOR FAILURE RP01 REACTOR TRIP POWER SUPPLY MOTOR GENERATOR (131, 141, OR BOTH)

RP02 CONTROL POWER SUPPLY BOTH MOTOR GENERATOR TRIPS (162, 172, OR BOTH)

RP03 REACTOR SCRAM RP04 REACTOR PROTECTION SYSTEM FAILURE TO SCRAM - AUTOMATIC RP05 REACTOR PROTECTION SYSTEM FAILURf TO SCRAM - COMPLETE RP06 REACTOR VESSEL ISOLATION RP07 PRIMARY CONTAINMENT ISOLATION RP08 ANTICIPATfD TRANSIENT WITHOUT SCRAM (ATWS)

R P09 EMERGENCY CONDENSER FAILS TO ISOLATE ( 11, 12, OR BOTH)

RR01 RECIRCULATION PUMP 11 DRIVE BREAKER TRIP RR02 RECIRCULATION PUMP 11 FIELD BREAKER TRIP RR03 RECIRCULATION PUMP 11 SEIZURE RR04 RECIRCULATION PUMP 11 CONTROL SIGNAL FAILURE RR05 RECIRCULATION PUMP 11 INCOMPLETE START SEQUENCE A3-10

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 RR06 RECIRCULATION PUMP 12 DRIVE BREAKER TRIP RR07 RECIRCULATION PUMP 12 FIELD BREAKER TRIP RR08 RECIRCULATION PUMP 12 SEIZURE RR09 RECIRCULATION PUMP 12 CONTROL SIGNAL FAILURE RR10 RECIRCULATION PUMP 12 INCOMPLETE START SEQUENCE RRll RECIRCULATION PUMP 13 DRIVE BREAKER TRIP RR12 RECIRCULATION PUMP 13 FIELD BREAKER TRIP RR13 RECIRCULATION PUMP 13 SEIZURf RR14 RECIRCULATION PUMP 13 CONTROL SIGNAL FAILURE RR15 RECIRCULATION PUMP 13 INCOMPLETE START SEQUENCE RR16 RECIRCULATION PUMP 14 DRIVE BREAKER TRIP RR17 RECIRCULATION PUMP 14 FIELD BREAKER TRIP RR18 RECIRCULATION PUMP 14 SEIZURE RR19 RECIRCULATION PUMP 14 CONTROL SIGNAL FAILURE RR20 RECIRCULATION PUMP 14 INCOMPLETE START SEQUfNCE RR21 RECIRCULATION PUMP 15 DRIVE BREAKfR TRIP RR22 RECIRCULATION PUMP 15 FIELD BREAKfR TRIP RR23 RfCIRCULATION PUMP 15 SEIZURE RR24 RECIRCULATION PUMP 15 CONTROL SIGNAL FAILURE RR25 RECIRCULATION PUMP 15 INCOMPLETE START SEQUENCE RR26 MASTER RECIRCULATION FLOW CONTROLLER FAILURE - HIGH RR27 MASTER RECIRCULATION FLOW CONTROLLER FAILURE LOW RR28 MASTER RECIRCULATION FLOW CONTROLLER FAILURE - AS IS RR30 REACTOR VESSEL PRfSSURE RECORDER FAILURE (ID77) - UPSCALE RR31 REACTOR VESSEL PRESSURE RECORDER FAILURE (ID77) - DOWNSCALE RR32 REACTOR VESSEL PRESSURE RECORDER FAILURE (ID77) - AS IS RR33 RECIRCULATION PUMP LOWER (INNER) SEAL FAILURE - PUMP 11 A3-11

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 RR34 RECIRCULATION PUMP UPPER (OUTER) SEAL FAILURE - PUMP ll RR35 REACTOR VESSEL PRfSSURE INDICATOR FAILURE (ID76C) - UPSCALE RR36 REACTOR VESSEL PRESSURE INDICATOR FAILURE (ID76C) - DOWNSCALE RR37 REACTOR VESSEL PRESSURE INDICATOR FAILURE (ID76C) - AS IS RR38 REACTOR VESSEL LEVfL RECORDER FAILURE (ID14) - UPSCALE RR39 REACTOR VESSEL LEVEL RfCORDER FAILURE ( Ill14) - DOWNSCALE RR40 REACTOR VESSEL LEVEL RECORDER FAILURE (ID14) - AS IS RR41 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE - UPSCALE

( ID59D)

RR42 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE - DOWNSCALE

( ID59D)

RR43 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE - AS IS

( ID59D)

RR44 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE-UPSCALE (LI 36-19, CH.12)

RR45 RfACTOR VfSSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE-DOWNSCALE (LI 36-19, CH.12)

RR46 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE-AS IS (LI 36-19, CH.12)

RR47 RECIRCULATION PUMP DISCHARGE VALVE STEM SEPARATES FROM VALVE GATE (11,12,13,14,15,0R ANY)

RR48 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE-UPSCALE RR49 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFfTY SYSTEM) FAILURE-DOWNSCALE RR50 REACTOR VfSSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE-AS IS RR51 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTfM INPUT) FAILS - HIGH RR52 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - LOW RR53 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - AS IS A3-12

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 3 March 1, 1986 RR54 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-HIGH RR55 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-LOW RR56 'REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-AS IS RR57 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - HIGH RR58 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - LOW RR59 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS AS IS RR60 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS - HIGH RR61 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS - LOW RR62 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS - AS IS RR63 REACTOR RECIRCULATION PUMP 12 INNER SEAL FAILURE RR64 REACTOR RECIRCUl.ATION PUMP 12 OUTER SEAL FAILURE RR65 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - HIGH RR66 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - LOW RR67 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - OSCILLATES RR68 REACTOR RECIRCULATION PUMP M/A STATION FAILURE - INCREASE (11, 12, 13, 14, 15, OR ANY)

RR69 REACTOR RECIRCULATION PUMP M/A STATION FAILURE - DECREASE (11, 12, 13, 14, 15, OR ANY)

RR70 REACTOR RECIRCULATION PUMP M/A STATION FAILURE - AS IS (11, 12, 13, 14, 15, OR ANY)

RR71 REACTOR SAFETY VALVE INADVERTENTLY OPENS (PSV NR28A)

RR72 LOSS OF LEVEL COMPENSATION TO FEEDWATER CONTROL SYSTEM (GEMAC) LEVEL TRANSMITTER RW01 ROD WORTH MINIMIZER FAILURE A3-13

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 RX01 FUE L CLADDING FAILURE RX02 INCREASED ROD WORTH FOR ANY CONTROL ROD SC01 SHUTDOWN COOLING PUMP TRIP (11, 12, 13, OR ANY)

SC02 SHUTDOWN COOLING HEAT EXCHANGER TUBE LEAK (ll, 12, 13, OR ANY)

TC01 MAIN TURBINE TRIP TC02 TURBINE GOVERNOR FAILS - HIGH TC03 TURBINE GOVERNOR FAILS - LOW TC04 ELECTRICAL PRESSURE REGULATOR FAILS - HIGH TC05 ELECTRICAL PRESSURE REGULATOR FAILS - LOW TC06 ELECTRICAL PRESSURE REGULATOR FAILS - OSCILLATES TC07 MECHANICAL PRESSURE REGULATOR FAILS - HIGH TCOB MECHANICAL PRESSURE REGULATOR FAILS - LOW TC09 MECHANICAL PRESSURE REGULATOR FAILS - OSCILLATES TC10 FIRST BYPASS VALVE STICKS OPEN TC11 ALL BYPASS VALVES FAIL - OPEN TC12 ALL BYPASS VALVES FAIL - CLOSED TC13 TURBINE CONTROL VALVE FAILS CLOSED (11, 12, 13, 14, OR ANY)

TU01 EXHAUST HOOD SPRAY VALVE FAILS CLOSED TU02 MAIN TURBINE HIGH VIBRATION BEARINGS g5 AND g6 TU03 MAIN TURBINE HIGH ECCENTRICITY TU04 MAIN TURBINE BEARING OIL LOW PRESSURE TU05 MAIN TURBINE BEARING HIGH TEMPERATURE TU06 MAIN TURBINE THRUST BEARING WEAR A3-14

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 REMOTE FUNCTIONS AD ADS NONE AN ANNUNCIATOR SYSTEM NONE CS CORE SPRAY NONE CT CONTAINMENT SPRAY RCT 1 80-43 TEST LINE TO TORUS BV OPEN CLOSE RCT 2 80-42 WASTE DISP MAN ISOLATION OPEN CLOSE CU REACTOR CLEANUP RCU1 CU-16 PCV ND37 MANUAL ISOLATION OPEN CLOSE RCU2 CU-19 FILTER BYPASS VALVE OPEN CLOSE RCU3 CU FILTER 11 INLET/OUTLET VALVES OPEN CLOSE RCU4 CU FILTER 12 INLET/OUTLET VALVES OPEN CLOSE RCU5 CU DEMIN ll INLET/OUTLET VALVES OPEN CLOSE RCU6 CU DEMIN 12 INLET/OUTLET VALVES OPEN CLOSE RCU7 CU-20 DEMIN BYPASS VALVFS OPEN CLOSE CW1 AUXILIARY WATER RCW1 INTAKE WATER TEMPERATURE 32/80 DEG 75. 00 RCW2 INTAKE TUNNEL REVERSE FLOW YES NO RCW3 UPPER WIND SPEED 0.100 MPH 52.00 RCW4 UPPER WIND SPEED VARIATION 0/30 MPH 5.00 RCW5 LOWER WIND SPEED 0/100 MPH 45.00 RCW6 LOWER WIND SPEED VARIATION 0/30 MPH 5.00 RCW7 UPPER WIND DIRECTION 0/360 MPH 5.00 RCW8 UPPER WIND DIRECTION VARIATION 0/90 DEG 150.00 RCW9 LOWER WIND DIRECTION 0/360 DEG 150.00 RCW10 LOWER WIND DIRECTION VARIATION 0/90 DEG 5.00 RCWll AMBIENT AIR TEMPERATURE -30/+120 DEG 90.00 RCW12 DELTA TEMPERATURE -10/+120 DEG 10.00 A4-1

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 CW2 AUXILIARY WATER DG DIESEL GENERATOR ROG1 DG 102 GOVERNOR SPEED DROOP SET RESET ROG2 DG 103 GOVERNOR SPEED DROOP SET RESET EC EMERGENCY COOLING REC1 IV 39-05 VALVE POSITION LIMIT 0/100$ 100.00 REC2 IV 39-06 VALVE POSITION LIMIT 0/100$ 100.00 EDl ELECTRICAL DISTRIB. RED1 SOUTH OSWEGO 115 KV BKR R10 OPEN CLOSE RED2 FITZ 115 KV BKR R40 OPEN CLOSE RED3 PB 13 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE

'RED4 PB 13 BUS TIE BKR SEC 8-SEC C OPEN CLOSE RED5 PB 14 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE ED1 ELECTRICAL DISTRIB. RED6 PB 14 BUS TIE BKR SEC 8-SEC C OPEN CLOSE RED7 PB 15 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE REDB PB 16 BUS TIE BKR SEC 8-SEC C OPEN CLOSE REO9 MG-SET 167 AC POWER SELECT PB16 PB17 REO10 MG-SET 167 DC POWER SELECT P811 PB12 REDll COMPUTER POWER SUPPLY SELECT NORM EHER RED12 IC BUS 130 NORM PWR BKR OPEN CLOSE RED13 IC BUS 130 ALT PWR BKR OPEN CLOSE RED14 P81671 BUS TIE BKR OPEN CLOSE RED15 P8131 CLOSE A-B, OPEN 13A SUPPLY <<

YES NO RED16 PB131 CLOSE A-B, OPEN 13C SljPPLY YES NO RED17 PB141 CLOSE A-B, OPEN 14A SUPPLY YES NO RED18 P8141 CLOSE A-B, OPEN 14C SUPPLY YES NO ED2 ELECTRICAL DISTRIB. RED19 PB151 CLOSE A-B, OPEN 15A SUPPLY YES NO RED20 P8151 CLOSE A-B, OPEN 15C SUPPLY YES NO a RED21 PB176 CLOSE A-B, OPEN 17A SljPPLY YES NO RED22 P8176 CLOSE AOB, OPEN 16A SUPPLY YES NO RED23 BAT BD11 EQUIP SW TO ALT 8811 8812 A4-2

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 E02 ELECTRICAL DISTRIB. RED24 BAT B012 EQUIP SW TO ALT BB12 BB11 RE025 PB143 FEEDER BREAKER 14A 14C E03 ELECTRICAL DISTRIB. NONE EG1 MAIN GENERATOR REGl 345 KV BKR 100 42 OPEN CLOSE REG2 345 KB MAN DISC 917 OPEN CLOSE REG3 345 KV MAN DISC 926, 927 OPEN CLOSE REG4 345 KV MOD SW 18 OPEN CLOSE REGS 345 KV BKR R915/10 OPEN CLOSE REG6 345 KV BKR R925/20 OPEN CLOSE REG7 MAIN SEAL OIL PMP STATUS STAT NEUT REG8 EMER SEAL OIL PMP STATUS START NEUT REG9 EMER SEAL OIL PMP STATUS TRIP AUTO REG10 GEN STATOR COOLING PMP ll START NEUT REG11 GEN STATOR COOLING PMP ll TRIP AUTO REG12 GEN STATOR COOLING PMP 12 START NEUT REG13 GFN STATOR COOLING PMP 12 TR I P AUTO EG1 MAIN GENERATOR REG14 GENERATOR OUTPUT LINKS OPEN CLOSE REG15 GEN HYOROGEN SUPPLY VALVE OPEN CLOSE REG16 BACKFEED INTERLOCKS ON OFF EG2 MAIN GENERATOR NONE FP FIRE PROTECTION RFP1 CITY WATER SUPPLY TO FP HDR OPEN CLOSE RFP2 SUPPLY TO EMER COOL MU TANK'l OPEN CLOSE RFP3 SUPPLY TO EMER COOL MU TANK 12 OPEN CLOSE RFP4 SUPPLY TO FEEDWATER SYSTEM OPEN CLOSE RFPS DIESEL FIRE PUMP STATUS OFF AUTO A4-3

GENERAL PHYSICS CORPORATION G P-R-115007 Attachment 4 March 1, 1986 FWl-FEEDWATER RFWl 50-10 COND PUMP 11 DISCH VLV OPEN CLOSE RFW2 50-11 COND PUMP 12 DISCH VLV OPEN CLOSE RFW3 50-12 COND PUMP 13 DISCH VLV OPEN CLOSE RFW4 50-31 COND DEMIN BYPASS VLV OPEN CLOSE RFWS COND DEMIN ll INLET/OUTLET VLV OPEN CLOSE RFW6 COND DEMIN 12 INLET/OUTLET VLV OPEN CLOSE RFW7 COND DEMIN 13 INLET/OUTLET VLV OPEN CLOSE RFW8 COND DEMIN 14 INLET/OUTLET VLV OPEN CLOSE RFW9 COND DEMIN 15 INLET/OUTLET VLV OPEN CLOSE RFW10 COND DEMIN 16 INLET/OUTLET VLV OPEN CLOSE RFW11 50-20 SJAE BYPASS FCB 0/10(5 50.00 RFW12 50-40 BOOSTER PUMP 11 SUCTION V OPEN CLOSE RFW13 50-39 BOOSTER PUMP 12 SUCTION V OPEN CLOSE RFW14 50-38 BOOSTER PUMP 13 SUCTION V OPEN CLOSE RFW15 FW HEATER STRING ll ISOL VLVS OPEN CLOSE RFW16 FW HEATER STRING 12 ISOL VLVS OPEN CLOSE RFW17 FW HEATER STRING 13 ISOL VLVS OPEN CLOSE RFW18 DEMIN WATER STORAGE TANK REFILL OPEN CLOSE FW2 FEEDWATER RFW19 50-16 BYPASS AROUND FCV 50-22 OPEN CLOSE RFW20 MANUAL OPERATION OF LCV50-15 0/100$ 0.00 RFW21 MANUAL OPERATION OF LCV50-07,08 0/100$ 0.00 RFW22 FW HEATER 135 ISOL VALVFS OPEN CLOSE RFW23 HOTWELL LEVEL CONTROL MAN AUTO FW3 FEEDWATER NONE HV HVAC NONE IA INSTRUMENT AIR RIAl INST AIR SUP TO BREATHING AIR OPEN CLOSE R IA2 BRW-G-6 WASTE DISPOSAL XTIE OPEN CLOSE RIA3 94-42 CONT SPRAY AIR RCVR ISOL OPEN CLOSE R IA4 SERV AIR TO INST AIR BV TRIP RESET A4-4

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 LIQUID POISON RLP1 LIQ POISON PMP 11 LOCAL START ON OFF RLP2 LIQ POISON PMP 12 LOCAL START ON OFF RLP3 DEMIN WATER TO LP PUMPS OPEN CLOSE MC CONDENSER RMCl OG-1,2 PRIM JET VAP SUCT VALVES OPEN CLOSE RMC2 OG-3,4 PRIM JET VAP SUCT VALVES OPEN CLOSE RMC3 MS 14,15 PRIM JET STEAM VLAVES OPEN CLOSE RMC4 MS-16, 17 PRIM JET STEAM VALVES OPEN CLOSE RMC5 OG-9, 10 SEC JET VAPOR SUCT VALVES OPEN CLOSE RMC6 MS-19,20 SEC JET STEAM VALVES OPEN CLOSE RMC7 MS-12 SJAE PCV BYPASS OPEN 'LOSE MS1 MAIN STEAM RMS1 HP FW HTR 115 RESET TRIP RESET RMS2 HP FW HTR 125 RESET TRIP RESET RMS3 HP FW HTR 135 RESET TRIP RESET RMS4 HP FW HTR STRING ll RESET TRIP RESET RMS5 HP FW HTR STRING 12 RESET TRIP RESET RMS6 HP FW HTR STRING 13 RESET TRIP RESET RMS7 MS-8 MAIN STEAM LINE ISOL OPEN CLOSE RMSB SPE 11 SUCTION VALVE OPEN CLOSE RMS9 SPE 12 SUCTION VALVE OPEN CLOSE RMS10 TR IP ALL FW HTR EXTR NRVS TRIP RESET MS1 MAIN STEAM NONE NM1 NEUTRON MONITOR RNM1 APRM ll GAIN 0/100$ 2.43 RNM2 APRM 1'2 GAIN 0/100% 2. 38 RNM3 APRM 13 GAIN 0/100$ 2.36 RNM4 APRM 14 GAIN 0/100$ 2.39 RNM5 APRM 15 GAIN 0/100$ 2.20 RNM6 APRM 16 GAIN 0/100$ 2.18 RNM7 APRM 17 GAIN 0/100$ 2.16 RNMB APRM 18 GAIN 0/IOOX 2.17 NEUTRON MONITOR NONE A4-5

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 NM3 NEUTRON MONITOR NONE OD ON DEMAND NON-FUNCTIONAL OG OFF-GAS/RAD WASTE RDG1 DG 102 GOVERNOR SPEED DROOP SET RESET RDG2 OG 103 GOVERNOR SPEED DROOP SET RESET PC CONTAINMENT" , RPC1 NITROGEN FROM VAPORIZER YES NO RPC2 201.7-13 DW CAM ISOL VLV ll OPEN CLOSE RPC3 201.7-29 DW CAM ISOL VLV 12 OPEN CLOSE RPC4 201-40,41 DW, TORUS TO VENT SYS OPEN CLOSE RPC5 201-44,46 DW, TORUS TO ATMOS OPEN CLOSE RPC6 BV201.2-135,136 INTERLOCK DEFEAT YES NO RPC7 IV201-31,32 ISOLATION DEFEAT YES NO PP PROCESS COMPUTER RPP01 MEMORY PROTECT PLAN NORM REMOVD RDl CONTROL -RODS RRD1 301-2A CRO PUMP 11 DISCH VLV OPEN CLOSE RRD2 301-28 CRD PUMP 12 DISCH VLV OPEN CLOSE RRD3 301-BA CRD PUMP ll HEAD SPRAY ISOL OPEN CLOSE RRD4 301-88 CRD PUMP 12 HEAD SPRAY ISOL OPEN. CLOSE RRD5 301-88 CRO FLOW CONTROL VLV ISOL NC30A NC308 RO2 CONTROL RODS NONE R03 CONTROL RODS NONE RM1 RAD MONITOR NONE RP RPS RRP1 RX TRIP BUS 131 PWR SOURCE NORM EMER RRP2 RX TRIP BUS 141 PWR SOURCE NORM EMER RRP3 RPS BUS ll PWR SOURCE NORM EMER RR P4 BUS 12 PWR SOURCE NORM EMER A4-6

GENERAL PHYSICS CORPORATION GP-R-115007 Attachment 4 March 1, 1986 RRl REACTOR RECIRC RRR1 RECIRC MG-SETS ll LOCKOUT RELAY TRIP RESET RRR2 RECIRC MG-SETS 12 LOCKOUT RELAY TRIP RESET RRR3 RECIRC MG-SETS 13 LOCKOUT RELAY TRIP RESET RRR4 RECIRC MG-SETS 14 LOCKOUT RELAY TRIP RESET RRR5 RECIRC MG-SETS 15 LOCKOUT RELAY TRIP RESET RR2 REACTOR REC IRC NONE RR3 REACTOR RECIRC NONE RR4 REACTOR RECIRC NONE RW ROD WORTH MINIMIZER RRW1 CONTROL ROD SEQUENCE SELECT RX REACTOR CORE NONE SC SHUTDOWN COOLING NONE TC TURBINE CONTROL RTC1 REACTOR FLOW LIMIT 0-120% 120.00 RTC2 CONTROL VALVE LIMIT 0-120% 100.00 TU MAIN TURBINE A4-7

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 MONITORED 'ARAMETERS CORE REACTIVITY DK/K 20 CORE THERMAL POWER, 5

3. CORE FLOW, LBM/HR CORE PLATE DIFFERENTIAL PRESSURE, PSIG
5. CORE BORON CONCENTRATION, PPM
6. CORE AVERAGE VOID FRACTION, $

7~ CORE MINIMUM CRITICAL POWER RATIO

8. CORE MAXIMUM LINEAR HEAT GENERATION, KW/F 9~ CORE INLET SUB COOLING, BTU/LBM
10. CORE AVERAGE FUEL TEMPERATURE, DEG F CORE AVERAGE CLADDING TEMPERATURE, DEG F
12. CORE AVERAGE EXIT DUALITY, 5
13. (SPARE)
14. (SPARE)
15. REACTOR COOLANT ACTIVITY, UCI/ML
16. . REACTOR COOLANT CONDUCTIVITY, <UMHO/CM
17. REACTOR HEATUP/COOLDOWN RATE, DEG F/HR
18. REACTOR LEVEL-NARROW RANGE, INCHES
19. REACTOR LEVEL-WIDE RANGE, FEET
20. REACTOR PRESSURE, PSIG
21. RECIRCULATION LOOP ll FLOW, LBM/HR
22. RECIRCULATION LOOP 12 FLOW, LBM/HR
23. RECIRCULATION LOOP 13 FLOW, LBM/HR
24. RECIRCULATION LOOP 14 FLOW, LBM/HR
25. RECIRCULATION LOOP 15 FLOW, LBM/HR
26. RECIRCULATION LOOP ll SUCTION TEMPERATURE , DEGF
27. RECIRCULATION LOOP 12 SUCTION TEMPERATURE DEG F
28. RECIRCULATION LOOP 13 SUCTION TEMPERATURE , DEGF
29. CRD SYSTEM FLOW, LBM/HR
30. DRYWELL PRESSURE, PSIG
31. DRYWELL AVERAGE TEMPERATURE, DEG F
32. ORYWELL HYDROGEN CONCENTRATION, g
33. DRYWELL OXYGEN CONCENTRATION, 5 5A-1

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986

34. SUPPRESSION CHAMBER PRESSURE, PSIG
35. SUPPRESSION POOL WATER TEMPERATURE, DEG F
36. SUPPRESSION POOL WATER LEVEL, FEET
37. SRM COUNT RATE, CPS
38. SRM Pf R IOD, SEC
39. APRM POWER LEVEL, 5
40. CORE XENON CONCENTRATION, $ OF FULL POWER E}U
41. RWCU SYSTEM PRESSURE, PSIG
42. RWCU SYSTEM FLOW, LBM/HR
43. RWCU NON-REGEN HEAT EXCHAN OUTLET TEMPERATURE, DEG F 44, RWCU DUMP FLOW, LBM/HR
45. TOTAL MAIN STEAM LINE FLOW, MLBM/HR
46. MAIN STEAM TUNNEL TEMPfRATURE, DEG F
47. MAIN STEAM LINE RADIATION LEVEL, MR/HR 48 ~ TOTAL MAIN STEAM RELIEF VALVE FLOW, LBM/HR
49. TURBINE SPEED, RPM
50. TURBINE INLET PRESSURE, PSIG
51. TURBINE STEAM FLOW, LBM/HR
52. TURBINE BYPASS VALVE STEAM FLOW, LBM/HR
53. TURBINE FIRST STAGE PRESSURE, PSIG
54. TURBINE EXHAUST HOOD TEMPERATURE, DEG F
55. SECOND STAGE REHfATER OUTLfT PRESSURE, PSIG
56. SECOND STAGE REHEATER OUTLET TEMPERATURE, DEG F
57. CONDENSER VACUUM, IN HG V
58. CONDENSER HOTWELL LEVEL, INCHES
59. CONDFNSER HOTWELL CONDUCTIVITY, UMHO/CM
60. CONDENSER VACUUM MAKEUP FLOW, LBM/HR
61. CONDENSER HOTWELL REJECT FLOW, LBM/HR
62. CONDENSATE DEPRESSION, BTU/LBM
63. CIRCULATING WATER INLFT TEMPERATURE, DEG F
64. CIRCULATING WATER OUTLET TEMPfRATURE, DEG F
65. TOTAL CIRCULATING WATER FLOW, GPM
66. CONDENSATE DEMINERA OUTLET CONDUC, UMHO/CM
67. TOTAL FEEDWATER SYSTfM FLOW, LBM/HR
68. FEEDWATER TEMPERATURE TO REACTOR, OEG F 5A-2

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986

69. GENERATOR LOAD, MWE
70. GENERATOR REACTIVE LOAD, MVAR
71. GENERATOR 'STATOR AMPS, AMP
72. GENERATOR TERMINAL VOLTS, VOLT
73. GENERATOR HYDROGEN PRESSURE, PSIG
74. DIESEL GENERATOR 102 LOAD, KWE
75. DIESEL GENERATOR 103 LOAD, KWE
76. OFF-GAS SYSTEM INLET FLOW, CFW
77. OFF-GAS SYSTEM OUTLET FLOW, CFW
78. OFF-GAS RECOMBINER INLfT HYDROGEN CONCENTRATION,
79. RECOMBINER OUTLET HYDROGEN CONCENTRATION, g 5'FF-GAS
80. OFF-GAS SYSTEM RADIATION LfVEL, MR/HR
81. CORE SPRAY LOOP 11 PRESSURf, PSIG
82. CORE SPRAY LOOP 12 PRfSSURE, PSIG
83. CORE SPRAY LOOP 11 FLOW, LBM/HR
84. CORE SPRAY LOOP 12 FLOW, LBM/HR
85. EMERGENCY CONDENSER LOOP 11 FLOW, LBM/HR
86. EMERGENCY CONDENSER LOOP 12 FLOW, LBM/HR

'7.

EMERGENCY CONDENSER LOOP 11 RETURN TEMPERATURE, DEG F

88. EMERGENCY CONDENSER "LOOP 12 RfTURN TEMPERATURE, DEG F
89. EMERGENCY CONDENSER LOOP ll VfNT RAO LEVEL, MR/HR
90. EMERGENCY CONDENSER LOOP 12 VENT RAD LEVEL, MR/HR 5A-3

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 NINE MILE POINT DATABASE CHANGES AS GENERATED BY PLANT MODIFICATIONS Modi ficat i on N18040 caused the fol l owing prints/drawings to move up in revision level:

PRINT NUMBER SHEET NUMBER S 18009 1,2 18041 7 19437 19438 6,8,9 19440 ll 19845 19854 19859 8,8A,9, 10, 11A, 13, 14, 17, 18A 19951 8,9 19954 2,3,5,8 22020 4,5 22373 2,4,5 22381 1,2,4,11 22382 2,4,5 22383 1,4,8 22387 2 23119 34841 1,2,3,4 34842 1 34845 1,2,3 Modification N18042 caused the following prints/drawings to move up in revi si on evel:

1 PRINT NUMBER SHEET NUMBER S 19409 6,9,10 19438 1,7,8 19440 2,8,9, 10 19854 1.3 22374 1,3,4 A6-1

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 Modi ficati on N18090 caused the following prints/drawings to move up in revision level:

PRINT NUMBER SHEET NUMBER S 19412 1,1A 19413 1,2 19416 2A 19417 19418 19847 22238 2 22239 7 22242 1 23126 1 23127 1 Modi fi cation N18266 caused the fol 1 owing prints/drawings to 'ove up in revision level:

PRINT NUMBER SHEET NUMBER S 18013 18014 19859 22 26726 Nl-OP-9 A6-2

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 Modification N182801 caused the following prints/dr awings to move up in revision level:

PRINT NUMBER SHEET NUMBER S 19437 2,6,10 19438 '3 19440 2,6,10,11 19842 2 19845 19859 2,3,5,6,8,8A,10,10A, 18,23 19951 10 22005 5,6,8,9,11,12,14,15 22302 1,2,6,7 22373 11,'12,13 22374 2,3,4,5 22379 3 22381 5,6,7,8,9,10 22382 1,3 22383 3,5,6 22386 1,2 Mal function N183582 caused the following prints/drawings to move up in revision level:

PRINT NUMBER SHEET NUMBER S 26726 1,2,3,4 Modi fi cation N183583 caused the fol lowing speci ficati ons and summary to be added to the database:

SPDS Functional Specification SPOS Software Oesign Specification Process Computer Point Summary dated 2/15/85 A6-3

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 Modi ficat i on N183586 caused the fol owing prints/drawings 1 to move up in revi si on 1 evel:

PRINT NUMBER SHEET NUMBER S 19859 18,18A 22373 3 22385 11,12,13,14,16,16A,17,17A, 18, 18A, 19,19A 23032 4 A6-4

GENERAL PHYSICS CORPORATION GP-R-115007 March 1, 1986 MALFUNCTIONS TESTED FOR ANSI/ANS 3.5 REPORT, 1986 AD01 FW17 RR09 AD05 FW21 RR13 AN01 FW25 RR17 CS02 FW29 RR21 CT03 IA01 RR25 CU04 LP01 RR28 CU08 MCOl RR29 CW01 MC04 RR32 CW02 MSOl RR36 CW05 MS02 RR40 CW09 MS04 RR44 DG02 MSO7 RR48 e- rn5ip EC01 NM03 RR52 EC02 NMll RR56 EC05 NM19 RR60 ED01 NM29 RR64 ED02 NM37 RR68 ED06 OG01 RR72 ED10 PC01 RX01 EA14 RD01 SC01 ED18 RD04 TC01 ED22 RDOS TC03 EGOl RD33 TC07 EG05 RD36 TC11 EG09 RD37 TU02 EG13 RD41 TU06 FP03 RM05 F P07 RP01 FW01 RP03 FW05 RP05 FW09 RR01 FW13 RR05 A7-1

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<:S nol) t book S4lvc S Figure 1 - Nine Mile Point Unit One Control Room

~Le end 2 Hain Fire Panel 82 A Panel A Electric/Turbine Controls

/ 8 Panel 8 Relays/Turbine Controls

/ F Plant Process Computer

// Equipment D D Chief Shift Operator Desk I

I E E Panel Hain Control Console I F F Panel NSSS I

I G G Panel Nuclear C

Instrumentation H H Panel Balance of Plant H I Instructors Console I

J J Panel Radiation Honitoring K K Panel ECCS I L L Panel Primary Containment 1

T T t1 H Panel RPS N

I N N Panel Turbine I

T T P Print Rack S Storage Cabinets T Tables Figure 2 - Nine Hile Point Unit One Simulator Control Room

C -M NINE MILE POINT NUCLEAR STATION UNIT 1 PLANT REFERENCED SIMULATOR ANNUAL REPORT: ANSI 3.5 1985 FOR THE YEAR 1987 Testin Conducted A ri1 Ma 1987 Re ort Prepared June 1987 Prepared Sy: George Roarick Reviewed 8

~ iz Zg is r, Unit ¹1 Operat ons ~ining Date Asst. Superintende t of Training Superintendent of Training o

0

/te j'7

SIMULATOR INFORMATION The purpose of this section is to provide familiarization with the Nine Mile Point Unit Plant Referenced Simulator and its applicability as an 1

operator training device.

A. General The simulator is owned by the General Physics Niagara Corporation which is a wholly-owned subsidiary of General Physics Corporation. It is used jointly by General Physics Corporation and Niagara Mohawk Power Corporation instructors.

It is maintained and modified by the General Physics Corporation 'under the 'direction of Niagara Mohawk Power Corporation. The simulator was built by Singer/Link.

2. The simulator 'is a full scope control room simulator that simulates the Nine Mile Unit ¹1 plant. The plant is an 1850 Megawatt Thermal, BNR-2 plant with an electrical output of 620 Megawatts.
3. The simulator was declare'd ready for training on September 1, 1984.
4. The initial report on the simulator was prepared in March 1986. This report was prepared to document modification since March 1984 and to document simulator performance subsequent to the modifications.
5. In addition to training, the simulator has been used by the Operations Department to validate procedures prior to their implementation at the plant.

-1 June 1987

B. Control Room (Phys i cal Fi de i ty) 1

1. The physical layout of the Nine Mile Point Unit ¹1 Control Room is shown in Attachment "A". The physical layout of the simulator is shown in Attachment "8". A comparison of these two drawings shows a high degree of similarity between the two rooms. The following differences exist:

An additional office was added next to the "SSS" office at the plant. This office does not exist in the simulator (Note 1).

The instructor's station occupies the area between Hain Fire Panel 2 and the "SSS" office. A desk occupies this area at the plant (Note 1).

C. A manual dose assessment calculator is mounted on the wall next to the stairwell in the plant's control room.

A meteorological computer is located next to i'his equipment does not exist in the simulator.

d. There are some minor differences in the amount and type of furniture in both rooms (Note 2).
e. A TV camera is mounted on the wall above the NSSS typer at the plant. An emergency lighting system occupies this location in the simulator.

NOTES:

Plans have been approved to move the instructor's station to an 18" high platform that is located in the same area that is occupied by the additional office Items to correct this discrepancy are on order.

I

-2 June 1987

2. Panels and Equipment The simulator contains all of the panels that are in the control room at the plant. All front panels and the "E" console are fully simulated. The back panels are fully simulated with the exception of the following:

RPS relays are installed, but are not functional.

b. Electrical protective relays are cosmetically simulated by photos mounted in the relay enclosures.

c ~ Only one of the four Transversing In-Core Probe control panels is functional.

d. Seismic Monitors are not functional.
e. Radwaste Solidification/Storage Building Area Radiations Monitors are not installed (Note 1).

The new Service Water, Radwaste Effluent and Control Room Vent Radiation Monitors are not installed (Note 2).

NOTES:

This modification is in the fabrication stage.

2. This modification is in the evaluation/approval stage but the older Service Water and Radwaste Effluent Monitors are still installed.
3. Systems All systems that are operable from the Nine Mi le Point Unit ¹1 Control Room are simulated. See Attachment "C" for a list of these systems.
4. Simulator Control Room Environment The Simulator Control Room was specifically designed to duplicate as nearly as possible the reference plant control room environment. Other than the discrepancies noted in Section 1 of this part, the following differences exist:

-3 June 1987

fven though the lighting is identical to that in the plant, the lighting system is not functionally interfaced with t he simulator. Plans to interface the lights and the simulator are being evaluated for training value.

b. The ambient noise level that exists in the plant's control room is not simulated. The hardware to do this exists, but is not functional. Neighing the training impact against the financial considerations have warranted not pursuing this "luxury".

Phase I of the "Detailed Control Room Design Review" was completed and implemented in both the simulator and Plant Control Room in 1986. The initial study utilized the simulator to identify potential problems and try new concepts'hase II is complete and scheduled for implementation in both the simulator and plant control room in early 1988.

C. Instructor Interface (Control Capabilities)

Initial Conditions The instructor has the capability to initialize to any one of fifty (50) sets of initial conditions . The first twenty of these sets of conditions are guarded and can only be changed by the proper code. These initial conditions are the foundation to approximately 951. of, the training. These twenty sets of initial conditions are -'listed in Attachment "D". The remaining thirty sets of i.ni tial conditions are set by the individual instructor and can be changed at any time.

2. Malfunctions Malfunctions vary from a discrete nature (i.e. pump trip) to ones of varying degrees of severity (i.e. leaks). They are listed in Attachment "E".

-4 June 1987

3. Remote Functions The instructor has the capability to simulate most in-plant operation required to back-up control room operation. The only remote function, wi th major training impact, that is not simulated is the ability to jumper out the Low-Low Reactor Hater Level MSIV isolation signal identified in Emergency Operating Procedures. Addition of this function is currently scheduled for late 1987 early 1988 time frame. The list of remote functions can be found in Attachment "F".
4. Instructor Overrides The instructor has the capability to override most functions that are simulated. This includes meter and chart recorder indications, indicating lights, annunci ators and switch functions. A few discrepancies still exist in the I/O program, but are being documented for evaluation and repair as they are identified.
5. Monitoring The instructor can monitor from the instructor station, up to sixteen (16) of any of ninty (90) parameters (See Attachment G). In addi tion, the instructor can select up to twelve (12) of the parameters to be plotted on the line printer in the control room.
6. Instructor Station Controls Consists of three (3) keyboard/CRT's,. two (2) for performing simulator functions, and one '1) .

which is used for minor troubleshooting. There are also various buttons which perform the following:

-6 June 1987

a. Freeze s tops (freezes) s imul at i on at any poi nt or restarts it.
b. Reset/Ready initializes the simulator to a selected set of initial conditions. This button wi 11 turn green when all controls are properly positioned.

C. Snapshot records and stores a set of conditions into any one of the IC's.

d. Malfunction Clear clears the entire malfunction tableau.
e. Backtrack sets the simulator to step up or back to any point in time, within 60 minutes. This includes buttons to step up or back one minute at a time and to step forward or reverse.
f. Manual Malfunction Control - Used to increase or decrease the severity level of up to three (3) of the variable malfunctions.

Annunciator Silence silences all annunciators.

h. Recorder Off shuts off power to all chart recorders.

Test and Lamp Test Testing and Troubleshooting.

Emergency Stop Kills all power to computer and simulator.

Computer Alarm/Acknowledge Hams of computer malfunctions.

Record and Replay Controls the tape recorder in the Computer Room.

m. Fast Time/Slow Time changes speed of simulator response.
7. Record/Replay

'f so desired, a scenario can be recorded on magnetic tape for replay at a later date. This function is controlled from the instructor station.

-6 June 1987

Reference Plant Operating Procedures The simulator is operated using the same procedures used to operate the plant. The operations department makes use of the simulator to help validate certain new procedures or procedural revisions.

Procedure steps that cannot be performed on the simulator are conspicuously identified in the simulator copy in accordance with established Nuclear Training Department Instructions.

Changes Since Last Report The following plant modifications have been incorporated into the simulator since the last report (March 1986):

78-04 CAD Valve N2 Supply 78-24 Remove FW Level Programmer 78-27 MSIV Monitoring (add Process Computer points) 78-32 CAD Alarm State Changes 79-06 Change TIP Controls 79-24 APRM Rod 8lock Scope 80-38 Fuel Zone Level Indicator Changes 80-41 Add Emergency Cooling Vent to Torus 80-74 Add Torus Temperature Indicators 80-84 Diesel Generator Annunciator Addition 81-14 "E" Gate Digital 81-29 Emergency Vent Valve Change 82-30 Reactor Recirc. Pump Monitoring 82-69 Feedwater Pump Low Flow Control .Valves and Recirc. Valves 82-71 Powerboards 102/103 Undervoltage Relays 82-80 Core Spray Valve Logic Changes/Emergency Cooling High Rad 82-93 TIP Changeout to Gamma Detectors 83-29 Control Room LED Displays 83-53 Scriba Sub-Station Retirement 83-58 Revisions to Integrated Cosmetic Package and SPD's 83-61 Control Room Ventilation 83-89 ¹13 IAC Cooling Hater Changeover 84-58 Feedwater Pump Control 85-26 ADS Inhibit

-7 June 1987

C II. SIMULATOR DESIGN DATA The initial design data for the simulator is listed in the original database document on file in the simulator library and Simulator Configuration Management System. Plant modification design data for all modifications incorporated in the simulator prior to March 1986 are contained in the last annual report (March 1986). Documentation, data changes and test results for modification incorporated since March 1986 (see Part I, Section E) are contained in Attachment "H".

III. SIMULATOR TESTS (PERFORMANCE TESTING)

During the April - May 1987 time frame, testing was done to verify real time operation, steady state and normal operation, transient performance and malfunction response. Documentation of these tests is avai 1 able in the simulator database and records under the title "Simulator Performance Test Data May 1987".

A. Computer Real Time Tests Simulator real time testing was performed by measuring individual model times during steady state and transient conditions. During this testing, no frame slippage or program overtimes occurred. The simulator contains safeguards which preclude operation outside of real time with the exception of two (2) instructor controlled functions. Slow time will slow all simulator responses to one-half real time. Fast time will step up response time to only three (3) functions, reactor xenon, condenser evacuation and turbine warming.

-8 June 1987

B. Steady State and Normal Operation

1. Simulator stability was verified by comparing heat balances (P-1') and cri tical parameter printout from the start and finish of a s i x ty (60) mi nute steady-state run. The only devi ation outs i de accceptab1 e 1 imi ts (+2/.) exi sted in the calculated Core therma,1 Power printout on the heat balance obta.ined at the start of the run. This was attributed to an slight feedwater transient and the nature of the simulators PPC programs, This problem has been documented and corrected. The Core Thermal Power printout on the final P-1 and on both the initial and final Computer Point Printouts were within one (1) Megawatt thermal.
2. Fidelity in performance was verified by comparing heat balances (P-1) and Balance of Plant parameters from the plant and simulator at 25/., 55'/. 76% and 100'/ of rated power.

25'/ Power one of the most significant differences was plant efficiency. Plant data shows 125 MWE at 467 MWT while the simulator shows 164 MWE at 484 MWT. This in 26.7/ eff. vs 33.8/. eff. Plant data shows mismatching flows between the five recirc. pumps which, on the simulator is exactly equal. Plant data also shows a mismatch between level columns ll and 12, which are also exactly equal on the simulator. Plant data for total steam flow is significantly different from simulator data, but the plant data appears incorrect.

b. 55/. Power The same problems with water level and plant efficiency were also evident at this. power level. The plant was able to achieve the same power level with significantly less recirc. flow. The problem needs further investigation. The simulator is modeled for middle of cycle and the plant data is beginning of cycle. The difference in total steam flow also fall outside of tolerances and will need further investigation.

-9 June 1987

c ~ 76% Power The same problems with plant efficiency and water level that were previously noted were also present at this power level. Again the plant achieved this power level with less recirc flow. As stated before, this will be further investigated. The APRM readings were also outside of the acceptable 'tolerances, but this is attributed to the differences in the gain adjustment factors.

d. 1001. Power - The same problems with plant efficiency and water levels still exist. Recirc Pumps 12 and 14 flows were outside the tolerances. These flows differed +2/

from the other three loops. The simulator shows all recirc. pump flows as being equal.

Except as noted above, all other parameters compared were within the acceptable tolerances. The problems with mismatching recirc. pump flows and water levels have no significnt training impact. The problems with plant efficiency and power to flow relationship are documented and wi 11 be corrected.

C. Trans i ent Tests

1. The following transients are FSAR analyzed and were run on the simulator in real time.

Simultaneous trip of all feedwater pumps (compared with FSAR transient "Feedwater Malfunction, Zero Flow). The FSAR shows an increase in recirc. flow due to changes in two-phase flow. The simulator doesn'0 model two-phase flow closely enough to produce this; therefore, it doesn't change. This discrepancy has no training impact and will not be discussed further.

Simultaneous Closure of MSIV's Consistent with FSAR.

-10 June 1987

c. Simultaneous Trip of all Recirc Pumps Consistant with FSAR.
d. Single Recirc. Pump Trip The simulator shows a much more pronounced decrease and recovery of recirc. flow than the FSAR. The other parameters responded as predicted with regards to recirc. flow. Reactor Pressure matches the FSAR analysis.
e. Design Bases Loss of Coolant Simulator response was consistent with the FSAR with the exception of one parameter. Drywell pressure spiked at approximately 22 psig in 31 seconds on the simulator. The FSAR predicts a pressure spike of 33 psig in two seconds. A discrepancy report has been wri tten and this problem wi 1 1 be addressed by the simulator support group.

Design Bases Main Steam Line Break The simulator response was consistent with FSAR results.

I

2. The following transients are not analyzed in the FSAR but are required by the standard <ANSI 3.5-1985).

'a ~ Manual Reactor Scram The simulator was consistent with predicted response.

b. Turbine Trip at 40/. Power <no scram) The simulator was consistent with predicted response C. Maximum Ramp of Power; 1001. to 75'/; 75/ to 100% The simulator was consistent to predicted response.
d. MSIV closure with a stuck open ERC '<No high pressure,ECCS systems available) Simulator was consistent with predicted response.

-ll June 1987

3. The following transients are not required by the standard (ANSI 3.5 1985) but were run because data exists in the FSAR.

a- Turbine Trip without Bypass

b. Recirc Pump Stall C. Inadvertant actuation of an ERV
d. Safety Valve Actuation
e. EPR/MPR Failure In all cases, the simulator response was consistent with FSAR predictions.

D. Malfunction Test During the course of testing to prepare this report, 25'/ of the existing malfunctions listed in Attachment "E" were tested in addition to those needed to obtai n the data in Section C of the report. The malfunctions tested are listed in Attachment "I".

During the course of performance testing, the malfunction and evolutions required for operator training by 10CFR55, "Operator Licenses", were conducted at least once. Minor discrepancies were identified while performance testing malfunctions. Hone of these problems have any significant training impact and will be addressed in accordance with NTI 4.5.3, "Simulator Configuration Management".

-12 June 1987

IV. SIMULATOR CONFIGURATION MANAGEMENT (DISCREPANCY RESOLUTION AND UPGRADING)

Discrepancy resolution and modifications are handled in accordance with NTI 4.5.3 "Simulator Configuration Management". A personal computer has been set up as a terminal for Niagara Mohawk Power Corporation's Nuclear Divisions'onfiguration Management System. This system tracks plant modifications and all plant documents. A special program has been set up in this system to document plant modifications that impact the simulator and track associated data base changes. In addition to the Configuration Management System, other computer programs exist for tracking discrepancy reports and simulator database changes (modifications). All discrepancies and modifications are evaluated for training impact.

NOTE: All data collected and analyzed to formulate this report are on file at the Nine Mile Point Nuclear Training Center and are available for review.

-13 June 1987

0

~Le end 2 Hain Fire Panel 82 A Panel A Electric/Turbine Controls B Panel B Relays/Turbine Controls C Plant Process Computer Equipment D Chief Shif t Operator Desk E E Panel Hain Control Console F F Panel NSSS G G Panel Nuclear c Instrumentation 11 H Panel Balance of Plant I Instructors Console J J Panel Radiation Honitoring K K Panel ECCS L L Panel Primary Containment c 0 H H Panel RPS g5 N N Panel Turbine P Print Rack S Storage Cabinets T Tables H/R Heteorological Computer R Hanual Dose Accessment Calculator R/T Radio/Telephone Equipment Procedures B/S Bookshelf D/K Desk Figure 1 - Nine Mile Poin One Control Room

~ I I ~ ~

~ ~

~ ~ ~

I~ ~

~ ~ ~ I ~ ~ ~

I ~ I ~ ~

Attachment C SYSTEMS fULLY SIMULATED

l. Nuclear Boiler and Instrumentation
2. Reactor Recirculation System
a. Reactor Recirculation Loops
b. Boiler Process Instrumentation
c. Recirculation flow Control
3. Control Rod Drive and Hydraulics System (CRDHS) 4, Reactor Manual Control System (RMCS)
5. Reactor Core (Physics and Thermodynamics)
a. Reactor Core Neutron Kinetics
b. Reactor Core Thermodynamics
6. Rod North Minimizer (RNM)
7. Main Steam Systems
a. Main Steam and Main Steam Bypass Systems
b. Moisture Separators Reheaters
c. Extraction Steam System
d. Auxiliary Steam System
8. Reactor Hater Cleanup System
9. Nuclear Instrumentation System
a. Source Range Monitor (SRM) System
b. Intermediate Range Monitor (IRM) System
c. Local Power Range Monitoring (LPRM) System
d. Average Power Range Monitoring (APRM) System
e. Rod Block Monitor (RBM) System
f. Traversing In-Core Probe (TIP) System
10. Reactor Protection System
11. Simulation oi'he Primary Containment and Isolation System
a. Primary Containment
b. Primary Containment Isolation System
12. Secondary Containment
13. Emergency Ventilation
a. Reactor Building Ventilation
b. Turbine Building Ventilation
c. Building Ventilation
14. Primary Containment Atmosphere Control and Sampling System ANSI 3.5 C-1 June 1987
15. Emergency Core Cooling Systems
a. Automatic Depressurization and pressure Relief System b., Core Spray
c. High Pressure Coolant Injection <HPCI) System
d. Containment Spray
e. Emergency Cooling System
16. Shutdown Cooling
17. Standby Liquid Control (SLC) System
18. Condensate and Feedwater System
a. Condensate System
b. Condensate Demineralizer System
c. Feedwater System
d. Condensate Storage and Transfer System
e. Reactor Vessel Level Control System
i. Feedwater Heaters, Vents and Drains
19. Off-Gas Recombiner and Condenser Air Removal
20. Main Condenser
21. Circulating Hater System
22. Reactor Building Closed Loop Cooling
23. Turbine Building Closed Loop Cooling
24. Service Hater System
25. Instrument, Service and Breathing Air
26. Area Radiation Monitoring System
27. Process Radiation Monitoring System
28. Ventilation Radiation Monitoring System
29. Main Turbine and Turbine Control
a. Turbine Oil System
b. Turbine Kinematics
c. Turbine Mechanics
d. Turbine Supervisory and Safety System
e. Gland Seal System
f. Low Pressure Hood Spray System
g. Moisture Separator and Reheat System
h. Main Turbine Electro-Hydraulic Control System ANSI 3.5 C-2 June 1987
30. Plant Electrical System
a. Main Generator and Auxiliary Systems
1. Main Generator Synchronous Machine
2. Excitation and Voltage Regulator System
3. Synchroscope
4. Hydrogen Cooling System
5. Stator and Iso-Phase Duct Cooling System
6. Hydrogen Seal Oil System
b. Electrical Distribution System
1. Buses and Transformers
2. Breakers
3. Currents, Voltages and Frequencies
4. DC Electrical Distribution and Control
5. Power System Electrical Grid
c. Diesel Generators
31. Containment Atmosphere Dilution, Vent and Purge System
32. Radiation Waste Disposal System Containment Equipment a nd Floor Drain Sump
33. Plant Carbon Dioxide System
34. Diesel Fire Pump and Pressurized Water Fire System
35. Fire Control Ventilation Systems
36. Control Room Heating, Ventilation and Air Conditioning
37. Communication System
38. Plant Process Computer System
a. Applicable Experience
39. Meteorological Experience
40. Plant Annunciators and Fire System Alarm ANSI 3.5 C-3 June 1987

ATTACHMENT D GUARDED INITIAL CONDITIONS IC ¹ DESCRIPTION Cold Iron Rx is S/D and C/D, all systems off-line but electrical di stribution Cold Startup All support systems on-line, ready to commence rod pull Cold Startup 5 rods subcritical Heatup 140 F Rx in heating range

5. Heatup 280 F Rx in heating range
6. Heatup 900 psig Rx in heating range
7. NOP, NOT, 1% Power
8. Shutdown inserting Rod Group 76
9. Shutdown Mode Switch in S/U
10. Shutdown All Rods In
11. Turbine Startup Turbine Harm
12. Feedwater Pump ¹13 Startup
13. 50% Power Precondi tioning
14. 100% Power - Middle of Cycle
15. Full Power End of Cycle
16. -Reserved for Future Use
17. -Reserved for Future Use
18. -Reserved for Future Use
19. Cool down 200 ps i g
20. Cooldown SDC in Service ANSI 3.5 D-1 June 1987

~

ATTACHMENT E NINE MILE POINT UNIT ONE MALFUNCTIONS A001 ADS FAILURE TO INITIATE PRIMARY VALVES AD02 ADS FAILURE TO INITIATE - COMPLETE AD03 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹111) FAILURE - SOLENOID AD04 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹111) FAILURE - VALVE LEAKS AD05 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹1 11) FAILURE OPENS INADVERTENTLY AD06 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹lll) FAILURE STUCK OPEN AN01 CONTROL ROOM ANNUNCIATOR SYSTEM FAILURE CS01 CORE SPRAY PUMP TRIP (111, 112, 121, 122 OR ANY)

CS02 CORE SPRAY TOPPING PUMP TRIP (111, 112, 121, 122 OR ANY)

CS03 CORE SPRAY INBOARD INJECTION VALVE FAILURE TO OPEN (IV40-01, IV40-09, IV40-11, IV40-10 OR ANY)

CT01 CONTAINMENT SPRAY PUMP TRIP ( 1 1 1, 112, 121, 122 OR SPRAY RAN WATER PUMP TRIP (111, 112, 121, 122 OR ANY)

ANY)'ONTAINMENT CT02 CT03 CONTAINMENT SPRAY HEAT EXCHANGER (111, 112, OR BOTH) TUBE LEAK CU01 COOLANT LEAKAGE INSIDE PRIMARY CONTAINMENT CU02 REACTOR WATER CLEANUP PUMP TRIP (11, 12 OR BOTH)

CU03 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-ND22) FAILS OPEN CU04 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-ND22) FAILS CLOSED CU05 REACTOR WATER CLEANUP HIGH PRESSURE CONTROL VALVE (PCV 33-39) FAILS OPEN CU06 REACTOR WATER CLEANUP HIGH PRESSURE CONTROL VALVE (PCV 33-39) FAILS CLOSED CU07 REACTOR WATER CLEANUP LOW PRESSURE CONTROL VALVE (PCV-ND37) FAILS OPEN CU08 REACTOR WATER CLEANUP LOW PRESSURE CONTROL VALVE (PCV-ND37) FAILS CLOSED CU09 REACTOR WATER CLEANUP NON-REGENERATIVE HEAT EXCHANGER TUBE. LEAK CU10 REACTOR WATER CLEANUP DEMINERALIZER RESIN DEPLETION (11, 12 OR BOTH)

CU11 COOLANT LEAKAGE OUTSIDE PRIMARY CONTAINMENT CW01 HIGH RADIATION IN SERVICE WATER CW02 SERVICE WATER PUMP TRIP (11, 12 QR BOTH)

ANSI 3.5 E-1 June 1987

I CN03 EMERGENCY SERVICE WATER PUMP TRIP (11, 12 OR BOTH)

CN04 REACTOR BUILDING CLOSED LOOP COOLING (11, 12, 13 OR ANY) PUMP TRIP CN05 TURBINE BUILDING CLOSED LOOP COOLING PUMP TRIP (11, 12 OR BOTH)

CN06 CIRCULATING HATER PUMP TRIP (11, 12 OR:BOTH)

CN07 CIRCULATING WATER EXPANSION JOINT LEAKAGE CW08 CIRCULATING WATER INTAKE STRUCTURE ICING CW09 LOSS OF DRYNELL COOLING CN10 MAIN CONDENSER TUBE LEAK DG01 DIESEL GENERATOR FAILURE TO START (102, 103 OR BOTH)

OG02 DIESEL GENERATOR TRIP (102, 103 OR BOTH)

EC01 STEAM LEAKAGE INSIDE PRIMARY CONTAINMENT EC02 STEAM LEAKAGE OUTSIDE PRIMARY CONTAINMENT EC03 EMERGENCY COOLING SYSTEM RETURN VALVE FAILS OPEN (IV39-05, IV39-06 OR BOTH)

EC04 EMERGENCY COOLING SYSTEM RETURN VALVE FAILS TO OPEN (IV39-05, IV39-06 OR BOTH)

EC05 EMERGENCY COOLING SYSTEM EMERGENCY CONDENSER MAKEUP CONTROL VALVE FAILS CLOSED (LCV60-17, LCV60-18 OR BOTH)

EC06 EMERGENCY CONDENSER TUBE LEAK (111, 121 QR BOTH)

ED01 LOSS OF OFF-SITE 115 KV POWER SOURCES (LIGHTHOUSE HILL-JAF, OSWEGO STEAM, OR BOTH)

ED02 BATTERY CHARGER AND EMERGENCY LIGHTING SUPPLY MOTOR GENERATOR TRIPS (161, 171 OR BOTH)

E003 COMPUTER POWER SUPPLY MOTOR GENERATOR TRIPS (167)

EO04 AC PONERBOARD ELECTRICAL FAULT <PB11)

EO05 AC POWERBOARO ELECTRICAL FAULT (PB12)

ED06 AC POWERBOARD ELECTRICAL FAULT <PB101)

E007 AC POWERBOARD ELECTRICAL FAULT (PB102)

ED08 AC POWERBOARD ELECTRICAL FAULT (PB103)

ED09 AC POWERBOARD ELECTRICAL FAULT (PB13 SECTION A)

ED10 AC PQWERSOARO ELECTRICAL FAULT (PS13 SECTION B)

EOll AC PONERSOARD ELECTRICAL FAULT (PB13 SECTION C)

ED12 AC PONERBOARD ELECTRICAL FAULT (PB14 SECTION A)

ED13 AC PONERSOARD ELECTRICAL FAULT (PB14 SECTION 8)

ED14 AC PONERBOARO ELECTRICAL FAULT (PB14 SECTION C)

ED15 AC POWERBOARO ELECTRICAL FAULT (PB15 SECTION A)

ED16 AC POWERSOARD ELECTRICAL FAULT (PB15 SECTION 8)

ED17 AC PONERBOARD ELECTRICAL FAULT (PB15 SECTION C)

ANSI 3.5 E-2 June 1987

ED18 AC POWERBOARD ELECTRICAL FAULT (PB16 SECTION A)

ED19 AC POWERBOARD ELECTRICAL FAULT (PB16 SECTION 8)

ED20 AC POWERBOARD ELECTRICAL FAULT (PB17 SECTION A)

ED21 AC POWERBOARD ELECTRICAL FAULT (PB18 SECTION B)

ED22 DC POWERBOARD ELECTRICAL FAULT ( 11, 12 OR BOTH)

ED23 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - NORMAL fD24 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 ALTERNATE ED25 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 NORMAL AND ALTERNATE EG01 MAIN GENERATOR TRIP ELECTRICAL FAULT EG02 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS - INCREASE EG03 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS - DECREASE EG04 MAIN GENERATOR CORE INTERNAL HEATING EG05 MAIN TRANSFORMER LOSS OF COOLING EG06 GENERATOR HYDROGEN COOLING SYSTEM LEAKAGE EG07 GENERATOR HYDROGEN MAIN SEAL OIL PUMP FAILURE EG08 GENERATOR HYDROGEN EMERGENCY SEAL OIL PUMP FAILURE EG09 STATOR COOLING PUMP TRIP <11, 12 OR BOTH)

EG10 LOSS OF CONTROL AIR TO 345 KV BREAKER (R-915, R-925 OR BOTH)

EG11 POWER GRID NETWORK LOAD TRANSIENT INCREASE EG12 POWER GRID NfTWORK LOAD TRANSIENT DECREASE EG13 STATOR WATER COOLING DEMINERALIZER RESIN DEPLETION FP01 DIESEL FIRE PUMP FAILURE FP02 ELECTRIC FIRE PUMP FAILURE FP03 AC FOAM PUMP FAILURE FP04 DC FOAM PUMP FAILURE FP05 TURBINE ISLAND FIRE DETFCTION (D-1195, D-1155, D-1165,'-1175, D-1061, D-1114, D-1131 OR ANY)

FP06 CONTROL ROOM FIR f DETECTION ( F IRf .PANEL 2, CONTROL, CONSOLE, "L" PANEL, "K" PANEL, "H" PANEL, "F" PANEL, "A" PANEL OR ANY)

FP07 TURBINE BUILDING FIRE DETECTION. (DA-22092MG, DA-2083M, DA-2081S, DA1092E, D-2102 OR ANY)

FP08 DIESEL ROOM FIRE DETECTION (DX-2113A, DX-21138, DX-02141A, DA-2141, DX-2151B, DA-2151, D-2151 OR ANY)

FP09 AUXILIARY CONTROL ROOM/cable spreading room fire detection (d-3031PL, DX-3031A, DX-3111B, WD-8131, WD-8082 OR ANY)

FP10 REACTOR BUILDING FIRE DETECTION <DX-4217A, DA-4116W, DA-4076E, D-4207, D-4156, SP-4126, D-4086 OR ANY)

ANSI 3.5 E-3 June 1987

FW01 CONDENSATE PUMP TRIP (11, 12, 13 OR ANY)

FW02 FEEDWATER BOOSTER PUMP TRIP <ll, 12, 13 OR ANY)

FW03 FEEDWATER PUMP TRIP <11, 12 OR BOTH)

FW04 SHAFT DRIVEN FEEDWATER PUMP 13 FAILURE FW05 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAI'LURE TO ENGAGE FW06 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAILURE TO DISENGAGE FW07 FEEDWATER CONTROL VALVE ll CONTROLLER FAILS HIGH FW08 FEEDWATER CONTROL VALVE 11 CONTROLLER FAILS LOW FW09 FEEDWATER CONTROL VALVE 12 CONTROLLER FAILS HIGH FW10 FEEDWATER CONTROL VALVE 12 CONTROLLER FAILS LOW FWll FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS HIGH FW12 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS LOW FW13 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS AS IS FW14 FEEDWATER MASTER CONTROLLER FAILS HIGH FW15 FEEDWATER MASTER CONTROLLER FAILS LOW FW16 FEEDWATER MASTER CONTROLLER FAILS AS IS FW17 CONDENSATE DEMINERALIZER DEPLETION FW18 FEEDWATER CONDUCTIVITY INCREASE FW19 CONDENSATE RECIRCULATION VALVE <FCV 50-24) FAILS OPEN FW20 CONDENSATE RECIRCULATION VALVE (FCV 50-24) FAILS CLOSED FW21 FEEDWATER BOOSTER PUMP RECIRCULATION VALVE FAILS OPEN (FCV 51-58, FCV 51-59, FCV 51-60 OR ANY)

FW22 FEEDWATER HEATER TUBE LEAK FW23 FEEDWATER PUMP RECIRCULATION VALVES FAIL OPEN (ll, 12, 13 OR ANY)

FW24 FEEDWATER CONTROL VALVE FAILS CLOSED (13A, 138 OR BOTH)

FW25 THREE MILE ISLAND ACCIDENT (BWR EQUIVALENT)

FW26 CONDENSATE BYPASS SPRAY TO MAIN CONDENSER FLOW CONTROL VALVE (FCV 50-22) FAILS CLOSED FW27 LOSS OF COMPENSATION TO FEEDWATER FLOW TRANSMITTER FW28 HPCI MODE FAILURE TO INITIATE (11, 12 OR BOTH)

FW29 HPCI MODE INADVERTANT INITIATION (11, 12 OR BOTH)

HV01 REACTOR BUILDING EXHAUST FAN TRIP (11, 12 OR BOTH)

HV02 EMERGENCY VENTILATION FAN TRIP <11, 12 OR BOTH)

IA01 LOSS OF INSTRUMENT AIR LP01 LIQUID POISON PUMP TRIP (A, 8 OR BOTH)

ANSI 3.5 E-4 June 1987

MC01 MAIN CONDENSER AIR IN LEAKAGE MC02 STEAM JET AIR EJECTOR STEAM SUPPLY VALVE FAILS CLOSED MC03 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL HIGH MC04 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL LOW MC05 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL AS IS MC06 EXPLOSION IN AIR EJECTOR DISCHARGE PIPING MS01 STEAM LEAK RUPTURE OUTSIDE PRIMARY CONTAINMENT (DESIGN BASIS)

MS02 MSIV DISC SEPARATES FROM STEM MS03 ONE MSIV FAILS CLOSED (VALVE 122)

MS04 STEAM LINE RUPTURE INSIDE PRIMARY CONTAINMENT (DESIGN BASIS)

MS05 TURBINE STEAM SEAL REGULATOR FAILS CLOSED MS06 MOISTURE SEPARATOR DRAIN TANK LEVEL CONTROL FAILS LOW MS07 FIRST STAGE REHEATER 111 STEAM SUPPLY VALVE CLOSES MS08 SECOND STAGE REHEATER 112 STEAM SUPPLY VALVE CLOSES MS09 SECOND STAGE REHEATER 112 DRAIN TANK LEVEL CONTROL FAILS LOW MS10 LOSS Of EXTRACTION STEAM TO HIGH PRESSURE FEEDWATER HEATER (115, 125, 135 OR ANY)

MS11 LOSS OF COMPENSATION TO STEAM FLOW TRANSMITTER NM01 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - UPSCALE NM02 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - DOWNSCALE NM03 SRM CHANNEL RECORDER FAILURE (RED, BLACK OR BOTH PENS)

NM04 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - INOPERATIVE NM05 SRM CHANNEL (ll, 12, 13, 14 OR ANY) fAILURE UPSCALE, RECORDER INOPERATIVE NM06 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - DOWNSCALE NM07 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - RECORDER NM08 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE - INOPERATIVE NM09 SRM CHANNEL (11, 12, 13, 14 OR ANY) DETECTOR STUCK NM10 IRM CHANNEL (ll, 12, 13, 14, 1'5, 16, 17, 18 OR ANY) FAILURE UPSCALE NM11 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM12 IRM/APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE RECORDER NM13 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM14 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE ANSI 3.5 E-5 June 1987

NM15 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM16 IRM/APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE-RECORDER NM17 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) INOPERATIVE NM18 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) DETECTOR STUCK NM19 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE NM20 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM21 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM22 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE - UPSCALE NM23 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM24 APRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM25 ANY LPRM (X-Y-J) FAILURE - UPSCALE NM26 ANY LPRM (X-Y-J) FAILURE UPSCALE NM27 ANY LPRM (X-Y-J) FAILURE UPSCALE NM28 ANY LPRM (X-Y-J) FAILURE - DOWNSCALE NM29 ANY LPRM (X-Y-J) FAILURE - DOWNSCALE NM30 ANY LP RM (X-Y-J ) FA I I URE DOWNSCALE NM31 ANY LPRM (X-Y-J) FAILURE DOWNSCALE NM33 TIP DETECTOR STUCK IN CORE NM34 ANY LPRM (X-Y-J) DRIFT +/- 25%

NM35 ANY LPRM (X-Y-J) DRIFT +/- 25%

NM36 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE UPSCALE NM37 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE DOWNSCALE NM38 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE AS IT NM39 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE INOPERATIVE NM40 RECIRC FLOW CONVERTER (11, 12 OR BOTH) FAILURE COMPARATOR OG01 OFF-GAS RECOMBINER PREHEATER STEAM SUPPLY FAILS CLOSED OG02 OFF-GAS RECOMBINER MIXING JET STEAM SUPPLY FAILS OPEN OG03 OFF-GAS RECOMBINER MIXING JET STEAM SUPPLY FAILS CLOSED OG04 OFF-GAS DISCHARGE TO STACK ISOLATION VALVE FAILS CLOSED PC01 DRYWELL TORUS DIFFERENTIAL PRESSURE CONTROL FAILURE INCREASE PC02 DRYWELL TORUS DIFFERENTIAL PRESSURE CONTROL FAILURE DECREASE PC03 PRIMARY CONTAIN LEAKAGE ANSI 3.5 E-6 June 1987

PP01 FAILURE OF PLANT PROCESS COMPUTER R001 CONTROL ROD XX-YY FAILURE - DRIFT IN R002 CONTROL ROD XX-YY FAILURE DRIFT OUT R003 CONTROL ROD XX-YY FAILURE - ACCUMULATOR STUCK R004 CONTROL ROD XX-YY FAILURE - STUCK R005 CONTROL ROD XX-YY FAILURE UNCOUPLED RD06 CONTROL ROD XX-YY FAILURE SCRAMMED R007 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME R008 CONTROL ROD XX-YY FAILURE - RPIS R009 CONTROL ROD XX-YY FAILURE - DRIFT IN R010 CONTROL ROD XX-YY FAILURE - DRIFT OUT R011 CONTROL ROD XX-YY FAILURE - ACCUMULATOR TROUBLE RD12 CONTROL ROD XX-YY FAILURE STUCK RD13 CONTROL ROD XX-YY FAILURE UNCOUPLED R014 CONTROL ROD XX-YY FAILURE - SCRAMMED R015 CONTROL ROD XX-YY FAILURE - SLOW SCRAM'IME R016 CONTROL ROD XX-YY FAILURE - RPIS RD17 CONTROL ROD XX-YY FAILURE DRIFT IN R018 CONTROL ROD XX-YY FAILURE DRIFT OUT RD19 CONTROL ROD XX-YY FAILURE ACCUMULATOR TROUBLE R020 CONTROL ROD XX-YY FAILURE - STUCK R021 CONTROL ROD XX-YY FAILURE - UNCOUPLED R022 CONTROL ROD XX-YY FAILURE - SCRAMMED R023 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME RD24 CONTROL ROD XX-YY FAILURE - RPIS RD25 CONTROL ROD XX-YY FAILURE DRIFT IN RD26 CONTROL ROD XX-YY FAILURE DRIFT OUT RD27 CONTROL ROD XX-YY FAILURE - ACCUMULATOR TROUBLE RD28 CONTROL ROD XX-YY FAILURE STUCK R029 CONTROL ROD XX-YY FAILURE - UNCOUPLED RD30 CONTROL ROD XX-YY FAILURE SCRAMMED R031 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME RD32 CONTROL ROD XX-YY FAILURE - RPIS R033 CONTROL ROD BANK FAILURE TO SCRAM (BANK I, II, III, IV, V OR:ANY)

RD34 LOSS OF CRD INSTRUMENT AIR PRESSURE RD35 CRD HYDRAULIC PUMP TRIP (11, 12 OR BOTH)

R036 CRD FLOW CONTROL VALVE FAILURE CLOSED (11, 12 OR BOTH)

ANSI 3.5 E-7 June 1987

RD37 RPIS FAILURE COMPLETE SYSTEM FAILURE RD38 REACTOR MANUAL CONTROL SYSTEM TithER MALFUNCTION WITHDRAWN RD39 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION INSERT RD40 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION SETTLE RO41 SCRAM DISCHARGE VOLUME RUPTURE RM01 DRAWER INOPERATIVE FOR ANY PROCESS RADIATION MONITOR SIMULATED (INSTRUCTOR SELECT)

RM02 DRAWER DOWNSCALE FOR ANY AREA RADIATION MONITOR SIMULATED (INSTRUCTOR SELECT)

RM03 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM04 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM05 CONTINUOUS AIR MONITOR FAILURE (TURBINE BUILDING, REACTOR BUILDING, WASTE BUILDING, ORYWELL)

RM06 ANY PROCESS RADIATION MONITOR FAILURE RP01 REACTOR TRIP POWER SUPPLY MOTOR GENERATOR (131, 141 OR BOTH)

RP02 CONTROL POWER SUPPLY BOTH MOTOR GENERATOR TRIPS (162, 172 OR BOTH)

RP03 REACTOR SCRAM RP04 REACTOR PROTECTION SYSTEM FAILURE TO SCRAM AUTOMATIC RP05 REACTOR PROTECTION SYSTEM FAILURE TO SCRAM COMPLETE RP06 REACTOR VESSEL ISOLATION RP07 PRIMARY CONTAINMENT ISOLATION RPOS ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS)

RP09 EMERGENCY CONDENSER FAILS TO ISOLATE (11, 12 OR BOTH)

RR01 RECIRCULATION PUMP 11 DRIVE BREAKER TRIP RR02 RECIRCULATION PUMP 11, FIELD BREAKER TRIP RR03 RECIRCULATION PUMP 11 SEIZURE RR04 RECIRCULATION PUMP 11 CONTROL SIGiVAL FAILURE RR05 RECIRCULATION PUMP 11 INCOMPLETE START SEQUENCE RR06 RECIRCULATION PUMP 12 DRIVE BREAKER TRIP RR07 RECIRCULATION PUMP 12 FIELD BREAKER TRIP RROS RECIRCULATION PUtdP 12 SEIZURE RR09 RECIRCULATiON PUMP 12 CONTROL SIGNAL FAILURE RR10 RECIRCULATION PUMP 12 INCOMPLETE START SEQUENCE RR11 RECIRCULATION PUMP 13 DRIVE BREAKER TRIP RR12 RECIRCULATION PUMP 13 FIELD BREAKER TRIP RR13 RECIRCULATION PUMP 13 SEIZURE RR14 RECIRCULATION PUMP 13 CONTROL SIGNAL FAILURE RR15 RECIRCULATION PUMP 13 INCOMPLETE START SEQUENCE ANSI 3.5 E-8 June 1987

RR17 RECIRCULATION PUMP 14 FIELD BREAKER TRIP RR18 RECIRCULATION PUMP 14 SEIZURE RR19 RECIRCULATION PUMP 14 CONTROL SIGNAL FAILURE RR20 RECIRCULATION PUMP 14 INCOMPLETE START SEQUENCE RR21 RfCIRCULATION PUMP 15 DRIVE BREAKER TRIP RR22 RECIRCULATION PUMP 15 FIELD BRfAKER TRIP RR23 RECIRCULATION PUMP 15 SEIZURE RR24 RECIRCULATION PUMP 15 CONTROL SIGNAL FAILURE RR25 RECIRCULATION PUMP 15 INCOMPLETE START SEQUENCE RR26 MASTER RECIRCULATION FLOW CONTROLLER FAILURE HIGH RR27 MASTER RECIRCULATION FLOW CONTROLLER FAILURE - LOW RR28 MASTER RECIRCULATION FLOW CONTROLLER FAILURE AS IS RR30 REACTOR VESSEL PRESSURE RECORDER FAILURE (ID77) UPSCALE RR31 REACTOR VESSfL PRESSURE RECORDER FAILURE (ID77) DOWNSCALE RR32 REACTOR VESSEL PRESSURE RECORDER FAILURE (I077) AS IS RR33 RECIRCULATION PUMP LOWER (INNER) SEAL FAILURE PUMP 11 RR34 RECIRCULATION PUMP UPPER <OUTER) SEAL FAILURE - PUMP 11 RR35 REACTOR VESSEL PRESSURE iNDICATOR FAILURE (I076C) UPSCALE RR36 REACTOR VESSEL PRESSURE INDICATOR FAILURE <ID76C) DOWNSCALE RR37 REACTOR VESSEL PRESSURE INDICATOR FAILURE (ID76C) AS IS RR38 REACTOR VESSEL LEVEL RECORDER FAILURE <ID14) UPSCALE RR39 REACTOR VESSEL LEVEL RECORDER FAILURE < ID14) DOWNSCALE RR40 REACTOR VESSEL LEVEL RECORDER FAILURE ( ID14) AS IS RR41 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE UPSCALE (I0590)

RR42 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE DOWNSCALE (I0590)

RR43 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE AS IS (I0590)

RR44 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE UPSCALE (LI 36-19, CH.12)

RR45 REACTOR VESSEL LEVEL INDICATION <WIDE RANGE SAFETY SYSTEM) FAILURE DOWNS CA L E ( LI 36-19, CH. 12)

RR46 ,REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE AS IS (LI 36-19, CH.12)

RR47 RECIRCULATION PUMP DISCHARGE VALVE STEM SfPARATES FROM VALVE GATE (11, 12, 13, 14, 15 OR ANY)

ANSI 3.5 E-9 June 1987

RR48 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFf TY SYSTEM) FAILURE UPSCALE RR49 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE-DOHNSCALE RR50 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE-AS IS RR51 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS HIGH RR52 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS LOH RR53 RfACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - AS IS RR54 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-HIGH RR55 REACTOR VESSEL LEVfL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-LOH RR56 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-AS IS RR57 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS HIGH RR58 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT> FAILS LON REACTOR VESSEL PRESSURE TRANSMITTER <LOCAL-REACTOR PROT;CTION SYSTEM INPUT) FAILS - AS IS RR60 REACTOR VESSEL PRESSURE TRANSiMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS - HIGH RR61 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS I OH RR62 REACTOR VESSEL PRESSURE TRANSMITTER < LOCAL-CONTROL'YSTEM INPUT)

FAILS AS IS RR63 REACTOR RECIRCULATION PUMP 12 INNER SEAL FAILURE RR64 REACTOR RECIRCULATION PUMP 12 OUTER SEAL FAILURE RR65 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAIL'S - HIGH RR66 REACTOR RfCIRCULATION PUMP 15 TACHOMETER FAILS - LON RR67 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS OSCILLATES RR68 REACTOR RECIRCULATION PUMP M/A STATION FAILURE INCREASE (11, 12, 13, 14, 15 OR ANY)

ANSI 3.5 E-10 June 1987

RR69 REACTOR RECIRCULATION PUMP M/A STATION FAILURE - DECREASE (11, 12, 13, 14, 15 OR ANY)

RR70 REACTOR RECIRCULATION PUMP M/A STATION FAILURE AS IS (11, 12, 13, 14, 15 OR ANY)

RR71 REACTOR SAFETY VALVE INADVERTENTLY OPENS (PSV NR28A)

RR72 LOSS OF LEVEL COMPENSATION TO FEEDWATER CONTROL SYSTEM (GEMAC) LEVEL TRANSMITTER RW01 ROD WORTH MINIMIZER FAILURE RX01 FUEL CLADDING FAILURE RX02 INCREASED ROD WORTH FOR ANY CONTROL ROD SC01 SHUTDOWN COOLING PUMP TRIP (ll, 12, 13 OR ANY)

SC02 SHUTDOWN COOLING HEAT EXCHANGER TUBE LEAK (11, 12, 13 OR ANY)

TC01 MAIN TURBINE TRIP TC02 TURBINE GOVERNOR FAILS HIGH TC03 TURBINE GOVERNOR FAILS LOW TC04 ELECTRICAL PRESSURE REGULATOR FAILS HIGH TC05 ELECTRICAL PRESSURE REGULATOR FAILS LOW TC06 ELECTRICAL PRESSURE REGULATOR FAILS OSCILLATES TC07 MECHANICAL PRESSURE REGULATOR FAILS HIGH TCOB MECHANICAL PRESSURE REGULATOR FAILS LOW TC09 MECHANICAL PRESSURE REGULATOR FAILS OSCILLATES TC10 FIRST BYPASS VALVE STICKS OPEN TC11 ALL BYPASS VALVES FAIL OPEN TC12 ALL BYPASS VALVES FAIL CLOSED TC13 TURBINE CONTROL VALVE FAILS CLOSED (11, 12, 13, 14 OR ANY)

TU01 EXHAUST HOOD SPRAY VALVE FAILS CLOSED TU02 MAIN TURBINE HIGH VIBRATION BEARINGS ¹5 AND ¹6 TU03 MAIN TURBINE HIGH ECCENTRICITY TU04 MAIN TURBINE BEARING OIL LOW PRESSURE TU05 MAIN TURBINE BEARING HIGH TEMPERATURE TU06 MAIN TURBINE THRUST BEARING WEAR ANSI 3.5 E-ll June 1987

ATTACHMENT F REMOTE FUNCTIONS AD ADS NONE AN ANNUNCIATOR SYSTEM NONE CS CORE SPRAY NONE CT CONTAINMENT SPRAY RCT 1 80-43 TEST LINE TO TORUS BV OPEN CLOSE RCT 2 80-42 HASTE DISP MAN ISOLATION OPEN CLOSE CU REACTOR CLEANUP RCU1 CU-16 PCV ND37 MANUAL ISOLATION OPEN CLOSE RCUZ CU-19 FILTER BYPASS VALVE OPEN CLOSE RCU3 CU FILTER 11 INLET/OUTLET VALVE OPEN CLOSE RCU4 CU FILTER 12 INLET/OUTL'ET VALVE OPEN CLOSE RCU5 CU DEMIN ll INLET/OUTLET VALVE OPEN CLOSE RCU6 CU DEMIN 12 INLET/OUTLET VALVE OPEN CLOSE RCU7 CU-20 DEMIN BYPASS VALVES OPEN CLOSE CHl AUXILIARY HATER RCW1 INTAKE HATER TEMPERATURE 32/80 DEG 75.00 RCW2 INTAKE TUNNEL REVERSE FLOW YES NO RCH3 UPPER HIND SPEED ,

0.100 MPH 52.00 RCH4 UPPER HIND SPED VARIATION 0/30 MPH ,5.00 RCHS LONER HIND SPEED 0/100 MPH 45.00 RCW6 LOWER HIND SPEED VARIATION 0/30 MPH 5.00 RCW7 UPPER HIND DIRECTION 0/360 MPH 5.00 RCH8 UPPER HIND DIRECTION VARIATION 0/90 DEG 5.0 RCW9 LOWER HIND DIRECTION 0/360 DEG 5.0 RCH10 LONER WIND DIRECTION VARIATION 0/90 DEG 5.00 RCH11 AMBIENT AIR TEMPERATURE -30/+120 DEG 90.00 RCH12 DELTA TEMPERATURE -10/+120 DEG 10.00 CW2 AUXILIARY HATER NONE ANSI 3.5 F-1 June 1987

DG DIESEL GENERATOR RDG1 DG 102 GOVERNOR SPEED DROOP SET RESET RDG2 DG 103 GOVERNOR SPEED DROOP SET RESET EC EMERGENCY COOLING REC1 IV 39-05 VALVE POSITION LIMIT 0/100/ 100.00 REC2 IV 39-06 VALVE POSITION LIMIT 0/100% 100.00, ED1 ELECT RICAL DISTRI8 REDl SOUTH OSWEGO 115 KV BKR R10 OPEN CLOSE RED2 FITZ 115 KV BKR R40 OPEN CLOSE RED3 PB 13 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE RED4 PB 13 BUS TIE BKR SEC 8-SEC C OPEN CLOSE RED5 PB 14 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE RED6 PB 14 BUS TIE BKR SEC 8-SEC 8 OPEN CLOSE RED7 PB 15 BUS TIE BKR SEC A-SEC 8 OPEN 'CLOSE REDB PB 16 BUS TIE BKR SEC A-SEC C OPEN CLOSE RED9 MG-SET 167 AC POWER SELECT PB16 PB17 RED10 MG-SET 167 DC POWER. SELECT P811 P812 REDll COMPUTER POWER SUPPLY SELECT NORM EMER RED12 IC BUS 130 NORM BWR BKR OPEN CLOSE RED13 IC BUS 130 ALT PWR BKR OPEN CLOSE RED14 P81671 8US TIE BKR OPEN CLOSE RED15 PB131 CLOSE A-B, OPEN 13A SUPPLY YES NO RED16 P8131 CLOSE A-B, OPEN 13C SUPPLY YES NO RED17 PB141 CLOSE A-B, OPEN 14A SUPPLY YES NO RED18 P8141 CLOSE-A-B, OPEN 14C SUPPLY YES NO ED2 ELECTRICAL DISTRIB RED19 P8151 CLOSE A-B, OPEN 15A SUPPLY YES NO RED20 PB151 CLOSE A-B, OPEN 15C SUPPLY YES NO RED21 P8176 CLOSE A-B, OPEN 17A SUPPLY YES NO RED22 PB176 CLOSE A-', OPEN 16A SUPPLY YES NO RED23 BAT BDll EQUIP SW TO ALT 8811 8812 RED24 BAT BD12 EQUIP SW TO ALT 8812 ll RED25 P8143 FEEDER BREAKER 14A 14C ED3 ELECTRICAL DISTRIB NONE ANSI 3.5 F-2 June 1987

EGl HAIN GENERATOR REG1 345 KV BKR 100 42 OPEN CLOSE REG2 345 KV MAN DISC 917 OPEN CLOSE REG3 345 KV HAN DISC 926, 927 OPEN CLOSE REG4 345 KV MOD SH 18 OPEN CLOSE REG5 345 KV BKR R915/10 OPEN CLOSE REG6 345 KV BKR R925/20 OPEN CLOSE REG7 MAIN SEAL OIL PMP STATUS START NEUT REGS EMER SEAL OIL PMP STATUS START NEUT REG9 EMER SEAL OIL PMP STATUS TRIP AUTO REG10 GEN STATOR COOLING PMP 11 START NEUT REG11 GEN STATOR COOLING PHP ll TRIP AUTO REG12 GEN STATOR COOLING PMP 12 START NEUT REG13 GEN STATOR COOLING PMP 12 TRIP AUTO REG14 GENERATOR OUTPUT LINKS OPEN CLOSE REG15 GEiN HYDROGEN SUPPLY VALVE OPEN CLOSE REG16 BACKFEED iNTERLOCKS ON OFF EG2 MAIN GENERATOR NONE FP FIRE PROTECTION RFPl CITY HATER SUPPLY TO FP HDR OPEN CLOSE RFP2 SUPPLY TO EMER COOL MU TANK 11 OPEN CLOSE RFP3 SUPPLY TO EMER COOL MU TANK 12 OPEN CLOSE RFP4 SUPPLY TO FEEDWATER SYSTEM OPEN CLOSE RFP5 DIESEL FIRE PUMP STATUS OFF AUTO ANSI 3.5 F-3 June 1987

FW1 FEEDWATER RFW1 50-10 COND PUMP 11 DISCH VLV OPEN CLOSE RFW2 50-11 COND PUMP 12 DISCH VLV OPEN CLOSE RFW3 50-12 COND PUMP 13 DISCH VLV OPEN CLOSE RFW4 50-31 COND DEMIN BYPASS VLV OPEN CLOSE RFW5 COND DEMIN 11 INLET/OUTLET VLV OPEN CLOSE RFW6 COND DEMIN 12 INLET/OUTLET VLV OPEN CLOSE RFW7 COND DEMIN 13 INLET/OUTLET VLV OPEN CLOSE RFW8 COND DEMIN 14 INLET/OUTLET VLV OPEN CLOSE RFW9 COND DEMIN 15 INLET/OUTLET VLV OPEN CLOSE RFW10 COND DEMIN 16 INLET/OUTLET VLV OPEN CLOSE RFW11 50-20 SJAE BYPASS FCB 0/100/ 50.00 RFW12 50-40 BOOSTER PUMP 11 SUCTiON V OPEN CLOSE RFW13 50-39 BOOSTER PUMP 12 SUCTION V OPEN CLOSE RFW14 50-38 BOOSTER PUMP 13 SUCTION V OPEN CLOSE RFW15 FW HEATER STRING ll ISOL VLVS OPEN CLOSE RFW17 FW HEATER STRING 12 ISOL VLVS OPEN CLOSE RFW18 DEMIN WATER STORAGE TANK REFILL OPEN CLOSE FW2 FEEDWATER RFW19 50-16 BYPASS AROUND FCV 50-22 OPEN CLOSE RFW20 MANUAL OPERATION OF LCV50-15 0/100'L 0.00 RFW21 MANUAL OPERATION OF LCV50-07,08 0/100'/ 0.00 RFW22 FW HEATER 135 ISOL VALVES OPEN CLOSE RFW23 HOTWELL LEVEL CONTROL MAN AUTO FW3 FEEDWATER NONE HV HVAC NONE IA INSTRUMENT AIR RIAl INST AIR SUP TO BREATHING AIR OPEN CLOSE RIA2 BRW-G-6 WASTE DISPOSAL XTIE OPEN CLOSE RIA3 94-42 CONT SPRAY AIR RCVR ISOL OPEN CLOSE RIA4 SERV AIR TO INST AIR BV TRIP RESET ANSI 3.5 F-4 June 1987

LP LIQUID POISON RLPl LIQ POISON PMP 1 1 LOCAL START ON OFF RLP2 LIQ POISON PUMP 12 LOCAL START ON OFF RLP3 DEMIN HATER TO LP PUMPS OPEN CLOSE MC CONDENSER RMC1 OG-1,2 PRIM JFT VAP SUCT VALVES OPEN CLOSE RMC2 DG-3,4 PRIM JET VAP SUCT VALVES OPEN CLOSE RMC3 MS 114,15 PRIM JET STEAM VALVES OPEN CLOSE RMC4 MS-16,1 PRIM JET STEAM VALVES OPEN CLOSE RMC5 OG-9,10 SEC JET VAPOR SUCT VALVES OPEN CLOSE RMC6 MS-19,20 SEC JET STEAM VALVES OPEN CLOSE RMC7 MS-12 SJAE PCV BYPASS OPEN CLOSE MS1 MAIN STEAM RMSl HP FN HTR 115 RESET TRIP RESET RMS2 HP FH HTR 125 RESET TRIP RESET RMS3 HP FN HTR 135 RESET TRIP RESET RMS4 HP FN HTR STRING 11 RESET TRIP RESET RMS5 HP FN HTR STRING 12 RESET TRIP RESET RMS6 HP FH HTR STRING 13 RESET TRIP RESET RMS7 MS-8 MAIN STEAM LINE ISOL OPEN CLOSE RMS8 SPE 11 SUCTION VALVE OPEN CLOSE RMS9 SPE 12 SUCTION VALVE OPEN CLOSE RMS10 TRIP ALL FN HTR EXTR NRVS TRIP RESET MS1 MAIN STEAM NONE NM1 NEUTRON MONITOR RNMl APRM 11 GAIN 0/100'/ 2.43 RNM2 APRM 12 GAIN 0/100'/. 2. 38 RNM3 APRM 13 GAIN 0/100/. 2 '6 RNM4 APRM 14 GAIN 0/100/ 2.39 RNM5 APRM 15 GAIN 0/100/ 2.20 RNM6 APRM 16 GAIN 0/100/ 2.18 RNM7 APRM 17 GAIN 0/100'/ 2.16

. RNM8 APRM 18 GAIN 0/1001. 2.17 NM2 NEUTRON MONITOR NONE NM3 NEUTRON MONITOR, NONE ANSI 3.5 F-5 June 1987

OD ON DEMAND NON-FUNCTIONAL PC CONTAINMENT RPC1 NITROGEN FROM VAPORIZER YES NO RPC2 201.7-13 DW CAM ISOL VLV ll OPEN CLOSE RPC3 201.7-29 DW CAM ISOL VLV 12 OPEN CLOSE RPC4 201.40,41 DW, TORUS TO VENT SYSTEM OPEN CLOSE RPC5 201.44, 46 DW, TORUS TO ATMOS OPEN CLOSE RPC6 BV201.2-135,136 INTERLOCK DEFEAT YES NO RPC7 IV201-31,32 ISOLATION DEFEAT YES NO PP PROCESS COMPUTER RPP01 MEMORY PROTECT PLAN NORM REMOVD RD1 CONTROL RODS RRD1 301.2A CRD PUMP 11 DISCH VLV OPEN CLOSE RRD2 301.2B CRD PUMP 12 DISCH VLV OPEN CLOSE RRD3 301-8A CRD PUMP 11 HEAD SPRAY ISOL OPEN CLOSE RRD4 301-8B CRD PUMP 12 HEAD SPRAY ISOL OPEN CLOSE RRD5 301.8B CRD FLOW CONTROL VLV ISOL , NC30A NC30B RD2 CONTROL RODS NONE RD3 CONTROL RODS . NONE RMl RAD MONITOR NONE RP RPS RRP1 RX TRIP BUS 131 PWR SOURCE NORM EMER RRP2 RX TRIP BUS 141 PWR SOURCE NORM . EMER RRP3 RPS BUS 11 PWR SOURCE NORM EMER RRP4 BUS 12 PWR SOURCE NORM EMER ANSI 3.5 F-6 June 1987

RR1 REACTOR RECIRC RRR1 RECIRC MG-SETS 11 LOCKOUT RELAY TRIP RESET RRR2 RECIRC MG-SETS 12 LOCKOUT RELAY TRIP RESET RRR3 RECIRC HG-SETS 13 LOCKOUT RELAY TRIP RESET RRR4 RECIRC MG-SETS 14 LOCKOUT RELAY TRIP RESET RRR5 RECIRC MG-SETS 15 LOCKOUT RELAY TRIP RESET RR2 REACTOR RECIRC NONE RR3 REACTOR RECIRC NONE RR4 REACTOR RECIRC NONE RW ROD NORTH MINIMIZER RRWl CONTROL ROD SEQUENCE SELECT RX REACTOR CORE NONE SC SHUTDOWN COOLING NONE TC TURBINE CONTROL RTC1 REACTOR FLOW LIMIT 0-120'L 120.00 RTC2 CONTROL VALVE LIMIT 0-120'L 100.00 TU MAIN TURBINE NONE ANSI 3.5 F-7 June 1987

ATTACHMENT "G" MONITORED PARAMETERS CORE REACTIVITY DK/K CORE THERMAL POWER, '/

3. CORE FLOW, LBM/HR CORE PLATE DIFFERENTIAL PRESSURE, PSIG
5. CORE BORON CONCENTRATION, PPM
6. CORE AVERAGE VOID FRACTION, /
7. CORE MINIMUM CRITICAL POWER RATIO
8. CORE MAXIMUM LINEAR HEAT GENERATION, KW/FT CORE INLET SUB COOLING, BTU/LBM
10. CORE AVERAGE FUEL TEMPERATURE, DEG F CORE AVERAGE CLADDING TEMPERATURE, DEG F
12. CORE AVERAGE EXIT'QUALITY, %
13. (SPARE)
14. (SPARE)
15. REACTOR COOLANT ACTIVITY, UCI/ML
16. REACTOR COOLANT CONDUCTIVITY, UMHO/CM
17. REACTOR HEATUP/COOLDOWN RATE, DEG F/HR
18. REACTOR LEVEL-NARROW RANGE, INCHES
19. REACTOR LEVEL-WIDE RANGE, FEET
20. REACTOR PRESSURE, PSIG
21. RECIRCULATION LOOP 11 FLOW, LBM/HR
22. RECIRCULATION LOOP 12 FLOW, LBM/HR
23. RECIRCULATION LOOP 13 FLOW, LBM/HR
24. RECIRCULATION LOOP 14 FLOW, LBM/HR
25. RECIRCULATION LOOP 15 FLOW, LBM/HR
26. RECIRCULATION LOOP 11 SUCTION TEMPERATURE, DEG F
27. RECIRCULATION LOOP 12 SUCTION TEMPERATURE, DEG F
28. RECIRCULATION LOOP 13 SUCTION TEMPERATURE, DEG F
29. CRD SYSTEM FLOW, LBM/HR
30. DRYWELL PRESSURE, PSIG
31. DRYWELL AVERAGE TEMPERATURE, DEG F
32. DRYWELL HYDROGEN CONCENTRATION, /
33. DRYWELL OXYGEN CONCENTRATION, '/

ANSI 3.5 G-1 June '1987

34. SUPPRESSION CHAMBER PRESSURE, PSIG
35. SUPPRESSION POOL WATER TEMPERATURE, DEG F
36. SUPPRESSION POOL WATER LEVEL, FEET
37. SRM COUNT RATE, CPS
38. SRM PERIOD, SEC
39. APRM POWER LEVEL, '/
40. CORE XENON CONCENTRATION, '/ OF FULL POWER EQU
41. RWCU SYSTEM PRESSURE, SPIG
42. RWCU SYSTEM FLOW, LBM/HR
43. RWCU NON-REGEN HEAT EXCHAN OUTLET TEMPERATURE, DEG F
44. RWCU DUMP FLOW, LBM/HR
45. TOTAL MAIN STEAM LINE FLOW, LBM/HR
46. MAIN STEAM TUNNEL TEMPERATURE, DEG F
47. MAIN STEAM LINE RADIATION LEVEL, MR/HR
48. TOTAL MAIN STEAM RELIEF VALVE FLOW, LBM/HR
49. TURBINE SPEED, RPM
50. TURBINE INLET PRESSURE, PSIG 5.1 TURBINE STEAM FLOW, LBM/HR
52. TURBINE BYPASS VALVE STEAM FLOW, LBM/HR 53 TURBINE FIRST STAGE PRESSURE, PSIG

'4.

TURBINE EXHAUST HOOD TEMPERATURE, DEG F

55. SECOND STAGE REHEATER OUTLET PRESSURE, PSIG
56. SECOND STAGE REHEATER OUTLET TEMPERATURE, DEG F 57., CONDENSER VACUUM, IN HG V
58. CONDENSER HOTWELL LEVEL, INCHES
59. CONDENSER HOTWELL CONDUCTIVITY, UMHO/CM
60. CONDENSER VACUUM MAKEUP FLOW, LBM/HR
61. CONDENSER HOTWELL REJECT FLOW, LBM/HR
62. CONDENSATE DEPRESSION, BTU/LBM
63. CIRCULATING WATER INLET TEMPERATURE, DEG F
64. CIRCULATING WATER OUTLET TEMPERATURE, DEG F 65 .TOTAL CIRCULATING WATER FLOW, GPM
66. CONDENSATE DEMINERA OUTLET CONDUC, UMHO/CM
67. TOTAL FEEDWATER SYSTEM FLOW, LBM/HR
68. FEEDWATER TEMPERATURE TO REACTOR, DEG F
69. GENERATOR LOAD, MWE
70. GENERATOR REACTIVE LOAD, MVAR ANSI 3.5 G-2 June 1987
71. GENERATOR STATOR AMPS, AMP
72. GENERATOR TERMINAL VOLTS, VOLT
73. GENERATOR HYDROGEN PRESSURE, PSIG
74. DIESEL GENERATOR 102 LOAD, KWE
75. DIESEL GENERATOR 103 LOAD, KWE
76. OFF-GAS SYSTEM INLET. FLOW, CFW
77. OFF-GAS SYSTEM OUTLET FLOW, CFW
78. OFF-GAS RECOMBINER INLET HYDROGEN CONCENTRATION, /
79. OFF-GAS RECOMBINER OUTLET HYDROGEN CONCENTRATION, /
80. OFF-GAS SYSTEM RADIATION LEVEL, MR/HR
81. CORE SPRAY LOOP ll PRESSURE, PSIG
82. CORE SPRAY LOOP 12 PRESSURE, PSIG
83. CORE SPRAY LOOP 11, FLOW, LBM/HR
84. CORE SPRAY LOOP 12 FLOW, LBM/HR
85. EMERGENCY CONDENSER LOOP 11 FLOW, LBM/HR
86. EMERGENCY CONDENSER LOOP 12 FLOW, LBM/HR I
87. EMERGENCY CONDENSER LOOP 11 RETURN TEMPERATURE, DEG F
88. EMERGENCY CONDENSER LOOP 12 RETURN TEMPERATURE, DEG F
89. EMERGENCY CONDENSER LOOP 1 1 VENT RAD LEVEL, MR/HR
90. EMERGENCY CONDENSER LOOP 12 VENT RAD LEVEL, MR/HR ANSI 3.5 G-3 June 1987

ATTACHMENT H SIMULATOR MODIFICATION DATA BASE CHANGES AND TEST RESULTS A. All the modifications listed in Part I, Section E of this report were operationally reverified on 27 May 1987. All modifications were functionally correct with the exception of:

l. 78-27 MSIV Monitoring MSIV repositioning will initiate alarm printouts, but not the right ones.
2. 83-61 Control Room Emergency Vents Only the hardware portion of this modification is presently incorporated.
3. 84-58 Feedwater Pump Control Low flow Bypass valves fail to close on HPCI initiation.

B. Below is a list of Data Base changes by Modification Number:

Mod ¹ Document ¹ Section To Rev ¹ Nl-7804 C-180140C 5 C-35628-C 9 C-35674-C 1 C-27120-C. 2 A-22110-C 36 2 II 23 3 8-2211 1-C 1 5 C-18014-C 1 1 II 2 20 C-22136-C 8 C-22137-C 10 C-18349-C 2 3 C-26990-C 2 2 C-22101-C 4A 1 Nl-7824 No Document Changes Nl-7827 C-19859-C 11 17 C-22374-C 1 28 C-22380-C 2 12 C-23088-C 1 16 C-22376-C 4 8 C-22025-C 3 6 C-220374C 5 8 II 7 18 C-22020-C 13 3 II 14 2 C-19859-C 4 21 II 7 21 II C-19437-C 7 18 Nl-7832 C-19859-C 14 18 II II 21 6 Nl-7906 C-22379-C 6 10 Nl-7924 N1-ST-W6 6 ANSI 3.5 H-1 June 1987

NQO ¹ PLACED ON DATASASE SHEET ¹ TO REV ¹ Nl-8038 C-19954-C 2 8 3 5 5' 5 7

C-19859-C 4 21 7 21 8 32 9 16 10 19 ll 17 12 21 13 20 14 21 C-22374-C 1 28 2 6 3 19 4 32 C-22382-C 1 18 2 8 3 18 4 1 5 2 C-19440-C 11 3 C-19438-C 9 0 C-19437-C 10 3 C-34854-C 1 0 2 0 3 0 4 1 5 0 6 0 C-34853-C 1 0 2 0 3 0 4 0 5 0 6 0 6 0 C-22374-C 5 8 C-19859-C 10A 5 C-19440-C 6 14 7 16 10 6 C-19438-C 8 1 C-19437-C 6 14 7 18 9 12 C-23146-C 9 7 11 5 II 15 1 C-23145-C 3 7 C-22025-C 3 6 C-19957-C 1 22 ANSI 3.5 H-2 June 1987

MOD ¹ PLACED ON DATABASE SHEET ¹ TO REV ¹ Nl-8038 C-19866-C 0 5 C-19425-C 6 22 C-19424-C 5 20 C-19423-C 5 20 C-18015-C 0 16 C-19951-C 8 13 9 4 C-19854-C 3 10 Nl -8 041 C-18017-C 14 C-18006-C 2 5 C-18055-C 14 C-18355-C 1 15 C-18357-C 16 B-34028-C 2 1 C-19913-C 4 42 5 41 7 75 8 79 9 60 10 76 ll 61 12 73 C-19914-C 7 29 C-19474-C 1 26 C-22442-C 2 11 3 16 4 12 C-23213-C 4 9 11 14 C-23214-C 10 8 11 5 18 3 12 4 13 25 12 26 5 C-23273-C 1 2 2 2 C-22381-C 10 12 8 18 II ll 1 C-34185-C C-18006-C 1 7 C-19437-C 9 8 C-19440-C 10 3 C-19859-C 9 ll C-19951-C 9 1 C-22381-C 5 12 ll 11 New C-22373-C 2 12 3 7 C-22381-C 2 12 4 10 10 10 ANSI 3.5 H-3 June 1987

MOD ¹ PLACED ON DATABASE SHEET ¹ TO REV ¹ Nl-8041 C-22383-C 14 C-22387-C ll C-23046-C 9 C-23093-C 10 C-23100-C 6 C-26982-C 2 2

C-26992-C 4 C-26993-C 3 C-26997-C 4 C-26998-C 3 C-23109-C 6 C-18355-C 1 C-18672-C 12 C-18103-C 3 C-34185-C 0 A-22110-C 29 1 A-22110-C 39 1 C-22141-C 3 C-22140-C 3 C-22105-C 3 C-26726-C 2 ll B-19741-C 12 A 13 14 8-19758-C 2 C 3 E B-34027-C 2 4 Il 2 2 C-19468-C 1 38 C-19472-C 34 C-19473-C 2 10 C-19476-C 12 C-19478-C 16 C-19909-C 1 56 2 50 3 48 4 57 5 52 C-19818-C 2 C-19408-C 10 6 C-19438-C 1 8 C-22302-C 2 7 8 2 C'-23107-C 2 6 C-19910-C ]0 21 12 27 C-19450-C 4 46 C-19859-C 9 12 C-22020-C 4 4 II 5 4 C-22383-C 1 16 4 16 8 10 ANSI 3.5 H-4 June 1987

0 I

MOO 4 PLACED ON DATABASE SHEET 4 TO REV rf Nl-8041 C-22383-C 1 16 1 14 8 10 C-23046-C 2 10 C-19440-C 6 13 C-19409-C 10 14 C-19859-C 10 19 IOA 5 8 32 17 10 C-19954-C 8 5

5 Nl-8074 C-34829-C 2 C-34831-C 0 0

C-18014-C 35 ll 26 C-19957-C 22 C-22020-C 7 12 3

.16 0 C-22382-C 1 18 3 18 C-23146-C 8 7 ll 5 15 1 C-26726-C 2 16 C-34853-C 1 0 4 1 5 0 6 0 C-34854-C 1 0 II 1

II 5 0 II

,6 0 Nl-8084 C-19410-C 10 7 C-19425-C 3 12 C-22239-C 1 15 C-22239-C 7 12 C-23145-C 4 5 II

'o 5 5 Nl-8114 Additions Nl-8129 No Additions Nl-8230 C-22374-C 19 C-23042-C 6 Il 5

II 6

C-19423-C 19 II 10 C-19424-C 19 C-19425-C 21 C-22364-C 13 ANSI 3.5 H-5 June 1987

MOD 4 PLACED ON DATABASE SHEET ¹ TO REV ¹ Nl-8230 C-22365-C 14 C-22366-C 14 C-22367-C 15 C-22368-C 12 Nl-8269 C-19473-C 1 C-99424-C 12 C-23076-C C-19897-C 2 2 C-19423-C 8 3 C-19424-C 6 3 13 C-22374-C 5 C-23077-C 3 5

C-22386-C 2 1

C-22372-C 2 1 11 C-22030-C 3 5 C-22004-C 5 5 C-19954-C 4 7 C-19859-C 2 23 5 23 C-26726-C 3 9 C-23077-C 5 6 C-18005-C 1 10 2 10 C-26727-C 1 2 C-26726-C 17 C-19423-C 3 15 C-23077-C 1 10 2 6 Nl-8271 C-23145-C 5 12 4 11 C-19409-C 3 17 C-19410-C 1 15 2 16 3 14 4 15 5 14 6 15 10 14 ll 4 12 Nl-8280 C-22005-C 11 5 12 4 14 6 C-22374-C 2 6 3 16 C-22382-C 1 16

3. 14 C-22356-C 1 7 2 6 ANSI,3. 5 H-6 June 1987

MOD ¹ PLACED ON DATABASE SHEET 4 TO REV 4 Nl-8280 C-19437-C 10 3 2 23 C-19440-C 11 3 C-22005-C 2 6 6 6 5 6 8 7 9 6 C-19859-C 3 24 6 24 8A ll 8 5 C-19437-C 10 3 C-19438-C 3 9 C-19842-C 2 4 C-19845-C 7 1 C-19859-C 2 24 5 24 10 16 10A 3 18 9 C-19440-C 2 22 6 10 10 6 Nl-8293 C-22380-C 3 2 II 1

I Nl-8329 No Additions Nl-8353 A-22217-C 20 4 II 47 4 C-22238-C 1 16 2 23 C-22239-C 7 18 C-23305-C 5 C-23311-C 4 C-19940-C 1 22 C-19411-C 1 8 C-19415-C 1 19 II 2 18 C-19417-C 1 16 II 2 5 C-19418-C 1 14 2 19 3 10 4 16 Nl-8 358 C-26726-C 1 16 2 16 3 10 II 4 21 C-19859-C 18 11 II 18A 1 C-22105-C 2 8 Nl-8361 C-22387-C 5 3 C-22388-C 1 6 II 3 2 ANSI 3. 5 H-7 June 1987

MOD 4 PLACED ON DATABASE SHEET ¹ TO REV ¹ Nl-8361 C-22387-C 1 7 I

C-?6726-C 6 8 Nl-8384 C-18022-C 1 15 2 8 C-18022-C 1 1 2 18 C-18027-C 2 5 Nl-8458 C-19951-C 3 13 C-22004-C 2 11 II 5 7 C-23077-C 4 8 C-23146-C 7 9 Nl-8526 C-19859-C 18 11 18A 1 C-22373-C ll 9 12 17 13 19 C-22374-C 4 28 4 32 5 8 C-26726-C 4 21 ANSI 3.5 H-8 June 1987

ATTACHMENT I MALFUNCTIONS TESTED IN 1987 FN14 RR37 EG06 FW18 RR41 EG10 FN22 RR45 EG14 FN26 RR49 FP04 HV01 RR53 FP08 MC01 RR57 FNOZ MC05 RR61 FN06 MS03 RR65 FN10 MS06 RR69 MSll RNOl NM04 SC02 NM12 R~TCqb (

NM20 TC08 NM33 TC12 NM38 TU03 OG02 AD02 R002 A006 R006 RD34 RD38 RM01 RM06

~

CS03 CU01 CN02 CN06 C'u oq RP02 CN10 RP06 EC02 RR02 EC06 RR06 ED03 BAR R>~>4 E007 RR14 EDll RR18 E015 RR22 E019 RR29 E023 RR33 EG02 ANSI 3.5 I-1 June 1987

NINE MILE POINT NUCLEAR STATION UNIT 1 PLANT REFERENCED SIMULATOR ANNUAL REPORT: ANSI 3.5 1985 FOR THE YEAR 1988 Testin Conducted Aorii - Ma 1988 Re ort Prepared June 1988 Prepared By: George Roarick Reviewed By Supervi , Uteri 0 era i ns Training Date Asst. Super'ende of Training Superintendent of Training Date

0 0-

I. SIMULATOR INFORMATION The purpose of this section is to provide familiarization with the Nine Mile Point Unit Plant Referenced Simulator and its applicability as an 1

operator training device.

A. General The simulator is owned by the General Physics Niagara Corporation and is a wholly owned subsidiary of the General Physics Corporation. It is used jointly by General Physics Corporation and Niagara Mohawk Power Corporation instructors.

It is maintained and modified by the General Physics Corporation under the direction of Niagara Mohawk Power Corporation. The simulator was built by Singer/Link.

2 ~ The simulator is a full scope control room simulator that simulates the Nine Mile Unit ¹1 plant. The plant is an 1850 Megawatt Thermal, BNR-2 plant with an electrical output of 620 Megawatts. The plant uses a GE BNR Mark I Primary Containment.

The simulator was declared ready for training on September 1, 1984.

The initial report on the simulator was prepared in March of 1986. The first annual report was completed in June of 1987.

In addition to training, the simulator has been used by the Operations Department to validate procedures prior to their implementation.

-1 June 1988 Unit 1 Ops/1S88

8. Control Room (Physical Fidelity)
1. The physical layout of the referenced plant's Control Room is shown in Attachment "A". The physical layout of the simulator is shown in Attachment "8". A comparison of these diagrams shows a high degree of similarity between the two rooms. The following differences exist:
a. The instructor's console in the simulator uses the same area occupied by the Control Room Clerk's Office in the reference plant' Control Room.

A manual dose assessment calculator is mounted on the wall next to the stairwell in the referenced plant's control room. A meteorological computer is located next to it. This equipment does not exist in the simulator.

A video camera is mounted on the wall above the NSSS typer. There are no video ca~eras permanently mounted in the simulator.

There are some minor differences in the amount and type of furniture in both rooms.

2. Panels and Equipment The simulator contains all of the panels that are in the referenced plant's Control Room. All front panels and the "E" console are fully simulated. The back panels are fully simulated with the exception of the following:
a. RPS relays are installed but are not functional.

Electrical protective relaying is cosmetically simulated by photos in the relay enclosures.

Only one of the four Transversing In-Core Probe panels is functional.

d. Seismic Monitors are not functional.

-2 June 1988 Unit 1 Ops/1588

3. Systems All systems that are operable from the referenced plant's Control Room are simulated. See Attachment "C" for a list of these systems.
4. Simulator Control Room Environment The Simulator Control Room was specifically designed to duplicate as nearly as possible the referenced plant's Control Room. Other than the discrepancies noted in Section 1 of this part, the following differences exist:
a. Even though the lighting is identical to that at the referenced plant, the lighting system is not functionally interfaced with the simulator. Plans to interface the lights and the simulator are being evaluated.
b. The ambient noise level of the referenced plant's Control Room is not simulated. The hardware in place to do this is installed but is not functional. Plans to do this are being evaluated.

C. Instructor Interface (Control Capabi 1 i ties)

l. Initial Conditions The instructor has the capability to initialize the simulator to any one of fifty (50) sets of initial conditions. The first twenty of these sets of conditions are guarded and can only be changed by the proper code. The "guarded" initial, conditions are the foundation for approximately 95% of the training. These twenty sets of initial conditions are listed in Attachment "0". The remaining thirty sets '.of initial conditions can be set at any time by the instructor.
2. Malfunctions Malfunctions vary from a discrete nature (i.e. pump trip) to ones of varying degrees of severity (i.e. leaks). Attachment "E" is a complete list of malfunctions.

-3 June 1988

3. Remote Functions The instructor has the capability to simulate most of the inplant operations needed to backup a Control Room evolution.

Attachment "F" is a complete list of remote functions.

4. Instructor Override The I/O function gives the instructor at the console the ability to override all switches, energize or deenergize any light or alarm, and drive all meters. A few problems exist in the I/O program, but are being documented for evaluation and repair as they are identified.
5. Monitoring The instructor can monitor on the console, up to sixteen (16) of ninety (90) parameters. In addition, the instructor can set up a line plotter to any of these same ninety (90) parameters. Attachment "G" is a list of these monitored parameters.
6. Instructor Station Control s Consists of three (3) keyboard/CRT's, two (2) are for

'performing simulator functions, one (1) is for minor trouble shooting and correction of the computer system. There are also various buttons which perform the following:

a. Freeze - stops <freezes) simulation at any point or restarts it.
b. Reset/Ready initfalizes the simulator to a selected set of initial conditions. This button turns green when all controls are properly positi'oned.

C. Snapshot records and stores a set of conditions into any "IC".

d. Malfunction Clear clears the entire malfunction tableau.

-4 June 1988 Unit 1 Ops/1588

e. Backtrack sets the simulator to step up or back to any point in time, within sixty 60 minutes. This includes buttons to step in one minute intervals and to step forward or reverse.

Manual Malfunction Control used to increase or decrease the severity of up to three (3) variable malfunctions.

g. Annunciator Silence silences, but does not acknowledge, all annunciators.

Recorder Off shuts off power to all chart recorders.

Test and Lamp Test testing and troubleshooting.

Emergency Stop - kills all power to the computer and simulator.

Computer Alarm/Acknowledge - warns of computer malfunctions.

Program Overtime - warns of computers inability to t

complete a cal cul ation wi thin an al lotted time frame.

Depressing the button resets the timer.

m. Record and Replay - starts and stops the tape recorder in the Computer Room when it is set put to record an exercise.
n. Slow Time - slows simulator response to where two (2) real time seconds equal one (1) problem time second.
o. Fast Time speeds up response time of Xenon, condenser evacuation, and turbine warming.
7. Record/Replay 1f so desired, a scenario can be recorded on magnetic tape for replay at a later date. This function is controlled from the instructor station.

-5 June 1988 Unit 1 Ops/1588

0. Reference Plant Operating Procedures The simulator is operated using the same procedures used to operate the plant. The Operations Department has used the simulator to validate new procedures or revisions.

E. Changes Since Last Report (June 1987)

1. Plant Modifications as of June 15, 1988 a ~ J-Panel Upgrades Installation of RSSB ARM meters (non-functional) and installation of new digital (Kaman) radiation monitoring panels for Control Room Ventilation, Service Water discharge, Rad Waste discharge.

(Functional)

b. N1Y86M057 Rod Worth Minimizer - Added new process computer points for RWM failures.

C. Mod ¹N1-80-072, Alternate Rod Injection (ARI) Alternate Rod insertion is a redundant method of rod injection, utilizing two one inch DC solenoid valves in series with the Control Rod Scram Air Header System. Initiation signal to operate these valves wi 11 be ATWS LoLo water level and ATWS high reactor pressure.

d. Mod ¹Nl-85-098, Containment Isolation on High Radiation-Shuts "N2 E,Air Vent 8 Purge valves on High Stack act>vity.
2. Physical Layout
a. The instructor's station was moved onto a twenty inch (20") platform adjacent to the SSS's office. (This area at the plant is occupied by the ASSS/STA office see Attachment "A").
b. A direct phone link between the instructor's station and the CSO's desk was'installed to simulate the direct phone link to the Energy Management Center an the NRC/state.

-6 June 1988 Unit 1 Ops/1588

II. SIMULATOR DESIGN DATA The current des i gn data base for the s imul ator i s on fi 1 e Training Center. A current listing of the design data base is available on the plant's Configuration Management System in the simulator data file. Changes to the design data base since June 1987 are listed in Attachment "H".

III. SIMULATOR TESTS (PERFORMANCE TESTING)

During May of 1988, testing was done to verify real time operation, steady state and normal operation, transient performance and malfunction response. Documentation of these tests is on file and available at the Training Center.

A. Computer Real Time Tests 4

Simulator real time testing was performed by measuring individual model times during steady state and transient conditions. During this testing, no frame slippage or program overtimes occurred. The simulator contains safeguards tha't preclude operation outside of real time with the exception of two (2) instructor controlled functions described in Section I.C.6 of this report (Fast Time/Slow Time).

B. Steady State and Normal Operation

1. Simulator stability was verified by comparing heat balances (P-1 edits) and printouts of selected "critical" parameters from the start and finish of a sixty (60) minute steady state run. There was no deviation outside of acceptable limits in regards to the "critical" parameters. There is, however, a problem with rising exhaust hood temperatures. This problem was found to be generic to all power operating conditions and C

was documented for correction.

-7 June 1988 Unit 1 Ops/1588

2. Fidelity in performance was verified by comparing heat bal ances and "cri ti cal" parameters from the pl ant at various power levels.

a ~ Generic Problems

1) Reactor Physics - Simulator was not able to duplicate the power'o flow relationship from a cold startup configuration, even with all rods out.
2) Plant Efficiency Even though the simulator was within tolerances at full power, during power ascension, the calculated efficiency was as much as five percent (5/.) lower than that of the plant for a similar configuration.
3) Off-Gas System During start up and power ascension, the Off-Gas System could not be operated in accordance with the procedure or ATP. The mechanical vacuum pumps were unable to achieve the vacuum they could at the plant within the actual time frame seen at the plant. Post recombine flow was erratic'nd too high and both the off-gas vacuum pumps were unable to maintain pressure. (It normally requires only one pump at the plant.)
4) Recirculation Pump Flows - Individual pump flows at the plant vary as much as 100,000 ibm/hr. On the simulator, they are almost equal.
5) Narrow Range GEMAC Level At the plant, it runs about a two inch (2") difference in channel ¹11 and

¹12 NR GEHAC's. In the simulator, they are equal.

Of the generic problems noted, items 1, 2, and 3 were documented for correction and assigned immediate priority. Items 4 and 5 have very little if any training impact and will not be addressed any further.

b. 25/ Power Plant data for total steam flow is significantly different than that from the simulator, but the plant data appears to be incorrect based on the rest of the data. There were no other problems other than the generic ones listed above.

-8 June 1988 Unit 1 Ops/1588

C. 50% Power - Simulator data compared favorably wi th the plant data except for the generic problems listed above.

One OG vacuum pump can handle the volume but flow is still too high and the slightest change causes off-gas to respond significantly.

d. 75% Power Simulator data compared favorably with plant data except for the generic problems listed above. OG system behavior was consistent to that experienced at 50%

power.

e. 100'/ Power Simulator data compared favorably with plant data except for the generic problems noted above. OG system behavior was consistent to that at 50% and 75%

power.

C. Transient Tests

1. The following transients are analyzed in the FSAR and were run on the simulator in real time a ~ Simultaneous Trip of all Feedwater Pumps (Feedwater Malfunction, Zero Flow). The FSAR shows an increase in recirculation control due to changes in two-phase flow.

The simulator does not model two-phase flow closely enou'gh to produce this; therefore, it does not change.

This discrepancy has little or no training impact and will not be further addres'sed.

b. Simultaneous Closure of MSIV's - Consistent with FSAR.

C. Simultaneous Trip of all Recirculation Pumps - consistent with FSAR.

d. Single Recirculation Pump Trip The simulator shows a much more pronounced decrease and recovery of recirculation flow than the FSAR. Parameters effected by flow respondent likewise.

-9 June 1988 Unit 1 Ops/1588

e. Loss of Coolant with a Loss of A/C Power At time + 3.5 seconds, Drywell pressured spiked to 10 psig then dropped to 8 psig with an increasing trend to a maximum pressure of about 22 psig over a period of 30 seconds. This is inconsistent with the FSAR which shows a pressure spike to 33 psig in 2 seconds. This inconsistency was identified and documented last year.
2. The following transients are not analyzed in the FSAR but are required by ANSI 3.5 1985.
a. Main Steam Line Break Inside the Drywell Drywell pressure in the first 3 seconds of the transient spiked to 11 psig. During the next second, it decreased to 8 psig before starting a steady increase to about 15 psig.

This problem is documented for further evaluation and resolution. Otherwise, the simulator was consistent with predicted response.

b. Turbine Trip at less than 40'/ Power The simulator was consistent with predicted response.
c. Power Change; 100 and 75'/..- 100'/ (Maximum Ramp) The simulator was consistent with predicted response.
d. MSIV closure with a stuck open ERV and no high pressure ECCS available.

The simulator was consistent with predicted response.

3. The following. transients are not required for the report by the standard but were conducted and compared to data from the FSAR.
a. Turbine Trip without Bypass
b. Inadvertent Actuation of an ERV
c. Safety Valve Actuation
d. EPR/MPR Failures In all cases, the simulator response was consistent to the FSAR.

-10 June 1988 Unit 1 Ops/1588

D. Malfunction Testing In addit'ion to the malfunctions used to create the transient tests above 251. of the available malfunctions (Attachment "E") were tested in accordance with the ATP. Attachment "I" is a list of the malfunctions tested. Operations Department personnel were on hand to he p va i date 1 1 the response of the s i mu1 ator to the malfunctions. A number of problems were documented. They will be handl ed in accordance wi th NTI-4.5. 3 "S imu1 ator Conf i gurat i on Management".

IV. SIMULATOR CONFIGURATION MANAGEMENT Discrepancy resolution and modifications are handled in accordance with NTI-4.5.3 "Simulator Configuration Management". This is a computer based tracking system, The system has identified seven modifications that fall outside the implementation guidelines of ANSI 3.5 1985.

Attachment "J" provides a list of these modifications and the status of each at the writing of this report.

In 1987 the Plant's Configuration Management System was implemented to track and maintain the plant's data base, this system also contains a program for tracking, maintaining, and generating reports on the simulator's data base. Once an audit of the hard copy is complete and a more complete list is available, this information wi 11 be inputed on the system.

                                                                                          • 0****
  • FINAL NOTE
  • All data used to compile this report as well *
  • as NTI-4.5.3 are on file and available for
  • inspection.

-11 June 1988 Unit 1 Ops/1588

ATTACHMENT A

~Le end Main Fire Panel ¹2 Panel A Electric/

Turbine Controls Panel B Relays/

Turbine Controls Plant Process Computer Equipment Chief Shift Operator Desk E Panel Main Control Console F Panel NSSS G Panel Nuclear Instrumentation H Panel Balance of Plant Instructors Console J Panel Radiation Monitoring K K Panel ECCS L L Panel Primary Containment M Panel RPS N Panel Turbine P Print Rack S Storage Cabinets T Tables M/R Meteorological Computer Manual Dose Assessment Calculator R/T Radio/Telephone Equipment 0 Procedures B/S Bookshelf 0/K Desk Figure 1 Nine Mile Point Unit 1 Control Room

-12 June 1988 Unit 1 Ops/1588

ATTACHMENT B teceend 2 Main Fire Panel 42 A Panel A El ectri c/

Turbine Controls B Panel 8 Relays/

Turbine Controls C Plant Process Computer Equipment 0 Chief Shift Operator Desk E E Panel Main Control Console F F Panel NSSS G G Panel Nuclear Instrumentation H H Panel Balance of Plant I Instructors Console J J Panel Radiation Monitoring K, K Panel ECCS L L Panel Primary Containment M M Panel RPS N N Panel Turbine P Print Rack S Storage Cabinets T Tables 0 Procedures B/S Bookshelf F/C False Column F/S False Stairway F/D False Door R/T Radio/Telephone Equipment Figure 2 Nine Mile Point Unit 1 Simulator Control Room

-13 June 1988 Unit 1 Ops/1588

Attachment C SYSTEMS FULLY SIMULATED

l. Nuclear Boiler and Instrumentation
2. Reactor Recirculation System
a. Reactor Recirculation Loops
b. Boiler Process Instrumentation
c. Recirculation Flow Control
3. Control Rod Drive and Hydraulics System (CRDHS)
4. Reactor Manual Control System (RMCS)
5. Reactor Core (Physics and Thermodynamics)
a. Reactor Core Neutron Kinetics
b. Reactor Core Thermodynamics
6. Rod Worth Minimizer (RWM)
7. Main Steam Systems
a. Main Steam and Main Steam Bypass Systems
b. Moisture Separators Reheaters
c. Extraction Steam System
d. Auxiliary Steam System
8. Reactor Water Cleanup System Nuclear Instrumentation System
a. Source Range Monitor (SRM) System
b. Intermediate Range Monitor (IRM) System
c. Local Power Range Monitoring (LPRM) System
d. Average Power Range Monitoring (APRM) System
e. Rod Block Monitor (RBM) System
f. Traversing In-Core Probe (TIP) System
10. Reactor Protection System
11. Simulation of the Primary Containment and Isolation System
a. Primary Containment
b. Primary Containment Isolation System
12. Secondary Containment
13. Emergency Ventilation
a. Reactor Building Ventilation
b. Turbine Building Ventilation
c. Building Ventilation
14. Primary Containment Atmosphere Control and Sampling System

-14 June 1988, Unit 1 Ops/1588

'h

15. Emergency Core Cooling Systems
a. Automatic Depressurization and pressure Relief System
b. Core Spray
c. High Pressure Coolant Injection (HPCI) System
d. Containment Spray
e. Emergency Cooling System
16. Shutdown Cooling
17. Standby Liquid Control (SLC) System
18. Condensate and Feedwater System
a. Condensate System
b. Condensate Demineralizer System
c. Feedwater System
d. Condensate Storage and Transfer System
e. Reactor Vessel Level Control System
f. Feedwater Heaters, Vents and Drains
19. Off-Gas Recombiner and Condenser Air Removal
20. Main Condenser
21. Circulating Water System
22. Reactor Building Closed Loop Cooling
23. Turbine Building Closed Loop Cooling
24. Service Water System
25. Instrument, Service and Breathing Air
26. Area Radiation Monitoring System
27. Process Radiation Monitoring System
28. Ventilation Radiation Monitoring System
29. Main Turbine and Turbine Control
a. Turbine Oil System
b. Turbine Kinematics
c. Turbine Mechanics
d. Turbine Supervisory and Safety System
e. Gland Seal System
f. Low Pressure Hood Spray System
g. Moisture Separator and Reheat System
h. Main Turbine Electro-Hydraulic Control System

-15 June 1988 Unit 1 Ops/1588

30. Plant Electrical System
a. Main Generator and Auxiliary Systems
1. Main Generator Synchronous Ma.chine
2. Excitation and Voltage Regulator System
3. Synchroscope
4. Hydrogen Cooling System
5. Stator and iso-Phase Duct Cooling System
6. Hydrogen Seal Oil System
b. Electrical Distribution System
1. Buses and Transformers
2. Breakers
3. Currents, Voltages and Frequencies
4. OC Electrical Distribution and Control
5. Power System Electrical Grid
c. Diesel Generators
31. Containment Atmosphere Dilution, Vent and Purge System
32. Radiation Waste Disposal System Containment Equipment and Floor Drain Sump
33. Plant Carbon Dioxide System
34. Diesel Fire Pump and Pressurized Water Fire System
35. Fire Control Ventilation Systems
36. Control Room Heating, Vent) lation and Air Conditioning
37. Communication System
38. Plant Process Computer System
a. Applicable Experience
39. Meteorological Experience
40. Plant Annunciators and Fire System Alarm

-16 June 1988 Unit 1 Ops/1588

0-ATTACHMENT D GUARDED INITIAL CONDITIONS IC ¹ DESCRIPTION Cold Iron - All systems are off line with the exception of Service Hater and Electrical Distribution. This is not an a roved lant lineu but is used to rovide trainin in s stem startu s in the Control Room.

2. Cold Startup - Shutdown Cooling in service with all systems needed for startup of the reactor on line and ready to pull rods. (After securing SDC.)
3. Cold Startup 5-10 rods subcritical, pulling in Group 4.

4, Heatup 100¹ - Pulling Group 9

5. Heatup 250¹ - Pulling Group 11
6. Heatup 950¹ - Pulling Group 23
7. Shutdown 80'1. Flow Start of controlled plant shutdown per OP-43
8. Shutdown Min Flow Inserting Group 76
9. Shutdown Min Flow - Mode Switch to "Shutdown", Inserting Group 33
10. Shutdown Min Flow All rods in Turbine Startup 3.5 Bypass Valves Open, Turbine Harmed
12. Feedwater Pump ¹13 Startup - 210 MHE, on two electric FNP's
13. 50K Power - Power accent to threshold and precondi tioning
14. 1001. Power - Normal full power configuration
15. Coastdown All rods out at end of cycle
16. Startup Rods at 100'/ target pattern, 301. flow
17. Reserved for Future Use
18. Reserved for Future Use
19. Cooldown 200¹ - Plant Cooldown
20. Cooldown 80¹ Shutdown Cooling In Service

-17 June 1988 Unit 1 Ops/1588

ATTACHMENT E NINE MILE POINT UNIT ONE MALFUNCTIONS AD01 ADS FAILURE TO INITIATE PRIMARY VALVES A002 ADS FAILURE TO INITIATE COMPLETE A003 SOLENOID ACTUATED PRESSURE RELIEF VALVE <¹111) FAILURf SOLENOID AD04 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹111) FAILURE VALVE LfAKS AD05 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹1 1 1 ) FAILURE OPENS INADVERTENTLY A006 SOLENOID ACTUATED PRESSURE RELIEF VALVE (¹111) FAILURE STUCK OPEN ANOl CONTROL ROOM ANNUNCIATOR SYSTEM FAILURE CS01 CORE SPRAY PUMP TRIP (111, 112, 121, 122 OR ANY)

CS02 CORE SPRAY TOPPING PUMP TRIP (111, 112, 121, 122 OR ANY)

CS03 CORE SPRAY INBOARD INJECTION VALVE FAILURE TO OPEN (IV40-01, IV40-09, IV40-11, IV40-10 OR ANY)

CT01 CONTAINMENT SPRAY PUMP TRIP (111, 112, 121, 122 OR ANY)

CT02 CONTAINMENT SPRAY RAW WATER PUMP TRIP (111, 112, 121, 122 OR ANY)

CT03 CONTAINMENT SPRAY HEAT EXCHANGER (111, 112, OR BOTH) TUBE LEAK CU01 COOLANT LEAKAGE INSIDE PRIMARY CONTAINMENT CU02 REACTOR WATER CLEANUP PUMP TRIP (11, 12 OR BOTH)

CU03 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-ND22) FAILS OPEN CU04 REACTOR WATER CLEANUP REJECT FLOW CONTROL VALVE (FCV-ND22) FAILS CLOSED CU05 REACTOR WATER CLEANUP HIGH PRESSURE CONTROL VALVE < PCV 33-39) FAILS OPEN CU06 REACTOR WATER CLEANUP HIGH PRESSURE CONTROL VALVE <PCV 33-39) FAILS CLOSED CU07 REACTOR WATER CLEANUP LOW PRfSSURE CONTROL '/ALVE (PCV-ND37) FAILS OPEN CU08 REACTOR WATER CLEANUP LOW PRESSURE CONTROL VALVE (PCV-ND37) FAILS CLOSED CU09 REACTOR WATER CLEANUP NON-REGENERATIVE HEAT EXCHANGER TUBE LEAK CU10 REACTOR WATER CLfANUP DEMINERALIZER RESIN DEPLETION (11, 12 OR BOTH)

CU11 COOLANT LEAKAGE OUTSIDE PRIMARY CONTAINMENT CW01 HIGH RADIATION IN SERVICE WATER CW02 SERVICE WATER PUMP TRIP (11, 12 OR BOTH)

-18 June 1988 Unit 1 Ops/1588

CW03 EMERGENCY SERVICE WATER PUMP TRIP (11, 12 OR BOTH)

CW04 REACTOR BUILDING CLOSED LOOP'OOLING (11, 12, 13 OR ANY) PUMP TRIP CW05 TURBINE BUILDING CLOSED LOOP COOLING PUMP TRIP (ll, 12 OR BOTH)

CW06 CIRCULATING WATER PUMP TRIP (11, 12 OR BOTH)

CW07 CIRCULATING WATER EXPANSION JOINT LEAKAGE CW08 CIRCULATING WATER INTAKE STRUCTURE ICING CW09 LOSS OF DRYWELL COOLING CW10 MAiN CONDENSER TUBE LEAK DG01 DIESEL GENERATOR FAILURE TO START (102, 103 OR BOTH)

DG02 DIESEL GENERATOR TRIP (102, 103 OR BOTH)

EC01 STEAM LEAKAGE INSIDE PRIMARY CONTAINMENT EC02 STEAM LEAKAGE OUTSIDE PRIMARY CONTAINMENT EC03 EMERGENCY COOLING SYSTEM RETURN VALVE FAILS OPEN (IV39-05, IV39-06 OR BOTH)

EC04 EMERGENCY COOLING SYSTEM RETURN BIVALVE FAILS TQ OPEN (IV39-05, IV39-06 OR BOTH)

EC05 EMERGENCY COOLING SYSTEM EMERGENCY CONDENSER MAKEUP CONTROL VALVE FAILS CLOSED (LCV60-17, LCV60-18 OR BOTH)

EC06 EMERGENCY CONDENSER TUBE LEAK <111, 121 OR BOTH)

E001 LOSS OF QFF-SITE 115 KV POWER SOURCES (LIGHTHOUSE HILL-JAF, OSWEGO STEAM, OR BOTH)

ED02 BATTERY CHARGER AND EMERGENCY LIGHTING SUPPLY MOTOR GEHERATOR TRIPS (161, 171 OR BOTH) ~

ED03 COMPUTER POWER SUPPLY MOTOR GENERATOR TRIPS (167)

ED04 AC POWERBOARD ELECTRICAL FAULT (PB11)

ED05 AC POWERBOARD ELECTRICAL FAULT (PB12)

ED06 AC POWERBOARD ELECTRICAL FAULT (PB101)

ED07 AC POWERBQARD ELECTRICAL FAULT (PB102)

ED08 AC POWERBOARD ELECTRICAL FAULT (PB103)

ED09 AC POWERBOARD ELECTRICAL FAULT <PB13 SECTION A)

ED10 AC POWERBQARD ELECTRICAL FAULT (PB13 SECTION B)

EDll AC POWERBOARD ELECTRICAL FAULT (PB13 SECTION C)

ED12 AC POWERBOARD ELECTRICAL FAULT (PB14 SECTION A)

ED13 AC POWERBOARD ELECTRICAL FAULT (PB14 SECTION 8)

ED14 AC POWERBOARD ELECTRICAL FAULT (PB14 SECTION C)

ED15 AC POWERBOARD ELECTRICAL FAULT (PB15 SECTION A)

ED16 AC POWERBOARD ELECTRICAL FAULT (PB15 SECTION 8)

ED17 AC POWERBOARD ELECTRICAL FAULT (PB15 SECTION C)

-19 June 1988 Unit 1 Ops/1588

0 ED18 AC POWERBOARD ELECTRICAL FAULT <P816 SECTION A)

ED19 AC POWERBOARD ELECTRICAL FAULT (P816 SECTION 8)

ED20 AC POWERBOARD ELECTRICAL FAULT (PB17 SECTION A)

ED21 AC POWERBOARD ELECTRICAL fAULT (PB18 SECTION 8)

E022 OC POWERBOARD ELECTRICAL FAULT J(11, 12 OR BOTH)

ED23 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - NORMAL ED24 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - ALTERNATE ED25 LOSS OF POWER TO INSTRUMENT CONTROL BUS 130 - NORMAL AND ALTERNATE EG01 'AIN GENERATOR TRIP ELECTRICAL FAULT EG02 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS INCREASE EG03 GENERATOR AUTOMATIC VOLTAGE REGULATOR FAILS DECREASE EG04 MAIN GENERATOR CORE INTERNAL HEATING EG05 MAIN TRANSFORMER LOSS OF COOLING EG06 GENERATOR HYDROGEN COOLING SYSTEM LEAKAGE EG07 GENERATOR HYDROGEN MAIN SEAL OIL PUMP FAILURE EGOS GENERATOR HYDROGEN EMERGENCY SEAL OIL PUMP FAILURE EG09 STATOR COOLING PUMP TRIP (11, 12 OR BOTH)

EG10 LOSS OF CONTROL AIR TO 345 KV BREAKER (R-915, R-925 OR BOTH)

EG11 POWER GRID NETWORK LOAD TRANSIENT INCREASE EG12 POWER GRID NETWORK LOAD TRANSIENT DECREASE EG13 STATOR WATER COOLING OEMINERALIZER RESIN DEPLETION FP01 DIESEL FIRE PUMP FAILURE FP02 ELECTRIC FIRE PUMP FAILURE FP03 AC FOAM PUMP FAILURE FP04 OC FOAM PUMP FAILURE FP05 TURBINE ISLAND FIRE DETECTION <0-1195, 0-1155, D-1165, 0-1175, D-1061, 0-1114, D-1131 OR ANY)

FP06 CONTROL ROOM FIRE DETECTION (FIRE PANEL 2, CONTROL CONSOLE, "L" PANEL, "K" PANEL, "H" PANEL, "F" PANEL, "A" PANEL OR ANY) fP07 TURBINE BUILDING FIRE DETECTION (DA-22092MG., DA-2083M, DA-2081S, OA1092E, D-2102 OR ANY)

FPOS DIESEL ROOM FIRE DETECTION (DX-2113A, DX-21138, DX-02141A, DA-2141, OX-21518, DA-2151, 0-2151 OR ANY) fP09 AUXILIARY CONTROL ROOM/cable spreading room fire detection (d-3031PL, DX-3031A, DX-31118, WD-8131, WD-8082 OR ANY)

FP10 REACTOR BUILDING FIRE DETECTION (DX-4217A, DA-4'116W, DA-4076E, 0-4207, 0-4156, SP-4126, 0-4086 OR ANY)

-20 June 1988 Unit 1 Ops/1588

0 FW01 CONDENSATE PUMP TRIP (11, 12, 13 OR ANY)

FW02 FEEDWATER BOOSTER PUMP TRIP (11, 12, 13 OR ANY)

FW03 FEEDWATER PUMP TRIP (11, 12 OR BOTH)

FW04 SHAFT DRIVEN FEEDWATER PUMP 13 FAILURE FW05 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAILURE TO ENGAGE fW06 SHAFT DRIVEN FEEDWATER PUMP CLUTCH FAILURE TO DISENGAGE FW07 FEEDWATER CONTROL VALVE 11 CONTROLLER FAILS HIGH FW08 FEEDWATER CONTROL VALVE 11 CONTROLLER FAILS LOW FW09 FEEDWATER CONTROL VALVE 12 CONTROLLER FAILS HIGH FW10 FEEDWATER CONTROL VALVE 12 CONTROLLER FAILS LOW FW11 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS HIGH FW12 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS LOW FW13 FEEDWATER CONTROL VALVE 13 CONTROLLER FAILS AS IS FW14 FEEDWATER MASTER CONTROLLER FAILS HIGH FW15 FEEDWATER MASTER CONTROLLER FAILS LOW FW16 FEEDWATER MASTER CONTROLLER FAILS AS IS FW17 CONDENSATE DEMINERALIZER DEPLETION FW18 FEEDWATER CONDUCTIVITY INCREASE FW19 CONDENSATE RECIRCULATION VALVE (FCV 50-24) FAILS OPEN FW20 CONDENSATE RECIRCULATION VALVE <FCV 50-24) FAILS CLOSED FW21 FEEDWATER BOOSTER PUMP RECIRCULATION VALVE FAILS OPEN <FCV 51-58, FCV 51-59, FCV 51-60 OR ANY)

FW22 FEEDWATER HEATER TUBE LEAK FW23 FEEDWATER PUMP RECIRCULATION VALVES FAIL OPEN (ll, 12, 13 0 R ANY)

FW24 FEEDWATER CONTROL VALVE FAILS CLOSED (13A, 138 OR BOTH)

FW25 THREE MILE ISLAND ACCIDENT (BWR EQUIVALENT)

FW26 CONDENSATE BYPASS SPRAY TO MAIN CONDENSER FLOW CONTROL VALVE (FCV 50-22) FAILS CLOSED FW27 LOSS OF COMPENSATION TO FEEDWATER FLOW TRANSMITTER FW28 HPCI MODE FAILURE TO INITIATE <11, 12 OR BOTH)

FW29 HPCI MODE Inadvertent INITIATION (11, 12 OR BOTH)

HV01 REACTOR BUILDING EXHAUST FAN TRIP (11, 12 OR BOTH)

HV02 EMERGENCY VENTILATION FAN TRIP (11, 12 OR BOTH)

IA01 LOSS OF INSTRUMENT AIR LP01 LIQUID POISON PUMP TRIP (A, 8 OR BOTH)

-21 June 1988 Unit 1 Ops/1588

MC01 MAIN CONDENSER AIR IN LEAKAGE MC02 STEAM JET AIR EJECTOR STEAM SUPPLY VALVE FAILS CLOSED MC03 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL HIGH MC04 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL LOW MC05 HOTWELL LEVEL CONTROLLERS IN AUTO FAIL AS IS MC06 EXPLOSION IN AIR EJECTOR DISCHARGE PIPING MSOl STEAM LEAK RUPTURE OUTSIDE PRIMARY CONTAINMENT (DESIGN BASIS)

MS02 MSIV DISC SEPARATES FROM STEM MS03 ONE MSIV FAILS CLOSED (VALVE 122)

MS04 STEAM LINE RUPTURE INSIDE PRIMARY CONTAINMENT (DESIGN BASIS)

MS05 TURBINE STEAM SEAL REGULATOR FAILS CLOSED MS06 MOISTURE SEPARATOR DRAIN TANK LEVEL CONTROL FAILS LOW MS07 FIRST STAGE REHEATER 111 STEAM SUPPLY VALVE CLOSES MS08 SECOND STAGE REHEATER 112 STEAM SUPPLY VALVE CLOSES MS09 SECOND STAGE REHEATER 112 DRAIN TANK LEVEL CONTROL FAILS LOW MS10 LOSS OF EXTRACTION STEAM TO HIGH PRESSURE FEEDWATER HEATER <115, 125, 135 OR ANY)

MSll LOSS OF COMPENSATION TO STEAM FLOW TRANSMITTER NM01 SRM CHANNEL (ll, 12, 13, 14 OR ANY) FAILURE UPSCALE NM02 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE DOWNSCALE NM03 SRM CHANNEL RECORDER FAILURE (RED, BLACK OR BOTH PENS)

NM04 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE INOPERATIVE NM05 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE UPSCALE, RECORDER INOPERATIVE NM06 SRM CHANNEL (11, 12, 13, 14 OR ANY) FAILURE DOWNSCALE NM07 SRM CHANNEL <11, 12, 13, 14 OR ANY) FAILURE RECORDER NM08 SRM CHANNEL (ll, 12, 13, 14 OR ANY) FAILURE INOPERATIVE NM09 SRM CHANNEL (ll, 12, 13, 14 OR ANY) DETECTOR STUCK NM10 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE NM11 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM12 IRM/APRM CHANNEL ( 11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE RECORDER NM13 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM14 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE

-22 June 1988 Unit 1 Ops/1588

NM15 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE I DOWNSCALE NM16 IRM/APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE RECORDER NM17 IRM CHANNEL (ll, 12, 13, 14, 15, 16, 17, 18 OR ANY) INOPERATIVE NM18 IRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) DETECTOR STUCK NM19 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE NM20 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM21 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM22 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE UPSCALE NM23 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE DOWNSCALE NM24 APRM CHANNEL (11, 12, 13, 14, 15, 16, 17, 18 OR ANY) FAILURE INOPERATIVE NM25 ANY LPRM (X-Y-J) FAILURE - UPSCALE NM26 ANY LPRM (X-Y-J) FAILURE - UPSCALE NM27 ANY LPRM (X-Y-J) FAILURE - UPSCALE NM28 ANY LPRM (X-Y-J) FAILURE DOWNSCALE NM29 ANY LPRM (X-Y-J) FAILURE DOWNSCALE NM30 ANY LPRM (X-Y-J) FAILURE - DOWNSCALE NM31 ANY LPRM (X-Y-J) FAILURE DOWNSCALE NM33 TIP DETECTOR STUCK IN CORE NM34 ANY LPRM (X-Y-J) DRIFT +/- 25/

NM35 ANY LPRM (X-Y-J) DRIFT +/- 25%

NM36 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE UPSCALE NM37 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE DOWNSCALE NM38 RECIRC FLOW CONVERTER CHANNEL ( 11, 12 OR BOTH) FAILURE - AS IT NM39 RECIRC FLOW CONVERTER CHANNEL (11, 12 OR BOTH) FAILURE INOPERATIVE NM40 RECIRC FLOW CONVERTER (11, 12 OR BOTH) FAILURE COMPARATOR OG01 OFF-GAS RECOMBINER PREHEATER STEAM SUPPLY FAILS CLOSED OG02 OFF-GAS RECOMBINER MIXING JET STEAM SUPPLY FAILS OPEN OG03 OFF-GAS RECOMBINER MIXING JET STEAM SUPPLY FAILS CLOSED OG04 OFF-GAS DISCHARGE TO STACK ISOLATION VALVE FAILS CLOSED PC01 DRYWELL TORUS DIFFERENTIAL PRESSURE CONTROL FAILURE INCREASE PC02 DRYWELL TORUS DIFFERENTIAL PRESSURE CONTROL, FAILURE DECREASE PC03 PRIMARY CONTAIN. LEAKAGE

-23 June 1988 Unit 1 Ops/1588

PP01 FAILURE OF PLANT PROCESS COMPUTER R001 CONTROL ROD XX-YY FAILURE DRIFT IN R002 CONTROL ROD XX-YY FAILURE DRIFT OUT R003 CONTROL ROD XX-YY FAILURE ACCUMULATOR STUCK R004 CONTROL ROD XX-YY FAILURE - STUCK RD05 CONTROL ROD XX-YY FAILURE - UNCOUPLED RD06 CONTROL ROD XX-YY FAILURE - SCRAMMED R007 CONTROL ROD XX-YY FAILURE SLOW SCRAM TIME RDOB CONTROL ROD XX-YY FAILURE RPIS R009 CONTROL ROD XX-YY FAILURE - DRIFT IN R010 CONTROL ROD XX-YY FAILURE DRIFT OUT RD11 CONTROL ROD XX-YY FAILURE ACCUMULATOR TROUBLE RD12 CONTROL ROD XX-YY FAILURE STUCK RD13 CONTROL ROD XX-YY FAILURE UNCOUPLED RD14 CONTROL ROD XX-YY FAILURE - SCRAMMED RD15 CONTROL ROD XX-YY FAILURE SLOW SCRAM TIME R016 CONTROL ROD XX-YY FAILURE RPIS RD17 CONTROL ROD XX-YY FAILURE - DRIFT IN RD18 CONTROL ROD XX-YY FAILURE - DRIFT OUT RD19 CONTROL ROD XX-YY FAILURE - ACCUMULATOR TROUBLE RDZO CONTROL ROD XX-YY FAILURE - STUCK R021 CONTROL ROD XX-YY FAILURE UNCOUPLED R022 CONTROL ROD XX-YY FAILURE - SCRAMMED RD23 CONTROL ROD XX-YY FAILURE - SLOW SCRAM TIME RD24 CONTROL ROD XX-YY FAILURE RPIS RD25 CONTROL ROD XX-YY FAILURE DRIFT IN R026 CONTROL ROD XX-YY FAILURE - DRIFT OUT R027 CONTROL ROD XX-YY FAILURE ACCUMULATOR TROUBLE RD28 CONTROL ROD XX-YY FAILURE - STUCK R029 CONTROL ROD XX-YY FAILURE -

UNCOUPLED'ONTROL RD30 ROD XX-YY FAILURE SCRAMMED R031 CONTROL ROD XX-YY FAILURE SLOW SCRAM TIME RD32 CONTROL ROD XX-YY FAILURE RPIS RD33 CONTROL ROD BANK, FAILURE TO SCRAM (BANK I, II, I II, IV, V OR :ANY)

RD34 LOSS OF CRD INSTRUMENT AIR PRESSURE RD35 CRD HYDRAULIC PUMP TRIP (ll, 12 OR BOTH)

R036 CRD FLOW CONTROL VALVE FAILURE CLOSED (11, 1Z OR BOTH)

-24 June 1988 Unit 1 Ops/1588

RD37 RPIS FAILURE COMPLETE SYSTEM FAILURE RD38 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION WITHDRAWN RD39 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION INSERT RD40 REACTOR MANUAL CONTROL SYSTEM TIMER MALFUNCTION - SETTLE RD41 SCRAM DISCHARGE VOLUME RUPTURE RM01 DRAWER INOPERATIVE FOR ANY PROCESS RADIATION MONITOR SIMULATED (INSTRUCTOR SELECT)

RM02 DRAWER DOWNSCALE FOR ANY AREA RADIATION MONITOR SIMULATED (INSTRUCTOR SELECT)

RM03 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM04 DRAWER UPSCALE FOR ANY AREA RADIATION MONITOR SIMULATED RM05 CONTINUOUS AIR MONITOR FAILURE (TURBINE BUILDING, REACTOR BUILDING, WASTE BUILDING, DRYWELL)

RM06 ANY PROCESS RADIATION MONITOR FAILURE RP01 REACTOR TRIP POWER SUPPLY MOTOR GENERATOR (131, 141 OR BOTH)

RP02 CONTROL POWER SUPPLY BOTH MOTOR GENERATOR TRIPS (162, 172 OR BOTH)

RP03 REACTOR SCRAM RP04 REACTOR PROTECTION SYSTEM FAILURE TO SCRAM - AUTOMATIC RP05 REACTOR PROTECTION SYSTEM FAILURE TO SCRAM COMPLETE RP06 REACTOR VESSEL ISOLATION RP07 PRIMARY CONTAINMENT ISOLATION RP08 ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS)

RP09 EMERGENCY CONDENSER FAILS TO ISOLATE (11, 12 OR BOTH)

RR01 RECIRCULATION PUMP 11 DRIVE BREAKER TRIP RR02 RECIRCULATION PUMP 11 FIELD BREAKER TRIP RR03 RECIRCULATION PUMP 11 SEIZURE RR04 RECIRCULATION PUMP 11 CONTROL SIGNAL FAILURE RR05 RECIRCULATION PUMP 11 INCOMPLETE START SEQUENCE RR06 RECIRCULATION PUMP 12 DRIVE BREAKER TRIP RR07 RECIRCULATION PUMP 12 FIELD BREAKER TRIP RR08 RECIRCULATION PUMP 12 SEIZURE RR09 RECIRCULATION PUMP 12 CONTROL SIGNAL FAILURE RR10 RECIRCULATION PUMP 12 INCOMPLETE START SEQUENCE RR11 RECIRCULATION PUMP 13 DRIVE BREAKER TRIP RR12 RECIRCULATION PUMP 13 FIELD BREAKER TRIP RR13 RECIRCULATION PUMP 1 I 3 S E ZURE RR14 RECIRCULATION PUMP 13 CONTROL SIGNAL FAILURE RR15 RECIRCULATION PUMP 13 INCOMPLETE START SEQUENCE

-25 June 1988 Un1t 1 Ops/1588

RR17 RECIRCULATION PUMP 14 FIELD BREAKER TRIP RRl 8 RECIRCULATION PUMP 14 SEIZURE RR19 RECIRCULATION PUMP 14 CONTROL SIGNAL FAILURE RR20 RECIRCULATION PUMP 14 INCOMPLETE START SEQUENCE RR21 RECIRCULATION PUMP 15 DRIVE BREAKER TRIP RR22 RECIRCULATION PUMP 15 FIELD BREAKER TRIP RR23 RECIRCULATION PUMP 15 SEIZURE RR24 RECIRCULATION PUMP 15 CONTROL SIGNAL FAILURE RR25 RECIRCULATION PUMP 15 INCOMPLETE START SEQUENCE RR26 MASTER RECIRCULATION FLOW CONTROLLER FAILURE HIGH RR27 MASTER RECIRCULATION FLOW CONTROLLER FAILURE - LOW RR28 MASTER RECIRCULATION FLOW CONTROLLER FAILURE AS IS RR30 REACTOR VESSEL PRESSURE RECORDER FAILURE (ID77) UPSCALE RR31 REACTOR VESSEL PRESSURE RECORDER FAILURE (I077) DOWNSCALE RR32 REACTOR VESSEL PRESSURE RECORDER FAILURE (ID77) AS IS RR33 RECIRCULATION PUMP LOWER (IiVNER) SEAL FAILURE PUMP 1 1 RR34 RECIRCULATION PUMP UPPER (OUTER) SEAL FAILURE PUMP 11 RR35 REACTOR VESSEL PRESSURE INDICATOR FAILURE (ID76C) UPSCALE RR36 REACTOR VESSEL PRESSURE INDICATOR, FAILURE (ID76C) DOWNSCALE RR37 REACTOR VESSEL PRESSURE INDICATOR FAILURE (ID76C) AS IS RR38 REACTOR VESSEL LEVEL RECORDER FAILURE (ID14) - UPSCALE RR39 REACTOR VESSEL LEVEL RECORDER FAILURE < I014) - DOWNSCALE RR40 REACTOR VESSEL LEVEL RECORDER FAILURE (ID14) - AS IS RR41 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE UPSCALE (I059D)

RR42 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE DOWNSCALE (I059D)

RR43 REACTOR VESSEL LEVEL INDICATION (CONTROL SYSTEM) FAILURE AS IS (ID590)

RR44 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE UPSCALE (L: 36-19, CH.12)

RR45 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE-DOWNSCALE (LI 36-19, CH.12)

RR46 REACTOR VESSEL LEVEL INDICATION (WIDE RANGE SAFETY SYSTEM) FAILURE-AS IS (LI 36-19, CH. 12)

RR47 RECIRCULATION PUMP DISCHARGE VALVE STEM SEPARATES FROM VALVE GATE

<11, 12, 13, 14, 15 OR ANY)

-26 June 1988 Unit 1 Ops/1588

RR48 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE UPSCALE RR49 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE DOWNSCALE RR50 REACTOR VESSEL LEVEL INDICATION (FUEL ZONE SAFETY SYSTEM) FAILURE-AS IS RR51 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS - HIGH RR52 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS LOW RR53 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS AS IS RR54 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS HIGH RR55 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-LOW RR56 REACTOR VESSEL LEVEL TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT) FAILS-AS IS RR57 REACTOR VESSEL PRESSURE TRANSMITTER < LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS HIGH RR58 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS LOW RR59 REACTOR VESSEL PRESSURE TRANSMITTER < LOCAL-REACTOR PROTECTION SYSTEM INPUT) FAILS AS IS RR60 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS HIGH RR61 REACTOR VESSEL PRE SSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS LOW RR62 REACTOR VESSEL PRESSURE TRANSMITTER (LOCAL-CONTROL SYSTEM INPUT)

FAILS AS IS RR63 REACTOR RECIRCULATION PUMP 12 INNER SEAL FAILURE RR64 REACTOR RECIRCULATION PUMP 12 OUTER SEAL FAILURE RR65 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - HIGH RR66 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - LOW RR67 REACTOR RECIRCULATION PUMP 15 TACHOMETER FAILS - OSCILLATES RR68 REACTOR RECIRCULATION PUMP M/A STATION FAILURE INCREASE (11, 12, 13, 14, 15 OR ANY)

-27 June 1988 Unit 1 Ops/1588

RR69 REACTOR RECIRCULATION PUMP M/A STATION FAILURE DECREASE (11, 12, 13, 14, 15 OR ANY)

RR70 REACTOR RECIRCULATION PUMP M/A STATION FAILURE - AS IS (11, 12, 13, 14, 15 OR ANY)

RR71 REACTOR SAFETY VALVE INADVERTENTLY OPENS (PSV NR28A)

RR72 LOSS OF LE>/EL COMPENSATION TO FEEDWATER CONTROL SYSTEM (GEMAC)

LEVEL'RANSMITTER RW01 ROD WORTH MINIMIZER FAILURE RX01 FUEL CLADDING FAILURE RX02 INCREASED ROD WORTH FOR ANY CONTROL ROD SC01 SHUTDOWN COOLING PUMP TRIP (11, 12, 13 OR ANY)

SC02 SHUTDOWN COOLING HEAT EXCHANGER TUBE LEAK (11, 12, 13 OR ANY)

TC01 MAIN TURBINE TRIP TC02 TURBINE GOVERNOR FAILS HIGH TC03 TURBINE GOVERNOR FAILS LOW TC04 ELECTRICAL PRESSURE REGULATOR FAILS HIGH TC05 ELECTRICAL PRESSURE REGULATOR FAILS LOW TC06 ELECTRICAL PRESSURE REGULATOR FAILS OSCILLATES TC07 MECHANICAL PRESSURE REGULATOR FAILS - HIGH TC08 MECHANICAL PRESSURE REGULATOR FAILS - LOW TC09 MECHANICAL PRESSURE REGULATOR FAILS - OSCILLATES TC10 FIRST BYPASS VALVE STICKS OPEN TC11 ALL BYPASS VALVES FAIL - OPEN TC12 ALL BYPASS VALVES FAIL - CLOSED TC13 TURBINE CONTROL VALVE FAILS CLOSED (11, 12, 13, 14 OR ANY)

TU01 EXHAUST HOOD SPRAY VALVE FAILS CLOSED TU02 MAIN TURBINE HIGH VIBRATION BEARINGS ¹5 AND ¹6 TU03 MAIN TURBINE HIGH ECCENTRICITY TU04 MAIN TURBINE BEARING OIL LOW PRESSURE TU05 MAIN TURBINE BEARING HIGH TEMPERATURE TU06 MAIN TURBINE THRUST BEARING WEAR

-28 June 1988 Unit 1 Ops/1588

ATTACHMENT F REMOTE FUNCTIONS AD ADS NONE AN ANNUNCIATOR SYSTEM NONE CS CORE SPRAY NONE CT CONTAINMENT SPRAY RCT 1 80-43 TEST LINE TO TORUS BV OPEN CLOSE RCT 2 80-42 HASTE DISP MAN ISOLATION OPEN, CLOSE CU REACTOR CLEANUP RCUl CU-16 PCV ND37 MANUAL ISOLATION OPEN CLOSE RCU2 CU-19 FILTER BYPASS VALVE OPEN CLOSE RCU3 CU FILTER 11 INLET/OUTLET VALVE OPEN CLOSE RCU4 CU FILTER 12 INLET/OUTLET VALVE OPEN CLOSE RCU5 CU DEMIN 11 INLET/OUTLET VALVE OPEN CLOSE RCU6 CU DEMIN 12 INLET/OUTLET VALVE OPEN CLOSE RCU7 CU-20 DEMIN BYPASS VALVES OPEN CLOSE CW1 AUXILIARY HATER RCH1 INTAKE WATER TEMPERATURE 32/80 DEG 75.00 RCH2 INTAKE TUNNEL REVERSE FLOW YES NO RCH3 UPPER HIND SPEED 0.100 MPH 52.00 RCW4 UPPER WIND SPED VARIATION 0/30 MPH 5.00 RCH5 LOWER WIND SPEED 0/100 MPH 45.00 RCW6 LOWER WIND SPEED VARIATION 0/30 MPH 5.00 RCW7 UPPER HIND DIRECTION 0/360 MPH 5.00 RCWB UPPER HIND DIRECTION VARIATION 0/90 DEG 5.0 RCW9 LOWER WIND DIRECTION 0/360 DEG 5.0 RCH10 LOWER WIND DIRECTION VARIATION 0/90 DEG 5.00 RCH11 AMBIENT AIR TEMPERATURE -30/+120 DEG 90.00 RCH12 DELTA TEMPERATURE -10/+120 DEG 10.00 CW2 AUXILIARY WATER NONE

-29 June 1988 Unit 1 Ops/1588

DG DIESEL GENERATOR RDG1 DG 102 GOVERNOR SPEED DROOP SET RESET RDG2 DG 103 GOVERNOR SPEED DROOP SET RESET EC EMERGENCY COOLING REC1 IV 39-05 VALVE POSITION LIMIT 0/100'/ 100.00 REC2 IV 39-06 VALVE POSITION LIMIT 0/100/ 100.00 ED1 ELECTRICAL DISTRIB REDl SOUTH OSWEGO 115 KV BKR R10 OPEN CLOSE RED2 FITl. 115 KV BKR R40 OPEN CLOSE RED3 PB 13 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE RED4 PB 13 BUS TIE BKR SEC 8-SEC C OPEN CLOSE RED5 PB 14 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE RED6 PB 14 BUS TIE BKR SEC 8-SEC 8 OPEN CLOSE RED7 PB 15 BUS TIE BKR SEC A-SEC 8 OPEN CLOSE REDB PB 16 BUS TIE BKR SEC A-SEC C OPEN CLOSE RED9 MG-SET 167 AC POWER SELECT P816 17 RED10 MG-SET,167 DC POWER SELECT PB11 12 REDll COMPUTER POWER SUPPLY SELECT NORM EMER RED12 IC BUS 130 NORM PWR BKR OPEN CLOSE RED13 IC BUS 130 ALT PWR BKR OPEN CLOSE RED14 P81671 BUS TIE BKR OPEN CLOSE RED15 PB131 CLOSE A-B, OPEN 13A SUPPLY YES NO RED16 PB131 CLOSE A-B, OPEN 13C SUPPLY YES NO RED17 PB141 CLOSE A-B, OPEN 14A SUPPLY YES NO RED18 P8141 CLOSE A-B, OPEN 14C SUPPLY YES NO ED2 ELECTRICAL DISTRIB RED19 PB151 CLOSE A-B, OPEN 15A SUPPLY YES NO RED20 P8151 CLOSE A-B, OPEN 15C SUPPLY YES 'NO RED21 P8176 CLOSE A-B, OPEN 17A SUPPLY YES NO RED22 PB176 CLOSE A-B, OPEN 16A SUPPLY YES NO RED23 BAT BD11 EQUIP SW TO ALT 8811 8812 RED24 BAT 8012 EQUIP SW TO ALT 8812 ll RED25 P8143 FEEDER BREAKER 14A 14C ED3 ELECTRICAL DISTRI 8 NONE

-30 June 1988 Unit 1 Ops/1588

EGl MAIN GENERATOR REG1 345 KV BKR 100 42 OPEN CLOSE REG2 345 KV MAN DISC 917, OPEN CLOSE REG3 345 KV MAN DISC 926, 927 OPEN CLOSE REG4 345 KV MOD SH 18 OPEN CLOSE REGS 345 KV BKR R915/10 OPEN CLOSE REG6 345 KV BKR R925/20 OPEN CLOSE REG7 HAIN SEAL OIL PMP STATUS START NEUT REG8 EHER SEAL OIL PMP STATUS START NEUT REG9 EHER SEAL OIL PMP STATUS TRIP AUTO REG10 GEN STATOR COOLING PMP ll START NEUT REG11 GEN STATOR COOLING PMP 11 TRIP AUTO REG12 GEN STATOR COOLING PMP 12 START NEUT REG13 GEN STATOR COOLING'PMP 12 TRIP AUTO REG14 GENERATOR OUTPUT LINKS OPEN CLOSE REG15 GEN HYDROGEN SUPPLY VALVE OPEN CLOSE REG16 BACKFEED INTERLOCKS ON OFF EG2 HAIN GENERATOR NONE FP FIRE PROTECTION RFP1 CITY HATER SUPPLY TO FP HDR OPEN CLOSE RFP2 SUPPLY TO EMER COOL MU TANK ll OPEN CLOSE RFP3 SUPPLY TO EMER COOL MU TANK 12 OPEN CLOSE RFP4 SUPPLY TO FEEDWATER SYSTEM OPEN CLOSE RFP5 DIESEL FIRE PUMP STATUS OFF AUTO

-31 June 1988 Unit 1 Ops/1588

FWl FEEDWATER RFW1 50-10 COND PUMP 11 DISCH VLV OPEN CLOSE RFW2 50-11 COND PUMP 12 DISCH VLV OPEN CLOSE RFW3 50-12 COND PUMP 13 DISCH VLV OPEN CLOSE RFW4 50-31 COND DEMIN BYPASS VLV OPEN CLOSE RFWS COND DEMIN 11 INLET/OUTLET VLV OPEN CLOSE RFW6 COND DEMIN 12 INLET/OUTLET VLV OPEN CLOSE RFW7 COND DEMIN 13 INLET/OUTLET VLV OPEN CLOSE RFW8 COND DEMIN 14 INLET/OUTLET VLV OPEN CLOSE RFW9 COND DEMIN 15 INLET/OUTLET VLV OPEN CLOSE RFW10 COND DEMIN 16 INLET/OUTLET VLV OPEN CLOSE RFW11 50-20 SJAE BYPASS FCB 0/100'/ 50.00 RFW12 50-40 BOOSTER PUMP 11 SUCTION V OPEN CLOSE RFW13 50-39 BOOSTER PUMP 12 SUCTION V OPEN CLOSE RFW14 50-38 BOOSTER PUMP 13 SUCTION V OPEN CLOSE RFW15 FW HEATER STRING ll ISOL VLVS OPEN CLOSE RFW17 FW HEATER STRING 12 ISOL VLVS OPEN CLOSE RFW18 DEMIN WATER STORAGE TANK'EFILL OPEN CLOSE FW2 FEEDWATER RFW19 50-16 BYPASS AROUND FCV 50-22 OPEN CLOSE RFW20 MANUAL OPERATION OF LCV50-15 0/100'/ 0.00 RFW21 MANUAL OPERATION OF LCV50-07,08 0/100/ 0.00 RFW22 FW HEATER 135 ISOL VALVES OPEN CLOSE RFW23 HOTWELL LEVEL CONTROL MAN AUTO FW3 FEEOWATER NONE HV HVAC NONE IA INSTRUMENT AIR RIA1 INST AIR SUP TO BREATHING AIR OPEN CLOSE RIA2 BRW-G-6 WASTE DISPOSAL XTIE OPEN CLOSE RIA3 94-42 CONT'SPRAY AIR RCVR ISOL OPEN CLOSE RIA4 SERV AIR TO, INST AIR BV TRIP RESET

-32 June 1988 Unit 1 Ops/1588

0 LP LIQUID POISON RLP1 LIQ POISON PMP 11 LOCAL START ON OFF RLP2 LIQ POISON PUMP 12 LOCAL START ON OFF RLP3 DEMIN WATER TO LP PUMPS OPEN CLOSE MC CONDENSER RMC1 OG-1,2 PRIM JET VAP SUCT VALVES OPEN CLOSE RMC2 DG-3,4 PRIM JET VAP SUCT VALVES OPEN CLOSE RMC3 MS 114,15 PRIM JET STEAM VALVES, OPEN CLOSE RMC4 MS-16,1 PRIM JET STEAM VALVES OPEN CLOSE

'RMC5 OG-9,10 SEC JET VAPOR SUCT VALVES OPEN CLOSE RMC6 MS-19,20 SEC JET STEAM VALVES OPEN CLOSE RMC7 MS-12 SJAE PCV BYPASS OPEN CLOSE MS1 MAIN STEAM RMS1 HP FW HTR 115 RESET TRIP RESET RMS2 HP FW HTR 125 RESET TRIP RESET RMS3 HP FW HTR 135 RESET TRIP RESET RMS4 HP FW HTR STRING ll RESET TRIP RESET RMSS HP FW HTR STRING 12 RESET TRIP RESET RMS6 HP FW HTR STRING 13 RESET TRIP RESET RMS7 MS-8 MAIN STEAM LINE ISOL OPEN CLOSE RMS8 SPE 11 SUCTION VALVE OPEN CLOSE RMS9 SPE 12 SUCTION VALVE OPEN CLOSE RMS10 TRIP ALL FW HTR EXTR NRVS TRIP RESET MS1 MAIN STEAM NONE NM1 NEUTRON MONITOR RNMl APRM 11 GAIN 0/100'/ 2.43 RNM2 APRM 12 GAIN 0/100/ 2.38 RNM3 APRM 13 GAIN 0/100% 2.36 RNM4 APRM 14 GAIN 0/100/ 2.39 RNM5 APRM 15 GAIN 0/100/ 2.20 RNM6 APRM 16 GAIN 0/100/ 2.18 RNM7 APRM 17 GAIN 0/100/ 2.16 RNM8 APRM 18 GAIN 0/100'/ 2.17 NM2 NEUTRON MONITOR NONE NM3 NEUTRON MONITOR NONE

-33 June 1988

0 OD ON DEMAND NON-FUNCTIONAL PC CONTAINMENT RPC1 NITROGEN FROM VAPORIZER YES NO RPCZ 201.7-13 DN CAM ISOL VLV 11 OPEN CLOSE RPC3 201.7-29 DN CAM ISOL VLV 12 OPEN CLOSE RPC4 201.40,41 DN, TORUS TO VENT SYSTEM OPEN CLOSE RPCS 201.44, 46 DN, TORUS TO ATMOS OPEN CLOSE RPC6 BV201.2-135,136 INTERLOCK DEFEAT YES NO RPC7 IV201-31,32 ISOLATION DEFEAT YES NO PP PROCESS COMPUTER RPP01 MEMORY PROTECT PLAN NORM REHOVD RD1 CONTROL RODS RRD1 301.2A CRD PUMP 11 DISCH VLV OPEN CLOSE RRD2 301.2B CRD PUMP 12 DISCH VLV OPEN CLOSE RRD3 301-BA CRD PUMP ll HEAD SPRAY ISOL OPEN CLOSE RRD4 301-8B CRD PUMP 12 HEAD SPRAY ISOL OPEN CLOSE RRD5 301.8B CRD FLOH CONTROL VLV ISOL NC30A NC308 RD2 CONTROL RODS NONE RD3 CONTROL RODS NONE RM1 RAD MONITOR NONE RP 0

RPS RRP1 RX TRIP BUS 131 PWR SOURCE NORM EHER RRP2 RX TRIP BUS 141 PNR SOURCE NORM EHER RRP3 RPS BUS 11 PNR SOURCE NORM EHER RRP4 BUS 12 PHR SOURCE NORM EHER

-34 June 1988 Unit 1 Ops/1588

RRl REACTOR RECIRC RRRl RECIRC MG-SETS 11 LOCKOUT RELAY TRIP RESET RRR2 RECIRC MG-SETS 12 LOCKOUT RELAY TRIP RESET RRR3 RECIRC MG-SETS 13 LOCKOUT RELAY TRIP RESET RRR4 RECIRC MG-SETS 14 LOCKOUT RELAY TRIP RESET RRR5 RECIRC MG-SETS 15 LOCKOUT RELAY TRIP RESET RR2 REACTOR RECIRC NONE RR3 REACTOR RECIRC NONE RR4 REACTOR RECIRC NONE RN ROD NORTH MINIMIZER RRNl CONTROL ROD SEQUENCE SELECT RX REACTOR CORE NONE SC SHUTDONN COOLING NONE TC TURBINE CONTROL RTC1 REACTOR FLON LIMIT 0-120'/ 120.00 RTC2 CONTROL VALVE LIMIT 0-120'/ 100.00 TU MAIN TURBINE NONE

-35 June 1988 Unit 1 Ops/1588

ATTACHMENT "G" MONITORED PARAMETERS

l. CORE REACTIVITY DK/K
2. CORE THERMAL POWER, '/
3. CORE FLOW, LBM/HR
4. CORE PLATE DIFFERENTIAL PRESSURE, PSIG
5. CORE BORON CONCENTRATION, PPM
6. CORE AVERAGE VOID FRACTION, /
7. CORE MINIMUM CRITICAL POWER RATIO
8. CORE MAXIMUM LINEAR HEAT GENERATION, KW/FT
9. CORE INLET SUB COOLING, BTU/LBM
10. CORE AVERAGE FUEL TEMPERATURE, DEG F
11. CORE AVERAGE CLADDING TEMPERATURE, DEG F
12. CORE AVERAGE EXIT QUALITY, /
13. (SPARE)
14. (SPARE)
15. REACTOR COOLANT ACTIVITY, UCI/ML
16. REACTOR COOLANT CONDUCTIVITY, UMHO/CM
17. REACTOR HEATUP/COOLDOWN RATE, DEG F/HR
18. REACTOR LEVEL-NARROW RANGE, INCHES
19. REACTOR LEVEL-WIDE RANGE, FEET
20. REACTOR PRESSURE, PSIG
21. RECIRCULATION LOOP 1 1 FLOW, LBM/HR
22. RECIRCULATION LOOP 12 FLOW, LBM/HR
23. RECIRCULATION LOOP 13 FLOW, LBM/HR
24. RECIRCULATION LOOP 14 FLOW, LBM/HR
25. RECIRCULATION LOOP 15 FLOW, LBM/HR
26. RECIRCULATION LOOP 11 SUCTlON TEMPERATURE, DEG F
27. RECIRCULATION LOOP 12 SUCTION TEMPERATURE, DEG F
28. RECIRCULATION LOOP 13 SUCTION TEMPERATURE, DEG F
29. CRD SYSTEM FLOW, LBM/HR
30. DRYWELL PRESSURE, PSIG
31. DRYWELL AVERAGE TEMPERATURE, DEG F
32. DRYWELL HYDROGEN CONCENTRATION, '/
33. DRYWELL OXYGEN CONCENTRATION, /.

-36 June 1988 Unit 1 Ops/1588

34. SUPPRESSION CHAMBER PRESSURE, PSIG
35. SUPPRESSION POOL WATER TEMPERATURE, DEG F
36. SUPPRESSION POOL WATER LEVEL, FEET
37. SRM COUNT RATE, CPS
38. SRM PERIOD, SEC
39. APRM POWER LEVEL, /
40. CORE XENON CONCENTRATION, / OF FULL POWER EQU
41. RWCU SYSTEM PRESSURE, SPIG
42. RWCU SYSTEM FLOW, LBM/HR
43. RWCU NON-REGEN HEAT EXCHAN OUTLET TEMPERATURE, DEG F
44. RWCU DUMP FLOW, LBM/HR
45. TOTAL MAIN STEAM LINE FLOW, LBM/HR
46. MAIN STEAM TUNNEL TEMPERATURE, DEG F
47. MAIN STEAM LINE RADIATION LEVEL, MR/HR 48: TOTAL MAIN STEAM RELIEF VALVE FLOW, LBM/HR
49. TURBINE SPEED, RPM
50. TURBINE INLET PRESSURE, PSIG 5.1 TURBINE STEAM FLOW, LBM/HR
52. TURBINE BYPASS VALVE STEAM FLOW, LBM/HR
53. TURBINE FIRST STAGE PRESSURE, PSIG
54. TURBINE EXHAUST HOOD TEMPERATURE, DEG F
55. SECOND STAGE REHEATER OUTLET PRESSURE, PSIG
56. SECOND STAGE REHEATER OUTLET TEMPERATURE, DEG F
57. CONDENSER VACUUM, IN HG V
58. CONDENSER HOTWELL LEVEL, INCHES
59. CONDFNSER HOTWELL CONDUCTIVITY, UMHO/CM
60. CONDENSER VACUUM MAKEUP FLOW, LBM/HR
61. CONDENSER HOTWELL REJECT FLOW, LBM/HR
62. CONDEilSATE DEPRESSION, BTU/LBM
63. CIRCULATING WATER INLET TEMPERATURE, DEG F
64. CIRCULATING WATER OUTLET TEMPERATURE, DEG F 65 TOTAL CIRCULATING WATER FLOW, GPM
66. CONDENSATE DEMINERA OUTLET CONDUC, UMHO/CM
67. TOTAL FEEDWATER SYSTEM FLOW, LBM/HR
68. FEEDWATER TEMPERATURE TO REACTOR, DEG F
69. GENERATOR LOAD, MWE
70. GENERATOR REACTIVE LOAD, MVAR

-37 June 1988 Unit 1 Ops/1588

71. GENERATOR STATOR AMPS, AMP
72. GENERATOR TERMINAL VOLTS, VOLT
73. GENERATOR HYDROGEN PRESSURE, PSIG
74. DIESEL GENERATOR 102 LOAD, KNE
75. DIESEL GENERATOR 103 LOAD, KNE
76. OFF-GAS SYSTEM INLET FLON, CFM
77. OFF-GAS SYSTEM OUTLET FLON, CFM
78. OFF-GAS RECOMBINER INLET HYDROGEN CONCENTRATION, /
79. OFF-GAS RECOMBINER OUTLET HYDROGEN CONCENTRATION, /.
80. OFF-GAS SYSTEM RADIATION LEVEL, MR/HR
81. CORE SPRAY LOOP 11 PRESSURE, PSIG
82. CORE SPRAY LOOP 12 PRESSURE, PSIG
83. CORE SPRAY LOOP 11, FLON, LBM/HR
84. CORE SPRAY LOOP 12 FLON, LBM/HR
85. EMERGENCY CONDENSER LOOP 11 FLOH, LBM/HR
86. EMERGENCY CONDENSER LOOP 12 FLOH, LBM/HR
87. EMERGENCY CONDENSER LOOP ll RETURN TEMPERATURE, DEG F
88. EMERGENCY CONDENSER LOOP 12 RETURN TEMPERATURE, DEG F
89. EMERGENCY CONDENSER LOOP 11 VENT RAD LEVEL, MR/HR
90. EMERGENCY CONDENSER LOOP 12 VENT RAD LEVEL, MR/HR

-38 June 1988 Unit 1 Ops/1588

0 ATTACHMENT H SIMULATOR MODIFICATION DATA BASE CHANGES AND TEST RESULTS A. Nl-86-057 RNM Inoperability

1) Test Results - Satisfactory
2) Data Base Changes RNM Drawings: NlY86M057 PRINT SHEET REV INDEX C-22032-C E21.2 C-22032-C E21.2 C-22032-C E21.2 C-22032-C E21.2 C-22032-C EZ1.2 C-22032-C E21.2 C-22032-C E21.2 C-22032-C E21. 2 B. Nl-80-07Z Alternate Rod Insertion
1) Test Results - Satisfactory
2) Data Base Changes ARI Prints: Nl-80-072 DHG SHEET REV INDEX C-34128-C C-34128-C C-34128-C C-18016-C C-22374-C 33 E9 C-22374-C 7 E9 C-22374-C 24 E9 C-22374-C 37 E9 C-22374-C 13 E9

-39 June 1988 Unit 1 Ops/1588

ATTACHMENT H (Cont'd)

SIMULATOR MODIFICATION DATA BASE CHANGES AND TEST RESULTS C. Nl-85-098 Containment Isolation on High Radiation

1) Test Results Satisfactory
2) Data Base -Changes Simulator Modification Data Base Changes and Test Results

(.2) DNG SHEET REV INDEX C-22025-C 4 7 E21 C-22379-C 3 23 E21 C-22383-C 2 17 E21 C-22383-C 6 16 E21 C-22383-C 7 13 E21 C-22385-C 1 30 E21 C-22385-C 2 21 E21 C-22385-C 3 21 E21 C-22385-C 3A 1 E21 C-22385-C 8 26 E21 C-22385-C 10 22 E21

-40 June 1988 Unit 1 Ops/1588

ANSI 3.5 REPORT 1988 ATTACHMENT I MALFUNCTIONS TESTED IN 1988 AD03 EG11 NM13 RR26 AN01 FP01 NM21 RR30 CT01 FP05 NM34 RR34 CU02 FP09 NM39 RR38 CU06 fW03 OG03 RR42 CU10 FW07 PC03 RR46 CH03 FW11 RD03 RR50 CW07 FW15 RD07 RR54 DG01 FW19 RD35 RR58 EC03 FW23 RD39 RR62 EC07 FW27 RM02 RR66 E004 HV02 RP03 RR70 ED08 MC02 RP07 RX01 ED12 MC06 RR03 TC01 ED16 MS04 RR07 TC05 ED20 MS08 RRll TC09 ED24 NM01 RR15 TC13 EG03 NM05 RR19 TU04 EG07 NM09 RR23

-41 June 1988 Unit 1 Ops/1588

0'

ATTACHMENT J LATE MODIFICATION STATUS

~

Nl-81-029 Emergency Venti 1 ation Ready for Testing

~~

Nl-81-038 Drywell Cooling Awaiting Data and Evaluation Nl-83-061 Control Room HVAC Upgrade Ready for Testing Nl-84-013 Replace IA 222/223 Ready for Testing Nl-85-016 Scriba Substation Phase II Awaiting Data & Evaluation

, Nl-85-017 Scriba Substation Phase III In Progress Nl-85-022 Electric and Diesel Fire Pumps Ready for Testing

, Reflects modification status as of July 1, 1988

-42 June 1988 Unit 1 Ops/1588

Oy-

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