ML20148Q290

Simulator Quadrennial Rept for 1993-1997

ML20148Q290
Person / Time
Site:	Millstone
Issue date:	06/24/1997
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20148Q280	List: ML20148Q290
References
NUDOCS 9707070045
Download: ML20148Q290 (145)

Text

. - - . . _ - - . . . - . _ - - - . . - . _ - - _ - . - -

4 4

l aucte g.

+

N *o e e

/

• g u

/

!:

• j rn 69

!  % 2 i o O e __l ll= _= A s  !

4

((OAi =l_gj ~

_ y l 9 j CC lN 0?

J i

i l

i i

MILLSTONE UNIT 2 SIMULATOR

QUADRENNIAL REPORT 4

1993 - 1997 4

W 9707070045 970630 PDR ADOCK 05000336

! R PDR

f

$ MILLSTONE UNIT 2 SIMULATOR QUADRENNIAL REPORT JUNE 1997 o

5 APPROVED:

MP'2 O ation. Manager OS b '

/ Date APPROVED:

l l'O Ma b~PorU r Training 4b+l97

/Cate/ j

! APPROVED:

4 i A LA./ /Lv n3s kb W ea,, ,

l Man.7,.3.e.p_Pn,.,.an.<_,a,.,.

i i

i i

1 w

h l

CERTSUB WHS

?

I

i TABLE OF CONTENTS  ;

.- [

Page No. }

Quadrennial Report 1  !

'1. Testing Goals, Methodology and Assumptions 7 [

r  ?

f 2. Normal Operation and Surveillance Testing 9 l

! 3. Malfunction Testing . 10 j

i

); 4. Yearly Operability Testing 12 ,

5. Physical Fidelity 13 1

4 .

d I

i 6. Initial Conditions Testing 14 i

7.. Simulator Operating Limits Testing 15

8. Instructor Station Testing 16 I i

9. Real Time Testing 17 s '

10. Ensuring Continued Performance of the MP2 Simulator 18

11. Open Deficiency Report (DR) List , 19

12. Next Four-Year Schedule 20

13. Administratalve Modifications to the Simulator Certification Quadrennial Reporting Program 21 i

\

O i

)

TABLE OF CONTENTS (continued)

Attachment 1 MP2 Normal Operations and Surveillance Testing Sequence

}

Attachment 2 MP2 Surveillances that can be performed on the Simulator Attachment 3 MP2 Malfunction Test Abstracts  !

i Attachment 4 initial Conditions Checklist and List of 25 Certified initial Conditions l f

Attachment 5 Student Feedback Survey Results  ;

i Attachment 6 Open Deficiency Report (DR) List  !

i Attachment 7 Schedule for Next Four Years of Testing I Attachment 8 Annual Operability Transient Testing Abstracts O Attachmente e h veicai Fid e'itv S # m m arv R e p ort Attachment 10 Sample Malfunction Test Procedures Attachment 11 List of Certified Remote Functions Attachment 12 Commonly Used Abbreviations and Definitions Attachment 13 Experience of Testing Personnel O

ii

QUADRENNIAL REPORT

SUMMARY

O The Millstone 2 simulator was initially certified on June 30,1989. Certification was accomplished through the Northeast Utilities Simulator Certification Program, which is also the vehicle for ensuring continued certification. Based on the results of the Performance Test, which was run over the past four years, the Millstone 2 simulator continues to demonstrate excellent physical and functional fidelity when compared to the reference unit. Our Simulator Certification Program contains a comprehensive testing program, as well as procedural controls to ensure that the Millstone 2 simulator retains its high degree of fidelity.

This submittal contains the following thirteen sections and thirteen attachments:

Section 1 provides testing, goals, methodology and assumptions.

Sections 2 through 10 review and summarize the individual tests, which make up the Millstone 2 Simulator Performance Test.

Section 10 reviews and summarizes the procedural controls for maintaining certification of the Millstone 2 Simulator.

Section 11 provides a list of open deficiencies on the Millstone 2 simulator. As of May 30,1997 there were 58 open deficiencies on the Millstone 2 simulator (12 PDCR, 13 dynamic,3 logic and 35 enhancements). A total of 255 deficienc:es were dispositioned over the past four years.

Section 12 discusses the +esting sequence for the next four (4) year certification period (July 1997 through June 2001).

Section 13 discusses changes which have been incorporated in the Millstone 2 simulator during the past four years.

Attachments 1 through 13 provide supporting documentation for sections 1 through

12. The attachments are referenced in the appropriate sections of this summary.

Attachment 12 provides a list of abbreviations and definitions. Attachment 13 details the experience levels of test personnel.

O 1

] The Performance Tests described in sections 2 through 9 were performed by NRC licensed SROs, l including two former Millstone 2 Shift Supervisors. As shown in the malfunction abstracts, l )

! computer code analytical predictions were used to the maximum extent possible for complex j

transients such as Loss of Coolant Accidents and Main Steam Line Breaks. Deficiencies identified l d

during the Performance Tests are identified in the attached Performance Test abstracts. i NU, in taking no exceptions to ANSI /ANS 3.5,1985 and Regulatory Guide 1.149,1987, takes the following positions:

i ,

I f (1) Modes of operation specifically prohibited by the Millstone 2 Plant design or by Technial Specifications need not be certified. An example of this is "startup, shutdown, and power j l operations with less than full reactor coolant flow" (ANS-3.5, Section 3.1.1(7)). This mode

, of operation is specifically prohibited by the Millstone 2 Technical Specifications.

J In those situations where the standard conflicts with the plant desigt. or operating

philosophy, the Millstone 2 plant design and/or Technical Specifications take precedence.

4 NU has taken this position to ensure that the conduct of simulator operations is in l accordance with approved plant procedures and the Technical Specifications.

! (2) Acceptance criteria and satisfactory performance are predicated upon the ability of the

! operator to discern differences between simulator response and reference performance data and the effects these differences may have on subsequent actions ano diagnostic

]

9 abilities.

j (3) With respect to the " Performance Testing Plan Change" section on Form 474, it should be recognized that NU's Performance Test Plan is an integral part of the Nuclear Simulator Engineering Manual. Changes to that manual will not be forwarded to the NRC unless they significantly alter the intent of the test program.

(4) Analog and digital process computer points not required for the conduct of training or examinations need not be tested.

(5) Testing of surveillances on redundant equipment or flowpaths is not required if the primary i piece of equipment or flowpath is tested.

l (6) In regard to the steady state testing portion of the yearly operability test, steam generator ,

temperature was not evaluated. This is acceptable because Millstone 2 control boards do not include steam generator temperature indication.

O

I

[

(7) When applying the Steady State Operation Performance Criteria of ANS 3.5, Section 4.1 l both instrument error and loop error are considered when evaluating the accuracy of a [

. particular reading.

l (8) ANS 3.5,1985 Appendix B, B2.2.4 requires Relief Valve Flow to be recorded if available.

Relief Valve Flow is not available and, therefore, is not recorded.

f (9) Testing of all input / Output (1/0) override capabilities is not required during testing of the I

! Instructor Station. Testing a sample of l/O overrides is sufficient to demonstrate the i simulator's capabilities. Specific I/O override points are to be tested, as required, during l curriculum testing.

l l l L  !

t l'

, i l

l t

i l

O l l

1 O  ;

SIMULATOR CERTIFICATION PROGRAM OVERVIEW The continuining mission of our certification program is to:

Ensure that the simulator has the capability to support the training program.

Ensure certification is maintained in a cost-effective manner and addressing the specific requirements NRC 10CFR55.45 (b), recommendations of Regulatory Guide 1.149 and the guidance of NUREG 1258.

Ensure compliance with the provisions set forth in ANSI /ANS 3.5,1985.

The effort required to accomplish this mission has been grouped into three main components:

Definition of the Scope of Simulation, Validation of the Scope of Simulation, and Configuration Management. NU has put in place a collection of formal procedures called the Nuclear Simulator Engineering Manual, to direct all aspects of the certification process and ensure compliance with the regulatory requirements. The NSEM is a departmentally controlled document to ensure consistent application.

The Scope of Simulation that NU is certifying is based upon the NU Simulator Instructor Guides, which encompass:

n U =

all events specified in ANSI /ANS-3.5,1985 and Regulatory Guide 1.149,1987 and 10 CFR55.59 (c)(3)(i),

the training requirements (LOIT/LORT) as specified in the various plant start-up, operating and emmgency procedures, and outside events (e.g., selected LERs, plant design changes, etc.) that affect the training programs and/or the trainer configuration.

Validation of the defined Scope of Simulation consists of two main groupings of activities: (1)

Performance Testing and (2) Verification. A specific performance test was developed for the Millstone 2 simulator, which fulfills the testing provisions of ANS 3.5,1985; INPO 86-026, Guidelines for Simulator Trairing; and NUREG 1258. Included are the following test categories: l

. Malfunctions (Major and Minor)

. Normal Operations

. Instructor Interface

. Operability

. Real-Time Simulation There are also activities, which are requirements of certification, but do not fit neatly within the .

context of performance testing, namely:

. Defined Simulator Operating Limits '

Plant-Referenced Physical Fidelity

. Approved initial Conditions i The Millstone 2 Simulator Performance Test is a dynamic document and is the primary  !

j mechanism for validating simulator performance and fidelity. As such, it is updated to reflect l modifications made to the simulator and/or new reference plant performance data. The entire performance test is repeated over 4 years. Malfunction testing is accomplished at a rate of l

_ approximately 25% per year. Enth Operability Testing and the Physical Fidelity report are done j annually. Section 1 through 11 proiide a detailed summary of the testing and results for the j period from July 1993 through June 1997.

NU's Certification Program provides control over the configuration of the Millstone 2 Simulator to

{ ensure that it can effectively support the training mission and that regulatory commitments are satisfied. 1 i

The main components of Configuration Management are: Design Data Base, Documentation,

, Modification Control and Scope of Simulation Expansion. l l O The intent of the Simulator Design Data Base is to have available the complete data on which the i

~

simulator is designed, and on which upgrading is based. The specific data which forms the design  !

basis for the current Millstone 2 simulator hardware configuration and software models has been j identified and validated it is NU's philosophy here not to create a separate, new data base, but to l utilize existing controlled reference plant design data bases. As such, NU utilizes the latest j revision of plant documents and relies on the formal plant design change process for notification of i

modifications and transmittal of pertinent information. Open Simulator Design Changes constitute the Update Design Data Base described in ANS 3.5,1985. In addition, there is also simuator-specific documentation which is needed for certification and/or maintenance of the simulator.

While this documentation is cortrolled and updated, it is not considered to be part of the Simulator Design Data Base.

NU has in place a modification control process that manages the implementation of design changes on the Millstone 2 simulator and complies with NRC recommendations and industry ,

standards regarding configuration control. Procedures within the NSEM control the coordination, resolution and documentation of identified differences between the simulator and reference plant.

A Simulator Configuration Control Committee (SCCC) has been established to be responsible for overall simulator design control and prioritizes deficiency reports (DRs). The SCCC is comprised of 1) Operations Manager, Millstone 2 (Chairman),2) Manager, Process Computers and O Simulators,3) Supervisor, Simulator Computer Engineering,4) Supervisor or Assistant Supervisor,  !

l 5-

MP2 Operator Training 5) MP2 Simulator Operations Assistant,6) MP2 Software Engineer, and l

7) MP2 Hardware Technician. A data base is used to track the status of identified simulator  !

discrepancies. l The need for expanding the Scope of Simulation is determined by monitoring outside events which ,

have the capability for affecting the training programs or simulator configuration. These include:

l l

[

. New reference plant performance data  !

. Student feedback *

= Curriculum testing i

. New simulator Instructor guides  !

- Major plant design changes  !

. Licensee Event Reports (LERs)

. Significant Operating Event Reports (SOERs)  !

. Plant Design Change (DCRs, Minor Mods, IEEs, etc) l I

I l

i l

O i 1

0 l

l

1. Testing Goals. Methodology and Assumptions -

, D The NU Simulator Certification Program, goals, methodologies and assumptions were l

k established to ensure an efficient, effective and comprehensive approach to testing.  !

Certain elements of this testing philosophy are: l J

. i Testing should be conducted during normal, abnormal and emergency conditions.  !

i The Simulator response, as verified by testing, during normal, abnormal and 4

emergency conditions shall meet the following criteria necessary to support the j c 4ntents of the training curriculum:

l l

f Correct operator diagnosis is possible. )

a -

Capabilities for operator intervention to mitigate events exist. ,

5 l i

Actions or inaction taken by operators results in similar response as in the  :

reference plant.  !

1 Alarms and automatic system nctuations shall occur such that operator diagnosis and response is not adversely affeted.

. Deficiencies found during testing which violate these criteria shall be documented by generating a Deficiency Report (DR), to be dispositioned in accordance with NSEM 5.01, Simulator Modification Cont ol.

. The requirements of ANS 3.5 shall be implemented.

. Simulator controls (used for training) such as switches, annunciators, meters, controllers, recorders, lights, keylocks, pushbuttons, etc., should be tested.  ;

• Process computer points needed to support training should be tested.

. A combination of operating experience, engineering judgment and analytical results should be used to test the simulator response to Major Malfunctions such as Large Break LOCA, Excess Steam Demand, etc.

During the development and conduct of specific testing it became necessary to establish additional guidance. This was done to more effectively apply the provisions of ANS 3.5 and respond to the unique attributes of each test. This additional guidance is summarized:

o

NORMAL OPERATIONS TRAINING I

(7 .

ANS 3.5, Section 3.1.1, item 7 (operation with less than full flow) need not be tested. l This evolution is prohibited by the Millstone 2 Technical Specifications.

Testing of surveillances on redundant equipment or flowpaths is not required if the primary piece of equipment or flowpath is tested. For example, if the Facility I Service Water Pump surveillance is performed, the Facility ll Service Water Pump surveillance need not be performed.

YEARLY OPERABILITY TESTING l

STEADY STATE TESTING

+ N. utoring Steam Generator Temperature as specified by ANS 3.5 Appendix B, B2.1, shall not be done. Millstone 2 has no control board indication for Steam Generator Temperature.

Calculation of simulator steady state tolerance allowances. For example:

RCS pressure RCS pressure, range 1500 to 2500 psia Loop range, 2500 -1500 = 1000 psia Allowed tolerance = 1%

Loop accuracy = 0.5%

At all power levels, RCS pressure reads 2250 psia Therefore, the tolerance applied to the simulator is:

1000 psia x 1.5% = 15 psia RCS temperature RCS Ts range 515 F to 615 F Loop range 615 - 515 = 100 F Allowed tolerance = 1%

Loop accuracy = 0.5%

Therefore, the tolerance applied to the simulator is:

< 100 F x 1.5% = 1.5 F O

TRANSIENT TESTING 4

During testing of the Maximum Rate Power Ramp (reference ANS 3.5, Appendix B, B2.2, item 7), the restoration to 100% power cannot be done without operator action and wi!! not be tested.

. All parameters required by ANS 3.5 Appendix B, B2.2.4 are to be tested at 0.25 second intervals except as noted below:

Relief Valve Flow is not to be recorded. Recording capabilities do not exist for this parameter.

This is acceptable because Appendix B, B2.2.4 requires recording these parameters if available.

In the case where the comparison between simulator response and reference plant response results in a discrepancy, that discrepancy is resolved via the Deficiency Report process and an appropriate retest conducted.

OTHER TESTING l

Testing of all input / Output (1/0) override capabilities is not required during testing of the Instructor Station. Testing a sample of 1/O overrides is sufficient to demonstrate the simulator's capabilities. Specific I/O override points are to be tested, as required, during curriculum testing.

2, Normal Operations T. ant l

J ANSI /ANS 3.5 (1985) Section 3.1.1 requires the simulator to be capable of performing normal plant evolutions and surveillances.

The normal operations and surveillances required by ANS 3.5 Section 3.1.1(1), (2), (3), (4),

(5), (6), (8) and (10) were performed using "then current" copies of Millstone 2 Operating procedures and Surveillances. As discussed, the operating condition specified in Section 7 is prohibited by Millstone 2 technical specifications and was therefore not tested. ANS 3.5 Section 3.1.1 (9) was tested by the Reactor Core System Test.

NSEM Procedure 4.10, " Normal Operations Verification" contains the generic guidance used to write the Millstone 2 Simulator Normal Operations and Surveillance Test.

i O

.g.

Attachment 1 contains the Millstone 2 Simulator test procedure used for Normal Operations and Surveillance testing. Using "then current" copies of Millstone 2 Operating Procedures  !

. the following sequence of operations was tested on the Millstone 2 Simulator:

The Simulator was initialized to Cold Shutdown conditions.  !

. A Plant Heatup was performed. f

. A Nuclear Startup was performed.

l

. A Plant Startup was performed. ,

. A Load increase to 100% power was performed.

. A Reactor Trip was initiated.  ;

. A Reactor Trip recovery was performed.  ;

. A Nuclear Startup was performed. j

. A Plant Startup was performed. j

- A Load increase was performed.  !

. The Simulator was reinitialized to 100% power.

. A Plant Shutdown was performed.  ;

. A Reactor Shutdown was performed. l

. A Plant Cooldown was performed until Cold Shutdown was reached.

The specific Millstone 2 Operating Procedure and Surveillance Procedure titles and numbers used in this test are listed in the individual steps of the test procedure shown in Attachment 1. Attachment 2 contains a concise list of 64 Surveillances that the Millstone 2

{

simulator is capable of performing. As shown on this list, some surveillances were exempted from testing if they were simply a repeat of the same type of test on a different Electrical Facility. '

I All Normal Operations and Surveillance Testing will be re-performed over a four year interval as described in Section 12 of this document. I

3. Malfunction Tests ANSI /ANS 3.5 (1985) Section 3.1.2 requires 25 types of malfunctions to be available on a Simulator. The Millstone 2 Simulator is capable of these 25 malfunctions that are applicable to PWRs.

Attachment 3 contains an index of all certified malfunctions available on the Millstone 2 Simulator. All malfunctions listed are certified malfunctions except those few that are annotated with an asterisk and associated note. This index is organized alpha-numerically by plant system. Attachment 12 contains a list of definitions for plant system abbreviations.

O i

i j Each certified malfunction has its own test procedure. Guidance for writing malfunction test

procedures and conducting tests is contained in

l NSEM Procedure 4.04, Major Malfunction Testing NSEM Procedure 4.05, Malfunction Testing 1

Malfunctions which cause major integrated plant effects, such as Large RCS Breaks, Main

} Steam Line Breaks, etc., have their respective malfunction test procedures written and tests conducted per the guidance in NSEM 4.04. For these " major" malfunctions, computer code analytical data or actual data (if available) is typically used to verify correct malfunction response. Analytical data was obtained from the following documents / sources:

CEN 128, " Response of a CE NSSS to Major Transients" CEN 268, " Justification of Trip 2/ Leave 2 RCP Strategy During Transients"

. Millstone 2 Final Safety Analysis Report (FSAR)

. Northeast Utilities generated "Best Estimate" cases, specific to Millstone 2 An example of a malfunction test written and conducted via the NSEM 4.04 process is contained in Attachment 10. The example in Attachment 10 is labeled "RC04: Unisolable RCS Leak on the Reactor Vessel Head Vent."

I Malfunctions which do not cause large integrated plant effects have their respective malfunction test procedures written and tests conducted per the guidance in NSEM 4.05.

This type of malfunction is typically an instrument malfunction, a controller malfunction, a pump trip, etc. Malfunction tests in this category are typically "Best Estimate" Analysis.  !

• Best Estimate" Analysis means a Millstone 2 Subject Matter Expert utilizes his experience, operating procedures, piping and instrument drawings, electrical drawings and possibly hand calculations to estimate proper Simulator response. An example of a malfunction test written and conducted via the NSEM 4.05 process is contained in Attachment 10. The example in Attachment 10 is labeled "CC11: RBCCW Pump Degraded Performance." i I

In summary, reb of the certified malfunctions has its own malfunction test procedure and  ;

its own "Cause and Effect" description. Due to their large volume, only 2 malfunction test l procedures are being provided in Attachment 10, as examples. These malfunction test procedures are available for review upon request.

I ANS 3.5 Section 3.1.2 requires 25 specific malfunctions to be available on a Simulator. In order to facilitate NRC review of this submittal, Attachment 3 contains a cross reference of C

these 25 specific malfunctions to the applicable Millstone 2 Simulator malfunctions. Listed in Attachment 3 under each of the 25 ANS 3.5 required malfunctions are:

Abstracts from Millstone 2 Simulator malfunction tests providing:

The name of the malfunction The date the test was performed Whether the malfunction is variable, and if so, its range and what severity was tested Starting conditions and end point conditions The source of baseline or reference data Details of identified deficiencies At the end of each of the 25 sections is a list of other Millstone 2 Simulator malfunctions which meet the ANS 3.5 requirement, but for which no abstract is provided.

Attachment 3 contains approximately 55 malfunction test abstracts of the 235 certified Millstone 2 Simulator malfunctions. These 55 malfunction abstracts were chosen to cover the major malfunctions of interest (Loss of Coolant, Steam Line Break, etc.) and to cover at least one malfunction in each of the 25 required ANS 3.5 malfunction types. The malfunction Cause and Effect" descriptions may be used as abstracts for other certified Millstone 2 Simulator malfunctions.

ANSI /ANS 3.5 (1985) Section 3.4.2 requires that provisions be available for incorporating additional malfunctions. The Millstone 2 Simulator has this capability.

All certified malfunctions will be retested over a four year interval, as described in Section 12 of this document.

4. Yearly Onorability Testing ANSI /ANS 3.5 (1985) Section 5.4.2 and Appendix B specify Annual Operability Testing requirements. The methodology used to write and conduct Yearly Operability Tests is described in NSEM Procedure 4.09, " Simulator Operability Testing". Using the guidance provided in NSEM-4.09, a Yearly Operability Test specific to the Millstone 2 Simulator was written. This Millstone 2 specific test procedure is not contained in this submittal, but is available for review on request.

Yearly Operability Testing will be performed on an annual basis.

O The Yearly Operability Testing performed consisted of the following items:

i i

l Steady State Testing at approximately 30% power,50% power and 96% power f Transient Performance Testing for ten (10) transients O Reference Plant data obtained at 30%,30% and 96% power during the August 1995 refuel startup was used as the basis for Steady Stata Testing. Utilizing the Reference Plant data, comparisons were made between the Simulator and Reference Plant for approximately 80 selected critical and nors:ritical points.

These 80 points include all those listed in ANS 3.5 Section B2.1 with one exception:

Millstone 2 has no control board Steam Generator Temperature indication.

g 2'

Acceptance criteria for the Steady Qate were based on ANS 3.5 Section 4.1. No deficiencies were identified.

Transient Performance Testing was performed fer ten transients. The ten transients tested were those described in ANS 3.5 Section B2.2, with one exception. This exception concerns ANS 3.5 Section B2.2 Transient #7, which is a " Max rate power ramp (100% down to 75%, backup to 100%)". The Millstone 2 Simulator test was performed by doing a maximum rate power ramp from 100% power down to 75% power only, and not back up to 100% power. A rapid rate power increase from 75% to 100% power would violate the physical design and trip criteria for Millstone 2. Millstone 2 does not have Automatic Rod Control, nor any automatic load reject logic, but does have a Variable High Power Trip. The Variable High Power Trip will actuate an automatic Reactor Protection System Trip on an approximate 9% power it' crease over the current setpoint. A rapid power increase would trip the plant by design, and therefore was not included in the Transient Performance Testing.

The ten transients described in ANS 3.5 Section B2.2 were analyzed using the parameters  ;

indicated in ANS 3.5 Sections 82.2.1,2,3, or 4, as appropriate. Attachment 8 contains l abstracts of each of the ten transient tests.

5. Physical Fidelity Comparisons of the MP2 simulator to the referenco plant are conducted annually in the i areas of panel simulation, instrument and cont;oi configuration and ambient operating environment. Identified discrepancies : .re e',aluated to determine the consequences to the simulator's ability to be used as an ef %tive training tool. This proceu is described in NSEM Procedure 4.12, " Simulator Phytical Fidelity / Human Factors Evaluation". A complete set of photographs was taken of the Reference Plant Control Room in November 1995 and compared to the Simulator. No deficiencies that affect training were identified. Those differences between the Simulator and Reference Plant Control Room which have been L dispositioned as "not affecting training" are described in Attachment 9, " Physical Fidelity Summary Report" To ensure continued physical fidelity, photographs will be taken as needed for Reference Plant Control Boards that have undergone changes.

NU has a strong commitment to maintain the Millstone 2 Simulator up to date with the Reference Plant Control Boards in a timely manner. NSEM Procedure 6.04," Major Plant Changes", addresses controls on major design changes (such as Control Room Desigt:

Review) that challenge a " plant referenced simulator" to remain an effective training tool.

Minor plant changes are addressed within the time constraints of ANS 3.5 Sections 5.2 and 5.3.

6. jnitial Conditions Testing Initial Conditions Testing was performed in February,1989. NSEM Procedure 4.02, " Initial Conditions", describes this process. The Millstone 2 Simulator has capabilities for storing 100 Initial Conditions.

All Certified initial Conditions (ICs) were reviewed to ensure equipment alignments, plant conditions, remote functions, etc., were reasonable for the stated IC conditions. The first four pages of Attachment 4 contain the checklist used for reviewing all certified ICs. The number of certified ICs may vary between 25 to 30 depending on simulator training requirements. Twenty-five ICs have been designated as the " base" group of ICs that will be maintained certified. These 25 certified ICs cover a broad range of conditions such as:

- Beginning of Core Life (BOL)

Middle of Core Life (MOL)

End of Core Life (EOL)

. Different Operating Modes such as Cold Shutdown, Hot Standby, Critical Approach,

etc.

Different Power Levels An IC used for training (or exam)is:

. A certified IC

. Developed from a certified IC using approved plant procedures.

Only certified ICs are used for training or exams. Listed in Attachment 4 for each of the 30 ICs are the following:

RCS TAVE ( F)

. Pressurizer Pressure (psig)

O .

Reactor Power (%)

U . RCS Boron Concentration (ppm)

. Xenon Reactivity (pcm)

Remarks section which describes the basic conditions of the IC O

V Certified ICs are maintained up-to-date as plant changes and procedure changes occur.

7. Simulator Operating Limits Testing )

l 1

The process used for identification and action concerning Simulator Operating Limits is  ;

described in NSEM Procedure 4.08," Simulator Operating Limits". .

Two methods are used to prevent negative training when Simulator Operating Limits are ,

reached: freezing the simulator or administrative controls. The Reference Plant design limits and/or Simulator model limits which csuse the Simulator to " freeze" are listed below.

Also listed are the administrative Simulator Operating Limits which are controlled by the l simulator instructor. These administrative limits are implemented through simulator  ;

instructor training and cautions placed in those Simulator Instructor Guides where such  !

situations could occur. I 4

- The Millstone 2 Simulator will go to " freeze"if any of the following conditions exist:

I RCS Pressure > 2750 psia  ;

Containment Pressure > 60 psig  !

- S/G Pressure > 1'00 psia i Fuel Temp > 5000 F Fuel Clad Temp > 3300 F ]

1 l

- The simulator instructor can determine which of these operating limits caused the j simulator to go to Freeze by reviewing a CRT display in the instruction station.

The following simulator operating limits are :lealt with administratively. These limits are all results of Simulator modellimitations:

. RCS Pressure should not exceed 2000 psi greater than Steam Generator Pressure (Steam Generator Tube /Tubesheet D P).

. Axial Xenon Oscillations are always divergent on the Simulator. This is not a problem in normal training sessions due to their short duration.

O l

The Chemical and Volume Control System (CVCS) if allowed to reach saturation, will not show any flashing or flow oscillation since it is modeled as a single-phase system.

The reactivity worth of stuck Control Element Assemblies (CEAs) is too small.

i The instructors are also informed of these operating limits and other abnormal responses by l Form 7.3 of NSEM-4.08 that follows.  !

The following simulator responses have been identified and not yet corrected: t High Pressure, high temperature water through the PORVs, i.e., once through cooling with the PZR full. The RCS pressure and RV level become abnormal with large

{

pressure and RV level swings.  ;

Reactor Vessel Level below the Hot Leg and RCS at saturated conditions, Incore Tc  ;

indicated.

End of cycle Large Main Steam Line break and a stuck out CEA the simulator might not show a short return to criticality.

8. Instructor Station Testing O During June 1994, Simulator instructor Station Testing was performed as described in NSEM procedure 4.11 " Instructor Station". No deficiencies were identified.

Instructor Station testing verified correct operation of the following features of the Millstone 2 Instructor Station.

. Backtrack Fast time, for each of the eight (8) modeled Fastime parameters

. Slow time

. Boolean Trigger

. Composite Malfunction

. Variab e Parameter Control

. Freeze

. SnFpshot To verify the I/O override feature of the Millstone 2 Simulator, a small number of the following points were tested to verify proper operation.

. Analog Outputs

. Analog Inputs

. Digital inputs

. Digital Outputs

[]' . "Crywolf" Annunciator feature

. Annunciator Override The purpose of the I/O override feature testing was to verify the feature itself, not every 1/O override point. The Millstone 2 simulator has the ability to 1/O override essentially every point in the simulator. Curriculum testing of a simulator lesson plan will require the testing of  :

any individual I/O override point to be used in training or exams, thereby verifying the individual I/O override points to be used. ,

Refer to NSEM Procedure 4.11 for the instructor Station Test Procedure. The data taken l from this test is not contained in this submittal, but is available upon request. l The Instructor Station test will be repeated once every four years. l l

i

9. Realtime Testing  !

l Real Time Testing was performed in June,1997, per NSEM procedure 4.13 "Real Time Simulator Verification".

l

The purpose of this test is to verify that all simulation models are running in real time and is accomplished by '

i i

. Installing a software frame counter in each of the 20 frames. l

. Running each of the following scenarios:

Turbine trip Steam-Line break at 100% severity RCS hot-leg double-ended LOCA RCP locked rotor

. Recording frame counters, simulator exercise time and the worst case execution time for each frame at the end of each scenario.

. Ensuring the frame counter values and simulator exercise time are consistent with the stop watch time (differing no more than 2 seconds for a 10-minute test run).

Ensurin6 the worst case execution time for each frame is less than 50 ms.

O 1

Th3 r:sults of th:st ts show that ths Millstonn 2 simulator p:rforms in real time and no deficiencies were idenNd.

U This test will be repeated once every four years or at any time a question exists that the Millstone 2 Simulator is not running in real time.

l

10. Ennuring continuing Performance of the MP2 Simulator To ensure that the MP2 Simulator performance remains in compliance with ANSl/ANS 3.5 (1985), Reg Guide 1.149 and 10 CFR 55.45 the following procedural controls have been I implemented:

hilajor Plant Modifications - The Millstone 2 Simulator was certified as a Plant Referenced Simulator. Significant Reference Plant Control Room changes, such as from Control Room Design Review modifications, must receive special consideration due to their potential major I impact. NSEM Procedure 6.04, " Major Plant Modifications", addresses this concern. This I procedure ensures that major plant modifications affecting the Reference Plant Control Room are reviewed and acted on in a timely manner. Inis ensures that training and exams continue to be performed on a valid plant referenced Simulator.

Elant Design Changes / Procedure Changes 1

O l V Plant Design Changes and Procedure Changes are sent to the Training Department to be reviewed for training impact and Simulator impact. This asswes that both training and the simulator are continually evaluated and updated as plant changes occur. Procedural controls covering this review process are in Training Procedures not provided in this submittal.

Plant Design Changes requiring Simulator modifications are handled within the time allowed by ANS 3.5 Section 5.2 and 5.3. A recently instituted self-assessment effort to review reference plant design changes which occurred after simulator delivery identified one minor modification which had previously not been identified. This modification involved replacement of two 3-position control board handswitches with 2-position handswitches and has subsequently been completed (outside of the time allotted by ANS 3.5). There was no impact on simulator operations or training.

Student Feedback - Student (licensee) feedback is an important input to Simulator Fidelity.

NSEM Procedure 6.01, " Student Feedback", describes how student feedback is requested.

Regular written feedback is requested from students on simulator training and fidelity.

Response has been frequent and favorable. Also, every one to two years a student survey on Simulator Fidelity is performed. Attachment 5 contains the results of the survey t

Q,)

performed in October,1996. The summary letter also contains the disposition of action to ,

be taken for each item. -

/*

Reference Plant Performance Data - As plant events occur, data will be retrieved and ,

evaluated to validate Simulator Fidelity. NSEM Procedure 6.03, Collection of Plant  !

Performance Data", covers the collection of reference plant performance data.

Develooment of New Simulator Training Guides - Simulator Certification Procedure NSEM 6.02, " Development of New Simulator Guides," covers requirements for new Simulator training guides. This ensures that new Simulator training guides use only certified remote functions, certified malfunctions, certified Initial Conditions and do not exceed any Simulatot ,

Operating Limits.

Simulator Certification Documentation - As the Millstone 2 Simulator is modified, appropriate simulator certification documentation needs to be updated. NSEM Procedure 5.02, " Retest Guidelines" covers updating of the Performance Test.

Reference Plant Design Changes may result in simulator changes such as:

. Adding or deleting remote functions t

. Adding or deleting malfunctions

. Changing remote functions or malfunctions '

. Changing Performance Tests or their criteria It is Northeast Utilities' interpretation that simulator documentation may be modified as the Reference Plant changes without requiring the submittal of an NRC Form 474 update.

Changes to simulator certification documentation will be made per the attached NSEM  !

procedures. Updated materials will be sent at the next regular certification quadrennial j report date, or upon NRC request. I i

11. Qpen Deficiency Report (DR) List Simulator Modification Control Procedure NSEM-5.01 controls for the coordination,  !

resolution, and documentation of identified differences between the simulator and its  !

reference plant, and to maintain the integrity of the simulator hardware, software and design  !

databases.  :

A Deficiency Report is a form used by the Operator Training Branch and the Simulator Comuputer Engineering Group to record all identified deficiencies between the simulator l and reference plant.

A Simulator Design Change is a documentation package, consisting of relevant Deficiency Reports and other forms, which is designed to track the resolution of Deficiency Reports and ensure that ANSI /ANS 3.5 is satisfied.

s l

- . - - - - . . . ~- .. - - - - . - - . . . .. . .-. . .-

I l

l Attachment 6 contains a current listing of all Millstone 2 simulator open Deficiency Reports

. m (DRs) associated with certification. These DRs will be dispositioned' in accordance with

! their importance to training.

12. Next 4-Year Schedule. (July 1997 to June 2001)

The entire MP2 performance test will be repeated over a four-year interval as described in Attachment 7. The schedule shown in Attacnment 7 has been written based on the guidance provided in NSEM Frocedure 4.07, " Master Test Schedule".

The following tests must be performed every year.

l I Annual Operability Testing

. Physical Fidelity Verification  ;

The following tests must be performed over a 4-year interval:

j Normal Plant Evolutions and Surveillance Testing All Certified Malfunctions Instructor Station Testing

. Real Time Testing

, l i

l i l i

"Dispositioned" means 1 of 2 things will occur. Either: 1) the DR will be fixed, or 2) If the DR l cannot be fixed, the problem may be added to the Simulator Operating Limits if it is significant. It is l i O important to recognize that prioritization of resolving DRs is a dynamic process. As new DRs are generated, their importance will be evaluated and the order of DR resolution appropriately changed, if necessary, to ensure the highest quality training is presented.

1 i

t

l

\

13. Administrative Modifications to the Simulator Certification Quadrennial Reporting l Program l t

The MP2 Simulator Performance Testing program has undergone administrative modifications since the last quadrennial report submitted in June 1993. The foliowing is a synopsis of those changes:

System Tests:

The MP2 Simulator has undergone over ten years of Performance Testir :.i support of

1) acceptance from t!'ie vendor; 2) acceptance of modifications; and 3) cer'.,iication and quadrennial reporting to the NRC. Performance Tests have been developed, selected, and used in accordance with the particular purpose to be achieved. One form of Performance Test was a detailed exercise of logic and flow paths associated with the simulated systems identified in the MP2 simulator design specification, e.g., Circulating Water System, Service Water System, Reactor Protection System, etc. This form of testing is called System Tests.

System Tests are resource intensive because of the simulator and personnel time required to achieve the level of detail necessary to test every element of the simulation. The value of these tests decreases as the individual elements of the simulation are confirmed to exist and perform correctly. The number of discrepancies identified reaches a point of diminishing marginal return on the effort expended. More importantly, the integrated response of the i V simulation elements becomes the key in determining simulator performance.

Within a direction to optimize the mission of the Simulator Certification Program, Simulator Computer Engineering (SCE) determined that System Tests were no longer justifiable, being neither timely, cost-effective, enhancing compliance, nor adding value to the training program. Thereby, the Systems Tests were discontinued as part of the Performance Tests, and are now used only when appropriate, for example, when new systems are installed.

Supoorting Facts:

The acceptance tests from the vendor achieved the initial acceptance of contract commitments through a thorough review of performance at an extreme level of detail. Both factory acceptance tests (FAT) and site acceptance tests (SAT) resulted in the identification of a significant number of discrepancies. The FAT was completed in March 1985. The SAT was completed in June 1985. System Tests were very valuable during this phase of the Simulator life cycle. They were used to achieve a firm baseline of performance for the individual elements of the simulation. This baseline remains firm through rigorous conformance to a simulator change control process that was instituted from the original FAT and remains active to this day.

O When the NRC required certification of plant referenced simulators the MP2 simulator was again tested in 1989 using the full Acceptance Test Procedure (ATP) that was used in the  :

FAT and SAT Performance Tests. Discrepancies other than those originally identified in FAT and SAT were found, but the number was much lower than the original ATPs.

. i The first quadrennial cycle (1989 - 1993) after certification also employed System Tests.

Very few discrepancies were found during conduct of the System Tests.

1 After the submittal of the 1993 four year report, it was determined that the performance of System Tests was no longer of value, that the other forms of Performance Testing were sufficient to achieve the mission, that System Tests were not a regula'ory requirement or industry standard, and that System Tests could always be conducted should the need arise.

l Therefore, System Tests have been discontinued as a necessary part of Performance Testing, i

_Comouter Platform Re-Host i

in October 1993, a computer platform re-host was completed which replaced the Gould 32/87 processors and peripherals with a SUN SPARCeenter 2000 system and new i peripherals. The re-host effort included complete benchmarking against the previous l platform by using the Simulator Operability test. These tests included 1) instructor station I test,2) all annual operability tests and 3) all major / minor malfunctions required by NSEM 4.07 for 1993-1994. Any differences in the two benchmarks were resolved before the simulator with the new platform was retumed to the Operator Training program.

MP2 Simulator Computer System Configuration (A) Hardware SPARCeenter 2000 40-MHz XDBus, Fast SCSI-2/ Buffered Ethemet Sbus Card (FSBE/S)

Internal SunCD Drive,14-Gbyte 8mm Tape Internal Drive Two 40MHz System Boards with Two 60MHz SuperSPARC-Il Modules each, No Memory 1

192-Mbyte of ECC Memory SIMMs DSCSI Drive Tray with Four 2.9GB drives fed by 2 Sbus Differential Fast / Wide Intelligent SCSI-2 Host Adapters (DWIS/S)

Three 20-inch Color Monitors, TurboGX Frame Buffers, and Cables A

U u

l

i (B) Software SUN Solaris 2.3 operating system Sybase 4.9.2 relational database management system Dataview 9.5 graphical tool runtime NUSE (Northeast Utilities Simulation Environment)

NUXIS (Northeast Utilities X-Window instructor Station)

NUSE The Northeast Utilities Simulation Environment (NUSE) was developed in-house by the SCE staff. The real time portion of NUSE includes the Real Time Executives (Rtexec), Interactive Debugging Task (IDT), I/O module, etc., and provides the model execution sequencing, scheduling, the panelinterfacing and on-line parameter monitoring The off-line portion of NUSE includes tools and utilities used by engineers to develop, debug and maintain the simulation models.

NUX13 Northust Utilities X-Window Instructor Station (NUXIS) was also developed in-house by the SCE staff using Dataview's graphic tools NUXIS provides a window-based, point and click, graphical user interface for instructors.

NSAs Model set Migration in January 1995, new advanced NSSS models were installed on the simulator and made available to the Operator Training program. This migration was from the original Singer-Link NSSS models ready-for-training in May 1985 to a model set developed by ABB-Combustion Engineering. The acquisition and installation of the ABB NSSS model set included a rigorous series of acceptance tests exceeding the testing of the original models. The acceptance tests included a preliminary Factory Acceptance Test (FAT), a FAT, a Site Acceptance Test, a RELAP Benchmark Test, a Simulator Operability Test benchmark, and Major Malfunction Tests. These tests constituted a System Test for the NSSS. Several reports have been generated and retained for review. All discrepancies were resolved.

The ABB NSSS simulator model set includes the neutronics of the reactor core and the thermohydraulics of the RCS, secondary side of the steam generators and the main steam header.

The neutronics modelis a three dimensional model based on a modified one and half group diffusion equation. Each fuel assembly is divided into ten axial nodes for a total of 2170 neutronic nodes. The thermohydraulics is a nonhomogeneous, non-equilibrium liquid and steam model. It models phase separation and occurrence of non-equilibrium conditions. The model solves mass conservation equations of liquid, mixture masses (liquid, liquid / steam,

. . - - . - ._. -- - - ._~ _ _ - - - - -- _ .

i

! steam), energy equations for the mixture and steam, and the conservation equation of the ,

i mixture momentum. The integration is an implicit integration of the coupled system. The momentum equation is solved for each path (junction). The pressure is solved locally based on masses and energies for each node. Phase separation is calculated with the use of drift '

flux model. ,

Physical Fidelity Verificction i The June 1993 to June 1997 four year test plan called for annual full panel photographs of the reference plant to be taken and compared to the simulator. This was not done.

Photographs and a re' port were prepared for the 93/94 period, pictures were taken for the l l 95/96 period, and a report was prepared for the 96/97 period. Since physical fidelity l discrepancies are identified and dispositioned within the normal simulator modification process, no physical fidelity discrepancies resulted from the full panel photograph comparisons. A record of full panel photographs is maintained and updated on an "as needed" basis for documentation of the reference plant comparison. A record of identified discrepancies is maintained with training value assessments of the impact on actions to be taken by the operator. These records are retained and updated as needed within the normal simulator modification process. NSEM procedere 4.12, ' Simulator Physical Fidelity / Human Factors Evaluation", has been updated to address these changes.

i l

l I

l I

5 O

i i

ATTACHMENT 1 10  !

MP2 NORMAL OPERATIONS AND SURVEILLANCE TESTING SEQUENCE  !

i k

r s

I i

l

> j l

l 1 I

l

'l I

l l

l O 1 l

I f

l

!O

. _ . -- . . _ _ - .-_ -- ~ . .

( \ Figure 7.1 Q)

NORMAL PLANT EVOLUTIONS LIST i

, 1. PLANT STARTUP '

o Will be tested by this procedure: YES[ NO []

l o If no, state reason:

o Operating Procedure (s) to be used: o P 22o3 o Unit specific Test Step, numbers which will complete this requirement: gPg

2. NUCLEAR STARTUP o Will be tested by this procedure: YES[ NO []

o If no, state reason:

j o Operating Procedure (s) to be used: oe22o2 o Unit specific Test Step, numbers which -sen z. e, will complete this requirement:

A

3. TURBINE STARTUP AND GENERATOR SYNCHRONIZATION o Will be tested by this procedure: YES p[NO[]

o If no, state reason:

o Operating Procedure (s) to be used: OP 22ch oP 23 23 A, WM of 232%

o Unit Specific Test Step numbers which will complete this requirement

4. REACTOR TRIP AND RECOVERY o Will be tested by this procedure: YES[ 'O [ ]

i o If no, state reason:

o Operating Procedure (s) to be used: 6oPZS2E p d GoP252G o Unit Specific Test Step numbers which will complete this requirement: TTcT6.O t

I Rev.: 1

' Date: 1/31/97 Page: 7.1-1 of 3

I l

Figure 7.1 O

V NORMAL PLANT EVOLUTIONS LIST

5. HOT STANDBY OPERATION o Will be tested by this procedure: YES [(NO[]

l o If no, state reason:

o Operating Procedure (s) to be used: c d P Z M , 2 7 0 ( 7 '2 c l o Unit Specific Test Step numbers which Is T will complete this requirement: eE l 6. LOAD CHANGES o Will be tested by this procedure: YES [hO[]

o If no, state reason:

o Operating Procedure (s) to be used: OP22o9 o Unit Specific Test Step numbers which will complete this requirement: P 9'O

7. PLANT OPERATIONS WITH LESS THAN FULL REACTOR COOLANT FLOW i o Will be tested by this procedure:

YES [] NO (( l o If no, state reason: .

o pe-rects wU \e%

o Operating Procedure (s) to be used:

g pg is l

o Unit Specific Te,st Step, numbers which will complete this requirement:

go7 go#5 (oy 1 l

8. PLANT SHUTDOWN () L 92. 'TbcM TMS  !

l o Will be tested by this procedure: YES ((NO []

o If no, state reason:

OP 2'loy aP2zos o Operating Procedure (s) to be used: I o Unit Specific Te,st Step, numbers which will complete this requirement: g4 G, o l

l Rev.: 1 v' Date: 1/31/97 Page: 7.1-2 of 3 i

l Figure 7.1 NORMAL PLANT EVOLUTIONS LIST

9. CORE PERFORMANCE TESTING o Will be tested by this procedure: YES [] NO[yf o if no, state reason: ccrcc TeSTMt. tc c o Operating Procedure (s) to be used: ^#" M 'N AT o Unit Specific Test Step numbers which mfchcx .

will complete this requirement:

10. SURVEILLANCE TESTING o Will be tested by this procedure:

YES 'M" NO [ ]

o If no, state reason:

A g ge g m (c- Z G C C W " 5 i o Operating Procedure (s) to be used:

5Ps wee Aced' i o Unit Specific Test Step numbers which d o <t % w s pro &

will complete this requirement:

m g c, ,;9 p a p  ;

G2Got A (be ON SUBMITTED BY: L M P SO ptmet . // TfwM TNM**#

APPROVED BY: " [M/C / ' '

ASOT / 4 g g fe,

&ccu-4.puSW 2" wsmCwy 2anA o^

$6Tc6PCWV I sp2w\ A Deli H 2- ME]

~

i Rev.: 1 Date: 1/31/97 Page: 7.1-3 of 3

i l L

N Figure 7.2 .

i NORMAL OPERATIONS TEST COVER SHEET i UNIT 2_ ATTACHME T NUMBER 8.2 Released for Performance By:

OT JATE 3

Performed By: f[2f[f/

DATE Verified By: ,

DATE Accepted By: '

^ ~

5 ff p ASOT 'DATE j 1. List of Operating Procedure Form!.; Attached ,

hh hSOl ff n

2. Comments Attached Yes [ ] No [ d I

i l

i

{

i i

b*q.

• Attachment 8.2 TEP PROCEDURE /RESULTS ACCEPT /DR

.0 Plant Heatup 1.1 Initialize to the following Simulator conditions:

o RCS Temperature at 100 F o "A" SDC Pump in service maintaining RCS temperatures o Pressurizer at 95% level, vents closed o PORV's in " low" o S/G levels at 60%

o Containment purge via purge supply fan and main exhaust o Atmospheric dumps in manual and 5% open o Pressurizer spray valve in manual closed o RCP bleedoff at 40 psig to EDST 1.2 Use OP2201, " Plant Heatup" ead Op A perform a normal plant heatup.

for;. 2tvi-1 Record the "

N following information at the specified times  !

dur'ng the heatup:

l 1.2.1 Record the revision level and number of changes against the OP2201 that is used:

Revision level M N Number of changes t,d ,D 1.2.2 Review each Prerequisite, Initial Condition and Precautions listed in OP 2201 to ensure that the simulator i is capable of supporting the use of  %

OP2201. //

1.2.3 During the heatup, comp *te (initial) or N/A steps in OP220 and attach a completed OP220lrT'to this proce' ure. /

t$ (A +

Note: Whenever valve alignments are specified to accomplish the heatup, the valve alignment shall be reviewed for any control board i

valves or remote functions that need to be manipulated to place them in the proper condition. Valve alignment sheets need not be attached or filled out.

(h.

L]

Rev.: 0 Date: 5/3/88 Page: 8.2-1 of 11 NSEM-4.10

J l

STEP PROCEDURE /RESULTS ACCEPT /DR I

x_) 1.2.4 Verify that it takes 350 gallons />

to overF ow the Pressurizer to the i

Quench Tank after the pressurizer

, indicates 100% level Record observed i number of gallons ~ 380 .

1.2.5 Verify that RCS heatup rates are as /$)

follows and record the requested data.

1.2.5.1 Two RCP's runnin , heatup rate /b> '

shall be 20-25*F r o Decay Heat /  % power o Observed heatup rate 2 6 F/hr o Observed heatup rate was at an average RCS temp of /dho *F 1.2.5.2 Three RCP's running, heatup rate /b shall be 30-35 F/hr o Observed hCatup ratej2l2s F/hr o Observed heatup rate was at an average RCS temp of A10 F 1.2.5.3 Four RCP's running, heatup rate y/I)

( 's

~

shall be 35-40 F/hr '

\ +

~' o Observed heatup rate 9/ / *F/hr o Observed heatup rate was at an average RCS temp of 9/00 F  !

Note: Fast time maybe used during heatup as long as it does not interfere with heatup rate determinations and use of fast time will not interfere with surveillance or other heatup related items.

1.2.6 Pressurizer heatup rate with all heaters j#52 on approximately 60 F/hr. Observed pressurizer heatup rate is 2L9 'F at an average pressurizer water temperature of 1

%@ F.

1.2.7 RCP "Bleedoff flow lo" alarms clear at /

)

RCS pressure of 1700 psia.

1.2.8 Atmospheric steam dumps maintain an /62 RCS temp of ~532*F. Record atmospheric steam dump pressure that maintains 532 F while in auto. 970 psia.

r~N  !

Rev.: 0 Date: 5/3/88 Page: 8.2-2 of 11 NSEM-4.10

i STEP PROCEDURE /RESULTS ACCEPT /DR j ~.

1.2.9 Perform the follow. ag surveillance procedures prior to starting the heatup and attach completed surveillance forms to this procedure:

o Service Water Valves o RBCCW Valves SP2612E # f; SP2611E A o SRAS Manual SP2616A W o CSAS Manual SP2606F p>P5 o SIAS Manual SP2604R N 1.2.10 Perform the following surveillance procedures prior to 200'F RCS '

O  !

temperature and attach completed -

l surveillance forms to their procedures.

o Transient Temp. Verif. SP2602B o RBCCW Pump Ops - FACI SP2611A //rv o RBCCW Valves - FACI SP2611C /v o SW Pump Ops - FACI SP2612A yh, _

o SW Valves - FACI SP2612C n _

o EBFAS and CR vent FACI SP2609A-1 6 o SIT outlet valve (SIAS) SP2603A /W7 o SIT outlet valve (Pressure) SP2603B r o Charg. Pmp. Op. SP2601G @

~

o Boric Acid Op. SP2601C 72

) o ECCS Valves o Borated Water Flow Ops SP2604N SP2601A

/>

o Ctnt. Isol. Valve Ops SP2605G Wo o D/G Op FAC I SP2613A /'b>

o Control Rm Wkly Checks SP2619C /r/

o HPSI Pump Op FACI SP2604A WV" o HPSI Valves FACI SP2604E /p' o Ctnt. Sump Isol. Vlv-FACI SP2604G /7,4Q o Control Room Daily Checks SP2619A /F 1.2.11 Perform the following surveillances #7 prior to reaching 300 F RCS temp and '

attach the completed Surveillance forms:

o Aux Feed Pump Ops o Aux Feed, Turbine Driven SP2610A h Ops SP2610B b o Aux Feed Valves SP2610C M o Ctnt Air Recirc Ops-FACI SP2607A rh o PORV Block Valve Ops SP2610F ytn i O Rev.: 0 Date: 5/3/88 Page: 8.2-3 of 11 NSEM-4.10

1 STEP PROCEDURE /RESULTS ACCEPT /DR N

1.2.12 Perform the following surveillance prior to reaching 1750 psia RCS

[

pressure and attach the completed surveillance forms:

o LPSI Pump Ops - FACI SP2604C /F o LPSI Valve Ops - FACI SP2604L //#  !

o Cont. Spray Pmp Ops-FACI SP2606A F o Cont. Spray Valves - FACI SP2606C F 1.2.13 Perform the following surveillance prior to exceeding 5% power and attach the M  !

completed surveillance forms:

o Hydrogen Purge SP2608E h '

o PIR Fan FACI SP2608C 'fr>

o Turbine Drives Aux Feedpump Ops SP2610B 1.3 Plant heatup is complete when RCS temperature is 532"F, RCS pressure is being maintained M

at 2260 psia and all 1.2.10, 1.2.11 surveillancescf sta k g m. 1 O l

-1.2.3, 7 1 ? 19 ?nd 1 A43 arecomplete.%d F7~ '  ;

(except MSIV surveillances) 1.4 Ensure all completed surveillance forms e and a completed ops 2201 i fe m are attached.

1.5 Take a snapshot if required Aendos9)/I.

,4 A/4 1

1 2.0 NUCLEAR STARTUP 2.1 Initial Conditions, Plant stable with RCS Tave at 532 F, 2260 psia, pressurizer '

N level at 40%, S/G 1evels at 65%, an ECP has been completed, boron in the RCS is at the ECP required boron concentration and CEDM cooling is in operation.

Note: It is the intent of Section 2.0 " Nuclear Startup to be performed from an initial condition that is a continuation of Section 1.0 Plant Heatup or from a snapshot taken from Step 1.5.

Rev.: 0 Date: 5/3/88 Page: 8.2-4 of 11 NSEM-4.10

STEP PROCEDURE /RESULTS ACCEPT /DR Q(~'s 2.2 Start the "A" CEDM MG Set ,M 2.3 Close all TCB's, Observe load of ~250 amps M on C-04 Record Observed value % fc amps.

2.4 Start and parallel the "B" MG set, observe load to be 125 amps each, Record actual values "A" MG set /Af~ amps and "B" MG set / d amps.

2.5 Using OP2202, " Reactor Startup",

withdraw CEA's to bring the reactor critical. Record the following information at the specified times ,

during the reactor startup:

2.5.1 Record the revision level and number of changes against the OP2 02 that is

/D used. Revision Level /

Number of Changes j_

2.5.2 Review each prerequisite, Initial th) ,

Conditions and Precautions listed in  !

OP2202 to ensure that the simulator i is capable of supporting the use of OP2202. 1

\

2.5.3 Ensure that OP2202-1 " Pre-Critical Checklist" is completed and attached M  !

to this test. Each item on OP2202-2 i should be initialized or N/A'd as appropriate.

2.5.4 At 1000 cps, indication shifts from CPS to % power.

[ j 2.5.5 Critical Position in observed to be within .9%

2.5.6 Complete and attach forms OP2619D-1 M and 2619D-2.

~

2.5.7 Perform the following surveillance I during the startup: I o CEA partial movement SP2620A o CEA Group Deviation Verification o CEA CMI Verification SPMr208 54Ws M WR

1. 2Q2^^p Aqua l 2.5.8 Stabilize conditions with the reactor O critical at ~1% power, Tave at ~532' F.

Perform and attach SP2619D-3 2.5.9 Take a snapshot if required ,Al[T A/b O

Rev.: 0 t

Date: 5/3/88 Page: 8.2-5 of 11 NSEM-4.10

STEP PROCEDURE /RESULTS ACCEPT /DR

(^N Q.0 PLANT STARTUP Note: It is the intent of Section 3.0 Plant Startup to be performed from an initial condition that is a continuation of Section 2.0 Nuclear Startup or from snapshot taken step 2.5.9.

3.1 Perform a plant startup to 20% power using OP 2203, " Plant Startup" procedure.

3.1.1 Record the revision level and number of changes against the OP2203 that is used:

Revision level //

Number of Changes 1. .

3.1.2 Review each prerequisite, Initial Condition and Precaution listed in OP2203 to ensure that the simulator is capable of supporting the use of OP2203. j 3.1.3 Ensure the Condensate System is in service on Short Recycle.

'M 3.1.4 Start one Circ Water Pump in each &

condenser.

(O 3.1.5 Start the Hydrogen Seal Oil System.

3.1.6 Place the Turbine on Turning Gear.  ;

3.1.7 Start the Stator Cooling System.

3.1.8 Start the EHC System. /

3.1.9 Heatup the Main Steam System and b open the MSIV's.

Perform SP2610E and 2610D attach the forms. I 3.1.10 Place Turbine Gland Seal System in service.

3.1.11 Increase reactor power as required to hold RCS Tave at 532" F.

3.1.12 Commence drawing a vacuum and observe at 15" hg Steam Dump / Bypass System N

may be put in service.

3.1.13 Observe Turbine Bypass PIC controls RCS T, at 532 F.

'd,A 3.1.14 Observe at condenser Vacuum of 20" Hg, a SGFP may be started.

A Rev.: 0 Date: 5/3/88 Page: 8.2-6 of 11 NSEM-4.10 ,

i

! STEP PROCEDURE /RESULTS ACCEPT /DR i n

\s_) 3.1.15 Increase Reactor Power to ' 7% Tave to db

, 53 5'F.

3.1.16 Start the Main Turbine per OP2323A /2 3.1.17 Synchronize and place the main generator on line.

/I)  !

3.1.18 Increase load to 20% power and at /b>

I 15% power, observe the LPD and Turbine  !

RPS trips are enabled. i 3.2 Final conditions are Main Turbine On Line, /Y i Reactor Power at ' 20%, NSST supplying In-House .

2 loads 3.3 Take a snapshot if required , d

4.0 LOAD CHANGE TO 100% POWER i Note

It is the intent of Section 4.0 " Load Changes to 4

100% power, to be performed from an initial

. condition that is a continuation of Section 3.0 .

Plant Startup or from a snapshot taken in Step 3.3. l l 4.1 Perform a power increase from 20% to 100%  !

power using OP2204, " Load Change" procedure. i 4.1.1 Record the revision level and number / ?>

of changes against the OP2204 that is used. Revision level /Y Number of I Changes O 4

4.1.2 Review each prerequisite, Initial Condition and Precaution listed in OP2204 to ensure the simulator is capable of supporting the use of OP2204.

4.1.3 Verify that Pressurizer Level follows /}

Ops form 2204-3 during the power increase. "

Note: Power increase rates used do not need to be restricted, as long as the power increase rate does not cause excessive pressurizer pressure ,

increases, pressurizer level problems, S/G Level Control Problems or T.. Control Problems.

4.1.4 Perform SP2619E and attach surveillance /hb7

~

forms.

Rev.: 0

^ Date: 5/3/88 Page: 8.2-7 of 11 NSEM-4.10 _

i

STEP PROCEDURE /RESULTS ACCEPT /DR o

, k) g, 4.2 Final conditions shall be T... '572" F, RCS pressure at '2260 psia and Generator

/ ,

Load at ~885 MWe.

5.0 Reactor Trip and Recovery ,

5.1 Initialize to 100% full power IC, normal - /#$)

steady state conditions, perform SP2601D, ,

SP2602A, and attach surveillance forms.

5.2 Record the revision level and number of changes against EOP 2525 and EOP 2526.

EOP 2525 Rev / 8- # of Changes c) A#PCP .

EOP 2526 Rev v # of Changes f 2rs 5.3 Manually trip the reactor from C04 TCB P.B. ,

Carryout, EOP 2525, Standard Post Trip '

Actions,  !

5.4 Verify the Startup Rate is -1/3 DPM ~ one minute after the trip - n p/Aq N i 5.5 Complete EOP 2526, Reactor Trip Recovery, complete and attach OPS Form EOP 2526-1.

() 5.6 Perform a reactor startup and plant startup to 20% full power using OP 2202, OP 2203, and OP 2204 as guides.

1 5.7 Final conditions are: the plant at ' 20%  !

full power with the Feed Reg. Valves in Auto. I 6.0 PLANT SHUTDOWN 6.1 Initialize to a 100% full power IC, normal 4k /

cruady state conditions. l 6.2 Using OP 2204, decrease load to 20% power. /)

6.3 Using OP 2205, perform a plant shutdown. 42) 6.3.1 Record the revision level and number d

/7 j of changes against the OP2205 that is '

used: j Revision Level /d Number of Changes F f i

? Review each prerequisite, initial /kh condition and precaution listed in OP2205 to ensure the simulator is capable l of supporting the use of OP2205.  ;

)

Rev.: 0 Date: 5/3/88 Page: 8.2-8 of 11 NSEM-4.10

STEP PROCEDURE /RESULTS ACCEPT /DR d

6.3.3 Durin the plant shutdown, perform //$p. figs surve llance SP2613B on the "B" D/G, '

SP2608A H2 Recombiner and attach the completed surveillances.

6.3.4 Verify that Heater Drain, _ Condensate and Feedwater Flow decrease as power

[

is decreased.

6.3.5 Verify that the RSST assumes the in-house loads when transferred from NSST.

6.3.6 Verify the Heater Drains Tank level is ~66% with no HD pumps and the High Level Dump in operation.

6.3.7 When the Main Turbine is tripped, verify the CV's, SV's, IV's and ISVS close.

6.3.8 Verify the Main Turbine rolls down to turning gear in ~20 minutes. '

M 6.3.9 Verify the Turbine Bypass / Steam Dump D Valves control RCS T.,. to ~532* F and S/G pressure to 880 psia.

6.4 Take a snapshot if required .

7.0 REACTOR SHUTDOWN 7.1 Obtain Initial Conditions to a Reactor Shutdown using OP2206, by continuing from Section 6.0 or initializing to an IC which is critical Tave ~532 F, RCS pressure d'2260 psia, Reactor Power < 5% with the turbine /

generator off line.

7.2 Record the revision level and number of changes against the OP2206 that is used:

Revision level M Number of Changes 8 7.3 Review each Prerequisite, Initial Condition and Precaution listed in OP2206 to ensure the simulator is capable of supporting the use of OP2206, 7.4 Use OP2206 to shut the reactor down verifying the following:

h 7.4.1 CEA's insert properly in Manual e sequential N

(

Rev.: 0 Date: 5/3/88 Page: P.2-9 of 11 NSEM-4.10

STEP PROCEDURE /RESULTS ACCEPT /DR

( i 7.4.2 Reactor Power level decreases at no greater than - 1/3 DPM

/h 7.4.3 Inward CEA motion for an individual CEA stops at the lower electrical limit.

7.4.4 CEA bottom lights go on after TCB's are opened. ,

7.5 Final Conditions shall be; all CEA's fully

/ b inserted, all TCB's open, Both MG sets shutdown, RCS boron concentration at Hot Standby concentration. ,

7.6 Take a snapshot if required 36 . //)

8.0 PLANT COOLDOWN 8.1 Reset to an Initial condition with the plant at ~532 RCS T,,,, RCS Pressure at

~2260 psia. All CEA's inserted, the Steam Dump Bypass System in service, one condensate  !

pump in service, one SGFP in service and RCS j boron concentration sufficient for meeting  :

shutdown margin. Alternatively using the IC l f) at the conclusion of section 7.0 is acceptable. l 8.2 Commence a cooldown using OP2207 " Plant '

Cooldown" verify or record the following during the Plant Cooldown:

8.2.1 Record the revision level and number of changes against the OP2207 that is used:

Revision Level / f Number of Changes D 8.2.2 Review each Prerequisite, Initial Conditions and Precaution listed in OP2207 to ensure the simulator is capable of supporting the use of OP2207.

8.2.3 Perform SP2614D for AEAS during the cooldown and attach the completed surveillance form.

8.2.4 Verify Pressurizer Pressure and Temperature decrease with all heaters '

h off.

\._)

Rev.: 0 Date: 5/3/88 Page: 8.2-10 of 11 NSEM-4.10

__. . . . . . _ _ . . _ - . _ . . . _ _ _ . . . . . . _ _ . ~ . . _ _ _ _ . -

. . . . _ _ . .- . . . _ - ._m

I STEP PROCEDURE /RESULTS ACCEPT /DR

\

~~

8.2.5 Verify Pressurizer Pressure and Temperature decrease when Aux Spray

/f?

is initiated.

8.2.6 Verify that MSI Block is permitted at Ab?>

< 600 psia S/G Pressure (3 out of 4).

8.2.7 Veriff750 that SIAS Block is permitted psia RCS pressure (3 out of 4) . /Y at <

8.2.8 Verify that SDC warmup on 1 LPSI pump / dC) l is 12" F/Hr. ' l

8.3 Final conditions are

RCS temperature sf 27

<200 F on SDC, i.e. in Mode 5. Pressurizer '

vents opened and Pressurizer corrected level in at ~50%, Pressurizer temperatures are

<200* F, excess JsedcW is in service at

~140gpmandS/Glevels{areat~80%.

-)

1 f

1 O

1 l

l 1

i 1

i i

1

, ~~

V l

Rev.: 0 Date: 5/3/88 Page: 8.2-11 of 11 NSEM-4.10

~_

l i

i i

ATTACHMENT 2 i

1 MP2 SURVEILLANCES THAT CAN BE PERFORMED ON THE SIMULATOR i

i .

i l l

i i

)

i  ;

i 4

i 4

d I

I i

T

! 1 kt  !

4 4

5 1

t i

4 i

I i  ;

i l e

I 3

I I

i J

t j

i e

1 i

t i

i l l

1 a

i 1

j a

a j Drafunsem\att2,ws0 i

Figure 7.3 Page1of1 i

l SURVEILLANCE LISTING I UNIT 2 (Yes/No)  !

l Seauential Number Iitic Procedure # To Be Tested )

( l. Borated Water Flow /Pamp Op SP2601A Yes  !

2.~ Boric Acid Ops SP2601C Yes

3. Power Range Calibration SP2601D Yes 'l

4. Charging Pump Ops Fac 1 SP260lG Yes  !

5. Charging Pump Ops Fac 2 SP260lH No .

6. Reactor Coolant Leakage SP2602A Yes .

7. Transient Temp. Verification SP2602B Yes l 8, SIT Valve Ops (SIAS) SP2603A- Yes I

9. SIT Valve Ops (Press) SP2603B Yes i

10. - HPSI Pump Ops Fac 1 SP2603B Yes i

11. HPSI Pump Ops Fac 2 SP2604B -No t

12. LPSI Pump Ops Fac 1 SP2604C . Yes i

13. LPSI Pump Ops Fac 2 SP2604D No '

14. HPSI Valves Ops Fac 1 SP2604E Yes  !

15. HPSI Valves Ops Fac 2 SP2604F No  !

16. Ctmt Sump Valve Fac 1 SP2604G Yes l

17. Ctmt Sump Valve Fac 2 SP2604H No I

18. LPSI Valve Ops Fac 1 SP2604L Yes  !

19. LPSI Valve Ops Fac 2 SP2604M No l

20. ECCS Valve Ops SP2604N Yes  ;

21. SIAS ManualTest SP2604R . Yes

22. Ctmt Valve Stroke Time SP2605G Yes  ;

23. Ctmt Purge Valve Power SP2605G Yes  !

' 24. Reactor Head & Pn Vent Sol v1 Oper. SP2605N Yes  :

25. Ctmt Spray Pump Fac 1 SP2606A Yes j

26. Ctmt Spray Pump Fac 2 SP2606B No  ;

27. Ctmt Spray Valve Ops Fac 1 SP2606C Yes .  !

28. Ctmt Spray Valve Ops Fac 2 SP2606D. No-  !

29. CSAS Manual Test SP2606F Yes 'I

30. CAR Fac 1 SP2607A Yes

31. CAR Fac 2 SP2607B- No

32. H Recombiner 2 SP268C Yes I

33. PIR Fan Fac 1 SP2608C Yes

34. PIR Fan Fac 2 SP2608D No

35. H Purge 2 Ops SP2608E Yes

36. EBFAS Fac 1 SP2609A Yes

37. EBFAS Fac 1 SP2609B- No

38. . Motor Driven Aux Feed Pumps SP2610A Yes Approved:

[- ,

7J ASOT O- -

Rev: 0 Date: 6/22/88 Page: 7.31 of 2 NSEM-4.10 l

. - - --. - - --~ -. ,

Figure 7.3 Page 2 of1 i

SURVEILLANCE LISTING - i UNIT 2 l

(Yes/No) .  !

Seouential Number litic Procedure # To Be Tested  !

i

39. Turb Driven Aux Feed Pump SP2610B Yes

40. : Aux Feed Valves SP2610C Yes l

41. MSIV Partial Closure SP2610D Yes  !

.42. MSIV Closure SP2610E Yes

43. PORV Block Valve Ops SP2610F Yes >

44. RBCCW Pp Fac 1 .

SP2611A Yes )

45. . RBCCW Pp Fac 2 SP2611B No j l

46. RBCCW Fac 1 Valves SP2611C Yes i

47. RBCCW Fac 2 Valves SP2611D No

48. RBCCW Valve Ops SP2611D Yes

49. SW Pp Fac 1 SP2612A Yes

50. SW Pp Fac 2 SP2612B No

51. Service Water Valves SP2612C Yes  ;

52. Service Water Valves SP2612D No

53. Service Water Valves - SP2612E Yes 54 D/G "A" Ops SP2613A Yes

55. D/G "B" Ops SP2613B Yes

. 56. AEAS Ops - SP2614D Yes l 57. CSAS Manual Test SP2616A - Yes

( 58. Control Room Shift Checks SP2619A Yes l 59. Control Room Weekly Checks SP2619C Yes

!- 60. Start UP Surveillance Checks SP2619D Yes

61. Control Room Monthly Checks SP2619E Yes

62. CEA Partial Movement SP2620A Yes

63. CEA Group Deviation Verification SP2411 Yes ,

64. CMI Verification SP2411A Yes l l

i l'

I l

Approved: \

i l

i

~O Rev

0

[ Date: 6/22/88  !

l ..

Page: 7.3-2 of 2 NSEM-4.10 l

1

.- .. . - . . - . . , - . - . . . - . - _ _ . . . . . - - . . . ~ - - . - . - . . . . . . . - _ _ . . . .

1

, l 3

l ATTACHMENT 3

! )

i MP2 MALFUNCTION TEST ABSTRACTS j i

3  !

i' i

i I e

1 i

j 5

6 i

t l

1 i

l O i I

l i

1 I

i l

I i

O M2MilTST.Wils .

l

. _ . .4

i i l

l

! 1. Loss of Coolant Malfunction Abstracts SG02 Malfunction - Steam Generator Tube Rupture This malfunction test was conducted in May,1997. This malfunction is capable ofinserting up to a 3000 gpm tube rupture to either Steam Generator. This malfunction was tested at 100%,30%, and 5% severity for both Steam Generators. All tests were started from 100%

Power, Middle of Life Core C,onditions, equilibrium Xenon, steady state Conditions for a period of ~ 12 minutes. Baseline data was from NU Retran model computer runs specific to MP2. No deficiencies were identified.

CV01 Malfunction - Unisolable letdown ime rupture in Containment This malfunction test was conducted in April,1995. This malfunction is capable ofinserting i up to a 200 gpm leak from the letdown line (RCS to CVCS) in Containment to the ,

l Containment Atmosphere / Floor. This malfunction was tested at 100%,50% and 10%  !

l severity. All tests were started from 100% Power, Middle of Core Life conditions,

equilibrium Xenon, steady state Conditions for a period of ~ 8 minutes. Baseline data was from CEN 128 Case E2. No deficiencies were idemified.

L r l CV02 Malfunction - Isolable letdown line rupture outside of Containment c This malftmetion test was conducted in April,1995. This malfunction is capable ofinserting up to a 200 gpm leak from the letdown line (RCS to CVCS) outside of Containment, in the Auxiliary Building. This malfunction was tested at 100%,50% and 10% severity. All tests were started from 100% Power, Middle of Life Core conditions, equilibrium Xenon, steady l state conditions for a period of' 8 minutes. Baseline data was from CEN 128 Case E2. No deficiencies were identified.

l l CV18 Malfunction - Isolable letdown line rupture inside Containment This malfunction was conducted in April,1995. This malfunction is capable ofinserting up to a 200 gpm leak from the letdown line (RCS to CVCS) in Containment to the Containment Atmosphere / Floor. This malfunction was tested at 100%,50% and 10% severity. All tests were started from 100% Power, Middle of Core Life conditions, Equilibrium Xenon, steady i

state conditions for a period of' 8 minutes. Baseline data was from CEN 128 Case E2. No j deficiencies were identified.

l l

l O

l 2 l

l

RCO2 Malfunction - RCS Hot Leg Break This malfunction test was conducted in April,1995. This malfunction is capable ofinserting up to a 180,000 lbm/sec leak from either RCS Hot Leg to Containment. 100% severity is equivalent to a guillotine rupture of a Hot Leg pipe. This malfunction was tested for each Hot Leg at 100%,50% and 10% severity. All tests were started from 100% Power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. All tests were for a  ;

duration of 5 minutes except the 100% severity case for #1 Hot Leg, which lasted 60 minutes,  !

to observe Containment Sump Recirculation, Containment Sump Level, RWST level, Containment Area Rad Monitor response and Core Exit Thermocouple (CET) response.

Baseline data was from the MP2 FSAR, sections 14.15 and 14.16. No deficiencies were identified.

RC03 Malfunction - RCS Cold Leg Break This malfunction test was conducted in April,1995. This malfunction is capable ofinserting up to a 100,000 lbm/sec leak from any of the four RCS Cold Legs to Containment.100%

severity is equivalent to a guillotine rupture of a Cold Leg pipe. This malfunction was tested at 100%,50% and 10% severity for all four cold legs. All tests were started from 100%

Power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. All tests were for a duration of 5 minutes. Baseline data was from the MP2 FSAR, sections 14.15 and 14.16. Three deficiencies were identified. They are: (1) Reactor Vessel level dropped to 0%

and stayed there. It should eventually have recovered to 7% level due to safety injection.

O No deficiencies were identified.

RC04 Malftmetion - Unisolable Reactor Head Vent Leak This malfunction test was conducted in May,1997. This realfunction is capable ofinserting up to a 400 gpm unisolable leak through the reactor head vent system. This malfunction was tested at 100%,50% and 10% severity for a duration of 8 minutes each. All tests were started from 100% Power, Middle of Core Life conditions, equilibrium Xenon, steady state  ;

conditions. Baseline was from the CEN 128 Case E2. No deficiencies were identified.

RC05 Malfunction - Pressurizer Safety Valve Failure This malfunction test was conducted in April,1995. This malfunction is capable ofinserting up to a 250,000 lbm/hr flow from either Pressurizer Safety Valve to the Quench Tank and  ;

eventually to the Containment Atmosphere if the Quench Tank Rupture Disc blows. This malfunction was tested at 100%,50% and 10% severity for each of the two pressurizer safeties for a duration of eight minutes each. All tests were started from 100% Power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. Baseline data j was from an NU Retran analytical case for MP2. No deGeiencies were identified.

3 l

RC06 Malftmetion - Pressurizer Relief Valve (PORV) Failure q

C/ This malfunction test was conducted in April,1995. This malfunction is capable ofinserting up to a 130,000 lbm/hr flow from either Pressurizer Relief Valve (PORV) to the Quench Tank and eventually to the Containment Atmosphere if the Quench Tank Rupture Disc blows. This malfunction was tested at 100%,50% and 10% severity for each of the two pressurizer relief valves (PORV's) for a duration of 13 minutes each. All tests were started from 100% Power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. Baseline data was from an NU Retran analytical case for MP2. No deficiencies were identified.

Other malfunctions which may be used to give Loss of Coolant conditions are: CV03, CV13, CV14, CV20, CV21, RC20, and SG01. All of these malfunctions give a maximum RCS leak rate of 100 gpm or less. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

2. Loss ofInstrument Air Malfunction Abstracts IA01 Malftmetion -Instrument Air Header Rupture This malfunction test was conducted in May,1996. This malfunction is capable of producing o a 5,000 Standard Cubic Feed per mi ute (SCFm) leak on the instrument air header. This O malfunction was tested at 10%,50%,85% and 100% severity. The test was started from l

100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions and was ended when a reactor trip occurred and Steam Generator safeties were removing heat.

Baseline data was from best estimate analysis and use of plant procedures and P&ID's. No deficiencies were identified.

IA03 Malfunction - Loss ofInstrument Air in Containment This malfunction test was conducted in August,1990. This malfunction is capable of producing a 300 SCFM leak in the Containment Receiver Tank, thus depressurizing the Containment Air Header. This malfunction was tested at 20% and then ramped up to 100%

severity. The test was started from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions and was ended when the Containment low air receiver alarm was received. Baseline data was from best estimate analysis and use of plant procedures and P&lD's. No deficiencies were identified.

It has been identified that this malftmetion is not required for either LORT or LOIT training and it will no longer be tested until it is required for training.

Other malfunctions which may be used to provide loss ofinstrument air conditions are:

IA01, IA04 and IA05. The Cause and Effects descriptions may be reference to for each of h these malfunctions to describe the malfunction characteristics. No deficiencies were identified in any of these malfunctions.

4

- 3. Loss or Degraded Electrical Power - Malfunction Abstracts ED05 Malfunction - Loss of 4160 Volt Bus This malfunction was tested in September,1990. This malfunction is capable ofinsening a loss of any of the 6 4160 Volt Buses due to ground fault. Each of the 6 bus malfunctions This malfunction was tested in April,1994. This malfunction is capable ofinserting a were individually tested starting from 100% power, Middle of Life Core conditions, equilibrium Xenon, steady state conditions. Each test was ended after verifying the loss of power to the correct components and verifying that malfunction removal (ground removal) allowed breakers to be reclosed. Baseline data was from best estimate analysis, use of plant procedure, electrical drawings and P&ID's. No deficiencies were identified.

ED08 Malfunction - 6.9 KV Bus failure to Auto Transfer This malfunction was tested in April,1994. This malfunction is capable ofinserting a failure for either 6.9 KV bus to auto transfer on a reactor trip. Each of the 2 6.9 KV bus failures to auto transfer were tested. Each malfunction was tested starting from 100% power, Middle of Life Core conditions, equilibrium Xenon, steady state conditions. Each test was ended after verifying correct breaker response, any verifying that manual action to close breakers was still possibl.:. Baseline data was from best estimate analysis, use of plant procedures, p electrical drawings, and P&ID's. No deficiencies were identified.

A i

ED10 Malfunction - Loss of 125 VDC Vital Bus l This malfunction was tested in April,1994. This malfunction is capable ofinserting a loss of either 125 volt DC Vital Bus. Each of the 2 malfunctions were individually testing starting at 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. I Each test ended afler DC loads were verified to be lost. Baseline data was from best estimate analysis, use of plant procedures, electrical drawings and P&lD's. No deficiencies were identified.

ED16 Malfunction - Loss of Vital 120 Volt Instrument AC This malfunction was tested in April,1994. This malfunction is capable ofinserting a loss of any of the 4 Vital 120 Volt AC buses. Each of the 4 malfunctions were individually tested starting from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions. Each test was ended after the correct loads were verified to be lost. Baseline data was from best estimate analysis, use of plant procedures, electrical drawings and P&ID's. No deficiencies were identified. 1 i

O I

(

5

t l EG08 Malfunction - Diesel Generator Output Breaker Failure l

pd This malfunction was tested in April,1994. This malfunction is capable of preventing either i DG output breaker to fail to close. Each of the 2 malfunctions were tested one at a time from 100% Power both with and without the emergency . -s energized, to verify the D/G output breaker would not close. Baseline data was from bes< ? stimate analysis and electrical drawings. No deficiencies identified.

EGil Malfunction - Diesel Generator Auto Start Failure This malfunction was tested in April,1994. This malfunction is capable of preventing either Diesel Generator from auto-starting following an LNP signal. Each of the 2 malfunctions were tested one at a time from 100% power. Each malfunction was also tested to ensure that a manual Diesel start could be done with the malfunction active. Baseline data was from best estimate analysis and electrical drawings. No deficiencies were found. l Other malftmetions which may be used to provide Loss or Degraded Electrical Power are: l l ED01, ED02, ED04, ED06, ED07, ED09, EDI 1, ED13, ED14, EDl5, ED17, EDI 8, EG02, i l EG07 EG09, EG10 and EG12. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were  ;

identified. l l l

4. Loss of Forced RCS Flow- Malfunction Abstracts RC11 Malfunction - RCP Locked Rotor l

This malfunction test was conducted in May,1995. '."his malfunction is capable of msertmg ,

l a Locked Rotor on any of the 4 RCP's. This malfunct on was tested for each of the 4 RCP's  !

l starting from 100% imwer, Middle of Core Life conditions, equilibrium Xenon, steady state conditions for a period of I minute each. Baseline data is from the MP2 FSAR section 14.6.

(One deficiency was identified. RCS temperatures, Tc, T ii, and T vEA did not increase enough  ;

on the Simulator and therefore Pressurizer level and pressure did not increase enough on the l Simulator. An RPS trip is correctly processed in ~ 1 second on the Simulator as it would in the Reference Plant. This is not a training issue since the reactor promptly trips on low RCS Flow, despite the inadequate RCS temperature and pressure response.)

j Other malfunctions which may be used to give Loss of RCS Flow conditions are: CC06, ED04 and RCl3. The Cause and Effect descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics.

Yearly Operability Transients 4 and 5 also tests loss of all RCP's and loss of a single RCP.

Refer to Attachment 10 for these test abstracts.

bJ 6

l l

5. Loss of Condenser Vacuum / Loss of Condenser Level Malfunction Abstrac1s p

d FWOI Malfunction - Loss of Condenser Vacuum This malfunction test was performed in June,1994. This malfunction inserts a loss of Condenser vacuum which at 100% severity will cause a Turbine trip in ~ 5 minutes. This ,

malfunction was tested at 10%,50% and 100% severity The test was started at 100%

Power, equilibrium Xenon, steady state conditions and continued until a Turbine trip occurred. Baseline data was from best estimate analysis and plant procedures. No deficiencies were identified.

6. Loss of Service Water Malfunctions Abstracts SWO5 Malfunction - Unisolable Service Water Header Rupture This malfunction was tested in May,1996. At 100% severity, a 5000 gpm leak can be  !

inserted on either Service Water header. This malfunction was tested at 20% and 100%

severity for both Service Water headers. All tests were started from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions and ended when ser ice water flow was lost to the major Service Water loads. Baseline data was from best estimate i

analysis and P&ID's. No deficiencies were identified.

hn Other malfimetions which may be used to give Loss of Service Water to Service Water headers or individual components are: SW01, SWO2, SWO3, SWO4, SWO6, SWO7, SWO8 and SWO9. Cause and Effects descriptions may be referenced to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified in any of these malfunctions.

7. Loss of Shutdown Cooling - Malfunction Abstracts RH01 Malfunction - Low Pressure Safety Injection (LPSI) Pump Trip This malfunction test was conducted in May,1997. This malfunction was started from RCS water level at the minimum level of Centerline of the RCE IMt Leg, with Shutdown Cooling in operation at Beginning of Life conditions with mr.ximum dtcay heat. The purpose of this test was ta verify proper response to loss of Shutdown Coolinr ; at the most limiting condition:,. The test was run for a period of 15 minutes. Baseline data was best estimate analysis. No deficiencies were identified.

Other malfunctions which may be used to give Loss of Shutdown Cooling related problems are: RH02, RH03, RH04, RH05 and RH07. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunctions characteristics. No hs deficiencies were identified in any of these malfunctions.

7

l i

8. L',ss of Comnonent Cooling - Malfunction Abstracts i

.C02 Malfunction - RBCCW Header Rupture 1

l This malfunction test was conducted in April,1997. This malfunction is capable ofinserting l up to 2000 gpm rupture on either Reactor Building Component Cooling Water (RBCCW) i Header. This malfunctior was tested at 100%,50% and 10% severity for each RBCCW Header. All tests were started from 100% power, Middle of Core Life, equilibrium Xenon, steady state conditions. The test was ended when the running RBCCW pump tripped, was isolated and the header restored with the "B" RBCCW pump (if available). Baseline data was from best estimate analysis and P&ID's. No deficiencies were identified.

l Other malfunctions which may be used to give a Loss of Component Cooling to headers or components are:

RBCCW (Reactor Building Closed Cooling Water): CC01, CC03, CC04, CC05, CC06, CC07, CC08, CC09 and CC10  !

TBCCW (Turbine Building Closed Cooling Water): TP01, TP02, TP03, and TP04 i l Circulating Water to Condenser: CW01, CWO2, CWO3, CWO4, CWO6, CWO7 and CC08.

l The Cause and Effects descriptions may be referred to for each of these malfunctions to l describe the malfunctions characteristics. No deficiencies were identified in any of these malfunctions.

O 6

9. Loss of Normal Feed / Feed System Failure - Malfunction Abstracts FWO6 Malfunction - Main Feed Pump Trip This malfunction test was conducted in lone,1994. This malfunction is capable ofinserting a trip signal on either Main Steam Generator Feed Pump (SGFP). Both main SGFP trips l were tested starting from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions and ending after a reactor trip occurred on Low Steam Generator l Level and verification that the tripped main SGFP could be restarted followirg malfunction removal. Baseline data was from best estimate analysis and P&lD's. No deficiencies were identified.

Other malfunctions which may be used to give Loss of Normal Feed or a normal feed system failure are:

FW23, FWO3, FWO4, FWOS, FWO7, FWO8, FWO9, FW10, FW11, FW13, FW14, FW15, FW16, FW18, FW22, FW28, FW29, MSI1, MS13 and MS16.

The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified in any of these malfunctions.

8

l 1

10. Loss of All Feedwater (Normal and Emergency)- Malfunction Abstracts This event was tested in the Yearly Operability Testing, Transient #2. Refer to Yearly Operability Testing Abstract for Transient #2 in Attachment 10.

I1. Loss of Protective System Channel Malfunctions listed here are those that give complete or partial channel failures of the

, Reactor Protection System (RPS) or Engineered Safeguards Actuation System (ESAS).

Nuclear Instrumentation Failures are listed separately later.

ED16 Malfunction - Loss of Vital 120 Volt Instrument AC Buses This malfunction was tested in April,1994. This malfunction is capable ofinserting a loss of I any of the 4 Vital 120 Volt AC buses. Loss of a single vital 120 V AC bus causes the loss of  !

an RPS channel and loss of an ESAS channel and depending on the malfunction, a loss of an l l

ESAS actuation cabinet. Each ofilie 4 malfunctions were individually tested starting from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions.  !

Each test was ended after the correct loads were verified to be lost. Baseline data was from i best estimate analysis, use of plant procedures, electrical drawings and P&lD's. No deficiencies were identified.

l ESO4 Malfunction - Loss of Vital 120 Volt Instrument AC to ESAS  ;

I '

l This malfunction was tested in May,1997. This malfunction is capable of tripping the input breakers to each of the 2 ESAS actuation cabinets. Each of the malfunctions were individue!!y tested starting from 100% power, Middle of Core Life conditions, equilibrium l

Xenon, steady state conditions. Each test was ended after the fuses were replaced to show they could be restored by procedure. Baseline data was from best estimate analysis, use of plant procedures, electrical drawings and P&ID's. No deficiencies were identified.

RP12 Malfunction - RPS Safety Channel Cold Leg Temperature Transmitter Failure This malfunction was tested in May,1997. This malfunction is capable of failing any of the 8 RCS Cold Leg Temperature transmitters from 465 to 615 F, corresponding to O to 100%

l severity. All 8 malfunctions were tested at 10% severity and then ramped up to 100%

severity. Testing was started at 100% power, equilibrium Xenon, steady state conditions.

Each test was ended when proper response of the RPS trips which use the malfunctioning Cold Leg Temperature was verified. Baseline data is from best estimate analysis and P&lD's. No deficiencies were identified.

Other malfunctions which may be used to give a loss of Protection System Channel are:

RCl4, RP01, RP05 RP06, RP07. RP08, RP09, RP10, RP11, RP12, RP13, RPl4, RP22, RP24 and RP25. The Cause and Effects descriptions may be referred to for each of these O meifeections te deecribe the maifunctien charecterietice. xe dericienciee were identiited.

9 l

l

l l 12. Control Rod Failure - Malfunctions Abstracts ,

RD01 Malfunction - Dropped Control Rod (CEA)

This malfunction test was performed in May,1997. This malfunction is capable of dropping t any of the 61 Control Rods (CEA's). When a CEA is dropped,it falls from its current position to the bottom of travel. There is no variable severity on this malfunction. Other j malfunctions may be used for positioning CEA's at other than full insertion. This  !

malfunction was tested in several stages. First, starting from 100% Power, Beginning of Life conditions, equilibrium Xenon, steady state, CEA's,14,15,16 and 17 were dropped one at a time to verify proper response. These 4 CEA's are each located in a different Core quadrant.

All 4 tests were run for 2 minutes a piece without operator intervention, which was suf6cient time for RCS Parameters to stabilize, following the CEA drop. ,

The next portion of the test was to drop CEA #14 starting at 100% power, Middle of Core Life, equilibrium Xenon, steady state conditions and 100% Power End of Life, equilibrium Xenon, steady state conditions to verify the differences between reactivity feedbacks at l various times in Core life (BOL vs. MOL vs. EOL). These tests were run for 2 minutes a piece to allow RCS parameters to stabilize without operation intervention. r The next portion of the test was to drop CEA #68, starting from 100% Power, Middle of Core Life, equilibrium Xenon, steady state conditions, remove the malfunction and withdraw the CEA back to pull out position.

O, i The next portion of the test was to drop 1 CEA in each of CEA groups B,1,2,3,4,5 and 7 starting from 100% Power, Middle of Core Life, steady state conditions, equilibrium Xenon to verify proper alarm response.

l Following the installation of the current core, testing identified that if a S/D group A CEA i was dropped (or uncoupled) a reactor trip would result. A DR was prepared (DR 97-2-0032) and to prevent negative training all S/D group 'A' CEAs cannot be dropped or uncoupled. l Baseline data is from Reference Plant dropped CEA events and best estimate analysis. Other than the above discussion on S/D 'A' CEAs no de6ciencies were identified.

RD02 Malfunction - Stuck Control kod (CEA) i This malfunction was tested in May,1997. This malfunction is capable of making any of the l l 61 CEA's stick at their current position. All 61 CEA's were tested by starting from 100%  !

l Power, Middle of Core Life, equilibrium Xenon, steady state conditions and verifying that a reactor trip did prevent the malfunctioning stuck CEA's from inserting. Baseline data was from Best Estimate analysis. No de6ciencies were identined.

O 10 i

t

i i

~

RD03 Malfunction - Partial CEA Drop  !

! This malfunction was tested in May,1997. This malfunction is capable of causing any of the 61 CEA's to drop from 0 to 100 steps below its current position. This malfunction was tested l at 10%,50% and 100% severity for CEA #1 and at 100% severity for all other CEA's. Tests l were started from 100% Power, Middle of Core Life conditions, steady state, equilibrium l Xenon conditions. Tests were ended when verifying that the individual CEA had dropped, l l alarm and position indication was correct, and RCS temperature and power had responded to l I

the partial CEA Drop. Baseline data was best estime.te analysis. No deficiencies were  !

identified.

l RD04 Malfunction - Ejected CEA '

This malfunction was tested in May,1997. This malfunction is capable of causing any of 4  !

CEA's to be ejected with a resulting 275 gpm RCS leak and reactivity effect dependent on 7

how far the CEA was inserted. All 4 CEA's were individually tested by starting from Middle  !

of Core Life conditions, Hot Zero Power, critical with Group 4 CEA's at 109 steps. The test was run for 5 minutes, allowing sufficient time for Pressurizer level and pressure trends to be j l established. Baseline data was from FSA R analysis of the Ejected CEA event for reactivity i

effects and NU Retran analysis to av':^-- RCS leak portion of this event. This is not a j training issue since we do not train on this .: vent ni the Simulator.

i RD09 Malfunction - Uncoupled CEA i O This malfunction was tested in May,1997. This malfunction is capable of uncoupling any of I

the 61 CEA's, from their driveGaft. This malfunction was tested by uncoupling all 61 CEA's starting from 100% Middle of Core Life conditions and verifying that a power (reactivity)  ;

drop took place as each CEA was uncoupled. Baseline data was best estimate analysis. No j deficiencies were identified. ,

l l

Following the installation of the current core, testing identified that if a S/D group A CEA )'

was dropped (or uncoupled) a reactor trip would result. A DR was prepared (DR 97-2-0032) l and to prevent negative training all S/D group 'A' CEAs cannot be dropped or uncoupled.

j Other malfunction which may be used to give Control Rod Failures are: RDOS, RD08, RD10 l and RDI1. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

1 l

i O l

< 1 11 i

i

13. Inability to Drive Control Rods (CEA's) Malfunction Abstract RD06 Malfunction - Switch Failure l

This malfunction was tested in April,1992. This malfunction causes the inability to move any control rods. This malfunction was tested at 100% power, steady state conditions and ended when it was verified that no CEA's could be moved. Baseline data was best estimate.

No deficiencies were identified. i l

14. Fuel Cladding Failure - Malfunction Abstracts CR01 Malfunction - Fuel Clad Failure <

l l This malfunction was tested in April,1995. This malfunction is equivalent to a fuel rod gap l release from 2% failed fuel at 100% malfunction severity. This malfunction was tested at

.1%,10% and 100% severity. The test was started from 100% power, steady state conditions, Middle of Core Life conditions. The test was ended when radiation monitor response had stabilized. Baseline data was from the FSAR, some reference plant data and some best estimate analysis. No deficiencies were identified.  ;

i I RC01 Malfunction - RCS Crud Burst P

l () This malfunction was tested in May,1995. This malfunction at 100% severity gives a full scale indication on the Letdown Radiation Monitor. This malfunction was tested at 20% and 100% severity. The test was started at 100% power, steady state, Middle of Core Life j conditions and ended when the Letdown Radiation Monitor reading had stabilized. Basehne I data was from the FSAR and best estimate analysis. No deficiencies were identified.

i

15. Turbine Trin Malfunction Abstracts TC01 Malfunction - Turbine Trip l

This malftmetion was tested in April,1994. This malfunction inserts a turbine trip. It was l tested at 100% power, steady state conditions and ended after verification of Reactor and  !

! Generator trips. Baseline data was best estimate. No deficiencies were identified.

TC09 Malfunction - Electro 11ydraulic Control (EllC) Pump Failure This malfunction was tested in April,1994. This malfunction causes either EllC pump to I trip. Tripping both EliC pumps eventually causes a Turbine trip. This test was run at 100%

power, steady state conditions and ended when the Turbine had tripped due to low EllC pressure. Baseline data was best estimate. No deficiencies were identified.

4

,]

12 l

l t

Other malfunctions which may be used to give a Turbine Trip are:

CW01, FW01, FW33, TP04, TU01, TUO3 and TUO4. The Cause and Effects descriptions bq may be referred to for each of these malfunctions to describe the malfunctions characteristics.

No deficiencies were identified.

f

16. Generator Trin - Malftmetion Abstracts l EG01 Malfunction - Main Generator Trip l

l This malfunction was tested in April,1994. This malfunction inserts a Main Generator trip.

l It was tested starting from 100% power, steady state conditions and ended after verification

of a Generator / Turbine and Reactor trip. Baseline data was best estimate. No deficiencies

were identified, i t

l ED01 Malfunction - Loss of Normal Station Service Transformer This malfunction was tested in April,1994. This malfunction causes a loss of the NSST. It was tested from 100% power, steady state conditions and ended after a Generator trip, i i

Reactor trip and Turbine trip was verified as well as a transfer of power to the RSST.

Baseline data was best estimate. No deficiencies were identified.

i . 17. Reactivity / Core Heat Removal Automatic Control System Failures - Malftmetion Abstracts CV09 Malfunction -Inadvertent Dilution  ;

i This malfunction was tested in April,1997. This malfunction inserts up to a 50 gpm dilution into the RCS at 100% severity. The malfunction was tested at 20% and 100% severity )

starting from 100% power, Middle of Core Life, equilibrium Xenon, steady state conditions  ;

and ending after a power increase was noticeable. Baseline data was best estimate analysis.

No deficiencies were identified.

Other malfunctions which affect reactivity related Automatic Control Systems are CV10, i CVl6 and CVl9 malfunctions. The Cause and Effects descriptions may be referred to for  !

each of these malfunctions to describe the malfunction characteristics. No deficiencies were )

identified.

FWO9 Malfunction - Feed Regulating Valve Position Failure l l

This malfunction was tested in June,1994. This malfunction allows either S/G main feed reg I valve to be positioned from 0% (full closed) to 100% (full open) depending on 0 100%

. malfunction severity. This malfunction was tested at 10% and 100% severity on each S/G starting from 100% power, steady state conditions. The test was ended when SG levels were

' O seen teirend grenerir deeie the m aifunctionins vaives. na eiine date wa be t cetimete analysis. No deficiencies were identified.

13 l

i l

i j Other malfunctions which affect Core heat removal Automatic Control Failures are the  !

FW10, RX10, RX11 and RX12 malfunctions. The Cause and Effects descriptions may be s

referred to for each of these malfunctions to describe the malfunction characteristics. No ,

, deficiencies were identified.

' Note that loss of or degraded Core heat removal from RCP failures or Shutdown Cooling  :

have been covered previously. Covered here was Core heat removal loss due to loss of i Steam Generator Water Inventory Control Systems and reactivity Automated Control Systems.

I

18. RCS PressureNolume Control Systems - Malfunction Abstracts RX01 Malfunction - Pressurizer Spray Valve Controller Failure This malfunction was tested in May,1996. This malfunction simulates an controller failure l on either spray valve. 0% to 100% malfunction severity corresponds to valve closed (0%)  !

and valve full open (100%). The malfunction was tested at 0%,10%,50% and 100%

severity for both spray valves. The malfunction was tested starting from 100% power, steady state conditions and ended when the plant tripped on low RCS Pressure. Baseline data was i best estimate analysis. No deficiencies were identified. t Other malfunctions which cause RCS PressureNolume Control Systems Failures are RX02,  ;

RX03, RX04, RXO6, RX07 and RX08. The Cause and Effects descriptions may be referred j to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

i i

i

19. Reactor Trin - Malfunction Abstracts l l

RP02 Malfunction - Spurious Reactor Trip This malfunction was tested in May,1997. This malfunction inserts a momentary, spurious Reactor Trip signal from the manual Reactor Trip System. This malfunction was tested starting from 100% power, steady state conditions and ended when a Reactor Trip was verified to occur. Baseline data was best estimate analysis. No deficiencies were identified.

Other malfunctions which can cause a Reactor Trip to occur are the RP05 and RP25 malftmetions. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

l O

14

i

20. Main Steam Line and Feed Line Breaks - Malfunction Abstracts

! 3

[ (V MS01 Malfunction - Main Steam Line Break - Inside Containment This malfunction was tested in May,1997. The malfunction is capable ofinserting up to a 1.9 x 10' lbm/hr Main Steam Line Break into Containment from either Steam Generator.

This malfunction was tested from 100%,50% and 5% severity for both Steam Generators.

l All tests were conducted from 100% power, Middle of Core Life conditions, equilibrium

! Xenon, steady state conditions for a period of 8 m'nutes. Baseline data is from CEN-128 Case A4D. No deficiencies were identified.

1 >

MS02 Malfunction - Main Steam Line Break - Outside Containment l This malfunction was tested in April,1995. This malfunction is similar to MS01 except it is outside Containment. MS02 is capable ofinserting up to a 1.9 x 10' lbm/hr Main Steam Line Break outside of Containment from either Steam Generator. This malfunction was tested from 100%,50% and 5% severity for both Steam Generators. All tests were conducted from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady stat ;onditions for a period of 8 minutes. Baseline data is from CEN 128 Case A4D. No deficienc:es were

! identified.

FW25 Malfunction - Feedline Break - Inside Containment (G) This malfunction was tested in April,1995. This malfunction is capable ofinserting up to a 1.0 x 10' lbm/hr Feedwater Line Break inside Containment for either Steam Generator. This malfunction was tested from 100%,50% and 10% severity for both Steam Generators. All tests were conducted from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions for a period of 5 minutes. Baseline data is from CEN 128 Case B5A.

No deficiencies were identified.

Other malfunctions which cause Main Steam or Feedwater Leaks are MS03 and FW27. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

21. Nuclear Instrumentation (ND - Malfunction Abstracts EPl7 Malfunction - Wide Range Detector liigh Voltage Failure This malfunction was tested in May,1997. This malfunction causes a loss ofIligh Voltage to any of the 4 Wide Range detectors causing loss ofindication. This malfunction was tested l for each of the 4 detectors starting from llot Zero Power, Critical at 10* power and ended when allindications were verified to go to the failed position. Baseline data was from best estimate analysis and electrical drawings. No deficiencies were identificd.

a w 15

l l RP23 Malfunction - Random Noise on Summed NI Power This malfunction was tested in May,1997. This malfunction causes a 10% power spike on any of the 4 RPS channels. At 100% severity the spike occurs every ~ 2 minutes. This ,

malfunction was tested at 25% and 100% severity on all 4 channels starting from 100%

power, steady state conditions. The test was ended when the spike was verified to occur and ,

l indications responded to the spike. Baseline data was from best estimate analysis. No

! deficiencies were identified.

Other malftmetions which cause Nuclear Instrumentation failures are malfunctions: RPl5, RP16, RP18, RP19, IU'20 and RP21. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunctions characteristics. No deficiencies were identified.

l

22. Process Instrumentation. Alarms & Control System Failures - Malfunction Abstracts l

CR02 Malfunction - Incore Detector Failure This malfunction was tested in April,1995. This malfunction causes any of the 180 Incore detectors to read 0 to 1100 millivolts corresponding to 0% to 100% malfunction severity.

This malfunction was tested for all 180 Incore detectors at 10%,50% and 100% malfunction severity. The test was started at 100% power, Middle of Core Life, steady state conditions J and ended when all expected alarms were received. Baseline data is from best estimate {

analysis. No deficiencies were identified.

Other malftmetions which cause Process Instrumentation, Alarm and Control System Failures are: Malfunctions CR03, C'J05, CV06, RM01, RM02, RM03, RXO9, RX15, SIO2, SIO1, i CV07, CV11, CV15, CWO3, EG02, EG07, EG10, EG12, FWO2, FW10, FW12, FW14,  !

FW16, FW28, MSO4, MSOL MS13, MS16, RH03, RH07, RX10, RX11, RX12, SWO7,

)

SWO8, TC02 TC04, TC07 a. ' TT 02. The Cause and Effects descriptions may be referred to for each of these malfunctions da describe the malfunction characteristics. Minor deficiencies were identified and are being corrected with the DR process.

23. Passive Malfunctions in Engineered Safeguards Systems (ESAS)- Malfunclio Abstracts  !

ES03 Malftmetion - Engineered Safety Feature (ESAS) Actuation Module - Failure to l Actuate i

This malfunction was tested in May,1997. This malfunction causes any of 7 Actuation Modules, to fail to start its components on an accident signal. All testing was started from 1 100% power, steady state conditions. The test was ended after each of the 7 actuation modules was verified not to start its equipment on the appropriate accident signal. Baseline O

amte wee rre m seete8 tim ete e eirsie ea eneretiec erecea res. we acticieeciee were identified.

i 16

l l

ES01 Malfunction - Automatic Auxiliary Feedwater Failure This malfunction was tested in May,1997. This malfunction causes either of the 2 channels of Auto Aux Feed not to actuate. Both malfunctions were tested starting from 100% power, tripping the Reactor and allowing Steam Generator levels to go low enough to cause an Auto Aux Feed actuation. The tests were ended when it was verified that Auto Aux Feed did not actuate on that channel. Baseline data was from best estimate analysis and operating procedures. No deficiencies were identified, t Other malfunctions which cause or can cause passive malfunctions in ESAS components are:

ES06, CC01, CH01, CV04, CV17, EG09, EGl1, FW20, RH06, RIIO7, S104 and SWO1. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction's characteristics. No deficiencies were identified.

24. Failure of Automatic Reactor Trio System (RPS)- Malfunction Abstracts There are 10 parameters monitored by the Reactor Protection System (RPS) that can cause the reactor to automatically trip. Any 1 or all 10 may be bypassed to fail to cause an automatic Reactor Trip. These bypasses are all remote functions on the MP2 Simulator, not malfunctions. As such, these 10 bypasses are all tested in the Reactor Protection System Test. This testing was performed in April,1991 and consisted of starting from 100% power, p steady state conditions, inserting the auto trip bypass for all 10 RPS monitored parameters V and causing conditions that should cause an automatic RPS trip. The test was ended when all 10 parameters had been verified to not initiate an automatic Reactor Trip. No deficiencies were identified.

RPO4 Malfunction - Manual Reactor Trip Failure This malfunction was tested in May,1997. This malfunction causes any of the 4 manual trip pushbuttons to not work. The automatic RPS trip is unaffected. The malfunction was tested by starting at 100% power, steady state conditions and testing each of the malfunctions to not allow its associated Trip Circuit Breakers to open. Baseline data was from best estimate analysis and electrical drawings. No deficiencies currently exist.

25. Reactor Pressure Control System Failure WWR's)

Millstone 2 is a PWR, this is not applicable.

l I

l l 17 l

l

Malf Test List For Certification (n)

CC01 A-C RBCCW Pp Trip CC02A-C RBCCW Pp Disch Rupture CC06A-D Blockage of RBCCW to RCP Cooler  !

CCI1 A-C Degraded RBCCW Pp Capacity 1 1

CH01 A-D CAR Fan Trip CR01 Fuel Clad Failure l CR02101-294 Incore NI Failure CR031-45 Incore CET Failure CV01

• L/D Line Rupture in CTMT (unisolable)

CV02

• L/D Line Rupture in Aux Bldg CV03 L/D Hx Tube Rupture CV04A-C Charging Pp Trip CV06 L/D Pressure Transmitter Failure CV09 Inadvertent Dilution CV10 Inadvertent Boration CV13 Charging Line Leak in CTMT

(] CV14 CV17A,B Charging Line Leak in Aux Bldg BA Pp Trip CV18

• L/D Line Rupture in CTMT (isolable) i CVl9 PMW Batch Counter Failure l

CWO1 A-D CW Pp Tnp i CWO2A-D Traveling Screen D/P (fouling) l CWO4A-D Condenser Tube Rupture (minor)

CWO6A-D Waterbox Fouling I CWO8A-D Condenser Tube Rupture (major) l CWO9A-D Cire Pp Screen D/P Trip Failure l

ED01 NSST Loss  :

ED02 RSST Loss ED04A,B 25A/B 6.9 KV Bus Loss ED05A-F 24A-F 4.16 KV Bus Loss ED06A-F 22A-F 480V Bus Loss '

ED07A-P 480V MCC's Loss ED08A,B Failure of 25A(25B) to transfer to the RSST following a plant trip ED09A,B Loss of VR-11/21 ED10A,B Loss of 201 A,B DC Bus Rev 1(06/16/97)

1 i

! ED11 A-F Loss ofINV-1,2,3,4,5,6 l f)

ED13A-D -

ED14 Loss of DV10,20,30,40 Loss of Annunciator Power Supply ED15 Loss of 201D DC Bus l ED16A-D Loss of VA10,20,30,40 ED17A,B 24C/D Auto Transfer Failure ,

EG01 Main Gen Trip i EG08A,B A/B DG Output Bkr Failure EG09A,B A/B DG Output Bkr Auto-Close Failure EG11 A,B A/B DG Auto-Start Failure EG12A,B A/B DG Load Control Oscillation ES01A,B A/B AAFAS Failure  !

ES02A,B CII 1,2 Esas Spurious Actuation  ;

ES03A AM503 Failure To Actuate ES03B AM515 Failure To Actuate ES03C AM523 Failure To Actuate ES03D AM602 Failure To Actuate ES03E AM604 Failure To Actuate ES03F AM615 Failure To Actuate  !

ES03G AM619 Failure To Actuate ES03H AM623 Failure To Actuate

[]

ESO4E,F Actuation Cabinet 5,6 Lcss of Power l

ES06AA-AD Z1 Sequencer Failure of Sequence 1,2,3,4 I ES06BA-BD Z2 Sequencer Failure of Sequence 1,2,3,4 l

FWOI Loss of Condenser Vacuum l FWO6A,B SGFP A,B Trip j FW10A,B A,B Main FRV Oscillation l FW13A,B A.B Heater Drains Pp Trip j FW20A,B MDAFP A,B Trip FW20C TDAFP Tr:p FW21A,B AFW Hdr A,B Rupture .

FW25A,B

• A,B Mn Feed Hdr Rupture In CTMT l FW30 A AFW Pp Degraded Performance FW31A/B Fail Open FRV Bypass FW32A/B Failure of SGFP to Trip on Low Vacuum j FW33 Large Vacuum Leak i FW34A/B Stuck Main FRV FW35A/B Stuck Aux FRV i FW36A/B AFW Leak Upstream FW43A/43B

(] Rev 1(06/16/97)

l '

l l

IA01 Instrument Air Ildr Rupture

(] IA05 Instrument Air Ildr Rupture In Turb Bldg MS01 A,B * #1,2 Main Steam lidr Rupture In CTMT MS02A,B * #1,2 Main Steam Ildr Rup:ure In EB (unisolable)

MS03

• Main Steam Ildr Rupture In Turb Bldg (isolable)

MS06A B #1,2 MSIV Spurious Closure MS07A,B #1,2 Main Steam Ildr Safety Fails Open PC01 Loss of PPC RC01 RCS Crud Burst RC02A B* #1,2 Ilot Leg Rupture RC03A-D * #1 A,B; 2A,B Cold Leg Rupture  ;

RC04

• Rx VesselIIcad Vent Leak RC05A,B

• Pzr Safety Viv 200/201 Fails Open RC06A,B

• A,B PORV Fails Open RC07A-D RCP A,B,C,D Lower Seal Failure RC08A-D RCP A,B,C,D Middle Seal Failure RC09A-D RCP A,B,C D Upper Seal Failure RC10A-D RCP A,B,C,D Vapor Seal Failure RCl1 A-D

• RCP A,B,C,D Locked Rotor O RCl3A-D RC20A-D RCP A,B,C,D Sheared Shaft RCP A,B,C,D Thermal Barrier Tube Rupture ,

RD0101-69

• Dropped CEA #1-69

)

RD0201-69 Stuck CEA #1-69 '

RD0301-69 CEA #1-69 Partial Drop RD0401,41,62,65

• CEA #1,41,62,65 Ejected RD06 Failure of"in/ hold /out" switch

' RD0701-69 CEA #1-69 Reed Switch Position ailure RD0901-69

• CEA #1-69 Uncoupled RD10A-I CEA Grp A,B,1-7 Continous Insertion RD!!A-1 CEA Grp A,B,1-7 Continous Withdrawal RD12 CMI Failure R1101 A,B

• A,B LPSI Pp Trip  !

RIiO2A,B A,B LPSI Pp Loss of Suction  ;

R1105A,B A,B SDC IIx Tube Rupture R1106A,B A,B CTMT Spray Pp Trip l

l RiiO8 A LPSI Pp Degraded Performance l l RIIO9 A CS Pp Degraded Performance l R1110A,B RWST Suction Check Valve CS-14A,B Failure To Open O Rev 1(06/16/97)  ;

l

RM01 A-S Variaus Area Radmonitors Failure RM02A-Q Various Process Radmonitors Failure RM03A-G Various Liquid Effluent Radmonitors Failure RM04 Waste Gas Decay Tank Rupture RP01A-P Noisy Th and Tc Temperature Instruments RP02 Spurious Rx Trip RPO4A-D Rx Trip Pushbuttons Failure RP05A-L RPS Trip Matrix Power Supply Failure RP06A-D A,B,C,D RPS TM/LP Calculator Failure RP09A-H RCS Loop Flow Transmitter Failure RP10A-D Pzr Pressure Transmitter Failure (safety channel) i RPl1 A-li RCS Th Temperature Transmitter Failure i RP12A-II RCS Tc Temperature Transmitter Failure RP13 A-li SG #1,2 LevelTransmitter Failure (safety channel)

RPl4A-H SG #1,2 Pressure Transmitter Failure (safety channel)

RPISA-D CH A,B,C,D Wide Range NI Failure RP18A-D Power Range CH A,B,C,D Failure j RPl9A-D Power Range CH A,B,C,D Lower Detector Failure l RPl9E,F Power Range CH X,Y Lower Detector Failure RP20A-D Power Range CH A,B,C,D Upper Detector Failure

(]

RP20E,F Power Range CH X,Y Upper Detector Failure RP24A-D CTMT Pressure Transmitter A,B,C,D Failure i RP26A-D CH A,B,C,D RPS Loss of Power RX01 A,B Pzr Main Spray Viv Fails Open l RX03A,B Pzr Pressure Controller X,Y Failure l RX04A,B Pzr Level Controller X,Y Failure RX06A,B Tavg Program Calculator X,Y Failure RXO7A,B Th Transmitter Failure (control)

RX08A-D Tc Transmitter Failure (control)

RX10A-D SG #1,2 Feed Flow Transmitter Main / Alt. Failure RX11A-D MS Hdr #1,2 Steam Flow Transmitter Main / Alt. Failure RX12A-D SG #1,2 Level Transmitter Main / Alt. Failure -

RX13A,B ADV #1,2 Fails Open l RX14 A-D SD+BV A-D Fails Open ]

SG01A,B SG #1,2 Tube Leak SG02A,B

• SG #1,2 Tube Rupture l S102A-D A,B,C,D ~ ' ; Leakage I S104A-C A,B,C HPat Pp Trip l S105A-C A,B,C HPSI Pp Degraded Performance O. Rev 1(06/16/97) l

SW0lA-C A,B,C SW Pp Trip SWO2A-C A,B,C TBCCW Hx Tube Side Flow Blockage SWO3A-C A,B,C RBCCW Hx Tube Side Flow Blockage SWO4A,B DG A,B SW Flow Blockage SWO5A,B A,B SW lidr Rupture in Turb Bldg SWO9A-C SW Pp A,B,C Strainer Blockage SW10 A SW Pp Degraded Performance SW11 A-C Failure of Switchgear Room Moisture Switch TC05A-D Failure of Turbine Stop Valve TC06A-D Failure of Turbine Control Valve WD01A,B CTMT Sump Disch Viv SSP-16.1,2 Failure to Open WD02A,B CTMT Sump Disch Viv SSP-16.1,2 Failure to Close

• Those malfunctions annotated with an asterisk are included in the list of Major Malfunctions.

Additionally, a composite malfunction consisting of a MSLB in CTMT with an LNP is tested as a Major Malfunction.

Deleted from Rev 1 of Certified List of Malfunctions which can be run for LOIT/LORT

1. RH04A,B A,B LPSI Pp Seal Failure O 2 eso4^-o ^ 8 c.o sees r cebi et 'ess err wer Added to certified list of malfunctions which can be run for LOIT/LORT

1. ED08A,B Failure of 25A(25B) to transfer to the RSST following a plant trip

2. FW31 A/B Failure open of FRV Bypasa

! 3. FW32A/B Failure of SGFP to trip on low vacuum

4. FW33 Large vacuum leak

5. FW34A/B Stuck main FRV

6. FW35A/B Stuck Aux FRV

7. FW36A/B Aux feed leak upstream of 43A/43B

8. RD06 Failure of"in/ hold /out" switch

9. TC05A-D Failure of Turbine Stop Valve 10.TC06A-D Failure of Turbine Control VJ de %

Developed:LodTWoncurren .

e LOFCoordinator 4( (Nb LOIT Coordinat7 SOA Qualified Instr.

This list contains all of the malfunctions required to create or support simulator conditions needed to provide training on the Task Training Requirements of the LOIT and LORT programs.

% bec e cm % %

. y _ .- _ .

i

! ATTACHMENT 4 1O t

l INITIAL CONDITIONS CHECKLIST AND LIST OF 30 CERTIFIED INITIAL CONDITIONS l 5 [

d 4

l a

l I

d l

f i

1 i

O l

l I

i O

l l

s

h l

2 fM ATTACHMENT 8.2 4

MILLSTONE 2 INITIAL CONDITIONS VERIFICATION '

CHECKLIST 1

1. Control Board Walkdown 9

With the Simulator in "run", at each of the following control boards, an SRO licensed or cenified instructor shall review switch positions, controller settings, meter indications, annunciator conditions, system alignments etc. to ensure they are consistent with the intended conditions of the Initial Condition.

J C-01 C-01R C-101 C-25A(B)

C-02 C-02R RC-05E RC - 14 C-03 C-03R RPS Ch A ESAS Ch A C-04 C-04R RPS Ch B ESAS Ch B C-05 C-05R RPS Ch C - ESAS Ch C i C-06 C-06R RPS Ch D ESAS Ch D C-07 C-07R C-OlX ESAS ActCABI C-08 C-08R C-80 ESAS ActCAB2 O C-21 SI-652 Wall Switch RC 100 d i II. Remote Functions Revics With the Simulator in "run", for each of the following remote function systems, review each remote function to ensure its condition is consistent with the intended conditions of j the Initial Condition.

CAR ESR RMR TCR CHR FWR RPR TPR i CVR IAR RXR TUR I CWR MSR SGR WDR )

EDR RCR SIR  !

EGR RHR SWR l

I 1

Rev.: 5 Date: 4/SO/97 Page: 8.2-1 of 4 NSEM-4.02 A

IIL Initial Condition Stability and Reasonability Perform either section A. B or C A. For Equilibrium Xenon, Steady State, Power Levels (50%,75%,100%, etc.) only, ensure the following parameters are stable and reasonable for the first 2 minutes after ;

resetting to the IC and going to run:  ;

o RCS Tc o CHG Flow o ASI reasonable o RCS Tave o L/D Flow and consistent -

o RCS Press o PZR LVL with power Ivl  ;

o S/G Lvl o Feed Flow and CEA posit '

o DT Power o Stm Flow o NI Power o Xe React Worth (Inst Station) i o CVMWTH  !

o Elec MWs o CEA Gp 7 posit I reasonable for l pwr Ivl Perform SP 2601D (when appropriate) to ensure consistency I o Using Xenon Fastime X60, and holding thermal power and CEA position constant, ensure that total xenon reactivity does not change by more than 20 pcm in 4 minutes of Xenon Fastime X60.

o. Usin Xenon Fastime X60, and holding thermal power and CEA position constant

, veri that ASI does not change ,by more than .03 from its initial value (if> 50%

, power) or greater than .05 from its initial value (if <50% power) during al minute period of Xenon Fastime X60. This can be done in parallel with the previous step.

o Ensure any items mentioned in IC description on instructor station are correct and any ke,y items not present on the instructor station IC descri ion are added. Key items m remarks section ofIC are BOL/MOL/EOL, Xenon rend, load limit pot setting, unusual CEA positions, unusual equipment lineups, etc.

O Rev.: 5 Date: 4/30/97 I- NSEM-4.02

i i 4

1 B.- For ICs which have the reactor critical, but do not fall i into category A above, verify the following parameters are ,

stable and/or reasonable for the first 2 minutes after  ;

resetting to the IC and going to run-o RCS Tc o CHG Flow o ASI reasonable and o RCS Tave o L/D Flow consistent with pwrIvl 4 o RCS Press o PZR Lvl and CEA position 1 o S/G Lvl o Feed Flow  :

o DT PwT. o Stm Flow i

o NI Pwr o Xe react worth o CVMWTH (inst station) )

o Electric Power o CEA G a 7 psit i

reasonable :or pwr Ivl o Using Xenon Fastime X60 for 4 minutes, and holding thermal power and CEA position constant, observe that the change in Xenon reactivity worth and ASI are consistent with the stated IC's conditions and will not cause unreasonable conditions to occur for the operator.

o Ensure any items mentioned in IC description on instructor station are correct and any key items not present on the instructor station IC description are added. Key items in the remarks section of the IC are BOL/MOL/EOL, Xenon Trend, load limit pot setting, unusual CEA position, unusual equipment lineups, etc.

O C. For ICs which the reactor is not critical and may be in various stages of Plant S.artup or Shutdown, verify the following parameters are stable and/or reasonable for the first 2 minutes after resetting to the IC and going to run:

o RCS Tc o PZR Lvl o If H/U nr C/D in ~

o RCS Tave o Feed Flow progress,# RCP's o RCS Press o Xe React Worth correct  ;

o S/G Lvl (inst Station) I o W/R Pwr o CHG Flow o Ifon SDC flow / temp is steady o Using Xenon Fastime X60 for 4 minutes, observe that the change in Xenon reactivity )

worth is consistent with the stated conditions of the IC description.

l Rev.: 5 Date: 4/30/97 Page: 8.2-3 of 4 t, NSEM-4.02

._.n, . - . - - - _ - - - _ - . - - - . - . - . . - .-.,-.. -. - -

4 L l 4

i b

V o - Ensure any items mentioned in the IC description on the instructor station are ' correct and any y items not present on the instructor station IC description are added. Key

}

items in e remarks section of the IC are BOL/MOL/EOL, Xenon trend, time after reactor trip, unusual CEA positions, unusual equipment lineups, etc.  !

IV. IC Reauirements to be Specifically Verified I i

l- ' Verify the following: -

l Remote Functions '

l CCR03 set to 50 F (to fail B HXTCV open) .

CVR03 set to open - ,

4

' CVR13 set to c osed l CWR01 set to 55"F i i

FWR55 set to stet gp 1(at-?owerICs)  !

FWR55 set to 0 gpm shutc own ICs) ,

.. IAR17 set to lead f

IAR15 set to start  ;

MSR03 set to auto MSR05 set to auto MSR07 set to auto  !

DPR31 set to full bdR01 set 10 9.4 f O, St K02 set to 38.0 S?'R04 set to own SWR 05 set to c)osed i

l SWR 06 set to closed i SWR 07 set to closed SWR 12 set to 0%

S WR24 set to 0%

TPR16 set to close TPRI8 set to TBCCW-WDR06 set to open WDR02, set to auto

- PPC Points SCBLDN1 consistent with SGR01 SCBLDN2 consistent with SGR01 1

i i

]

I Rev.: 5 Date: 4/30/97 Page: 8.2-4 of 4 NSEM-4.02

O Initial Conditions IC RCS PZR RX BORON XE REMARKS NO TAVG PRESS PWR CONC REACT 1 On SDC, ready to draw bubble

• • 1 2 On SDC, =40% in pzr, ready to start

• • RCPs 3 532 2267 0 1382 -1 Initial Rx S/U after refueling ECP GP 7 0 90 Steps 4 532 2263 0 1383 -1 BOL, Rx S/U with S/D A and S/D B out, ECP GP 7 @ 90 Steps 5 533 2269 1- 1382 -1 BOL, AFW to both S/G, Turbine at 0 rpm 6 538 2270 10 1381 -18 BOL, 'A' SGFP to both S/G, Turbine at 1800 rpm P 7 540 2265 19 1377 -31 BOL, ready to transfer load to NSST U

8 551 2265 40 1358 -91 BOL, ready to start 2"# condensate pump 9 554 2265 49 1345 -118 BOL, ready to start 2"d SGFP 10 572 2265 100 1016 --2313 BOL, 100% Eq xe 11 535 2265 0 434 -1 EOL, Rx S/U S/D A and B out, Reg GP 1 thru 4 out, ECP GP7 0 90 steps 12 572 2265 100 20 -2632 EOL, 100% Eq xe 13 532 2260 0 1000 -2389 EOL, Rx SD, OP 2207 step 4.1 =24 hr after S/D 14 252 370 0 -2000 -2686 EOL, warm-up of SDC system prior to initiation 15 EOL, on SDC, Pzr level @ 40% ready

,, to drain down J

Rev.: 2 Date: 3/12/97 Page: 1 of 2

_m - -

.______.-.____..__.______--_.-._.__._.-_.._._.m._ _

. > _ _ - _ _ .m. .

lO Initial Conditions IC RCS PER RX BORON XE REMARKS NO TAVG PRESS PWR CONC REACT i 16 88 13 0 2000 -1 EOL, on SDC draining down, at top I 1

of Hot Leg j 17 87 14 0 2000 -1 EOL, on SDC at centerline of Hot Leg l 18 EOL, comming off line at 10% per

,,, hour 19 534 2265 0 952 -1 MOL, Rx S/U SD A and B out, Reg group 1 thru 4 out, ECP Gp 7 0 90 steps 20 534 2265 0 567 -3567 MOL, =4 Hr post trip (Xe inc) 21 534 2265 0 567 -3514 MOL, =12 Hr post . rip (Xe dec) 22 569 2265 89 583 -2444 MOL preparing for turbine valve test 23 541 2265 0 567 -2914 MOL, RCS in stable N/C AEW to both S/G 24 572 2265 100 567 -2424 MOL, 100% Eq xe

• 25 532 2265 0 1125 -5 Conditions which will exist for l

plant S/U @ 5895 MWD /MTU j

• 26 572 2272 100 721 -2423 5895 MWD /MTU, 100%, Eq xe )

• 27 545 2255 28 1004 -75 MOL, =30% pwr - Stability

• 28 552 2255 48 978 -154 MOL, =50% pwr - Stability

• 29 571 2270 96 571 -2432 MOL, =96% pwr - Stability

• Not certified - not for training use.

l

• • Currently under development.

l Rev.: 2 1

Date: 3/12/97

, Page: 2 of 2

}

i j

4 I

i.

! ATTACHMENT 5

!O 4

STUDENT FEEDBACK SURVEY RESULTS

)

t t

i t

i 1

i 1

1 l

3 I

i e

O O

I k File l l

From: W.II. Souder, III MP2 Simulator Operations Assistant

Subject:

Student Feedback 1

Attachment:

1) Memo W.H. Souder dated 10/16/96 l

2) NSEM 6.01 form 7.01

)

l l

1. Twice in the past seven months I have requested student feedback on the simulator.

l

\

• In October 1996, A memo (Att.1) was sent to all MP2 License IIolders asking for input in l

identifying areas in which the simulator model was in error (or could be improved).  !

= 20% of the operators responded and the greatest percentage of comments concerned the modeling of various cooling systems (RBCCW, TBCCW, CW and SW). DRs were generated and work will begin on these model improvements. '

DR 96-2-0050 CW System l

DR 96-2-0051 RBCCW System I DR 96-2-0052 TBCCW System DR 96-2-0053 SW System I

• In April /May 1997 student feedback was requested using nsem 6.01 form 7.01 from four of the six shifts in training during that time.

= 25% responded and has provided useful input in identifying differences between the simulator and the plant.

Items mentioned included:

• Annunciator response for door alarms (C.0lX) is incorrect. (DR-96-2-0021)

• • Inability of the STA to use the SM SPDS terminal to select safety functions or EOPs.

Corrected.

• Use of metal trays for computer keyboards /IDT terminals. Corrected j

• Inability of the operators to put alarms in alarm defeat on the simulator. Currently being evaluated.

• Too much room exists on the US desk, there in an additional PC on this desk which is used throughout the shift. A PC was installed on the desk 5/30/97.

l *

• The alignment of the AFW reset switches on C-05 alignment (position) is not correct.

l Corrected.

1

, 2. Both methods of request have worked well for me. I plan on using both in the future:

,

• The memo to all license holders will be sent out annually.

• I will personally request input from two of the training shifts each training cycle.

I draftWsem\sfb.wsl l

Attachment 1

) 16 Oct 1996 To: ALL MP2 License Holders From: Bill Souder, MP2 Simulator Operations Assistant

Subject:

MP2 Simulator Modeling

1. Over tile past few months, much has been said about the accuracy of the MP2 Simulator in modeling the response of the plant during both normal and accident situations.

2. Since you operate the plant and know how it responds better than me, I am asking for your help in identifying areas that the simulator modeling could be improved.

3. Please complete the attached form and return to me via inter-office mail by 10/24/19.

When completing the form:

Ensure that your name is on this form. This is necessary ifI need to ask any additional information to ensure that I understand exactly what Make any comments as detailed as necessary. This again will ensure that I understand

]C what you are saying. So that I can understand what you are saying.

4. At the completion of this data gathering, I will provide a summation of all problems which have been identified.

Name Date L

l i MP2 Simulator Modeling Questionnaire  !

t I

. Simulator systems which are modeled

PBCCW, CVCS, Electrical Distribution / Generation, Feed / Condensate, Instrument Air, Main Steam, RCS, SDC, Rad Monitors, RPS, Reactor Reg, S/G,  !

Safety Injection, SW, Turbine, TBCCW and Waste Disposal

1. How closely does the MP2 Simulator model the response of the plant?

l l ,

I l

l 6

2. List any areas in these system which you feel could be modeled differently to improve the overall training.

l l

i Use additional paper to answer either question i O

1

- Att:chment 2 FORM 7.1 l

i i

1 I

To: All Simulator Users i Unit  ;

i c From- I Simulator Operations Assistant j

Subject:

Accuracy of Simulator Unit Ifyou have noticed issues concerning the accuracy of the simulator please complete the appropriate sections of this form. Unit We are interested in knowing of any deficiencies you may have observed between the simulator and the actual control room / control boards. This fonn should be used to address i fidelity issues only. Do not use this form to address training topics or issues.  ;

t Please return the completed form to either the instructor or the unit SOA. Please include your name in the space provided if you wish to have a written response forwarded to you regarding the comments you have provided.

If you have and questions please contact the unit SOA at EXT .

Comments Submitted by: DATE:

Rev.: 3 Date: 4/18/97 Page: 7.1-1 of 5 NSEM-6.01

FORM 7.1

(^') Instructions v

For the following areas of simulator fidelity please indicate the problem you observed including the following information:

The simulator's condition / response (what evolutions were in progress and what the simulator did or did not do)  ;

• The condition or response you expected Any special plant conditions needed to reproduce the condition or response including any ,

temporary ICs that may have been produced to capture the condition )

• Any plant procedures in use including step numbers significant to the problem and Any applicable reference sources (if known and applicable) to assist in the problem investigation The more specific information you can provide to the Simulator Computer Engineering group the faster we can resolve the problems you have identified. t (3 t s

V i

k

/^S 1

/

Rev.: 3 Date: 4/18/97 Page: 7.1-2 of 5 NSEM-6.01

. _ _ _ . . . _ . _ - - _ . ~ . _ _ . _ _ _ - . . . _ _ _ _ . _ _ _ _ . _ . . _ _ - _ _ . . _ _ _ _ . . .

l i

l FORM 7.1 l l ,

\

i l SIMULATOR FIDELITY FEEDBACK i i

4 a A. PANELS. INDICATION AND CONTROLS ,

i

1. Use this area to report observed differences between the plant and the simulator regarding  ;

l panels, meters, switches, lights, scales, ranges, locations, etc.  !

}

i l

1

)

e i

I I

2. Use this area to report any observed difference between the plant and the simulator j regarding mimic, back shading, name tags, labels, etc. 1

.O 1 B. COMMUNICATION EOUIPMENT

1. Use this area to report any observed differences between the plant and the simulator regarding the communications devices available.

O Rev.: 3 Date: 4/18/97 Page: 7.13 of 5 NSEM-6.01 I

1

_- - . . . .. , -; .. . . . - - - - .- - - - - -- ,,-.~ -- '

l l

l FORM 7.1 1

i l r's C. ENVIRONMENT l !

Q '

1. Use this space to report any observed differences between the plani and the simulator l regarding the amount and type of normal and/or emergency lightmg.

l l

l l l

l J l

l 1

2. Use this space to report and observed differences between the piant and the simulator {

regarding the amount, type and arrangement of furniture or incidental equipment (timers, I print racks, calculators, etc.).

l t

D. PLANT COMPUTER i

1. Use this space to report any observed differences between the plant and the simulator regarding the PPC input and output devices (CRTs, keyboards, printers, etc.). j i

l 2. Use this space to report any observed differences between the plant and the simulator regarding the PPC functions, capabilities and responses.

l l

Rev.: 3 l Date: 4/18/97 l Page: 7.1-4 of 5 l NSEM-6.01 l

l

1 FORM 7.1 l

<' E. SIMULATOR RESPONSE

1. Use this space to report any simulator response that you believe to be incorrect or l questionable.

l 1

9 l

l

2. Use this space to report any procedure section, step or operation you were unable to perform because oflimitations of the simulator. l l

l s .

l 1

l l

l t0 l kl -

Rev.: 3 i Date: 4/18/97 j Page: 7.1-5 of 5 :

NSEM-6.01 l

ATTACHMENT 6 OPEN DEFICIENCY REPORT (DR) LIST O

O

[ I l

l

' MP2 SIMULATOR OPEN DR STATUS - ASSOCIATED WITH CERTIFICATION l O l DRf DESCRIPTION WHEN SCHEDULED {

95-2-0044 MS14, MS15 and MS16 response 12/97 l

1 95-2-0042 MSI1 A affects suspect 12/97 l i

i 97-2-0032 SD "A" CEA worth (or modeling) 7/97 l 97-2-0033 Incore temperatures inaccurate 7/97  !

97-2-0040 RM01B and RM01E response (no inst fall light 8/97 when fail) i 97-2-0041 RD03 response (incomplete annunciator actuation) 7/97 l 0 97-2-0043 EG07 response (no visible effect on generation) 12/97 l

97-2-0044 RH05 response (no increase in RBCCW activity) 10/97 97-2-0046 RD04 response (leak flow much greater than 275 12/97 gpm) (Reactivity effect questionable) (RCS pressure response questionable) 97-2-0047 RP22 response (not all RPS trip units actuated by 12/97 actuation of PTT) 97-2-0049 PC01 response 8/97 97-2-0050 RM02A response 8/97 l

. 97-2-0051 RM02P response 8/97

\ l l

l draftinsem\sodrs.wsl

- ..4.--

a ^7 3-1 4

f L

s

{

j ATTACHMENT 7 e

SCHEDULE FOR NEXT FOUR YEARS OF TESTING e

l 4 i 8 l f

i  !

e a  :

I, i f

i I

i. ,

l l  !

r I l l i

• l l

l l .

l I

l l

!O 1

i l I

\

4 1

)

l i

O

_ _ . . . - . _ .. . _. _ _ . _ . ~ _ . _ _ . _ . _ _ _ . . _ _ . . _ . . _ . . _ _ _ _ _ _ . . . _ . _ _ . _ _ _ _ _ _ . -

i 4

J. ,

1... ,

f I i- .!

ATTACHMENT 8.2 j.

MP2 .

PERFORMANCE TEST .

I i SCHEDULE I l

l J

)

i START- ENQ t

i Performance Test (4 Year Cycle): Julv 199L June 2001 -  ;

i Year One: ' July 1997 June 1998 - )

I

-Year Two: July 1998 June 1999 l l

Year Three: Juiv 1999 June 2000 i Y July 2000 June 2001  !

O earFour: i i

i i

l l

-i :

- t

/ . APPROVED BY ASOT' O Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-1 of 11

YEAR ONE f ( IESI SCHEDULING WORK AREA Annual Operability NSEM-4.09 28% Steady State Accuracy 50% Steady State Accuracy 100% Steady State Accuracy 100% Stability Transient #1: Manual Reactor Trip Transient #2: Simultaneous Trip of All Feedwater Pumps l

 ,s Transient #3: Simultaneous Closure of

( All Main Steam isolation Valves l 1 Transient #4: Simultaneous Trip of All I Reactor Coolant Pumps l Transient #5: Trip of the "A" Reactor Coolant Pump Transient #6: Main Turbine Trip From Power Level Not Resulting in immediate i Reactor Trip Transient #7: Rapid Ramp Rate Decrease in Plant Power  : i Transient #8: Maximum LOCA with LNP l i 1 i O l Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-2 of 11

YEAR ONE i l

l ! (,)

 /                 TEST                                SCHEDULING WORK AREA Transient #9: Maximum Unisolable MSLB inside Containment Transient #10: Slow RCS Depressurization to Saturated Conditions Using "A" PORV Stuck Full Open (1 HPSI Defeated)

Physical Fidelity Verification ! (NSEM-4.12) Note: SHM requires approximately 2 months advance notice to schedule photographer for photo book update. Instructor Station Test - (NSEM-4.11) Maior Malfunction Tests NSEM-4.04 RD09XX Uncoupled CEA RC11A (B,C,D) RCP Locked Rotor l Minor Mal functions NSEM-4.05 (Listed by System) Electrical Distribution - ED l Reactor Protection System - RP  ; Electrical Generation - EG l Condensate /Feedwater- FW Turbine Controls - TC TBCCW - TP l Plant Computer - PC Rev: 3 Date: 5/28/97 NSEM-4. 07 Page: 8.2-3 of 11

YEAR TWO TEST SCHEDULING WORK AREA Annual Operability NSEM-4.09 28% Steady State Accuracy 50% Steady State Accuracy 100% Steady State Accuracy 100% Stability Transient #1: Manual Reactor Trip Transient #2: Simultaneous Trip of 1 All Feedwater Pumps { Transient #3: Simultaneous Closure of All Main Steam isolation Valves 1 Transient #4: Simultaneous Trip of All Reactor Coolant Pumps Transient #5: Trip of the "A" Reactor Coolant Pump Transient #6: Main Turbine Trip From Power Level Not Resulting in immediate Reactor Trip Transient #7: Rapid Ramp Rate Decrease in Plant Power Transient #8: Maximum LOCA with LNP Transient #9: Maximum Unisolable MSLB Inside Containment O. l Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-4 of 11 we*

YEAR TWO l TEST SCHEDULING WORK AREA 1

Transient #10

Slow RCS Depressurization  :

4 to Saturated Conditions Using "A" PORV i Stuck Full Open (1 HPSI Defeated)  ! Physical Fidelity Verification (NSEM-4.12)  ! Note: SHM requires approximately 2 ' l

     - months advance notice to schedule                                                                                             r j      photographer for photo book update.

{ 1 Normal Plant Evolutions NSEM-4.10  : Plant Startup f Nuclear Startup , 2 Turbine Startup and Generator Synchronization j OReactor Trin and Recoverv i  ! Hot Standby Operation Load Changes  ! Plant Shutdown [ l Surveillance Testing  ! Note: Individual surveillance tests are  ; performed during normal operations. l l 1 J O Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-5 of 11 3 j

YEAR TWO I l f IESI SCHEDULING WORK AREA Maior Malfunction Tests NSEM-4.04 1 RC05A(B) RCS SV Failure  ; I RC06A(B) PORV Failure l MS01 A(B) MSLB in CTMT - i 1 RD01XX Dropped CEA l l RCO2A(B) Ts LOCA j l RD04XX Ejectea CEA l

                                                                                    \

l l RC03A(B,C,D) Tc LOCA r,  ! (g'MS02A(B) MSLB out CTMT Minor Malfunctions NSEM-4.05

{ Listed by System) l Main Steam - MS

! Reactor Coolant System - RC Core - CR l l j Shutdown Cooling - RH Control Rods - RD Steam Generators - SG 3 (J Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-6 of 11

l YEAR THREE TEST SCHEDULING WORK AREA Annual Ooerability NSEM-4.09 28% Steady State Accuracy  ; 50% Steady State Accuracy

    . 100% Steady State Accuracy 100% Stability
   - Transient #1: Manual Reactor Trip Transient #2: Simultaneous Trip of All Feedwater Pumps Transient #3: Simultaneous Closure of AllM ain Steam Isolation Valves O

Transient #4: Simultaneous Trip of All Reactor Coolant Pumps

    - Transient #5: Trip of the "A" Reactor Coolant Pump Transient #6: Main Turbine _ Trip From Power Level Not Resulting in immediate Reactor Trip Transient #7: Rapid Ramp Rate Decrease in Plant Power Transient #8: Maximum LOCA with LNP Transient #9: Maximum Unisolable MSLB
    . Inside Containment O                                                                                                                                                   ,

Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-7 of 11 l

                                                                                                                                                      \

    .         .    . _ . -    . . _... .~ _ _ __ _ _. _ ._ _ __.___ _ . .                                    . _. . _ _ . _ .

YEAR THREE [ TEST SCHEDULING WORK AREA l Transient #10: Slow RCS Depressuriz-ction to Saturated Conditions Using "A" PORV Stuck Full Open (1 HPSI Defeated) l Real Time Test - MP2 (NSEM-4.13)

      , Physical Fidelity Verification (NSEM-4.12) l       Note: SHM requires approximately 2 months advance notice to schedule photographer for photo book update.

Ma_ior' Malfunction Tests NSEM-4.04 ' FW25A (B) FW Line Break in Ctmt CV01 A LD Line Leak in Ctmt. L CV02A LD Line Leak in Aux Bldg. CV18A LD Line Leak in Ctmt. (Isolable) Minor Malfunctions NSEM-4.05 (Listed by System) CTMT/ Heating / Vent. - CH Instrument / Station Air -IR l Reactor Regulating System - RX i Waste Disposal- WD

Service Water- SW Turbine - TU O  !

l l l Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-8 of 11

_ _ . . _ . - .__ . _ .._. _. _ _ _ _ . . _ . . _ . . _ _ _ _ . . _ . _ - ~ .. _- -_ _ . _ _ . _ . _ t l l i i l l YEAR FOUR  ! i i TEST SCHEDULING WORK AREA O. i l Annual Ooerability NSEM-4.09 -  ! I i i 28% Steady State Accuracy l 50% Steady State Accuracy 100% Steady State Accuracy l 100% Stability Transient #1: Manual Reactor Trip  ! Transient 1/2. Simultaneous Trip of All  : Feedwater Pumps l Transient #3: Simultaneous Closure of i All Main Steam isolation Valves O Transient #4: Simultaneous Trip of All Reactor Coolant Pumps Transient #5: Trip of the "A" Reactor  ! Coolant Pump Transient #6:' Main Turbine Trip From Power Level Not Resulting in immediate Reactor Trip Transient #7: Rapid Ramp Rate Decrease in Plant Power l' Transient #8: Maximum LOCA with LNP Transient #9: Maximum Unisolable MSLB Inside Containment O l Rev: 3 Date: 5/28/97 1 NSEM-4.07 Page: 8.2-9 of 11

YEAR COUR p TEST SCHEDULING WORK AREA V Transient #10: Slow RCS Depressuriz-ation to Saturated Conditions Using "A" PORV Stuck Full Open (1 HPSI Defeated) Physical Fidelity Verification (NSEM-4.12) Note: SHM requires approximately 2 months advance notice to schedule photographer for photo book update. Maior Malfunction Tests NSEM-4.04 MSO3 MSLB in Turb. Bldg i MSLB W/LNP Compof.ie RC04 RCS Leak r~)RH01 A (B) LPSI Pp Trip (Loss of SDC) %J , SG02A (B) SGTR I Minor Malfunctions NSEM-4.04 l (Listed by System) RBCCW- CC CVCS-CV Circulating Water- CW Engineered Safeguards - ES Radiation Monitoring - RM 4 Safety injection - SI l O b Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-10 of 11

4 MARGINAL NOTE DIRECTORY ATTACHMENT 8.2

v

1. Deleted the requirement to perform approximately 25% of all system test on an annual basis. System

, testing has proven to produce very few discrepancies and has bs 7 deleted. m i i E 1 (3 (/ Rev: 3 Date: 5/28/97 NSEM-4.07 Page: 8.2-11 of 11

, _ . . . - _ _ _ _ . _ . _ . _ _ . . _ _ _ . . _ _ _ . . _ . . _ _ . _ _ _ _ _ _ . _ . . _ . __ ___ ___m-4 5 3 I i ATTACHMENT 8

ANNUAL OPERABILITY TRANSIENT TESTING ABSTRACTS 3-i J

i l i i i > J l. 5 i a 4 1

O

1 i I 1 i O Drabsennatts.ws;

i l; 4 4 YEARLY OPERABILITY TRANSIENT TESTING ABSTRACTS i n. D)%. The following 10 transients were all run in A,-il/May,1996. All parameters discussed below were recorded at a .5 second time interval as required by ANSI /ANS 3.5 (1985) Appendix B. No exceptions to p' ANSI / ANSI. 5 (1985) are taken. 1 Transient #1 - Manual Reactor Trip l i As required by ANSI /ANS 3.5 (1985) Appendix B, a manual reactor trip was performed from 100% { j power, steady state (Middle of Life conditions), equilibrium Xenon. All parameters listed in ANSUANS

'                                                                                                                                      l 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 2 minutes, which is sufficient time for RCS                  !
               . temperature to stabilize and Pressurizer pressure, level and Steam Generator level to ramp towards                    l recovery. Baseline data for comparison is from a April 16,1987 reference plant Turbine Trip. Graphical

! comparisons were made for each of the monitored parameters from Appendix B.2.2.1. All acceptance j

criteria were met, no deficiencies were identified. I I

, Transient #2 - Simultaneous Trip of all Feedwater Pumps  ! ( j As required by AhSI/ANS 3.5 (1985) Appendix B, a simultaneous trip of all Feedwater pumps were l

performed from 100% power, steady state (Middle of Life conditions), equilibrium Xenon. No Main or ,

[ Auxiliary feed ws allowed to the Steam Generator; All parameters listed in ANSI /ANS 3.5 (1985) i Appendix B.2.2.1 were recorded for a period of 4 minutes, which is sufficient time for RCS temperature j , to stabilize and Pressurizer pressure and level to ramp towards recovery. Baseline data for comparison is ' i from CEN 128 Case b.4. Graphi. 'l comparisons were made for each of the monitored parameters from  ! Appendix B.2.2.1. All acceptance - wria were met, no deficiencies were identified. j $ I j TransienL#_'l- Siin %nwus Closure s . All Main Steam Isolation Valves (MSIV's) 1 i- As required by ANSI . JS 3.5 (198.i) Appendix B,9 simultaneous closure of all MSIV's was performed  ! !- - from 100% power, ste. dy state (Middle of Life conditions), e,quilibrium Xenon. To be consistent with  !

anzlytical data to be used for comparison, Atmospheric Steam Demps were not allowed to open,  !

]; Preasurizer PORV's were blocked closed and a loss of all feed occu. red at the same time. All parameters l [ listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded fo. y eriod of 1 minute, which is  ! ! . sufficient time for RCS temperature to stabilize due to the Steam Gi nerator Safety Valves and Pressurizer  ! L Pressure and Level to bottom out and start recovery. Baseline data ist comparison is from a FSAR Case, j Section 14.9, Figures 14.9-1,2,3 and 4 and some best estimate at slyi- Granhical comparisons were

~ made for each of the monitored parameters from Appendix B.2.2.! M. uptance criteria were met, no y deficiencies were identified. i i.

.!O Rev: 1 i i t Date: $/27/97 r Page: I of 4 s

                                                                                                           , - - ,---w              -

Transient #4 - Simultaneous Trip of all Reactor Coolant Pumps (RCP's) b As required by ANSI /ANS 3.5 (1985) Appendix b, a simultaneous trip of all RCP's was performed from 100% power, steady state (Middle of Life conditions) equilibrium Xenon. All four RCP's and 3 . Condensate Pumps were lost by a electrical fault /on both 6.9 KV buses. All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 5 minutes, which is sufficient time for RCS Tave to be controlled on the Condenser Dump Valves. ~ Baseline data for comparison is from an NU generated RETRAN run for MP2, which is a best estimate run for a loss of 4 RCP's at 100% power. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.1. All acceptance criteria were met, no deficiencies were identified. Transient #5 - Trip of any Single Reactor Coolant Pump (RCP) j As required by ANSI /ANS 3.5 (1985) Appendix B, a trip of a single RCP (the "A" RCP) was performed from 100% power, steady state (Middle of Life conditions), equilibrium Xenon. All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.2 were recorded for a period of 4 minutes, which is sufficient time for RCS temperature to stabilize and Pressurizer level and pressure to ramp towards recovery. Baseline data for comparison is from CEN 128 Case C.I. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.2. All acceptance criteria were met, no deficiencies were identified. Transient #6 - Main Turbine Trip from Power Level not resulting in immediate Reactor Trip. As required by ANSI /ANS 3.5 (1985) Appendix B, a Main Turbine Trip from =15% power was performed.15% power is the highest Reactor power that will not cause an immediate Reactor Trip from a Turbine Trip. All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 2 minutes, which is sufficient time for RCS temperature to stabilize and Pressurizer pressure, level and Steam Ger,erator level to ramp towards recovery. Baseline data for comparison is from best estimate andysis. Waphical comparisons were made for each of the monitored parameters from Appendix B.2.2.1. Ali acce,tr r.ce criteria were met, no deficiencies were identified. 'f Transient #7 - Rapid Ramp Rate Decrease in Plant Pcwer from 100% to 75% power. As required by ANSI /ANS 3.5 (1985) Appendix B, a rapid ramp rate decrease in plant power from 100% power to 75% power was performed from 100% power steady state (Middle of Life conditions), equilibrium Xenon. All parameters ': 'ed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 10 minutes, which is sufneau to reduce power to < 75% power by use of boric acid addition to the RCS and use of the Throttle Pres etre Limiter for the Turbine Control valves. Baseline data for comparison is from best estimate analysis. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.1. All acceptance criteria were met, no deficiencies were identified. 1 O Rev: 1 Date: 5/27/97 Page: 2 of 4

Transient #8 - Maximum Size Reactor Coolant System Rupture combined with loss of all offsite power. As required by ANSI /ANS 3.5 (1985) Appendix B, a maximum size Reactor Coolant System Rupture I with a complete loss of offsite power was performed from 100% power, steady state (Middle of Life conditions), equilibrium Xenon. This was done by causing the equivalent of a double ended Hot Leg pipe  ! guillotine rupture coincident with a complete loss of all offsite power, such that only the Emergency l Diesel Generator Buses and battery power were available. 1 l All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.3 were recorded for a period of 5 minutes, which is sufficient time for Containment Temperature and Pressure to peak. Baseline data for comparison is from FSAR Section 14.15 and 14.16; figures 14.5-11 and Figures 14.16-6 & 7. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.3. 1 1 I Following the original transient #8, it was identified that both CTMT pressure and temperature response was unsat. This response was identified as unacceptable, work was done on the simulator model and the 1 transient was run again. The results of this run were, all acceptance criteria were met, no deficiencies l were identified.

 ] Transient #9 - Maximum Unisolable Main Steam Line Rupture l

As required by ANSI /ANS 3.5 (1985) Appendix B, a maximum size unisolable main Steam Line Rupture was performed from 100% power, steady state (Middle of Life Conditions), equilibrium Xenon. This was done by causing the equivalent of a double ended Main Steam Line pipe guillotine rupture in

                                                                                                                       )

Containment. All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.3 were recorded for a period l of 7 minutes, which is sufficient time for Pressurizer Pressure and Level to begin recovery and I Containment Temperature and Pressure to have peaked. baseline data for comparison is from CEN 128 ) Case A4A, CEN 268 Case 5.3 and some Best Estimate Analysis. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.3. Following the original transient #8, it was identified that both CTMT pressure and temperature response was unsat. This response was identified as unacceptable, work was done on the simulator model and the transient was run again. The results of this run were, all acceptance criteria were met, no deficiencies were identified. O Rev: 1 Date: 5/27/97 Page: 3 of 4

i a i Transient #10 - Slow Primary System Depressurization to Saturated Condition using Pressurizer Relief i OValve Stuck Open (HPSI inhibited). As required by ANSI /ANS 3.5 (1985) Appendix B, a PORV was stuck fully open and the RCS was allowed to depressurize with all High Pressure Safety Injection Disabled. This event was initiated from 100% power, steady state (Middle of Life conditions), equilibrium Xenon. All parameters, listed in 4 ANSI /ANS 3.5 91985) Appendix B.2.2.4 were recorded for a period of 11 minutes which was sufficient

time to allow the RCS to reach saturation. With regard to Appendix B.2.2.4, relief valve Gow was not available for recording and Reactor Vessel Level and Saturation Margin were available but were recorded

at 5 second intervals versus .5 second intervals. baseline data for comparison is from a i NU RETRAN  :

l analytical case specific to Millstone 2, CEN 268, Case 5.1.2, and some Best Estimate analysis. Graphical i comparisons were made for each of the monitored parameters from Appendix B.2.2.4 (except relief valve i flow). All acceptance criteria were met, no deficiencies were identified. j i s 1 i 5  ! i O i i 4 l I J l i 1 Rev: 1 Date: 5/27/97 l Page: 4 of 4

  . . . . - ~ . . . - . . . - - - . -
                                             ....- - - - - ~ . _ -- . . . - - - - . _ . . - . - - ~ . - . . . - - = _ .

4 k 4 i i l 4 ATTACIIMENT 9 e l PHYSICAL FIDELITY

SUMMARY

REPORT i 1 4 4 3 1 1 i i i i i 1 1 O + I 2 d 1 1 i 1 i i 4 1 4 I O

Pags 1 of 2 FORM 7.1 EXCEPTIONS - CONTROL ROOM LAYOUT UNIT: ._2_ l

1. During the 1994 refueling outage, the physical appearance of the real control room was changed. An STA office was added, the Shift Managers office was enlarged, the PPO/SPO/US desks were separated then relocated. Many of the storage locations for both procedures, prints, etc. were changed. The size of the simulator room will not allow all these changes to be made.  !

The following is a listing of the major differences between the real control room and the simulator together with a statement as to any effect on training.

            . The Shift Managers office is not modeled. (no effect on training)
            . The STAS office is not modeled. (no effect on training)
            . In the real control room, there are many more procedures (Chemistry, Health Physics, l&C, Maintenance, etc.) which are frequently referenced during daily operation. Dunng            l typical simulator training, these procedures are not necessary so are not stored and             '

maintained on the simulator. (no effect on training) 1 In the real control room, the phone system has been replaced with a newer and more j modern system. It is the desire of the MP2 OPS MGR that differences be minimized. ' An SDC has been generated to replace the simulator phone system.

            . In the real control room, the operating procedures are stored in a different type of filing cabinet. On the simulator the location of these cabinets are also different. (no impact          i on training)                                                                                     l

2. The emergency plan communications consoles (radio pager), Tech Support Center  !

(TSC) Phone, Waterford Police Phone, Operational Support Center (OSC) Phone, Berlin l Phone, Emergency Operations Facility (EOF) Phone and NRC red phone are not ) present on the Simulator. Push buttons are provided on the simulator control room desk phone console for EOF and NRC. Since all communications from the operators would be to a limited number of simulator instructors in the instructor booth anyway, there is DQ  ; significant training imoact on whether they communicate thru the phone console at the 1 Simulator Operator Desk versus the real EOF and NRC phones in the refarence plant control room. The simulator is not used to provide training on the radiopager. This is l not a licensed operator task. When E-Plan drills are run on the simulator, additional l phones are installed to allow the Shift Manager, On Site Duty Officer and other drill participants a chance to deal with other members of the Station Emergency Response Organization (SERO). i Completed by: b oo m u. . n--- Date: M i fC Reviewed by: .ASOT ' 8// Date:

                                                                       '     /

M Rev: 4 Date: 5/28/97 NSEM-4.12 Page: 7.1-1 of 2

' l P:gs 2 of 2 i FORM 7.1 l l EXCEPTIONS - CONTROL ROOM LAYOUT UNIT: ._2_

3. In the real control room, there are three (3) laserjet printers. Two are used to print alarms and periodic reports while the third is used to print when requested by various control room personnel computers (SM, US, STA and PPO/SPO).

The first two are modeled on the simulator while the third is not.

4. The location and size of the key locker is different between the real control room and the i simulator. The number of keys required to operate the simulator is quite small when compared to the real plant. To improve realism, the simulator is using the same index and key numbering system as the plant. (no effect on training)

5. The chairs located in the real control room are of a slightly different design and color as those on the simulator. (no effect on training)

6. The P&lD table in the real control room has 1/4" thick sheet of plexiglass used to j protect the various drawing indexes. Laminated drawing indexes are available on the -

simulator. (no effect on training) I

7. On the simulator, Fire Protection Panel (C26) is not modeled. In the real control room, this panel provides alarm indication only and all subsequent actions are taken in the  !'

plant. No control room training can be received with this panel. (no effect on training) l Completed by: D h ( d . h 'T Pate: [G [Tl '

                                                   /A /n                                 ,

Reviewed by: "

                                             >         /          Date: m             /7/9 7 ASOT                                      @             '

4 O Rev: 4 6 Date: 5/28/97 NSEM-4.12 Page: 7.1-2 of 2

Page 1 of1 FORM 7.2 EXCEPTIONS - PANEL LAYOUT UNIT: _2.

1. C-08 in reference plant has Electrical Protection Relaying, Simulator does not, except for Main Generator 86/87 relay resets. This has no sianificant training imoact.

2. C-04 Rear, C-03 Rear and C-02 Rear in reference plant has bistables, simulator does not. The bistables serve no direct operator control purpose and therefore have no kainino imoact.

3. Recorders FR 150,160,170 and 180 on C-05 Rear in the refuence plant are not  !

simulated. These recorders display RCP Parameters for trend purposes. These RCP points are available on the simulator, as in the reference plant, on C-04 Rear. The only thing lost is long term trending by not having these recorders, but simulator sessions would not last long enough to make the recorders of use to the operators in the simulator. There is no trainina imoact. j

4. Tl 6897 on C-01 Rear in reference plant not simulated. Lack of Reactor Vessel Support I Leg temperatures in Containment has no significant imoact on training.

5. The reference plant has an oscillograph next to RC-14, the simulator does not. The  ;

operators do not use the osci!lograph for any control or monitoring function, therefore  ; there is no trainina imoact.

6. On C-05R, the reference plant has an "EDAN Flow Monitor" which consists of a switch i and fuse, the simulator does not. Operators are not trained on this switch. There is no O training imoact.

j

7. On C-05R, the reference plant has the CEAPDS computer above the CEAPDS 1/O I chassis, at the simulator these are reversed. These units are not manipulated by the operators, therefore there is not training imoact.

8. In the real control room, modifications to the ESAS sensor and actuation cabinets were accomplished in 1992. Many of these hardware changes were installed on the simulator in an effort to maintain fidelity. The completion of the SDC was put "on-hold" due to the potential of replacing the sensor and actuation cabinets. There are currently minor differences in appearance between the real control room and the simulator. This has minimal training impact, i

I 1 l Completed by: . W ate: G b ~l

                                                                                                             )

Reviewed by: ASOT X Date: d' '/ ' 7 I Rev: 4 1 Date: 5/28/97 NSEM-4.12 Page: 7.2-1 of 1

         -       .   .      .-     - . _ - . ~.           .    .-       -     ..                 -     .

A

,                                                                                          Pcgs 1 of 3 EXCEPTIONS - COMPONENTS
    . UNIT: _2_

4

1. The recorders listed below are GMAC recorders in the reference plant but in the simulator are TRACOR recorders made to appear similar to GMAC recorders since they have a clear cover. This was done because GMAC recorders are practically unattainable. The GMAC reference plant recorders have pointers which point up, the simulator Tracor recorders have pointers which point down. Scales and chart paper are identical. Recorders affected are:

C-02 RR202 C-04 FRC210X C-05 PR4215 FRC210Y LR5282 JR011  ; C-06/7 TR7030 JR009 UR4501 The below listed recorders are YOKOGAWA recorders in the plant but are currently TRACOR on the simulator. (These are scheduled to be replaced on the simulator) i C-05 URS265 C-Q_4 UR7660 i TR 111 TR 121 Having different recorder produces no adverse training impact. 3 2. Recorders RJR 9129,9373 and AJR 7837 on RC-14 are different in reference plant versus simulator. Difference is different model of Leeds and Northrup recorder. The plant recorders have print wheels, the simulator recorders have thermal paper. There is no training imoact.

3. On reference plant panel C-02, controllers PIC 201, TIC 223 and PlC 215 have a green

stripe over the setpoint band, the simulator does not have the green stripe. The green

stripe has no special meaning, therefore there is no training imoact.

1

i

, i l a l Completed by: bLLO3% - ate: b 1/ T 7 l Reviewed by: ASOT Date: #/

                                                                       /

M / l 4 Rev: 5 Date: 5/28/97 NSEM-4.12 Page: 7.3-1 of 3

.__ _ ._.._. _ . . _ . _ _ _ _ . _ . _ . - _ _ . _ ._. _ ,_. m . _ . _ _ . _ . - _ . FORM 7.3 l i EXCEPTIONS - COMPONENTS i

              . UNIT: _2.

4. On Control Panel C-04 at the reference plant, the backup scanner digital CEA position indicator and selector switches are slightly different than the simulator. The actual '

digital display numbers have a different degree of " sharpness" versus the reference  ; plant. The selector switches are thumo' wheel type on the simulator versus dials in the reference plant. They are functionally identical and therefore there is no training imoact.

5. On Control Panel C-05, the "B" Condensate Pump Ammeter is different between the i simulator and reference plant. The difference is slight, and since there is no opportunity for an operator to misread the scale there is no training imoact.

6. The reference plant control panel C-21, " Hot Shutdown Panel", located in West 480 Volt Switchgear room, has several meters which are " SIGMA" meters, where as the <

Simulator C-21 panel, located in a connecting room off the simulator has G-E style meters The meters are the same size and scale. These meters are TI-115 TI-125, P1103, P1103-1, LI 110X and LI 110Y. .There is no trainina imoact since the information obtained from the meters is identical.  !

7. There are some holes in the reference plant C-21 panel but not in the simulator C-21 panel. This has no training imoact.

i

8. On reference plant control panel C-06/7, the EHC Insert has the "G-E" nameplate, the f simulator does not. This has no training imoact. .

9. On reference plant control panel RC-14, recorders RR8123A/B and RR 8262 A/B are Esterline Angus recorders, on simulator they are TRACOR recorders. The recorder scales and chart paper are otherwise identical. The make of the recorder has no ,

training imoact. i I l Completed by: ktaCo w . er Date: il /O Reviewed by: Date: #/ 7

                                                                                          /

7 ASOT ' Rev: 5 i Date: 5/28/97  ! NSEM-4.12 Page: 7.3-2 of 3

Pcg3 3 of 3 FORM 7.3 EXCEPTIONS - COMPONENTS I UNIT: _2. I l 10. The reference plant has 22 " SIGMA" style meters on C-03 and C-05, but the simulator

has "G-E" style meters. Same ranges on simulator and reference plant, just different l l style meter. Affected meters are

.

C-03: PI-103, PI-103-1. TI-112CA, TI-112CC, TI-122CA, TI-122CC, TI-112HA, l TI-112HB, TI-112HC, TI-112HD, TI-122HA, TI-122HB, TI-122HC, TI-122HD, TI-111Y,  ; TI-121Y, TI-112CB, TI-112CD, TI-122CB, TI-122CD C-05: Li-5271, LI-5273. . There is no training imoact since the information obtained from the meters is identical. l l 11. The reference plant has calibration stickers on various meters, controllers, etc.; the l simulator does not. Calibration stickers have no imoact to trainina.

12. The 2-SI-652 manual disconnect switch in the reference plant is a different style than in the simulator. The switch positions are the same.' There is no imoact to trainina. i

13. On Control Panel C-03 at the reference plant, the meters for RCS pressure and TM LP setpoint Channels A, B, C, and D have both pointers colored red. At the simulator the RCS pressure meter has a red pointer and the TMLP setpoint pointer is colored yellow. i These meters are PI-102A, PI-102B, PI-102C and PI-102D. ,

l i l' There is no training imoact due to this difference in pointer colors. l O i i i Completed by: . . Date: [O/97 Reviewed by: Date: 7 F7 ASOT ' ' l i l O  ! i i Rev: 5 Date: 5/28/97 NSEM-4.12 Page: 7.3-3 of 3

P gs 1 of f FORM 7.4 EXCEPTIONS - AMBIENT ENVIRONMENT IJN;T: _2.

1. In the reference plant there is a steady background noise of Control Room Air Conditioning (CRAC). Also when changing CRAC equipment in the reference plant there is a change in background noise. At the simulator, there is no CRAC background noise and therefore no audio cue of deliberate or non-deliberate changes to CRAC.

There is no significant training imoact.

2. In the reference plant, a page may be performed by dialing at the operator's desk i phone, in the simulator a pushbutton is used for paging at the operator's desk phone.  ;

i This has no training imoad. I

3. There is no radiopager, NRC, Waterford Police, Operational Support Center (OSC), 1

Berlin, EOF and TSC phones on the simulator. This is discussed on Form 7.1, item #2. 1 This causes no significant training imoad. I i

I t i ! i l l i i O l l l l Completed by: -i- - W - M ate; b[C7 [ P i i n Reviewed by: ASDT Date: //7/f7 ' O Rev: 3 Date: 5/28/97 NSEM-4.12 Page: 7.4-1 of 1 l l

l P:ga 1 of 1 j l i ATTACHMENT 8.2 l n MARGINAL NOTE DIRECTORY  ! U 1997 ) SIMULATOR PHYSICAL FIDEllTY , REPORT CHANGES l l i

1. This Physical Fidelity report is a modification of the report submitted in June 1993. It is j current with respect to the simulator and the control room.

l l 1 l 1 l ( l 1 l l l l l l Rev: 0 Date: 5/28/97 NSEM-4.12 Page: 8.2-1 of 1 l

P t t i ATTACllMENT 10 SAMPLE MALFUNCTION TEST PROCEDURES I i i h 4 i l l , l l ! I I r I I t t O , i i O

  ~

l .

          }               RC04 Ubisolable RCS Leak on the Rx Vessel Head Vent (Generic or Variable)

STEP # PROCEDURE / RESULT PANEL TAG # ACCEPT /DR 12.0 RC04: Unisolable RCS leak on the Rx vessel head vent. Range: 0-100%, 100% = 400 gpm at 2250 psid. Establish a Trend Group (RC04) which contains the following points and will trend for 8 minutes: PZR Pressure P102B1 PZR Lever L110X CTMT temperature T8096

       ~'

CTMT pressure P8113 J CTMT sump level L9155 1 l' , N

l. PPC points to be verifj ed: none 12.1 Reset simulator to 100%.

12.2 Insert mal RC04 at 100% severity to actuate at 2 seconds then star +. the recording program. 12.3 Place simulator in run. 12.4 When RC04 actuates, verify: 12.4.1 Pzr level decreases which cause:

• L/D flow to be reduced PPC F212
• Both first and second back up charging pumps to start.
• CTMT sump level to rise ladlA7$

Rev.: 2 Date: 3/12/97 Page: 8.2-48 of 91 i l

         ~

i i I

 ;      STEP #  PROCEDURE / RESULT                    PANEL   TAG #

o ACCEPT /DR ( v) 12.4.2 With maximum charging and minimum PPC L110X L/D, Pzr level will continue to lower at approximately 5%/ min. LdATS 12.4.3 Pzr pressure will decrease and TM/LP trip will occur in approximately 6 minutes. I GELD $ , 12.4.4 VCT level decreases. 12.4.5 RCS leak rate program shows PPC approximately 400 gpm leak. Lukts 12.4.6 CTMT conditions indicate RCS l leakage: j

• CTMT normal sump PPC L9155
• CTMT temperatures {

PPC T8096 l

• CTMT pressure PPC P8117

   /~~T

• CTMT gas / particulate RAD monitors PPC R8123A

(_,/ (will not show any radiation R8123B effects for approximately 10 minutes) 12.4.7 The following alarms will actuate when their setpoints are exceeded: Pzr Press deviation CO2/3 D37 Pzr level X (Y) Hi/ low CO2/3 A38(39)

• SIAS Actuation CH1(2) Col A(B)34 htk!6 12.5 After the recording program stops, place simulator in freeze.

Exit Data recording program. [

   \/                                                         Rev.: 2 Date: 3/12/97 Page: 8.2-49 of 91 NSEM-4.05

l

 ;p STEP #  PROCEDURE / RESULT                            PANEL    TAG #

t ACCEPT /DR x--) Plot the transient as follows: l

• On any Xterm window, type-mp2 plot _rco4, then Hit return
• On that CRT, the "WINGZ" program will come up. Sel,ect All when prompted l
• In approximately 5 minutes the l plots will be available in the SCE office. I l

l 12.6 Reset the simulator to 100%, then reset the leak rate program. 12.7 Insert RC04 at 50% severity, parameter recording is not required. Run transient for 4 minutes. 12.8 When RC04 actuates, verify: (^'% x# 12.8.1 RCS leak rate approximately 200 PPC gpm.

• Actual leak rate 2'47 gpm (M6 l

12.8.2 RCS pressure decreases but TM\LP PPC P102B1 I trip does not occur.

• Minimum RCS pressure 22cy4 psia btdU$

12.8.3 Pressurizer level decreases at approximately 1% per minute.

• Minimum Pzr level S'(.\ % bChh5 12.8.4 RCS pressure will be maintained above 2200 psia by pzr heaters.

12.9 Reset simulator to 100%, then reset j the RCS leakrate. C)

  %/

Rev.: 2 l Date: 3/12/97 Page: 3.2-50 of 91 NSEM-4.05 i t

STEP # PROCEDURE / RESULT PANEL TAG # ACCEPT /DR

  /%

O 12.10 Insert RC04 at 10% severity, parameter recording is not required. Run transient for i approximately 4 minutes 12.11 At completion of 4 minute run, verify 12.11.1 RCS leak rate approximately 40 l gpm.

• Actual leak rate 4\ gpm PPC M8 12.11.2 RCS pressure maintained 2225 psia-2275 psia.
• Minimum RCS pressure 72c2 psia PPC P102B1 M 12.11.? eressurizer level maintained by cycling of first back up CHG pump.

[]

• Minimum pressurizer level G5.7 %

                ,wTc    T$(o c etc p-sQ cM uoT PPC    L110X      lEik 12.12     End of RC04 Test. S c- m ted&M i

Performed by: beunu.M . T Date: "! 7N7 o Rev.: 2 Date: 3/12/97 Page: 8.2-51 of 91 NSEM-4.05 I l j

e- , . i

                           .CC11A (B and C) RBCCW Pump Degraded Performance Variable l

l STEP # PROCEDURE / RESULT PANEL TAG # ACCEPT /DR 1.0 Reset to 100% power. PCM M l-l 2.0 start or verify running the RBCCW pump.

                                                        'A'                         M Actual pressure LLB         psig    PPC       P6008-Actual flow        (oc6"F- GPM      PPC       F6035       M 3.0     Insert malfunction CC11A at a 20%      PCM       CC11        M         '

severity and observe discharge pressure and flow decrease: I r Pressure ~ 108 psig PPC P6008 l Flow ~ 5600 GPM PPC F6035 d Actual pressure- LoS psig PPC P6008 Actual flow 9,,o*4 GPM PPC F6035 M 4.0 Increase severity to 40% and observe: Pressure ~ 96 psig l Flow ~ 5075 GPM i l Actual pressure O psig PPC P6008 ! Actual flow  %"~l '{ GPM PPC F6035 W l l I 4 1 Rev.: 1 Date: 3/27/97 Page: 8.2-1 of 2 NSEM-4.05

STEP # PROCEDURE / RESULT PANEL TAG # ACCEPT /DR  ; 'r's i 5.0 Increase severity to 100% and Gxf(75 i observe:  ! Pressure - 60.8 psig  ; Flow ~ 2660 GPM l Actual pressure Gi psig PPC P6008 } Actual flow 2( Gt GPM PPC F6035 @ l i 6.0 Remove malfunction CC11 and observe

                   'A'  RBCCW pump parameters return to normal.

i Pressure and flow as indicated in 1tep 2.0 7.0 Reset to 100% power and complete the remaining CC11 malfunctions on the remaining RBCCW pumps. The s

 ,s               effects will be similar to those                                           !

[

        )         seen in CC11A with the following exceptions:

CC11B affects the "B" RBCCW pump C11B PPC P6009 PPC F6035 CC11C affects the 'C' RBCCW pump CC11C PPC P6010 PPC F6034 8.0 End of CC 11 test. Performed by: h. Date: YM*t7

 ,,)  ,                                                          Rev.:    1 Date:    3/27/97
                                                                 . age:   8.2-2 of 2 NSEM-4.05

1 1 l 0 ^rr^c"ueur ii L.lST OF CERTIFIED REMOTE FUNCTIONS l i 1 1 l l i O O Cl.RTR) L W}ts

FIGURE 7.1 { CERTIFIED REMOTE FUNCTIONS LIST 5 LCul 2 PAGE 1 OF 10 3 CCR01 RBCCW SURGE TK FILL BYPASS 2-RB-53  ! 4 CCR02 TIC-6308 IlX-18A TO TV-6308,6380A(DEGF) l CCR03 TIC-6307 liX-18B TO TV-6307,6307A(DEGF) CCR04 TIC-6306 liX-18C TO TV-6306,6306A(DEGF) i I CCR05 RBCCW SURGE TK DRAINS 2-RB-54A/54B  : j CCR06 ISOL VLV 2-RB-3 A DOWNSTREAM P-11 A(%)  ! , CCR07 ISOL VLV 2-RB-3A DOWNSTREAM P-11B(%)  ! CCR08 ISOL VLV 2-RB-3C DOWNSTREAM P-11C(%)

CCR10 l TliROTTLE SFP CLG FLOW VLV 2-RB-8B(%)

j CCRll TilRO1TLE SFP CLG FLOW VLV 2-RB-8A(%)  ! CCR12 SFP llEAT LOAD (E6 BTU /IIR)  ; CCR13 SFP PUMP A START l 4 CCR14 SFP PUMP B START i CCR15 SFP IfX-20A VLV 2-RW-6A > i CCR16 SFP llX-20B VLV 2-RW-6B I ? CCR17 EVAP IIEAT LOAD (ON/OFF)

CCR18 2-RB-66 SURGE TANK MAKEUP ISOLATION VALVE

I CCR31 IIPSI PUMP B llDR SEAL COOLING SUPPLY J

CCR32 R-RB-14A SD HX A OUTLET (%) '

CCR33 R-RB-14B SD llX B OUTLET (%)  ; CCR34 CC PMP B BREAKER STATUS - i CCR35 RBCCW PMP A RAD &RECIRC ISOL 2RB43&l07A i CCR36 RBCCW PMP B RAD &RECIRC ISOL 2RB41&l07B ' ! CCR37 RBCCW PMP C RAD &RECIRC ISOL 2RB39&l07C l CCR38 RBCCW PMP A BREAKER 1 CCR39 RBCCW PMP C BREAKER i CIIR01 OUTSIDE AIR TEMPERATURE VARIATION (DEGF)  : CllR02 112 PURGE AIR SOURCE 2-IA-27.2 ' CilRC, CNTMT PURGE VALVES L,0CK OUT(AC-4,5,6&7) l CllR05 RESET ClilLLERS 169A/B,196A/B,170 CVR01 C11G PMPS OUTLET TO IIPSI 2-Cll-440,340 , CVR02 LETDOWN FLOWPATil TilRU DEBORATION IX CVR03 BORIC ACID PUMP DISCIIARGE VALVES (2-CII-152/153) CVR05 B.A. FROM BATCil TK TO TK 8A 2-Cll-124  ; CVR06 B.A. FROM BATCli TK TO TK 8B 2-Cll-135 1 CVR07 VCT LETDOWN ISL 2-Cil-397 CVR08 CilARGING PUMP, P-18A, RACKOUT

            ' CVR09                                     CilARGING PUMP, P-18B, RACKOUT CVR10                                     CllARGING PUMP, P-18C, RACKOUT CVR11                                     CilARGING PUMP, P-18B, KIRK KEY CVR12                                     CliARG PUMP P-18A DIS IIDR ISOL (2-Cil-338)

CVR13 VCT TO WASTEGAS IIEADER (2-Cll-102) CVR14 BORIC ACID MAKEUP ISOLATION (2-Cll-172) CVR15 2-Cll-201P ISOLATION VALVE 2-Cll-350 CVR16 2-Cli-201Q ISOLATION VALVE 2-CH-349 CVR17 PSI IIDR TO CVCS PURIF LINE 2-Cit-603 ASOT # igiiature/Date S Y7

                                                                          /   '     /

O cmmns NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-1 of 10 .

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 2 OF 10 CVR18 PURIF LINE TO LPSI SUCTION 2-Cll-024 CVR20 2-Cll-110P ISOLATION VALVE 2-Cil-342 CVR21 2-Clli 10Q ISOLATION VALVE 2-Cll-344 CVR22 PMW TRANSFER PUMP P-22B CVR23 PMW TRANSFER PUMP P-22A CVR21 2-Cil-210Y BORIC ACID MAKEUP VLV LEAKAGE CVR25 B.A. TANKS SUCTION CROSS-TIE 2-Cil-144 l CVR26 B.A. TANK 8A TO BA PUMP SUCTION 2-CH-131 CVR27 B.A. TANK 8B TO BA PUMP SUCTION 2-CH-142 CVR28 DEAERATOR WATER TRANSFER PUMP CVR29 BORIC ACID TANK 8A CONC CIIANGE (% WT) CVR30 BORIC ACID TANK 8B CONC CilANGE(% WT) CVR31 N2 TO VCT 2-CH-109 CVR32 112 TO VCT 2-CH-107 l CVR33 PURIF IX OUTLET 2-CH-378 CVR34 CIIARGING PUMP A SUCTION VALVE 2-CH-316 CVR35 CHARGING PUMP B SUCTION VALVE 2-CH-319 CVR36 CllARGING PUMP C SUCTION VALVE 2-CH-322 CVR37 PMW STOP VALVE 2-Cll-195 CVR38 BA GRAVITY FEED ISOLATION 2-Cll-130 (TANK A) i CVR39 BA GRAVITY FEED ISOLATION 2-CH-140 (TANK B)  ! CWR01 SEA WATER TEMPERATURE (DEG F) CWR02 WATER BOX A PRIME & VENT 2-VP-1 A,14A

 ^- CWR03           WATER BOX B PRIME & VENT 2-VP-1B,14B CWR04           WATER BOX C PRIME & VENT 2-VP-lC,14C CWR05           WATER BOX D PRIME & VENT 2-VP-lD,14D CWR06           SCREENWASil PUMP P-8A CWR07           SCREENWASH PUMP P-8B CWR08           CIR WATER PUMP A BREAKER                                             !

CWR09 CIR WATER PUMP C BREAKER CWRIO CIR WATER PUMP B BREAKER CWRll CIR WATER PUMP D BREAKER EDR01 DISCONNECT SWITCH (1SG-22S1-4) EDR02 BKR A505 IN TEST & INTERLOCKS DEFEATED EDR03 NSST LOCKOUT DEVICE l EDR04 RSST LOCKOUT DEVICE EDR05 STATIC TRANSFER SWITCH VS1 EDR06 STATIC TRANSFER SWITCH VS2 EDR07 STATIC TRANSFER SWITCH VS3 EDR08 STATIC TRANSFER SWITCil VS4 EDR09 IAC 1(VRll) TilROW OVER SWITCil(RSI) EDR10 I AC 2(VR21) THROW OVER SWITCil(RS2) EDRll DISCllARGE RATE FOR BATT dbl (X NORMAL) EDR12 DISCHARGE RATE FOR BATT DB2(X NORMAL) EDR13 DISCHARGE RATE FOR BATT DB3(X NORMAL) EDR14 SUPPLY BREAKER FOR 24E FROM 24C (A305) EDR15 SUPPLY BREAKER FOR 24E FROM 24D (A408) EDR16 BATF CilGR 201 A AC/DC BKRS EDR17 BATT CilGR 201B AC/DC BKRS EDR18 BATT CilGR 201C AC/DC BKRS U CIRTRFL WHS NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-2 of 10

, FIGURE 7.1 4 a CERTIFIED REMOTE FUNCTIONS LIST A UNIT 2 PAGE OF U 3 10 EDR19 D.C. BREAKER (D0103) CONTROL EDR20 D.C. BREAKER (D0104) CONTROL EDR21 D.C. BREAKER (D0204) CONTROL EDR22 D.C. BREAKER (D0203) CONTROL EDR23 345 KV SWITCil BREAKERS CONTROL EDR24 GRID VOLTAGE VARIANCE (KV) EDR25 GRID FREQUENCY VARIANCE (llZ) EDR26 MOTOR OVERCURRENT RELAY RESET EDR27 6.9 KV NSST(ll101) BREAKER Rl/RO EDR28 6.9 KV NSST(ll286) BREAKER Rl/RO EDR29 6.9 KV RSST(H103) BREAKER Rl/RO EDR30 6.9 KV RSST(il204) BREAKER Rl/RO EDR31 ENERGlZE VRll FROM (B32) MCC-2C EDR32 ENERGlZE VR21 FROM (B41 A) MCC-lDA EDR33 ALL BREAKERS ON 25 A (6.9 KV) EDR34 ALL BREAKERS ON 25B (6.9 KV) EDR35 ALL BREAKERS ON 24 A (4.16 KV) EDR36 ALL BREAKERS ON 24B (4.16 KV) EDR37 ALL BREAKERS ON 24C (4.16 KV) EDR38 ALL BREAKERS ON 24D (4.16 KV) EDR39 ALL BREAKERS ON 24E (4.16 KV) EDR40 NSST TO 24A (A102) BREAKER RI/RO EDR41 NSST TO 24B (A206) BREAKER Rl/RO p EDR42 480V TO 22A (A103) BREAKER Rl/RO Q EDR43 EDR44 480V TO 22C (A104) BREAKER Rl/RO 480V TO 22B (A204) BREAKER Rl/RO EDR45 480V TO 22D(A205) BREAKER Rl/RO EDR46 24E TO 24F (A505) BREAKER Rl/RO i EDR47 UNIT 1 TO 24E (A602) BREAKER RI/RO l EDR48 RSST TO 24D (A411) BREAKER Rl/RO i EDR49 RSST TO 24C (A302) BR.EAKER Rl/RO , EDR50 480V TO 22E (A303) BREAKER Rl/P,0  ! EDR51 480V TO 22F (A409) BREAKER RI/RO EDR52 24C TO 24A (A304) BREAKER Rl/RO EDR53 24D TO 24B (A410) BREAKER Rl/RO EDR54 24G TO 24C/D ( A702) BREAKER Rl/RO EDR55 PPC POWER EDRS6 APPENDIX R INTERLOCK DEFECT SWITCil EDR57 UNIT 1 BREAKER (A602) CONTROL EDR58 UNIT 1 RSST TO 1411 (A603) CONTROL EDR59 UNIT 1 LOAD TilRU A601 (AMP) EDR60 141-1/14C,D (A601) CONTROL EDR61 UNIT 1 NSST/DG POWER EDR62 UNIT 1 RSST POWER EGR01 MOTOR OPERATED DISCONN SWITCll(15G-2XI-4) EGR02 EMERG SEAL OIL PUMP (Rl/RO) EGR05 MAIN GEN BREAKER ST OPERATION SELECTED EGR06 MAIN GEN BREAKER 9T OPERATION SELECTED EGR07 DG 12U MECilANICAL RESET SWITCll EGROS DG 13U MECllANICAL RESET SWITCil a amt wus NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-3 of 10

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST

UNIT 2 PAGE 4 OF 10 EGR09 EMERG SO PUMP TEST (11-33,11-34)

EGR10 MAIN GEN BKR 8T/9T GETAC CONTROL EGRII MAIN GEN LOCKOUT (86GCDl,86GCD2) EGR12 DIESEL GEN A STARTING AIR ISOLATION t EGR13 DIESEL GEN B STARTING AIR ISOLATION EGR14 DIESEL GEN A MANUAL TRIP EGR15 DIESEL GEN B MANUAL TRIP EGR16 DIESEL GEN A/B TROUBLE / ACKNOWLEDGE EGR17 DIESEL GEN A BREAKER (A312) Rl/RO EGR18 DIESEL GEN B BREAKER (A401) Rl/RO EGR19 F-38A LOCAL CONTROL EGR20 F-38B LOCAL CONTROL ESR01 8A BLOCK FUSE STATUS ON ACT CAB 5 ESR02 8A BLOCK FUSE STATUS ON ACT CAB 6 ESR03 8A SIAS FUSE ON ACT CAB 5 ESR04 8A SIAS FUSE ON ACT CAB 6 FWR01 SJAEA STM AND COND INLET AR2A & MS59A FWR02 SJAEB STM AND COND INLET & MS59B FWR03 COND PUMP P2A DISC 11 VLV 2-CN-4A(%) FWR04 COND PUMP P2B DISCil VLV 2-CN-4B(%) FWR05 COND PUMP P2C DISC 11 VLV 2-CN-4C(%) FWR06 COND PUMP T CW DISCII 2-CN-503(%) FWR07 LP llEATERS BYPASS VLV 2-CN-15(%) q V FWR08 FWR09 DRAINS COOLER "A" INLET 2-CN-11 A(%) DRAINS COOLER "B" INLET 2-CN-11 B(%) FWRIO MFWP'S BYP VLV 2-CN-51(%) FWRll ifP llTR l A,1B BYP VLV 2-FW-6(%) FWR12 IIP llTR 1 A IN& OUTLET ISOL 2-FW-2A,3A(%) FWR13 11P liTR 1B IN& OUTLET ISOL 2-FW-2B,3B(%) FWR14 MAKEUP TO TDAFP P4 2-CN-30,2-FIRE-94C FWR15 MAKEUP TO AFW P9A 2-CN-29A,2-FIRE-94A FWR16 MAKEUP TO AFW P98 2-CN-29B,2-FIRE-948 l FWR17 #1 SG AUX FRV BYP2-FW-56A (%) . FWR18 #2 SG AUX FRV BYP2-FW-56B (%) I FWR19 VACUUM PUMP F5A SUCTION ISOL 2-AR-12A FWR20 VACUUM PUMP FSB SUCTION ISOL 2-AR-12B FWR21 liOT WELL REJECT LCV ISOL 2-CN-21 FWR22 AUX FEED SV4188 LOCAL OPERATION FWR23 AUX FEED GOVERNOR VALVE POS (%) FWR24 DRN CLR A INLET EQUALIZER 2-CN-214A(%) j FWR25 DRN CLR B INLET EQUALIZER 2-CN-214B(%) FWR26 IITR DRAIN PUMP P3A DISCilISOL 2-IID-91(%) FWR27 IITR DRAIN PUMP P3B DISCH ISOL 2-IID-9B(%) , FWR28 liTR DRN PUMP P3A RECIRC STOP 2-IID-45A l FWR29 lITR DRN PUMP P3B RECIRC STOP 2-IID-45B FWR30 COND TRANSFER PUMP P71 FWR31 COND SRG TK FILL FROM COND TRANS CN-206 FWR32 COND FILL FROM COND TRANS 2-CN-223 . FWR36 MECilANICAL VACUUM PUMP SET POINT (INilG) l FWR37 LONG RE-CYCLE VALVE AOV20(%) O V FWR38 COND DEMINS 1 A,1B,1C,lD,lE,lF,1G IN SERV FWR39 CST / SURGE TK M/U TO llW 2-CN-100,34 , Cf.RTRFL WH5 NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-4 of 10

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST O UNIT 2 PAGE 5 OF 10 FWR41 FW REG VLV 2FW51 A NORMAL / HAND CONTROL  ! FWR42 FW REG VLV 2FW51 A MANUAL VLV POSITION FWR43 FW REG VLV 2FW51 B NORMAL /11AND CONTROL 1 FWR44 FW REG VLV 2FW51B MANUAL VLV POSITION FWR45 FRV 2FW51 A LEAKAGE (VLV POS %) FWR46 FRV 2FW51B LEAKAGE (VLV POS %) FWR47 FPT 1 A AUX Olt PUMP TEST PB(HS-7185) FWR48 FPT I A EMERG OIL PUMP TEST PB(llS-7186) FWR49 FPT IB AUX OIL PUMP TEST PB(IIS-7135) FWR50 FPT IB EMERG OIL PUMP TEST PB(IIS-7136) FWR51 COND l A EVAC SUCT VLV 2ARI A  % FWR52 COND l A EVAC SUCT VLV 2ARlB ' FWR53 COND IB EVAC SUCT VLV 2ARIC FWR54 COND IB EVAC SUCT VLV 2ARlD i FWR55 MAKEUP TO CST (GPM) i FWR56 TRAIN A LP llEATER ISOL 2-CN-12A,43A POS (%) FWR57 i TRAIN B LP llEATER ISOL 2-CN-128,43B POS (%) i FWR58 FIRE PUMP ALARM STATUS FWR59 #1 SG AUX FRV 2-FW-43A FWR60 #2 SG AUX FRV 2-FW-43B l FWR61 #1 SG AUX FW ISO VLV 2-FW-18A FWR62 #2 SG AUX FW ISO VLV 2-FW-10B FWR63 #1 SG FRV 2-FW-43A FWR64 #2 SG FRV 2-FW-43B <m FWR65 COND PUMP A BREAKER (il106) Rl/RO (') FWR66 FWR67 COND PUMP B BREAKER (11107) Rl/RO COND PUMP C BREAKER (11203) Rl/RO l FWR68 AUX FEED PMP A BREAKER (A307) Rl/RO  ! FWR69 AUX FEED PMP B BREAKER (A406) Rl/RO FWR70 IITR DRN PLIMP A BREAKER (A105) RI/RO FWR71 IITR DRN PUMP B BREAKER (A203) Rl/RO FWR72 REFUEL LOAD CENTER TRANSFER SWITCli I AR01 SA CTMT llDR ISOL 2-SA-19 IAR02 TURBINE BLDG 1 A ISOL (2-I A-25) IAR03 IA CMPR F3A SW llS-7046 IAR04 IA CMPR F3B SW 11S-7055 lAR05 IA CMPR F3A,B SEL SW llS-7046A IAR06 SA CMPR F2 SW llS-7086A 1AR07 TURB BLDG IA IIDR AIR LEAKAGE (SCFM) IAR08 ENCL & AUX BLDG 1A IIDR AIR LEAKAGE (SCFM) IAR09 CONTAINMENT IA 11DR AIR LEAKAGE (SCFM) lARIO UNIT 1 SA SPLY BV 2 sal 2,2 sal 2A IARll IA CMPR F3A RESET SW IAR12 lA CMPR F3B RESET SW IAR13 SA CMPR F2 CNROL llS-7086B 1AR14 CW PRIMING EDUCTOR LOAD (SCFM) IARl5 IA CMPR F3C START /STOP SWITCil IARl6 IA CMPR F3C RESET SWITCil IAR17 I A CMPR F3C SEL SWITCII 1 AR18 1A TO STEAM DUMP VLV I A-47 p N.] CI RTRI1 WHs NSEM-4.03 REV/ 5 DATE: 5/27/97 PAGE: 7.1-5 of 10

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 6 OF 10 IAR19 IA ALT SUPPLY TO FW VLV IA-101 IAR20 IA TO AUX BLDG VLV IA-28 IAR21 C COMPRESSOR TO RECEIVER I A-542 IAR22 I A TO SA CROSSTIE VLV I A-641 MSR01 #1 ATM STM DUMP ISOL 2-MS-3A MSR02 #2 ATM STM DUMP ISOL 2-MS-3B MSR03 MSR 1 A/IB HI LOAD VL 20-MS-85A,85B MSRO4 MSR 1 A/lB lil LOAD A/B POS(%) MSR05 MSR 1 A LOW LOAD VALVE 2-MS-79A MSR06 MSR l A LOW LOAD VALVE 2-MS-79A POS(%) MSR07 MSR IB LOW LOAD VALVE 2-MS-79B MSR08 MSR IB LOW LOAD VALVE 2-MS-79B POS(%) MSR09 AUX STEAM SUPPLY MSRll TURBINE SHELL WARMING MODE VLV STATUS MSR12 2-MS-201 CLOSING COLL /RACKOUT MSR13 2-MS-202 CLOSING COIL /RACKOUT MSR14 #1 ATM STM DUMP 2-MS-190A MSR15 #1 ATM STM DUMP 2-MS-190A POS (%) MSR16 #2 ATM STM DUMP 2-MS-190B MSR17 #2 ATM STM DUMP 2-MS-190B POS (%) l RCR01 PZR VENT VALVES 2-RC-021, & 2-RC-421 ' RCR02 PZR/RCS/CVCS BORON CONC (PPM) RCR03 PRESSURIZER BORON CONC (PPM) i l w RCR04 RCP-40A RACKOUT I RCR05 RCP-40B RACKOUT RCR06 RCP-40C RACKOUT RCR07 RCP-40D RACKOUT RCR08 PZR N2 SUPPLY 2-RC-030 & 2-RC-015 RCR10 IIOT LEG LOOP 1 LEVEL VALVE 2-RC-214 RCR11 PZR PROPORTIONAL HTR BKRS I RCR12 PZR B/U liTR BKRS l RCR13 VENT VALVE 2-EB-86 RCR14 2-RC-039,413 AND 447 RCR15 2-RC-448 RilR01 SDC SUCTION HDR MAN STOP 2-SI-709 RIIR02 A LPSI PUMP SUCT FROM SDC 2-SI-441 RHR03 B LPSI PUMP SUCT FROM SDC 2-SI-440 RHR04 A LPSI PUMP SUCT FROM RWST 2-SI-444 RilR05 B LPSI PUMP SUCT FROM RWST 2-SI-432 RIIR06 LPSI PP P-42A RECIRC VLV 2-SI-449 RIIR07 LPSI PP P-42B RECIRC VLV 2-SI-450 RHR08 CTMT SPRAY PP P-43A RECIRC VLV 2-CS-7A R!lR09 CTMT SPRAY PP P-43B RECIRC VLV 2-CS-78 RilR10 CTMT SPRAY PP P-43A DISCH VLV 2-CS-3A I RHR11 CTMT SPRAY PP P-43B DISCH VLV 2-CS-3B l RilR12 LPSI PP DISCil TO SDC llX A 2-SI-452 I RiiR13 LPSI PP DISCll TO SDC HX B 2-SI-453 l PilR14 CTMT SPRAY 11DR A VLV 2-CS-4A & 2-S1-456 R11R15

                                                                                 )

CTMT SPRAY 11DR B VLV 2-CS-4B & 2-SI-457 l RHR16 SlT RECIRC llDR STOP 2-SI-463 O L/ RHR17 RHRIS CTMT SPRAY PP P-43A RACKOUT CTMT SPR/.Y PP P-43B RACKOUT narnrtwas NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-6 of 10 j I

l FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST f] UNIT 2 PAGE 7 OF 10 kJ RHR19 LPSI PP P-42A RACKOUT RilR20 LPSI PP P-42B RACKOUT RilR21 SIT RECIRC HDR STOP 2-SI-459 RIIR22 CS HDR A TO SI TEST HDR ISO 2-CS-51 RHR23 CS HDR B TO SI TEST HDR ISO 2-CS-50 RilR24 SDC WARMUP LINE 2-SI-400 POS % l RilR25 SDC DISCHG TO CVCS 2-SI-040 RHR26 SDC DISCHG TO RWST 2-SI-460 POS % RiiR27 2-SI-306 AIR SUPPLY, FUSE BLOCK RHR28 SDC FLOW CONTROL VALVE 2-S1-657 RHR29 SDC FLOW CONTROL VALVE 2 SI-657 POS % RilR30 SIAS REMOVED FROM LPSI VALVES RHR31 CORE DELAY HEAT FACTOR RMR01 RIT-4262 ALARM DEFEAT SWITCH l RMR03 RIT-6038 ALARM DEFEAT SWITCH RMR04 RIT-7890 ALARM DEFEAT SWITCH RMR05 RIT-7891 ALARM DEFEATSWITCH 4 RMR06 RIT-7892 ALARM DEFEAT SWITCil 1 RMR07 RIT-7894 ALARM DEFEAT SWITCH RMR08 RIT-7895 ALARM DEFEAT SWITCH RMR09 RIT-7896 ALARM DEFEAT SWITCH RMR10 RIT-7897 ALARM DEFEAT SWITCll RMRll RIT-7899 ALARM DEFEAT SWITCll RMR12 RIT-80ll ALARM DEFEAT SWITCH p RMR13 RIT-8123A ALARM DEFEAT SWITCH V RMR14 RMR15 RIT-8123B ALARM DEFEAT SWITCil RIT-8132A ALARM DEFEAT SWITCll l RMR16 RIT-8132B ALARM DEFEAT SWITCH l RMR17 RIT-8139 ALARM DEFEAT SWITCH RMR18 RIT-8142 ALARM DEFEAT SWITC11 l RMR19 RIT-8145A ALARM DEFEAT SWITCH RMR20 RIT-8145B ALARM DEFEAT SWITCH RMR21 RIT-8156 ALARM DEFEAT SWITCil RMR22 RIT-8157 ALARM DEFEAT SWITCH 1 4 RMR23 RIT-8262A ALARM DEFEAT SWITCH I RMR24 RIT-8262B ALARM DEFEAT SWITCli 1 RMR25 RIT-8434A ALARM DEFEAT SWITCH RMR26 RIT-8434B ALARM DEFEAT SWITCH RMR27 RIT-8997 ALARM DEFEAT SWITCH RMR28 RIT-8998 ALARM DEFEAT SWITCH RMR29 RIT-8999 ALARM DEFEAT SWITCH I RMR30 RIT-9049 ALARM DEFEAT SWITCll I RMR31 RIT-9095 ALARM DEFEAT SWITCH RMR32 RIT-9116 ALARM DEFEAT SWITCH RMR33 RIT-9327 ALARM DEFEAT SWITCH RMR34 RI-8123A ALARM DEFEAT SWITCil RMR35 RI-8132A ALARM DEFEAT SWITCII RMR36 RI-8262A ALARM DEFEAT SWITCll RPR01 TCB-9 BREAKER crnrest wns NSEM-4.03 REV.: 5 DATE: 5/27/97 l PAGE: 7.1-7 of 10 I l

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 8 OF 10 l RPR02 MG SET A INPUT CONTACTOR RPR03 MG SET A OUTPUT BREAKER  ! RPR04 MG SET A OUTPUT CONTACTOR I RPR05 MG SET B INPUT CONTACTOR RPR06 MG SET B OUTPUT BREAKER RPR07 MG SET B OUTPUT CONTACTOR RPR12 COMPARATOR AVERAGER SELECT CHANNEL-A RPR13 COMPARATOR AVERAGER SELECT CHANNEL-B RPR14 COMPARATOR AVERAGER SELECT Cil ANNEL-C RPR15 COMPARATOR AVERAGER SELECT CilANNEL-D l RPRiG Cil-X INPUT TO POWER RATIO CALCULATOR RPR17 Cil-Y INPUT TO POWER RATIO CALCULATOR RPR18 C11 A&B HI PZR PRESS MODULE DESCONN RPR19 VARIABLE 111 POW TRIP RPR20 RCP LO SPEED TRIP , RPR21 RCS LO FLOW TRIP ' RPR22 SG #1 LO LEVEL TRIP i RPR23 SG #2 LO LEVEL TRIP RPR24 SG #1 LO PRESS TRIP RPR25 SG #2 LO PRESS TRIP I RPR26 111 PZR PRESS TRIP RPR27 TM/LP TRIP RPR28 TURBINE / REACTOR TRIP fs RPR29 111 CONTAINMENT PRESS TRIP lj RPR30 LOCAL POWER DENSITY TRIP RPR31 POWER RATIO RECORDER TENT SELECTED RPR32 ATWS TRIP DEFEAT RPR33 A CEDM MG SET BKR RPR34 B CEDM MG SET BKR RPR35 PWR RATIO Hl/LO ALARM BYPASS RXR01 LOOP 1 SEL FOR TAVG CALC IN RRS-UNIT I RXR02 LOOP 2 SEL FOR TAVG CAL.C IN RRS-UNIT 1 RX203 LOOP 1 SEL FOR TAVG CALC IN RRS-UNIT 2 RXR04 LOOP 2 SEL FOR TAVG CALC IN RRS-UNIT 2 RXR05 LO LO PZR LEVEL HEATER TRIP DEFEATED RXR06 PZR PRCP llTR BKRS IN TEST POSITION SGR01 BD TK INLETS 2-MS-381 A&B(100%=6 TURNS) SGR02 BD Q TK INLETS 2-MS-218,219(100%=6 TURNS) SIR 01 IIPSI PP P-41 A RACKOUT SIR 02 HPSI PP P-41B RACKOUT SIR 03 IIPSI PP P-41C RACKOUT SIR 04 T-39A SI-614 RACKOUT/CL COIL REMOVAL SIR 05 T-39B SI-624 RACKOUT/CL COIL REMOVAL SIR 06 T-39C SI-634 RACKOUT/CL COIL REMOVAL SIR 07 T-39D SI-644 RACKOUT/CL COLL REMOVAL SIR 08 Si TANK VENT LINE CAPS SIR 09 2-SI-654 BKR SIP 10 2-SI-656 BKR SIRll 2-SI-617/27/37/47 BKRS SIR 12 2-SI-616/26/36/46 BKRS SWR 01 SW PUMP A L)lSCHG 2-SW-2A O' SWR 02 SW PUMP B DISCHG 2-SW-2B CLRTRFL Wlts NSEM-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-8 of 10

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIS T UNIT 2 PAGE 9 OF 10 SWR 03 SW PUMP C DISCilG 2-SW-2C SWR 04 RBCCW llX A/B XCONN 2-SW-7B SWR 05 RBCCW IIX B/C XCONN 2-SW-7A SWR 06 TPCCW llX A/B XCONN 2-SW-4B ' SWP J7 TPCCW IIX B/C XCONN 2-SW-4A SLTm RBCCW HX A OUTLET VLV MODE fiW1 RBCCW HX B OUTLET VLV MODE

                                                                                            ]

Si RBCCW HX C OUTLET VLV MODE ' , SW K11 ISOL VLV 2-SW-5A TO TP llX-17A(%) SWRl2 ISOL VLV 2-SW-5B TO TP HX-17B(%) SWR 13 ISOL VLV 2-SW-5C TO TP IlX-17C(%) SWR 16 VITAL AC CLR 181 ISOL OUTLET 2-SW-181 A% SWR 17 VITAL AC CLR 182 ISOL OUTLET 2-SW-181B% , SWR 18 VITAL AC CLR 183 ISOL OUTLET 2-SW-181C% l SWR 19 "A" SW HDR TO ClilLLERS 2-SW-194 ) SWR 20 "B" SW HDR TO CHILLERS 2-SW-195 SWR 21 SW TO AC SWGR CLRS XTIE 2-SW-175 SWR 22 SW PUMP B/ STRAINER POWER SUPPLY  ; i ' SWR 23 A RBCCW llX OUTLET 2-SW-9A (%) SWR 24 B RBCCW IiX OUTLET 2-SW-9B (%) SWR 25 C RBCCW liX OUTLET 2-SW-9C (%) l SWR 26 SW PUMP "B" BREAKER STATUS SWR 27 SW PUMP "A" BREAKER l SWR 28 SW PUMP "C" BREAKER I n SWR 29 FIRE WATER SUPPLY TO A D/6 Q SWR 30 TCR01 FIRE WATER SUPPLY TO B D/6 EllC BYP VALVE POS. (FV-1)(%) TCR02 EHC OIL PUMP TEST (FV9/10) TCR03 CONDENSER LOW VAC TRIP TCR04 BRG OIL PRESS LO TRIP TCR05 SHAFT PUMP PRESS LO TRIP TCR06 LOSS OF STA' A COOLANT TRIP TCR07 EHC HYD PRLSS LO TRIP TCR08 MSR lil LVL TRIP TCR09 UNIT ELECTRICAL PROTECTION TRIP TCR10 REACTOR /TURB TRIP YCRII MANUAL EMERG PB TRIP TCR12 SG #1 til LVL TRIP TCR13 SG #2111 LVL TRIP TCR14 EXHAUST HOOD TEMP HI TRIP TCR15 PMG MALFUNCTION LIGHT RESET TPR01 TBCCW SURGE TGK SUPPLY 2-PMW-219 TPR02 TBCCW PUMP A DISCHG 2-TB-3A (%)

TPR03 TBCCW PUMP A DISCHG 2-TB-3B (%)

4 TPR04 TBCCW PUMP C DISCilG 2-TB-3C (%) TPR05 TIC-6305 IlX-17A TO VLV 2-SH-88A(F) TPR06 TIC-6304 IlX-17B TO VLV 2-SW-88B(F) TPR07 TIC-6303 IlX-17C TO VLV 2-SW-88C(F)

;    TPR08              TBCCW SURGE TK DRAIN 2-TB-211 TPR09              HYDROGEN SUPPLY VLV SETPOINT(PSIG)

TPR10 TBCCW PUMP SEAL LEAKAGE (GPH) i p V TPRI1 TPR12 ISO-PHASE BUS FAN A C5 ISO-PliASE BUS FAN B C5 CI:RTRrl WHS 'l " 1-4.03 REV.: 5 DATE: 5/27/97 PAGE: 7.1-9 of 10

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 10 OF 10

               .TPR15                         TBCCW HX A OUTLET 2-TB-5A
              -TPR16                          TBCCW HX B OUTLET 2-TB-5B TPR17                         TBCCW HX C OUTLET 2-TB-5C TPR18                         ALTERNATE WATER SUPPLY TO I A AFT / COMP TPR19                         SLO STBY PUMP TEST (Y-76,Y-45)

TPR20 ALL REMOTE PANEL ALARM RESET TUR01 'MSP AUTO START TEST (2-LO-60) TUR02 TGOP HYD OIL TEST (2-LO-55A) TUR03 EBOP OIL TEST (2-LO-103A/55B) TUR04 EMERG BEARING OIL PUMP (Rl/RO) WDR01 AERATED WASTE DRAIN TANK PUMP P31 A WDR02 EQUIP. DRN TNK SUMP PMPS P94A,B WDR03 DEGASIFIER 3 WAY,VLV 2-LRR-7.1 WDR04 RBCCW SUMP PUMPS P38A,B WDR05 RBCCW SUMP TO L1 SOUND 2-SSP-21 WDR06 DEGASIFIER BYPASS VALVE 2-LRR-69 l l O t l l l i l t ! O. I I i crnran..wns NSEM-4.03 REV.: 5

j. DATE: 5/27/97 PAGE: 7.1-10 of 10

 -.   .      . . . .      .- . . . . . ~ . ..--. - . . ... .- .. .-- - ..- ~ ~-- - --. .- -._.- - ~. ~ . -. .
                                                                                                              .u...-_ _ ..

t, 1 r d l ATTACHMENT 12 1 r 4 COMMONLY USED ABBREVIATIONS AND DEFINITIONS .l i e 3 t. l t h I i l l l 1 O Draftnsem\attl2 ws!

i l SIMULATOR SYSTEM ABBREVIATIONS l O 1 CC Component Cooling - Reactor Building Component Cooling Water (RBCCW) l CH Containment Heating and Ventilating (HVAC) l CR Reactor Core  ; i CV Chemical and Volume Control System (CVCS)  ; CW- Circulating Water . I ED Electrical Distribution  ; l EG Electrical Generation ! ES Engineered Safety Actuation System (ESAS) l FW Feedwater  ! i IA Instrument Air , MS Main Steam l PC Process Computer RC Reactor Coolant System l RD Control Rod Drive System o O RH. Residual Heat Removal / Shutdown Cooling System i i RM Radiation Monitoring I RP Reactor Protection System (RPS) RX Reactor Regulating System (RRS)  ! SG Steam Generators SI Safety Injection SW Service Water TC Turbine Controls TP Turbine Building Components Cooling Water (TBCCW) TU. Turbine l WD Waste Disposal i I Rev: 1 Date: 5/27/97 Page: 1 of 7 i

l l i COMMON ABBREVIATIONS t] { I l Abbreviations Full Name I 1 l Al Analog Input AO Analog Output ANS3.5 ANSI /ANS 3.5 (1985)  ! ARO All Rods Out BOL Beginning of Life CAR Containment Air Recirculation i CEA Control Element Assembly l CEDS Control Element Drive System  ; CIAS Containment Isolation Actuation Signal

CRAC Control Room Air Conditioning CVCS Chemical and Volume Control System DG Diesel Generator l

' n DigitalInput (jDI DO Digital Output l DR Deficiency Report EBFAS Enclosure Building Filtration Actuation Signal EDG Emergency Diesel Generator EOL End of Life l EOP Emergency Operating Procedure ESAS Emergency Nafeguards Actuation System 11FP Ilot Full Pcwer l IIPSI High Press tre Safety Injection i 11ZP Hot Zero Power IC Initial Condition ITC Isothermal Temperature Coefficient LNP Loss ofNormal Power I LPSI Low Pressure Safety Injection hMOL Middle of Life i Rev: 1 Date: 5/27/97 Page: 2 of 7

1 i COMMON ABBREVIATIONS  : b Abbreviations Full Name I MP2 Millstone Point Unit #2 MSIV Main Steam Isolation Valve MLSB Main Steam Line Break MTC Moderator Temperature Coefficient MWE Megawatts Electric

                                                                                   ]

NSEM Nuclear Simulator Engineering Manual PDCR Plant Design Change Request PPC Plant Process Computer PTL Pull-To-Lock RBCCW Reactor Building Component Cooling Water j RCP Reactor Coolant Pump RCS Reactor Coolant System l i RPS Reactor Protection System t q e j l V RRS Reactor Regulating System i SCCC Simulator Configuration Control Committee  ; SDC l Shutdown Cooling System or Simulator Design Change l SIAS Safety Injection Actuation Signal l 1 SIG Simulator Instructor Guide i SIT Safety Injection Guide  ; SOER Significant Operating Event Report SSD Simulator System Diagram l S/G Steam Generator  ; SW Service Water l l TBCCW Turbine Building Component Cooling Water TM/LP Thermal Margin / Low Pressure i l f% l Q i Rev: 1  ! Date: S/27/97 ! Page: 3 of 7

                                                                                   ),

l DEFINITIONS Anomalous Resnonse - Simtdator response which violates the physical laws of nature or differs greatly from expected response . Expected response may be based on plant data, accident analysis, or best estimate evaluation. ANS 3.5 - Anytime ANS 3.5 is listed in this document, it refers to ANSI /ANS 3.5 (1985). Axial Shane Index (A.S.11 - Common term used in reactor core axial power distributions measurement. It is an index which describes the relative amount of power between the top and bottom halves of the reactor core. l Backtrack - The ability to move the simulator back in time to conditions which had previously existed. l This is accomplished by the automatic storage (at one minute intervals) of the simulators I/O's over the past hour. l 1 Best Estimate Evaluation - A method used, (in the absence of plant data, engineering analysis, or accident analysis), to determine the direction, rate, and magnitude of response for critical plant parameters during transient and accident conditions. Experience, rough engineering calculations and mass / energy balances, and table-top discussion may all be used to determine best estimate response. Boolean Trigger - An algebraic expression which is used to automatically activate a malfunction when its { I (J" value becomes true.

   "Cause & Effect" Document - A description of the simulator response (effect) to the insertion of a specific malftmetion or malfunctions. Each malfunction description also contains they physical "cause" of the malfunction as well as a description of the significant effects on plant operation due to the malfunction.       l Certi6ed IC - An IC which has been reviewed by an SRO qualified instructor and verified to have consistent control board and remote function conditions as the reference plant would under the same conditions.                                                                                                      l Certified Remote Function - Those remote functions which will be tested to work correct1 , 'nd may be used in simulator training and exams.

Comnosite Malftmetion - A combination of up to 10 predefined simple malfunctions which ( .oe arranged in a logical sequence. Once built, this composite malfunction is stored and can be used at any time. 1 l Core Performance Testing - Plant heat balance, determination of shutdown margin, measurement of i reactivity coefTicients and control rod worth using permanently installed instrumentation. Critical Parameters - Those parameters, specific to a given major malfunction, which are driven directly q by the initiating event, required for diagnosis, or required to verify proper plant response to safety V equipment actuations and/or operators' corrective actions. l Rev: 1 Date: 5/27/97 l Page: 4 of 7 l 1

l l l p Denciency - An identified difference in a simulator quality or element (hardware and/or software) that requires review and resolution. Deficiency Renort (DR) - Form (STS-Bi-FI A) used by the Operator Training Branch (OTB) and the Process Computers and Simulators (PC&S) to record all identified simulator deficiencies between the  ; simulator and reference plant. 1 Design Limits - Extreme values for specified plant parameters. Design limits are obtained from engineering design and accident analysis documents, e.g.: maximum RCS pressure, peak containment j pressure, etc. I EOP's - Emergency Operating Procedures address the required response by operations personnel to  ; emergency conditmns. Fast Time - The increase in the speed at which certain parameters (such as Xenon, condenser air evacuation, RCS heatup, RCS cooldown, turbine metal heatup, turbine metal cooldown, and decay heat) are modeled to change.  ! Freeze - The stopping of all simulator dynamic modeling. When the simulator is taken out of freeze, the model will continue to run from the time that it was placed in freeze. 1 o Hot Standby Operations - Maintaining stable plant conditions at hot standby. l O i input / Output H/0)- Any digital or analog computer inputs / outputs. l Initial Conditions dC's) - A set of analog / digital points that are stored on the Simulator's Computers so l that a starting point is available for a Simulator session. Physical components (handswitches, relays, etc.) must also be manipulated to match the analog / digital initialization points (switchcheck). Load Changes - Increasing and decreasing plant load. Maior Malfunction - Those malfunctions which produce extensive integrated effects in a number of plant systems which requires complicated analysis to verify acceptable response. i Major Plant Modification - A significant change made to the reference plant which cannot be trained around on the simulator and would result in negative training. Major plant modifications such as the extensive component relocations /changeouts associated with a Control Room Design Review, seriously challenge the ability of the simulator to ftmetion as a plant-referenced training / examining tool. Malfunction - A specific equipment failure which produces discemible indications in the Control Room l that replicates the same equipment failure should it occur in the reference plant. Specific preprogrammed malfunctions are available at the simulator instructor station. O Rev: I l

                                                                                                                         )
                                                                                .                          Date: 5/27/97 Page: 5 of 7

l I Model Limits - Physical conditions which cannot be simulated by the model coding, e.g.: critical pressure l and temperature, core melt, clad melt, etc. Normal Plant Evolutions - Evolutions that the simulator shall be capable of performing, in real time, that simulate routine reference plant evolutions. NSEM - The Nuclear Simulator Engineering Manual contains all of the procedures necessary for the i development and implementation of the certification program. To insure consistent application of the l certification process. This manual is controlled by the PC&S to ensure all copies are current. Nuclear Start-Un - From all CEA's fully inserted to going critical at hot standby conditions. Onerability Testing - A defined group of tests conducted to verify: 1

1. The overall completeness ofintegration of the simulation model, 1

2. Steady state performance for a benchmark set of transients

3. Simulator performance for a benchmark set of transients against established criteria. Operability testing is a subset of the Performance Test and is required annually for maintenance of certification.

Performance Test - A defined group of tests conducted to verify a simulation facility's performance as (] compared to actual or predicted reference plant performance. A performance test is required for initial U certification and for every subsequent four year period in order to maintain certification. Performance l testing for certification maintenance is intended to be an on-going process with approximately 25% of the testing performed during each year of the four year cycle. 1 PDCR - A Plant Design Change Record which contains all necessary information and forms to accomplish in an orderly manner, the modification of a plant system, structure or component. Plant Shutdown - Shutdown from rated power to hot standby, then cooldown to cold shutdown conditions. Plant Startun - The starting conditions shall be cold shutdown temperature and pressure to hot standby temperature and pressure. The Reactor Vessel Head need not be removed for cold shutdown. Reference Phnt.D_ata Book (PDB)- A compilation of reference plant data for specific plant transients / evolutions. The data defines plant parameters response to specific initiating events or l evolutions. Reference plant data may be used to verify simulator response for certification testing, for training development, or as supporting data for DR submittal. Remote Function - An instructor initiated input to the simulator model which will provide the same discernible effects as the corresponding manual operation in the reference plant. O Rev: 1 Date: 5/27/97 Page: 6 of 7

Simulator Configuration control Committee (SCCC) - The committee responsible for overall simulator p design control and management of NTD resources involved in the simulator modification effort. The d SCCC shall include 1) MP2 OPS MGR (Chairman),2) MGR Process Computers and Simulators,

3) Supervisor Software Computer Engineering,4) Supervisor / Assistant Supervisor MP2 Operator Training,5) MP2 Simulator Operations Assistant,6) MP2 Software Engineer,7) MP2 Hardware Technician.

Simulator Design Change (SDC)- A documentation package consisting of relevant DR's and all forms indicated on STS-BI-FIE which is designed to track the resolution of DR's and ensure that ANSl/ANS 3.5-1985, and NRC Reg.1.149 requirements are satisfied. l Simulator Instructor Guide (SIG)- A training document outlining the sequence of events for a simulator j training session. SIG's also contain additional infonnation for the instructor conducting the session. Simulator Operating Limit - A given simulator condition beyond which simulation is unrealistic or inaccurate and negative training may be provided. Simulator operating limits may be imposed due to plant design limits, computer code model limits, or observed anomalous response. Simulation System Diagram (SSD)- Functional representation of the simulator modeling for a given system. Slow Time - In reality, this is the expansion of real time which produces the appearance that a uansient is r b' occurring at a slower speed. The slow time which can be selected can vary from 5% to 95% o (at 5% increments). Snanshot - The recording of the present status of all simulator digital / analog I/O's. After this snapshot is l taken, the simulator may be initialized to this condition at some later time. l 1 SOER - Significant Operating Event Report is generated by INPO and distributed to industry members. It i includes recommendations concerning the event which must be addressed by concerned facilities. j SRO Oualified Instructor - An instructor who is (or was in the past) an NRC licensed (or certified) Senior Reactor Operator, who by nature of his training and experience, has the knowledge to make decisions on proper plant system alignments for given operating conditions. Surveillance Testing - Operation conducted surveillance testing on safety related equipment or systems. Turbine Generator Start-Un - Turbine Generator at zero RPM, to rated speed and synchronization to grid. i O V Rev: 1 Date: 5/27/97 Page: 7 of 7

ATTACHMENT 13 O QUALIFICATIONS OF PERSONNEL DEVELOPING AND PERFORMING SIMULATOR TESTING O I I O I Drafhnsem'attl3551

               .         .      .  .    -.     -    _ -      . . .         .-        .    - -  .   - - . . . .. ~-

Y 1 This attachment lists the Nuclear Training Department personnel involved in developing and conducting Millstone 2 Simulator Recertification Testing. A briefdescription of the relevant qualifications is provided with each individual. As is readily apparent, an excellent mix of operating, training, engineering and simulation software experience is represented by these individuals. Only those personnel directly involved are included. 4 1 I l 4 5 i O l 'V i ) l 1 i 1 l 4 j 1 i i i  ! 1 1 Page: 1 of 9 1

SIMULATOR TECHNICAL SUPPORT BRANCH Lung-Rui Huang - Supervisor, Simulator Computer Engineering NU Experience: Twelve years in Simulation Computer Engineering as both a Senior Engineer and Supervisor. Two years in Probabilistic Risk Assessment and Safety Analysis with Reactor Engineering Other Related Experience:

• Four years simulation experience including Technical Staff Leader with Electronic Associates, Incorporated.

Six years university teaching experience at National Tsing Hua University and Iowa State University in Nuclear and Electrical Engineering Departments. Education: O + PhD in Nuclear Engineering O Page: 2 of 9

_ _ _ _ _ _ _ _ . . _ _ _ _ . . _ . _ _ _ _ . _ _ . _ . _ _ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ ~ _

i i  !

l Shih-Kno Ch=== - Senior Engineer, Simulation Computer Engineering  ;

)

j - NU Experience: l i  ! Eleven years in Simulation Computer Engineering as an Engineer and Senior Engineer, with

;                                              lead responsibility for NU Reactor Core and RCS modeling disciplines.                                                 ,

4 -; l i i ! Other Related Exnerience: I e

                                =                                                                                                                                     I

} Four years simulation experience with Singer Link-Miles Simulation Corporation, including  : ] the position of Section head, PWR NSSS Modeling. i-j- Education: 1 i * . PhD in Nuclear Engineering 2 J 1 3 1 i A O i i ? l I I 4 i i i e 7 e i 4 i N 1 i 1 !O l. l Page: 3 0f 9 i

    ,            -             . . . . - , - ,              -      r        -
                                                                                              ,,-r               ,     , . --   -- . , - , - . . . - ,----,--r ~. ,

                 . _ . _ . . - . _ _ _ . _ . . . ~ . . . . _ . . . . . _ _ . _ _ _ _ _ _ _ _ _ . _ . - . _._

i Hsin-Cheng Huang - Engineer, Simulation Computer Engmeenng  ; O i NU Funerience: i i Ten years in Simulation Computer Engineering as an Engineer. Other Related Ernerience:  ! Three years simulation experience with Singer Link-Miles Simulation Corporation. j i Education: , t PhD in Mechanical Engineering with specialization in Thermofluids and Thermohydraulics. i i t i O i j l O Page: 4 of 9

  . . _ - -         -. - - . - - - - .                - - - . . . - . - . - . - _ - -              --. -             . . ~ . . - .

t John H. Rein - Supervisor, Operator Training, Millstone 2 NU Exnerience: Three and one-half years in Operator Training as a supervisor responsible for simulator  ! performance and all aspects of operator training, l 1 e l Ten years in Operator Training as an operator instructor responsible for coordinating, l development, and delivery of all aspects of operator training. This also included simulator construction, startup and testing responsibilities. l e .Two years as a Connecticut Yankee Auxiliary Operator. Other Related Exnerience: One and one-half years as an operator instructor at Louisiana Power & Ught responsible for all development, and delivery of all aspects ofoperator training. NRC Licenses: RO License (CY) - SRO Instructor Certification (Waterford 3) SRO Instructor Certification (CY) i l i l' i l l l I i Page: 5 of 9

Bonald G. Stotts - Assistant Supervisor, Operator Training Millstone 2 N.] NU Experience:

• Six years at MPl in both Operations and in Operator Training.

Nine years at MP3 Operator Training in various positions from Instructor to Supervisor MP3 Operator Training.  ; i Three years at MP2 Uperator Training as Assistant Supervisor MP2 Operator Training. j Diher Related Experience:

• Four years, U.S. Navy I NRC Licenses:

1

• SRO Licensed MP1

('T

• SRO Licensed MP3 (J

l A NY i l Page: 6 of 9 l

                                    .-   . - .      . . - .           - - , . .       - . - . . . . - ~ . , - . - . . ,-

Richard N. Snurr - LORT Coordinator, Operator Training, Millstone 2 O NU Experience: Three years in Operator Training as the supervisor responsible for simulator performance and training. Ten years in Operator Training as a Senior Operator Instructor, including simulator construction, startup and testing responsibilities.

• Three years as a Millstone 2 Shift Supervisor.
• Five years as a Millstone 2 Supervising Control Operator.

Other Related Exnerience:

        .       Six years, U.S. Navy NRC Licenses:

O

• Currently SRO Licensed, MP2 (21 years) l l

l O Page: 7 of 9

4

Robert H. Burnside - Simulator Operations Assistant j NU Exnerience

3 l

• Six and one-half years in Operator Training as a Senior Operator Instructor, ,

j e- Two years in Operating Training as a Simulator Operations Assistant.

• Eleven years as a Millstone 2 Shift Supervisor.

l , ll j e Twelve years additional operating experience, including Control Operator at the Haddam  : 1 Neck Plant. 1 NRC Licenses: 4 I i e Previously SRO Licensed, MP2 (19 years) J e Currently retired 4 i O

}

l t e t i, if I k i. 4 1-i- O Page: 8 of 9

          . ~            _ .                                                                                        _              ,

[ l t William H. Souder - Simulator Operations Assistant t NU Experienen

             . Twelve and one-half years in Operator Training as an Operator Instructor, including simulator         I construction, startup and testing responsibilities.
             . One year as MP2 Simulator Operations Assistant Other R:iated Experience:
             . Four years as a Simulator Instructor at the Combustion Engineering Simulator.
             . Fourteen years, U.S. Navy, including Engineering Officer of the Watch, Qualifications and power plant protype training experience.

N_RC Licenses:

              . Currently SRO Licensed, MP2 (eleven years)

O l l 1 i l O l Page: 9 of 9}}