ML20245K877

Simulator Certification Submittal,Vol 3: 'Performance Test Rept.'

ML20245K877
Person / Time
Site:	Millstone
Issue date:	06/29/1989
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20245K850	List: B13287, Forwards Millstone Unit 2 Simulator Certification Submittal, Vols 1-4 ML20245K865 ML20245K870 ML20245K877 ML20245K883
References
NUDOCS 8907050289
Download: ML20245K877 (251)

Text

- - - - - - - - - _ - - - - - , - - - - - - , - - , - -

gd

VOLUME 3 1

MILLSTONE UNIT 2 SIMULATOR PERFORMANCE TEST REPORT O

O grusa ns,PDC

.\

l MILLSTONE UNIT 2 SIMULATOR PERFORMANCE TEST

SUMMARY

O O

1

-s VOLUME 3

)

w/

TABLE OF CONTENTS  !

i l

I i

i Page No. 1 i

l Summary of Performance Testing 4

1. Testing Goals, Methodology and Assumptions 7 2, System Tests 12

3. Normal Operation and Surveillance Testing 13

4. Malfunction Testing 15 i

5. Yearly Operability Testing 19

6. Physical Fidelity Testing 21

7. Initial Conditions Testing 24

,r' 8. $1mulator Operating 1.imits Testing 25

' ', _ _ 9. Instructor Station Tasting 29

10. Realtime Testing 31

11. Ensuring Continued Performance of the MP2 Simulator 32

12. Open Deficiency Report (DR) List 34

13. Next Four-Year Schedule 37 l

\

uJ

f l

TABLE OF CONTENTS (continued) d Attachment 1 MP2 Turbine System Test Attachment 2 MP2 System Test Abstracts Attachment 3 MP2 Normal Operations and Surveillance Testing Sequence Attachment 4 MP2 Suiveillances that can be performed on the Simulator Attachment 5 MP2 Malfunction Test Abstracts Attachment 6 initial Conditions Checklist and List of 25 Certified Initial Conditions Attachment 7 Student Feedback Survey Results k./ Attachment 8 Open Deficiency Report (DR) List Attachment 9 Schedule for Next Four Years of Testing Attachment 10 Annual Operability Transient Testing Abstracts Attachment 11 Physical Fidelity Summary Report Attachment 12 Sample Malfunction Test Procedures Attachment 13 List of Certifled Remote Functions j I

Attachment 14 Commonly Used Abbreviations and Definitions l

Attachment 15 Experience of Testing Personnel l

't

1

SUMMARY

OF PERFORMANCE TESTING

(

U)

The Millstone 2 Simulator was declared " Ready for Training" in May,1985. An Acceptance Test Procedure (ATP) was performed prior to using the Millstone 2 Simulator for operator training.

In March, 1987,10 CFR 55.45 was issued requiring certification of Simulation Facilities.

Northeast Utilities (NU) formulated a comprehensive testing program to meet the reautre-ments of 10 CFR 55.45, ANSI /ANS 3.5 (1985) and Regulatory Guide 1.149 (1987).

The purpose of this program was to provide systematic and detailed assurance that NU Simulators meet or exceed regulatory requirements and provide the best possible simulator training to its licensed operators. The generic procedures which make up the Northeast Utilities Simulator Certification Program are included in Volume 2 labeled " Nuclear Simulator Engineering Manual" (NSEM).

Utilizing the generic procedures in the NSEM for guidance, a Millstone 2 specific Performance Test was written. This Performance Test was conducted on the Millstone 2 Simulator starting in May,1987 and finishing in April,1989. Volumes 3 and 4 of this submittal describe the Millstone 2 Performance Test and its results. While the generic procedures covering Simulator Certification are contained in Volume 2, the Millstone 2 specific Performance Test procedures are not included in this submittal. The Millstone 2 specific Performance Test procedures and results are available for review upon request.

i Abstracts of these procedures together with test results are presented in this submittal.

The Mllistone 2 Performance Test will be repeated over a four year interval as described in Section 13 of this submittal.

l f

(..

This volume contains the following 13 sections and 15 attachments.

%)

o Section 1 provides a description of testing rnethodology and assumptions.

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

o Section 11 reviews and summarizes the procedural controls for maintaining certification of the Millstone 2 Simulator, o Section 12 provides a time. cble for resolving open deficiencies on the Mllistone 2 Simulator. A 4 of 3-28-89 there are 67 open deficiencies on the Mllistone 2 Simulator.

o Section 13 discusses the testing sequence for the next four (4) year certification period (June 1989 through June 1993).

o Attachments 1 through 13 provide supporting documentation for sections 1 through 12. The attachments are referenced in the appropriate sections of this summary. Attachment 14 provides a list of abbreviations and definitions. Attachment 15 contains a list of experience levels of test personnel.

Volume 4 contains an index of all malfunctions certified on the Mllistone 2 Simulator.

Following the index is an Individual "Cause and Effects" description for each certified malfunction. Each "Cause and Effects" description is from 1 to 3 pages in length and describes the basic characteristics of the malfunction, what "Causes" the malfunction, and what the " Effects" of the malfunctions are for a given " Plant Status". The malfunction "Cause and Effect" descriptions are organized by plant system in alpha numeric order as shown by the index. Approximately 225 malfunctions were certified.

O V

.s.

l

' Based on the results of the completed Performance Test, the Millstone 2 Simulator has b

Q demonstrated excellent physical and functional fidelity when compared to the Millstone 2 reference plant. A comprehensive testing program has been performed to reach this conclusion and procedural controls are in place to ensure that the Millstone 2 Simulator retains the demonstrated fidelity.

The Performance Tests described in sections 2 through 9 were all performed by NRC licensed SROs, including three former Millstone 2 Shift Supervisors and a former Millstone 2 Reactor Engineer. As shown in the malfunction abstracts, computer code analytical predictions were used to the maximum extent possible for complex transients such as Loss of Coolant Accidents and Main Steam Line Breaks. Any deficiencies identified during the Performance Tests are identified in the attached ,

Performance Test abstracts, including dates for resolution.

l O

l 1

l

I.

L  !

t

1. Testing Goals, Methodology and Assumptions During the development and implementation of the Millstone 2 Simulator certification program, goals, methodologies and assumptions were established to.

ensure an efficient, effective and comprehensive approach to testing. Certain elements of this testing philosophy are worthy of mention here: ,

o Testing should be conducted during normal, abnormal and emergency conditions, o- The Simulator response, as verifled by testing, during normal, abnormal and emergency conditions shall meet the following criterla necessary to support the contents of the training curriculum:

Correct operator diagnosis is possible.

O ~-

Capabilities for operator intervention to mitigate events exist.

Actions or inaction taken by operators result in similar response as in the reference plant.

Alarms and automatic system actuations shall occur such that operator diagnosis and response is not adversely affected.

o Any deficiencies found during testing which violate these criteria shall be documented by generating a Deficiency Report (DR), to be dispositioned in accordance with the NSEM.

o The requirements of ANS 3.5 shall be implemented. '

-7

in

/

't,-) . o All Simulator controls such as switches, annunciators, meters, l controllers, recorders, lights, keylocks, pushbuttons, etc.,

should be tested.

o All process computer points needed to support training should be l' tested.

o To the maximum extent possible, former Millstone 2 operators should be used for Normal Operations, System and less-severe Malfunction testing.

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

m o Two experienced observers should be used whenever possible to improve the observations made during testing, i o Acceptance Testing Procedure (ATP) documents should be used as a basis for certification testing procedures and expanded upon where necessary.

o Provisions for the revision of NSEM procedures shall be available.

Changes to that manual will not be forwarded to the NRC unless they significantly alter the intent of the test program.

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

This additional guidance, or deviation from the general philosophy, is summarized:

,r x

8

I i,

l SYSTEM TESTS o Some analog and digital process computer points need not be tested if they do not impact the ability to conduct training or examinations.

o Digital plant process computer points need not be included on the Simulation Diagrams. Their absence improves the clarity of the diagrams.

NORMAL OPERATIONS TRAINING i

o ANS 3.5, Section 3.1.1, item 7 need not be tested. This evolution is orchibited by the Millstone 2 Technical Specifications.

o Testing of surveillance on redundant equipment or flowpaths is not

(. required if the primary piece of equipment or flowpath is tested. For

( example, if the Facility 1 Service Water Pump surveillance is performed, the Facility 11 Service Water Pump surveillance need not be performed.

o The simulator's capability of performing a reactor trip followed by recovery to rated (full) power (ANS 3.5, Section 3.1.1, item 4) may be tested by testing:

an increase in power from 20% to 100%, followed by the reactor trip, then an increase in power to 20%

There is no need to test the power ascension to 100% twice.

O C

I

)

i, / YEARLY OPERABILITY TESTING STEADY STATE TESTING o Monitoring Steam Generator Temperature and Pressurizer 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.

Additionally, the accuracy of the reference plant's Pressurizer Temperature Indication is not considered sufficient to be meaningful.

o Application of both instrument error and allowed tolerance is to be made as follows. The ANS 3.5 allowed tolerance (Section 4.1(3) or (4)) is to be applied as a function of the reference plant Indicated value above the minimum scale value. The allowable instrument error is to be a function of the instrument range, not its reference plant Indicated p value, when the instrument error is given as a percentage. For example:

For a pressure instrument with a range of 1500 to 2500 psia, reading 2250 psia in the reference plant, the simulator accuracy, as indicated on the Instrument, is expected to be:

(2500-1500) (.01) + (2250-1500) (.02) - 25 psia i

This assumes an Instrument accuracy of 1 % and that this parameter is a critical parameter.

O g TRANSIENT TESTING o During testing of the Maximum Rate Power Ramp (reference ANS 3.5, Appendix B, B2.2, item 7), returning power to 100% will not be tested.

Raising power at the maximum rate from an initial power level of 75%

violates Millstone 2 current design and trip criteria. This is more fully explained later.

o All parameters required by ANS 3.5, Appendix B, B2.2.4 are to be tested at .5 second intervals except as noted below:

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

Reactor Vessel Level and Saturation Margin are to be recorded at 5 second intervals. Data collection of these two points at .5 second intervals is not within current

( capabilities.

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

o 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. Replotting of the transient is not required until the next scheduled yearly testing of the transient.

I o Documenting the difference between the response of a simulator parameter and predicted reference plant response is not necessary for those differences of an obvious nature.

A

.(

i

[ I V OTHER TESTING j

A eTesting of allinput/ Output (1/0) override capabilities is not required during testing of the Instructor Station. Testing a sample of 1/O overrides is insufficient to' demonstrate the simulator':; capabilities. Specific I/O override i points are to be tested, as required, during curriculum testing, g ,

- 2. System Tests-

'L l The Millstone.2 Simulator models 23* Plant Systems. A separate test was

. conducted for each of these 23 Systems to ensure that all of the following operate correctly:

o. Control Board Hardware such as handswitches, meters, controllers,

- recorders, Indicating Ilghts, keylocks and pushbuttons -

o~ . Annunciators and Process Computer Points o Remote Functions (These are tasks performed by an instructor at the instructor station to simulate local actions; typically these are locally operated valves) o- Flowpaths, both normal and abnormal 4 k

NSEM Procedure 4.01 " Verifying Simulator Capabilities via System Tests" in Volume 2 of this submittalis the generic procedure which governs the writing and perform- i ance of SystemTests. Page 17 of this procedure lists all 23 Millstone 2 Systems that are modeled. Examples of Millstone 2 Systems tested are: Service Water, Main Steam and Turbine.

• Volume 1 of this submittallists > 23 simulated systems. For purposes of -

Simulator Performance Testing, some systems have been combined into a single system test, hence the value of 23 simulated systems.

O  ;

i i

( ,/ The Turbine System Test has been provided in Attachment 1 as an example. ,

Due to their large volume, the other 22 System Tests are not being transmitted. f Abstracts of all 23 System Tests are contained in Attachment 2. The only deficiencies found by these tests and still open are contained in the Reactor Core System Test (System Test #23 in . Attachment 2). Refer to the Reactor Core System Test for details and dates for resolving these deficiencies.

Each simulated plant system has its own performance test to ensure:

1 o That all components of a specific system are checked at the same j time for consistency.

o That a consistent set of performance requirements are applied to each system.

fm o That as Plant Design Changes are implemented, the System Test will act as a benchmark for proper system response.

The System Tests satisfy the requirements of ANS 3.5 Section 3.3. System Tests will be repeated over a four year Interval as detalled in Section 13 of this summary.

The System Tests are listed first in this performance test summary because their performance is a logical prerequisite to Normal Operations Testing, Surveillance Testing and Malfunction Testing.

Attachment 13 contains a list of all certified remote functions. All certified remote functions have been tested in the System Tests.

3. Normal Operations Test ANSI /ANS 3.5 (1985) Section 3.1.1 requires the simulator to be capable of performing n normal plant evolutions and surveillance.

L__________________. _ _ _ . . _ _ _ _ _ . . _ ._

,3

, The normal operations and surveillance required by ANS 3.5 Section 3.1.1(1), (2),

'(, ) (3), (4), (5), (6), (8) and (10) were performed using controlled copies of Millstone 2 Operating Procedures and Surveillance. As discussed in Volume 1 (Exceptions), the operating condition specified in Section 7 is prohibited by Mllistone 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" (See Volume 2) contains the generic guidance used to write the Mllistone 2 Simulator Normal Operations and Surveillance Test.

Attachment 3 contains the Millstone 2 Simulator test procedure used for Normal Operations and Surveillance testing. Using controlled copies of Millstone 2 Operating Procedures the following sequence of operations was tested on the Millstone 2 Simulator:

o The Simulator was initialized to Cold Shutdown conditions.

o A Plant Heatup was performed.

.O o A Nuclear Startup was performed.

o A Plant Startup was performed.

o A Load increase to 100% power was performed.

o A Reactor Trip was initiated.

o A Reactor Trip recovery was performed.

o A Nuclear Startup was performed.

o A Plant Startup was performed, o A Load increase was performed.

o The Simulator was relnitialized to 100% power, o- A Plant Shutdown was performed.

I' o- A Reactor Shutdown was performed.

o 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 3. Attachment 4 contains a concise list of 64 Surveillance that the Millstone 2 Simulator is capable of performing. As shown on this list, some surveillance were exempted from testing if they were simply a repeat of the same type of test on a different Electrical Facility. Given the comprehensive nature of the System tests, testing of only one Facility is sufficient assurance of correct operation.

Normal Operations and Surveillance Testing was performed in April and May,1988. No j deficiencies were identified. The Reactor Core System Test used to meet the ANS 3.5 Section 3.1.1 (9) requirement did have some deficiencies. These deficiencies are discussed in Attachment 2, System Test #23. All Normal Operations and Surveillance Testing will be re. performed over a four year interval as described in Section 13 of this document.

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

O

I m The Millstone 2 Simulator is certified for approximately 225 malfunctions. Volume 4 )

of this submittal contains an index of all malfunctions available on the Millstone 2 i

Simulator. All malfunctions listed in the Volume 4 Index 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 14 contains a list of )

definitions for plant system abbreviations. Also contained in Volume 4 of this l submittal for each certified malfunction is a brief summary (1 to 3 pages) of the malfunction "Cause and Effect" Each "Cause and Effect" description contains:

o The malfunction description / title ,

o Whether or not it is a variable malfunction o The "Cause" of the malfunction o The initial Plant Status that the malfunction " Effects" are written for k o What the " Effects" of the malfunction are on plant operations o References showing where information was obtained Each certified malfunction has its own test procedure. Guidance for writing malfunction test procedures and conducting tests is contained in:

o NSEM Procedure 4.04, Major Malfunction Testing (See Volume 2) o NSEM Procedure 4.05, Malfunction Testing (See Volume 2)

Malfunctions which cause major integrated plant effects, such as Large RCS Breaks, Main Steam Line Breaks, etc., have their respective malfunction test procedures l

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 O following documents / sources:

1 p o CEN 128, " Response of a CE NSSS to_ Major Transients" u

o CEN 268, " Justification of Trip 2/ Leave 2 RCP Strategy During Transients" o o. Millstone 2 Final Safety Analysis Report (FSAR) o ~ Northeast Utilities generated "RETRAN" cases, specific to

Mllistone 2 An example of a malfunction test written and conducted via the NSEM 4.04 process is contained in Attachment 12. The example in Attachment 12 is labeled "CV01: Unisolable letdown line rupture in CTMT between the RCS and 2-CH-515."

Malfunctions which do not cause large integrated plant effects have their respective malfunction test procedures written and tests conducted per the guidance in NSEM O- 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 NRC licensed SRO Instructor 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 12. The example in Attachment 12 is labeled "FWO1: Loss of Condenser Vacuum" In summary, each of the approximately 225 certified malfunctions has its own i malfunction test procedure and its own "Cause and Effect" description. All of the "Cause and Effects" descriptions are contained in Volume 4 of this submittal. Due to their large volume, only 2 malfunction test procedures are being provided in Attachment 12, as examples. Several additional volumes would be required if all the malfunction test procedures were to be submitted. These malfunction test procedures are available for review upon request.

l

ANS 3.5 Section 3.1' 2 requires 25 specific malfunctions to be available on a Simulator.

()

7-q .

In order to facilitale NRC review of this submittal, Attachment 5 contains a cross reference of these 25 specific malfunctions to the applicable Millstone 2 Simulator malfunctions. Listed in Attachment 5 under each of the 25 ANS 3.5 required malfunctions are:

o Abstracts from Mllistone 2 Simulator malfunction tests providing:

The name of the malfunction The date the lest was performed Whether the malfunction is variable, and if so, what its range is and what severity was tested Starting conditions and end point conditions The source of baseline or reference data  !

Details of identified deficiencies, if any, including dates for resolution (b o At the end of each of the 25 sections is a list of other Millstone

' 2 Simulator malfurictions which meet the ANS 3.5 requirement, but for which no abstract is provided. Open deficiencies and a schedule for resolution is also provided.

Attachment 5 contains approximately 55 malfunction test abstracts of the 225 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 (available in Volume 4 of this submittal) may be used as abstracts for other certified Millstone 2 Simulator malfunctions.

O

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

'As an example,' malfunction CV21 was added to the Simulator about one year ago based

'on observation of student performance on the Simulator.

All certlfled malfunctions will be retested over a four year interval, as described in Section 13 of this document.

l

5. . Yearly Operability 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 was performed on the Millstone 2 Simulator in April and May,1988 and will be re-performed on an annual basis.

The Yearly Operability Testing performed in 1988 consisted of the following items:

o Steady State Testing at 28% power,50% power and 100% power o Stability Testing at 100% power o Transient Performance Testing for ten (10) transients Reference Plant data obtained at 28%,50% and 100% power during the February 1988 refuel startup was used as the basis for Steady State Testing. Utilizing the Reference Plant data, comparisons were made between the Simulator and Reference Plant for approximately 80 selected critical and non-critical points.

O

L L

l=

l l

, These 80 points include all those listed in ANS 3.5 Section B2.1 with 2 differences:

-t V) (1) Mllistone 2 has no control board Steam Generator Temperature Indication and (2) the Pressurizer Temperature Indication in the Reference Plant data was not accurate and therefore was not used.

A Stability Test was performed at 100% power for approximately 36 points over a one hour period. This test was in conformance with ANS 3.5 Section B2.1.

Acceptance criteria for the Steady State and Stability Tests were based on ANS 3.5 Section 4.1. No deficiencies were identified.

Transient Performance Testing was performed for ten transients. The ten transients tested were those described in ANS 3.5 Section B2.2, with one difference. The one difference 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, p and not back up to 100% power. A rapid rate power increase from 75% to 100% power k, 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 increase 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 B2.2.1,2,3, or 4, as appropriate. Attachment 10 contains abstracts of each of the ten transient tests. This testing was performed in April,1988. As described in the abstracts, five of ten transients passed without deficiencies. Five of the ten transients had some deficiencies identified. As of the date of this submittal, three deficiencies are still outstanding and will be resolved by May 1,1991. Refer to Attachment 10 for details of these three open deficiencies. ,

O 1

._-__-_______-___D

1 i

l

. 6.' . Physical Rdelity h.g ANSI /ANS 3.5 (1985) Sections 3.2 and 3.3.1 require sufficient panel and controls '

simulation to conduct normal operations and malfunction response. Further, the Simulator controls are required to duplicate the physical characteristics of the Reference Plant. In response to the issuance of 10 CFR 50.45, a two step evaluation process was employed for the existing Millstone 2 Simulator to ensure compliance with the ANS 3.5 Section 3.2 and 3.3.1 requirements. The first step was verification that the existing Millstone 2 Simulator contained sufficient panels, controls and instrumentation to perform normal operations and respond to malfunctions. The second step was verification that the Millstone 2 Simulator panels and controls have

, physical fidellty with the Reference Plant. These two steps in the Physical Fidellty evaluation are summarized below, as sections A and B.

A. The Millstone 2 Simulator was constructed prior to the issuance of 10 CFR 55.45. NU considered it appropriate to conduct a review of the adequacy of the Simulator to meet the needs of lleensed operators in f-performing normal operations and responding to malfunctions. This review process was governed by the following 2 procedures:

o NSEM Procedure 2.01, " Defining Training Requirements" o NSEM Procedure 2.02, " Defining the Certified Trainer" Utilizing the above procedures, a systematic review was conducted of all tasks contained within the Simulator Training Guides and all ANS 3.5 requirements in order to compile a comprehensive list of:

o All required control board hardware (control board switches, meters, controllers, etc.) needed for Simulator Tralning o All regulied software flowpaths, both normal and abnormal, needed for Simulator Training ,

I

-O

o All remote functions (locally operated valves, etc.)

needed for Simulator Training o All malfunctions needed for Simulator Training The Simulator Training Guldes utilized in this process were developed using a Systematic Approach to Training. The few identified deficiencies resulting from this systematic review have all been corrected. The documentation of this systematic review is not in this submittal, but is available for review, upon request.

The small number of deficiencies (all corrected) Identified by this process confirmed the adequacy of the decision-making process used for the initial Millstone 2 Simulator design. This systematic review as defined in NSEM Procedures 2.01 and 2.02 was a one-time process to verify the adequacy of the scope of simulation to support training. The Training Department receives sufficient notification of procedure changes, LERs, Plant Design Changes, etc.,

V to ensure that training programs and the Millstone 2 Simulator are modified to keep pace with Reference Plant changes.

B. The next step was a compare the Reference Plant and Simulator Control Rooms for physical fidelity. This process is described in NSEM Procedure 4.12,

" Simulator Physical Fidelity / Human Factors Evaluation". A complete set of photographs was taken of the Reference Plant Control Room in December,1988 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 dispositioned as "not affecting training" are I described in Attachment 11, labeled " Physical Fidelity Report" To ensure continued physical fidelity, photographs will be taken annually for Reference Plant Control Boards that have undergone changes.

1 1 - _ _ _ _ _ _ _ _ - _ _

I 7x NU has a strong commitment to maintain the Millstone 2 Simulator up to date with the Reference Plant Control Boards in a timely manner. This commitment has been demonstrated by the fact that the Millstone 2 Reference Plant has undergone extensive changes due to the implementation of the Control Room Design Review over the last two refueling outages (Early 1988 and Early 1989).

The Millstone 2 Simulator was updated after each of these two outages within one month of the Reference Plant allowing licensed operators to receive up to date training on the new control boards. The timeliness of the simulator modifications following the early 1988 refuel outage is documented in the following NRC inspection report: Report No. 50-336/88-07, Docket No. 50-336, License No. DPR-65, Inspection Period 3/22/88 - 5/2/88, Page 4, which states:

"The C01 panel on the Unit 2 Simulator was found to be almost identical to the control room panel C01 with regard to recent modifications. Differences were that the simulator panel did nct have the new flow mimic and the digital discharge pressure Instrumentation for the LPSI pumps. Otherwise, the improvements b to control room panel C01 performed during the most recent refueling outage were reflected on the Unit 2 simulator panel."

This NRC Inspection was conducted about one month after Reference Plant changes. The differences mentioned in the above report were corrected with the completion of the simulator modifications shortly after the NRC Inspectors left.

NSEM Procedure 6.04, " Major Plant Changes", addresses controls on major design changes (such as Control Room Design 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.

l

)

l

y. -

b g 7. . initial Conditions Testing

{

inillal Conditions Testing was performed in February,1989. NSEM Procedure 4.02,  !

" Initial Conditions", describes this process. The Millstone 2 Simulator has

' capabilities for storing 60 Initial Conditions.

All Certified Initial Conditions (ICs) were reviewed to ensure equipment alignments, plant conditions, remote functions, etc., were reasonable for the stated IC condi- I tions. The first four pages of Attachment 6 contain the checklist used for reviewing all certified ICs. The number of certified ICs may vary between 25 to 60 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:

o Beginning of Core Life (BOL) o Middle'of Core Life (MOL) o . End of Core Life (EOL) o Different Operating Modes such as Cold Shutdown,

.( Hot Standby, Critical Approach, etc.

o Different Power Levels Only cedified ICs are used for training or exarns. The last two pages of Attachment 6

. lists the " base" group of 25 certified ICs. Listed in Attachment 6 for each of the 25 ICs are the following:

3 o RCS T,y, ( F) o . Pressurizer Pressure (psig) o Reactor Power (%)

o RCS Boron Concentration (ppm) o Xenon Reactivity (pcm) o Date/ Time when the IC was last modified o Remarks section which describes the basic conditions of theIC I

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

_4.-,. __ __ --_. __ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . ______._-_-..m _ ____ . _m m___.m-._m.. ._____u--_-- _ - --

L f} Simulator Operating Limits Testing

" Q 8.

. Simulator Operating Limits Testing was performed between February,1988 and February,1989. The purpose of this testing was to identify any areas of possible negative training and to take actions to prevent such negative training.

The process used for identification and action concerning Simulator Operating

. Limits is described in NSEM Procedure 4.08, "Straulator Operating Limits".

Two methods are used to prevent negative training when Simulator Operating Limits are reached: a) freezing the simulator and, b) administrative controls. The Reference Plant design limits and/or Simulator model limits which cause the l I

Simulator to " freeze" are listed below.

Also listed are the administrative Simulator Operating Limits which are controlled by the simulator instructor. These administrative limits are p implemented through simulator instructor training and cautions placed in

\ those Simulator Guides where such situations could occur.

Additionally, problems currently covered by Deficiency Reports (DRs), which have been judged to be of such significance that they are dealt with administratively as if they were Simulator Operating Limits, are indicated. As these DRs are dispositioned,- the deficiency will either be corrected or become an administrative Simulator Operating Limit.

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

RCS Pressure 2750 psia Containment Pressure 60 psig S/G Pressure > 1200 psia

- Fuel Temp 4980 F

- Fuel Clad Temp 3310 F l j

1 1

(

,e m The simulator instructor can determine which of these operating limits

(, caused the simulator to go to Freeze by reviewing a CRT display in the instruction station.

The following simulator operating limits are dealt with administratively.

These limits are all results of Simulator model limitations:

o RCS Pressure should not exceed 2000 pst greater than Steam Generator Pressure.

o A drain down of the RCS from a given pressurizer level to below the bottom of the pressurizer (or vice versa) at atmospheric pressure is not permitted. Note: The simulator can be initialized for Shutdown Cooling operation with RCS Level below the Pressurizer.

rm o Axial Xenon Oscillations are always divergent on the Simulator. This is not a problem in normal training sessions due to their short duration. However, should a training session be done in Xenon Fastime, it would have to be done carefully to prevent negative training.

o 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.

o The reactivity worth of stuck Control Element Assemblies (CEAs) is too small. I 1

l I

J

((/ The following limits are addressed administratively. These limits are a result of currently existing Deficiency Reports (DRs). If the deficiency cannot be corrected, the limit will be added to the non-DR Administrative Controls list.

o For Large RCS Breaks (Large Break LOCAs), the Reactor Vessel Water Level response, Core Exit Thermocouple (CET) response and Unheated Junction Thermocouple response is not always accurate. These deficiencies are tracked via Deficiency Reports (DRs) 87-2-144, 87-2-185, 88-2-55 and 87-2-28. These DRs will be dispositioned

• by 5-1-91.

o The MP2 Simulator can be initialized to provide training for a Loss of Shutdown Cooling from centerline of the Hot Leg conditions. On a loss of Shutdown Cooling, the Simulator is accurate only up to the point at which saturated conditions are reached. Past saturation, the p

V Simulator gives inaccurate CET response and Reactor Vessel Water Level response. This deficiency is tracked by DR 89-2-15, which will be dispositioned

• by 5-1-91.

"Dispositioned" means 1 of 2 things will occur. Either: 1) the DR will be fixed, or 2) If the DR cannot be fixed. the problem may be added to the Simuistor Operating Limits ifit is significant. It is important to recog-nize 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

- _ _ -_____ _ - l

i 1

o For dropped CEA events, the incore radial tilt Indicates zero lm 5

s f after CEA recovery. For most dropped CEAs, even after recovery, the actual incore radial tilt will still be significant. This deficiency is tracked by DR # 87-2-90 and will be dispositioned

• by 5-1-91. l o For dropped CEA events, the Simulator dropped CEA reactivity worth is too large, causing too much of a power decrease.

This deficiency is tracked by DR # 89-2-16 and will be dispositioned

• by 5-1-91.

o Sudden changes in S/G steam flow at or near 100% power do not cause sufficiently large changes in RCS parameters. Shutting both MSIVs does trip the reactor but does not cause sufficient RCS temperature and pressure increase. Also, causing a Steam Line Rupture downstream of the MSIVs does not cause sufficient s RCS cooldown. These deficiencies are tracked by DRs 88-2-77 and 89-2-17 and will be dispositioned

• by 5-1-91.

i o Loss of one or more RCPs at power does not result in sufficient RCS temperature and pressure increase, although the reacter is automatically tripped promptly as required. This deficiency is tracked by DR # 89-2-14 and will be dispositioned

• by 5-1-91.

"Dispositioned" means 1 of 2 things will occur. Either: 1) the DR will be fixed, or 2) If the DR cannot be fixed, the problem may be added to the Simulator Operating Limits if it is significant. It is importar.t to recog-nize 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 j quality training is presented.

1 l

1 i

(

I f

- o The Ejected CEA worth is too small for Ejected CEA malfunc-() tions. This malfunction is not used in training. This deficiency is tracked by DR 88-2-114 and will be dispositioned

• by 5-1-91.

o At saturated RCS conditions, void formation in the RCS and the l

Reactor Vessel Upper Head is not severe enough. Similarly for

(

excessive cooldowns of the RCS, upper head volding is not severe enough. This deficiency is tracked by DR 89-2-41, and will be dispositioned by 5-1-91.

The deficiencies listed in Section C above are discuss.ed in more detail in the Performance Test abstracts contained in this submittal.

9. Instructor Station Testing i

(%)

\/ During June 1988, 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.

l

• j "Dispositioned" means 1 of 2 things will occur. Either: 1) the DR wl!!

be fixed, or 2) If the DR cannot be fixed, the problem may be added to q the Simulator Operating Limits if it is significant. It is important to  !

recognize that priorltiration of resoh'ing DRs is a dynamic process. As i 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.

i G

1

L L-p

, ,s o Backtrack o Fastime, for each of the eight (8) modeled Fastime

. parameters o Slowtime o Boolean Trigger o Composite Malfunction o Variable Parameter Contro!

o Freeze o Snapshot 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.

o Analog Outputs o Analog Inputs o Digitallnputs

o Digital Outputs O o "Crywolf" Annunciator feature o Annunciator Override The purpose of the I/O override feature testing was to verify the feature itself, not every 1/O override point. The Mllistone 2 simulator has the ability to 1/0 override essentially every point in the simulator. While this is a great capability, there are therefore thousands of 1/O override points. Curriculum testing of a simulator lesson plan will require the testing of any individual 1/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 from this test is not contained in this submittal, but is available upon request.

The Instructor Station test will be repesc d once every four years.

O

~s 10. Rea! Time Testing

[ )

%J Real Time Testing was performed in November,1988, per NSEM procedure 4.13, "Real Time Simulator Verification".

The purpose of this test was to verify that all simulation models are running in real time. Verification was accomp!!shed by:

o Monitoring the operations of the Internal Computer Clock and Interrupt Timers and comparing them against the vendor's specifications.

o Ensuring that the spare time remaining in the simulation computer for each of the following complex scenarios was > 10k i Turbine load reject / trip Steam-Line Break RCS Hot-Leg double-ended LOCA O -

RCP locked rotor o Installing software counters in the Reactor Core, RCS and Feedwater models and comparing their actual values to expected values for each of the above scenarios.

The results of these tests show that the Millstone 2 Simulator performs in real time.

In ae 'llon, an Internal software timer continuously monliors computer usage and will automatically bump out any task that slips two consecutive frames.

No deficiencies were identified. 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

[ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

I

11. Ensuring' Continuing Performance of the MP2 Simulator l

O To ensure that the MP2 Simulator performance remains in compliance with

' ANSI /ANS 3.5 (1985), Reg Guide 1.149 and 10 CFR 55.45 the following procedural controls have been implemented:

Major Plant Modifications The Millstone 2 Simulator was certilled as a Plant i

l - Referenced Simulator. Significant Reference Plant Control Room changes, such as from Control Room Design Review modifications, must receive special consider-l atton due to their potential major impact. NSEM Procedure 6.04, " Major Plant l

Modifications", addresses this concern. This procedure ensures that major plant modifications affecting the Reference Plant Control Room are reviewed and acted on in a timely manner. This ensures that training and exams continue to be performed on a valid plant referenced Simulator. The Control Room Design Review modifica-tions implemented in the Reference Plant Control Room during the last 2 refueling outages fit this category. The Millstone 2 Simulator was updated within A approximately one (1) month of the Reference Plant changes after each of these

' 2 refueling outages. This demonstrates a commitment to keeping the Millstone 2 Simulator a Plant Referenced Simulator.

Plant Design Changes / Procedure Changes .

All Plant Design Changes and Procedure Changes are sent to the Training Department to be reviewed for training impact and Simulator impact. This assures that both training and the simulator are continually evalue.ted and updated as plant changes occur. Procedural controls covering this review process are in Training Procedures not provided in this submittal. Plant Des!gn Changes requiring Simulator modifications are handled within the time allowed by ANS 3.5 Section 5.2 and 5.3. 1 l

O 32-

l' L

l l

- Student Feedback - Student (licensee) feedback is an important input to Simulator (m\ Fidellty. NSEM Procedure 6.01, " Student Feedback", describes how student feedback C/

L 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 7 contains the results of the survey performed in February,1989. The summary letter also contains the disposition of action to be taken for each item. The few items Indicated as open requiring further investigation will be investigated during 1989.

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.

As an example, on September 7,1988, Millstone 2 had a loss of Vital 120 volt bus VA20 at 100% power Data was obtained and the Simulator response was compared.

Results were excellent with a few minor exceptions. All exceptions have been corrected.

f~

- b} Development of New Simulator Training Guides - Simulator Certification Procedure NSEM 6.02, " Development of New Simulator Guides," covers requirements for new l Simulator training guidct This ensures that new Simulator training guides use only certified remote functions, certified malfunctions, certified initial Conditions and do not exceed any Simulator 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:

o Adding or deleting remote functions o Adding or deleting malfunctions o Changing remote functions or malfunctions o Changing Performance Tests or their criteria q

l l l 1 l

i

, it is Northeast Utilities' interpretation that simulator documentation may be modified Q 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 l l

l attached NSEM procedures. Updated materials will be sent at the next regular )

certification report date, or upon NRC request.

12. Open Deficiency Report (DR) List Attachment 8 contains a current listing of all Millstone 2 simulator open Deficiency l

Reports (DRs). This list contained 66 open DRs as of 3-28-89.

DRs have been placed in three categories, based on their importance.

A. The following 17 DRs will be cilspositioned

• by 5-1-91. These DRs have all been previously discussed in this submittal. The 17 DRs, by DR# are:

87-2-144,87-2-90,87-2-185,89-2-16,88-2-55,89-2-14,89-2-15,89-2-17, f3 88-2-77,87-2-28,88-2-115,88-2-138,88-2-140,88-2-114,89-2-18, U 88-2-122 and 89-2-41 * . Refer to Attachment 8 for the title that corresponds to a given DR number. These 17 DRs are the most important open DRs to be dispositioned

• based on their significance described in this submittal.

"Dispositioned" means 1 of 2 things will occur. Elther: 1) the DR will be fixed, or 2) If the DR cannot be fixed, the problem may be added to the Simulator Operating Limits ifit is significant. liis important io i recognize that prioritization of resolving DRs Is a dynamic process. As I 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. i l

• DR 89-2-411s not listed on the Attachment 8 DR printout. This DR was opened up after the 3-28-89 printout. Refer to Attachment 10, Transient

• 10, for an explanation of this DR.

O(N 1

- -- - - - - _- --_D

L (O

\'j B. The following 18 DRs will be dispositioned

• by 5-1-91. Some of these DRs may have been discussed in this submittal. These DRs have less impact

! than the above group of 17. The 18 DRs by DR # are 88-2-81,88-2-142, 88-2-109,89-2-26,88-2-125,88-2-12,88-2-23,88-2-40, 88-2-86,88-2-112, 89-2-20, 89-2-2,88-2-70,88-2-108,88-2-147,89-2-7,88-2-134,89-2-25.

l Refer to Attachment 8 for the title that corresponds to a given DR number.

DRs in this category do not have significant impact on training or exams.

C. The following 32 DRs will not have a specific date for being dispositioned

• based on the reasons stated below:

o 11 DRs are plant design changes which are in progress in the reference plant and are not yet operational as of the writing of this document.

They are, by DR#, 89-2-24,88-2 145,88-2-130,88-2 131,89-2-23, 88-2-132,88-2-133,89-2-21,89-2-22,88-2-127,88-2-143.

O V o 8 DRs are for deficiencies that have no impact on training from the students point of view. These DRs, by DR # are 88-2-100,89-2-5, 88-2-137,87-2-209,87-2-195,87-2-77,89-2-3 and 89-2-13, i

o Work on 2 DRs has been completed on the simulator, but they are awaiting administrative closcout. These DRs, by DR # are 87-2-66 ar'd 88-2-128.

"Dispositioned" means 1 of 2 things will occur. Etther: 1) the DR will be fixed, or 2) If the DR cannot be fixed, the problem may be added to the Simulator Operating Limits if it is significant, it is 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.

t v

1 o 11 DRs are of extremely minor impact to training and will be 73

() handled on a time available basis. These DRs by DR # are:

87-2-191, 88-2-85,88-2-90, 87-2-42,87-2-78,87 2-26, 87-2-220,89-2-12,89-2-19,89-2-27 and 89-2-28.

o Refer to Attachment 8 for the title of the DR that corresponds to the given DR number. These 32 DRs will be dispositioned

• by:

- Implementing the 11 DRs corresponding to pending plant design changes within ANS 3.5 guidelines once they are made operationalin the reference plant.

Resolving the balance of these 32 DRs as time permits.

Priortly would be given to the DRs listed above in Category .

A or B or the 11 DRs corresponding to design changes, if operational in the reference plant.

'v -

v

13. Next 4-Year Schedule,- (June 1989 to June 1993) .

The entire MP2 performance test will be repeated over a four-year interval -

as described in Attachment 9. The schedule shown in Attachment 9 has been written based on the guidance provided in NSEM Procedure 4.07, " Master Test Schedule". This 4-year Interval will start on the date of this submittal.

p.

,1 .

The following tests must be performed each year:

o Annual Operability Testing o Physical Fidelity Verification The following tests must be performed over a 4-year interval:

o Normal Plant Evolutions and Surveillance Testing o All System Tests J

o' All Certified Malfunctions o- Instructor Station Testing o Real Time Testing O

k

(

's ATTACHMENT.1 MP2 TURBINE SYSTEM TEST This attachment is referenced by Section 2 of the Performance Test Summary.

NRC #1 (41 O

4 Figure 7.5 1

SIMULATOR SYSTEM TEST COVER SHEET l

l SYSTEM TITLE Turbine Test ATTACHMENT 8.2.9 REV. O Unit-Sys i

UNIT 2 l

m ?n f .K 4A/A$

Developed By nhL '#f0f00

' Date pg , Y-2E'S$

// pproved By Date A(sistan/ pervisor Simulator V. Training i

i

- - Rev.: 0 Date: 04/18/88 Page: 7.5-1 of 1

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ - _ _ _ _ _ -)

ATTACHMENT 8.2.9 Turbina Test STEP PROCEDURE /RESULTS PANEL TAG #

V)' -

1.0 The following PPC points shall be verified during the course of this test 1.1 - Analog Points -

MN GEN STR WTR IN TEMP PPC TE005C '

MN GEN STR WTR OUT TEMP PPC TE005D MN TURB BRG #1 OIL TEMP PPC T4401 MN TURB BRG #2 OIL TEMP PPC T4402 MN TURB BRG #3 OIL TEMP PPC T4403 MN TURB BRG #4 OIL TEMP PPC T4404 )

MN TURB BRG #5 OIL TEMP PPC T4405  !

MN TURB BRG #6 OIL TEMP PPC T4406 i MN TURB BRG #7 OIL TEMP PPC T4407  :

MN TURB BRG #8 OIL TEMP PPC T4408 MN TURB'BRG #9 OIL TEMP PPC T4409 MN TURB BRG #10 OIL TEMP PPC T4410  :

MN TURB THRUS BRG TEMP PPC T4411 MN TURB THRUS BRG TEMP PPC T4412 MN TURB THRUS BRG TEMP PPC T4413 MN TURB THRUS BRG TEMP PPC T4414 MN TURB THRUS OIL TEMP PPC T4608 MN TURB THRUS OIL TEMP PPC T4609 f MN TURB L/O CLR OUT TEMP PPC T7208

.fr "g P7210 MN TURB L/O CLR INLT PPC t,,_) PPC SE439 MN TURB SPEED MN TURB BRG #1 VIB PPC X4391 MN TURB BRG #2 VIB PPC X4392 MN TURB BRG #3'VIB PPC X4393 4 MN TURB BRG #4 VIB PPC X4394 MN TURB BRG #5 VIB PPC X4395 MN TURB BRG #6 VIB PPC X4396 MN TURB BRG #7 VIB PPC X4397 MN TURB BRG #8 VIB PPC X4398 MN TURB BRG #9 VIB PPC X4399 MN TURB BRG #10 VIB PPC X4400 1.2 Digital Points -

TURBINE LO HYD FLUID PRES PPC Z4352 MN TURB BKUP OVERSPEED PPC Z4354 MN TURB SPEED LOSS SIG PPC Z4355 MN TURB OVERSPEED PPC Z4356 EHC HYD PP A AUTO PPC Z4359A EHC HYD PP B AUTO PPC Z4359B MN TURB HYD TRANSFER PP PPC Z4386 MN TURB CONDR VAC PPC Z4386 MN TURB MASTER RELAY PPC Z4389 MN TURB MANUAL PPC 24489 l

n 1b Rev.: 0 Date: 4/18/88 Page: 8.2.9-1 of 47 NSEM-4.01 i

STEP PROCEDURE /RESULTS PANEL TAG #

\# MN TURB-START-UP ACC SLO PPC Z4501 MN TURB-START-UP ACC MED PPC Z4502 MN TURB-START-UP ACC FAS PPC Z4503 MN TURB EHC SYSTEM PPC ZE430 TURB HYD FLUID PP A PROT PPC ZE814A TURB HYD FLUID PP B PROT PPC ZE814B MN TURB THRUS BRG WEAR PPC Z4350 MN TURB-LO SPEED HOLD PPC ZE504 -

MN TURB-HI SPEED-HOLD I PPC Z4505 MN TURB-RATED SPEED HOLD PPC Z4506 MN TURB-PRI. SPEED LOSS PPC Z4508 MN TURB-SEC SPEED LOSS PPC Z4509 MN TURB-REM AUTO LOAD PPC Z4513 MN TURB-LOAD LIMIT PPC Z4514 MN TURB-EHC 24 VDC PPC Z4515 MN TURB STOP VV 4 PPC 2MS60A#

MN TURB STOP VV 4 PPC 2MS60A=

MN TURB STOP VV 3 PPC 2MS60B#

MN TURB STOP VV 3 PPC 2MS60D=

MN TURB STOP VV 2 PPC 2MS60C#

MN TURB STOP VV 2 PPC 2MS60C= l MN TURB STOP VV 1 PPC 2MS60D#

MN TURB STOP VV 1 PPC 2MS60D=

MN TURB CNTL VV 4 PPC 2MS61A#

MN TURB CNTL VV 4 PPC 2MS61A=

MN TURB CNTL VV 2 PPC 2MS61B#

7)-

(_ MN TURB CNTL VV 2 PPC 2MS61B-MN TURB CNTL VV 3 PPC 2MS61C#

MN TURB CNTL VV 3 PPC 2M561C=

MN TURB CNTL VV 1 PPC 2MS61D#

MN TURB CNTL VV 1 PPC 2MS61D=

MN TURB INTERCEPT VV 3 PPC 2MS62A#

MN TURB INTERCEPT VV 3 PPC 2MS62A=

MN TURB REHEAT STOP VV 3 PPC 2MS62Al#

MN TURB REHEAT STOP VV 3 PPC 2MS62Al=

MN TURB INTERCEPT VV 2 PPC 2MS62Bt MN TURB INTERCEPT VV 2 PPC 2MS62B=

MN TURB REHEAT STOP VV 2 PPC 2MS62Bl#

MN TURB REHEAT STOP VV 2 PPC 2MS62Bl=

MN TURB INTERCEPT VV 4 PPC 2MS63A#

MN TURB INTERCEPT VV 4 PPC 2MS63A=

MN TURB REHEAT STOP VV 4 PPC 2MS63Al#

MN TURB REHEAT STOP VV 4 PPC 2MS63Al=

MN TURB INTERCEPT VV 1 PPC 2MS63B#

MN TURB INTERCEPT VV 1 PPC 2MS63B= .

l MN TURB REHEAT STOP VV 1 PPC 2MS63Bl#

MN TURB REHEAT STOP VV 1 PPC 2MS63Bl=

TURB HYD FLUID RSVR LVL PPC L7289 TURB HYD FLUID RSVP LVL PPC L7291 EHC HYD PP A RUNNING PPC P108A Rev.: 0 Date: 4/18/88 Page: 8.2.9-2 of 47 NSEM-4.01

STEP. PROCEDURE /RESULTS PANEL TAG G EHC HYD PP B RUNNING PPC P108B EHC HYD FLUID'FLTR DP. PPC P4693 MN TURB HYD FLUID PRES PPC P7056' MN.TURB HYD FLUID TEMP- PPC T7290

.TURB TRIP ON RX/ GEN; PROT PPC ZE184

TURB RNBK - PPC ZE297 TURB.EHC EMER PRES PPC ZE429 MN TURB BRG LIFT PP A PPC Z4375 MN TURB BRG LIFT PP B PPC Z4376A MN TURB BRG LIFT PP C PPC Z4376B MN TURB.BRG, LIFT PP D PPC Z4377 MN TURB BRG LIFT PP E PPC Z4378 MN TURB'BRG LIFT PP F PPC Z4379 SHAFT VOLT-TEST DEVICE' PPC Z4380.

MN TURB HPU HTRS FAN PPC Z4384

'NEW/ DIRTY.L/O TK.LVL PPC L7152

.MN TURB L/O RSVR LVL PPC L7195 BOWSER CLEAN OIL LVL PPC L7206 MN TURB BRG L/O RSVR LVL PPC L7212 BOWSER CLEAR OIL LVL PPC L8996 MN TURB LIFT PP PRES PPC P4638 MN TURB EMER BRG OIL PP PPC P7199 MN TURB TGR-OIL PP PPC P7200 MN TURB MTR SUCT PP PPC P7201 MN TURB BRG L/O'HDR PRES PPC P7215 MN TURB VIB PPC ZE428 MN TURB TSI PNL'. VOLT PPC ZE431 t

O/ '- L MN.TURB EXPANSION PPC. ZE493 MN TURB QUILL SHAFT- PPC ZE791 MN TURB MTR-SUCT PP PROT PPC ZE810 MN TURB EMER OIL PP PROT PPC ZE811

'MN TURB TGR OIL PP PROT PPC ZE812 OIL VAPOR EXT PP PROT PPC ZE813

' MN TURB SHAFT PP PRES PPC Z4348 MN TURB THRUS BRG WEAR PPC Z4350 MN TURB HI VIB PPC Z4353 i MN TURB TGR PPC Z4360 MN TURB OIL VAPOR EXT PP PPC Z4363 2.0 The following Control Panel Annunciators shall be verified during the course of this test HYDRAULIC FLUID PUMP A AUTO START CO6/7 A31 HYDRAULIC FLUID RSVR LEVEL HI/LO CO6/7 A32 HYDRAULIC FLUID TEMP HI/LO CO6/7 A33 EHC EMER PRESS SWITCHES TURB TRIP CO6/7 A35 TURBINE THRUST BRG WEAR TRIP CO6/7 B27 HYDRAULIC FLUID PUMP A CO6/7 B31 OVERLOAD / TEMP HI Rev.: 0 Date: 4/18/88 )

Page: 8.2.9-3 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

[)

\/

HYDRAULIC FLUID.RSVR LEVEL HI/HI CO6/7 B32 EHC MONITOR PANL TROUBLE CO6/7 B33 "A" EHC PUMP RUNNING CO6/7 BA24 "B" EHC PUMP RUNNING CO6/7 BB24 HYDRAULIC FLUID PUMP B AUTO. START CO6/7 C31 l -, HYDRAULIC FLUID PRESSURE LO CO6/7 C32 TSI PANEL VOLTAGE LO CO6/7 C33 LOSS OF SPEED SIGNAL TRIP CO6/7 C35 HYDRAULIC PUMP FILT DELTA P HI CO6/7 D29 HYDRAULIC FLUID PUMP B- CO6/7 D31 OVERLOAD / TEMP HI HYDRAULIC' FLUID PRESS LO-LO TRIP CO6/7 D32 OVERSPEED TRIP CO6/7 D35 BACKUP OVERSPEED TRIP CO6/7 A36 TURBINE' MASTER RELAY TRIP CO6/7 B36 j CONDENSER VACUUM LO-LO TRIP CO6/7 B37 1 REACTOR / GEN PROT TURBINE TRIP CO6/7 C48 l MOTOR SUCTION OIL PUMP RUNNING A25 CO6/7 TURB ROTOR STOPPED CO6/7 A26 TURBINE SHAFT PUMP' PRESS LO-LO CO6/7 A27 TRIP TURBINE-BRG OIL PRESSURE LO CO6/7 A28 TURBINE BRG OIL RSVR LEVEL HI/LO CO6/7 A29 TURBINE BRG OIL VAPOR EXTRACTOR CO6/7 A30 NOT RUN TURBINE VIBRATION HI CO6/7 A34

' /'N . MOTOR SUCTION OIL PUMP CO6/7 B25

,(_) OVERLOAD / TEMP HI TURNING TCR QUILL SHAFT FAILURE CO6/7 B26 TURBINE BRG LIFT PUMP PRESSURE LOW. CO6/7 B28 TURBINE BRG OIL RSVR LEVEL HI-HI CO6/7 B29 TURBINE BRG OIL VAPOR CO6/7 B30 EXTRACTOR OVERLOAD TURBINE VIBRATION HI-HI TRIP CO6/7 B34

. TURBINE BRG EMER OIL PUMP RUNNING CO6/7 B25 TURBINE TGR OIL PUMP RUNNING CO6/7 C26 BOWSER CLEAN OIL COMPARTMENT CO6/7 C27 LEVEL LO BOWSER FILTER BAG COMPARTMENT CO6/7 C28 LEVEL HI TURBINE BRG EMER OIL PUMP OVERLOAD CO6/7 D25 TURBINE TGR OIL PUMP OVERLOAD / CO6/7 D26 TEMP HI BOWSER OIL FILTER FLOW LO CO6/7 D27 NEW/ DIRTY TURBINE OIL STORAGE CO6/7 D28 TANR LEVEL HI TURBINE' EXPANSION ABNORMAL CO6/7 D33 TURBINE MOISTURE SEP. LEVEL CO6/7 B35 HI HI TRIP CONDENSER VACUUM LO CO6/7 A37 (1) {

Rev.: 0 Date: 4/18/88 Page: 8.2.9-4 of 47 NSEM-4.01 e i

i i f

STEP PROCEDURE /RESULTS PANEL TAG #

,o CO6/7 C36 i *: H2 SEAL OIL STATOR CLG CABINET ,

'- TROUBLE LOSS OF STATOR COOLANT TRIP CO6/7 D36 I TURBINE EXHAUST HOOD TEMP HI CO6/7 C34 {

i TURBINE EXHAUST HOOD TEMP HI HI TRIP CO6/7 D34 I

TURBINE GENERATOR' LOCKOUT TRIP CO6/7 A48 CHANNEL 1 TURBINE GENERATOR LOCKOUT TRIP CO6/7 B48 CHANNEL 2 EXCITER FIELD BREAKER TRIP CO6/7 A45

-GENERATOR FIELD BREAKER TRIP CO6/7 A46 3.0 Reset simulator to any IC with PCM Rx TCB's open 4'.0 verify turbine will not reset C07 TRPRESETPB when the Reset Pushbutton is depressed for at least 5 seconds 5.0 Reset simulator to Rx critical, PCM MSIV's closed 6.0 ' Verify the following.EHC Panel lineups prior to resetting the Turbine " Closed Valves" CO7 PBCLSV selected as speed set o Chest warming in "Off" CO7 CHST/SHLLOFF with its associated CO7 CHST/SHLLPT Potentiometer at minimum o Load Limit Potentiometer C07 LOADLMTPDT set at zero Load set at zero C07 LOADSETMTR o

o Initial pressure "Off" C07 TPLOFF and set at zero percent C07 TPLPT o First Stage Pressure C07 SPF0FF Feedback "Off" and set C07 SPFPT at zero (out) ,

Rev.: 0 Date: 4/18/88 Page: 8.2.9-5 of 47 NSEM-4.01

f

- STEP PROCEDURE /RESULTS PANEL TAG #

/ 7.0 Verify that NSST and Main L- Generator Lockout Devices

, are. reset with their associated handswitches C07R-NOTE: If the Reset Push-button is prematurely released, the reset is aborted and all trip alarms are reactivated 8.0 Reset the Emergency Trip System by depressing the Reset Pushbutton for CO7 TRPRESETPB at least 5 seconds o Mechanical and C07 MECHTRIP (G)

Emergency Trip C07 MECHRESET (R)

Systems reset as indicated by o Trip lights deenergizing C07 EMERGTRIP (G) and reset lights energicing. CO7 EMERGRST (R)

(Trip out after 1 sec/ reset on af ter 5 see)

-( o Mechanical Trip Resetting Light on momentarily C07 MECHRESETTIN (A) during reset (o 2 see after being dep.essed/

off 3 see after release o " Turbine Master Relay CO6/7 B36 Trip" alarm clears PPC Z4389 o "Overspeed Trip" alarm CO6/7 D35

. clears PPC Z4356 o "EHC Emer Press Switches CO6/7 A35 Turb Trip" alarm clears PPC ZE429 o Control valves remain shut CO7 CVNO.lPOS PPC 2MS61D#

C07 CVNO.2POS PPC 2MS61Bt C07 CVNO.3POS PPC 2MS61C#

C07 CVNO.4POS PPC 2MS61A#

1 l

O Rev.: O i

l I Date: 4/18/88 l Page: 8.2.9-6 of 47 NSEM-4.01

1 STEP. PROCEDURE /RESULTS PANEL TAG #

f J. o Stop Valves remain shut 'C07 MSVNO.1POS

' ' ' PPC- 2MS60D#

C07 MSVNO.2POS l PPC 2MS60C#

C07 MSVNO.3POS PPC 2MS60Bt  ;

C07 MSVNO.4POS PPC 2MS60A#

o Intercept Valves remain C07 IV-1POS shut PPC 2MS63Bt C07 IV-2POS PPC 2MS62B#

C07 IV-3POS PPC 2MS62Af C07 IV-4POS PPC 2MS63A#

o Intermediate Stop Valves CO7 ISV-1POS open PPC 2MS6391 i I

C07 ISV-2POS PPC 2MS62B1 C07 ISV-3POS PPC 2MS62A1 207 ISV-4POS PPC 2MS63Al O 9.0 Trip the Turbine by depressing C07 TRPPB the Turbine-Trip Pushbutton o Mechanical and Emer- C07 MECHRESET (R) gency Trip Systems trip C07 MECHTRIP (G) as indicated by reset C07 EMERGTRIP (G) !

lights deenergizing and PPC Z4489 trip' lights energizing o " Turbine Master Relay CO6/7 B36 Trip" alarms PPC Z4389 PPC ZE184 o "Overspeed Trip" alarms CO6/7 D35 PPC Z4356 o "EHC Emer Press Switches CO6/7 A35 Turb Trip" alarms PPC ZE429 ks. l Rev.: 0 Date: 4/18/69 Page: 8.2.9-7 of 47 NSEM-4.01 i

l-I STEP PROCEDURE /RESULTS PANEL TAG #

'f

, ~.I . o Intermediate Stop Valves CO7 ISV-1POS

'- shut PPC 2MS63B1 C07 ISV-2POS PPC 2MS62B1 C07 ISV-3POS o TCB's remain shut PPC 2MS62A1 CO7 ISV-4POS PPC 2MS63Al 10.0 Reset simulator to Rx at 3% PCM

. power, main steam lines warmed, condenser vacuum established 11.0 Prepare for Turbine Startup ,

as follows:

11.1 Verify that Combined Control valve before Seat Drain Valve (HV-4039A) C07 HS-4309A and Control Valve Lead CO7 HS-4309B Drain Valves (HV-4309C, C07 ZI-4310/4311 4309D, 4310 and 4311) are open. HS indication shows

("

s all valves open. HV-4310 and HV-4311 only have position indication 11.2 Verify an EHC Pump is C07 HS-4466 running, with one.in C97 HS-4467 standby with handswitch in auto position 11.3 verify " Hydraulic Fluid CO6/7 C32 Pressure Lo" and " Hydraulic CO6/7 D32 Fluid Pressure Lo-Lo Trip" alarms clear 12.0 verify the following EHC Panel CO7 EHC PANEL lineup prior to resetting the turbine o "Close Valves" selected C07 PBCLSV as speed set O Rev.: 0 Date: 4/18/88 Page: 8.2.9-8 of 47 NSEM-4.01

1 i

STEP PROCEDURE /RESULTS PANEL TAG 4 A C07 CHST/SHLLOFF t,- ) o Chest warming in "OFF" with its associated CO7 CHST/SHLLPT Potentiometer at minimum 1 o Load Limit' Potentiometer C07 LOADLMTPDT set at zero o Load set at zero CO7 LOADSETMTR {

o Initial Pressure Limiter C07 TPLOFF in "OFF" with its CO7 TPLPT associated Potentiometer at minimum (out) o First Stage Pressure C07 SPF0FF Feedback in "OFF" SPFPT 1 i

with its associated Potentiometer at minimum (out) 13.0 Reset the Turbine Emergency CO7 TRPRESETPB Trip system by momentarily depressing the " Reset" push-

-button at the EHC Panel NOTE: The Reset Pushbutton must be held depressed for at least 3 seconds to reset trip system 13.1 The following annunciators will clear in the order listed:-

"EHC Emer Press CO6/7 A35 Switches Turb Trip" PPC ZE429

- " Turbine Master Relay CO6/7 B36 Trip" PPC Z4389

- "Overspeed Trip" CO6/7 D35 PPC Z4356

- Mechanical and Emergency Trip Systems reset as indicated by their associated trip lights extinguishing and reset i lights illuminating Rev.: 0 Date: 4/18/88 Page: 8.2.9-9 of 47 NSEM-4.01

i l'

STEP PROCEDURE /RESULTS PANEL TAG #-

All Control Valves, Stop

- /^N) .

l 1,

-- valves and Intercept Valves i remain shut

- Intermediate Stop Valves C07 ISV-lPOS ISV-1,2,3 and 4 open PPC 2MS63B1 C07 ISV-2POS PPC 2MS62B1

' C07 ISV-3POS PPC 2MS62Al C07 ISV-4POS PPC 2MS63Al l 14.0 Isolate drain valves necessary PCM MSR11 for turbine pre-warming (IAW Ops Form 2323A-1) with associated Remote Function 15.0 set the Load Limit Potentiometer C07 LOADLMTPOT at its maximum value (10) 16.0 Select " FAST" starting rate FASTRATEPB

> o Associated light illuminates FASTRATEPB(W)

Z4503 k PPC 17.0 Select shell warming with Chest /Shell Warming Push-button at EHC Panel o Shell Warming' Light illuminates o All Control Valves open C07 CVNO.1POS fully PPC 2MS61D#

C07 CVNO.2POS PPC 2MS61B#

C07 CVNO.3POS PPC 2MS61C#

C07 CVNO.4POS PPC 2MS61A9 O Rev.: 0 Date: 4/18/88 Page: 8.2.9-10 of 47 NSEM-4.01 4

i STEP PROCEDURE /RESULTS PANEL TAG #

,[~h' o All Intermediate Stop CO7 ISV-1POS N/ Valves go closed PPC 2MS63B1 .

C07 ISV-2POS l PPC 2MS62B1 C07 ISV-3POS PPC '2MS62A1  ;

C07 ISV-4POS PPC 2MS63Al 18.0 Slowly position the Chest /Shell C07 CHST/SHLLPT Potentiometer to_open #2 Main Stop Valve Bypass to pressurize and warmup the High Pressurt Turbine. Pressure should be increased to 60-100 PSIG as indicated by intermediate pressure o Intermediate pressure C07 PI-INTPRES60-100 PSIG o Differential Expansion C07 UR-4500(P1) controllable within UR-4500(P2) normal operating band UR-4500(P3) '

o Off light extinguishes C07 CST /SHLLOFF o Main Turbine Steam Chest PPC P4321 Pressure, H.P. Turbine PPC P4298

-First Stage Pressure and PPC T4315 HP. Turbine to the M.S.R.'s PPC T4316 Temperature and pressure PPC P4167 reflect the turbine warmup PPC P4169 PPC P4176 PPC P4178 PPC T4166 PPC T4168 PPC T4177 PPC T4179 o MSR temperatures and MSR PPC T4606 steam temperatures and PPC T4607 pressures reflect warmup PPC P4161 of MSR's PPC P4162 PPC P4172 PPC P4174 PPC T4155 CO7 UR-4500(P13)

PPC T4160 C07 UR-4500(P12)

Rev.: 0 Date: 4/18/88 Page: 8.2.9-11 of 47 NSEM-4.01

-_-_-___g -. _ _ _ . . . _ . - _ . _ - _ - - _ _ - _ _ _ _ - _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ , _ - .

STEP , PROCEDURE /RESULTS PANEL TAG #

PPC T4173 (y -

C07 'UR-4500(P10)

PPC -T4175 C07 UR-4500(P11) o Uniform temperatures.will. PCM be achieved in-approximately four. hours.. Turbine warmup fast time may be utilized as desired 19.0 rirst-Stage-Shell lower inner C07 UR-4500(P4) temperature (point.4) and C07 UR-4500(P6)

External Valve Chest inner C07 PI-INTPRES temperatures (point 6) increase to.near saturation values for H

-existing intermediate pressure 20.0. First stage Shell lower outer C07 UR-4500(PS)

(point 5).and External Valve C07 UR-4500(P7)

Chest-outer temperatures (point 7) increase r~ 21.0- Util'ze i Fast Time Turbine PCM

( ' Warmup function as necessary to achieve stable and-acceptable turbine differential and shell-temperatures 22.0 Stop shell warming by shutting C07 CHST/SHLLPT

1. 2 Main Stop Valve Bypass with the. Chest /Shell-Potentiometer o Intermediate pressure CO7 PI-INTPRES decreases when valve is shut o' Off light illuminates CO7 CHST/SHLLOFF 23.0 Turn-off Chest /Shell Warming CO7 CHST/ShaLOFF by depressing its "OFF" pushbutton o Main Turbine Control valve close O Rev.: 0 Date: 4/18/88 Page: 8.2.9-12 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

!yA-} o Intermediate Stop Valves open o Chest warming light CO7 CHSTWARM illuminates 24.0. Unisolate Turbine Drain valves PCM MSRll with associated Remote Function 25.0. Set the' Load Limit Potentiometer C07 LOADLMTPOT at one (1) 26.0 Select Chest Warming by depressing C07 CHST/SHLLPB Chest /Shell pushbutton o Chest Warming light illuminates 27.0 Increase Steam Chest pressure PPC P4321 by opening #2 Main stop Valve PPC T4317 Bypass with the Chest /Shell Potentiometer f']

o Increase Steam Chest C07 UR-4500(P6) pressure to rated PPC T4318

-temperature and' pressure C07 UR-4500(P7) observing inner and outer differential temperature.

limitations on the Steam Chest o Fast Time Turbine Warmup C07 UR-4500 may be used to achieve a stable and acceptable shell metal temperature 28.0 Stop Chest warming by returning C07 CHST/SHLLPOT the Chest /shell Warming Potentiometer to minimum o "OFF" light illuminates when Warming Potentiometer is placed at the fully counter-clockwise.(CCW) position l(

Rev.: 0 Date: 4/18/88 Page: 8.2.9-13 of 47 NSEM-4.01

-STEP PROCEDURE /RESULTS PANEL TAG #

/~N-i

\~';- 29.0- Increase reactor power to 5%

on the Steam Dump Bypass Valve 30.0 Test'for proper operation of the Motor Suction Pump / Hand-switch as follows:

o Place Motor Suction Pump C07 HS-7191 Handswitch in the start ,

position. Verify red l light illuminates o Motor Suction Pump starts automatically when hand-switch'is dropped down from Pull-To-Lock position o " Motor Suction Pump Running" CO6/7 A-25 ,

annunciator received PPC P7201 i o Place Motor Suction Pump CO6/7 A25 Handswitch in the Pull- PPC P7201 To-Lock position, green  !

light extinguishes, i

" Motor Suction Pump fNs. Running" annunciator clears

\~sl ,

1 31.0 Select " Medium" acceleration C07 MEDRATEPB rate by depressing the " Medium" PPC Z4502 pushbutton o Associated pushbutton light illuminates while previously selected rate light extinguishes 32.0 Roll the turbine by momentarily PPC Z4504 depressing the "100" Speed Set C07 PB100 RPM pushbutton o #2 Main Stop Valve opens C07 MSVNO.lPOS fully followed by the C07 MSVNO.2POS opening of the remaining C07 MSVNO.3POS Main Stop Valves CO7 MSVNO.4POS 1 O Rev.: 0 Date: 4/18/88 Page: 8.2.9-14 of 47 NSEM-4.01 L

STEP PROCEDURE /RESULTS PANEL TAG #

~

o #1 and #2 Intercept valves C07 IV-lPOS

~/ open fully followed by the C07 IV-2POS opening of #3 and #4 C07 IV-3POS Intercept Valves C07 IV-4POS o All four Main Control C07 CVNO.lPOS Valves open to bring the C07 CVNO.2POS turbine up to 100 RPM at C07 CVNO.3POS a 90 RPM acceleration rate C07 CVNO.4POS o " Increasing Speed" speed C07 LPINCSPD monitoring light illuminated C07 LPATSETSPD until turbine is at 100 RPM.

When at 100 RPM, the

" Increasing Speed" light extinguishes and "AT SET SPEED", light illuminates o Turning Gear autmaatically CO7 HS-4463 disengages when the turbine PPC Z4360 is rolled as observed by the Turning Gear Engaged Light Deenergizing o As the turbine warms the rotors lengths and position

,r's will change. The rotor

(' " '

) growth will be different than the shell growth which will be reflected as Differential Expansion o Outer Shell and Steam Chest temperatures will increase at a slower rate than the inner temperatures until equilibrium points are reached when pressures stabilize o Turbine Bearing Oil Drain PPC T4401 thru temperatures increase T4410 o Turbine Bearing temperatures PPC T4411 begin to increase PPC T4412 PPC T4413 PPC T4414

- j G

Rev.: 0 Date: 4/18/88 ,

Page: 8.2.9-15 of 47 )

NSEM-4.01 1

i STEP PROCEDURE /RESULTS PANEL TAG #

[' o Main Turbine Lube Oil PPC T7208 k~ Cooler Outlet temperature maintained at approximately 100 degrees farenheit o Turbine speed held at C07 SI-TURB 100 RPM PPC SE-439 33.0 Stop the Turning Gear Motor PPC Z4360 by momentarily placing its handswitch in the "Stop" position and then allow it to spring return to " Auto" o Red light extinguishes, green light illuminates when in auto o Turning Gear remains off PPC Z4360 when in auto 34.0 Select " SLOW" acceleration C07 SLORATEPB rate by depressing the " SLOW" PPC Z4501 pushbutton I)'-'

o Associated Pushbutton Light illuminates while the previously selected rate light extinguishes 35.0 Increase turbine speed to 800 RPM CO7 PB800 by momentarily depressing the PPC SE-439 "800" Speed Set pushbutton o Speed increases to 800 RPM at selected rate o "AT SET SPEED" light deenergizes until unit reaches 800 RPM o Increasing speed light illuminated o "800" Speed Set light energizes while previously selected Speed Set light extinguishes

, (~s Rev.: 0 Date: 4/18/88 Page: 8.2.9-16 of 47 NSEM-4.01

i

-STEP PROCEDURE /RESULTS PANEL TAG #

Turbine speed held at 800 RPM

("}

k-o PPC C07 Z4504 SI-TURB l

36.0 Select fast acceleration rate by CO7 FSTRATEPB depressing the " FAST" pushbutton o Turbine speed will increase at 180 RPM o Associated pushbutton light illuminates while the previously selected rate light extinguishes 37.0 Increase turbine speed to 1500 RPM CO7 PB 1500 by momenarily depressing the "1500" Speed set pushbutton o Speed increases to 1500 RPM at 180 RPM rate o "AT SET SPEED" light extinguishes until unit is at 1500 RPM o "1500" Speed Set light illuminates while previously selected speed set light extinguishes NOTE: Critical speed between C07 UR-4502 BYU and 1000 RPM will cause vibration levels on Generator Bearing to increase.

Cascading effect observed on remaining bearings NOTE: Critical speed between C07 UR-4502 T7UU and 1400 RPM will cause vibration levels on HP Turbine Bearings to increase.

Cascading effects observed on remaining bearings.

o Turbine speed held at 1500 RPM CO7 SI-TURB o High speed hold at 1500 RPM PPC Z4505 Rev.: 0 Date: 4/18/88 Page: 8.2.9-17 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

LN--

[') .o Wobbulator Varies speed PPC SE439

+/-50 RPM at a 6 min half cycle 38.0 Increase turbine speed to 1800 RPM C07 PB1800 by momentarily depressing the "1800" RPM Speed Set pushbutton o Speed increases to 1800 RPM at selected rate o Recorder follows speed C07 UR4501 indication o "AT SET SPEED" light extinguishes until at 1800 RPM o "1800" Speed Set light illuminates while previously selected speed set light extinguishes o Turbine speed held at C07 SI-TURB 1800 RPM I o Rated speed hold at PPC Z4506 1800 RPM o PMG malfunction switch light illuminated o "EHC Monitor Panel Trouble" CO6/7 B33 annunciator in alarm 39.0 Take a snapshot for subsequent PCM use 40.0 Utilize remote function to PCM TCR15 reset PMG malfunction light on EHC Panel, observe:

o "EHC Monitor Panel CO6/7 B33 Trouble" anunciator clears

,f) l \~J Rev.: 0 l

Date: 4/18/88 Page: 8.2.9-18 of 47 NSEM-4.01

STEP ~ PROCEDURE /RESULTS PANEL TAG #

' 41.0 Verify auto start functions k- . of EHC Pumps 41.1 Place 'A' EHC Pump Hand- C07 HS-4466 switch in run and obseve:

"'A' EHC Pump Running" CO6/7 BA24 annunciator alarms. PPC P108A Red light illuminates, C07 HS-4466(R) green light extinguishes, CO7 HS-4466(G) white light remains off CO7 HS-4466(W)

'EHC pressure by computer PPC P7056  !

indicates normal pressure 41.2 Place 'B' EHC Pump Handswitch C07 HS-4467 in auto position and observe:

"'B' EHC Pump Running" CO6/7 BB24 annunciator clear- PPC P108B Red light extinguished, CO7 HS-4467(R) green light illuminates, C07 HS-4467(G) white light illuminates C07 HS-4467(W) s 41.3 Stop the 'A' EHC Pump by C07 HS-4466 q~') placing its handswitch in PPC P108A off and observe:

"'Ar EHC Pump Running" CO6/7 BA24 annunciator clears Red light extinguishes, C07 HS-4466(R) green light illuminates, CO7 HS-4466(G) white light extinguished CO7 HS-4466(W)

" Hydraulic Fluid Pressure CO6/7 C32 Lo" annunciator alarms as PPC P7056 pressure decreases below 1300 PSIG

'B' EHC Pump starts in PPC P108 B auto when EHC pressure decreases below 1300 PSIG

" Hydraulic Fluid Pump 'B' CO6/7 C31 Auto Start" annunciator PPC Z43598 alarms 1

Rev.: 0 Date: 4/18/88 Page: 8.2.9-19 of 47 NSEM-4.01

LSTEP PROCEDURE /RESULTS PANEL TAG #

! -- "'B' EHC Pump Running" CO6/7 BB24 1s , . annunciator alarms-Red light illuminates, CO7 HS-4467(R) green light extinguishes, C07 HS-4467(G) white light illuminated CO7 HS-4467(W)

-. " Hydraulic Fluid Pressure CO6/7 C32 q Lo" annunciator clears PPC P7056 )

when-EHC pressure increases above 1300 PSIG-41.4 Place 'B' EHC Pump Handswitch in run and observe:

C07 HS-4467 ]

" Hydraulic fluid Pump 'B' CO6/7 C31 i Auto Start" annunciator PPC Z4359B clears White: light extinguishes C07 HS-4467(W) 41.5 Place 'A'.EHC. Pump Handswitch C07 HS-4466 in auto and observe:

White light illuminates C07 HS-4466(W) l 41.6 stop theB' EHC Pump by C07 HS-4467 placing its handswitch in the stop position and observe:

"'B' EHC Pump Running" CO6/7 BB24 annunciator clears PPC P108B Red light extinguishes, C07 HS-4467(R) green light illuminated, CO7 HS-4467(G) white light extinguished C07 HS-4467(W).

" Hydraulic Fluid Pressure CO6/7 C32 Lo" annunciator alarms as PPC P7056 pressure decreases below 1300 PSIG

'A' EHC Pump Starts in PPC P108A auto when EHC pressure decreases below 1300 PSIG

" Hydraulic Fluid Pump 'A' CO6/7 A31 Auto Start" annunciator PPC Z4359A alarms O Rev.: 0 Date: 4/18/88 Page: 8.2.9-20 of 47 NSEM-4.01

m;

' STEP PROCEDURE /RESULTS PANEL TAG #

j

"'A' EHC Pump Running" CO6/7 BA24 annunciator alarms Red light illuminates, C07 HS-4466(R)

l. green light extinguishes, CO7 HS-4466(G) p white 2ight illuminated CO7 HS-4466(W) 1

" Hydraulic Fluid Pressure CO6/7 C32 Lo" annunciator clears PPC P7056 when EHC pressure increases above 1300 PSIG 41.7 Place 'A' EHC Pump Handswitch C07 HS-4466 in run and observe:

White light extinguished CO7 HF-4466(W)

" Hydraulic Fluid Pump 'A' CO6/7 A31 Auto Start" annunciator PPC Z4359A clears 41.8 Place 'B' EHC Pump Handswitch C07 HS-4467 in auto and observe:

White light illuminates CO7 HS-4467(W)

?~

42.0 Test the Standby EHC Pump auto start capability as follows:

42.1 start the 'B' EHC Stby PCM TCR02 Pump using Remote Function TCR02 in Cycle 2 position for 'B' EHC Pump 42.2 Verify the Standby EHC Pump starts by observing:

"'B' EHC Pump Running" CO6/7' BB24 annunciator alarms PPC P108B

" Hydraulic Fluid Pump CO6/7 C31

'B' Auto Start" PPC Z4359B annunciator alarms Red light illuminates, CO7 HS-4467(R) green light extinguishes, C07 HS-4467(G) white light illuminated CO7 HS-4467(W)

O Rev.: 0 Date: 4/18/88 Page: 8.2.9-21 of 47 NSEM-4.01

(__

l

. STEP PROCEDURE /RESULTS PANEL TAG #

l s-t 42.3 Return Remote Function to PCM TCR02

' As to " Pres" position

'B' EHC pump remains running 42.4 Place 'B' EHC Pump Hand- C07 HS-4467 switch in run position, HS-4466

! place 'A' EHC Pump Hand-switch in "OFF" then " AUTO",

observe:

" Hydraulic Fluid Pump 'B' CO6/7 C31

-Auto Start" annunciator PPC Z4359B clears

"'A' EHC Pump Running" .CO6/7 BA24 anunciator clears PPC P108A 42.5 Start the 'A' EHC Pump PCM TCR02 using remote function in Cycle:1 position 42.6 verify the Standby EHC Pump starts by observing:

e' -

"'A' EHC Pump Running" CO6/7 BA24

( annunciator alarms PPC P108A

" Hydraulic Fluid Pump CO6/7 A31

'A' Auto Start" PPC Z4359A Red light illuminates, C07 HS-4466(R) green light extinguishes, C07 HS-4466(G) white light illuminated C07 HS-4466(W) 42.7 Return Remote Function to PCM TCR02 j

" Pres" position 42.8 Place 'A' EHC Pump Handswitch C07 HS4466  !

in 'Run' position, place 'B' C07 HS4467 EHC Pump handswitch in "OFF" then " AUTO", observe:

" Hydraulic Fluid Pump 'A' CO6/7 A31 Auto Start" annunciator PPC Z4359A clears

"'B' EHC Pump Running" CO6/7 BB24 annunciator clears PPC P108B O Rev.: 0 Date: 4/18/88 Page: 8.2.9-22 of 47 NSEM-4.01

- - - - _ _ _ _ _ _ - - _ _ _ l

l STEP PROCEDURE /RESULTS PANEL TAG #

TC09A

(~) 42.9 Insert malfunction to trip the 'A' EHC pump, observe:

PCM

"'A' EHC Pump Running" CO6/7 BA24 annunciator clears PPC P108A

" Hydraulic Fluid Pump CO6/7 B31

'A' Overload / Temp Hi" PPC ZE814A annunciator alarms

'A' EHC pump handswitch C07 HS-4466(R) red light remains energized

'B' EHC pump starts 42.10 Remove malfunction, observe: PCM TC09A

'A' EHC pump starts

" Hydraulic Fluid Pump CO6/7 D31  ;

'A' Overload / Temp Hi" PPC ZE814A '

annunciator clears j 42.11 Insert malfunction to trip PCM TC09B

'B' EHC pump, observe:

"'B' EHC' Pump Running" CO6/7 BB24 O

annunciator clears PPC P108B

" Hydraulic Fluid Pump CO6/7 D31

'B' Overload / Temp Hi" PPC ZE814B annunciator alarms

~

'B' EHC pump handswitch CO7 HS-4467(R) )

red light remains energized )

42.12 Remove malfunction, observe: PCM TC09B

'B' EHC pump starts PPC P108B 1 1

" Hydraulic Fluid Pump CO6/7 D31

'B' Overlord / Temp Hi" PPC ZE814B , i annunciator clears  !

43.0 Start Hydraulic Fluid Filter C07 HS-4468 and Transfer Pump and observe:

o Red light illuminates, CO7 HS-4468(R) green light extinguishes C07 HS-4468(G)

PPC Z4386 l l

Rev.: 0 Date: 4/18/88 Page: 8.2.9-23 of 47 NSEM-4.01

--__m_-____-___mm_m_ __. -

i STEP PROCEDURE /RESULTS PANEL TAG #

'[ 44.0 Operate EHC Heaters and Tans C07 HS-4464 Handswitch and observe: I o Fans and heaters cycle PPC Z4384  ;

periodically when in auto j i

o Fans run continuously c when handswitch is placed l<

in fan-o Green light extinguishes when handswitch is placed in "OFF" 45.0 Test the overspeed trip and mechanical trip valve from rated speed 45.1 Momentarily depress the C07 LKOUTPB

" Locked Out" pushbutton, observe:

" Locked Out" light illuminates 45.2 Select " Medium" starting C07 MEDRATEPB(W)

-rx rate by depressing the PPC Z4502

(,) " Medium" pushbutton, observe:

Pushbutton light illuminates and computer point displays medium rate selected 45.3 Simultaneously depress the C07 OSTESTPB "Overspeed Test" and " Oil Trip" C07 OILTRIPPB pushbuttons allowing turbine speed to increase slightly above 1980 RPM, observe:

"Overspeed Test" light C07 OSTESTa'? f W) illuminates MTV trips immediately

" Increasing-Speed" light C07 INCSPEED(W) illuminates "AT SET SPEED" light C07 ATSPEED(W) extinguishes  ;

Rev.: 0 Date: 4/18/88 Page: 8.2.9-24 of 47 NSEM-4.01

J L I l STEP PROCEDURE /RESULTS PANEL TAG #

[ ) -

Turbine speed increases C07 SI-TURB at medium rate C07 UR-4501(P1)

(90 RPM / min.) to PPC SE439 approximately 1980 RPM

)

.I "Overspeed Trip" alarm CO6/7 D35 i received PPC Z4356 45.4 Release the "Overspeed Test" l

and " Oil Trip" pushbuttons l and observe: 3 Turbine speed will reduce i to and be held at 1800 RPM ]

i 45.5 Reset the MTV and observe: I Lockout automatically disarms 1

46.0 Repeat step 45.0 without locking out the mechanical trip and observe:

o Results similar to Step 44.0 C07 SI-TURB

(~N except at 1980 RPM the PPC SE439

(_) Emergency Trip System will trip causing all Turbine Stop Valves to close 47.0 Reset to snapshot taken with turbine at 1800 RPM ready to synch 48.0 Test the Turbine Backup Overspeed Trip 48.1 Set Load Set at or above 10% to permit reaching of Backup Overspeed Trip Setpoint NOTE: Overspeed Trip Set-point of 111.5% (2007 RPM) cannot be achieved in overspeed test with Load Set less than 10%

q Rev.: 0 Date: 4/18/88 Page: 8.2.9-25 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

1 48.2 Momentarily depress the C07 LOCKOUTPB N._ " Locked Out" pushbutton to override the Mechanical

.cip 48.3 Select " FAST" as the CO7 FASTRATEPB starting rate and observe:

Pushbutton light C07 FASTRATEPB(W) illuminates and computer PPC Z4503 point displays selection of fast rate 48.4 Depress and hold in the "Overspeed Test" pushbutton until the unit trips on backup overspeed at 2007 RPM the following will occur:

Lockout clears Mechanical and Emergency Trips occur

" Backup Overspeed Trip" CO6/7 A36 alarm received PPC Z4354

() -

All Turbine Steam Valves shut 48.5 Verify turbine coastdown conditions by observing:

o Turbine Main Stop and Control valves shut ,

o Turbine speed decreasing C07 SI-TURB o When turbine speed reaches C07 HS-7191 )

approximately 1500 RPM, CO6/7 A25 l Motor Suction Pump starts PPC P7201 with handswitch in the auto position. " Motor Suction l Oil Pump Running" annunciator alarms i

o Turbine Turning Gear Oil C07 HS-7190 l Pump running with hand-switch in auto position.

" Turning Gear Oil Pump CO6/7 C26 Running" annunciator alarming i i

Rev.: O l Date: 4/18/88 l Page: 8.2.9-26 of 47 NSEM-4.01 1 -

' STEP PROCEDURE /RESULTS PANEL TAG #

'E~ i o Lift pumps with handswitches b

in auto position start, and CO6/7 B28

" Lift Pump Low Pressure" PPC P4638 annunciator clears

-; o Turbine Turning Gear engages and starts 49.0 Perform the fo11ern.g steps to test the Turbine Lube Oil, Lift Pumps and Turning Gear response:

49.1 Stop the Turning Gear Oil ,

Pump and observe: f i

" Turning Gear Oil Pump CO6/7 C26 i Running" annunciator PPC P7200 I clears

{

Bearing Oil Pressure C07 PI-7210 initially decreases and {

PPC P7210 <

is restored by Emergency l Bearing Oil Pump operation at 10 PSIG { !

J

" Turbine Brg Emer Oil Pump CO6/7 C25 g~ Running" alarm received PPC P7199 when pump starts

" Turbine Brg Oil Pressure CO6/7 A28 Lo" annunciator alarms PPC P7215 with pressure less than 15 PSIG Emergency Bearing Oil C07 PI-7193 Pressure increases to approximately 25 PSIG while the D.C. Pump is running All Turbine Bearing Lift C07 HS-4453-57 Pumps stop

" Lift Pump Low Pressure" CO6/7 B28 annunciator alarms PPC P4638 Turning Gear stops and C07 HS-4463 remains engaged l l

lO Rev.: 0 Date: 4/18/88 Page: 8.2.9-27 of 47 NSEM-4.01

' STEP PROCEDURE /RESULTS PANEL TAG #

()L '

- " Turbine Rotor Stopped" alarms 5 seconds after CO6/7 PPC A26 Z4360 turbine speed is zero RPM 49.2 Restart Turbine Bearing Lift CO6R PB-MP80A Pumps by momentarily depressing CO6R PB-MP80B their reset pushbuttons and CO6R PB-MP80C observe: CO6R PB-MP80D CO6R PB-MP80E

- Lift pumps. remain off CO6R PB-MP80r until reset with associated pushbutton

- Turning Gear Motor remains off until any lift pump is started

- All lift pumps running and  !

turbine turning on gear upon step completion 50.0 Reset the Turbine )

50.1 Stop the Emergency Bearing Oil C07 HS-7189 i Pump and observe:

/~ - Bearing oil Header pressure C07 PI-7210 (h / decreases to zero i

- Turning gear stops but remains engaged

- All Turbine Bearing Lift Pumps stop

- " Turbine Rotor Stopped" CO6/7 A26 annunciator does not PPC Z4360 alarm due to no oil PPC SE439 pressure

- " Lift Pump Low Pressure" CO6/7 B28 anunciator alarms when PPC P4638 lift pumps are off

- Turbine trips on low CO6/7 B27 bearing oil pressure PPC Z4350 P7215 yielding the " Turbine PPC l

Thrust Bearing Wear Trip Alarm" Rev.: 0 Date: 4/18/88 Page: 8.2.9-28 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

n

'( ) 50.2' Restart the Emergency or Turning Gear Oil Pumps to restore lube oil header pressure and observe:

- Turbine Bearing Lift Pumps CO6R PB-MP80A remain off after oil CO6R PB-MP80B pressure is restored, until CO6R PB-MP80C reset with associated push- CO6R PB-MP80D button CO6R PB-MP80E CO6R PB-MP80r

- Turning Gear Motor remains deenergized until a Turbine Bearing Lift Pump is started 50.3 Operate Turbine Bearing Lift Pumps NOTE: Results noted for Pump A operation are similar for pumps B,C,D,E and F. Pumps will only auto start with zero speed signal 50.4 Place Lift Pump Handswitchs CO7 HS-4454 from auto position to Pull-

[^} -

, To-Stop position and observe:

'A' Lift Pump Oil pressure C07 P80A(R) indicating light illuminates PPC Z4375

- Pump Breaker status CO7 P80A(R)(G) '

indication extinguishes when handswitch is placed in Pull-To-Lock position

- Lift Pumps B and C operate CO7 HS-4453 PPC Z4376A together with associated handswitches PPC Z4376B C07 P80B(R)

C07 P80C(R)

- Lift Pump D operates C07 HS-4455 with associated hand- PPC Z4377 switch CO7 P80D(R)'

- Lift Rg59 E op9tates CO7 HS-4456 with Kosociated hand- PPC Z4378 switch C07 P80E(R)

O Rev.: 0  !

Date: 4/18/88  !

l Page: 8.2.9-29 of 47 NSEM-4.01 I

- _ - - - _ - - _ - _ __ -_a

-_-___---_ -= _ _ _ _ _ _ _ - - _

STEP PROCEDURE /RESULTS PANEL TAG 4

'[N /-}I - Lift Pump F operates-with associated hand-CO7 PPC HS-4457 Z4379 switch CO7 P80F(R) )

- Turning Gear Motor stops when.all lift pumps are stopped

- " Turbine Bearing Lift CO6/7 B28 Pump Pressure Lo" annunciator alarms ,

when all lift pumps  :

are stopped 50.5' Place any Lift Pump Hand-switch in the start position, spring return.to auto

-position, and observe:

- " Turbine Bearing Lift CO6/7 B28 Pump Pressure Lo" annunciator clears when any lift pump starts

- Turning Gear Motor

/ automatically restarts '

( when lift pump pressure is restored 50.6 Operate the Turning Gear C07 HS-4463 Handswitch and verify proper operation by observing Turning Gear Motor and CO6/7 A26  !

Turbine Shaft stop when j handswitch is placed in

" Pull-To-Stop" Turning Gear restarts in

" FAST" speed when hand-switr is placed in sta:t, sprng return to auto 51.0 Stop the Turbine Bearing Oil C07 HS-7423 l Vapor Extractor by placing {

HS handswitch in the stop position, observe:

r f~

Rev.: 0 Date: 4/18/88 Page: 8.2.9-30 of 47 NSEM-4.01

a .y 1 STEP 1 PROCEDURE /RESULTS PANEL TAG #

A30 I)

k-s o " Turbine Bearing Oil Vapor Extractor Not Running" CO6/7 PPC Z4363 annunciator alarms when stopped 51.1 Restart the. Turbine Bearing C07 HS-7423 Vapor Extractor by placing its associated handswitch i

in the start position and observe:

i

- -" Turbine Bearing oil CO6/7 A30 Vapor Extractor Not PPC Z4363 Running" annunciator clears 52.0 Reset simulator to 100% power, PCM NOP, NOT condition 52.1 Perform the Mechanical Trip valve Test 52.2 Momentarily depress the C07 LKOUTPB

" Locked Out" pushbutton observe:

" Locked Out" pushbutton C07 LKOUTPB(A) light illuminates after 1 sec delay, " Normal" CO7 NORMAL (R) light extinguished 52.3 Momentarily depress the C07 OILTRIPB

" Oil Trip" pushbutton for at least 3 seconds,  !

observe:

MTV " Tripped" light C07 MECHTIP(G) illuminates i 52.4 Momentarily depress the MTV C07 MTVR

" Reset" pushbutton for at -

least 3 seconds, observe:

O Rev.: 0 Date: 4/18/88 Page: 8.2.9-31 of 47 NSEM-4.01

STEPL ' PROCEDURE /RESULTS PANEL TAG 8

.(' -); -

Lockout' automatically disares in 10. seconds as indicated by'the j

" Locked Out" light extinguishing and the

" Normal" light illuminating

- MTV light goes from tripped to resetting and then reset 53.0 Test each Main Stop Valve as follows:

53.1 Momentarily depress each C07 MSVlTESTPB Stop Valve Test switch C07 MSV2TESTPB and observe: C07 MSv3TESTPB C07 MSV4TESTPB Stop valve strokes at medium rate until 12%

open, then rapidly closes I

f

1. 2 Min Stop Valve Servo

'q Valve Milliamp Indicator goes full scale low when test pushbutton is depressed

- Valve reopens when test PPC 2MS60D=

pushbutton is released PPC 2MS60C=

PPC 2MS60B=

PPC 2MS60A=

PPC 2MS60D6 PPC 2MS60C#

PPC 2MS60Bt PPC 2MS60At

- #2 Main Stop Valve Servo Valve Milliamp Indicator goes full scale high until valve is fully open, then milliamp indicator returns to zero i k

(_-

Rev.: 0 Date: 4/18/88 Page: 8.2.9-32 of 47 NSEM-4.01

_ - . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

1

. STEP PROCEDURE /RESULTS PANEL TAG #

[~'

'b-)E No electrical load is lost from full load j value when any Main Stop valve is closed 54.0 Test each Intercept and Intermediate-Stop Valve as follows!.

-54.1 Momentarily depress each C07 ISVITESTPB l Intercept and Intermediate .C07 ISV2TESTPB !

Stop Valve Switch and C07 ISV3TESTPB observe: C07 ISV4TESTPB Intercept Valve closes I at medium rate, until 10% open, then rapidly closes Intermediate Stop Valve closes at medium rate when the Intercept Valve is fully shut then rapidly shuts when it is at 10%

Intercept valves 1 and 2 Servo Valves Milliamp

~

Meters go full scale low

.w hen pushbutton is depressed, then full scale high when pushbutton is released until Intercept valves are fully open

- Releasing the pushbutton causes the ISV to open fully followed by tha IV opening fully 40 to 50 MW electrical load loss from full power when any set of valves are shut l-O Rev.: 0 Date: 4/18/88 Page: 8.2.9-33 of 47 NSEM-4.01 j.

STEP PROCEDURE /RESULTS PANEL TAG #

. 3

' I' e\- 55.0 Establish conditions for Control

\" valve testing as follows:

o Rx Power / Turbine Load at 90%

o 'A' Turbine Bypass Valve in in auto and open approx 5%

55.1 Momentarily depress each C07 CVITESTPB Control Valve Test Switch C07 CV2TESTPB C07 Cv3TESTPB C07 CV4TESTPB 55.2 Verify Control valves positilon does not change NOTE: This step verifies First Stage Pressure Feed- ,

back (FSPF) Interlock 55.3 Place turbine on load set C07 LOADSTPB control 55.4 Ensure Load Limit Potentio- C07 LOADLMTPOT meter is set at "10" 55.5 Place First Stage Pressure Feedback in service as follows:

Ensure that FSPF C07 FSPFPOT Potentiometer is positioned to "OUT",

check "OUT" light lit Depress the manual C07 FSPFKANPB pushbutton, observe the " Manual" light on and.the "OFF" light out Turn the FSPF C07 FSPFPOT Potentiometer slowly from "OUT" to "IN" observe "IN" light illuminated Rev.: 0 Date: 4/18/88 Page: 8.2.9-34 of 47 NSEM-4.01

l l

STEP PROCEDURE /RESULTS PANEL TAG #

'[^

55.6 Depress each control CO7 CVITESTPB N-) Valve Test Switch, C07 CV2TESTPB j hold in and observe: C07 CV3TESTPB C07 CV4TESTPB

- Moderate velocity C07 CVIPOS closure of Control PPC 2MS61D=

Valve PPC 2MS61D# 1 C07 CV2POS {

l At approximately.15% PPC 2MS61B=

open, rapid closure C07 Cv3POS l PPC 2MS61C= '

PPC 2MS61C#

- RPS Turbine Trip C07 CV4POS Bistable trips when PPC 2MS61A associated Co:4 trol PPC 2MS61A# j Valve Disk Dump Valve actuates (CV-1,2,3,4 affects RPS Channel D,B,C and A respectively)

Valve reopens when push-button is released p

( ,) 56.0 start the Motor Suction with PCM TUR01 Remote Function to operate 2-LO-60 and observe:

o Pump auto starts when

" Test" is selected by Remote Function o " Motor Suction Pump CO6/7 A25 Running" annunciator PPC P7201 alarms 56.1 Place Motor Suction Pump CO7 HS-7191  !

Handswitch in the Pull-To-Stop position and obsvrve:

- Pump stops and CO6/7 A25 annunciator clears PPC P7201  ;

(

l l

l l

O Rev.: 0 Date: 4/18/88 Page: 8.2.9-35 of 47 NSEM-4.01 i

STEP PROCEDURE /RESULTS PANEL TAG #

\ 56.2 Return Remote Function to PCM TUR01

[~'/

N- " Normal" and place Motor C07 HS-7191 Suction. Pump Handswitch in auto, observe:

Pump remains off 56.3 Start the Motor Suction CO7 HS-7191 Pump with handswitch 56.4 Insert malfunction to PCM TUO4A trip the Motor Suction Pump, obse rve:

Pump trips

" Motor Suction Pump Running" annunciator clears

" Motor Suction Oil Pump CO6/7 B25 Overload / Temp Hi" PPC ZE810 annunciator alarms 56.5 Remove malfunction, verify PCM TUO4A f" -

Motor Suction Pump does not restart 57.0 start the Turning Gear Oil PCM TUR02 Pump with its Remote Function to operate 2-LO-55A and observe:

o Turning Gear Oil Pump auto starts when " Test" is selected by Remote Function o " Turning Gear Oil Pump CO6/7 C26 Running" annunciator PPC P7200 alarms-57.1 Place Turning Gear Oil CO7 HS-7190 Pump Handswitch in Pull-To-Stop position and observe:

Pump stops and CO6/7 C26 annunciator clears PPC P7200 1

O Rev.: 0 Date: 4/18/88 Page: 8.2.9-36 of 47 NSEM-4.01 L. . _ _ _ - _ _ - - - - _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ )

STEP PROCEDURE /RESULTS PANEL TAG #

I'T 57.2 Return Remote Function to PCM TUR02

\l " Normal" and place CO7 HS-7190 Turning Gear Oil Pump Handswitch to auto, observe:

Pump remains off 57.3 Start Turning Gear Oil CO7 HS-7190 Pump using handswitch 57.4 Insert malfunction to PCM TUO4C trip the Turning Gear Oil Pump, observe:

Pump trips

" Turbine TGR Oil CO6/7 D26 Pump Overload / Temp Hi" PPC P7200 anunciator alarms PPC ZE812 l

57.5 Remove malfunction, verify PCM TUO4C j Pump remains off

(~3 58.0 Start the Emergency Bearing Oil PCM TUR03

(",) Pump with its Remote Function to operate 2-LO-103A and 2-LO-55B and observe:

o Emergency Bearing Oil Pump auto starts when

" Test" is selected by Remote Function o " Emergency Bearing Oil CO6/7 C25 Pump Running" annunciator alarms o Turning Gear Oil Pump auto starts o " Turning Gear Oil Pump CO6/7 C26 Running" annunciator alarms 58.1 Place Emergency Bearing Oil CO7 HS-7189 Pump Handswitch in Pull-To-Stop position and observe:

(~T V

Rev.: 0 Date: 4/18/88 Page: 8.2.9-37 of 47 )

NSEM-4.01 j l

)

u______--._ 1

t a I

STEP PROCEDURE /RESULTS PANEL TAG 4 Pump stops and_ CO6/7 C25 1 '

)!

annunciator clears PPC P7199 58.2 Start EBOP using handswitch. C07 HS-7189 58.3 Insert malfunction to PCM TUO4B trip the EBOP, observe:

- EBOP trips

" Turbine BRG Emer CO6/7 D25 Oil Pump" overload PPC ZE811 58.4 Remove malfunction PCM TUO4B i

59.0 Place Turning Gear Oil Pump C07 HS-7190 Handswitch in Pull-To-Stop position and observe:

o Pump stops and CO6/7 C26 annunciator clears PPC P7200 59.1 Return Remote Function to

" Normal" then place

. Emergency Baaring Oil Pump CO7 HS-7189 Turning Gear Oil Pump Hand- C07 HS-7190 O, switches in auto positions, observe:

- Pumps remain off 60.0 Reset simulator to 100% power, PCM NOP, NOT condition 60.1 Bypass the Turbine EHC PCM TCR07 i Hydraulic Pressure Low Trip using remote function in the " Bypass" position 60.2 Place both EHC Pumps C07 HS-4466 Handswitches in "OFF" C07 HS-4467 and observet

" Hydraulic Fluid CO6/7 C32 f Pressure Lo" l annunciator alarms j when EHC System pressure decreases ,

l to 1300 PSI O

v Rev.: 0 Date: 4/18/88 Page: 8.2.9-38 of 47 NSEM-4.01

!' L f

" STEP PROCEDURE /RESULTS PANEL TAG # l j

Turbine Lo Hyd Fluid PPC Z4352

?(7~)

Pres

- Mn.Turb Hyd Fluid Pres PPC P7056 l i

- Mn Turb EHC System PPC ZE430 ZE429 I Turb EHC Emer Pres PPC 60.3 Verify turbine does not trip when EHC pressure decreases to 1100 PSIG ]

i by observing:

" Hydraulic Fluid Press CO6/7 D32 j Lo Lo Trip" annunciator 1 alarms 60.4 Return the Turbine EHC PCM TCR07 Hydraulic Pressure Low Trip to " Normal" and observe:

- Turbine trips and asociated annun-ciators alarm

. 61.0- Reset simulator to 100% power, NOP, NOT condition 61.1 Bypass the Turbine Low PCM TCR03 Condenser vacuum Trip using Remote Function in the " Bypass" position 61.2 Cause a loss of condenser PCM FWO1 vacuum using malfunction at 100% severity, observe:

- Condenser vacuum decreases to turbine trip setpoint

- Turbine does not trip Rev.: 0 Date: 4/18/88 Page: 8.2.9-39 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

i 61.3 Return the Turbine. Low 9NJ' -

Vacuum trip to " Normal" and observe:

- Turbine trips

" Condenser Vacuum Lo" CO6/7 A37 and " Condenser vacuum CO6/7 B37 Lo-Lo Trip" annunciators alarm 62.0 Reset simulator to 100% power, PCM NOP, NOT condition 62.1 Place Motor Suction Pump, C07 HS-7191 Turning Gear Oil Pump and C07 HS-7190 i Emergency Bearing Oil Pump C07 HS-7189 l Handswitches in Full-To-Stop position- l 62.2 Bypass the Turbine Bearing PCM TCR04 Oil Low Pressure Trip and PCM TCR05 Turbine Driven Shaft Pump Low Pressure Trip using Remote Functions in the rT " Bypass" position U 62.3 Cause a decrease of Bearing PCM TUO3 Oil Supply Pressure using malfunction at 100%,

observe:

- Decrease in Bearing Oil C07 PI-7210 i Supply Pressure PPC P7210

- " Turbine Bearing Oil CO6/7 A28 Pressure.Lo" annunciator alarms when pressure is less than 15 psi 62.4 Return the Turbine Bearing PCM TCR04 Oil Low Pressure Trip to

" Normal" when Bearing Oil Pressure is less than 8 PSI, observe:

- Turbine trips and associated annun-clators alarm O Rev.: 0 Date: 4/18/88 Page: 8.2.9-40 of 47 NSEM-4.01

-_a_-______-_-_-

l l

STEP PROCEDURE /RESULTS PANEL TAG I 1J 'l 63.0 Reset simulator to 100% power, PCM

\~/' NOP, NOT condition 63.1 Place Motor Suction Pump, C07 HS-7191 Turning Gear Oil Pump and C07 HS-7190 Emergency Bearing Oil Pump C07 HS-7189 Handswitches in Pull-To-Stop position 63.2 Bypass the Turbine Bearing Oil PCM TCR04 Low Pressure Trip and Turbine PCM TCR05 Driven Shaft Pump Low Pressure Trip using Remote Functions in the " Bypass" position 63.3 Cause a decrease of Shaft Pump PCM TUO3 Oil Supply Pressure using malfunction at 100%, observe:

- Decrease in Shaft Pump PPC Z4348 Oil Supply pressure

- " Turbine Shaft Pump CO6/7 A27 Pressure Lo Lo Trip" annunciator alarms r~3 63.4 Return the Turbine Shaft PCM TCR05

(,) Pump Low Pressure Trip to

" Normal" when Shaft Pump pressure decreases below 105 PSIG, observe:

- Turbine trips and associated annun-ciators alarm 64.0 Reset simulator to 100% power, PCM NOP, NOT condition 64.1 Bypass the Loss of Stator PCM TCR06 Coolant Trip using Remote Function in the " Bypass" position 64.2 Cause a Loss of Stator PCM TPO4 .

l Coolant Flow using malfunction at 100%

severity, observe:

)

l s l

,d Rev.: 0 Date: 4/18/88 Page: 8.2.9-41 of 47 NSEM-4.01

_ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ ]

, . STEP PROCEDURE /RESULTS PANEL TAG i l

l l)

L./

"H2 Seal Oil Stator

'Clg Cabinet Trouble" CO6/7 PPC C36 ZE433 annunciator alarms Stator Coolant Inlet C07 TR-4510(P9) and Outlet temperatures C07 TR-4510(P8) increase PPC TE005C PPC TE005D 64.3 Return Remote Function to PCM TCR06

" Normal" when Stator Coolant temperature exceeds 81 C for greater than 70 seconds, observe:

Turbine trips

" Loss of Stator Coolant CO6/7 D36 l Trip" annunciator alarms l 65.0 Reset simulator to 100% power, l NOP, NOT conditions 65.1 Bypass the MSR High Level PCM TCR08 Trip using Remote Function rs in the " Bypass" position

\'.) 65.2 Cause an MSR high level PCM MS13A&c(0) condition using malfunction MS16A(0) at 0% severity, observe:

" Moisture Separator COS AA22 Drain Tank 1A Level PPC L4131 High" annunciator PPC Z4488 alarms 65.3 Return Remote Function PCM TCR08 to " Normal" when MSR Drain Tank level increases to greater than 3 inches below MSR shell, observe:  :

- Turbine trips

" Turbine Moisture CO6/7 B35 Separator Level Hi-Hi Trip" annunciator alarms A

U Rev.: 0 Date: 4/18/88 Page: 8.2.9-42 of 47 NSEM-4.01 l

STEP. PROCEDURE /RESULTS PANEL TAG #

Reset simulator to 100% power, PCM (s \ 66.0 NOP, NOT conditions 66.1 Bypass the Unit Electrical PCM TCR09 Protection Trip using Remote Function in the

" Bypass" position 66.2 _Cause a Main Generator Trip PCM EG01 using malfunction, observe:

" Turbine Generator CO6/7 A48 Lockout Trip Channel B48 1/2" annunciator alarms

" Exciter Field Breaker CO6/7 A45 Trip" annunciator alarms

" Generator Field Breaker CO6/7 A46 Trip" annunciator alarms PPC ZE184 Turbine does not trip 66.3 Return Remote Function to PCM TCR09

" Normal" and observe:

/~' -

Turbine trips on over-(_) speed and associated annunciators alarm 67.0 Reset simulator to 100% power, PCM NOP, NOT conditions 67.1 Bypass the Reactor / Turbine PCM TCR10 Trip using Remote Function in the " Bypass" position 67.2 Manually trip the Reactor using the.Rx Trip Push-buttons on C04, observe:

Turbine does not trip 67.3 Return Remote Function to PCM TCR10

" Normal" and manually trip the Reactor Turbine trips, Rx trips O Rev.: 0 Date: 4/18/88 Page: 8.2.9-43 of 47 NSEM-4.01

STEP PROCEDURE /RESULTS PANEL TAG #

" Reactor / Gen Prot Turbine CO6/7 C48

./\~/) Trip" annunciator alarms 68.0 Reset simulator to 100% power, PCM NOP, NOT conditions 68.1 Bypass the Turbine Manual PCM TCR11 Emergency Pushbutton Trip using Remote Function in the " Bypass" position 68.2 Depress the Turbine co7 Emergency Trip Push-button, observe:

Turbine does not trip 68.3 Return Remote Function to PCM TCR11

" Normal" 68.4 Depress the Turbine Manual C07 Emergency Pushbutton, observe:

/ - Turbine trips and i,, associated annun-ciators alarm 69.0 Reset simulator to 100% power, PCM NOP, NOT conditions 69.1 Bypass the Steam Generator PCM TCR12

1. 1 High Level Trip using Remote Function in the

" Bypass" position 69.2 Fail #1 S/G FRV at 100% PCM FWO9A(100) and increase S/G level to greaer than 90% level, observe:

- Turbine does not trip i

k Rev.: 0 Date: 4/18/88 Page: 8.2.9-44 of 47 NSEM-4.01

STEP ~ PROCEDURE /RESULTS PANEL TAG #

[)

\d 69.3 Return Remote Function to

" Normal and observe:

PCM TCR12

- Turbine trips and associated annun-ciators alarm 70.0 Reset simulator to 100% power, PCM NOP, NOT conditions 70.1 Bypass the Steam Generator PCM TCR13

1. 2 High Level Trip using Remote Function in the

" Bypass" position 70.2 Fail #2 S/G FRV at 100% PCM FWO9B(100) and increase S/G level to greater than 90% level, .

observe: )

1 Turbine does not trip

'70.3 Return Remote Function to PCM TCR13

" Normal" and observe:

Associated annunciators O

alarm 71.0 Reset simulator to 100% power, PCM NOP, NOT conditions l 71.1 Bypass the Turbine Exhaust PCM TCR14 Hood High Temperature Trip using Remote Function in the " Bypass" position j 71.2 Set Turbine Exhaust Hood HZL MSTHOOD temperature to 225*F at l the Hazeitine terminal, observe:

- Turbine does not trip

" Turbine Exhaust Hood CO6/7 C34 Temp Hi" annunciator PPC T7158 alarms when temp j

exceeds 175*F O Rev.: 0 Date: 4/18/88 Page: 8.2.9-45 of 47 .

NSEM-4.01 l

J >

L STEP: PROCEDURE /RESULTS . PANEL TAG #

li' E 71.3 Return Remote Function to PCM TCR14 l -

" Normal" and observe:-

- Turbine trips l - " Turbine Exhaust Hood CO6/7 D34 Temp Hi_Hi. Trip"- PPC Z4346 3 L' annunciator alarms L

72.0 Reset simulator to 100% power PCM NOP, NOT and place in Run 72.1' Insert malfunction in PCM TUO2(A,B,C,0,E)

, 10% increments (1 mil /10%)

to increase Turbine I vibration levels, observe:

- Turbine bearing's C07 UR4501  !

vibration values PPC ZE428 increase PPC X4391

- " Turbine Vibration Hi" PPC X4392 annunciator alarms when PPC X4393 any bearing vibration PPC X4394 value exceeds 7-8 mils PPC X4395

' ~

PPC X4396

- " Turbine Vibration Hi Hi" PPC X4397 annunciator alarms when PPC X4398 any bearing vibration PPC X4399 value exceeds 12 mils PPC X4400 PPC Z4353 CO6/7 A34 CO6/7 B34 73.0 Depress master trip test C07 TESTA (R) solonoid " Test A" pushbutton, observe:

o Switch light extinguishes C07 TESTA (R) 73.1. Release " Test A" push-button, observe:

- Switch light C07 TESTA (R) illuminates l

l

'()

Rev.: 0 Date: 4/18/88 Page: 8.2.9-46 of 47 l

I NSEM-4.01 .J 1

lI

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ ]

H STEP' PROCEDURE /RESULTS PANEL TAG # )

- H .

I

, 73.2 Depress master. trip test l L'  ;

solonoid " Test B" push-button, observe:

- Switch light C07 'TESTB(R) extinguishes 73.3 Release." Test B" push-button, observe:

l L --

Switch light C07 TESTB(R) illuminates

-74.0 Reset to IC-24 and place in PCM

'Run. Verify the-following non-modeled annunciators are clear: CO6/7 A32 CO6/7 D32 CO6/7 C33 CO6/7 C35 CO6/7 D29 CO6/7 A29 CO6/7 B26 CO6/7 .B29 CO6/7 D30

/V

-(_)

CO6/7 CO6/7 C27 C28 l

.CO6/7 D27 i CO6/7 D28 l CO6/7 D33 l

Rev.: 0 Date: 4/18/88 ,

Page: 8.2.9-47 of 47 l' NSEM-4.01

O ATTACHMENT 2 MP2 SYSTEM TEST ABSTRACTS i

This attachment is referenced by section 2 of the Performance Test Summary.

O 1

4 O '

..c ,, ,,,

1

/

f SYSTEM TEST ABSTRACTS

' V(

.I

1. Service Water System Test This System Test was performed in December, 1987. The Service Water System is a. safety related system which takes water from Long Island Sound, cools the Reactor Building Component Cooling Water (RBCCW) Heat Exchangers, Turbine Building Component Cooling Water.(TBCCW) Heat Exchangers, Emergency Diesel Generator Heat Exchangers and various chillers and then discharges back to Long Island Sound.

This System Test was performed at 100% power. The following  ;

areas were tested:  !

o Correct response of Plant Process Computer (PPC) Analog and .

Digital Points related to the Service Water System. l 0 Annunciators associated with the Service Water System were verified to alarm at the correct setpoints, o Service Water System Lights on Control Panel C-Olx (Safeguards  !

Actuation Status) were verified to light properly. 1 o All' remote functions associated with the Service Water System were tested for correct operation, o All control board hardware associated with Service Water Systems were. cycled, i.e., valve hand switches, reset pushbuttons, etc.

-o Service Water Pumps were started and stopped.

o Normal Service Water flowpaths and backup Service Water flowpaths were tested.

o Service Water valve interlocks were tested.

o Service Water related temperature control valves were cycled.

o Stroke times of Service Water safety related valves were tested.

Data used to support this test was obtained from P&ID's (Piping and Instrumentation Drawings), Electrical Drawings, Plant Surveillance Data and Annunciator Setpoints from Plant Procedures.

No deficiencies were identified, although 4 Digital Plant Process Computer Points (PPC) were identified to work incorrectly. These points would not be used by operators or by a PPC Program operators would use, therefore, no further action is planned, une ei ts' 2

? y-L L

.p-sq 2. Circulating Water System Test

_This test was performed in November, 1987. The Circulating Water-System takes water from Long Island Sound cools the Main Condenser and returns the water to Long Island Sound.

The following areas were tested:

o. Correct-response of Plant Process Computer (PPC) Analog and Digital Points related to the Cire Water System.

o. Annunciators associated with Cire Water System were verified to alarm at the correct.setpoint.

oL All control board hardware associated with the.Cire Water-System were cycled, i.e., valve handswitches, pump 3 handswitches, etc.

o At Het Zero Power, all pump and valve interlocks associated with the Cire Water System were tested.

o At 75% power, all 4 Circ Pumps were cycled, all 4 waterboxes were sequentially isolated, vented and backwashed.

o At 100% power,'the following was tested: condenser vacuum relationship to the number of running Circ Water Pumps,  ;

Traveling Screen System response, Screen Wash System response, ;

sensitivity to Sea Water temperature and proper operation of f the vacuum Priming System.

o All. remote functions associated with the circ Water System  !

were tested.

Data used to support this test was obtained from P&ID's, Electrical Drawings and Annunciator Setpoints from plant procedures. No deficiencies were identified.

3. Turbine Building Component Cooling Water (TBCCW) System This test was performed in December, 1987. The TBCCW System is a i closed loop that removes heat from Turbine Building components  !

(Main Generator, Lube Oil, etc.) and rejects that heat to the Service Water System through the TBCCW to Service Water Heat Exchangers. l The following areas were tested:

o Correct response of Plant Process Computer (PPC) Analog +

Digital Points related to the TBCCW System.  :

i o Annunciators associated with TBCCW were verified to alarm at I the correct setpoint.

O wac si es 3 1 l

i-t f

I As1 N. o- All. control. board hardware associated with the TBCCW System were cycled, 1.e., valve handswitches, pump handswitches, etc.

o- At 20% power, TBCCW pumps were started and stopped to ensure correct. valve and breaker. interlocks.

o At 100% power, TBCCW heat'exchangers were swapped to check flowpaths and the Temperature Control valves (TCv's) were checked to have a correct relationship between valve position and' Service Water Flow. Degraded or Lost TBCCW flow was verified to affect components served by TBCCW, including Air Compressors.

o At 100%~ power, the Isophase Bus cooling System and Stator Water Cooling System were also tested.

o All remote functions associated with the TBCCW System were tested.-

Data used to support this t>st was obtained from P&ID's,' {

Electrical Drawings and Annunciator Setpoints from plant procedures. No deficiencies were identified.

4. Reactor Building' Component Cooling Water (RBCCW)' System g This test was performed in January, 1988. The RBCCW System is a 6

safety related system which is'a closed loop that removes heat

\- .from Containment and Auxiliary Building components and rejects that heat to the Service Water System through the RBCCW to Service Weter Heat Exchangers.

The following areas were. tested:

o Correct response of Plant Process Computer Analog and Digital Points related to the RBCCW System.

o Annunciators associated with RBCCW were verified to alarm at the correct setpoint.

o All control board hardware associated with the RBCCW System were cycled, i.e., valve handswitches, pump handswitches, reset pushbuttons, etc.

At 100% power, all the following were tested:

. o RBCCW pumps were stopped and started to verify interlocks.

o Safety Related RBCCW valve stroke times were checked.

o RBCCW fi*wpaths thru various pumps / heat exchanger combinations were checked.

O - NRC 01 (5) 4

l

[' o- Using Temperature Control Valves, Service Water Flow

[ (\~-} relationships to valve positions were verified..

o Heat loads on.each RBCCW header were-checked, such as safeguards Room Cooling, Shutdown Cooling Heat'Exchangers, Spent Fuel Pool' Cooling, Containment' Air Coolers, Quench Tank / Primary Drain Tank Coolers,-Degassifier Coolers and the Letdown Heat Exchanger.

o Remote functions associated with the RBCCW System were verified to work correctly.

o C-O1x'(Safeguards Status) lights were checked for proper response.  !

I Data used to support this test was obtained from P&ID's, Electrical Drawings, Annunciator Setpoints from plant i procedures and Plant Surveillance Data.

No deficiencies currently exist for this System Test.

5. Safety Injection / Containment Spray-(SI/CS) Test This System Test wasiperformed in November, 1987. This is a safety related system.- This system included High Pressure and Low Pressure Safety Injection, Safety Injection Tanks and the

( Containment' Spray System.

The following areas were tested:

o Plant Process Computer (PPC) Analog and Digital Points related to the SI/CS System were verified to work correctly, o Annunciators associated with the SI/CS System were verified to operate at their correct setpoints.

i o All control board hardware associated with the SI/CS System were cycled, i.e., valve handswitches, pump handswitches, etc.

At 100% power:

o various safety related SI/CS valves were cycled and interlocks checked.

o Stroke times of safety related SI/CS valves were verified.

o C-Olx (Safeguards status) lights were verified.

o Flowpaths between components were verified.

o SI/CS pumps were started / stopped and breaker /handswitch I 1 operation checked.

)

nac e2 <s' 5

i

=____ _ _ _ _ _ _ _ _ . _ _ _ _ l

1 f

,_, o SI/CS pump suction flowpaths were verified.

ks J o Safety Injection Tanks were filled. i o All SI/CS remote functions were verified to work correctly.

In a Post-LOCA condition the following was verified:

o Both methods of Boron Precipitation control were tested.

Data'used to support this test was obtained from P&ID's, Electrical Drawings, Plant Surveillance data and annunciator setpoints from plant procedures.

I No deficiencies were identified.

.I

6. Chemical and Volume Control System (CVCS) Tast This System Test was performed in April, 1988. The CVCS is a safety related system which services the Reactor Coolant System. {

This test was. performed at 100%. power. The following areas were tested:

o Correct Response of. Plant Process Computer (PPC) Analog and Digital Points related to the CVCS.

o Annunciators associated with the CVCS were checked for proper alarm setpoints.

(

o All remote functions associated with the CVCS were checked.

o All control board hardware associated with the CVCS were cycled, such as valve handswitches, controllers, reset pushbuttons, etc.

o Flowpaths to and from the Volume Control Tank (VCT) were I tested.

o . Safety related valve stroke times were tested.

1 o Makeup methods to the VCT were tested.

o The letdown backpressure controller was tested.

o C-01X (Safeguards status) lights were tested.

o Flcwpaths from the Boric Acid Storage Tanks (BAST's) to the Charging Pumps were tested.

o Charging Pump Handswitch/ Selector Switch operation was tested.

i o The flowpath from the Refuel Water Storage Tank (RWST) to the CVCS was tested.  !

O NRC '91 (5) g L

)

a _ - - _ - -

i I j o The HPSI to charging flowpath was checked.

"o The Pressurizer Level and letdown control system was checked, o The RCP Bleedoff, Letdown Temperature and Precise Dilution Flow Controllers were all checked.

o The.Boronometer. Package was checked.

o The-Pressurizer heater cutout was checked.

Data used to support this test was obtained from P&ID's, Electrical Drawings, Plant Surveillance Data and Annunciator setpoints from plant procedures.

No deficiencies were identified, although 2 Digital Plant. Process j Computer Points (PPC) were identified to work incorrectly. These ,

points would-not be used by operators or by a PPC program  !

operators would use, therefore, no further action is planned.

7. Condensate /Feedwater System Test f This System Test was performed in February, 1988. Portions of the Feedwater System are safety related. The Condensate /-

Feedwater System delivers feedwater flow to the 2 Steam

("'i Generators for RCS heat removal. The following areas were V tested:

o Correct response of Plant Process Computer (PPC) Analog and j Digital Points related to Condensate /Feedwater. '

o Annunciators associated with the Condensate /Feedwater System

were checked for proper alarm setpoints.

o All remote functions associated with the Condensate /Feedwater Systems were checked.

o All control board hardware associated with the cond/ Feed System were cycled such as valve handswitches, controllers, reset pushbuttons, etc.

o With the Condensate System / Main Feed System out of service, the Condensate and Feedwater System were started up.

o Stroke times of safety related valves were checked.

o C-Olx (Safeguards Status) lights were checked.

O Auxiliary Feed Valves and Pumps were cycled to check operation.

o Automatic Auxiliary Feed Actuation was initiated and verified.

nne et isi 7

L____

r-i-

1 t ol The: Steam Driven Auxiliary Feed Pump (Terry Turbine) was started,

~

o- The Fire Water System supply to Auxiliary Feedwater was verified.

o- Main Steam Generator Feed Pumps'(SGFP's) were started and auxiliary equipment verified. I L o The Short Recycle flowpath was verified and condenser vacuum capabilities were verified.

o At'40% power, proper operation af Heater Drains Pumps and valves were verified.

o At 70% power, Main Feed Pump Trips were tested.

~

o At 100% power, Loss of TBCCW and SGFP oil pumps were tested for proper effects on the'SGFP's.

Data used to support this test was obtained from P&ID's, electrical drawings, plant surveillance data and annunciator setpoints from plant procedures. No deficiencies were identified.

8. Main Steam System Test This System Test was performed in April,- 1988. Portions of the Main Steam System are safety related. The Main Steam System removes heat from the Steam Generators.

The following areas were tested:

o Correct response of Plant Process Computer Analog and Digital-Points related to the Main Steam System were verified. ,

o Annunciators associated with Main Steam were verified to alarm at their correct setpoint.

o Remote functions associated with Main Steam were verified for proper operation, o All control board hardware associated with Main Steam was cycled to verify proper operation (i.e., valve handswitches, controllers, reset pushbuttons, etc.).

o Normal and abnormal flowpaths were tested.

o At ~ 0% power, the Terry Turbine was started and steam supply flowpaths to the Terry Turbine verified, MSIV handswitches )

logic was checked, safety valve flowpaths were checked, the  !

Turbine Stop/ Control valve logic and flowpaths were checked j

,/ and steam supply to the MSR's was checked. j

)

NRC 91 (5) g

\

y ..

f a o C-OlX-(Safeguards status) lights were' checked.

L i

o- At1100% power, flowpaths and valves to the MSR, 1st= Stage l Reheater and 2nd Stage Reheater Tanks were verified, Extraction Steam flowpaths/ valves to the Feedwater Heaters were: verified, Main Turbine Intercept Valve testing was .

performed, stroke testing of.MSIV's was performed, i verification of Main Steam Isolation (MSI) actuation was I performed and verification of flowpath to condenser thru'the Turbine Bypass valves was performed.

o Stroke time of various. safety related valves were tested.

Data used to. support'this test was obtained from P&ID's, electrical drawings, plant surveillance data and annunciator setpoints from' plant procedures.

No deficiencies correctly exist for this System Test.

9. Turbine System Test This System Test was performed in February, 1988.

The following areas were tested. 1 o

O Correct response of Plant Process Computer Analog and Digital Points'related to'the Turbine.

o Annunciators associated with the Turbine were verified to alarm at their correct setpoint.

o Remote functions associated with the Turbine were verified to work correctly.

o All control board hardware associated with the Turbine was I cycled to verify proper operation (i.e. valve handswitches, "

i controllers, reset pushbuttons, etc.).

o Normal and abnormal flowpaths were tested.

o The turbine was taken from a cold shutdown condition, warmed I up and rolltd up to 1000 RPM with auxiliary equipment tested during the terbine startup. j o Auxiliary equipment tested included Electro Hydraulic Control l (EHC) Pumps and System, Lift Pumps, Turning Gear Pump / Motor i and Emergency Oil Pumps.

o Various Turbine Surveillance were performed such as Main stop Valve Testing, Intermediate Stop Valve (ISV) Testing and Control Valve Testing.

O ... ., m, 1

[ o Turbine Trip bypasses were checked for proper operation.

Data used to support this test was obtained from P&ID's, electrical drawings, plant surveillance data, annunciator setpoints from plant procedures and the Turbine technical manual.

No deficiencies currently exist for this system test.

10. Electrical Generation System Test This System Test was performed in January, 1988. This test includes the Main Generator and the Emergency Diesel Generators i (EDG's). The EDG's are safety related.

The following areas were tested:

o Correct response of Plant Process Computer Analog and Digital Points related to the Electrical Generation Systems, o Annunciators associated with the Electrical Generation Systems were verified to alarm at their correct setpoint.

o Remote functions associated with the Electrical Generation Systems were verified to work correctly.

/ o All control board hardware associated with the Electrical Generation Systems were cycled to verify proper operation (i.e., breaker handswitches, pushbuttons, etc.).

o The EDG's were started and paralleled to their Emergency Busses and emergency shutdown's performed.

o The EDG's by themselves were tested to carry their respective emergency busses'with loads started and stopped to verify proper LOAD / VAR response.

o An Loss of Normal Power (LNP) was initiated to test proper Diesel performance.

o various electrical faults were tested to verify EDG logic.

o The Main Generator Hydrogen Seal Oil System was tested, o At 8% power, the Main Generator was " synchronized" to the grid.

o At 100% power, Main Generator controls were tested and a manual trip performed to verify correct response.

Data used to support this test was obtained from P&ID's, electrical drawings and annunciator setpoints from plant procedures.

No deficiencies currently exist for this test.

nec e2- es) 10 l

,as 11. Electrical Distribution System Test

! Y

\_/c This System Test was performed in January, 1988. This test includes'the AC 345 KV, 6.9 KV, 4.16 KV, 480 V and 120 Volt (AC &

DC) Distribution. Systems. Some of these AC 4.16 KV, 480 V and 120 volt (AC + DC) busbes are safety related.

The following areas were tested:

o . Correct: response of Plant Process Computer Analog and Digital Points related to the Electrical Distribution System.

o Annunciators associated with the Electrical Distribution System were verified to alarm at their correct setpoint.

o Remote' functions associated with the Electrical Distribution System were. verified to work correctly.

o All control board' hardware associated with the Electrical Distribution System was cycled to verify proper operation (i.e., breaker handswitches).

o 480-Volt buses were cross tied to verify correct response.

o 4160. Volt buses were tested to ensure that they could be fed from the NSST, RSST and Unit 1.

o. Verify that on a reactor trip the normal fast transfer occurs O'

t correctly for the Normal Station Service Transformer (NSST) to the Reserve Station Service Transformer (RSST).

o Verify' Electrical Breaker interlocks.

o Verify NSST/RSST indication on the control boards.

o verify vital 120 volt Static Switch operation.

o verify non vital 120 volt (VRll, VR21) transfer switches.

o Verify DC Battery Charger operation.

Data used to support this test was obtained from P&ID's, electrical drawings and Annunciator setpoints from plant procedures.

l l

NRC el ($)

11

1 p .. 12. Reactor Coolant System (RCS) Test

~'

This' System Test was performed in January, 1988.

The following areas'were tested:

o Correct response of Plant Process Computer Analog and Digital Points related to the RCS.

o Annunciators associated with the RCS were verified to alarm at their correct setpoint.

o Remote functions associated with the RCS were verified'to work correctly, o 'All control board hardware associated with the RCS was cycled to verify proper _ operation (i.e., valve handswitches, controllers, reset pushbuttons, etc.).

o Normal and abnormal RCS flowpaths were tested.

o During Shutdown Cooling (SDC) operations, verified Pressurizer Nitrogen overpressure capability and ability to start RCP's when leaving SDC. Also RCP breaker interlocks were checked.

o At Hot Zero Power, RCS flowpaths were verified including back flow depending on which RCP's were in operation.

() o Verified Quench Tank and Primary Drain Tank Operations.

o . Verified RCP Bleedoff flowpaths.

o Tested main and auxiliary spray.

o Tested Pressurizer PORV and Safety Valve operations.

o Verified Natural Circulation capability.

-o At 100% power, verified Pressurizer Heater Operation, Spray Valve Operation, Pressure Control System operation and Low Temperature Over Pressure (LTOP) operation.

o Verified RCP response to RCP seal problems, RCP oil problems and RBCCW flow problems.

Data used to support this test was obtained from P&ID's, electrical drawings, plant surveillance data and annunciator setpoints from plant procedures.

No deficiencies were identified.

O 12

e ~v

,s

, 13. Reactor Protection System / Nuclear Instrumentation (RPS/NI) Test I

^~- )

i This System Test was performed in April, 1988.

The following areas were tested:

o Correct response of Plant Process Computer Analog and Digital Points related to the RPS/NI Systems.

o Annunciators associated with the RPS/NI Systems were verified to alarm at their correct setpoint.

o Remote functions associated with the RPS/NI Systems were verified to work correctly.

o All control board hardware associated with the RPS/NI Systems was cycled to verify proper operation (i.e., handswitches, controllers, reset pushbuttons, etc.).

O For.each RPS channel, correct trip and pretrip values / lights for all 10 RPS trips were tested, proper operation of channel bypass keys were verified, RPS trip bypasses from the instructor station were verified for proper operation and all switches on the RPS front drawers were tested.

o Power Ratio Recorder operation was verified to function properly.

() o Surveillance OP 2601D was performed on the 4 RPS channels.

o All 4 Wide Range detectors front panel switches were verified to work correctly.

o The Trip Circuit Breakers (TCB's) were verified to open/Close correctly.

o Control Element Drive Mechanism (CEDM) Motor Generator (MG) sets were shutdown and started.  !

o RPS Matrix testing was performed.

o RPS auctioneering input testing, zero power mode bypass switch testing and Core Protection Calculator testing was performed.

o Contrcl Valve Testing was performed (at 90% power).

o The variable High Power Trip was tested for proper operation.

Data used to support this test was obtained from electrical drawings, annunciator setpoints were from plant procedures and the RPS Technical Manual was also used.

No deficiencies were identified, however, 6 Digital Plant Process Computer (PPC) Points did not work properly. These 6 points are

(~' not used by operators or by a PPC program operators would use,

(_s} therefore, no further action is planned.

NRC el (5) 3 l

l L_ _ _ _ _ _ _

j

.f x 14. Engineered Safety Features Actuation System Test (ESAS)

!' ') This System Test was performed in January, 1988.

The following areas were tested:

o Correct response of Plant Process Computer Analog and Digital Points related to the CSAS.

1 o

Annunciators associated with the ESAS were verified to alarm at their correct setpoint.

o Remote functions associated with the ESAS were verified to work correctly, o All control board' hardware associated with the ESAS was cycled to verify proper operation (i.e., handswitches, pushbuttons, etc.). I o For each of the 4 ESAS sensor cabinets, the Trip Test Switch was tested for proper response.

o For Actuation Cabinets 5 & 6, a "] out of 5" test signal was initiated to verify proper response.

o The Loss of Normal Power (LNP) trip setpoints were verified for each sensor cabinet.

(], ' ,

o Each actuation module was tested to ensure the correct emergency equipment actuated, o The Diesel Generator Sequencer was tested, o The Test Inhibit / Bypass keys and 2/3 logic were checked.

o The "ATI" (Automatic Test Inserter) Bypass was checked.

o Manual Initiation and Blocks of ESAS were tested.

o ESAS Fuse Replacement was tested.

Data used to support this test was obtained from P&ID's, electrical drawings and annunciator setpoints/ESAS setpoints are from plant procedures.

No deficiencies currently exist for this System Test, however 4 digital Plant Process Computer Points (PPC) were identified to work incorrectly. These points are not used by the operators or by a PPC program the operators would use. Therefore, no further action is planned on these 4 points.

l x,,)

(D NRC #1 (5) 74

i:

c t i-h l :7 s 15. Reactor' Regulating System'(RRS) Test

% IS This System Test was performed in January, 1988. The Reactor L

L Pegulating System affects Pressurizer Pressure Control, Pressurizer Level Control, the Power Ratio Recorder, and under certain conditions, automatic control of the Condenser Steam g Bypass valves and Atmospheric Steam Dumps.

The following areas were tested:

o Correct response of' Plant Process Computer Analog and Digital Points related to the RRS.

o Annunciators associated with the RRS were verified to alarm at their correct setpoint.

o Remote functions associated with the RRS were verified to work correctly.

o All control board hardware associated with the RRS was cycled to verify. proper operation (i.e., valve handswitches, controllers,.pushbuttons, etc.).

o Proper operation-of Hand Switch 111 and 121 for temperature inputs to RRS was verified.

o Proper operation of the Pressurizer Pressure Control System

/N. was-verified. =

(~ '

o. Proper operation of the Pressurizer Level control System was verified.

o Proper operation of the Linear control channel Switches was ve r i.f i ed .

o. Proper operation of the Power Ratio Recorder was verified.

o Proper Operation of the " quick open" signal from RRS to the Atmospheric Dump and Condenser Bypass valves was verified. l o At < 275 F.RCS temperature, proper " Low Temp Over Pressure" operation was verified.

o At 5% power, the temperature input to the 3 element feed control station was verified. .

o At 100% power, the proper feed station response to a reactor trip was tested.

)

Data used to support this test was obtained from P&ID's, I electrical drawings, annunciator setpoints from plant procedures  !

and the Reactor Regulating System Technical Manual. l 1

No deficiencies currently exist for this system. I

_p v

NRC 41 (5) 15

oN ]

% 1

16. Control Element Drive System (CEDS) Test i jJ/~S7p 1

%/ .This System Test was-performed in~ January, 1988. This system covers all ~ the controls, interlocks, alarms and displays ' 1 associated.with moving Control Rods (CEA's)..

.The following areas were tested:

o Correct response offPlant Process Computer Analog and Digital '

Points related to the CEDS.

o Annunciators associated with the CEDS vere verified to alarm atltheir correct setpoint.  ;

o Remote functions associated with the CEDS were verified to work correctly.

o All control board hardware associated with the CEDS was1 cycled to verify proper operation (i.e., handswitches, pushbuttons, etc.).

o All CEA's were moved in Manual Individual mode to verify _

proper operation, indication and alarm.

o All Shutdown Group CEA's were withdrawn in Manual Group (MG) 4 to verify proper operation, indication and alarm. '

("T o All-Regulating Group CEA's were withdrawn in Manual Group and

.(,) Manual Sequential to verify proper operation, indication and alarm.

o The Emergency Evacuation Siren was tested (not part of CEDS, but located with CEDS controls).

Data used to support this test was obtained from Electrical Drawings, and setpoints for the CEDS and associated annunciators

-are from plant procedures.

17. Containment, Heating / Ventilation and Air Conditioning (CTMT/HVAC)

System Test This System Test was performed in January, 1988. Portions of this System are safety related. This system consists primarily of Containment HVAC equipment.

The following areas were tested:

o Correct. response of Plant Process Computer Analog and Digital Points related to the CTMT/HVAC System.

o Annunciators associated with the CTMT/HVAC System were verified to alarm at their correct setpoint.

o Remote functions associated with the CTMT/HVAC System were verified to work correctly.

NRC #1 (5)

(L f~ . '

4 o> All. control board hardware associated with the CTMT/HVAC

. System-was cycled to verify proper operation (i.e., valve handswitches, controllers, pushbuttons, etc.).

o Normal and abnormal.flowpaths were tested.

o Containment Air' Cooler Fans were Operated and verified to work properly.

o Post Incident Recirculation (PIR) Fans were verified to work properly. j o Hydrogen Recombiners and Hydrogen-Analyzers-were verified to work properly.

o CEDM' Cooling Fans, Auxiliary Circulation Fans and Main Exhaust Fans were verified to-work properly.

)

o- The Enclosure Building Filtration System (EBFAS) was verified to work' properly, i

, o The Containment was purged via EBFAS by several available i flowpaths.

o various miscellaneous HVAC Fans were run to verify proper ,

. ._ operation. '

l o The Control Room Air Conditioning (CRAC) was operated in all.

its modes'to verify correct operation.

Data used'to support this test was obtained from P&ID's, electrical drawings, plant surveillance data and annunciator setpoints from plant procedures.

No deficiencies currently exist on this System Test.

18. Instrument Air / Station Air System Test This System Test was performed in January, 1988.

The following areas were tes:ed at 100% Power:

o Correct response of Plant Process Computer Analog and Digital Points related to the Instrument Air and Station Air System.

o Annunciators associated with the Instrument and Station Air system were verified to alarm at their correct setpoint.

o Remote functions associated with the Instrument and Station Air System were verified to work correctly.

l l' /

t A  !

NRC e2 15) 77

(

ll -

U.

o All control board hardware associated with the Instrument Air and Station Air System was cycled to verify proper operation (i.e., valve handswitches, etc.).

t o Normal and abnormal flowpaths were tested.

o All 4 Air Compressors were started and' stopped to verify proper loading sequences and unloading sequences.

o The Unit 1 to Unit 2 Cross Tie capability was tested.

o Alternate cooling supplies to Air Compressors were tested.

o Power Supplies to the air compressors were tested, o Interaction between Station Air and eductint Circulating Water was tested.

Data used to support this test was obtained from P&ID's, electrical drawings, and annunciator setpoints are from plant procedures.

No deficiencies currently exist for this System Test.

19. Radiation Monitoring System Test This 3ystem Test was performed in January, 1988. Most of this system is safety related.

The.following areas were tested at 100% power:

o Correct response of Plant Process Computer Analog and. Digital Points related to the Radiation Monitoring System.

o Annunciators associated with the Radiation Monitoring System were verified to alarm at their correct setpoint.

o Remote functions associated with the Raciation Monitoring System were verified to work correctly.

o All control board hardware associated with the Radiation Monitoring System was cycled to verify proper operation (i.e.,

handswitches, pushbuttons, etc.).

o All Process Radiation Monitors were verified to work correctly including cycling their respective sample fans.

l-o All Area Radiation Monitors ware verified to work correctly.

o The Failed Fuel Radiation Monitor and Unit 1 Stack Monitor were tested for proper operation.

b'"%

.Q NRC 01 (5) yg l

l

E

[ o The' Steam Generator Blowdown and Steam Jet Air Ejector

~ Radiation Monitors were verified to isolate Steam Generator-Blowdown at the correct setpoint.

Data used to support this test was obtained from P&ID's, electrical drawings and annunciator setpoints are from plant procedures.

No deficiencies currently exist on this System Test. ,

20. Waste Disposal System Test This test was. performed in February, 1988. Some portions of this  !

system are safety related, i

The following areas were' tested at 100% Power:

o Correct response of Plant Process Computer Analog and Digital Points related to the Waste Disposal System.

o Annunciators associated with the Waste Disposal System were l verified to alarm at their correct.setpoint.

o- Remote functions associated with the Waste Disposal System were verified to work correctly.  ;

j"%)

(_ o All control board hardware associated with the Waste Disposal L System was cycled to verify proper operation (i.e., valve ,

handswitches., controllers, pushbuttons, etc.). l o Normal and abnormal flowpaths associated with the Waste Disposal System were tested.

o Primary Drain Tank and puench Tank Level and Pressure were adjusted to. verify proper operation.

o RCP bleedoff was switched to the Equipment Droin Sump Tank (EDST) to verify correct operation.

o The Degassifier operation was tested, o various floor sumps were tested such as the Safety Injection Room Sump, RBCCW Floor Sump, Containment Sump, Pipeway Sump and Condenser Pit Sumps, to verify correct operation.

o The Aerated Waste Drain Tanks, Pumps and Waste Gas Surge Tank was tested for proper operatien.

Data used to support this test was obtained from P&ID's, ]

electrical drawings and annunciator setpoints are from plant j procedures. 1 l

() No deficiencies currently exist for this System Test.

NRC 61 (5) g 1

U

[

f

21. Plant ~ Process Computer (PPC) Test E('/'~'Y ;This test'was performed in January, 1988.

The following areas were tested at 100% power :

o Selected Analog Process Computer points that may not have been tested in either system tests were. verified to operate correctly.

o All Annunciators driven by.or related to the Plant. Process ,

Computer (PPC) were verified to operate correctly. l o All. Control Board Hardware associated with the PPC was tested for proper operation.

o All' Programs available to the operator were reviewed for proper operation. Certain programs were identified as-unavailable to the. operator. Those PPC Programs unavailable to the operator are of no significance to training.

oL The interface between the PPC and Control Rod Drive System was tested for proper operation.

o The interface between the Inadequate Core Cooling (ICC) System and PPC was tested for proper operation.

(s s

o The. Safety-Parameter Display System (SPDS) was tested for proper' operation. This testing was conducted by testing all of the following events: Normal Reactor Trip, Large Break LOCA, Main Steam Line Break, Steam Generator Tube Rupture, Los;'of all Feedwater, Complete Loss of AC Power and some DC, and an event which resulted in a loss of several safety functions. All of these events were tested-(one at a time) to verify that the PPC SPDS Displays correctly and properly relayed information to the operator.

Data.used to support this test was obtained from PPC manuals, SPDS manuals, Plant Procedures for Annunciator Setpoints and selected reference plant data.

No deficiencies currently exist for this system.

22. Shutdown Cooling (SDC) System Test This test was performed in October, 1987. This is a safety related system.

The following areas were tested:

o Correct response of Plant Process Computer (PPC) Analog and Digital points related to the SDC System.

O une a2 isi 20

[ o Annunciators associated with the SDC system were verified to

alarm at:their correct setpoint.

o Remote functions associated with the SDC System were verified

.'to work correctly.

o Normal and abnormal flowpaths associated with the SDC System were tested.

o With the RCS at > 300 psia, Boron Equalization and SDC System warmup were verilied.

o With the'SDC System in operation, various' combinations of equipment operation and flowpaths were tested to verify proper response.

Data used to. support'this test was obtained from P&ID's, electrical drawings, Plant Surveillance Data and Annunciator Setpoints from plant procedures.

No-deficiencies were. identified.

23. Reactor Core System Test This test was performed in May, 1988. This test covers all the reactor core physics-characteristics to ensure proper response of O

.s the Simulator to operator actions.

The-following areas were tested:

o Analog Process Computer points which are outputs from the Reactor Core Model were tested for proper operations.

o Annunciators which are outputs from core Model related parameters were tested for proper operation.  ;

o At Beginning of Life (BOL) Reactor Core conditions, the following was tested:

At 10~'% power, the following was measured: Core Reactivity, Isothermal Temperature Coefficient (ITC) and Group 7 CEA g Worth. 1 At 25% power, Axial Shape Index (ASI) was measured.

At 50% power, Equilibrium Xenon Reactivity, Doppler Only Power Coefficient (DOPC), Isothermal Temperature coefficient (ITC). i Core Reactivity Defect and ASI were measured.

At 75% power, ASI was measured.

O' l

i) s/

NPC 41 (5) y L- _ ____- _ _ _ _ _ _ _ _ -

At 100% power, Equilibrium Xenon Reactivity, ASI, Core (N-1 Reactivity Defect, Isothermal Temperature coefficient (ITC),

DOPC,, Core oT and RCS Flow were all measured. The Power Ratio Recorder was also tested, a dropped CEA event was tested and Xenon' Reactivity was tracked following a full power trip.

At an RCS-temperature of 532 F with all control rods inserted

=and 150*r with all control rods inserted, dilutions to criticality were performed to verify core reactivity response to inadvertent dilutions.

o 'At Middle of Life (MOL) and End of Life (EOL) Reactor Core conditions, the followino was tested:

At 10% power, Core Reactivity and Isothermal Temperature Coefficient (ITC) were measured.

At 25% pcwer, ASI was measured.

At 50% pcwe r , . ASI, DOPC and Isothermal Temperature Coefficient (ITC) were measured.

At 100% power, . Equilibrium Xenon Reactivity, ASI, Reactivity Defect, Isothermal Temperature. Coefficient (ITC) and DOPC were all measured.

.'jN o Temperature Effects of RCS Cold Leg temperature on Nuclear (j Excore Detectors was measured at 50% and 100% power, MOL conditions.

o MOL' Dropped CEA measurements were made at 100% power.

Data used to support this test was obtained from the Cycle 9 Westinghouse Nuclear Design Report and reference . plant historical  :

data.

One deficiency was identified and is still open. The Simulator shows no incore power tilt after a dropped CEA is recovered. In the real plant, after recovery of a dropped CEA, a real incore tilt will be present. Depending on the specific dropped CEA, this could cause an annunciator to alarm and Tech Spec Action Statement to be entered. This deficiency is documented by DR 87-2-90. This DR will be resolved by 5-1-91.

Two other minor problems were also identified. While within acceptable limits, the total power defect from 0 to 100% power was 'slightly low. This is documented by DR 88-2-138. Also, the  ;

Doppler Only Power Coefficient (DOPC) as a function of power  !

level were acceptable within applied limits, but could be improved to match predicted / measured DOPC's. This is documented  !

by DR 88-2-140. These 2 DR's are minor adjustments and will be resolved by 5-1-91.

NRC 01 (5) 22

4% -

A) ,

1 ATTACHMENT 3 l

MP2 NORMAL OPERATIONS AND SURVEILLANCE TESTING SEQUENCE This attachment is_ referenced by Section 3 )

~

of the Performance Test Summary.

O l

L l

unces < <, i O

l 1

Attachment 8.2 Step

.f'>i Completed (Initials)

DR Number 1.0 Plant Heatup 1.1 Initialize to the following Simulator

~

conditions:

o RCS Temperature at ~ 100* r o "A" SDC Pung in service maintaining RCS temperatures o Pressurizer at ~ 95% level, vents closed o PORV's in " low" o S/G 1evels 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" and Ops form 2201-1 and perform a normal plant heatup. Record the .

following information at the specified times  !

during the heatup:

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

O aevisioa tevet 'T #

Number of changes ( /f) 1.2.2 Review each Prerequisite, Initial Condition and Precautions listed in OP 2201 to ensure that the simulator is capable of supporting the use of OP2201.

1.2.3 During the heatup, complete (initial) /i '

or N/A steps in OP2201-1 and attach a >

completed OP2201-1 to this procedure. /M Note: Whenever valvo alignments are specified to accomplish the heatup, the valve alignment shall be reviewed for any control board 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.

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

1.2.4 verify that it takes ~ 350 gallons [I to overflow the Pressurizer to the

/l Quench Tank after the pressurizer

'/

- indicates 100% level Record observed number of gallons Jd'O .

1.2.5 Verify that RCS heatup rates are as _

follows and record the requested data: 7 ~/ ~

1.2.5.1 Two RCP's running, heatup rate M -

shall be 20-25 F/hr / 7 "'

o Decay Heat () % power o Observed heatup rate Z Q 'F/hr o Observed heatup rate was at an average RCS temp of /fD F 1.2.5.2 Three RCP's running, heatup rate Ix shall be 30-35 F/hr / 0 ~'

o Observed heatup rate 32 F/hr .

o Observed heatup rate was at an I average RCS temp of yfo, F 1.2.5.3 Four RCP's running, heatup rate d shall be 35-40 F/hr o observed heatup rate 26 *F/hr

-s o Observed heatup rate was at an

() average RCS temp of f30 F Note: Fast time maybe used during heatup as long as it does not interfere with heatup rate determinations i and use of fast time will not interfere with surveillance or other heatup related items.

1.2.6 Pressurizer heatup rate with all heaters d on approximately 60 F/hr. Observed pressurizer heatup rate is 6f F at an average pressurizer water temperature of

2cb F.

1.2.7 RCP "Bleedoff flow lo" alarms clear at l RCS pressure of ~ 1700 psia.

1.2.8 Atmospheric steam dumps maintain an [

RCS temp of ~532 F. Record atmospheric steam dump pressure that maintans 532 F while in auto. fw psia. j l

n j U

Rev.: 0 Date: 5/3/38 l Page: 8.2-2 of 11 ,

NSEM-4.10 I

J

c____ _ _

1.2.9 Perform the following surveillance d procedures prior to starting the a heatup and attach completed surveillance

-] forms to this procedure:

l o service Water Valves SP2612E d o RBCCW Valves SP2611E M l

o SRAS Manual SP2616A A o CSAS Manual SP2606F /5 o SIAS Manual SP2604R /'5 1.2.10 Perform the following surveillance h procedures prior to 200 F RCS /I7 temperature and attach completed surveillance forms to their procedures.

o Transient Temp. Verif. SP2602B _

o RBCCW Pump Ops - FACI SP2611A /Wfli_

o RBCCW Valves - FACI SP2611C r/5 7 ~

o SW Pump Ops - FACI SP2612A #

o SW Valves - FACI SP2612C /3 o EBFAS and CR vent FACI SP2609A-1 A .

o SIT outlet valve (SIAS) SP2603A F

,a o SIT outlet valve (Pressure) SP2603B o Charg. Pmp. Cp. SP260lG /7t51% _

o Boric Acid Op. SP2601C heftf\ _

o ECCS Valves SP2604N #5' ~ ' '

o Borated Water Flow Ops SP2601A /h o Ctnt. Isol. Valve Ops SP2605G 7t/M f- //y '

o D/G Op FAC I SP2613A o Control Rm Wkly Checks SP2619C M ,

o HPSI Pump Op FACI SP2604A  %&\

o HPSI Valves FACI SP2604E WJ/ f o Ctnt. Sump Isol. Viv-FACI SP2604G Ndf3 _

o Control Room Daily Checks SP2619A # 5' 1.2.11 Perform the following surveillance '

295 prior to reaching 300 F RCS temp and attach the completed Surveillance forms:

o Aux Feed Pump Ops SP2610A #f o Aux Feed, Turbine Driven i Ops SP2610B A3 o Aux Feed Valves SP2610C #3 o Ctnt Air Recirc Ops-FACI SP2607A W o PORV Block Valve Ops SP2610F 7#5 i

I I

I O i l

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

1.2.12 Perform the~following surveillance prior to reaching 1750 psia RCS

P. pressure and attach the completed

, ?. =

surveillance forms:

o LPSI Pump Ops - FACI SP2604C M o LPSI Valve Ops - FACI SP2604L #5 o Cor.t. Spray Pmp Ops-FACI SP2606A A o Cont. Spray Valves - FACI SP2606C A 1.2.13 Perform the following surveillance prior to exceeding 5% power and attach the

[

completed surveillance forms:

o Hydrogen Purge SP2608E M o PIR Fan FACI ' SP2608C A o Turbine Drives Aux Feedpunp Ops SP2610B d 1.3 Plant heatup is complete when RCS temperature 8 is ~ 532 F, RCS pressure is being maintained at ~ 2260 psia and all surveillance of step 1.2.9, 1.2.10, 1.2.11, 1.2.12 and 1.2.13 (except MSIV surveillance) are complete.

1.4 Ensure all completed surveillance forms //$

and a completed ops 2201-1 form are attached.

~

1.5 Take a snapshot if required d. //$

2.0 NUCLEAR STARTUP 2.1 Initial Conditons, Plant stable with RCS //$

Tave at ~ 532* F, ~ 2260 psia, pressurizer

~

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.

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

1

i 2.2 Start the "A" CEDM M Set 2.3 Close all KB's, Observe load of ~250 amps on C-04 Record Observed value 7/@ amps. /(7

^

2.4 Start end parallel the "B" M set, observe load to L,a ~ 125 amps each. Record actual //Y~

values "A" M se jZQ amps and {

-"B" M set . lZy,tamos, j 2.5 Using OP2202, " Reactor Startup", withdraw CFA'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 OP2202 that is /

at 15" hg Steam Dump / Bypass System may be put in service.

3.1.13 observe Turbine BypaLG PIC controls

/

RCS T, at 532'r. l 3.1.14 Observe at condenser vacuum of 20" Hg, l a SGTP may be started. [/

/7 1 kJ Rev.: 0 Date: 5/3/88 Page: 8.2-6 of 11 NSEM-4.10

'. i ,.

~

3.1.15 Increase Reactor Power to ~ 7% Tave to l 535*r.

3.1.16' Start the Main Turbine per OP2323A 3.1.17 Synchronize and place the main generator

( /

.on line.

of nj 3.1.18. Increase load to ~ 20% power and at 15% power, observe the LPD and Turbine [ l' 7 RPS trips are enabled.

3.2 Final conditions'are Main Turbine On Line, Reactor Power at ~ 20%, NSST supplying In-House / /. / ~

loads 3.3 Take a snapshot if required. S . [

4.0 IDAD CHANGE 'IO 100% POWER f

Note:- It is the intent of Section 4.0 " Load Changes to 100% power, to be performed from an initial 1 condition that is a continuation of Section 3.0 Plant Startup or from a snapshot taken in Step 3.3.

4.1 Perform a power increase from 20% to 100%

power using OP2204, " Load Change" procedure.

[

4.1.1 Record the revision level and number of changes against the OP2204 that is used. Revision level 8 Number of Changes /

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.

l 4.1.3 Verify that Pressurizer Level follows ~~

Ops form 2204-3 during the power increase.

l Note: Power increase rates used do not need to be restricted, as long as the ji

. power increase rate does not cause excessive pressurizer pressure increases, pressurizer level problems, S/G Level control Problems or i'

.T,,, Control Problems.

4.1.4 Perform SP2619E and attach surveillance forms.

O- I Rev.: 0 Date: 5/3/88 Page: 8.2-7 of 11 NSEM-4.10

4.2 Final conditions shall be T ~572* r, RCS pressure at ~2260 psia afd Generator Load at ~885 MWe.

/ ( 7" 5.0 Reactor Trip and Recovery 5.1 Initialize to 100% full power IC, normal 8 steady state conditions, perform SP26010, SP2602A, and attach surveillance forms.

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

EOP 2525 Rev d # of Changes O . A EOP 2526 Rev I- # of Changes g . /f]

5.3 Manually trip the reactor from C04 7CB P.B.

Carryout, EOP 2525, Standard Post Trip Actions. M 5.4 verify the Startup Rate is -1/3 DPM ~ one minute after the trip - 75 >441. M 5.5 Complete EOP 2526, Reactor Trip Recovery, complete and attach OPS Form EOP 2526-1.  ;

5.6 Perform a reactor startup and plant 6tartup  !

to ~ 20% full power using OP 2202, OP 2203, and OP 2204 as guides.

Q 5.7 Final conditions are: the plant at ~ 20% i full power with th6 reed Reg. Valves in Auto.

6.0 PLANT SHlffDOWN l'( '

6.1 Initialize to a 100% full power IC, normal steady state conditions. 7

/(

6.2 Using OP 2204, decrease load to 20% power. N 6.3 Using OP 2205, perform a plant shutdown. #3 6.3.1 Record the revision level and number M of changes against the OP2205 that is used: E Revision Level J Number of Changes 6.3.2 Review each prerequisite, initial M condition and precaution listed in OP2205 to ensure the simulator is capable '

of supporting the use of OP2205.

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

y <

6.3.3 During the plant shutdown, perform

. surveillance SP2613B on the "B" D/G,

[

/7 SP2608A H2 Recombiner and attach the N> completed surveillance.

6.3.4 Verify that Heater Drain, Condensate ~

/75 and reedwater 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 y

is.~66% with no HD pumps and the High l

Level Dump in operation.

l 6.3.7 When the Main Turbine is tripped, j

l 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.

[

6.3.9 verify the Turbine Bypass / Steam Dump to. ~532* r Valves and S/Gcontrol.

pressureRCSto T"880 psia.

6.4 Take a snapshot if required ~

. 4/# '

O 7.0 REACIOR SHUTDONN 7.1 Obtain Initial Conditions co a Reactor Shutdown using OP2206, by continuing from Section 6.0 or initializing to an IC which is critical Tave ~532" F, RCS pressure = 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- b Number of Changes K 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:

7.4.1 CFA's insert properly in Manual Sequential l

O Rev.: 0 Date: 5/3/88 Page: 8.2-9 of 11 NSEM-4.10

1 7.4.2 Reactor Power level decreases at no greater than - 1/3 DPM 7.4.3.Inward CEA motion for an individual CEA stops at the lower electrical limit.

7.4.4 CFA bottom lights go on af ter 'ICB's //3 are opened.

7.5 Final Conditons shall be; all CEA's fully. [-

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

7.6 Take a snapshot if required ~

M 8.0. PLAffr COOLDOWN 8.1 Reset to an Initial condition with the ._

plant at.~532 RCS T RCS Pressure at

~2260 psia. All CF1 4 , inserted, the Steam Dcusp Bypass System in service,' one condensate pump in service, one SGFP in service and RCS boron concentration sufficient for meeting shutdown margin. Alterr.atively using the IC at the conclusion of section 7.0 is acceptable.

8.2 Connence a cooldown using OP2207 " Plant Cooldown" verify or record the following Q during the Plant Cooldown:

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

Revision Level M Number of Changes 7 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 off.

O.

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

l

.a 8.2.5 verify Pressurizer Pressure and [

Temperature decrease when Aux Spray

\ .

is initiated.

8.2.6 Verify that MSI Block is permitted at

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

8.2.7 verify that SIAS Block is permitted M at < 1750 psia RCS pressure (3 out of 4).

8.2.8 verify that SDC warmup on 1 LPSI pump M is ~12 F/Br.

8.3 Final conditions are: RCS temperature M

<200 F on SDC, i.e. in Mode 5. Pressurizer vents opened and Pressurizer corrected level in at ~50%, Pressurizer temperatures are

<200* ' r, excess letdown is in service at

~140gpm and S/G-levels are at ~80%.

O

('

(s Rev.: 0 Date: 5/3/88 Page: 8.2-11 of 11 NSE:M-4.10

i l

l c

o- l ATTACHMENT 4 MP2 SURVEILLANCE THAT CAN BE PERFORMED OW THE SIMULATOR l

l l

i This attachment is referenced by Section 3 of the Performance Test Summary.

O 4 NRC91 (7)

%)

I

Figure 7.3 Page 1 of 2 SURVEILLANCE LISTING UNIT 2 O

(Yes/No)

Sequential Number .litle Procedure # To Be Tested

1. Borated Water Flow / Pump Op SP2601A Yes

2. Boric Acid Ops SP2601C Yes

3. Power Range Calibration SP2601D Yes

4. Charging Pump Ops Fac 1 SP2601G Yes

5. Charging Pump Ops Fac 2 SP2601H No

6. Reactor Coolant Leakage SP2602A Yes

7. Transient Temp. Verification SP2602B Yes

8. SIT Valve Ops (SIAS) SP2603A Yes

9. SIT Valve Ops (Press) SP2603B Yes

10. HPSI. Pump Ops Fac 1 SP2604A Yes

'11. HPSI Pump Ops Fac 2- SP2604B No

12. LPSI Pump Ops Fac 1 SP2604C Yes

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

17. Ctmt Sump valve Fac 2 SP2604H No

18. LPSI, Valve Ops Fac 1 SP2604L Yes

'19. LPSI Valve Ops Fac 2 SP2604M No

.20. ECCS Valve Ops SP2604N- Yes ,

21.

SIAS Manual Test SP2604R Yes I SP2605G Yes

)2.Ctmt'ValveStrokeTime

3. Ctmt-Purge Valve Power SP2605G Yes

-('24.ReactorHead&

Pzr Vent Sol V1 Oper. SP2605N Yes l

25. Ctat Spray Pump Fac 1 SP2606A Yes

26. Ctat Spray Pump Fac 2 SP2606B No

27. Ctat Spray Valve Ops Fac 1 SP2606C Yes 28.-Ctat Spray Valve Ops Fac 2 SP2606D No

29. CSAS Manual Test SP2606F Yes

30. CAR Fac 1 SP2607A Yes f31. CAR Fac 2 SP2607B No

32. H, Recombiner SP2608A Yes c 33. PIR Fan-Fac 1 SP2608C Yes j

34. PIR Fan Fac 2 SP2608D No '

35. H Purge Ops SP2608E Yes

36. E,FAS B Fac 1 SP2609A Yes

'~37. EBFAS Fac 2 SP2609B No

38. Motor Driven Aux Feed Pumps SP2610A Yes k L% nf Approved: hd< 14911' l (ASOT

() Rev.: 0 Date: 6/29/88 Page: 7.3-1 1 of 2 h;EM-4.10 L-____ _ __ _ -

Figure 1.3 Page 2 of 2 SURVEILLANCE LISTING UNIT 2

<i .c (Yes/No)

Sequential Number Title Procedure # To Be Tested

39. Turb Driven Aux Feed Pump SP2610B Yes ,

40. Aux Feed Valves SP2610C Yes  !

41. MSIV Partial Closure SP2601D Yes l

42. MSIV Closure SP2601E Yes  !

43. PORV Bloc.; Valve Ops SP2610F Yes l

44. RBCCW Pp Fac 1 SP2611A Yes j

45. RBCCW Pp Fac 2 SP2611B No  !

46. RBCCW Fac 1 Valves SP2611C Yes {

47. RBCCW Fac 2 Valves SP2611D No  !

48. RBCCW Valve-Ops SP2611E Yes l

49. SW Pp Fac 1 SP2612A Yes

50. SW Pp FAc 2 SP2612B No

51. Service Water Valves SP2612C Yes 1

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 1

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

58. Control Room Shift checks SP2619A Yes

59. Control Room Weekly Checks SP2619C Yes

60. Start UP Surveillance Checks SP2619D Yes 7-,51. Control Roc:m Monthly Checks SP2619E Yes

)2. CEA Partial Movement SP2620A Yes

("63. CEA Group Deviation Verification

~

SP2620B Yer,

64. CMI Verification SP2620C Yes l

{

Approved: a 7-D M i ASOT l ]'

(~

t Rev.: 0 Date: 6/29/88  !

Page: 7.3-1 2 of 2 NSEM-4.10 1

l. .

o -ATTACHMENT 5 l l l

MP2 MALFUNCTION TEST ABSTRACTS l

l l

This attachment is referenced by Section 4 of the Performance Test Summary.

r3 V

O nment tai y

g-) 1. Loss of Coolant Malfunction Abstracts

\J SG02 Malfunction - Steam Generator Tube Rupture This malfunction test was conducted in July, 1988. This malfunction. is capab? of inserting up to a 3000 gpm tube rupture to either Steam Genet tor. 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 Conditions, equilibrium xenon, steady state Conditions for a period of ~ 12 ,

I minutes. Baseline data was from NU Retran model computer runs specific to MP2. No deficiencies were identified.

CV01 Malfr7ction - Unisolable letdown line rupture in Containment This malfunction test was conducted in July, 1988. This malfunction is capable of inserting up to a 200 gpm leak from the letdown line (RCS toThis CVCS) in Containment to the Containment malfunction was tested at 100%, SOS and Atmosphere / Floor.

10% 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 identified.

CV02 Malfunction - Isolable letdown line rupture outside of Containment This malfunction test was conducted in July, 1988. This gg malfunction is capable of inserting up to a 200 gpm leak from the (s/

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, Midd', of Lite Core conditions, equilibrium Xenon, steady state conditions for a period of ~ 8 minutes. Baseline data was from CEN 128 Case E2. No deficiencies were identified.

CVl8 Malfunction - Isolable letdown line rupture inside Containment This malfunction was conducted in July, 1988. This malfunction is capable of inserting up to a 200 gpm leak from the letdown line (RCS to CVCS) This in Containment to the Containment malfunction was tested at 100%, 50% and Atmosphere / Floor.

10% severity. All tests were started from 100% Power, Middle of Core Life conditions, Equilibrium Xenon, steady state conditions for a peried of ~ 8 minutes. Baseline data was from CEN 128 Case E2. No deficiencies were identified.

J l

($) I wac.1 <a' 2 i

{

l r3 RC02 Malfunction - RCS Hot Leg Break U This malfunction test was conducted in July, 1988. This l

malfunction is capable of inserting 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. Two deficiencies were identified. They are: (1) Reactor Vessel Level dropped to 0%

and stayed there. It should eventually have recovered to > 19%

level due to safety injection. This deficiency is covered by DR's 87-2-144, 87-2-28 and 87-2-185 which will be resolved by 5-1-91. (2) The Containment Area Radiation Monitors went too high. This deficiency is covered by DR 89-2-18 which will be resolved by 5-1-91.

RC03 Malfunction - RCS Cold Leg Break This malfunction test was conducted in September, 1988. This malfunction is capable of inserting up to a 100,000 lbm/see leak

,f x from any of the fcur RCS Cold Legs to Containment. 100% severity 1 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. This deficiency is covered by DR's 88-2-55, 87-2-144, 87-2-28 and 87-2-185 which will be resolved by 5-1-91. (2) The Containment Area Radiation Monitors went too high. This deficiency is covered by DR 89-2-18 which will be resolved by 5-1-91. (3) Core Exit Thermocouple (CET's) stayed ~ 100 F superheated through the event. As the core refloods with safety injection, they should eventually go subcooled, they do not. This deficiency is covered by DR 87-2-185, L9-2-55 and 87-2-28, which will be resolved by 5-1-91.

RC04 Malfunction - Unisolable Reactor Head Vent Leak This malfunction test was conducted in July, 1908. This malfunction is capable of inserting up to a 170 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 s identified NEC#1 (8) 3

c .RC05 Malfunction - Pressurizer Safety valve railure b.5) This malfunction test was conducted in July, 1988. This malfunction is capable of inserting 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 was from an NU Retran analytical case for MP2. No deficiencies were identified.

RC06 Malfunction - Pressurizer Relief Valve (PC2V) railure l

This malfunction test was conducted in August, 1988. This malfunction is capable of inserting 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 x conditions.are: CV03, CV13, Cvl4, CV20, CV21, RC20, and SG01.

All of these malfunctions give a maximum RCS leak rate of 100 gpm or less. The Cause and Ef fects dese'.iptions ma; be referred to  !

for each of these malfunctions to describe the malfunction l characteristics. One deficiency was identified during these l tests. The charging header pressure indication for varicus l severities of malfunctions CV13, CV14, CV20 and CV21 was not accurate. This deficiency is tracked by Deficiency Report 2-88-0142, which will be resolved by 5-1-91.

I

2. Loss of Instrument Air Malfunction Abstracts IA01 Malfunction - Instrument Air Header Rupture This malfunction test was conducted in January, 1989. This malfunction is capable of producing a 1500 Standard Cubic reed per minute (Scrm) leak on the instrument air header. This malfunction was tested at 10%, 50%, 85% and 100% severity. The test was started from 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.

(

..cm o, ,

IA03 Malfunction - Loss of Instrument Air in Containment 7( )

This malfunction test was conducted in January, 1989. 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, i 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&ID's. No deficiencies

~ were identified.

Other malfunctions which may be used to provide loss of instrument air conditions are: IA02, IA04 and IA05. The Cause and Effects-descriptions may be reference to for each of these malfunctions to describe the malfunction characteristics. No deficiencies were identified in any of these malfunctions.

3. Loss or Degraded Electrical Power - Malfunction Abstracts

'ED03 Malfunction . Loss of offsite 345KV Electrical Lines This malfunction.was tested in January, 1989. This malfunction is capable of inserting a loss of any one of the four offsite 345KV transmission lines serving Millstone Station. This

'O malfunction may also be used to give a loss of any combination of

'2 or 3 lines or a loss of all 4 lines giving an Station Loss of

-Normal. Power (LNP). This malfunction was tested by individually tripping each of the 4 offsite lines one at a time and then tripping out all 4 lines at once (LNP). All Tests were conducted starting from 100% Power, Middle of Life Core conditions, equilibrium Xenon, steady state conditions. For the individual offsite line-trips, the tests were ended when all 345 KV breakers had gone to the correct position. For the LNP test, the test was ended when the Emergency Diesels had energized the 4160 Volt Emergency Buses. Baseline data was from best estimate analysis, use of plant procedures, electrJcal drawings and P&ID's. No deficiencies were identified.

ED05 Malfunction - Loss of 4160 Volt Bus This malfunction was tested in January, 1989. This malfunction is capable of inserting a loss of any of the 6 4160 Volt Buses due to ground fault. Each of the 6 bus malfunctions were individually tested starting from 100% power, Middle of Life Core f conditions, equilibrium Xenon, steady state conditions. Each test was ended after verifying the loss of power to the correct i 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.

O NRC#1 (Bl 5

)

- - - - - _ _ _ - _ _ - - _ _ - _ - _ _ _ _ _ _ _ _ __ _}

ri ED08 Malfunction - 6.9 KV Bus failure to Auto Transfer l,

This malfunction was tested in January, 1989. This malfunction is capable of inserting 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 possible. Baseline data was from best estimate analysis, use of plant procedures, electrical drawings, and P&ID's. No deficiencies were identified.

ED10 Malfunction - Loss of 125 VDC Vital Bus This malfunction was tested in January, 1989. This malfunction is capable of inserting 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. Each test ended after DC loads were verified to be lost. Baseline data was from best estimate l analysis, use of plant procedures, electrical drawings and P&ID's. No deficiencies were identified.

EDl6 Malfunction - Loss of Vital 120 volt Instrument AC This malfunction was tested in January, 1989. This malfunction f is capable of inserting a loss of any of the 4 Vital 120 Volt AC <

()

r+

buses. Each of the 4 malfunctions were individually tested starting from 100% power, Middle of Core Life conditions,

{

l equilibrium Xenon, steady state conditions. Each test was ended j 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.

EG08 Malfunction - Diesel Generator Output Breaker Failure This malfunction was tested in January, 1989. This malfunction is capable of preventing either 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 bus energized, to verify the D/G output breaker would not close. Baseline data was from best estimate analysis and electrical drawings. No deficiencies identified.

l l

l wneet (s) 6 i

. EGil Malfunction - Diesel Generator Auto Start Failure This malfunction was tested in January, 1989. 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.

Other malfunctions which may be used to provide Loss or Degraded Electrical Power are: ED01, ED02, ED04, ED06, ED07, ED09, EDil, ED13, ED14, ED15, ED17, ED18, EG02, 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.

4. Loss of Forced RCS Flow - Malfunction Abstracts RC11 Malfunction - RCP Locked Rotor This malfunction test was conducted in September, 1988. This malfunction is capable of inserting a Locked Rotor on any of the 4 RCP's. This malfunction was tested for each of the 4 RCP's starting from 100% power, Middle of Core Life conditions, r~ equilibrium Xenon, steady state conditions for a period of 1 x minute each. Baseline data is from the MP2 FSAR section 14.6.

One deficiency was identified. RCS temperatures, T T and T yy did not increase enough on the Simulator and the,relo,re Pre,ssurizer level and pressure did not increase enough on the 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.

This deficiency is tracked by DR 89-2-14, which will be resolved by 5-1-91.

Other malfunctions which may be used to give Loss of RCS Flow conditions are: CC06, ED04 and RC13. The cause and Effect descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. One deficiency was identified for RCl3, RCP Shaft Break. For the shaft break on the "A" and "B" RCP's, RCP Amps for the running RCP's were not always accurate. This deficiency is tracked by DR 89-2-2 and will be  !

resolved by 5-1-91.

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 I test abstracts.

f

.co o, ,

i

m 5. Loss of Condenser vacuum / Loss of Condenser Level Malfunction y,) Abstracts FW01 Malfunction - Loss of Condenser Vacuum This malfunction test was performed in January, 1989. 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.

FWO2 Malfunction - Condenser Level Transmitter Failure This malfunction was tested in January, 1989. O to 100%

malfunction severity causes the Condenser Hotwell Level transmitter to indicate O to 100% level to the Condenser reject valve. The malfunction was tested at the end point conditions 1%

(reject valve open) and 100% (reject valve closed). The test was started at 100% power, equilibrium Xenon, steady state condition and ended when actual Hotwell level had stabilized. Baseline data used was from best estimate analysis and P&ID's. No deficiencies were identified.

FW12 Malfunction - Condensate Surge Tank Level Transmitter

< Failure

(

This malfunction was tested in January, 1989. O to 100% j malfunction severity causes the Condensate Surge Tank (CST) level transmitter to indicate 0 to 100% level to the Hotwell makeup valve. This malfunction was tested at the endpoint conditions 10% severity (makeup valve shut) and 100% severity (makeup valve open). The test was started at 100% power, equilibrium Xenon, steady state conditions and ended when Hotwell level stopped .

changing either by Condensate Pump trip or Hotwell reject valve I action. Baseline data used was from best estimate analysis and P&ID's. No deficiencies were identified.

6. Loss of Service Water Malfunctions Abstracts l

SWO5 Malfunction - Unisolable Service Water Header Rupture This malfunction was tested in January, 1989. 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 Watet headers. All tests were started from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions and ended when service water flow was lost to the major Service Water loads. Baseline data was from best estimate analysis and P&ID's. No deficiencies were identified.

O NRC#1 (8) g l

j

l r- Other malfunctions which may be used to give Loss of Service

't Water to Service Water headers or individual components are:

SWOl, SWO2, SWO3, SWO4, SWO6, SWO7, SWOB 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 l

RH01 Malfunction - Low Pressure Safety Injection (LPSI) Pump Trip This malfunction test was conducted in September, 1988. This malfunction was started from RCS water level at the minimum level of Centarline of the RCS Hot Leg, with Shutdown Cooling in operation at Beginning of Life conditions with maximum decay heat. The purpose of this test was to verify proper response to loss of Shutdown Cooling at the most limiting conditions. The test was run for a period of 15 minutes. Baseline data was best estimate analysis. Two deficiencies were identified. (1) After saturation was reached in the RCS, Core Exit Thermocouple (CET's) temperatures continued to increase, but should not, since once boiling is reached, CET's should stabilize. This deficiency is tracked by DR 89-2-15 and will be resolved by 5-1-90. In the interim, training is administrative 1y controlled, to preclude negative training. (2) Wide Range Hot Leg temperature does not increase after the loss of Shutdown Cooling. This deficiency is tracked by DR # 88-2-112 and will be resolved by 5-1-91.

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 deficiencies were identified in any of these malfunctions.

8. Loss of Component Cooling - Malfunction Abstracts CC02 Malfunction - RBCCW Header Rupture This malfunction test was conducted in January, 1989. This malfunction is capable of inserting up to 2000 gpm rupture on either Reactor Building Component Cooling Water (RBCCW) Header.

This malfunction 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.

Other malfunctions which may be used o give a Loss of Component Cooling to headers or me.i.p_nents are:

O V

une.1 (e> 9

k L

'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 TPO4 Circulating Water to Condenser: CW01, CWO2, CWO3, CWO4, CWO6, CWO7 and.CC08.

The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunctions characteristics.

No deficiencies were identified in any of these malfunctions.

9. Loss of Normal Feed / Feed System Failure - Malfunction Abstracts FWO6 Malfunction - Main Feed Pump Trip This malfunction test was conducted in. January, 1989. This.

malfunction isfcapable of inserting a trip signal on either Main Steam Generator Feed Pump (SGFP). Both main SGFP trips were tested starting from 100% power, Middle of Core Life conditions, equilibrium Xenon,. steady state conditions and ending after a.

1 reactor trip' occurred on Low Steam Generator Level and i verification that the tripped main SGFP could be restarted following malfunction removal. Baseline data was from best estimate analysis and P&ID's. No deficiencies were identified.

Other malfunctions which may be used to give Loss of Normal Feed

qq' or a normal feed system failure are:

V FW23, FWO3, FWO4, FWOS, FWO7, FWO8, FWO9, FW10, FW11, FW13, FW14, FW15, FW16, FW18, FW22, FW28, FW29, MS11, 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.

= 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.

4

11. 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.

O mace 2 (s) 10

l p'

i 1'

() ED16 Malfunction - Loss of Vital 120 volt Instrument AC Buses This malfunction was tested in January, 1989. This malfunction is capable of inserting a loss of 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 ESAS actuation cabinet. 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.

ESO4 Malfunction - Loss of Vital 120 Volt Instrument AC to ESAS This malfunction was tested in January, 1989. This malfunction is capable of tripping the input breakers to each of the 4 ESAS sensor channels and depending on the malfunction, 2 ESAS actuation cabinets. 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 lost and for malfunctions which lost actuation cabinets, the fuses were replaced to show they could be restored by procedure. Baseline data was from best

(~g 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 FiTIure j This malfunction was tested in January, 1989. 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%

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 i is from best estimate analysis and P&ID's. No deficiencies were f identified. j I

Other malfunctions which may be used to give a loss of Protection j System Channel are: RCl4, RP01, RP05, RP06, RP07, RP08, RP09, z RP10, RPil, RP12, RP13, RP14, RP22, RP24 and RP25. The Cause and j Effects descriptions may be referred to for each of these i malfunctions to describe the malfunction characteristics. No  !

deficiencies were identified.

i NRCt1 (8) y

i l

n

(_) 12. Control Rod Failure - Malfunctions Abstracts RD01 Malfunction - Dropped Control Rod (CEA)

{

This malfunction test was performed in September, 1988. This malfunction is capable of dropping 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 malfunctions may be used for positioning CEA's at other than full insertion. "his malfunction .

was tested in several stages. First, starting f'om 100% Power, Beginning of Core Life conditions, equilibrium Y.enon, steady state, CEA's, 42, 48, 44 and 46 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 sufficient time  ;

for RCS Parameters to stabilize, following the CEA drop.

The next portion of the test was to drop CEA #42 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 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.

The next portion of the test was to drop CEA #50 starting from 100% Power, Beginning of Core Life, equilibrium Xenon, steady state conditions and verify the response of CEA #50 was bounded  ;

by CEA #42. .

l The next portion of the test was to drop 1 CEA in each of CEA groups B, 1, 2, 3, 4, 5, 6 and 7 starting from 100% Power, Middle of Core Life, steady state conditions, equilibrium Xenon to I verify proper alarm response.

Finally, all CEA's not yet tested were dropped, one at a time starting from 100% power, Middle of Core Life conditions, steady state, equilibrium Xenon conditions to verify proper alarm and RPS response.

Baseline data is from Reference Plant dropped CEA events and best i estimate analysis. Three deficiencies were identified. They i are:

(1) The Simulator shows no incore radial power tilt after a  !

dropped CEA is recovered. In the real plant, after recovery of a dropped CEA, this could cause an annunciator to alarm and Tech Spec Action Statement to be entered. This Os deficiency is documented by DR 87-2-90. This DR will be

,,c,, ,,

resolved by 5-1-91.

12

L 1 \

l l

e- l

(,) (2) The Simulator shows the reactivity worth of the dropped CEA l

to be too large, resulting in an excessive power drop due to i the dropped CEA. This deficiency is tracked by DR 89-2-16 and will be resolved by 5-1-91.

(3) The Simulator excore detector response due to rod shadowing I from the dropped CEA is not sufficiently accurate. This deficiency is documented by DR 88-2-122, and will be  ;

resolved by 5-1-91. l I

RD02 Malfunction - Stuck Control Rod (CEA) j This malfunction was tested in January, 1989. This malfunction is capable of making any of the 61 CEA's stick at their current  ;

position. All 61 CEA's were tested by starting from 100% Power, I Middle of Core Life, equilibrium Xenon, steady state conditions l and verifying that a reactor trip did prevent the malfunctioning j stuck CEA's from inserting. Baseline data was from Best Estimate analysis. No deficiencies were identified. 3 RD03 Malfunction - Partial CEA Drop This malfunction was tested in January, 1989. 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 at 10%, 50% and 100% severity for CEA #1 and at 100% severity for all other CEA's. Tests were started from 100% Power, Middle of O,s Core Life conditions, steady state, equilibrium Xenon conditions.

Tests were ended when verifying that the individual CEA had dropped, alarm and position indication was correct, and RCS

-temperature and power had responded to the partial CEA Drop.

Baseline data was best estimate analysis. No deficiencies were identified.

RD04 Malfunction - Ejected CEA i This malfunction was tested in September, 1988. 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 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 established. Baseline data was from FSAR analysis of the Ejected CEA event for reactivity effects and NU Retran analysis to analyze the RCS leak portion of this event.

One deficiency was identified. The reactivity worth of the Ejected CEA is too small, and does not cause a reactor trip if I the CEA was ejected from the fully inserted position. This deficiency is tracked by DR 88-2-114 and will be resolved by 5-1-91. This is not a training issue since we do not train on this event in the Simulator.

O uncei ts' 13

a

() RD09 Malfunction - Uncoupled CEA  !

l This' malfunction was tested in September, 1988. This malfunction q L is capable of uncoupling any of the 61 CEA's, from their i driveshaft. This malfunction was tested by uncoupling all 61 CEA's starting from Hot Zero Power, Beginning of Core Life conditions and verifying that a power (reactivity) drop took j Baseline data was best estimate place as each CEA was uncoupled.

analysis. One deficiency was identified. For only CEA #1, an L

improper " dropped CEA" alarm was received. This deficiency is tracked by DR # 88-2-122, which will be resolved by 5-1-91. l

)

Other malfunction which may be used to give control Rod Failures ]

are: RDOS, RD08, RD10 and RD11. The cause and Effects  !

descriptions may be referred to for each of these malfunctions to  !

describe the malfunction characteristics. No deficiencies were I identified. )!

13. Inability to Drive Control Rods (CEA's) Malfunction Abstract RD06 Malfunction - Switch Failure This malfunction was tested in January, 1989. 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.

14. Fuel Cladding Failure - Malfunction Abstracts CR01 Malfunction - Fuel Clad Failure This malfunction was tested in February, 1989. This malfunction is equivalent to a fuel rod gap release from 2% failed fuel at.1%,

100% malfunction severity. This malfunction was tested at 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.

RC01 Malfunction - RCS Crud Burst This malfunction was tested in January, 1989. 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 conditions and ended when the Letdown Radiation Monitor reading had stabilized. Baseline data was from the FSAR and best estimate analysis. One deficiency was written.

O NRCt1 (8) 4

(-

100% severity did not quite allow full scale to be reached on the L

, ()

,s Letdown Radiation Monitor.

89-2-3.

Thic deficiency is tracked by DR Since this deficiency is minor and is transparent to the

(

i examinee, no specific date for resolution will be given. It will i be resolved as time permits as discussed in section 11 of this document.

L 15. Turbine Trip Malfunction Abstracts l

l TC01 Malfunction - Turbine Trip This malfunction was tested in January, 1989. This malfunction inserts a turbine trip. It was tested at 100% power, steady state conditions and ended after verification of Reactor and Generator trips. Baseline data was best estimate. No L deficiencies were identified.

TC09 Malfunction - Electro Hydraulic Control (EHC) Pump Failure This malfunction was tested in January, 1989. This malfunction i causes either EHC pump to trip. Tripping both EHC 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 EHC pressure. Baseline data was best estimate. No deficiencies were identified.

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

(_) CW01, FWOl, TPO4, TUO1, TUO3 and TUO4. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunctions characteristics. One deficiency was identified. In malfunction TUO4, with no Turbine Motor Suction Pump, the main Turbine should trip when Turbine shaft speed exceeds 1300 RPM during a Turbine startup. This deficiency is tracked by DR 89-2-7 and will be resolved by 5-1-91.

16. Generator Trip - Malfunction Abstracts EG01 Malfunction - Main Generator Trip This malfunction was tested in January, 1989. This malfunction inserts a Main Generator trip. It was tested starting from 100%

power, steady state conditions and ended after verification of a J Generator / Turbine and Reactor trip. Baseline data was best ]

estimate. No deficiencies were identified.

l

)

O nacet iei 15

(

() ED01 Malfunction - Loss of Normal Station Service Transformer (NSST)

This malfunction was tested in January, 1989. This malfunction causes a loss of the NSST. It was tested from 100% power, steady state conditions and ended after a Generator trip, 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.

17. Reactivity / Core Heat Removal Automatic Control System Failures -

Malfunction Abstracts CV09 Malfunction - Inadvertent Dilution This malfunction was tested in January, 1989. 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, Cvi6 and CV19 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 This malfunction was tested in January, 1989. This malfunction allows either S/G main feed reg valve to be positioned from 0%

(full closed) to 100% (full open) depending on 0 4 100%

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 seen to trend properly due to the malfunctioning valves. Baseline data was best estimate analysis. No deficiencies were identified.

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 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 Steam Generator Water Inventory Control Systems and reactivity Automated Control 3

Systems.

O NRCal (8) 16

l 1

l i

() 18. RCS Pressure / Volume Control Systems - Malfunction Abstracts RX01 Malfunction - Pressurizer Spray Valve Controller Failure This malfunction was tested in January, 1989. This malfunction simulates an controller failure 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 best estimate analysis. No deficiencies were identified.

Other malfunctions which cause RCS Pressure / Volume Control Systems Failures are RX02, RXO3, RXO4, RXO6, RX07 and RX08. The cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics.

No deficiencies were identified.

19. Reactor Trip - Malfunction Abstracts RP02 Malfunction - Spurious Reactor Trip This malfunction was tested in January, 1989. This malfunction inserts a momentary, spurious Reactor Trip signal from the manual Reactor Trip System. This malfunction was tested starting from f-)5

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

20. Main Steam Line and reed Line Breaks - Malfunction Abstracts MS01 Malfunction - Main Steam Line Break - Inside Containment This malfunction was tested in August 9 1988. The malfunction is capable of inserting 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. All tests were conducted from 100% power, Middle of Core Life conditions, equilibrium Xenon, steady state conditions for a period of 8 minutes. Baseline data is from CEN-128 Case A4D. No deficiencies were identified.

O

..<m m, 1,

)_ MS02 Malfunction - Main Steam Line Break - Outside Containment This' malfunction was tested in August, 1988. This malfunction is similar to MS01 except'it is outside containment. MS02.is capable of inserting 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 state conditions for a period of 8 minutes. Baseline data is from CEN 128 Case A4D. No' deficiencies were identified.

FW25 Malfunction - Feedline Break - Inside Containment This malfunction was: tested in September, 1988. This malfunction is capable of inserting 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 Generatorn. 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. One deficiency was identified. At severities > I 50%, the amount of RCS cooldown is not sufficient. This deficiency is tracked by DR.88-2-115 which will be resolved by 5-1-91.

- Other malfunctions which cause Main Steam or Feedwater Leaks are

$m) MS03 and FW27. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunction characteristics. One deficiency was identified for l MS03. For 100% severity the amount of RCS Cooldown is not fast enough to generate a Reactor Trip due to the High Power Trip.

j This deficiency is tracked by DR 89-2-17, which will be resolved by 5-1-91. .

21. Nuclear Instrumentation (NI) - Malfunction Abstracts RP17 Malfunction - Wide Range Detector High Voltage Failure This malfunction was tested in January, 1989. This malfunction causes a loss of High Voltage to any of the 4 Wide Range detectors causing loss of indication. This malfunction was tested for each Critical at 10-4 pf the 4 detectors power and ended starting when allf rom Hot Zero indications Power, were verified to go to the failed position. Baseline data was from best estimate analysis and electrical drawings. No deficiencies were identified.

O

""#'* '*' 18

I l) RP23 Malfunction - Random Noise on Summed NI Power This malfunction was tested in January, 1989. 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 concitions. The test was ended when the spike was verified to occur and indications responded No to the spike. Baseline data was from best estimate analysis.

deficiencies were identified, other malfunctions which cause Nuclear Instrumentation failures are malfunctions: RP15, RP16, RP18, RP19, RP20 and RP21. The Cause and Effects descriptions may be referred to for each of these malfunctions to describe the malfunctions characteristics.

No deficiencies 'Jere identified.

22. Process Instrumentation, Alarms & Control System Failures -

Malfunction Abstracts CR02 Malfunction - Incore Detector Failure This malfunction was tested in February, 1989. This malfunction causes any of the 180 Incore detectors to read 0 to 1100 millivolts corresponding to 0% to 100% malfunction severity.

l

' - This malfunction was tested for all 180 Incore detectors at 10%,

l 50% and 100% malfunction severity. The test was started at 100%

power, Middle of Core Life, steady state conditions and ended when C 1 expected alarms were received. Baseline data is from best estimate analysis. No deficiencies were identified.

CC05 Malfunction - RBCCW Temperature Control railure This malfunction was tested in January, 1989. This malfunction causes the RBCCW Temperature Transmitter to fail to O to 100 F corresponding to O to 100% malfunction severity for any of the 3 RBCCW Heat Exchangers. This malfunction was tested at 101 and 100% severity for all 3 RBCCW Heat Exchangers. The test was started from 100% power, steady state conditions and ended when trends were established for key parameters on the RBCCW and Service Water Systems, Baseline data was from best estimate analysis. No deficiencies were identified.

SIO3 Malfunction - Safety Injection Tank (SIT) Lew Pressure Malfunction This malfunction was tested in January, 1989. This malfunction causes up to a 20 psi / minute drop in any of the 4 SIT tanks, causing indication and alarm to the operators. This malfunction was tested at 10%, 50% and 100% severity for all 4 SIT tanks.

O

""#'* '*' 19 l

E _ - - - ___ -_- - _ - _ - _ _ ___ _

f- The test was started from 100% power, steady state conditions and

'(j) ended when SIT pressure decrease rates were verified. Baseline data was from best estimate analysis. No deficiencies were identified.

Other malfunctions which cause Process Instrumentation, Alarm and Control System Failures are: Malfunctions CR03, CV05, Cv06, RM01, RM02, RM03, RX09, RX15, SIO2, SIO1, CV07, Cvil, CV15, CWO3, EG02, EG07, EG10, EG12, FWO2, FW10, FW12, FW14, FW16, FW28, MSO4, MS08, MS13, MS16, RH03, RH07, RX10, RX11, RX12, SWO7, SWO8, TC02, TC04, TC07 and TP02. The Cause and Effects descriptions may be referred to for eacn of these malfunctions to describe the malfunction characteristics. No deficiencies were identified.

23. Passive Malfunctions in Engineered Safeguards SyLtems (ESAS) -

Malfunction Abstracts ES03 Malfunction - Engineered Safety Feature (ESAS) Actuation Medule - Failure to Actuate This malfunction was tested in Janne;y, 1989. This malfunction causes any of 7 Actuation Modules, to fail to start its components on an accident signal. All testing was started from 100% power, steady state conditions. The tes; was ended after each of the 7 actuation modules was verified not to start its equipment on the appropriate accident signal. Baseline data was

(~ from best estimate analysis and operating procedures. No

( deficiencies were identified.

ES,01 Malfunction - Automatic Auxiliary Feedwater Failure This malfunction was tested in January, 1989. 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.

Other malfunctions which cause or can cause passive malfunctions in ESAS components are: ES06, CC01, CH01, Cv04, CV17, EG09, EGil, FW2C, RH06, RH07, SIO4 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.

..r., o, 2e

g s' Malfunction

(_) 24. Failure of Automatic Reactor Trip System (RPS) -

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, 1989 and consisted of starting from 100%

power, steady state conditions, inserting the auto trip. bypass for all 10 RPS monitored parameters 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 January, 1989. This malfunction causes any of the 4 manual trip pushbuttons to not vnrk. The automatic RPS trip is unaffected. The malfunction was tested by ctarting 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.

,Q

(> 25. Reactor Pressure Control System Failure (BWR's)

Millstone 2 is a PWR, this.is not applicable.

(

l l

0 NRC#1 (8) 21

L i

(3 w.)

ATTACHMENT 6 INITIAL CONDITIONS CHECKLIST AND LIST OF 25 CERTIFIED INITIAL CONDITIONS This attachment is referenced by Section 7 j of the Performance Test Summary. I NRC #1 (9) y O

1

ATTACHMENT 8.2 MILLSTONE 2 INITIAL CONDITIONS VERIFICATION CHECKLIST

. l I. CONTROL BOARD WALKDOWN With the Simulator in "run", at each of the following control f boards, an SRO licensed or certified 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.

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 C-06 C-06R RPS Ch D ESAS Ch D C-07 C-07R C-OlX ESAS ActCAB1 C-08 C-08R C-80 ESAS ActCAB2 2-SI-652 Wall Switch RC 100 Hot Shutdown Panel REMOTE FUNCTIONS REVIEW

( ) II.

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 the Initial Condition.

CCR E S F. RMR TCR CHR FWR RPR TPR CVR IAR RXR TUR CWR MSR SGR WDR EDR RCR SIR EGR RHR SWR Rev.: 0 Date: 1/12/89 Page: 8.2-1 of 4 NSEM 4.02

l III. INITIAL' CONDITION STABILITY AND REASONABILITY f 3

l

['N Perform either section A, B or C 1 is) (A) For Equilibrium Xenon, Steady State, Power Levels (30%, 50%,

l i

75%, 100% power, 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 Charging Flow o ASI reasonable o RCS Tave o Letdown Flow and consistent o RCS Pressure o Pressurizer Level with power level o S/G Levels o S/G Feed Flows and CEA position o oT Power o S/G Steam Flows (RPS) o NI Power o Xenon Reactivity Worth (RPS) from Instructor Station o Calorimetric Power is equilibrium worth on PPC CVMWTH o Electric Power o CEA Group 7 position ,

is reasonable for power level

• Perform SP 2601D (when appropriate) to ensure consistency 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

(}N- Fastime X60.

o Using Xenon Fastime X60, and holding thermal power and CEA position constant verify 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 750% power) during a 4 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 key items not present on the instructor station IC description are added. Key items in remarks section of IC are BOL/MOL/EOL, Xenon Trend, load limit pot setting, unusual CEA positions, unusual equipment lineups, etc.

) Rev.: 0 Date: 1/12/89 Page: 8.2-2 of 4 NSEM 4.02

1 J

o (B) For IC's which have the reactor critical, but do not fall into category A above, verify the following parameters are stable and/or reasonable for the first 2 minutes after f'N resetting to the IC and going to run:

3v ]

o RCS Tc o Charging Flow o ASI is reasonable o RCS Tave o Letdown Flow for [ower level o RCS Pressure o Pressurizer Level and CEA position o S/G Levels o S/G Feed Flows .

o 6T Power o S/G Steam Flows o NI Power o Xenon reactivity worth  !

o Calorimetric Power for the instructor station on PPC CVMWTH is reasonable o Electric Power o Group 7 CEA position is reasonable for power level 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 consisten,t 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 i 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 f~ ; equipment lineups, etc.

'%)

o (C) For IC's which the reactor is not critical and may be in various stages of Plant Startup 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 Pressurizer Level o If Heatup or o RCS Tave o S/G Feed Flows cooldown is o RCS Pressure o Xenon reactivity worth in progress, from the instructor # of running o S/G Levels ECP's is o Wide Range Power station is reasonable or CPS reasonable o Charging Flow o If SDC in operation, SDC flow 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.

0( Rev.: 0 Date: 1/12/89 Page: 8.2-3 of 4 NSEM 4.02

o Ensure any items mentioned in the IC description on the

(; .

' instructor station are correct and any key items not present on the instructor station IC description are added. Key items in

(~3 the remarks section of the IC are BOL/MOL/EOL, Xenon trend, time i s ,,/

l after reactor trip, unusual CEA positions, unusual equipment lineups, etc.

IV. IC REQUIREMENTS TO BE SPECIFICALLY VERIFIED verify the following:

Remote Functions CCR03 set to 50 F (to fail B HXTCV open)

CVR03 set to open CVR13 set to closed CWR01 set to 55 F FWR55 set to 100 gpm (at-power IC's)

FWR55 set to 0 gpm (shutdown IC's) ,

IAR17 set to lead IAR15 set to start MSR03 set to auto MSR05 set to auto MSR07 set to auto RPR31 set to full SGR01 set to 51.5 SGR02 set to 38.0 1 SVR04 set to open l

\-)

/

SHR05 set to closed SWR 06 set to closed SWR 07 set to closed .

SWR 12 set to 0%

SWR 24 set to 10% ("B" RBCCW ~1500 gpm)

TPR16 set to close TPR18 set to TBCCW WDR06 set to open WDR02 set to auto PPC Points SCBLDN1 consistent with SGR01 SCBLDN2 consistent with SGR02 t'~T Rev.: 0

( )

Date: 1/12/89 Page: 8.2-4 of 4 NSEM 4.02 ,

1

t O

l INITIAL CONDITION 5  !

N$$

4 i

, u 8 09:12:56,82/21/89 ROL COLD SHU DOWN SEC SYSTEMS EMPTY C 2l 112 292 8 2006 0 99:13:49,02/21/89 BOL PZR S MSX l ON 500 3 172 376 8 2006 0 64:27:53,82/21/89 80L PER 8 48%

READY TC START RCP's 4l 536 2236 0 1286 0 15:47:25,02/17/89 BOL 18-4 (RO-29)

STEAM OUT GP 7 8 71 l

5l 537 2237 5 1284 58 15:23 98,82/17/89 BOL 4X "A* SGrP

() 6 l

333 2237 4 1269 53 15:36:38,82/17/89 HOT TURBINE (7 8 96)

BOL 4X "A"5GrP LL 1.0 TURBINE 1899 RPM 7l F43 2255 26 1254 258 13:42:24,92/17/89 50L 25X (XE INC) NS$T l LL 1.57 7 9 148 4 572 2255 ite 949 2743 12:44:20,82/23/89 BOL 190X Et XE LL 7.30 RC5 2270 PSI l Cl 9 536 2248 1 891 2 55:51:23,92/17/89 MOL 10-4 GP 7 8 182 STEAM Ct8T itl 573 2243 100 514 2716 12:30:35,82/17/89 MOL 100X Et XC I LL 7.30

. t . ., . , 7, ;. , - .. . . . .. . .-

l O

- ~

l l

0  !

INITIAL CONDITIONS

~ ,, ,1 w. .

l .' E'-  ; F F t: 4 -

ft ' I t1E . ~ 0 -. :E":-J 4 2m p1 i 535 2 5,. 43,8 78,88,.. 2,2 u t, g,,gs,gR ,..,. 4 534 2233 2 516 4247 47:51:35,82/21/89 MOL 14 NR5 AFTER.ite% E I(2ll TRIP (XE DEC) q 542 2228 29 768 183 13:38:53,02/17/89 MOL 30*/ (XE INC) E 113ll LL 1.57 NSST q 542 2047 1 579 3497 68:01:46,92/14/81 M0L LNP LOS5'tW R55T 114 NATURAL CIRC.

535 2236 4 789 314 15:27:09,82/17/89 MOL 3% "A".Sarr E 115 NOT TURBINE (7 8 151) Cg MOL 6% "A"EGrP'LL 1.4 536 2218 7 784 238 15:32:34,42/17/89 116 TURSINE 1899 RPM 554 2242 52 614 2233 12:44:15,02/17/89 MOL 50% EG XP...

117ll LL 3.95 GP 7 8 169 l18 l 535 2251 1 339 2 15:54:52,92/17/89 EOL 10-4 GP 7 8 SS g STEAM OUT l

543 2231 27 321 112 13:35:57,02/17/89 E0L 25% (XE INC) E 119 LL 1.48 N55T Cg 9 573 2241 100 46 2718 12:35:27,42/17/89 L  % Et XE

. . . , g g .p.. 7;,

ll i

154 8 0 2000 9 98:14:31,02/21/89lBOL PZR SEX SDC(RO-1) E l22 l Cg g l lT/C 162 VENT 5 5 NUT 3 570 2229 92 529 2652 12:48:52,02/17/89 L . Et XE 4 573 2239 itt 514 2716 14:17:36,02/22/89 L1 X, O l25l554 l222 53 1 55 22..

13:*.856,.2/iv/89 gLa;. E. xE J, I

. \

1 ATT, MENT 7 l

STUDENT FEEDBACK SURVEY RESULTS l

i l

l 1

l I

l This attachment is referenced by Section 6 of the Performance Test Summary.

l unt .2 <io, l .

1 o

l L..

l

s L

noanmaastununes E-

..__._._m wtgetan maappeart?S (JCtauc C0erm

.......em feDRf84agf MS$ SEM CCtPANT k sognegast usucuam gastslavCosaw April 11, 1989 pJ27p9-047 TO: All MP2 Licensees

.. \

'FROM: J. J. Parillo;s.\. "

(Ext. 622) [

SUBJECT:

MP2 Simulator Fidelity Evaluation Results During January, simulator fidelity evaluation questionnaires were distributed to all MP2 licensees. The attachment contains the results of this survey.

All comments with their associated rating have been included.

The disposition of items brought to our actention is stated for your information. If you have any comments or questions about l T the survey, L) please contact me at extension survey results, or622.

disposition of particular items,

'If you. identify any other simulato; fidelity discrepancies during the year you may use the student assessment forms provided during the conduct of training.

JJP/pab c: File 4.2.6.6

/" s.

l f '

'x ) -

1 l

I

> os7o r.rv s-as

)

M'P2 Licensees .

April 11, 1989

.oT2-89-047 Page 2 of 2

}8%

(_)'I 63 surveys distributed to MP2 licensees 25 (39.6%) surveys returned 189 (68.7%) responses among all questions rated as N 66 (24%) responses among all questions rated as 1 9 (3.2%) responses among all questions rated as 2 2 (.7%) responses among all questions rated as 3 ,

9 (3.2%) responses among all questions rated as S E 1 2 3 S Question A1 76% 16% 81-Question A2 72% 20% 4 t- 4%

Question B1 64% 28% 4%

Question C1 48% 52%

Question D1 48% 36% 8% 8%

t Question El 72% 28% .

l Question E2 76% 24%

Question F1 56% 16% 12% 16%

Question F2 84% 12% 4%

Question G1 88% 12%

Question G2 72% 20% 4% 4%

N= No observable difference between simulator and actual plant 1- The difference between simulator and actual plant is observable but has LITTLE OR NO AFFECT on the operators actions or diagnostic ability.

2- The difference between simulator and actual plant may cause confusion or impair the operator's ability to diagnose or take the required actions PROMPTLY.

3= The difference between simulator and at.tual plant causes confusion. The difference may cause an INCORRECT DIAGNOSIS and/or

( cause the operator to take INCORRECT ACTIONS.

S= The difference between simulator and actual plant may not affect the operator's actions or diagnostic ability, but I FEEL STRONGLY that the difference should be corrected.

- _ - _ _ - _ = _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _

,: _ . _ _ . _ _ _ --__ m

A. PANELS, INDICATION AND CONTROLS

/( )

1. Did you observe any differences between the plant and the simulator regarding panels, meters, switches, lights, scales, ranges, locations, etc.?' ,

Rating Comment

1) -VCT Press in Simulator at 30-40 psi', in-plant maintained at-13 21 psi.

Response: We will attempt to maintain VCT pressure at approximately 20 psig at initialization.

1) Large Recorders on RC 14 are difficult to check for trends.

Response: To allow adequate trends to be developed and recorded, the simulator will be "run" as long as feasible prior to commencing a training exercise. This may not always be possible in sessions with limited time between scenarios.

1) . .. . some annunciator windows have red lights in simulator white lights at plant.

Response: This issue is resolved with the simulator now the same as the plant.

2 )- FRV position is different at 1001 between plant and simulator. Plant at 65%.

Response: We are investigating this item, i

2) Operator cannot put radmonitors in alarm defeat at simulator . . .

Response: See response to this item in Question G2 response.

L O. .... ..: ,

1 l

i

! 2) RCP seal recorder C04R.

)

n Response: The new RCP seal recorders will be installed during the simulator outage. The recorders should be ready for Cycle 1 training.

2. Did you observe any differences between the plant and the simulator regarding mimic, back shading, tags, labels, etc.?

Rating- Comment 1). RPS Mimic Ground Indication Lights

' Response: This item has been corrected.

3) Words " withdraw" and " Insert" are not. labeled.at simulator for'CEA joystick.

Response: The labels will be added.

-s Rating Comment

\I 1) H2 Chromatograph labeling.

Response: This item has been corrected.

1) At times caution tags are not up to date. No orange TR/AWO stickers.

Response: We will not attempt to. maintain the simulator current with short term conditions or alignments. Long term alignments such as: RCP Bleedoff controller caution tag, or caution tag on RCP Bleedoff isolation valve will be reflected on the simulator. TR/AWO stickers are too transient in nature to be of value to training.

l B. INFORMATIONAL AIDS l

1. Did you observe any differences between the plant and the ]

simulator regarding the availability of aids and reference l materials such as procedures, forms, prints, drawing, i operator aids, etc.?

.n ..sv. o>

2 i

m_1____.12_______-- _ _ J

O.

4 1

0 O

I

Rating , Comment f

1) There are many bits of info under the Plexiglas. (at the plant) . . ..

Response: This item was investigated and found that the many bits of. info were not useful for training, with 1 exception.

The operator aid for the PPC was added to the l

simulator.

1) Simulator is somewhat' sterile in appearance, but this should have no effect on its use.

Response: We recognize this to be true. This may be attributed to the relatively short time the simulator has existed, the controlled environment for computer protection, .and the limited activities that take place in the simulator. We do not see this as a problem. I N) Drawings are not kept in plastic.

R esponse: Because the P&ID drawings are infrequently used in the simulator for developing tagout boundaries, we see

-O

~

t relatively little value in keeping them in plastic!

1) Black book' binders not the same, no copies of Maint./ Chem./I&C/H.P. Procedures, no telephone number

-list (as actual) .

Response: The black binders containing the ops Department procedures are essentially the same between the plant )

and the simulator. Maintenance, Chemistry, I&C l procedures are not seen as necessary for effective j t ra ining. The HP procedure for Containment entry is  !

available in the simulator. The telephone number list in the control room would only add confusion due to the limi?ations of the phone system in the simulator.

1) . . . procedures are not always located in the same place as the plant . . . i Response: The instructors will attempt to ensure the proper placement of procedures prior to training sessions.

O neem (w.10) 3

._._____-_.--___-._______w

O 1

O'

)

0 l

L O S) . Computer inputs operator aid is not posted at simulator.

Response: We have obtained-a copy of the computer. inputs aid, and p will post this list.

C. AUDYBLE

1. Did you observe any differences between the plant and the simulator regarding the types and level of noise such as annunciators, printers, background, turbine, steam or incidental sounds?

Rating Comment 1)' As a whole the simulator is a more sterile environment as. compared with the unit.

Response: As stated earlier, we recognize this to be a true statement.

1) Control Room Air Conditioning background noise not

-g present.

'lv- J.

Response:. We are aware that this difference between the simulator and control room exists. Earlier attempts at replicating the sound of control room ventilation have proven unsuccessful. We feel the benefits of simulating thisodo not justify the efforts necessary.

1) Lack of background noise / vibration, ambient noise on phones.

Response: Due to physical limitations, we are unable to simulate background noises on the dial phones. We are, however, currently experimenting with ambient noises on the GAITRONICS in-plant phones and background noise on the PA system.

1) Control Room A.C. units are more noisy.

Response: See comment on Control Room Ventilation System background noises above.

nsentw.10) 4

l l

O, i

l l

O i

t O, 1

it:

t s

)[h

. q)

Rating , Comment

1) There are many bits of info under the Plexiglas. (at the plant) . . .

Response: This item was investigated and found that the many bits of info were not useful for training, with 1 exception.

The operator aid for the PPC was added to the simulator, u

1) Simulator is somewhat sterile in appearance, but this l should have no effect on its use.

i Response: We recognize this to be~true. This may be attributed to the relatively short time the simulator has existed, the controlled environment for computer protection, and the limited activities that take place in the simulator. We do not see this as a problem.

N) Drawings are not kept in plastic.

Response: .Because the P&ID drawings are infrequently used in the g simulator for developing tagout boundaries, we see relatively little value in keeping them in' plastic!

1) Black book binders not the same, no copies of

'Maint./ Chem./I&C/H.P. Procedures, no telephone number list (as actual).

Response: The black binders containing the Ops Department 3 procedures are essentially the same between the plant and the simulator. Maintenance, Chemistry, I&C procedures are not seen as necessary for ef fective training. The HP procedure for Containment entry is available in the simulator. The telephone number list in the control room would only add confusion due to the limitations of the phone system in the simulator.

1) . . . procedures are not always located in the same place as the plant . . .

l Response: The instructors will attempt to ensure the proper '

placement of procedures prior to training sessions.

I n ..tw.to 3

I

O l

1 0

O I

l~

l S) . Computer' inputs operator aid is not posted at simulator.

Response: We have obtained a copy of the computer inputs aid, and will post this list.

-C. AUDIBLE

1. Did you observe any differences between the plant and the

~

L simulator regarding the types and level of noise such as

= annunciators, printers, background, turbine, steam or incidental sounds?

Rating Comment  ;

1)' As a whole the simulator is a more sterile environment  !

as compared with the unit.

Response: As stated earlier, we recognize this to be a true statement.

1) Control Room Air Conditioning background noise not present.

O i Pesponse: We are aware that this difference between the simulator and control room exists. Earlier attempts at replicating the sound of control room ventilation have proven unsuccessful. We feel the benefits of simulating this do not justify the efforts necessary.

1) Lack of background noise / vibration, ambient noise on phones.

Response: Due to physical limitations, we are unable to simulate background noises on the dial phones. We are, however, currently experimenting with ambient noises on the GAITRONICS in-plant phones and background noise on the PA system.

1) Control Room A.C. units are more noisy.

Response: See comment on Control Room Ventilation System background noises above.

O neemtw.30) 4

- - - - - - - - - - - - , _ _ ~ _ - _ _ _ _ _ _ _ _ , , _ _ _ , _ _ _ _ _ _ __

O l

l 4

i l

. 1 1

1

)

i I

O,

)

4' 8 E 80 s9 )@,4k/

t

• / IMAGE EVALUATION it ////gf

%[ %ff TEST TARGET (MT-3) 4g

\/o//7/

y,,,, 9 /{q KNN/

l[f8(,

+

< 6" >

/I[>%zz #f'<bh

. g % ,,

o ,,,,,

4<gy<<y w

4& to IMAGE EVALUATION ///j//f <b%

o'/kt*d@#

'\Rgpf N; TEST TARGET (MT-3) 4

</ Th' #

s Vs4+gy $4,%v,,,p

,.o , .

W 1

Fc in llll!21

' = =

C lhb l,l l.8 1.25 l i.6

_.4-4 150mm

• V 44 e 4& A 4>A > */////;6#'e r ' ~

• M? c3 obf7 [ $*

, w

9

. 4r

%;6>9.

• kyy, IMAGE EVALUATION j/j/'

O

, \g//// Yg,% ' Vg ,, TEST TARGET (MT-3) /g ['$j; ;t

,g

%\

$f>

kr jfp l

1.0 t

lFmm

@ El

e lills ii =

I l .25 1.4 11.6 I I

• 150mm >

• 6" > i h'

. ,y , ,,,,/

, , ,u ,4 o  ;

4y

.p h --_ _

j

N) Slight differences in sound of annunciators.

Response: This may be the case. A great deal of effort went into accurately simulating the tone of the annunciators.

Any differences that exist are not seen as impacting the trainees ability to respond to annunciator as required.

1) The level of background noise is different at the simulator than it is ih the plant on day shift.

Response

This is true simulator is aand can environment.

training be attributed to the fact that the Rating Comment

1) Simulator is quieter.

Response: See response to preceding comment.

D. COMMUNICATION

1. Did you observe any differences between the plant and the simulator regarding the amount and type of communication devices available?

Rating Comment

1) Radiopager is unavailable at the simulator.

Response:  !

The use of the aadiopager console is primarily a SSSA function. SSSA's are generally not with the shift during training sessions on the simulator. The limited training benefits do not justify the expense of obtaining a Radiopager Console.

i

1) Communications with instructor station via phones I causes phones to ring unintentionally.

i Response: I This may be attributed to instructors inadvertently leaving Convex, EOF, or NRC buttons selected. We will attempt to minimize unintentional ringing of phones.

O noem(w.10) 5 l

1) No differences that would warrant changing the simulator.

Response: None

1) E-plan phones.

Response: Presently the phone system provides for dedicated lines to NRC and EOF. No changes to the phone system are currently planned.

1) Is the vibrating page available? or do you simulate its use? This could provide faster PEO response.

Response: Because the simulator has no radiopager it has no vibrating page. PEO response time is based on the instructor's judgement of reasonable "real-time" response.

1) Page communication is not the same as plant.

O Response: This is due to the limitations of the dit.1 phone system in the simulator. No changes to the phone system are planned.

2) Plant Page System at simulator not correct. (no dial-a-page and only one button page)

Response: See preceding response concerning plant page capabilities at the simulator.

Rating Comment

1) . . . the communication console is missing along with numerous other phones for incident communications . . .

The paging system is broadcast in the simulator at a  :

higher volume than in the plant.

Response: See preceding responses concerning Radiopager console and E-Plan phones. The volume of the page system will I be compared between the simulator and control room, and l adjusted as necessary. i nsentw.10) g

1) CONVEX, NRC phones, additional computer terminal.

Response: The simulator provides dedicated phone lines to Convex and the NRC. Refer to response to question F1 concerning additional PPC terminal.

S) Phone System - only one line available . . . this can impair prompt action.

Response: There are now two phone lines provided in the instructor booth for improved communications with trainees.

E. ENVIRONMENT

1. Did you observe any differences between the plant and the siniula to r regarding the amount and type of normal and/or Emergency Lighting?

Rating Comment

1) The simulator is better illuminated than the unit.

Response: Lighting checks will be performed. If a difference is found to exist between the control room and simulator, the lighting will be modified.

2) Slightly brighter in the simulator.

Response: Lighting checks will be performed. If a difference is found to exist between the control room and simulator, the lighting will be modified.

1) With an LNP I don't think the simulator room went dark the same as the plant would.

Response: Some lights in addition to the emergency lights do remain illuminated, these are left on to comply with fire code requirements.

l 9 n...iv. ion 7

1) The lighting in the simulator seems to be a little brighter than the Centrol Room.

Response: Lighting checks will be performed. If a difference is found to exist between the control room and simulator, the lighting will be modified.

1) Simulator brighter.

Response: Lighting checks will be performed. If a difference is found to exist between the control room and simulator, the lighting will be modified.

2. Did you observe any difference between the amount, type, and arrangement of the furniture?

Rating Comment

1) Yes, but of no consequence.

Response: None

1) One metal desk to many - The real plant has removed desks from surv. area.

Response: This item will be resolved by placing the desk in a different location in the simulator.

1) The desk by the trend typer does not exist in plant.

The drawing table is not in the same place.

Response: This item will be resolved by placin? the deca in a different location in the simulator.

1) Simulator better.

Response: None n ..tw.:oi 8

F. PLANT COMPUTER h 1. Did you observe any difference between the plant and the simulator regarding the PPC input and output devices (CRT's, keyboards, printers, etc.)?

Rating Comment

1) SS PPC station is in the instructor booth.

Response: The PPC station in the instructor's booth was not intended to be the SS CRT.

1) Simulator lacking SS computer station; I do not know if this means anything to the operators using the machine.

Response: We intend to provide the additional computer station.

2) The 3rd monitor by the SS office is not available.

~It's usually on ICC Summary.

Response: See preceding response.

~

2) Additional terminal needed - plant has one by SS office.

Response: See response above.

Rating Comment

2) There's one missing in the sinalator (the one by the communication console). The letters on the function key pad are wearing / worn away.

Response: See response above. The letters on the function key pads have been replaced.

O n...tw.: ci g

/ S) Simulator has no SS CRT and keyboard.

Response:- An additional CRT and keyboard will be provided.

i S) It.tekes a significant amount of time for.the. computer people in the plant to send the updated displays to1the {

simulator.

Response: It is a significant ef fort, to. update the ' PPC functions.

We. attempt to maintain the PPC as up to date as:

possible. Due to the resources / effort required one to two updates each year is the maximum feasible.

S) There are three CRTs in the Control Room, two.in simulator.

Response:- An additional CRT and keyboard will be provided.

2. Did you observe any differences between the' plant and the simulator regarding the PPC functions, capabilities and responses?

Rating Comment

1) Some of the PPC screens are not yet available on the simulator.

Response: PPC will be updated during the simulator, outage in March, 1989.

O........., ,,

G. SIMbLATOR RESPONSE

1. Did you observe any simulator response which you believe to be incorrect?

Rating Comment ICC level response during accident conditions don't correctly respond to a head void. See attached example of expected vessel liquid volume response for a steam line break.

Response: This is under review.

1) None, other than the fact that no water has to be shot to maintain power.

Response: Because of the transient conditions typically presented at the simulator, maintaining power via dilution is felt to provide minimal training benefits.

Rating Comment

1) S/G level response at low power seems eisier at simulator. The load limit pot doesn't seem to have 9 same response as the plent. Letdown back pressure is not as sensitive at the simulator. Annunciators are not immediately reset or acknowledged as soon as the button is pushed.

Response: We agree that S/G level response is probably easier at the simulator. So far, we have been unable to determine why, and therefore do not know what to modify to correct this condition. Load limit potentiometer response will be compared during the next turbine startup. We agree that letdown back pressure control is less sensitive at the simulator. We feel that the effort required to exactly match the plant character-istics is not justified due to marginal improvements in training. The annunciator acknowledge / reset response is limited by the computer's time delay.

Heatup of H, Recombiner is too fast. "

Response: This item will be checked after the simulator outage.

If found to be true, we will ensure trainees are made aware of this particular artificiality.

O n ...<w. ion yy

.]

i i

l

,.( G2. Was there any procedural section, step or operation-you were i j' unable to perform because of limitations of the simulator? l Rating Comment The student'does not have the capability of placing a ,

module in alarm defeat at RC 14. This reduces the #

in-plant feeling you may have when it is required'te l ask an instructor to perform this function.

Response: We will not simulate ahy hardware items which are located within control Room panels. Alarm defeating  !

radiation monitors will continue to be performed by the instructor.

I

1) Taking RPS Channel out of Comparator Averager Alxim defeat Radmonitor from RC 14.

Response: We will not simulate any hardware items which are located within Control Room panels. Alarm defeating radiation monitors will continue to be_ performed by the j instructor. '

() 1) The alarm disable on RC 14.

Response: See comment above. ^

1) Alarm defeat of RLdmonitors. Recording and resetting  ;

all relay target drops (EOP 2526 Step 3.11). Reset  !

turbine protection relays (EOP 2526 Step 3.149). j Response: See preceding comment. The resources necessary to install all back panel relays or target drops is not seen to be justified by training benefits.

H2 Monitor span zero gas admission Response: This item is being investigated. Span Gas not used in l Control Room-nuestw.10) 7

_____-_m _ _ _ _ _ _ _ _ _ _ _

S) RCS drain down to SDC entry. No' alarm. defeat-capability.-for RC 14 monitors.

Response RCS drain down to below pressurizer is.outside the

~

scope of simulation.

Simulator model. does not consider burnup.

' Response:' Burnup has no training'value during the re'latively short duration training sessions.

O n...<w.to' 13

'l i

O 4 1

j l

ATTACHMENT 8 OPEN DEFICIENCY REPORT LIST o

J This attachment is referenced by section 12 of the Performance Test Summary.

I l

J I

l l

l 4

NRC #1 (111 1

.O

___.______-__.________m __- M

_ C I X I OG LE OS A E' E CE

_ S D H WI F D F R C RP C 0L

_ E F H E E O) UC( C E _ D SO O A RE 2 L LN SR

) L _ R C EO R L .O S R N S T _ L OS T CO OO DO NI E4O W Y I _ E TT T TC DR OT R T O S T _ D CC G D AT A PNO DF D _ O EE N ES S0 N EC E R TO S _ M TF I PI E1 O SI SH U Y S _ EF DA P G R C NF C '. D H R L _ L DE A C OR NE D OI R E SN A / _ E EO R E AL RW PT EK M E _ S NT RL DT HC CO7 S R FSC F M U _ OC A CY L2 EE OAO O U2D _ / RE SE rL C /F - RC EL S E _ TL ' G 1 U E 2 ( KR ST TT _ C UF TR 0M R2 5 N - T CCP S GI A _ C F E EA DI OP 0I8 EI ANC OE NND _ I NR CL R S CM CM8 CS LI R LC IU/

K Y IE _

CET CN _

ANI SI _

ROR IL _

TTO DP _ S S S S S B S S S SI YLR D _

CLP SS _

NI SY _ R R R R W C C C EMY LS _ C C C C F R R R I B C _

I S R _

F ' O _ O -

E R T _ L O O T O O D D A _ L R L L H R L L I H R L L G R L L N N __

I R T U I I I U I I E G _ A U P R R R P R R P I _ P R S A A h S A A O O

(

R O

_ J 6

R P

J P

J D

D P

J P

J E _ / / / / / / / / /

TE _ -

AU _ -

DD _ / / / / / / / / /

_ 7 7 7  ? 9 5 5 v 9

_ 3 3 5 s 3 3 0 5 5 E _ / / / / / / / / /

T _ 1 $ 2 3 1 1 2 3 3 A _ 3 1 2 1 0 0 1 1 1 D _ / / / / / / / / /

_ 2 $ 9 2 3 1 4 2 2

_ 1 0 0 0 0 1 0 0 0

_ 0 5 8 6 4 2 2 4 5 O _ 3 3 3 1 2 4 5 1 1 N _ 1 0 1 0 0 1 0 0 0

_ 0 0 0 0 0 0 0 0 0 C _ - - - - - - - - -

D _ 0 2 2 2 2 2 2 2 2 2 S _ - - - - - - - - -

_ = 7 7 7 9 9 3 3 9 9

_ 8 8 5 8 8 5 6 3 6

)

_ 0

• 0 5 o 4 5 5 4 5

_ = 4 9 d 1 2 4 5 1 1 i O _ N 21 0 1 0 0 1 0 0 0 N _ E 0 0 0 0 0 0 0 0 0

_ P = - - - - - - - - -

- R _ O 2 2 2 2 2 2 2 2 2 D _ ( Y - - - - - - - - -

- a6 9

/ N _

S U

T A

T7 I 8 R

O 7

o 7

S 7

9 5

8 5

8 8

9 8

9 o

e2 d . I _ T I G _ S R G/ I _ P R C C R R A3 R _ = E D D E E

,C O _ =

• C P P C C c  ;

( C c

( C (; C  ?

( C G

)

( C L

. MR I S ET GT E O DA I A oG EL C S C T2 R C

• I EI A U 0F DF EH SV 0I SC SI D ' 3 N ES O 00 OW R M) LM 0*

P O P N O 0 3 S L NO R) ) 1 EF &OI TH3 LR -

R T ND1 OD2 MR PEA OA - R R -

R A E UTC CC2 TCi OE W T A I ( - N( B TTO A UF 3 5 O A S P EQI 0d 8 C W R L/ H ET / E EC UD% D3 2I R 4 I D ME0 S A E 0VC 0VP I E0 CNC - ED - E S F 1 RI ( C R P C R 3 S 9 3 S D R C C A A R R V V O C O O L L L L L L L L I I I I R 8 P R A A A

< A P P P P

J J J J

/ / / /

/ / / /

0 3 8 8 8 3 3 3

~

/ / / /

3 5 3 3

- 1 1 1 1

/ / / /

2 e 0 0 0 0 1 1 7 4 7 3

., 1 7 2 2

- 0 0 1 1 0 0 0 0 n

a 2 2

2 2

9 8 8 8 w 3 7

8 7

8 0

8 1

7 3 3 w

1 0

0 0

0 1

0 1

0 v 2 2 2 2 9 3 S 3

)!,. 8 8 S 8

"(

R C C

-- E D D C P P

) ) ) ) ) ) -

1 to i e- G.  ;

1 I tu > i I (n N W V) .J H ll 1 Z l H z E hi & Z  :

8 O tr; I O e Om u .C I ZM o O Z Q. tu E E% O D-1 OH  % M O M tr uJ 4O Q. so I H et M J M uJ 3W J .J 2 I to u Nu F a v) < rD .J D  % O

.. I tu H UQ > *C in us O us O Q. U

( l 4

OE M

I Q.

v u

z 6-z O

au .J a

rr (L

Cs O c

3 6- Q su 4 EO O W uJ >= n .J E JE a 3 m .J l OO 6- e4 C>- O E 4 _J us O =C Z m >= 0 0A 'o a 3 v E ait a tu >- 2

> M i & MC 1 tu eq g M (D uJ E 3 gH 44 3 to >= 1 ^ to O O e- E rw 6- O .J O in t- OM (D O 1 .J K e- O 9 I uJ B to Zm aC 4 u O to I OO l EuM N Z ME WO z .J e- us u

> 44 1 CE E N  % W 4 M I 6- W Lu J K Q.

E .J l >= u 9 E DC tu @ a, ed O. I at us O CL 4  % 6 2 v or) O ,- > x en J E O tas E p- >- a .J u  ;

E us 8 o cc (r O M M at us 4 z sn s to su u co O l r o i us u v >- O= w ru Hz .u- wz* xm w e- '

3NO I tu fr %v Zu v) O OO >M 3 u9 O m uJ e to M u o e- us uJ O 6 .J w CL tu o Q. I 4 2>

6- t- 9 O>0 WOA CL n. 4. F at to tr EOE N PS *E Z OM at 1 4 ua (L M i eA Q. U >- O I aJ O uJ Z O a 1 I tu 72O 4 u& E v nu et a u t- u tr u E aC u EN u to 1 M 3%

W > H su 1 u W 6- UE 4 4 ZM in M i 2 O tr M .J G

>= t- O Oa 8 'D sD Z W to W M (D W W to M

> .J tr Q l u .J tL m in I w w 2M to > $ 4 4 u I I tr (r tk a su E > Jmi > > v v v v v v v H (D u 1 W H en E 6 _J u . O4 O O .g uJ N 6- 4 .J J >= Q O O O O *E 4 .J .J Z >= E so .J .J 2 1 M M C I E W O .J J 2 *-e 4 & (L M O 3 p E H *~e to O 4 *1 *k cr M CL tr cr 4 W fr Q. M 6 Q. & 3 E m p- (L 3 as et O fY l 1 3 I & a

- O f n(j l "4 7

O O

E W

. a M C s 4 I

I us i  %  %  %  %  %  %  % N  %  %

>= LAJ l 43 I OO I  %  %  %  %  %  %  %  %  %  %

8 I

I fo 4, t (K to en n o o &

1 m in ,o <o .O r.n n <v3 n en Lu 1  %  %  %  % N  %  %  %  %  %

b- 1 FI N sA N e- sA e- eu tw N

• 8 4 e= e- O N N N U Q N fJ O 1  %  %  %  %  %  %  %  %  %  %

0 O o O N o O e- N N N I e- e e- O O e- e- O O O 9 & O to eq to f> N m - N O 4 N ** \ N N N W1 4 e= f an l 'J Z l e" e- e- O O e- O O O O I O O O O O O O O O O U l 1 1 1 8 8 I i 9 8 I O 1 N N N N N N f'J N N N W 9 1 0 $ 1 0 9 l 4 4 0 6 cn in en & to ,o tw & <> tw I 'o en @ (O 00 to rn to en to i N to to PS e- N in 'O e- N 1 to en N N to 4 rd e- N N 9 O I e- e- eq e- O O e- O O O O l Z l O O O O O O O O O O I

i i 11 0 1 4 8 I i 1 I tr i N N N N N N N N N N O i >l i l i

.n 1,

,i, ,

,i ,

.n M n

,n .I

.o

,I ,

,o

.n

,o N

,o

,i,

.n e

.o

.i

.o T

O.

,O 7 0 O

, .o M t M

.u N . O B -

k 49 % M I U U a. u u 2 v v 4 .a wi O O O O >- O O Q. O Of Q. <t . Q. Q. O Q. Q.

O e * * * * * * * * * * * * *

• O A O RI 0O F S I

T A

w L

B U

ET ( O PC S R

OA NL T T OA P RC I U U A) o EI TN J LT C A E TR NM K LO T LA A ET A OR U R RR F D

• DI E N E T E D TO N N EE E TF I UM OE C C G CT O .

A Y A MY YT L YN ET E RI C RI F I S V I R L I M A A E DI DA R OL DR 0 UU I 9 QG D ME AG r'

5 5 S 5 5 V W G V V E C C C C

E D

T I R H S T L H L R G N I R G O A M

S U

P S

R k

d B

I R

W D 4 D K P

/ / /

/ /

/ / / /

/

7 5 7 9 7 6 5 d 8 6 /

/ / /

/ 3 0 3 3 7 0 1 2 1 0

/ / /

/ /

0 2 3 3 0 1 0 0 0 0

6 4 e 3 0 v C 5 2 5 9 1 1 0 w 1 1 0 0 0 0 0 - -

2 2 2 2 2 -

7 B 7 9 7 8 8 8 3 8 6

- 6 1 9 8 6 9 0 7 2 1 0 3 0 1 0 0 0 0 0

- - 2 2 2 2 2 -

- 9 7 3 7 O

7 8 8 8 3 8 H H T T O O

> ') )

) )

i 0- l 8 O 2 to 9 m E Z u O I I W W 4 N M & e=

I 0 0 N u  % t- e- W I i J K N m M e- "'3 4 o 4 O ^ O M M to to 9 O 2 i sn u m o o u eu O ru a z O '

l' 0 > W w a m N o. 9 O M i W E 4 m E 4 to a t4J >=

1 2 in C m > 0 & mm u W to u l M > tu 4 I O CL

• CL 2 Z2

( 6 m - 3 6- u e t- ><< a 4 Os us 4 4 M w >- 4 u to Cr u w "E CL ta.

n J l u > O M W K mo >- an .,J m 9- 8 M O M 3 2 4 W Q. E w ui 4

> H l Z >- en >- O in O w O W W O E E.

in >= I O u O u W W D >= m O I M 3 O W M O O O M 3 us o in i H O E en E 2 4 W e u m A. U Em

> m 9 4 Z Z 4 4 O F CL ua Z E 0. 3 E J l D O CD O _J Z .J 4E N >0 0 C. sn o 4  % 1 >= m uW CL oW tr M u an 2 E uA 9 uW to me so W W = MH W 4H W W. M T 3 4 4 3 >z W D u u 6- u o4 >> c a3 i 3NO I O wO 4 m M 4 . en m 4 2u wI 4 a. O i en W I sn CL M CL & > .J 4 WW J4 OM M Q. E J

>= >- 1 4 O to O m O Q. O u > 6- CL N uO O CL o an .J oM 4 9 in s OW s U cr Wo 4 zW WN wz On o. vo i rzO 1 WO ra w a n. cr .: M M3 ms mM Eo 3 m <

M D% 'l V > M us 4 i u us >- U2 1 {

( zM mM 4 l MOa M J l 6- >- O O CL i m m an en I I m m rn (D m to m MM

> Ja O 8 )

u J CL (n m $ u u )

ZM M> 0 m 3 3 3 3 3 4 m to an O. Q. u )

WL> .J to I us w w w 2 I H E E E Q. O. M M CD U 1 M to CY l taf E O O $

• Ol O O W m 6- 8 J M 6- >- >- .J .J Q O 4 I >= J _J a >- >- sn Z Z Z .J -,J 2 8 I J e-e fr T Z 2 t9 O O M M , j Z M 9 O 4 a 3 c d a M M M E Q W C4 M E 4 Q, M M 3 E M Cr 4 4 i CL M i a en CL M (K n LD 3 3 3 CL O. 2 O tr I J 3 1 M A w O8 e e e e e e o e e o e e I i O V O O o o o O 7 u lO i 9

7 2 7 l Y  !

W 6- us I 8 %  %  %  %  %  %  %  %  %  %  %  %  % f, l I f3 8 hl OO I N N  %  %  %  %  %  % g )

8  :.

l .. :6 J 0 0 is ex> to & <> 50 co en to m m o l m

i to m n m 'O en m on en eu en *c tu 9 N  %  %  %  %  %  %  %  %  %  %  %  % , .

WG (D fs n 4 N N (sJ o en at t C 4 e= l

• • 1 O O e- c- O e- O tu e O e- O O O 1  %  %  %  %  %  %  %  %  %  %  %  %  % j i ro fu c- & N N N c N O N O e-I O " O O O O O O O e- O e- e-I h4 PO N N N 0- P4 N tw 4 N N O -

l O 1 N N e- v-

• e- N C0 (X) #9 rq ru 4 l 7 i O N O v. O O O O O v. O w= O I O O O O O O O O O O O O O '

i u9 0 I 1 8 8 0 4 0 1 1 1 1 8 I O I fJ N N N N N N N N ru fu N N mI l i i l l 1 1 8 0 8 0 0 1 I

I m rs co to & & cn m a3 :n n) n N i to to r0 to (O (O CO CO m en (O to (O 4 an O N in N & en in O N O no C -

I N N v= e- e- e- N CO (% N =$ N N M O I O N O e- U O O O O + o e- O

• 7 i O O O O O O O O O O O O O O I I 6 I I 1 4 I I I i l i M 4 N N N N N N N N ru N N N N f3 0 0 $ $ 1 I i 9 0 8 8 8 I i 8 & N cn en & & rn ro en to at 'n N

• • I % m r0 to m t) 40 m to to En to 'n O to #

Z% 7 1 en M l $.

r

'i&A N

O4 N e4 1 T L1' T I u I k,

4 en er i p. tal M >= fg p.,

< n. O O 4 O u O O n. O y. .

! s L*'**********I

.)

-l' l

m w- l o, 4

<, vi 2 U r. >

< J O O  ! j y H a ^

D o A st

>- . mO '

O I t-iL us ' O 09 t- at 2 a .J O 'm it A H sn 2- t OD '

O tr N 3E-FF w 4 O O- OI m u. N /m UM al O MF us E' 6/5 E2 j OM- OO 7O WW Y nd U m w2 .O ~ >-

3 OO N fL G8 U F

(L U >- uh un O O )

O .) U 4M 4O O&

l

>= Lu . Z as tha wO

.J > - 4 . a >- U ^ ..

su 4 tt <J O ~a m n Oa u3 tu O D 4AJ to O OO 4

> fE tu )j CL >= 23V U4 N 66 Ot 3O 4U N to V i W to

.]

m m .n . m U U U J LL W & S.

O O J J 8

tz K .J O M N m E E G 3 D 4 (L CL 4 m to (L - (L e e e e

& "1 *1 TI (K) CO m P e en e 'O 4 N &n Gi i N v- e N N  %  %

N

@ eb O O O e- e-O n & e- nn I

'O O n n O e-O O O O 8

I e 8 N

N N N  !

[  : e a0 i  !

i 0) to e I

@ fo 00 (O e N 4 eo 00 e n n O y- e. e-O O O O 8 8 8 8 N N' N N f 8 8 8 I e ai co as 10 f0 e EO g U pr the O the U IL U

1 I I N I 8 fr F O N .,2 8 tu E O ^O O M l ) (L O F to I to tu o I tu O 3 3N a, to I > tr W >- I CE CL Z >-

I au a 4 fr *= to 4N su q O 6a.

1 J Z au oD Z v ro >- 3 a. 4 l 1 v 9 i i

3 tr W

3 M

(D m

o J O tu 4

tu

>= >=

uW 0

n.

in to

>- rr 2v WO 4

W I

to Z

O o

Z M

w us E

m I

3 eu

(*

l EO^ e- O >- W _J 2 M CL E M T J O 3 W l tu CL >= u >= u w tu en - W 6- fr 4 .J .J

^ Jl "J E E W wO E Z N 4 4 E M M >- 1 O to W W 3 W O at M u 3 tr Z

> M i CL > u O W W  % >= CE E M4 O O E en >= 8 v Z Z O En H >- En 3 O e- .J Z O 1 > En 4 4 >- Z* 2 , Z an Z e # CD Q. O en g _J >- >- E Z H *C O OO en MC

> tu 4 M W tn 1 2ZZ 4 Lu 3 to E WH tu e- Z E >-

Cr _J B O tu tu m e- M Ov >- >- E tu to N >- 3 4  % 8 ME e- W O v O O 44 (L >e O W CL

! E 10 0 CE u W v en u M 4O 4 Q. u J J Lu u to tr r E D f tu M > CE 2 tr E I tu E MM >O 4m Z P- Z3 O D ev O - t JwO O w M Q. u4 aE >v H M an v m tu a CL E rr w (L w 3 CL b- I MM M t- (D E CD m F >- 3 CL W CL .J m .J m OO Eo >- 0 OZ ue tr J cr W .J l cM 4 OOE 4 uJ 4 Z OE >- M ZO 00 CL 3 4 3E t ZZO I. OuH c& r M 4O ur 4r Eu a. N >- E >- Q. .u M 3%

v > MW I u W 6- UZ l 4 2M nn H 8 W O cr- M .J G HFO O CL 8 to in M m so m *D en to to to m sn MM

> .J m o i U .J a Mm0 ZH m>0 v v v O O E CL >C H v v 3 CL to 1 > .J m $ & 2 T E E E E E M $= >= 9- >

M (D u I W H M tr i

u. .J au

= 00 0 W tu O 4 W .,J W tr >- 8 .J O O O p-O O

.J O O 4 >-

4 0 J M M E E J Z >- .J to M O I Z& M m to CE Cr .J O I M W so to e Z H l W Z Z 3 3 M M c &

W c4 4 cr a 3 2 tal M CL CL E cr M 4 E a 3 cm CL M i a 3 3 en to *f J (L 9 O w

& I nI 4

e rn E

o m

e CL e

Q e

CL

-1 o

O e

Ex O

3 e e e

F V

e 3

Cn e

LL ca 6-V o

J O

e 4

4 W 9  %  %  %  %  %  %  %  %  %  %  %  %  %

t- to 8 4 *J l OO 1  %  %  %  %  %  %  %  %  %  %  %  %  %

1 i

s o o es as en (> <n o en o un tw N g en en 'n P n en m m er. cn m m en WG  %  %  %  %  %  %  %  %  %  %  %  %  %

6 $ 6A 4 s 4 fn m an N & e- O O m g i N O O e- N e- tv e. e- O e- m fu o4  %  %  %  %  %  %  %  %  %  %  %  %  %

i O e- e- & O N O N &n & e- e- s i e- O O O O O e- O O O O e- O h i N N m w & C O O N LA 4 N O I en O U 4

e e- e- s N <> O s O &

2 I e- O O e- e- O e- O O e- e- O O 4 O O O O O O O O O O O O O vi 4 8 8 8 i i 1 8 I 1 4 0 0 O t N N N N N N N N N N N N N g en i 1 1 0 8 8 8 I I I i 3 I I I u0 C> & W (U O' m o to :D (O & N l *n M) to 'D 10 in to 90 'O en 'o 'O *O i

8 O N 'S 4 N O M O O .n N h. N O -s O O e- N e- 4 tv N O 4 O N 4 O I w U O e- e- O e Q O O

e- r- O O 2 8 O O O O O O

, O. O. i O, a i N 8 0 i O. I 4 O. O. I N N N N N N N tv tv N N N O 4 0 4 0 0 8 8

, ,0 , .> n - m & . ,8 ,1 ,8n ,8, 4 0 n c ,. N ,

.. , , ,o . ,o e m ,o ,o . .o ,o <<>

Oo Z .

O r% *t)

W FV M l c$

c% M I tr tr W E fr u E tr

<m .Y I ,,, .o ... .., .., O .I- ,o W CL o OI u u u u u CL O u u O 9 9 9 9 9 9 e 9 9 9 9 9 9 9 9 9

O D O N C T

N I A ,

A T I A C SS C N RE I U EH F N .

LC I N 4 LT D A 0 OI O C RW M 7 T TS 0 N E C N OR D / O CE I 6 I L R 0 T SL R C C UO E N OR V P U I T O A F R N M L A O I E A VC A R M S H 5 S P P P P Y Y Y Y-T T H E H T O I R

G W

F F

A G

I E

H G

I J w D W

D A D D

/ / / /

/ / / /

7 7 5 9 8 3 d 3

/ / / /

6 3 5 3 1 1 2 2

/ / / /

0 1 0 1 1 1 1 0 3 2 4 5 9 1 3 0 1 2 1 0 0 0 0 0 2 2 2 2

, 7 7 8 9 8 8 8 8 5 9 7 5 9 0 3

- 1 0

2 1 0 0 0 0 0 2 2 2 2 O 7 3

7 3

8 8

9 8

H H T T 0 O

o _

_ O R

N O I

N NE A R MB

_ I T OV

_ T C T K E _ C N 6

) _ N U S1 S

Y TL_ U F N .

I _ F O4 S T _ E I P D _ E T TD E S _ T OT CN E Y S _ O MN NA W R L _ M EE U S A / _ E RM F 9 M E _ R N . 9 M U _ EI E6 9 U2D _ d VA T 2 S E _ C OT OP 4 TT _ C MA MI R GI A _ b EO ER M NND _ d RC RT R IU/

K Y IE _

CET CN _

ANI SI _

ROR I L _

TTO DP _ S S 5 S SI YLR D _

CLP SS _

NI SY _ C H D P EMY LS _ C C E Y I B C _

I S R _

F ' O _

E R T _

D D A _ T T N _ H H o N E

P O

C I

I G .

R _

O _

G I

D 4

G I

o w

P S

P U

T G

I R

H

_ D D S =

E _ / / / /

TE _

A U _

DD _ / / / /

_ 0 7 7 E

_ 3 5 5 3 E _ / / / /

T _ 4 4 2 3 A _ 1 1 2 0 D _ / / / /

_ 3 3 1 3

_ 0 0 0 0

_ 6 7 1 7 O _ 2 2 5 9 N _ 0 0 1 0

_ 0 0 0 0 C _ - - - -

D _ 2 2 2 2 S _ - - - -

_ 7 9 7 S

_ 3 8 8 S

_ 6 7 2 0

_ 2 2 4 0 5 O _ 40 0 0 1 N _ 0 0 0 0

_ = - - - -

R _ 2 2 2 2 D _ Y - - - -

_ T9 7 7 3 9 _ I3 8 S 8 O O8 R

N/ N _ O 5 I _ I E2 G _ R G/ I _ Pe R H A3 R _ i E T P0 O _

• C C 0

) ) ) ) ) ) 5 ) > > ) )

I f

O ATTACHMENT 9 SCHEDULE FOR NEXT FOUR YEARS OF TESTING i

This attachment is referenced by Section 13 of the Performance Test Summary. 4 O

NRC #3 l12) y O

i l

( ATTACHMENT 8.2 1

MP2 1

PERFORMANCE TEST I SCHEDULE l

\

l 1

START END .

1 Performance Test (4 Year Cycle): June 1989 June 1993  !

Year One: June 1989 June 1990 Year Two: June 1990 June 1991 Year Three: June 1991 June 1992 Year Four: June 1992 June 1993 A

/l '

I/ m

• i )f1d APPROVED BY: , JW/)N $l{})la]/yJ,

' SOT '

- ~

/ ,

1 j

1 lu l .i j 4!

1.

L l

Rev: 0 Date: 4/1/89 NSEM-4.07 Page: 8.2-1 of 13 .j

--_-___-__---__Q

YEAR ONE l

TEST SCHEDULING WORK AREA.

EO.

i Annual Operability NSEM-4.09 l 28% Steady State Accuracy i

50% Steady State Accuracy '

1

-100% Steady State Accuracy I

i 100% Stability. I Transient #1: Manual-Reactor Trip Transient #2: Simultaneous Trip of All Feedwater Pumps j l

I Transient #3: simultdheous closure of '

All Main Steam Isolation Valves

() Transient #4: Simultaneous Trip of All Reactor Coolant Pumps Transient #5: Trip of the "A" Reactor Coolant Pump 1 Transient #6: Main Turbine Trip From Power Level Not Resulting in Immediate  ;

Reactor Trip 1 Transient #7: Rapid Ramp Rate Decrease in Plant Power Transient #8:- Maximum LOCA with LNP l

Transient #9: Maximum Unisolable MSLB Inside Containment O

Rev.: 0  ;

Date: 4/1/89 NSEM-4.07 Page:- 8.2-2 of 13

YEAR ONE I TES" SCHEDULING WORK AREA 'l Transient #10: Slow RCS Depressurization to Saturated Conditions Using "A" PORV 3 Stuck Full Open (1 HPSI Defeated) f

)

Physical ridelity verification (NSEM-4.12) ]

l Note: SHM requires approximately 2 months advance notice to schedule photographer for photo book update.

Normal Plant Evolutions NSEM-4.10  !

Plant Startup I

Nuclear Startup Turbine Startup and Generator l Synchronization j l

Reactor Trip and Recovery l

(:)

Hot Standby Operation Load Changes Plant Shutdown Surveillance Testing Note: Individual surveillance tests are performed during normal operations. 4 l

System Tests NSEM-4.01 i I

Safety Injection / Containment Spray Test '

Plant Process Computer Test  ;

1

(

Rev.: 0 Date: 4/1/89 j NSEM-4.07 Page: 8.2-3-of 13 j

YEAR ONE  !

TEST SCHEDULING WORK AREA

()

1

' Shutdown cooling System _ Test Control Element Drive System Test Containment / Heating, Ventilation & Air Conditioning Test

)

l

, i Reactor Regulating. Test

)

Major' Malfunction Tests NSEM-4.04 CV01 LD Line Leak in Ctmt (Unisolable)

CV02 LD Line Leak in Aux Bldg CV18 LD Line Leak in Ctmt (Isolable)  !

i j

RD0lXX Dropped CEA

() RD04XX Ejected CEA RD09xX Uncoupled CEA Minor Malfunctions NSEM-4.05 (Listed by System)

CTMT/ Heating / Vent. - CH Electrical Distribution - ED Electrical Generation - EG Condensate /Feedwater - FW Instrument /f'.ation Air - IA Radiation Monitoring - RM

) Waste Disposal - WD Rev.:' O Date: 4/1/89 NSEM-4.07 Page:' 8.2-4 of 13

1

-l YEAR TWO-l TEST SCHEDULING-WORK AREA-Annual' Operability NSEM-4.09 '

28% Steady State Accuracy'

-l 50% Steady State Accuracy 3 100% Steady State Accuracy .h i

il 1004 Stability  !

l i

Transient #1: Manual Reactor Trip. 1 i

i Transient #2: Simultaneous Trip'of All i Feedwater Pumps  !

i i

l Transient #3: ' Simultaneous Closure of I All Main Steam Isolation Valves l 7'

Transient #4: Simultaneous Trip of All l Reactor Coolant Pumps  !

1 i

Transient #5: Trip of-the "A" Reactor  :

Coolant Pump 4 Transient #6: Main Turbine Trip From  !

Powe r - Level Not Resulting in Immediate )'

Reactor Trip Transient #7: Rapid Ramp Rate Decrease in Plant Power l

Transient #8: Maximum LOCA with LNP  !

l

-i i

1

( '

Rev.: 0 I Date: 4/1/89 {

NSEM-4.07 Page: 8.2-5 of 13  ;

l l

YEAR TWO TEST SCHEDULING WORK AREA-Transient #9: Maximum Unisolable MSLB Inside Containment 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.

Instructor Station Test - (NSEM-4.11)

System Tests NSEM-4.01 Reactor Protection / Nuclear Instrumen-tation Test Engineered Safe:y Features Actuation

() System Test chemical & Volume Control System Tect Main Stiam Test Turbine Test Turbine Building closed Cooling Water Test Major Malfunction Tests NSEM-4.04 RCOSA (B) RCS SV Failure RC06A (B) PORV Failure  !

1 RCllA (B,C,D) RCP Locked Rotor

( MS01A (B) MSLB in Ctmt Rev.: 0 NSEM-4.07 Date: 4/1/89 Page: 8.2-6 of 13

YEAR TWO TEST SCHEDULING WORK AREA Minor Malfunctions NSEM-4.05 (Listed by System)

Reactor Coolant System - RC Reactor Protection System - RP Core - CR Shutdown Cooling - RH Turbine Controls - TC I,

TBCCW - TP j

i I

O l

I O

Rev.: 0 Date: 4/1/89 NSEM-4.07 Page: 8.2-7 of 13

YEAR THREE  :

TEST SCHEDULING WORK AREA 1

Annual Operability NSEM-4.09 <

l 28% Steady State Accuracy i l

50% Steady State Accuracy {

1

(

100% Steady State Accuracy  !

100). Stability l

1 Transient #1: Manual Reactor Trip l Transient #2: Simultaneous Trip of All reedwater Pumps l I

I l

Transient #3: Simultaneous closure of '

All Main Steam Isolation Valves Transient #4: Simultaneous Trip of All

'\_,) Reactor Coolant Pumps i Transient #5: Trip of the "A" Reactor Coolant Pump {

Transient #6: Main Turbine Trip From Powe r Level Not Resulting in Immediate  ;

Reactor Trip Transient #7: Rapid Ramp Rate Decrease i in Plant Power Transient #8: Maximum LOCA with LNP Transient #9: Maximum Unisolable MSLB Inside Containment O

Rev.: 0 Date: 4/1/89 NSEM-4.07 Page: 8.2-8 of 13

YEAR THREE TEST SCHEDULING WORK AREA C) . Transient #10: Slow RCS Depressuriz-ation to Saturated Conditions Using  ?

"A" PORV Stuck Full Open (1 HPSI Defeated)

Real Time Test - MP2 (NSEM-4.13) 1 Physical Fidelity verification j (NSEM-4.12)

Note: SHM requires approximately 2 )

1 months advance notice to schedule photographer for photo book update.

System Tests NSEM-4.01 Condensate /Feedwater Test i Electrical Distribution Test Reactor Coolant System / Steam Generator Test Reactor Building Closed Cooling Water Test Electrical Generation Test  ;

l Instrument / Station Air Test i i

Major Malfunction Tests NSEM-4.04 '

FW25A (B) FW Line Break in Ctmt MS02A (B) MSLB Out Ctmt RCO2A (B) Th LOCA RCO3A (B,C,D) Tc LOCA I

'O i

Rev.: 0 ,-

NSEM-4.07-Date: 4/1/89 '

Page: 8.2-9 of 13

k

-I YEAR.THREE

-TEST SCHEDULING WORK AREA Minor Malfunctions NSEM-4.05 (Listed by System)

Main Steam - MS Control' Rods - RD Reactor Regulating System - RX Steam Generators - SG Service Water - SW Turbine - TU

(

j

)

i i

i i

j 4

i i

)

i l

i Rev.: 0 NSEM-4.07 Date:- 4/1/89' .

Page: .8.2-10 of 13- l

YEAR!FOUR TEST SCHEDULING WORK AREA' 1

Annual Operability NSEM-4.09 28% Steady State Accuracy 4 1

l 50% Steady State Accuracy j 100% Steady State Accuracy. ,

100% Stability l

-i Transient #1: Manual Reactor Trip Transient #2: Simultaneous Trip of All i i

Feedwater Pumps Transient #3: Simultaneous Closure of All Main Steam Isolation Valves 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 i Reactor. Trip I Transient #7: Rapid Ramp Rate Decrease in Plant Power 1 Transient #8: Maximum LOCA with LNP Transient #9: Maximum Unisolable MSLB Inside Containment l

i l

.O Rev.: 0 I Date:- 4/1/89 ,

NSEM-4 07 Page: 8. 2-11. of 13 l

YEAR FOUR 1

i TEST SCHEDULING WORK AREA' O Transien't #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.  ;

' System Tests NSEM-4.01 Reactor Core Test Service Water Test Circulating Water Test Radiation Monitors Test

() Waste Disposal Test Major Malfunction Tests NSEM-4.04 MSO3 MSLB in Turb. Bldg MSLB W/LNP Composite RC04 RCS Leak RH01A (B) LPSI Pp Trip (Loss of SDC)

SG02A (B) SGTR I

Minor Malfunctions NSEM-4.04 i (Listed by System)  ;

RBCCW - CC

(

Rev.: 0 Date: 4/1/89 ,

NSEM-4.07 Page: 8.2-12 of 13 j A

YEAR FOUR TEST SCHEDULING WORK AREA CVCS - CV Circulating Water - CW i

Engineered Safeguards - ES 4

Plant Computer - PC Safety Injection - SI O

O Rev.: 0 Date: 4/1/89 NSEM-4.07 Page: 8.2-13 of 13

O ATTACHMENT 10 ANNUAL OPERABILITY TRANSIENT TESTING ABSTRACTS This attachment is referenced by Section 5 of the Performance Test Summary.

O e.c .o,,

1

e fg i

YEARLY OPERABILITY TRANSIENT TESTING ABSTRACTS

%/

The following 10 transients were all run in April, 1988. 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 ANSI /ANS 3.5 (1985) are taken.

Transient #1 - Manual Reactor Trip As required by ANSI /ANS 3.5 (1985) Appendix B, a manual reactor trip was performed from 100% power, steady state (Middle of Life Core conditions), equilibrium Xenon. 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 Generator level to ramp towards 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 criteria were met, no deficiencies were identified.

Transient #2 - S_multaneous Trip of all Feedwater Pumps As required by ANSI /ANS 3.5 (1985) Appendix B, a simultaneous trip of all Feedwater pumps were performed from 100% power, steady state (Middle of Life Core conditions), equilibrium Xenon. No Main feed or

(~T Auxiliary feed were allowed to the Steam Generators. All parameters

(_) listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 4 minutes, which is sufficient time for RCS temperature to  !

stabilize and Pressurizer pressure and level to ramp towards recovery.

Baseline data for comparison is from CEN 128 Case B.4. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.1. All acceptance criteria were met except one. It was found that the Simulator took an excessive time for Steam Generator level to decrease to its trip setpoint, ~ 54 seconds versus 20 to 25 seconds. This was found to be a problem with the Main Feedpumps coasting down to slow. This deficiency has been fixed and retested. Graphical comparisons were not re-run, but will be regraphed at the next yearly Operability Test.

Transient (MSIV's)

1. 3 - Simultaneous closure of All Main Steam Isolation Valves As required by ANSI /ANS 3.5 (1985) Appendix B, a simultaneous closure of all MSIV's was performed from 100% power, steady state (Middle of Core Life conditions), equilibrium Xenon. To be consistent with analytical data to be used for comparison, Atmospheric Steam Dumps were not allowed to open, Pressurizer PORV's were blocked closed and a loss of all feed occurred at the same time. All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.1 were recorded for a period of 1 minute, which is sufficient time for RCS temperature to stabilize due to the Steam Generator Safety Valves and Pressurizer Pressure and 7s Level to bottom out and start recovery. Baseline data for comparison u/

') is from a FSAR Case, Section 14.9, Figures 14.9-1, 2, 3 and 4 and some NRC #1 (13) 2

best estimate analysis. Graphical comparisons were made for each of

' the monitored parameters from Appendix B.2.2.1. Many of the parameters failed to meet acceptance criteria since the MSIV closure at the Simulator did not produce a sufficiently large temperature increase in Tc, Th or Tave and therefore did not produce a sufficiently large increase in Pressurizer Level and pressure. In both the analytical case and simulator case, a Reactor Trip occurred within ~ 7 secor.ds, but f rom dif ferent trips. The lack of sufficient Simulator RCS heatup is documented in Deficiency Report (DR) 88-2-77, and is still open. This DR will be resolved by 5-1-91. This is not a significant training issue since from a students perspective, the reactor is automatically tripped by the S/G low level trip at almost the same time it should have by the High Pressurizer _ Pressure trip.

The post-trip response is similar regardless of the reason for the RPS tril.

Transient #4 - Simultaneous Trip of all Reactor Coolant Pumps (RCP's)

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 Core 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

((,)N 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. Many of the parameters failed to meet acceptance criteria due to 3. problems, 2 of which have now been resolved. These 3 problems are:

(1) The RPS trip on RCP underspeed was taking approximately 5 seconds {

on the simulator versus the actual Reactor Trip time of less than i l second. This has been fixed.

(2) Feedwater Flow was observed to decrease at too slow a rate at the Simulator. This is the same problem discussed in Transient #2 )

j and has been fixed. '

(3) The Simulator RCS does not show sufficient increase in Tc, Th, and Tave and therefore Pressurizer Level and Pressure do not increase enough, in response to the loss of RCS flow. This has not been fixed and is documented in Deficiency Report (DR) 89-2-14. This t w will be resolved by 5-1-91. 1 This open DR is not a significant training issue since from a students perspective, the Reactor is tripped immediately from the RCP underspeed trip and the post-trip response is similar regardless of the lack of elevated RCS temperature and pressures immediately following the RCP trips.

l NRC #1 (13)

s Transient #5 - Trip of any Single Reactor Coolant Pump.(RCP)

\/ 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 Core 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.l. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.2.

All acceptance criteria would have been met except for 1 problem. It was found that the Simulator time for an RPS trip to' occur on RCS Low Flow was excessive. This deficiency has been fixed and retested.

Graphical comparisons were not re-run, but will be regraphed at the next yearly Operability Test.

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 Trip.

that will not cause an immediate Reactor. Trip from a Turbine 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 Generator level to ramp towards recovery. Baseline data for comparison is from best estimate analysis. Graphical comparisons were O made for each of the monitored parameters from Appendix B.2.2.1. All acceptance criteria were met, no deficiencies were identified.

Transient 75% power.

1. 7 - Rapid Ramp Rate Decrease in Plant Power from 100% to 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 Core Life conditions),

equilibrium Xenon.

All parameters listed in ANSI /ANS 3.5 (1985)

Appendix B.2.2.1 were recorded for a period of 7 minutes, which is i sufficient to reduce power to < 75% power by use of boric acid i additionControl Turbine to the RCS and use of the Throttle Pressure Limiter for the valves.

estimate analysis. Graphical Baseline data for comparison is from best comparisons were made for each of the monitored parameters from Appendix B.2.2.1. All acceptance criteria j were met, no deficiencies were identified.

i Transient #8 - Maximum Size Reactor Coolant System Rupture combined '

l with loss of all offsite power.

As required by ANSI /ANS 3.5 (1985) Appendix B, a maximum size Reactot Coolant System Rupture with a complete loss of offsite power was performed from 100% power, steady state (Middle of Core 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 Diesel Generator Buses and battery power were available.

NRC 41 (13) 4 l

1 l

I All parameters listed in ANSI /ANS 3.5 (1985) Appendix B.2.2.3 were .j (lll recorded for a period of.5 minutes, which is sufficient time for Containment Temperature and Pressure to peak. Baseline data for i

l comparison is from FSAR Section 14.15 and 14.16; Figures 14.5-11 and Figures 14.16-6 4 7. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.3. All acceptance criteria were met, no deficiencies were identified.

Transient #9 - Maximum Unisolable Main Steam Line Rupture 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 Core 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 i

ANSI /ANS 3.5 (1985) Appendix B.2.2.3 were recorded.for a period of 7 l minutes, which is sufficient time for Pressurizer Pressure and Level ,

to begin recovery and Containment Temperature and Pressure to have j 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. l All acceptance criteria were met, no deficiencies were identified. d l Transient #10 - Slow Primary System Depressurization to Saturated 1 Condition using Pressurizer Relief Valve 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 O Safety Injection Disabled. This event was initiated from 100% power, steady state (Middle of Core Life conditions), equilibrium Xenon. All parameters, listed in ANSI /ANS 3.5 (1985) 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 flow 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 an NU RETRAN analytical case specific to Millstone 2, CEN 268, Case 5.1.2, and some Test Estimate analysis. Graphical comparisons were made for each of the monitored parameters from Appendix B.2.2.4 (except relief valve flow). One deficiency was identified. After saturation was reached in the RCS, insufficient voiding took place.

Expected increases in Pressurizer Level due to void formation did not occur to any significant extent. DR 89-2-41 has been written to document this problem. This DR will be dispositioned by 5-1-91.

O NRC #1 (13) 5 r

O ATTACHMENT 11 PHYSICAL FIDELITY

SUMMARY

REPORT

\

This attachment is referenced by Section 6 of the Performance Test Summary.

O O

.<1 m.,

1

1 FORM 7.5

.O C/

SIMULATOR PHYSICAL FIDELITY / HUMAN FACTORS REPORT i

UNIT: 2 REVISION: 0' The MP2 simulator control room has been compared to the MP2 real control room and has'been found to be the same except as described on the attached pages.

1 1

)

O [ N Approved By: 87/ Date:

SOT //~'

Concurrence: Date: [ /!$h Mof // )

Concurrence: 8' Date:

Uyt Superi'ntendent SCCC Mtg. No. 2 f"OO I .

i I

Rev: 0 fO Date:

Page:

4/13/89 7.5-1 of 1 'i

)

I

=* _ _ . _ _ - _ _ _ _ _ _ - _ - - _ _

Pags l_ of 3_

FORM 7.1 EXCEPTIONS - CONTROL ROOM LAYOUT UNIT: 2

)

I

1. The 4 PPC printers in the reference plant control room are brown; in the simulator all 4 are gray. The color of the printers has no impact on training. The printers are otherwise functionally identical.

2. The emergency plan communications consoles (radio pager), Tech j Support Center (TSC) Phone, Waterford. Police Phone, Operational '

Support Center (OSC) Phone, Berlin 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 no significant training impact on whether they communicate thru thli~~

phone console at the Simulator Operator Desk versus the real EOF and NRC phones in the reference plant control room. The simulator is not used to provide training on the radiopager.

This is not a licensed operator task. Large scale exercises on the emergency plan are not performed on the simulator, therefore the lack of identical radiopager, TSC, OSC, Berlin, Waterford

, Police, NRC and EOF phones do not present a problem.

I

3. The Shift Supervisor's (SS) office is not simulated. There is no 1

i training value and therefore no training impact to having an SS office in the simulator room.

4. The key locker on the simulator is not identical to the reference plant. The reference plant key locker is located in the SS office and contains a very large number of keys for use in the entire reference plant. The simulator key locker is located on a wall and has very few keys in it. Key numbers are not the same as reference plant. The reference plant key locker has a very good index in the locker and no training value and therefore no training impact is seen having a student locate a key in the -

locker on t.1e simulator.

Completed by: 2 an Date: [ 2[ [j> h 7 '

Reviewed by: As d Date: //- 2 bd'9 9 f 9**

Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.1-1 of 1

Page 2 of 3 FORM 7.1 EXCEPTIONS - CONTROL ROOM LAYOUT O

UNIT: 2

5. 'The reference plant has a Plant Process Computer (PPC) terminal just outside the SS office, the simulator does not. There is no training value in the missing PPC station since two other identical PPC stations are present at the simulator operators desk. There is a minor loss of convenience in having 2 PPC stations available versus 3 PPC stations. This has not caused any noticable degradation of operator performance during EOP/AOP operations on the simulator. There is no significant training impact.

6. The P&ID Table, the P&ID cabinet and the Operating Procedures File Cabinets are in slightly different locations on the simulator versus the reference plant. All P&ID's present in the Reference Plant Control Room are present in the simulator. All operating procedures present in the reference plant control room are present in the simulator. The difference in location of these P&ID's and procedures has no training impact.

7. The simulator has a desk near panel C-101, the reference plant does not. The reference plant used to have this desk, but moved it out. We have found it convenient to keep the desk on the

("} simulator and this has no training impact.

ud

8. The simulator and reference plant have chairs with different colors. This has no training impocL. -

9. The operators desk in the reference plant has a plexiglass cover on it with various written material underneath it, the simulator does not have the plexiglass cover or the written material underneath it. A review of the material underneath it found only 1 document that had any training or convenience value to it and it will be placed at the Simulator Operators Desk. Therefore, there is no training impact on not having the plexiglass cover or written material beneata it.

Completed by: -

M , Date: 8 b r / 1 l Reviewed by:  % Date: S?PM

/ 7 o Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.1-1 of 1

Page 3 of 3 FORM 7.1 EXCEPTIONS - CONTROL ROOM LAYOUT UNIT: 2

10. The reference plant has Maintenance, Chemistry, I&C and HP procedures located at the Operators Desk, the simulator does not have these procedures. Lack of I&C, Maintenance, Chemistry and HP procedures on the simulator have been found to have no impact on simulator training. The I&C, Maintenance, Chemistry and HP procedures are not used to operate the plant in normal, abnormal 1

or emergency conditions. One HP procedure (Containment Entry) is kept on the simulator since this is an exception. Therefore there is no training impact.

11. Control Panel C-26 is not simulated. The Fire Protection Panel provides alarm indication only and all subsequent actions are local in the plant, therefore there is no control room training that could be provided. There is no training impact.

12. The Fire Shutdown Panel (C-10), and Bottle-up panels (C70 A/B) ]

used by Appendix R AOP's are not simulated. C-10 and C70 A/B are '

located out in the plant not in the control room. The simulator is not used as a training environment for these Appendix R procedures. There is no significant training impact.

()

.1 a

d Completed by: 6's Date: l/ # # I

'/

jf ' I/  !

Reviewed by: ur Date: P-tr-h

}ll ( 90T O Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.1-1 of 1 i

l Page 1.of 2 FORM 7.2 EXCEPTIONS - PANEL LAYOUT O

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 significant training impact.

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 training impact. j

3. Recorders FR 150, 160, 170 and 180 on C-05 Rear.in the' reference  !'

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 training impact.  !

4. TI 6897 on C-01 Rear in' reference plant not simulated. Lack of {

Reactor Vessel Support Log temperatures in Containment has no~~

significant impact on training.

LA 5. RC05E Security' Door Panel Controls in reference plant, not V simulated. Since security doors not simulated, these controls l would have no use, no training impact.

l

6. On Panel C-04, a small Boron concentration status board exists in '

the reference plant, but not on the simulator. This board is  !

updated by the operator with a grease pencil. There is no-~

training impact to not having it.  !

7. The reference plant has an oscillograph next to RC-14, the simulator does not. The operators do not use the oscillograph for any control or monitoring function, therefore there is no training impact.  !

Completed by:

/ Cmc '

Date: '/ O f N p'

// (  !

Reviewed by: y ) n Date: %2['f9

/  !

l

() Rev: 0 '

Date: 4/13/89 NSEM-4.12 Page: 7.2-1 of 1

J Page 2 of 2 FORM 7.2 EXCEPTIONS - PANEL LAYOUT UNIT: 2

8. On C-05R, the reference plant has an "EDAN Flow Monitor" which consists of a switch and fuse, the simulator does not. Operators are not trained on this switch. There is no training impact.

9. On C-05R, the reference plant has a small panel above and related The switches on this panel are not to the Water Quality Monitor manipulated by the operators. There is no' training impact. l l

10. Above the RPS, the reference' plant has a' temperature gauge for control room temperature, the simulator does not. This gauge is used for shift logs, but since there will never be an intentional change to simulator control room temperature, there is no training impact.

l O

i Completed by: At

/

Date: t/ C'f fi i Reviewed by: '/ r Jo Date: 4/-lPfj

1. ASOT

//

r O Rev: 0 Date: 4/13/89 l NSEM-4.12 Page: 7.2-1 of 1  !

I

~~

7 Page 1 of 3 FORM 7.3 EXCEPTIONS - COMPONENTS UNIT: 2-

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 rRC210Y UR5265 UR243 i LR5282 TRlli >

C-06/7 TR7030 TR121  !'

UR4501 JR0ll JR009 There is no training impact. -

2. Recorders RJR 9129, 9373 and AJR 7837 on.RC-14 are different in reference plant versus simulator. . Difference is different'model 1 of Leeds and Northrup recorder. The plant recorders have print j wheels, the simulator recorders have thermal paper. There is no O trainine imo ct. - ,

i

3. The reference plant has various annunciators on all control panels labeled " Computer Test Point". This is for testing which is not necessary on the simulator, therefore the simulator has blank windows versus windows that say " Computer Test Point".

There is no training impact.

4. On reference plant panel C-02, controllers PIC 201, TIC 223 and PIC 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 impact.

Completed by: C eft Date: .2 d#k I

f" / /

o

/

Reviewed by: M d> Date: F-?f'f3 ffj! l (,h*O ,

i Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.3-1 of 1 i

~

Page 2 of 3 FORM 7.3 EXCEPTIONS - COMPONENTS iO l UNIT: 2

5. 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 thumbwheel type on the simulator versus dials in the reference plant. They are functionally identical and therefore there is no training impact.

6. On Control Panel C-05, the "B" Condensate Pump Ammeter is  ;

different between the simulator and reference plant. The i difference is slight, and since there is no opportunity for an operator to misread the scale there is no training Ampact.

i

7. The reference plant control panel C-21, " Hot Shutdown Panel", l located in West 480 Volt Switchgear room, has several meters  !

which are " SIGMA" meters, where as the Simulator C-21 panel, j 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, PIl03, PIl03-1, LI 110X and LI 110Y. There I is no training impact since the information obtained from the  !

meters is identical. l

8. There are some holes in the reference plant C-21 panel but not in l the simulator C-21 panel. This has no training impact. i

9. On reference plant control panel C-06/7, the EHC Insert has the "G-E" nameplate, the simulator does not. This has no training impact.

10. 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 impact.

Completed by: 4M d/ Date: I/> [

/

p n j) l G/

Reviewed by: .LfjA-f Date: Y-24 $M

(}//(6***

O Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.3-1 of 1

Page 3 of 3 FORM 7.3 EXCEPTIONS - COMPONENTS O

UNIT: 2

11. 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 style meter.

Affected meters are:

C-03: PI-103,-PI-103-1, TI-112CA, TI-112CC, TI-122CA, TI-122CC, TI-112HA, 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 impact since the information obtained from the meters is identical.

12. The ref erence plant has calibration stickers on various meters, controllers, etc.; the simulator does not. Calibration stickers have no impact to training.

13. The Main Generator Megawatt /MegaVAR recorder on C-08R is of-a slightly different style as in the reference plant. Ranges are the same. There is no impact to training.

O 14. The Mdin Generator Voltage Recorder on C-08R is of a slightly different style as in the reference plant. Ranges are the same.

There is no impact to training.

15. 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 impact to training.

Comp 1eted by: m er Date: h3 /fk

/ , (

Reviewed by: fv;v Date: I'- 2 I U

/ As

//

0 pT

() '

Rev: 0 Date: 4/13/89 NSEM-4.12 .Page: 7.3-1 of 1

Page 1, of 1 FORM 7.4 EXCEPTIONS - AMBIENT ENVIRONMENT UNIT: 2

1. In the referenet plant there is a steady background noise of-Control Room Air Conditioning (CRAC). Also when changing CRAC l

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 impact.

2. In the reference plant, a page may be performed by dialing at the operator's desk phone, in the simulator a pushbutton is used for paging at the operator's desk phone. This has no training impact.

3. In the simulator, the maintenance jack phones have background noise available, but the. dial phones at the operator's desk do not. If the operator is conversing with someone in the plant, the lack of background noise causes no training impact.

4. There is no radiopager, NRC, Waterford Police, Operational Support Center (OSC), Berlin, EOF and TSC phones on the simulator. This is discussed on Form 7.1, item #2. This causes no significant training impact.

O completed by: o#AG Date: I 6 Reviewed by: /M 'I; , Date: N8Nj

- /A OT /

O-Rev: 0 Date: 4/13/89 NSEM-4.12 Page: 7.4-1 of 1

^\

'd ATTACHMENT 12 SAMPLE MALFUNCTION TEST PROCEDURES l

This attachment is referenced by Section 4 of the Performance Test Summary.

O O

.- ., os, 1

() STEP G PROCEDURE / RESULT PANEL TAG S ACCEPT /DR 1.0 CV01: Unisolable letdown line rupture in CTMT between the RCS and 2-CH-515.

Range: 0-100%, 100% = 200 gpa at 2250 psid.

Critical Record Parameters Method Duration Interval Pzt Level RP 8 min. 15 sec.

LI-110X Pzr Press. RP 8 min. 15. see PI-102A-B LD Flow RP 8 min. 15 sec.

FI-202 LD Temp. TT 8 min. 15 sec.

T221 Chg. Temp RP 8 min. 15 sec.

TI-229 LD Backpres- RP 8 min. 15 sec. i sure PIC-201-C PPC points to be verified: None 1.1 Reset simulator to IC-24 1 1.2 Insert malfunction Cv01 at 100%

severity to actuate at 30 seconds exercise time.

1.3 Activate the record program under case name "MM" for an 8 minute duration. Select a print interval of 15 seconds.

1.4 Select trend group 861 for typer  !

trending.

1.5 Simultaneously activate the typer trend and place the simulator in "Run".

)

Rev.: 0 Date: 9/9/88 Page: 8.2-4 of 90 NSEM-4.04

STEP $ PROCEDURE / RESULT PANEL TAG t ACCEPT /DR 1.6 when Cv01 actuates, verify the following results:

1.6.1 Pressurizer level decreases CO3 LIC-110X _

rapidly and the first, then second back up charging pumps start.

1.6.2 Letdown flow rapidly decreases CO2 FI-212 d to less than 20 gpm in less than 1 minste.

1.6.3 Letdown temperature rapidly drops CO2 TI-221 .

to approx. VCT temperature. TI-225 1.6.4 Charging temperature rapidly CO2 TI-229 -

drops to approx. VCT temperature. TI-225 ,

1.6.5 Letdown backpressure initially CO2 FIC-201 drops, then the backpressure controller responds to restore normal backpressure.

1.6.6 When charging is at maximum and CO2 FI-202 letdown at minimum, pressurizer CO2 FI-212 Q 1evel continues to drop approx.

1% per minute.

CO3 LR-110 4

1.6.7 Pressurizer pressure slowly CO3 PR-100 decreases to less than 2200 psia, the backup heaters energize to  !

cause a slow increase in pressur-  !

izer pressure.

1.6.8 With maximum charging and minimum CO2 FI-202 letdown, VCT level decreases FI-212 approx. 3% per minute. LI-226 1.6.9 RCS leak rate program shows uniden- PPC N/A tified RCS leakage increasing, approaching 200 gpm as data is accumulated.

i 1.6.10 Containment conditions slowly l increase to indicate a small RCS leak:

o CTMT Sump Level CO6 LI-9155 o CTMT Temperatures C01 TI-8096 o CTMT Dewpoint Col MI-8064 o CTMT Narrow Range Pressure C01 FI-8117,.

Rev.: 0 Date: 9/9/88 Page: 8.2-5 of 90 NSEM-4.04

1 l

I i

O STor e raocrouac acsu'r >^ue' r^o e ^cccer oa o CTMT Particulate and Gaseous RC14A RIT-8123A Radmonitors respond af ter RC14C RIT-81238 a delay for transport time. RC148 RIT-9262A RC14D RIT-8262B  ;

1.6.11 After the record program times out, secure charging and letdown by stopping all charging pumps 4 and shutting 2-CH-515.  :

1.6.12 The rJte of pressurizer level CO3 LR-110 $

decrease increases to approx.

3% per minute.

1.6.13 The following alarms, key to b ,

event diagnosis, will occur as 1 their setpoints are exceeded: I o " Letdown riow Lo" CO2/3 C-15 o "Pzr Pressure Selected CO2/3 D-37 Channel Deviation Hi/Lo" i o "Pzr Ch 'X' Level Hi/Lo" CO2/3 A-38 l o "Pzr Ch 'Y' Level Hi/Lo" CO2/3 A-39 i m

1.6.14 Remove malfunction CV01. '

O verify that pzt level stops decreasing (except 4 gpm RCP CO3 LR-110 _

bleedoff). l N9te: Increasing RCS pressure will af fect pzr level decrease 1 (up to 12 gpm is acceptable).

1.7 Stop the typer trend program i and reset the simulator to IC-24.  ;

1.8 Collect the trend typer and record program printouts and mark for malfunction identity and severity.

1.9 Insert CV01 at 50% severity.

Parameter recording is not ,

l required.

1.10 verify similar results as 100%  !

severity but of a lesser mag-  ;

nitude, with the following exception:

Pressurizer level is maintained CO3 LR-110 with the backup charging pumps.

1.11 Reset to IC-24 and insert Cv01 '

(]) at 10% severity, Rev.: 0 Date: 9/9/88 Page 8.2-6 of 90 NSeM-4.04

\

O sre, . r Octou raesotr emo xxo . accreuo. 1 1.12 verify similar results of a lesser <

magnitude, with the.following i

exception:

Pressurizer level is maintained CO3 LR-110 by cycling of the first backup charging pump. ,

l 1.13 End of cv01 test.

o 1

l l

1 l

I O l!

l l

i l

l l

l O -

Rev.: 0 i Date: 9/9/88  :

Page: 8.2-7 of 90 NSEM-4.04

q rWOI Loss of Condenser

(/ Vacuum variable STEP # PROCEDURE / RESULT PANEL TAG 4 ACCEPT /DR 1.0 Reset to IC-24. PCM 2.0 Insert malfunction rWOI with PCM rW01 a 10% severity.

Condenser vacuum begins to decrease slowly.

COS UR-5265 ) /fI

///'

Generator load slowly C07 WM-15G-20 decreases. COBR 15G-20/W-VAR REC j/

f f Hotwell and condensate T5128 temps begin to increase.

PPC T5129 k[f ~

PPC PPC T5186 j/ ( /

r 3.0 Place 2nd set of air ejectors PCM TWR01 ,

in service. rWR02 jr //*

Verify at this time that O' at low severities of this malfunction vacuum may be improved by operating an additional set of air ejectors.

4.0 Increase malfunction severity PCM rWO1 to 50%.

vacuum decreases with all COS UR-5265 air removal equipment .

operating. /Y/"

/

5.0 Remove malfunction rW01 and PCM rWO1 allow vacuum to restore to normal.

Vacuum restores to normal 1 as a function of load and / rr

_, J j

operating air removal /

equipment. l 6.0 Reset to ICIO and reinsert PCM TWO1 FWOI at 100% severity. .

() Rev.: 0 Date: 1/31/89 Page: 8.2-1 of 2 i NSEM-4.05 i

STEP # PROCEDURE / RESULT PANEL TAG $ ACCEPT /DR 100% severity will cause vacuum to decrease to the f ~( 7 turbine trip setpoint /

within 5 minutes.

Condenser vacuum rapidly C06/7 B37 decreases to cause a /l/

turbine trip at 22.5" /

within 5 minutes.

Low vacuum alarm at 25" hg. CO6/7 A37 Turbine bypass valve op-eration inhibited below 15" hg.

SGrP turbines trip at 10" hg.

7.0 End of TWO1 test. / (/

O]c /

O Performed by: gd/

f ~( -

'- Date: / 3/

g g 9

O ev. 0 Date: 1/31/89 Page: 8.2-2 of 2

N ATTACHMENT 13 LIST OF CERTIFIED REMOTE FUNCTIONS This attachment is referenced by Section 2 of the Performance Test Summary..

O l

O noe et (toi y s

l J

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST 1

l

{) UNIT 2 PAGE 1 OF 9 CCR01 RBCCW SURGE TK FILL BYPASS 2-MB-53 CCR02 TIC-6308 HX-18A TO TV-6308,6380A(DEGF) i CCR03 TIC-6307 HX-18B TO TV-6307,6307A(DEGF)

CCR04 TIC-6306 HX-18C TO TV-6306,6306A(DEGF)

CCR05 RECCW SURGE TK DRAINS 2-RB-54A/54B CCR06 ISOL VLV 2-RB-3A DOWNSTREAM P-11A(%)

CCR07 ]

ISOL VLV 2-RB-3A DOWNSTREAM P-11B(%)

CCR08 ISOL VLV 2-RB-3C DOWNSTREAM P-11C(%)

CCR10 THROTTLE SFP CLG FLOW VLV 2-RB-8B(%)

CCR11 THROTTLE SFP CLG FLOW VLV 2-RB-8A( %)

CCR12 SFP HEAT LOAD (E6 BTU /HR)

CCR13 SFP PUMP A START l CCR14 SFP PUMP B START l CCR15 SFP HX-20A VLV 2-RW-6A  :

CCR16 SFP HX-20B VLV 2-RW-6B CCR18 2-RB-66 SURGE TANK MAKEUP ISOLATION VALVE '

CCR31 HPSI PUMP B HDR SEAL COOLING SUPPLY i CCR32 R-RB-14 A SD HX A OUTLET ( % ) l CCR33 R-RB-14B SD HX B OUTLET (%)

CCR34 CC PMP B BREAKER STATUS CCR35 RBCCW PMP A RAD &RECIRC ISOL 2RB43&I07A CCR36 RBCCW PMP B RAD &RECIRC ISOL 2RB41&l07B

() CCR37 CHR01 RBCCW PMP C RAD &RECIRC ISOL 2RB39&l07C OUTSIDE AIR TEMPERATURE VARIATION (DEGF) 4 CHR02 H2 PURGE AIR SOURCE 2-IA-27.2 CHR04 CNTMT PURGE VALVES LOCK OUT( AC-4,5,6&7 )

CHn05 RESET CHILLERS 169A/B,196A/B,170 CVR01 CHG PMPS OUTLET NTO EPSI 2-CH-440,340 CVR02 LETDOWN FLOWPATH THRU DEBORATION IX CVR03 BORIC ACID PUMP DISCHARGE VALVES ( 2-CH- 152/153 )

CVR05 B.A. FROM BATCH TK TO TK 8A 2-CH-124 CVR06 B.A. FROM BATCH TK TO TK 8B 2-CH-135 CVR07 VCT LETDOWN ISL 2-CH-397 CVR08 CHARGING PUMP, P-18A, RACKOUT CVR09 CHARGING PUMP, P-18B, RACKOUT CVR10 CHARGING PUMP, P-18C, RACKOUT CVR11 CHARGING PUMP, P-18B, KIRK KEY CVR12 CHARG PUMP P-18A DIS HDR ISOL (2-CH-338 )

CVR13 VCT TO WASTEGAS HEADER (2-CH-102)

CVR14 BORIC ACID MAKEUP ISOLATION (2-CH-172)

CVR15 2-CH-201P ISOLATION VALVE 2-CH-350 CVR16 2-CH-201Q ISOLATION VALVE 2-CH-349 CVR17 LPSI HDR TO CVCS PURIT LINE 2-CH-603 O MISCSIM#6 (3) NSEM-4.03 REV.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

i FIGURE 7.1 l CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 2 OF 9 l

(~) )

l l

CVR18 PURIT LINE TO LPSI SUCTION 2-CH-024 CVR20 2-CH-110P ISOLATION VALVE 2-CH-3 4 2 i CVR21 2-CH1100 ISOLATION VALVE 2-CH-344 l CVR22 PMW TRANSFER PUMP P-22B CVR23 PMW TRANSFER PUMP P-22A CVR24 2-CH-210Y BORIC ACID MAKEUP VLV LEAKAGE CVR25 B.A. TANKS SUCTION CROSS-TIE 2-CH-144 CVR26 B.A. TANK 8A TO BA PUMP SUCTION 2-CH-131 CVR27 B.A. TANK BB TO BA PUMP SUCTION 2-CH-142 CVR28 DEAERATOR WATER TRANSFER PUMP l CVR29 BORIC ACID TANK 8A CONC CHANGE (% WT)  !

CVR30 BORIC ACID TANK BB CONC CHANGE (% WT) i CVR31 N2 TO VCT 2-CH-109 )

CVR32 H2 TO VCT 2-CH-107 l CVR34 CHARGING PUMP A SUCTION VALVE 2-CH-316 CVR35 CHARGING PUMP B SUCTION VALVE 2-CH-319  ;

CVR36 CHARGING PUMP C SUCTION VALVE 2-CH-322  :

CVR37 PMW STOP VALVE 2-CH-195 i CVR38 BA GRAVITY FEED ISOLATION 2-CH-130 (TANK A)

CVR39 BA GRAVITY FEED ISOLATION 2-CH-140 (TANK B) i CWR01 SEA WATER TEMPERATURE (DEG F)

CWR02 WATER BOX A PRIME & VENT 2-VP-1A,14A i CWR03 WATER BOX B PRIME & VENT 2-VP-1B,14B p)

(_ CWR04 WATER BOX C PRIME & VENT 2-VP-1C,14C CWR05 WATER BOX D PRIME & VENT 2-VP-1D,14D CWR06 SCREENWASH PUMP P-8A j CWR07 SCREENWASH PUMP P-8B EDR01 DISCONNECT SWITCH (15G-22S1-4)

EDR02 BKR A505 IN TEST & INTERLOCKS DEFEATED EDR03 NSST LOCKOUT DEVICE EDR04 RSST LOCKOUT DEVICE EDR05 STATIC TRANSFER SWITCH VS1 EDR06 STATIC TRANSFER SWITCH VS2 EDR07 STATIC TRANSFER SWITCH VS3 EDR08 STATIC TRANSFER SWITCH VS4 EDR09 IAC 1(VR11) THROW OVER SWITCH (RS1)

EDR10 IAC 2(VR21) THROW OVER SWITCH (RS2)

MDR11 DISCHARGE' RATE FOR BATT DB1(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 BATT CHGA 201A AC/DC BKRS EDR17 BATT CHGR 201B AC/DC BKR$

EDR18 BATT CHGR 201C AC/DC BKRS EDR19 D.C. BREAKER (D0103) CONTROL EDR20 D.C. BREAKER (D0104) CONTROL

[D

\/ MISCSIMI6 ( 3 ) NSEM-4.03 REV.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 3 OF 9 l

EDR?1 D.C. BREAKER (D0204) CONTROL 1 EDR22 D.C. BREAKER (D0203) CONTROL EDR23 345 KV SWITCH BREAKER 3 CONTROL EDR24 GRID VOLTAGE VARIANCE (KV)

EDR25 GRID FREQUENCY VARIANCE (HZ)  ;

EDR26 MOTOR OVERCURRENT RELAY RESET j EDR27 6.9 KV NSST(H101) BREAKER RI/RO j EDR28 6.9 KV NSST(H286) BREAKER RI/RO l EDR29 6.9 KV RSST(H103) BREAKER RI/RO f EDR30 6.9 KV RSST(H204) BREAKER RI/RO {

EDR31 ENERGIZE VRll FROM (B32) MCC-2C /

EDR32 ENERGIZE VR21 FROM (B41A) MCC-1DA EGR01 MOTOR OPERATED DISCONN SWITCH (15G-2X1-4)

EGR02 EMERG SEAL OIL PUMP (RI/RO)

EGR05 MAIN GEN BREAKER ST OPERATION SELECTED EGR06 MAIN GEN BREAKER 9T OPERATION SELECTED EGR07 DG 12U MECHANICAL RESET SWITCH l EGR08 DG 13U MECHANICAL RESET SWITCH EGR09 EMERG SO PUMP TEST (H-33,H-34)

EGR10 MAIN GEN BKR 8T/9T GETAC CONTROL 1 EGR11 MAIN GEN LOCKOUT (86GCD1,86GCD2 ) l EGR12 DIESEL GEN A STARTING AIR ISOLATION l EGR13 DIESEL GEN B STARTING AIR ISOLATION Oi- EGR14 DIESEL GEN A MANUAL TRIP EGR15 DIESEL GEN B MANUAL TRIP ESRO1 8A FUSE STATUS ON ACT CAB 5 )

ESR02 8A FUSE STATUS ON ACT CAB 6  !

FWR01 SJAEA STM AND COND INLET AR2A & MS59A FWR02 SJAEB STM AND COND INLET AR2B & MS59B FWR03 COND PUMP P2A DISCH VLV 2-CN-4A(%)

FWR04 COND PUMP P2B DISCH VLV 2-CN-4B(%)

FWR05 COND PUMP P2C DISCH VLV 2-CN-4C(%)

FWR06 COND PUMP T CW DISCH 2-CN-503(%)

FWR07 LP HEATERS BYPASS VLV 2-CN-15(%)

FWR08 DRAINS COOLER "A" INLET 2-CN-11A(%)

FWR09 DRAINS COOLER "B" INLET 2-CN-11B(%)

FWR10 MFWP'S BYP VLV 2-CN-51(%)

FWRll HP HTR 1A, 1B BYP VLV 2-FW-6(%)

FWR12 HP HTR 1A IN& OUTLET ISOL 2-FW-2A,3A(%)

FWR13 HP HTR 1B IN& OUTLET ISOL 2-FW-2B,3B(%)

FWR14 MAKEUP TO TDAFP P4 2-CN-30,2-FIRE-94C i

FWR15 MAKEUP TO AFW P9A 2-CN-29A,2-FIRE-94A FWR16 MAKEUP TO AFW P9B 2-CN-298,2-FIRE-94B FWR17 #1 SG AUX FRV BYP2-FW-56A (%)

FWR18 52 SG AUX FRV BYP2-FW-56B (%)

FWR19 VACUUM PUMP FSA SUCTION ISOL 2-AR-12A FWR20 VACUUM PUMP FSB SUCTION ISOL 2-AR-12B MISCSIMN6 (3) NSEM-4.03 REV.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 4 OF 9 FWR21 HOT WELL REJECT LCV ISOL 2-CN-21 FWR24 DRN CLR A INLET EQUALIZER 2-CN-214A(%)

FWR25 DRN CLR B INLET EQUALIZER 2-CN-214B(%)

FWR26 HTR DRAIN PUMP P3A DISCH ISOL 2-HD-91(%)

FWR27 HTR DRAIN PUMP P3B DISCH ISOL 2-HD-9B(%)

FWR28 HTR DRN PUMP P3A RECIRC STOP 2-HD-45A FWR29 HTR DRN PUMP P3B RECIRC STOP 2-HD-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 FWR33 SGFP A RECIRC VLV 2FW36A A/M CNTRL FWR34 SGTP A RECIRC 2FW36A FIC5237 S.P.(PSIG)

FWR35 SGFP B RECIRC VLV 2FW36B A/M CNTRL FWR36 SGFP B RECIRC VLV 2FW36B FIC5240 S.P. ( PSIG)

FWR37 LONG RE-CYCLE VALVE AOV20( % )

FWR38 COND DEMINS 1 A,1B,1C,1D,1E,lF,1G IN SERV FWR39 CST / SURGE TK M/U TO HW 2-CN-100,34 IWR41 FW REG VLV 2FW51A NORMAL / HAND CONTROL FWR42 FW REG VLV 2FW51A MANUAL VLV POSITION FWR43 FW REG VLV 2FW51B NORMAL / HAND CONTROL FWR44 FW REG VLV 2FW51B MANUAL VLV POSITION FWR45 FRV 2FW51A LEAKAGE (VLV POS %)

FWR46 FRV 2FW51B LEAKAGE (VLV POS %)

FWR47 FPT 1A AUX OIL PUMP TEST PB(HS-7185)

FWR48 FPT 1A EMERG OIL PUMP TEST PB(HS-7186)

FWR49 FPT 1B AUX OIL PUMP TEST PB(HS-7135)

FWR50 FPT 1B EMERG OIL PUMP TEST PB(HS-7136)

FWR51 COND 1A EVAC SUCT VLV 2 ARIA FWR52 COND 1A EVAC SUCT VLV 2AR1B FWR53 COND 1B EVAC SUCT VLV 2ARIC FWR54 COND 1B EVAC SUCT VLV 2AR1D FWR55 MAKEUP TO CST (GPM)

FWRS6 TRAIN A LP HEATER ISOL 2-CN-12A,43A(%)

FWR57 TRAIN B LP HEATER ISOL 2-CN-12B,43B(%)

FWR58 FIRE PUMP ALARM STATUS FWR59 #1 SG AUX FRV 2-FW-43A IWR60 #2 SG AUX FRV 2-FW-43B FWR61 il SG AUX FW ISO VLV 2-FW-18A FWR62 5 2 SG AUX FW ISO VLV 2-FW-10B FWR63 il SG FRV 2-FW-43A FWR64 #2 SG FRV 2-FW-43B O MISCSIMl6 (3) NSEM-4.03 REv.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 5 Or 9

+O l

IAR01 SA CTMT HDR ISOL 2-SA-19 IAR02 TURBINE BLDG IA ISOL (2-IA-25)

IAR03 IA CMPR F3A SW HS-7046 IAR04 IA CMPR F3B SW HS-7055 IAR05 IA CMPR F3A,B SEL SW HS-7046A i IAR06 SA CMPR F2 SW HS-7086A IAR07 TURB BLDG IA HDR AIR LEAKAGE ( SCFM)

IAR08 ENCL & AUX BLDG IA HDR AIR LEAKAGE (SCFM)

IAR09 CONTAINMENT IA HDR AIR LEAKAGE (SCFM)

IAR10 UNIT 1 SA SPLY BV 2SA12, 2SA12A j IAR11 IA CMPR F3A RESET SW '

IAR12 IA CMPR F3B RESET SW IAR13 SA CMPR F2 CNROL HS-7086B IAR14 CW PRIMING EDUCTOR LOAD (SCFM)

IAR15 IA CMPR F3C START /STOP SWITCH IAR16 IA CMPR F3C RESET SWITCH IAR17 IA CMPR F3C SEL SWITCH MSR01 #1 ATM STM DUMP ISOL 2-MS-3A i MSR02 #2 ATM STM DUMP ISOL 2-MS-3B MSR03 MSR 1A/1B HI LOAD VL 20-MS-85A,85B MSRO4 MSR 1 A/1B HI LOAD A/B POS( % )

MSR05 MSR 1A LOW LOAD VALVE 2-MS-79A MSR06 MSR 1A LOW LOAD VALVE 2-MS-79A POS(%)

(k MSR07 MSR 1B LOW LOAD VALVE 2-MS-79B MSR08 MSR 1B LOW LOAD VALVE 2-MS-79B POS(%)

MSR09 AUX STEAM SUPPLY MSR11 TURPINE SHELL WARMING MODE VLV STATUS MSR12 2-MS-201 CLOSING COIF >/RACKOUT MSR13 2-MS-202 CLOSING COIL /RACKOUT RCR01 PZR VENT VALVES 2-RC-021, & 2-RC- 421 RCR02 PZR/RCS/CVCS BORON CONC ( PPM)

RCR03 PRESSURIZER BORON CONC (PPM)

RCR04 RCP-40A RACKOUT RCR05 RCP-40B RACKOUT RCR06 RCP-40C RACKOUT RCR07 RCP-40D RACKOUT RCR08 PZR N2 SUPPLY 2-RC-030 & 2-RC-015 RHR01 SDC SUCTION HDR MAN STOP 2-SI-709 RHR02 A LPSI PUMP SUCT FROM SDC 2-SI-441 RHR03 B LPSI PUMP SUCT FROM SDC 2-SI-4 40 RHR04 A LPSI PUMP SUCT FROM RWST 2-SI-444 RHR05 B LPSI PUMP SUCT FROM RWST 2-SI-432 RHR06 LPSI PP P-42A RECIRC VLV 2-SI-449 RHR07 LPSI PP P-42B RECIRC VLV 2-SI-450 MISCSIM#6 (3) NSEM-4.03 REV. 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 j CERTIFIED REMOTE FUNCTIONS LIST UNIT 2 PAGE 6 OF 9 RER08 CTMT SPRAY PP P-43A RECIRC VLV 2-CS-7A RER09 CTMT SPRAY PP P-43B RECIRC VLV 2-CS-7B  ;

RER10 CTMT SPRAY PP P-43A DISCH VLV 2-CS-3A I RER11 CTMT SPRAY PP P-43B DISCH VLV 2-CS-3B ,

RER12 LPSI PP DISCH TO SDC HX A 2-SI-452 j RER13 LPSI PP DISCH TO SDC.HX B 2-SI-453 RER14 CTMT SPRAY HDR A VLV 2-CS-4A & 2-SI-456  ;

FER15 CTMT SPRAY HDR B VLV 2-CS-4B & 2-SI-457 l RER16 SIT RECIRC HDR STOP 2-SI-463 RER17 CTMT SPRAY PP P-4 3A RAC? OUT RER18 CTMT SPRAY PP P-43B RACKOUT RER19 LPSI PP P-42A RACKOUT i RER20 LPSI PP P-42B RACKOUT l RER21 SIT RECIRC HDR STOP 2-SI-459 RER22 CS HDR A TO SI TEST HDR ISO 2-CS-51 RER23 CS HDR B TO SI TEST HDR ISO 2-CS-50  ;

RER24 SDC WARMUP LINE 2-SI-400 POS %

RER25 SDC DISCHG TO CVCS 2-SI-040 RER26 SDC DISCHG TO RWST 2-SI-460 POS %

RER27 2-SI-306 AIR SUPPLY, FUSE BLOCK RER28 SDC FLOW CONTROL VALVE 2-SI-657 RER29 SDC FLOW CONTROL VALVE 2-SI-657 POS %

w RMR01 RIT-4262 ALARM DEFEAT SWITCH f

RKR02 RIT-5099 ALARM DEFEAT SWITCH RER03 RIT-6038 ALARM DEFEAT SWITCH IUER0 4 RIT-7890 ALARM DEFEAT SWITCH RER05 RIT-7891 ALARM DEFEAT SWITCH RER06 RIT-7892 ALARM DEFEAT SWITCH RMR07 RIT-7894 ALARM DEFEAT SWITCH RKR08 RIT-7895 ALARM DEFEAT SWITCH RMR09 RIT-7896 ALARM DEFEAT SWITCH futR10 RIT-1897 ALARM DEFEAT SWITCH RMR11 RIT-7899 ALARM DEFEAT SWITCH FUER12 RIT-8011 ALARM DEFEAT SWITCH IUtR13 RIT-8123A ALARM DEFEAT SWITCH IUtRi d RIT-8123B ALARM DEFEAT SWITCH RPutL 5 RIT-8132A ALARM DEFEAT SWITCH RMR16 RIT-8132B ALARM DEFEAT SWITCH RMR17 RIT-8139 ALARM DEFEAT SWITCH RMR18 RIT-8142 ALARM DEFEAT SWITCH futR19 RIT-8145A ALARM DEFEAT SWITCH RKR20 RIT-8145B ALARM DEFEAT SWITCH RKR21 RIT-8156 ALARM DEFEAT SWITCH RMR22 RIT-8157 ALARM DEFEAT SWITCH

/

MISCSIMf 6 ( 3) NSEM-4.03 REv.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST

() UNIT 2 PAGE 7 OF 9 1

RMR23 RIT-8262A ALARM DEFEAT SWITCH RMR24 RIT-8262B ALARM DEFEAT SWITCH RMR25 RIT-8434A ALARM DEFEAT SWITCH RMR26 RIT-84 34B ALARM DEFEAT P4 ITCH RMR27 RIT-8997 ALARM DEFEAT SOITCH RMR28 RIT-8998 ALARM DEFEAT SWITCH RMR29 RIT-8999 ALARM DEFEAT SWITCH RMR30 RIT-9049 ALARM DEFEAT SWITCH RMR31 RIT-9095 ALARM DEFEAT SWITCH RMR32 RIT-9116 ALARM DEFEAT SWITCH RMR33 RIT-9327 ALARM DEFEAT SWITCH RMR34 RI-8123A ALARM DEFEAT SWITCH RMR35 RI-8132A ALARM DEFEAT SWITCH RMR36 RI-8262A ALARM DEFEAT SWITCH RPR01 TCB-9 BREAKER RPR02 MG SET A INPUT CONTACTOR RPR03 MG SET A OUTPUT BREAKER RPR04 MG SET A OUTPUT CONTACTOR 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 RPR14 COMPARATOR AVERAGER SELECT CHANNEL-B COMPARATOR AVERAGER SELECT CHANNEL-C RPR15 COMPARATOR AVERAGER SELECT CHANNEL-D RPR16 CH-X INPUT TO POWER RATIO CALCULATOR RPR17 CH-Y INPUT TO POWER RATIO CALCULATOR RPR18 CH A&B HI PZR PRESS MODULE DESCONN RPR19 VARIABLE HI POW TRIP RPR20 RCP LO SPEED TRIP RPR21 RCS LO FLOW TRIP RPR22 SG 81 LO LEVEL TRIP RPR23 SG #2 LO LEVEL TRIP RPR24 SG #1 LO PRESS TRIP RPR25 SG #2 LO PRESS TRIP RPR26 HI PZR PRESS TRIP RPR27 TM/LP TRIP RPR28 TURBINE / REACTOR TRIP RPR29 HI CONTAINMENT PRESS TRIP RPR30 LOCAL POWER DENSITY TRIP RPR31 POWER RATIO RECORDER TENT SELECTED j O MISCSIMI6 (3) NSEM-4.03 REV.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 l CERTIFIED REMOTE FUNCTIONS LIST  !

l UNIT 2 PAGE 8 OF 9 l

l RXR01 LOOP 1 SEL FOR TAVG CALC IN RRS-UNIT 1 )

RXR02 LOOP 2 SEL FOR TAVG CALC 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 PROP HTR BKRS IN TEST POSITION i SGR01 BD TK INLETS 2-MS-381A&B(100%=6 TURNS) l SGR02 BD Q TK INLETS 2-MS-218,219(100%=6 TURNS)

SIRO 1 HPSI PP P-41A RACKOUT SIR 02 HPSI PP P-41B RACKOUT SIR 03 HPSI 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 i SIR 07 T-39D SI-644 RACKOUT/CL COIL REMOVAL SIR 08 SI TANK VENT LINE CAPS SWR 01 SW PUMP A DISCHG 2-SW-2A SWR 02 SW PUMP B DISCHG 2-SW-2B SWR 03 SW PUMP C DISCHG 2-SW-2C SWR 04 RBCCW HX A/B XCONN 2-SW-7B SWR 05 RBCCW HX B/C XCONN 2-SW-7A SWR 06 TPCCW HX A/B XCONN 2-SW-4B

\

SWR 07 TPCCW HX B/C XCONN 2-SW-4A SWR 08 RBCCW HX A OUTLET VLV MODE l SWR 09 RBCCW HX B OUTLET VLV MODE I SWR 10 RBCCW HX C OUTLET VLV MODE I SWR 11 ISOL VLV 2-SW-5A TO TP HX-17A(%) I SWR 12 ISOL VLV 2-SW-5B TO TP HX-17B(%)

SWR 13 ISOL VLV 2-SW-5C TO TP HX-17C(%)

SWR 16 VITAL AC CLR 181 ISOL OUTLET 2-SW-181A%

SWR 17 VITAL AC CLR 182 ISOL OUTLET 2-SW-181B%

SWR 18 VITAL AC CLR 183 ISOL OUTLET 2-SW-181C%

SWR 19 "A" SW HDR TO CHILLERS 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 SWR 23 A RBCCW HX OUTLET 2-SW-9A (%)

SWR 24 B RBCCW HX OUTLET 2-SW-9B (%)

SWR 25 C RBCCW HX OUTLET 2-SW-9C (%)

SWR 26 SW PUMP "B" BREAKER STATUS TCR01 EHC BYP VALVE POS. (TV-1) (%)

TCR02 EHC OIL PUMP TEST (FV9/10)  !

O MISCSIMf6 (3) NSEM-4.03 REV.: 0 1

DATE: 2/7/89 PAGE: 7.1-1 of 1

FIGURE 7.1 CERTIFIED REMOTE FUNCTIONS LIST

- UNIT 2 PAGE 9 OF 9 TCR03 CONDENSER LOW VAC TRIP TCR04 BRG OIL PRESS LO TRIP TCR05 SHAFT PUMP PRESS LO TRIP TCR06 LOSS OF STATOR COOLANT TRIP TCR07 EHC HYD PRESS LO TRIP TCR08 MSR HI LVL TRIP TCR09 UNIT ELECTRICAL PROTECTION TRIP TCR10 REACTOR /TURB TRIP TCR11 MANUAL EMERG PB TRIP TCR12 SG #1 HI LVL TRIP TCR13 SG #2 HI LVL TRIP TCR14 EXHAUST HOOD TEMP HI TRIP TCR15 PMG MALFUNCTION LIGHT RESET TPR01 TBCCW SURGE TGK SUPPLY 2-FMW-219 TPR02 TBCCW PUMP A DISCHG 2-TB-3A (%)

TPR03 TBCCW PUMP A DISCHG 2-TB-3B (%)

TPR04 TBCCW PUMP C DISCHG 2-TB-3C (%)

TPR05 TIC-6305 HX-17A TO VLV 2-SH-88A(F)

TPR06 TIC-6304 HX-17B TO VLV 2-SW-88B(F)

TPR07 TIC-6303 HX-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)

_) TPR11 ISO-PHASE BUS FAN A C5 l TPR12 ISO-PHASE BUS FAN B C5 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 IA 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 (RI/RO)

WDR01 AERATED WASTE DRAIN TANK PUMP P31A 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 LI SOUND 2-SSP-21 WDR06 DEGASIFIER BYPASS VALVE 2-LRR-69 l

MISCSIM86 (3) NSEM-4.03 REV.: 0 DATE: 2/7/89 PAGE: 7.1-1 of 1

O ATTACHMENT 14 COMMONLY USED ABBREVIATIONS AND DEFINITIONS l

1

)

i i

I I

i l

1 O

O

SIMULATOR SYSTEM ABBREVIATIONS CC Component Cooling - Reactor Building Component Cooling Water (RBCCW)

CH Containment Heating and Ventilating (HVAC)

CR Reactor Core CV Chemical and Volume Control System (CVCS)

CW Circulating Water ED Electrical Distribution EG Electrical Generation ES Engineered Safety Actuation System (ESAS)

FW Feedwater IA Instrument Air MS Main Steam PC Process Computer RC Reactor Coolant System RD Control Rod Drive System RH Residual Heat Removal / Shutdown Cooling System RM Radiation Monitoring

() RP RX Reactor Protection System (RPS)

Reactor Regulating System (RRS)

SG Steam Generators SI Safety Injection SW Service Water l 1

TC Turbine Controls TP Turbine Building Component Cooling Water (TBCCW)

TU Turbine WD Waste Disposal o

NRC #2 (4) 2

COMMON ABBREVIATIONS Abbreviation Full'Name

jl AI -Analog Input AO ' Analog Output

- ANS 3.5 -ANSI /ANS 3.5 (1985)

ARO All Rods Out BOL Beginning of Life CAR Containment Air Recirculation-CEA Control Element Assembly CEDS Control Element Drive. System CIAS Containment _ Isolation Actuation Signal CRAC Control Room Air Conditioning CVCS . Chemical'and Volume Control System DG Diesel Generator DI Digital Input j DU Digital Output DR Deficiency Report j EBFAS Enclosure Building Filtration Actuation Signal EBFS Enclosure Building Filtration System ,

EDG Emergency Diesel Generator. ,

EOL End of Life EOP Emergency Operating Procedure i ESAS Emergency Safeguards Actuation System HFP Hot-Full Power HPSI High_ Pressure Safety Injection HZP Hot Zero Power IC Initial Condition ITC Isothermal Temperature Coefficient LNP Loss of Normal Power LPSI Low Pressure Safety Injection MOL Middle of Life MP2 Millstone Point Unit #2 WRC #2 (48 3

i COMMON ABBREVIATIONS. .]

Abbreviation Full'Name

)

MSIV Main Steam Isolation Valve MLSB Main Steam Line Break j MTC Modcswtor. Temperature Coefficient --

MWE Mega47tts Electric- j NSEM Nuclear Simulator Engineering Manual PDCR Plant Design Change Request l

PPC Plant Process Computer' PTL Pull-To-Lock ]'

RBCCW Reactor Building' Component Cooling' Water l RCP Reactor Coolant. Pump RCS Reactor Coolant System RPS Reactor Protection' System RRS Reactor Regulating System SCCC Simulator Configuration Control Committee SDC Shutdown Cooling System or Simulator Design- .

Change SIAS Safety Injection Actuation Signal SIG Simulator Instructor Guide

. SIT Safety Injection Guide 1

SOER Significant Operating Event Report SSD Simulator System Diagram' S/G Steam Generator SW Service Water TBCCW Turbine Building Component Cooling Water TM/LP Thermal Margin / Low Pressure O

NRC #2 (4) 4

DEFINITIONS

)

Anomalous Response - Simulator response which violates the physical l 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 shape Index (A.S.I) - 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.

Backtrack - The ability to move the simulator back in time to conditions which had previously existed. This is accomplished-by the automatic storage (at one minute intervals) of the simulator.s I/O's over the past hour.

Best Estimate Evaluation - A method used, (in the absence of plant j 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, j

() Boolean Trigger - an algebraic expression which is used to automatically activate a malfunction when its value beenmes true.

"Cause & Effect" Document - A description of the simulator response (effect) to the insertion of a specific malfunction or malfunctions.

Each malfunction description also contains the physical "cause" of.the l malfunction as well as a description of the significant effects on 1 i

plant operation due to the malfunction.  !

Certified 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.

Certified Remote Function - Those remote functions which will be tested to work correctly-and may be used in simulator training and exams.

Composite Malfunction - A combination of up to 10 predefined simple malfunctions which can be arranged in a logical sequence. Once built, this composite malfunction is stored and can be used at any time.

Core Performance Testing - Plant heat balance, determination of' shutdown margin, measurement of reactivity coefficients'and control rod worth using permanently installed instrumentation.

O une e2 <*' 5

i 1

)

Critical Parameters --Those parameters, specific to a given major O malfunction, which are driven directly by the initiating event, required for diagnosis, or required to verify proper plant response to safety equipment actuations and/or operators' corrective actions.

Deficiency - An identified difference in a simulator quality or element (hardware and/or software) that requires review and resolution.

Deficiency Report (DR) - Form (STS-BI-FlA) used by the Operator Training Branch.(OTB'i and the Simulator Technical Support' Branch (STSB) to record all identified simulator deficiencies between the simulator and reference plant.

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 pressure, J etc.

1 EOP's - Emergency Operatin. Procedures address the required response l by operations personnel to emergency conditions. l Fast Time - The increase in the speed at which certain parameters (such as Aenon, 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 1 O simulator is taken out of f reeze, the model will continue to run from the time that it was placed in freeze.

Hot Standby Operations - Maintaining stable plant conditions at hot standby.

Input / Output (I/0) - Any digital or analog computer inputs / outputs. j Initial Conditions (IC's) - A set of analog / digital points that are stored on the Simulator's Computers so 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.

Major Malfunction - Those malfunctions which produce extensive integrated effects in a number of plant systems which requires complicated analysis to verify acceptable response.

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 function as a plant-referenced training / examining tool.

O wne *2 (4i 6

l

Malfunction - A specific equipment failure which produces discernible indications in the Control Room that replicates the same equipment

- () failure should it occur in the reference plant. Specific preprogrammed malfunctions are available at the simulator instructor station.

Model Limits - Physical conditions which cannot be simulated by the model coding, e.g.: critical pressure 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 development and implementation of the certification program. It is a controlled document and its purpose is to insure consistent application of the certification process.

Nuclear Start-Up - From all CEA's fully inserted to going critical at hot standby conditions.

Operability Testing - A defined group of tests conducted to verify:

1. The overall completeness and integration of'the simulation model,

2. Steady state performance of the simulator to that of the reference plant, O 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 certification and for every subsequent four year period in l

order to maintain certification. Performance 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.

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 pos:er to hot standby, then cooldown to cold shutdown conditions.

Plant Startup - 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.

(

nw ,, o,

Reference Plant Data Book (PDB) - A compilation of reference plant ]

.O. -data for specific plant transients / evolutions._ The data defines plant'

[q parameter. response to specific initiating events or 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.

Simulator Configuration Control Committee (SCCC)'- The committee responsible for overall simulator design control and management of NTD t resources involved in the simulator modification effort. -The -

committee shall include as permanent members; the Director of NTD, the Managers of OTB and STS, Supervisor - SCE, Supervisor - HW Maintenance, and the four Unit Operations Consultants.

Simulator Design Change (SDC) - A documentation package consisting of relevant DR's and all forms indicated on STS-BI-FlE which is designed to track the resolution of DR's and ensure that ANSI /ANS 3.5-1985, and NRC Reg. 1.149 requirements are satisfied.

Simulator Instructor Guide (SIG) - A training document outlining the sequence of events for a simulator training session. SIG's also contain additional information for the instructor conducting the i 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.

i l

Simulation System Diagram (SSD) - Functional representation of the simulator modeling for a given system, q Slow Time - In reality, this is the expansion of real time which produces the appearance that a transient is occurring at a slower speed. The slow time which can be selected can vary from 5% to 95% of i real time (at 5% increments). i Snapshot - The recording of the present status of all simulator digital / analog I/O's. After this snapshot is taken, the simulator may I be initialized to this condition at some later time.

SOER - Significant Operating Event Report is generated by INPO and distributed to industry members. It includes recommendations concerning the event which must be addressed by concerned facilities. _

SRO Oualified Instructor - An instructor who is (or was in the past) an NRC licensed Senior Reactor Operator (or certified), who by nature of his training and experience, has the knowledge to make decisions on proper plant system alignments for given operating conditions.

O nne e2 (4) 8 L______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Surveillance Testing - Operation conducted surveillance testing on safety related equipment or systems.

System Test - A test developed for each modeled plant system that ensures proper response of all control board instrumentation, controls, annunciators, PPC points, remore functions, flowpaths and components that are associated with an individual plant system.

Turbine Generator Start-Up - Turbine Generator at zero RPM, to rated speed and synchronization to grid.

O i

O nne e2 (4) 9 l

i l

I n

U .

ATTACHMENT 15 <

QUALIFICATIONS OF PERSONNEL DEVELOPING AND PERFORMING SIMULATOR CERTIFICATION TESTING l

• s This attachment is referenced by section 1 j of the Performance Test Summary.

( l i

i l

L -(

1 t'

l l This attachment lists the Nuclear Training Department personnel involved in developing and conducting Millstone 2 Simulator Certification Testing. A brief description of the relevant qualifications is provided with each individual.

As is readily apparent, an excellent mix of operating, training, engineering

]I and simulation software experience is represented by these Individuals.

Only those personnel directly involved are included. Additional experienced NTD personnel were involved in the certification effort, but to a lesser degree.

l l

O 2-

SIMULATOR TECHNICAL SUPPORT BRANCH Lung-Rui Huang - Supervisor, Simulation Computer Engineering NU Experience:

o Five years in Simulation Computer Engineering as both a Senior Engineer and Supervisor.

o Two years in Probabilistic Risk Assessment and Safety Analysis with Reactor Engineering.

Other Related Experience:

O.

o Four years simulation experience including Technical Staff Leader with Electronic Associates, incorporated.

o Six years university teaching experience at National Tsing Hua University and towa State University in Nuclear and Electrical Engineering Departments.

Education:

o PhD in Nuclear Engineering O

Shih-Kao Chang - Senior Engineer, Simulation Computer Engineering NU Experience:

o Three years in Simulation Computer Engineering as an Engineer and Senior Engineer, with lead responsibility for NU Reactor Core and RCS modeling disciplines.

Other Related Experience:

o Four years simulation experience with Singer Link-Miles Simulation Corporation, including the position of Section Head, PWR NSSS Modeling.

Education:

o PhD in Nuclear Engineering O

Hsin-Cheng Huang - Engineer, Simulation Computer Engineering i

l NU Experience: l i.

o Two years in Simulation Computer Engineering as an engineer. l Other Related Experience: )

o Three years simulation experience with Singer Link-Miles Simulation Corporation.

1 Education- l o PhD in Mechanical Engineering with specialization in Thermofluids and Thermohydraulics I

i

OPERATOR TRAINING BRANCH O

r Joseph J. Parillo - Assistant Supervisor, Operator Training, Millstone 2 ,

1 NU Experience:

o Three years in Operator Training as the supervisor responsible for simulator performance and training.

o Five years in Millstone 2 Engineering as the Millstone 2 Reactor i Engineer and Assistant Engineering Supervisor with the following )

f responsibilities:

Reactor Core Secondary Plant Performances Duty Officer 6

\ o Six years in Reactor Engineering including three years in Nuclear Fuel Planning and three years in Reactor Core Analysis with responsibilities for:  ;

Core Design Codes, including CASMO and SIMULATE i

- the transition from Combustion Engineering to Westinghouse fuel j including safety analysis, physics testing and incore analysis. i Education: ,

o BS in Nuclear Engineering NRC Licenses:  !

o Currently SRO licensed, MP2 (four years)

O i

l

o Richard N. Spurr - Senior Operator Instructor, Millstone 2  !

NU Experience: -) i i

i l

o Five years in Operator Training as a Senior Operator Instructor, including simulator construction, startup and testing responsibilities.

l I l o Three years as a Millstone 2 Shift Supervisor.

] i j

o Five years as a Millstone 2 Supervising Control Operator.

• i i

Other Related Experience: '

l o Six years, U. S. Navy  ;

( l i

NRC Licenses: 'j o Currently SRO Licensed, MP2 (13 years) i l

)

O d a t

a i

J Donald H. Wright - Operations Consultant, Millstone.2 Operator Training -

. NU Experience: 2 i

o Five years in Operator Training as a Senior Operator Instructor and l Simulator Operations Consultant, including simulator construction, startup and testing responsibilities.

o Five years as a Millstone 2 Shift Supervisor.

o Three years as a Millstone 2 Supervising Control Operator. 1 Other Related Experience: .

o 22 years, U. S. Navy, including qualification as Engineering Officer of ]

the Watch.

1 NRC Licenses:

o Currently SRO Licensed, MP2 (15 years)

• 1 l i I

I l

'I

.j O  !

Kenneth G. Truesdale - Operator instructor, Millstone 2 NU Experience- >

l o Two and one-half years in Operator Training as an Operator instructor.

o One year as a Millstone 2 Supervising Control Operatoi o Three years as a Millstone 2 Control Operator.

o Two years as a Millstone 2 Plant Equipment Operator.

I Other Related Experience:

o Eight years, U. S. Navy, including qualification as Engine Room j Supervisor. j 0-NRC Licenses:

l o Currently SRO Licensed, MP2 (six years with RO or SRO license) l O  !

8-

l

,m Robert H. Burnside - Senior Operator Instructor, Millstone 2 l k) m

.l

,k NU Experience:

i o Two and one-half years in Operator Training as a Senior Operator j Instructor.

o Eleven years as a Millstone 2 Shift Supervisor. ]

l o Twelve years additional operating experience, including Control Operator 1 at the Haddam Neck Plant. ]

i NRC Licenses:

o Currently SRO Licensed, MP2 (13 years) 3 l

I i

l t

\ )

s__/

9 i

l

l William H. Sou' der - Operator Instructor, Millstone 2 1

NU Experience:

o Four t.nd one-half years in Operator Training as an Operator Instructor, .

Includitig simulator construction, startup and testing responsibilities.-

Other Related Experience:

o Four years as a Simulator Instructor at the Combustion Engineering Simulator.

o Foutteen years, U. S. Navy, including Engineering Officer of the Watch,-

Qualifications and power plant prototype training experience.

NRC Licenses:

o Currently SRO Licensed, MP2 (three years)

L__ ___ _ _- _-_ _ - - _ _ _ _