ML20202F339
| ML20202F339 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/31/1986 |
| From: | CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| Shared Package | |
| ML20202F335 | List: |
| References | |
| NUDOCS 8607150144 | |
| Download: ML20202F339 (123) | |
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i PERRY NUCLEAR POWER PLANT t 1
CONTROL ROOM VALIDATION
SUMMARY
REPORT 4
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.' PREPARED FOR ! <
t TH2 NUCLEAR REGULATORY COMMISSION l l :
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1 BY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY I O
j JULY, 1986 I
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i 8607150144 860711 PDR ADOCK 05000440
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O TA3LE OF CONTENTS PAGE 1.0 INTh0 DUCTION 1 2.0 FLANNING 2 3.0 REVIEW 5 ,
4.0 l ASSESSMENT 7 5.0 IMPLEMENTATION 16
6.0 CONCLUSION
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LIST OF ATTACHMENTS ATTACHMENT DESCRIPTION A PNPP CONTROL ROOM VALIDATION PLAN B NOV. 1985 AND JAN., 1986 VALIDATION SCENARIOS C CONTROL ROOM AND VALIDATION HED'S .
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PAGE 1
1.0 INTRODUCTION
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4 The Perry Nuclear Power Plant has successfully completed the Control Room Validation which was performed to ensure that the NUREG 0737, Supplemeat 1, individual initiatives, have been integrated sufficiently to meet the needs of the Control Room operators and provide adequate emergency i response capabilities. This summary report is submitted to l the NRC to document the planning, review, assessment and 1
l implementation phases of the November, 1985 and January, 1986 Control Room Validation exercises.
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2.0 PLANNING 4
! The objective of the PNPP Control Room Validation was to evaluate the adequacy and usefulness of the Control Room canels with Detailed Control Room Design Review (DCRDR) and l
R.G. l.97 improvements implemented, the Plant Emergency Instructions (PEI's), and the Safety Parameter Display System (SPDS), in enabling the trained operating crew to respond to emergency situations. The validation process was defined as an iterative effort, since if design modifications or instructions improvements were determined to be necessary, then consequent re-performance of discrete portions of the validation with additional exercises may be required. The i
Control Room Validation Plan was developed by the Human
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] Factors multi-disciplinary team, and is provided as Attachment A.
l Ine simulator cnd walk-through validation methodology was selected whereby the Control Room crew performs functions I
and tasks on the plant specific simulator in real time during a scenario. Validation evaluation is performed by a ,
i multi-disciplinary review team and the operating crew for each scenario.
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-m s Prior to the validation, the Plant Simulator and Control Room were evaluated to identify the significant design deviations. 'The simulator was modified to incorporate these deviations which included design changes from the DCRDR and other plant design change packages. Differences between the simulator panel design and actual plant Control Room panel design at fuel load were identified and were determined to be of low significance relative to enacting the scenarios.
The validation scenarios were selected in advance such that the major or representative paths in all sections of the PEI's were utilized. Simulator response limitations for scenarios were only identified in the modeling of extremely high containment pressures and temperatures, and of plant radiation release related to core degradation. In these two scenarios, these " limited" parameters, and trends were provided to the operators as discrete static values by the simulator operator either verbally or on " post-it" stickers.
This static simulation of a few limited parameters did not detract from the real time scenario enactments. A simulator scenario form for each scenario was prepared detailing the PEI's evaluated, initial plant conditions, and scenario milestone events including event descriptions and expected operator actions. The eight scenarios used in the November, 1985, and the two scenarios used in the January, 1986, validation exercises a e included as Attachment B.
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PAGE 4 I
! The data collection techniques used for validation
- evaluation includes the use of observer checklists and video i
recording for each scenario enactment.
i j The validation review team conducted a pre-validation 1
exercise in October, 1985, observing a total of ten scenarios i
performed by two crews to familiarize all members of the review team with Control Room activities, verify the validation methodology, and verify the placement of the l
audio-visual equipment. Following the January Control Room i i Validation, a PEi Validation was conducted in March to
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] validate a specific revision to the PEI's. The March PEI Validation is discussed in detail in the PNPP PEI Validation
{} Summary Report (Ref. CEI/NRR-0447L).
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i 3.0 REVIEW 4
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4 Two Control Room validation exercises were conducted.
i The first one in November, 1985, which evaluated 15 scenario
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enactments performed by two crews and the second one in
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January, 1986, with four scenarios enacted by two crews. The
, following is applicable to both exercises unless otherwise noted.
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4 The multi-disciplinary validation review team consisted i of the following personnel:
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- Operations Engineer - Wally Colvin
- - Human Factors Unit I&C Engineer - Larry Lawrence j - License Training Instructor - Mike Morrow e
j - Human Factors Specialist - Sharon Eckert 1
] - General Electric Systems Engineer - Jim Howard i
4 In the January validation exercise a Shift Supervisor,
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4 Don Cobb, was added to the Review Team to provide additional perspective. During the exercises the Review Team took notes 4
t on the Simulator Scenario Form concentrating on behaviors and l,
operational activity that would indicate a potential problem.
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PAGE 6 Each validation exercise consisted of two operating crews, designated Crew 1 and Crew 2. Each Operating Crew i t
consisted of five individuals designated as the minimum shift crew composition in the Technical Specifications:
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r j - Two licensed or certified Senior Reactor Operators (Shift and Unit Supervisor) l
- Two licensed or certified Reactor Operators
- One Shift Technical Advisor i
l The Operating Crew performed or simulated all actions i
- including, communications with individuals in other areas of j the plant or offsite, notification of local or federal I
i governmental agencies as if they were under actual operating conditions. All required Emergency Plan notification forms were filled out and simulated phone calls were logged, t I
l Because the SPDS may not always be available in i emergency situations, each scenario was enacted once with SPDS available and once with SPDS unavailable. Crew 1 had SPDS l available on odd numbered scenarios and Crew 2 had SPDS available on even numbered scenarios. !
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PAGE 7 4.0 ASSESSMENT fs
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Immediately following the enactment of each scenario, the members of the review team and the operating crew completed the respective check list included as Attachment A:
- Review Team Evaluation Checklist-Attachment 2
- Operating Crew Evaluation Checklist-Attachment 3 After completion of the checklists, the operating crew met with the review team for a critique of the scenario just completed. First the operating crew identified any major comments including positive or negative aspects of the scenario
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enactment. Next each member of the operating crew and review team identified observations in each of the check list categories (ie: PEI's, SPDS, and Control Room). Each observation was discussed with the operating crew to resolve the explanation of possible causes and potential solutions.
Following the completion of the critique, the review team analyzed the evaluation checklists, records of critique sessions, and scenario comments to provide an overall validation summary and subsequent recommendations. Validation observations pertaining to the Control Room design and SPDS
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were assessed for HED's, and improvements. selected by the team
! in accordance with the DCRDR Assessment and Selection of Design Improvement methodology as described in the DCRDR Summary Report (PY-CEI/NRR-0167L). PEI observations were assessed as described in the PNPP PEI Validation Summary Report (CEI/NRR-0447L). A summary results of the November, 1985, and J January, 1986, validation exercises along with the PEI, Control 1
Room and SPDS findings is provided below. Control Room and 1
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SPDS HED Sheets with the description, review and implementation j dates have been provided in Attachment C.
t NOVEMBER, 1985 VALIDATION This exercise evaluated 15 scenario enactments performed by two crews. One crew demonstrated an overall acceptable performance, but the other crew experienced difficulties related to training. Training problems were resolved by providing additional operator and STA training in the PEI 4
basis, procedural flowpaths and operational limits along with
{ ERIS SPDS screens, dynamic codes, parameters, algorithms and alarm limits.
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A. PEI On the basis of the one crews' performance, the PEI's were demonstrated to be effective, workable and compatible with the Perry Unit 1 Control Room Design.
The PEI instructions were proven to be capable of i
effective implementation when the major flow paths of the Perry Plant Specific Guidelines were tested with dynamic scenarios, although areas were identified where instruction improvement was required. PEI procedural discrepancies were resolved and revisions I
j issued to provide clear and accurate detail while allowing the desired operational flexibility. A list of the PEI comments and resolutions may be found in j the PNPP PEI Validation Summary Report l
(Ret. CEI/NRR-0447L).
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! B. CONTROL ROOM i
I' The validation scenarios performed by both operating crews showed that the improved Control Room
! design adequately supported the operating crews' emergency tasks. However, three new HED's were identified as follows:
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PAGE 10 The upper pool level scales for meters G43-R022A(B), and recorders G43-R073A(B),
have a non-standard zero reference to the normal surface level (HED-614).
The RCIC manual initiation pushbutton must be held in for two second before RCIC will initiate to ensure that the main and feedpump turbines have tripped. However the RCIC seal in light comes in immediately which causes operator confusion (HED-616).
The digital displays on H13-P680 do not indicate when they are off range and the range selector switch should be changed (HED-613).
C. SPDS The ERIS displays were utilized in many ways by the two crews throughout the six evaluated scenarios.
Generally, the ERIS screens were not monitored periodically to assess plant status. The validation review team concluded that the ERIS displays could not
PAGE 11 3 be validated due to limited usage. ERIS was not a d hindrance to Control Room operations, but with limited usage could not be substantiated. No Control room crew demonstrated a thorough understanding of the scope of the displays or how to effectively use the displays to implement the PEI's. Therefore, the review team recommended that the following SPDS training be performed for all crews prior to the January, 1986, validation exercises:
- The Shift Technical Advisor should be thoroughly familiar with the ERIS, SPDS display system: all screens, parameters, algorithms, and alarmed units.
- The Control Room crew should have training to provide direction on how the SPDS displays can be effectively utilized.
One SPDS HED was identified as follows:
- RPV Temperature Display should show RPV cool-down rate in units of degrees F/Hr. versus degrees F/30 Min. to make it consistent with the procedure.
PAGE 12 JANUARY, 1986 VALIDATION 3
(J This second validation consisted of four scenarios enacted by two crews, and evaluated the revised PEI's and training improvements implemented to resolve discrepancies identified during the November, 1986, validation. This exercise observed both crews using ERIS to gain maximum data on SPDS screen usage. The validation review team and the Operations General Supervisor rated the overall performance as very good.
A. PEI On the basis of both crews' scenario performances, the PEI's were demonstrated to be usable and compatible with the Control Room panels and the operating crew.
The minimum shift crew manning was adequate to perform the required PEI tasks. While most of the comments and discrepancies noted were of minor significance, difficulty was observed in performing one task in PEI-B13, Reactor Pressure Vessel Control. After a procedure correction was made, a PEI Validation was performed in March, 1986, to prove the adequacy of the PEI Revision. A total list of the PEI comments and resolution may be found in the PNPP PEI Validation Summary Report, dated April 24, 1986,(Ref CEI/NRR-0447L).
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B. CONTROL ROOM i
In general, both operating crews were able to perform the required emergency functions and tasks effectively within the improved control room design.
i The results of the validation showed that ove'rall, the Control Room has sufficient instrumentation, controls and other equipment available and suitable from a human engineering standpoint to effectively accomplish the emergency tasks. One Control Room HED was identified as follows:
- P680 Reactor Operator attempted to restore the feedwater system to maintain vessel level, j however, he had not realized why the RFPT's were l
3 tripped. The reason for the trip was because of 1
i a RCIC initiation (HED-615)
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j C. SPDS 1
1 Both crews were observed during the performance .
-i j of the two scenarios with the ERIS SPDS displays j available. Operator actions were enhanced by the use J
of the safety oisplay information. In no case did the ERIS SPDS displays disable or misinform the crews.
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PAGE 14 The nature of ERIS usage during the exercised varied depending on the crew. In general, Crew 1 i
utilized the US workstation ERIS terminal as a command post where the SS, US, and the STA were all able to monitor several parameters simultaneously to arrive at the proper course of action. Crew 2 preferred to use ERIS as a supplement to panel information as well as trending PEI conditional parameters while the crew including the US pursed appropriate action from the horseshoe. In all cases, the STA's found ERIS helpful in providing current plant status with respect to PEI figures.
Several displays were utilized effectively by the i
crews. During the majority of time, the top level displays were used to provide overview information.
However,when required by the PEI response, other more
! specific information was retrieved. Especially 1
j helpful, were the trend plot for RPV Level, the RPV a
- Level displays, and various 2-D Plots associated with RPV and containment parameters. Of special note was i
the primary usage of the following three top level l 1
j displays:
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I 000 Critical Plant variables 003 RPV Control WR & Power 031 Containment Control NR There were no SPDS HED's identified in this validation.
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- 5.0 IMPLEMENTATION
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The complete Control Room Validation identified a total of 29 deficiencies and which broke down as follows:
Deficiencies
- 1. PEI 24
- 2. Control Room 4
- 3. SPDS 1 Implementation schedules for PEI deficiencies are contained in the PNPP PEI Validation Summary Report
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(CEI/NRR-0447L). Control Room and SPDS HED implementation dates are shown in the planned completion section of the HED report sheets in Attachment C. Currently all HED's are scheduled to be implemented prior to startup following first refuel outage.
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6.0 CONCLUSION
In conclusion, the PNPP Control Room Validation has successfully evaluated the adequacy and usefulness of the Control Room panels, PEI's and SPDS in enabling the trained operating crew to respond to emergency situations. The results of the validation show that overall, Perry has sufficiently integrated these individual initiatives in order to meet the needs of the Control Room operators and to provide adequate emergency response capabilities. We believe that Perry has met the requirements of NUREG 0737, Supplement 1, for coordination and integration of these individual initiatives.
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O ATTACl! MENT A i
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l PNPP CONTROL ROOM VALIDATION PLAN l 1
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- 1. INTRODUCTION The FNPP Control Room Validation will be performed to evaluate the adequacy and usefulness of the Control Room Panels with DCRDR and RG 1.97 improvements, the Plant Emergency Instructions (PEIs), and
' the Safety Parameter Display System in enabling the trained oper-ating crew to respond to emergency situations. This procedure decribes the method to be used to accomplish the PNPP Control Room Validation.
- 2. VALIDATION PROCESS
- a. The Perry Control Room Validation process consists of the following phases:
- 1) Scenario Development
- 2) Validation Preparation
- 3) Simulator Exercises
- 4) Evaluation
- 5) Resolution
- 6) Reporting
- b. Due to the fact that design modifications and instruction improvements may be necessary, the validation is an iterative effort and consequently reperformance of discrete portions of the validation may be required.
- 3. SCENARIO DEVELOPMENT 3.1 Criteria Scenarios of suf ficient number and variety to adequately test the major procedural flow paths addressed in the PEIs shall be devel-oped. Scenarios will be numbered sequentially for ease of identi-fication. Each scenario will detail the PEI flow paths followed.
3.2 Re spo ns ibility The Control Room Validation Operations Engineer shall be responsible for the development and selection of scenarios to be evaluated. A Simulator Scenario Form Attachment 1, shall be prepared for each scenario. A detailed description of scenario milestones, including event descriptions, expected operator actions and space for comments shall be attached to the Simulator Scenario Form.
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- 4. VALIDATION PREPARATION i
The Review Team will prepare for the validation simulator exercises by performing a trial review of an operating crew responding to a scenario in the simulator. Preferably the crew observed will be 8 different from the operating crews evaluated. Any validation methodology changes determined to be required shall be documented.
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- 5. SIMULATOR EXERCISES 5.1 Operating Crews There will be two operating crews, designated Crew I and Crew 2.
Each Operating Crew shall consist of five individuals designated as the minimum shif t crew composition in the Technical Specifications:
- a. Two licensed or certified Senior Reactor Operators, SS and US.
- b. Two licensed or certified Reactor Operators, 20s.
- c. One Shif t Technical Advisor, STA.
i Each Operating Crew will enact and be evalurted on each scenario, using the Perry simulator. Because the SPD3 may not always be available in emergency situations, each scenario shall be enacted once with SPDS available and once with SPDS unavailable. Crew I will have SPDS available only on odd-numbered scenarios; Crew 2 will have SPDS available only on even-numbered scenarios.
The Operating Crew should perform or simulate all actions (including co=munications with individuals in other areas of the plant or of f-site., notification of local or federal govern = ental agencies, etc.) as if they were under actual operating conditions. All required emergency plan notification forms will be filled out and simulated phone calls may be logged. Each exercise will commence with one SRO and one RO in the control area. The remainder of the Operating Crew will wait outside the simulator room until summoned by the operators in the control area.
5.2 Review Team i The multidisciplinary Review Team shall minimally consist of the following five individuals:
- a. Operations Engineer
- b. Human Factors Unit I&C Engineer
- c. Nuclear Training Instructor
- d. Human Factors Specialist
( e. General Electric Systems Engineer 1
HED REPORT SHEET HED-617 HED DESCRIPTION: Validation - RPV temperature display should show RPV cooldown rate in units of *F/HR vice *F/30 min which makes it consistent with the procedure.
HUMAN FACTORS REVIEW: Cooldown rate will be displayed in units of *F/HR and will alarm whIn the 100 *F/HR cooldown rate is exceeded.
H.F. GUIDELINES: NUREG 0700 Section 7, computers IMPLEMENTATION: Fix SAFETY AND OPERABILITV ASSESSMENT: CATEGORY I SAFETY CONSEQUENCE 7 No INCREASE ERROR POTEhTIAL7 No VERIFICATION: CORRECTS HED? NO NEW HED?
ORRECTION SCHEDULE: PLANNED COMPLETION Prior To Startup Following First Refuel WORK COMPLETE 7-3-86
REFERENCES:
DW187A/A/397/cp
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1 Page: 3 Rev.: 2 l The composition' of the Review Team may be expanded at the discretion of the Review Team Leader. The Review Team shall observe the Operating Crews' performances on each scenario and evalucte their responses in accordance with Section 6 of this procedure. The Operations Engineer will be the Review Team Leader. With the a exception of the Nuclear Trainin5 Instructor, members of the Review Team shall not assist nor interfere with the Operating Crew in any way. ,
5.3 Simulator Operator The Simulator Operator is the Nuclear Training Instructor, who will program the initial conditions and the various f ailure modes for each scenario as described on the Simulator Scenario Form, Attach-ment 1. The scenarios are to be enacted in real-time. The Simula-tor Operator will also act as the interface for the Operating Crew on any communications outside the control room. .
5.4 Video Documentation Each exercise will be recorded on video tape to provide a permanent record.
5.5 Simulator Design Documentation Differences between the simulator panel design and actual plant control room panel design at fuel load shall be documented by the Human Factors Unit.
- 6. EVALUATION An evaluation shall be conducted by the Review Team and Operating Crew for each scenario. The Control Room Validation is not intended to evaluate the abilities of the Operating Crew or its members, but to evaluate the usefulness and adequacy of the operating tools (i.e., control room design and components, PEIs and associated procedures, and the SPDS) in enabling the trained Operating Crew to respond correctly.
Prior to performing the first scenario, the Review Team and Operat- ,
ing Crew should be briefed on the purpose and objectives of the l Control Room Validation. They should become f amiliar with the '
review forms that will be used.
6.1 Review Team Checklist )
( Each member of the Review Team will take notes on the attachment to the Simulator Scenario Form during the exercise, and immediately l
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Page: 4 Rev.: 2 af terwards complete the appropriate checklists. The evaluation by l the Review Team will utilise the three part checklist included as 1 Attachment 2: j
- a. Control Room Evaluation.
' b. Plant Emergency Instruction (PEI) Evaluation. .
It is also important to note problems involving the interaction of crew members. Some indicatora of difficulties are as follows-1
- a. Congested traffic patterns . ,
- b. Uncertainty about areas of responsibility I
- c. Uncertainty about lines of authority l
- d. Difficulties with communications and interaction
- e. Confusion or uncertainty in performing tasks which require more than one operator Insuf ficient work space or an inadequate number of operating f.
tools (e.g., procedures, reference materials, telephones, etc.)
, During the exercises, the Review Team will concentrate on beh'aviors and operational activity that would indicate a potential 3roblem.
The assessment of these observations will be made during Phe de-briefing to determine causality, signigicance, etc.
6.2 Operating Crew Evaluation Checklist Immediately following the enactment of each scenario, each member :
of the Operating Crew will complete an Operating Crew Evaluation Checklist, included as Attachment 3. Responses should be as de-tailed and complete as possible. Part C will be completed only for those scenarios where the SPDS is available.
6.3 Critique Immediately following completion of each exercise, the Operating Crew shall meet with the Review Team for a critique of the scenario :
just completed. The discussion should focus on those problem areas identified in the checklists which are deemed to be most signifi-cant. If the Operating Crew responded incorrectly or too slowly at any time, the reason (s) should be determined and corrective measures recommended. A suggested sequence for conducting the critiquing l follows:
- a. The Operating Crew presents problems and discrepancies identi-fied during the enactment with possible causes and potential )
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- b. The Review Team presents other observations identified during the enactment.
- c. The Operating Crew reviews the identified observations, if ,
applicable, describes possible cause and potential solution for problems.
- d. The Review Team resolves the explanation of possible causes and potential solutions with the Operating Crew.
The Control Room Validation Operations Engineer shall prepare a summary of each exercise critique in accordan::e with Attachment 4.
- 7. RESOLUTION Af ter the evaluation process has been completed, the Review Team will analyze all the evaluation sheets, records of critique sessions, and scenario comments. The video tape may be used to clarify comments during the assessment phase. Human Engineering Observations for the Control Room Design and SPDS will be processed according to Figure 1. Discrepancies in PEI usage shall be noted and resolved in accordance with the PEI Validation Plan of the Procedures Generation Package.
- 8. REPORTING Two reports of the FNPP Control Room Validation will be prepared for the NRC. These reports should include the following:
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- a. PEI Validation Report I
- 1) Brief summary of the validation process.
- 2) Description of scenarios used.,
- 3) List of discrepancies.
- 4) List of procedural changes resolution of discrepancies.
- 5) Report shall be submitted prior to initial criticality.
. b. Control Room Validatio 77oort
- 1) Brief summary of the validation process.
- 2) Description of scenarios used.
- 3) Provide list of Human Engineering Deficiencies, proposed fixes and implementation schedules.
- 4) Report shall be submitted prior to receipt of a Full Power License.
A copy of this procedure and the document detailing the differences
() between the simulator design and the actual plant control room design will be attached to the report, l
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- 9. REFERENCES 9.1 PNPP Procedures Generation Package.
9.2 PNPP Detailed Control Room Design Review Summary Report.
3 9.3 NUREG /CR-3557 CRT Display Evaluation: The Checklist Evaluation of CRT-Cenerated Displays.
9.4 BWR Owners' Group Control Room Survey Workshop (Tulsa, Oklahoma -
October 18-20, 1983).
9.3 INPO 83-047 (NUIAC) Component Verification and System Validation Guideline.
9.6 ALO-1019 Simulator Evaluation of the Boiling Water Reactor Owners' Group (BWROG) Graphics Display System (GDS).
9.7 NSAC/39 Verification and Validation for Safety Parameter Display Sys t ems.
9.8 NSAC/61 Verification and Validation of the Yankee Plant Safety Parameter Display System.
- 10. FIGURES 10.1 Figure 1: DCRDR Validation Review Process to Resolve Control Room Design and SPDS HEOs.
. 11. ATTACHMENTS 11.1 Attachment 1: Simulator Scenario Form 11.2 Attachment 2: Review Team Evaluation Checklist .
11.3 Attachment 3: Operating Crew Evaluation Checklist 11.4 Attachment 4: Scenario Evaluation Summary O
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DCRDR Validation Review Process to Resolve Control Room Design and SPDS HEOs s
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Attachment 1 Simulator Scenario Form t
- PEI - Revision:
Title:
Date:
Purpose:
Scenario
Description:
Initial Plant Conditions:
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Simulator Sequence (to be completed by the simulator supervisor):
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TIME HR: MIN:SEC EVENT / REMARKS / COMMENT
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F 33 9 Rev.: 2 Scenario No. Attachment 2 j Crew No.
Review Team Evaluation Checklist j l
A. Control Room Evaluation i
- 1. Did you observe operators walking to the incorrect area of the ,
control room on the first try? If so, please explain.
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- 2. Did you observe operators looking at an incorrect display or looking in the wrong direction or area on the first try? If so, please explain.
- 3. Did you observe operators having dif ficulty locating a particu-lar control or reaching for an incorrect control on the first try? If so, please explain.
- 4. Did you observe operators setting a control on an incorrect value or moving it in the wrong direction on the first try?
If so, please explain.
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- 5. Did you observe operators failing to observe a key signal? 1 If so, please explain.
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Attachment 2 (Cont.)
- 6. Did you observe operators having to convert displayed values to operational process units? If so, please explain. c S
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- 7. Did you observe operators opening the door of a recorder in order to " read a parametric value? If so, please. explain.
- 8. Did you observe operators having particular difficulty with any particular panel or task? If so, please explain.
- 9. Did you observe operators needing or desiring more controls or displays to respond to this situation? If so, please explain?
- 10. Did you observe operators having dif ficulties reading some displays or reaching / operating some controls? If so, please explain.
1
- 11. Did you observe operators having difficulty locating and operating emergency controls? If so, please explain.
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Page: 11 Rev.: 2 Attachment 2 (Cont.)
- 12. Did you observe any situations where an emergency control could have been inadvertently activated? If so, please explain.
- 13. Did you observe operator (s) having difficulty interpreting or responding to the information presented on the annunciator system? If so, please explain.
- 14. Did you observe operators misinterpreting any information or drawing erroneous conclusions from controls and instrumentation?
If so, please explain.
- 15. Did you observe any congestion problems during this exercise?
If so, please explain.
- 16. Did you observe any uncertainty among the operators as to their area of responsibility? If so, please explain.
- 17. Did you observe any uncertainty among operators as to the lines of authority? If so, please explain.
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Pag 3: 12 Rev.: 2 Attachment 2 (Cont.)
- 18. Did you observe sny problems in executing tasks that require more than one operator to perform? If so, please explain.
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- 19. Did you observe any problems in terms of insufficient work space or an inadequate nerber of operating tools (e.g.,
procedures, reference materials, telephones, etc.) to support the response to this scenario? If so, please explain. j i
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- 20. Did you observe any verbal instructions between operators having to be repeated? If so, please explain.
- 21. Did you observe any verbal instructions between operators not being carried out? If so, please explain.
- 22. Did you observe any problems with operators communicating an insuf ficient amount of information to support the coordination of the activities of the entire crew? If so, please explain.
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- 23. Did you make any other observations about the usefulness and adequacy of the control room and instrumentation in enabling
(. the trained operating crew to respond correctly to this scenario? Please describe.
1.
P 33 13 Rev.: 2 1
l Attachment 2 (Cont.) ;
- 3. Plant Emersency Instruction (FEI) Evaluation
- 1. Did you observe operators selecting the wrong procedure? If so, please explain.
e t
- 2. Did you observe operators selecting too many procedures? If so, please explain.
- 3. Did you observe operators taking an excessive amount of time to read procedures? If so, please explain.
- 4. Did you observe operators re-reading procedures? If so,
'please explain.
- 5. Did you observe operators misunderstanding or having difficulties interpreting procedures. If so, please explain, i
- 6. Did you observe operators not using procedures, when they were available to support a particular operation? If so, please exp la in.
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F: gas 14 Rev.: 2 W Attachment 2 (Cont.)
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- 7. Did you observe operators having difficulty locating key infor-nation in the procedures? If so, please erplain.
- 8. Did you observe operators having dif ficulty interpreting infor-nation in charts, graphs, etc. in the procedures? If so, explain.
s
- 9. Did you observe operators performing procedural steps out of sequence? If so, please explain.
- 10. Did you observe operators neglecting to perform actions or steps detailed in the procedures? 11 so, please explain.
- 11. Did you observe the operators having difficulty obtaining the necessary information specified in the procedures from the installed plant ins trumentation? If so, please explain.
- 12. Did you observe the operators using information or equipment not specified in the procedures to accomplish a task? If so, please explain.
O C.
Pcgs: 15 Rev.: 2 i
Attachment 2 (Cont.)
- 13. Did you observe operators having difficulty in using the procedures in the control room? If so, please explain. ,
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- 14. Did you observe operators having difficulty moving from one procedure to another, as operational conditions required?
If so, please explain.
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- 15. Did you observe operators having difficulty using procedures concurrently? If so, please explain.
- 16. Did you make any other observations about the usefulness and adequacy of the PEls in enabling the trained operating crew to respond correctly to this scenario? Please describe.
C. SPDS Evaluation
- 1. Did you observe operators having difficulty accessing or reading the systen then stationed at the controls? If so, please explain.
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Page: 16 Rev.: 2 Attachment 2 (Cont.)
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- 2. Did you observe operators having to wait an unacceptably long time to re.trieve requested information or displays? If so, please explain, t ,
a o -
i 3. Did you observe operators having difficulty in using the SPDS keyboard and controls? If so, please explain. =
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- 4. Did you observe operators having dif ficulty accessing desired information and displays? If so, please explain.
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- 5. Did you observe operators being confused by or misinterpreting information presented? _If so, please explain. _
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- 6. Did you observe operators having difficulty reading the CRT l j because of glare, inadequate brightness, viewing angle, etc.? ]
If so, please explain. ,
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I 1 l l i j 7. Did you observe operators having to convert or interpret informatior prior to usage? If so, please explain.
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. - , _ . _ _ _ , . . , . . , , , ..,__,,....m., . , _ _ _ . _ - . _ , , , _ . . , ..m_, - , _ _ . . _ . , , , _ _ _ . , , . , , _ _ , , , _ _ , - , . - . _ . _ , , _ _ _ _ . . . - - . .
Page: 17 Rev.: 2 Attachment 2 (Cont.)
- 8. Did you observe any situations where you thought SPDS was of particular value to the operators in responding to this ,
scenario? If so, plasse explain.
4
- 9. Did you observe any situations where you thought SPDS hindered the operators in responding to this scenario? If so, please explain.
j
- 10. Did you make any other observations about the usefulness and adequacy of the SPDS in enabling the trained operating crew
- to respond correctly to this scenario? Please describe.
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i P; gar 18 Rev.: 2 Scenario No. Attachment 3 Crew No. j i
I operatina crew Evaluation checklist
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A. Control Roou Evaluation i
I
- 1. What problems arose due to the physical layout of the control i room (panel locations, storage space, work space, etc.)? ,
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- 2. Identify any problems with locating or operating components j ,
due to inadequate labeling, location, etc.
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- 3. Identify any components that were confusing or hard to read or operate due to design or location.
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Fage: 19 Rev.: 2 Attachment 3 (Cout.)
- 4. Could you obtain the necessary information specified in the '
procedures from the installed plant instrumentation? Identify any problem areas.
t
- 5. Is there a particular panel which you consider more difficult or confusing to operate than the others?
- 6. Does the control room lack any controls or displays needed in I your response to this situation?
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Page: 20 Rev.: 2 l
Attachment 3 (Cont.)
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l 7. How ef fective do you consider the annunciator system to be in I conveying important information to you?
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- 8. List any problems locating or using procedures, instructions or reference asterial. {
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- 9. Identify any problems with communication systems. "
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Pagos 21 j Rev.: 2 ,
Attachment 3 (Cont.)
- 10. Identify any problems you encountered in responding to this '
situation Wich could have been averted through improved control room design.
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i B. Plant Emergency Instruction (PEI) Evaluation !
l 1. Identify any instances where the PEI did not supply sufficient information to complete a given task. ;
- 2. Are the alternatives adequately described at each decision I point?
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w-u ----- ._-,- _ , _--_ - - - - . - - - - . . , ,,y, .__,,--__- y_,, _ ---, ,,.m. - - - --_ -y -._--- , _ -g-,.-y -, ,-,.-,.-e.,.-.,.w--wg-,ww-,
l .
Page: 22 Rev.: 2 Attachment 3 (Cont.)
- 3. Are the labeling, abbreviations, and location information as provided in the PEI suf ficient to enable the operator to find the needed equipment?
8
- 4. Is the PEI missing information needed to annage the emergency 1
condition?
- 5. Are the contingency actions suf ficient to address the symptoms?
G
l Fage: 23 r Rev.: 2 Attachment 3 (Cont.)
- 6. Are the titles and numbers sufficiently descriptive to enable the operator to find referenced and branched instructions?
- 7. Is the PEI easy to read? .
- 8. Were tables and charts easy to read with accuracy? Cite any '
problems.
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- 9. Are the figures and tables easy to read with accuracy?
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Page: 24 Rev.: 2 Attachment 3 (Cont.)
- 10. Can the values on figures and charts be easily determined?
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- 11. Are CAUTION and NOTE statements readily understandable?
- 12. Are the FEI steps readily understandable?
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- 13. Can the actions specified in the instruction be perforned in I
the designated sequence?
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Page: 25 Esv.: 2 Attachment 4 (Cont.) 1 l
- 14. Are there alternate success paths that are not includs.d in the Fels?
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i 15. Can the information from the plant instrumentation be obtained
- as specified by the FEI?
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- 16. Are the plant symptoms specified by the PEI adequate to enable the operator to select the applicable FEIT r
l 1.-_--. _
7 Page: 26 Rev.: 2 Attachment 3 (Cont.)
- 17. Are the FEI entry conditions appropriate for the plant symptoms '
displayed to the operator?
- 18. Is information or equipment not specified in the FEI required to accomplish the task?
- 19. Do the plant responses agree with the FEI basis?
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- 20. Are the instrument readings and tolerances stated in the PEI consistent with the instrument values displayed on the instru-ments?
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Faga 27 Eev.: 2 Attachment 3 (Cont.)
- 21. Is the FEI physically compatible with the work situation (too j bulky to hold, binding would not allow them to lay flat in worir, space, no place to lay the Fels down to use)?
I
- 22. If time intervals are specified, can the action steps be per-formed on the plant within or at the designated time inter-vals?
I
- 23. Can the action steps be performed by the operating shif t?
1 I i
-.% _.. - - - --_g -- .,,,.-,-,,p..,---,-_-.-...e -
,. _,-- , ,- ,,, , - - - ..e+--,,m. -_..._ ._. _ ...-., .---._,.--e -- 3 . - . - - _-e
Page: 28 Rev.: 2 Attachment 3 (Cont.) !
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- 24. If specific actions are assigned to individual shif t personnel, does the PEI adequately aid in the coordination of actions ;
among shif t personnel where necessary? l
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- 25. Can the operating shif t follow the designated action step sequences?
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- 26. Can the particular steps or sets of steps be readily located when required? l l
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- -- - ____,_----,-~,,__._.-------.,_.-__-.__,.----._,___y_
F; gat 29 Rev.: 2 Attachment 3 (Cont.)
- 27. Can the exit point be returned to without omitting steps when required? ,
- 28. Can instruction branches be entered at the correct point?
- 29. Are PEI exit points specified adequately?
- 30. Identify any problems you encountered responding to this situation which could have been averted through improvements to the PEI.
Pages 30 Rev.: 2 Attachment 3 (Cont.)
- 31. List any specific reconnendations for improvement to the FEI.
5 e i C. SPDS Review (To be completed only if SPDS we avsilable.)
- 1. Did you encounter any difficulty operating the system? Cite speciffe problems.
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- 2. Please comment on the amount of information available. (Any information that was not available that should be, too such redundant information, etc.).
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P 33: 31 Rev.: 2 Attachment 3 (Cont.)
- 3. Were there any problems understanding the information pre-sented? Cite specific instances. ;
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- 4. Were there any problems with the way information was organized?
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- 5. Could you access inform .cion as quickly as you needed it?
Cite any delays.
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- 6. Were there any inconsistencies between SPDS and the control '
panels regarding color codes, abbreviations, acronyms, equip-ment terminology and symbology?
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Pag 3: 32 l Rev.: 2 l l
Attachment 3 (Cout.) i
- 6. Were you able to use all information as it as presented?
Identify cases dere conversion or interpolation was required.
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- 7. How much did you use the SPDS in responding to this situation?
Identify those areas in sich it was the most helpful.
S. Identify any problems you encountered in responding to this ,
situation sich could have been averted through improvements to the SPDS.
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Pag 3: 33 Rev. 2 Attachment 4 Scenario Evaluation Summary Scenario No. .
s Crew Ilo.
Significant Human Engineering Observations and Discrepancies v
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ATTACHMENT B NOVEMBER, 1985 AND JANUARY, 1986 CONTROL ROOM VALIDATION SCENARIOS I'
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NOVEMBER, 1985' EXERCISE SCENARIOS I
NUMBER TITLE 1 Small Leak in Drywell i t
- 2 Small Increasing LOCA in Drywell Loss of ALL Injection Systems i 3 Small Break.LOCA in Drywell, Loss !
of High Pressure Make-up, Loss ;
of Drywell Cooling, Loss of RPV i Level Indication 1 a 4 Small Break LObA in Drywell, Failure 1 of Four Rods to Insert, Loss of I High Pressure Make-up, Loss of 1 Drywell Cooling, Loss of RPV Level Indication
- 5 Startup in Progress, Inadvertent HPCS i Initiation, Power Transient with Fuel Damage, MSIV Hi Rad Isolation Small Leak in Drywell, RCIC Initiation, and Steam Line Break, Rad Release Out of Containment 6 MSIV Isolation with Stuck Open SRV, Failure to Scram (ARI ok) j 7 SRVs Stuck Open, Failure to Scram 8 SRVs Stuck Open, Failure to Scram, i
Crack in LPCI-B Injection Line,
- Containment Pressurization, l Flooding Required
} JANUARY, 1986, EXERCISE SCENARIOS i
^
l MSIV Isolation, Rods Out, LOCA
- 2 Low Vacuum Main Steam Line Isolation, !
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SCENARIO: I SMALL LEAK IN DRYWELL h PEIs EVALUATED: 1 (RPV CONTROL) 2 (SUPPRESSION POOL TEMPERATURE) 3 (DRYWELL TEPPERATutE) -
4 (CONTAINMENT TEMPERATURE) -
5 (DRYWELL PRESSURE) - -
6 (SUPPRESSION POOL LEVEL) s INITIAL CONDITIONS: 100 PERCENT POWER, BEGINNING OF CYCLE i
RCIC out of service while replacing turbine exhaust rupture disks (Simulator malfunction 511 - RCIC turbine trip -
is active). NCC Pump C is out of service (Simulator Malfunction 655)
TIME EVENT - REMARKS - EVALUATOR COMMENT )
EVENT: A small.recirc system suction pipe crack occurs, !
0000 causing a buildup in drywell pressure and temperature I (Simulator malfunction 704). Reactor scrams on high drywell pressure unless operator shuts down the reactor ,
before hand. l
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', REMARKS: Operator should anticipate the reactor scram j and manually scram the reactor.
Operator should i recognize leakage as pressure boundary leakege, which is an Unusual Event per EPI-Al (Emergency Action Levels) . j Operator should enter PEI-3 (Drywell Temperature Control) and PEI-5 (Drywell-Pressure Control). ,
1 COMMENTS:
l
- l 0004 EVENT: LOCA initiation signals and LOCA isolation signals on high drywell pressure occur. NCC and instrument air to the drywell and containment are lost.
All divisions of ECCS initiate, with HPCS injecting into the vessel.
REMARKS: Operator enters PEI-1 power, level, and
( , pressure contrcl sections. Pressure will still be
- O) controlled by the Steam Bypass System, and Feedwater through the Motor Feed Pump will still be available in L
I excess of leakage from vessel. Operator's main concern
( will be with controlling the rising temperatures in the containment and drywell, as well as drywell pressure.
O Ta a to- tic ias-=tioa =< *c> * = a uau 1 ev at a r EPI-A1. Operator may estimate leakage in excess of 50 gpe (f rom information f our;d on Leakage Detection System panel P642), in which case an Alert should be called per EPI-A1.
a COMMENT: i l
1 0010 EVENT: Drywell temperature and pressure continue to ,
rise. Containment tempera.ture begins to rise slowly due I to loss of Containment Vessel Cooling. I REMARKS: Operator uses PEI-3 and PEI-5 to restore !
Drywell Cooling System operation. This will require l bypassing the stub bus LOCA trips, re-energizing XH11 C, and/or XH-12 stub busses, and restarting NCC pump A
)
6 and/or B (Pump C is out of service).
NCC LOCA isolation signals will have to be bypassed.
Additionally, the j COMMENTS:
(
0015 EVENT: RPV pressure and temperature remain fairly constant, controlling through Bypass Valves. Drywell ,
temperature and pressure continue to rise (Drywell Temperature is still less than 212 F).
REMARKS: Operator should consider use of backup hydrogen purge line to the Annulus Exhaust Gas Treatment System to bleed off drywell pressure. This method becomes un-available if drywell temperature exceeds 212 F. Containment temperature is not being controlled, due k'- to loss of Containment Chill Weter System because of Balance of Plant Isolation which cannot be reset as long
() )
as drywell pressure is above 1.68 psig.
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COMMENTS:
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,I 0025 EVENT: Drywell temperature increases above 212 F. !
(I.mcrease simulator malfunction 704 if necessary).
REMARKS: Operator must observe drywell temperature lin.itation of 330 F. If temperature cannot be maintained below this point, then emergency depressurization is required.
COMMENT: ,
0030 EVENT: Drywell temperature continues increase toward 330 F (Increase simulator malfunction 704). Suppression Pool level will increase past 18.5 feet. Suppression -
Pool temperature is increasing, but is still below 95 F.
REMARKS: Operator will notice need f or increasing amounts of make up water, which may be coming from the motor feed pump or HPCS. If operatoe decides to emergency depressurize, operator should deepressurize using the Steam Bypass Valves to the main condenser.
Operator will enter PEI-6 (Suppression Pool Level Control), but will not be able to use Suppression Pool Cleanup System to lower level due to SPCU LDCA isolation signals. Operator may enter PEI-2 (Suppression Pool Temperature Control) and divert a loop of RHR to the Suppression Pool Cooling Mode of operation if 10 minutes have elapsed since the RHR LOCA initiation signal has occurred.
COMMENTS:
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0035 EVENT: Manual depressurization in progress.
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REMARKS: During depressurization process, vessel a temperature will decrease rapidly. As heat input to the drywell decreases, the restored drywell cooling system .
will begin to lower drywell temperature and pressure. i COMMENTS:
e a 0040 EVENT: END OF SCENARIO REMARKS: Scenario my be terminated at any point after the vessel has been depressurized.
n.
COMMENTS:
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SCENARIO: 2 SMALL INCREASING LOCA IN DRYWELL, LOSS OF ALL INJECTION SYSTEMS.
1 PEIs EVALUATED 1 (RPV CONTROL) '
1 ATT. ~ 1 .(LEVEL RESTORATION) -
-, . 1 ATT. 2 '(EPERGENCY DEPRESSURIZATION)
-1 ATT. 3 (STEAM COOLING)
- 2 (SUPPRESSION POOL TEMPERATURE)
- 3 (DRYWELL TEMPERATURE) #
4 (CONTAINMENT TEMPERATURE) 5 (DRYWELL PRESSURE) 6 (SUPPRESSION POOL LEVEL) l INITIAL CONDITIONS: 100 PERCENT POWER, BOC (IC-12),
MOTOR FEED PUMP FLOW CONTROL VALVE SIGNAL FAILURE PRESENT (MF 606)
HPCS OUT OF SERVICE FOR BREAKER REPAIR.
(MF 517)
NUCLEAR CLOSED COOLING PUMP C OUT OF SERVICE (MF 655)
TIME EVENT - REMARKS - EVALUATOR COMMENT 0000 EVENT: A small crack developes in a recirc loop suction line (MF 704). Drywell pressure begins to increase.
I REMARKS: Operator will diagnose problem from drywell pressure high and drywell cooler drain flow high alarms.
Operator should plan for fast reactor shutdown as drywell pressure approaches the scram setpoint Of 1.68.
psig.
COMMENTS:
0005 EVENT: Drywell pressure reaches 1.68 psig. Heactor scrams if not already shutdown by operator. All ECCS receive start signals but LPCS does not start (MF 518),
HPCS is out of service. Stub busses XH-11, XH-12 trip causing a loss of CRDH and NCC pumps A and B. Balance of Plant and RHR LOCA isolations take place. Drywell
) temperature is high. Probable trip of feedpumps occurs on high RPV water level.
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REMARKS: Operator should enter PEI-1 (power, pressure, and level control), PEI-3, PEI-4, and PEI-5. Since there is no immediate problem with water level, the operator will place highest priority on the restoration
- ' Any of NCC and air supply to containment and drywell.
attempts to restore CKDH will be prevented for this scenario by a spuriods pump trip (MF 497 and 498).
Operator should evaluate leakage and determine that s leakage is greater than 50 gpm, requiring the declaration of a ALERT per EPI-A1 (Emergency Action Levels). -
COMMENTS:
I 0010 EVENT: Water level begins to decrease as leak gets larger. Suppression Pool temperature and level are both increasing.
REMARKS: Operator should restore turbine feedwater pumps to make up water to the RPV, or use motor feedpump with a plant operator controlling the flow control
) valves with the manual handwheels as directed by the coatrol room. If operator has initiated RCIC to provide makeup, this will trip the turbine feedpumps. Oberator should give consideration to lowering reactor pressure by using the Bypass Valves to blowdown to the Main Condenser. Operator should take steps to ensure that condenser vacuum is maintained. Operator should contin,ue to attempt to control the degrading conditions of the containment systems with PEI-2, 3, 4, 5, and 6. j Operator will not be able to lower suppression pool l 1evel due to the LOCA isolation of the Suppression Pool Cleanup system. Oper ator may decide to override the '
automatic dump of the Suppression Pool Makeup system '
which will occur 30 aninutes af ter the high drywell pressure condition occurred.
COMMENTS:
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l 0015 EVENT: Reactor Feed Booster Pumps trip on spurious hot surge tank low level (MF 841 and 601), causing a loss of
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feedpump operation.
REMARKS: Wat'er level will drop at a moderate rate.
Operator will start'RCIC system to provide make up to a the' vessel.
COMMENTS:
0018 EVENT: After several minutes of use, the RCIC flow controller fails low (MF 510) and prevents further l injection in to the RPV.
. REMARKS: Operator should enter PEI-1 Att. 1 for level restoration. Since only low pressure injection systems are available. the operator thould anticipate the need for emergency depressurization and proceed to rapidly depressurize using the Bypass Valves to the Main Condenser.
COMMENT: .
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1 0025 EVENT: A ground fault on emergency bus EH-12 (MF 676) i I
results in a loss of division 2 ECCS (RHR-B, RHR-C).
REMARKS: The operator is lef t with only one injection system (RHR-A), which is adequate f or the operator to proceed with depressurization. If water level has reached 16.5 inches by this point, the MSIVs will isolate. Emergency Depressurization using the SRVs per
, ( PEI-1 Att. 2 will be required.
} COMMENT:
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0030 EVENT: RHR-A pump trips (MF 514). Any attempts to use SLC as an injection system fail as both SLC systems are found not to be operable (MF 853 and 854).
REMARKS: Operator will likely call a Site Area Emergency per EPI-Al based on loss of vessel make up capability. Operator will attempt to line up alternate injection systems while proceeding to PEI-1 Att. 3 (Steam Cooling).
COMMENTS:
0035 EVENT: Maintenance informs control room that LPCS is available for injection (remove MF 518) .
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REMARKS: Operator will commence injecti on with LPCS.
Water level will begin to increase. Operator will exit PEI-1 Att.1 and return to the level control portion of PEl-1.
COMMEN'TS:
0040 EVENT: Recirc loop suction line totally ruptures, leak increases resulting in decreasing water level.
) REMARK: Operator re-enters PEI-1 Att. 1. and verifies LPCS flow rate greater than 6250 and RPV pressure less
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than 330 psig. If operator had managed to line up any injection sourens external to containment, then these systems would be stopped.
COMMENT:
t i
0045 EVENT: Maintenance informs control room that RHR-A is available (remove MF 514), that HPCS is available (remove MF 517), and that the reactor feed booster pumps are available (remove MF 841) .
RENARKS: Operator will' inject with HPCS and/or RHR-A, but not with RFBPs, since the RFBPs are sources of make up <?xternal t'o containment.
COMMENTS:
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0055 EVENT: Water level has increased to above 16.5 inches.
REMARKS: Operator exits to PEl-1 level control section.
Operator continues to pursue solutions for drywell temperature and pressure control. Operator should line up RHR-A to Suppression Pool Cooling per PEI-2.
COMMENTS-t e
) 0100 EVENT: END OF SCENARIO
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O COMMENTS:
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i C SCENARIO: 3 SMALL BREAK LOCA IN DRYWELL.
'oss oe assa easssuas aa*=-ue.
O LOSS OF DRYWELL COOLING.
LOSS OF RPV LEVEL INDICATION. -
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PEIs EVALUATED: 1 (RPV CONTROL) .
. 1 ATT. 1 (LEVEL RESTORATION) 1 ATT. 2 (RPV DEPRESSURIZATION) 1 ATT. 4 (RPV FLOODING) l 2 (SUPPRESSION POOL TEMPERATURE CONTROL) l 3 (DRYWELL TEMPERATURE CONTROL) 5 (CONTAINMENT PRESSURE CONTROL) . .
6 (SUPPRESSION POOL LEVEL CONTROL)
INITIAL CONDITIONS: 100 PERCENT POWER, BOC (IC-12),
HPCS OUT OF SERVICE FOR BREAKER REPAIR (MF 517) .
TIfE EVENT - REMARKS - EVALUATOR COMMENT * ,
0000 EVENT: Instructor inserts Recirc Loop Crack (MF 704).
Ci '
A Drywell Cooler Drain Flow High alarm is received on P601. A moderate rate of increase in drywell pressure O is observed.
REMARKS: Operator will diagnose the existence of a
. leak. In the absence of other indicators, he should assume that the leakage is pressure boundary leakage and declare an Unusual Event per EPI-Al (Emergency Actiod I
Levels). Operator should take actions to situt d'own thq reactor as drywell pressure approaches the scram point ,
1 of 1,68 psig.
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1 COMMENT: )
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i 0004 EVENT: Reacter is scrammed by high drywell pressure or e by operator action. Main Turbine trips on reverse power. High drywell pressure causes initiation of Division 1 3 2, and 3 ECCS (but HPCS is out of service).
O seaare ioa eoen a = a s v te- a a ei er <so ia t >
begins counting. Suppression Pool High alarm may be l
received as drywell pressure pushes down on weir wall l
- l. .
surface. Drywell floor drain sumps indicate greater 9 than 50 gpa leakage. A Balance of Plant Isolation also occurs on High Drywell Pressure. The stub busses XH-11 and XH-12 trip when Division 1 and 2 ECCS activate. ~
causing a loss of Nuclear Closed Cooling supply to the drywell
- coolers, and a loss of. air supply to the
, containment and drywell. Control rod hydraulic pumps, c also supplied from the stub busses, are lost.
REMARKS: High drywell pressure requires entry into PEI-1. Operator will ensure that all rods have inserted past notch 02, that RPV pressure is being controlled ,
below 1033 psig by the Steam Bypass and Pressure Control System, and that RPV water level is under manual control using the'feedwater system. Operator may begin 100 F/HR cooldown and depressuriiation using the Bypass Valves in order to reduce the leak rate. High drywell pressure requires entry into PEI-5. Operator will attempt to restore drywell cooling per PEI-5. High drywell-
- temperature requires entry into PEI-3, which will .n1so direct restoration of drywell cooling. Operator should select one stub bus f or re-energization f rom its 9
associated ECCS divisional bus, after ensuring that ECCS
- from that division are not required for maintenance of adequate core cooling. Operator may use authority of )
()
( PAP-0205 (Overriding Safety Systems) Eto prevent automatic SPMU pool dump. Operator should declare an
()6 Alert per EPI-Al when floor drain sumps in drywell and contaiment indicated greater than 50 gpm of leakage.
~
COMMENTS
(
A 0010 EVENT: Operator is unable to open inboard NCC isolation J valves (Instructor f ails to bypass LOCA isolation signal 1 when requested by operator). z REMARK: Sustained loss of drywell cooling will cause continued drywell heatup. Isolation valves are not accessible for local operation. Containment pressure will begin to rise slowly as drywell pressure rises.
- Operator will use PEI-5 to control both containment and k drywell pressure, observing the requirement to use containment spray if containment pressure approaches the O- Pressure Suppression Limit.
COMMENT:
_ _ _ - - _ _ _ _ _ , _ ~ _ _ _ _ _ _ _ __ _ _ _ . . _ _ _ _ _ _ _ - _ _ .
l
)
1 s ..
.f 0015 EVENT: Failure of Hot Surge Tank level interlock (MF 841) causes a trip of any turbine feedpump that is operating. Motor feedpump, if operating, cavitates and trips 30 seconds later (MF 601).
REMARKS: Operator will investigate loss of feedsystem.
I Operator will notice slow decrease in RPV water level.
Operator may halt cooldown until makeup water can be supplied to the vessel at high pressure. Reactor Core- ~
Isolation Cooling (RCIC) will be started for this purpose.
COMMENT: l I
0018 EVENT: RCIC flow control system malfunctions (MN 510).
- and does not allow RCIC pump to deliver water to the RPV.
1 REMARKS: Operator will pursue two options - continue to depressurize with Bypass Valves until Low Pressure Core Spray injects at aprroximately 450 psig, or enter Attachment 1 in preparation for possible emergency depressurization so low pressure make up systems can inject. j 1
COMMENTS:
_ . = ux-
~
l
.,.- -.,,.y ----v
0025 EVENT: Size of leak increases (Instructor increases MF 704).
REMARKS: Operator observes quickly decreasing water
. level, but RPV pressure drops much more slowly. This will indicate definite need to enter PEI-1. attachment 1.
Until RPV level 1 (16.5 inches) is reached, operatt' will use maximum Bypass Valve capacity to direct steam flow to the main condenser, as allowed in PEI-1 pressure control section. ;
. i COMMENT:
0030 EVENT: RPV level'1 is reached, resulting in MSIV
- isolation and slow repressurization of RPV.
REMARKS: Operator is required to enter Emergency Depressurization (PEl-1, Att. 2) to continue depressurization. This will require eight ADS SRVs to be opened by the operator. Low pressure ECCS will inject.
COMMENT: (
4 0035 EVENT: RPV water level begins to increase after dropping below top of active fuel. Suppression pool temperature increases as result of emergency depressurization. Pool level may be either high (as a result of depressurization or automatic pool dump) or low (as a result of ECCS drawdown).
REMARKS: Operator will leave attachment 1 and begin return to normal water level range of 183 to 218 inches, while SRVs remain open to continue depressurization.
l Operator must enter PEI-2 to control suppression pool temperature if 95 F is reached. Operator must enter PEI-6 to control suppression pool level if l evel is out
l l
l Of band.
COMr*ENT:
l ea se
)
l 0040 EVENT: Drywell temperature has increased to approximately 270 F. RPV pressure has decreased to zero. Water level indications become erratic (Instructor must. verbally present this indication problem to the operator).
REMARKS: 'The possibility of RPV water level instrumentation ref erence leg flashing becomes a concern. The operator will enter attachment 4 (RPV flooding). This will require injecting to the vessel until RPV pressure is 130 psig above containment pressure. Injection continues until drywell temperature i is reduced below 212 F.
COMMENT:
)
i l
1 L
4 0055 EVENT: NCC inboard isolation valves are returned to service and may be opened.
REMARKS: Operator should immediately restore drywall cooling per PEI-3 and/or PEI-5.
COMMENT: l
C 0100 EVENT: Drywell temperature decrease below 212 F.
O REMARKS: Operator terminates injection for.the time allowad by the Maximum Core Uncovery Time L:imit in Att.
4 to observe developement of on-scale RPV level '
indication (on Shutdown Range).
COMMENT: ,,
I I
r I
0110 EVENT: On-scale indication observed.
)
REMARKS: Level indication will~ allow the operator to exit PEI-1. Operator continues to use PEI-3 and PEI-5 to reduce drywell temperaturn and pressure. Operator uses PEI-2 to reduce suppression pool temperature and PEI-6 to reduce wppression pool level.
C? COMMENT:
O .
1 L 0115 EVENT: End of scenario.
I COMMENT:
i O
.=
w-. -,.-.,.w. ,, ,.m.yg , , ,, , , , , ,, , _ _ _ _ , __
s SCENARIO: 4 SMALL BREAK LOCA IN DRYWELL, g FAILURE OF FOUR RODS TO INSERT, LOSS OF HXGH PRESSURE MAKE-UP,
, LOSS OF DRYWELL COOLING, ;
LOSS OF RPV LEVEL INDICATION. !
l i
PEIs EVbLUATED: 1 (RPV CONTRCL) [
1 ATT. 2 (RPV DEPRESSURIZATION) l 1 ATT. 4 (RPV FLOODING) ;
i 1 ATT. 5 (POWER CONTROL WITH LEVEL) 2 (SUPPRESSION POOL TEMPERATURE CONTROL) I 3 (DRYWELL TEMPERATURE CONTRGL) l
- 5 (CONTAINMENT PRESSURE CONTROL) )
6 (SUPPRESSION POOL LEVEL CONTROL) l INITIkL CONDITIONS: 100 PERCENT POWER, BOC (IC-12).
HPCS OUT OF SERVICE FOR BREAKER REPAIR (MF 517)
],
)
TIME EVENT - REMARKS - EVALUATOR COMMENT l l
0000 EVENT: Instructor inserts Recire Loop Crack (MF 704).
C.- A Drywell Cooler Drain Flow High alarm is received on O e601. a moderate rate of increase in dev-et1 aressure is observed.
REMARKS: Operator will diagnose the existence of a ,
leak. In the absence of other indicators. he should assume that the leakage is pressure boundary leakage'and declare an Unusual Event per EPI-Al (Emergency Action .
3 Levels). Operator should take actions to shut down the l reactor as drywell pressure approaches the scram point l of 1.68 psig. '
COMMENT:
J 0005 EVENT: Reactor is scrammed by high drywell pressure or by operator action. Four Control Rods do not insert (MF
( 424, 426, 427, 429). Main Turbine trips on reverse power. High drywell pressure causes initiation of Division 1, 2, and 3 ECCS (but HPCS is out of service) .
LPCS fails to start (PF 518) . Suppression Pool Makeup
Cyntes c. ump timer ("30 cinuto) begino counting.
Suppression Poc1 Level High alarm may be received as C drywell pressure pushes down on weir wall surface.
Drywell floor drain sumps indicate greater than 50 gpm O leakage. A Balance of Plant Isolation also occurs on High Drywell Pressure. The stub busses XH-11 and XH-12 trip when Division 1 and 2 ECCS activate, ecusing a loss of Nuclear Closed Cooling supply to t5e drywell coolers, and a loss of Mir supply to the containment and drywell. <
Control rod hydraulic pumps, also supplied from the stub busses, are lost.
REMARKS: High drywell pressure rey. sires entry into l PEI-1. Operator will observe that four rods have NOT inserted past notch 02, that RPV pressure is being (
controlled below 1033 psig by the Steam Bypass and Pressure Control System, and that RPV water level is '
under manual centrol using the feedwater system.
Operator should notice upon entry into level control section of PEI-1 that LPCS has failed to start.
Operator will then exit level control and enter Att. 5 to control water level at 183 to 218 inches. Operator will use PEI-1 power control section to attempt further ,
methods of Anserting stuck rods. Operator may begin 100 F/HR cooldoven and depress.trization using the Bypass
- Valves in order to reduce the leak rate. High drywell pressure requires entry into PEI-5. Operator will h' '
attempt to restore drywell cooling per PEI-5. High drywell t'.emperature requires entry into PEI-3, which ,
O -iii 2 = air ci r teratiaa =+ dev 12 ===ria.-
Operator should select one stub bus for re-energization from its associated ECCS divisional bus, after ensuring that ECCS from that division are not required for maintenance of adequate core cooling. NCC, CRDj and ,
contaiment air should be restored per PEI-1. Operator may use authority of PAP-0205 '(Overriding Saf ety Systems) to prevent automatic SPMU pool dump. Operator should declare an Alert per EDI-A1 when floor drain sumps,in drywell and containment indicate greater than
- . 50 gpm of leakage.
COMMENT:
i l
i EVENT: Operator is unable to open inboard NCC isolation
( 0010 valves (Instructor fails to bypass LOCA isolation signal when requested by operator).
l . . _ . _ . . _ .
l
, REMARK: Sustained loss of drywell cooling will causz r:antinued drywell heatup. Isolation valves are not accessible for local operation. Containment pressure will begin to rise slowly as drywell pressure rises,.
Operator will use PEI-5 to control both containment and drywell pressure, observing the requirement to use
, containment spray if containment pressure approaches "
the i Pressure Suppression Limit.
' COMMENT: I l
1 l
l 1 EVENT: Failure of Hot Surge Tank level interlock (MF 0015 i 841) causes a trip of any turbine feedpump that is ,
operating. Motor feedpump, if operating, cavitates and trips 30 seconds later (MF 601).
l
- g. ' REMARKS: Operator will investigate loss of feedsystem.
Operator will notice slow decrease in RPV water level.
Operator may halt cooldown until makeup water can be supplied to the vessel at high pressure. Reactor Core Isolation Cooling (RCIC) will be started for this purpose.
COMMENT: .
j 1
j 0019 EVENT: RCIC flow control system malfunctions (MF 510)
) and does not allaw RCIC pump to deliver water to the RPV.
I REMARKS: Operater will proceed to attempt to maintain level above 16.5 inches per PEI-1 Att. 5 using systems that inject outside the core shroud. Except for CRDH
, flow to vessel, all such means are unavailable.
l ( COMMENTS:
l
)
I
s 0020 EVENT: Size of leak increases (Instructor increases MF 704).
REMARKS: Operator observes quickly decreasing water level. It should be apparent to the operator that he is not able to maintain level above 16.5 inches. Att. 5 requires the operator to override the injection valves shut on all ECCS, along with concurrent emergency depressurization. ,
COMMENT:
8 0025 EVENT: Eight ADS SRVs are opened by operator. RPV level 1 is reached, resulting in MSIV isolation.
REMARKS: Operator uses Att. 2 to emergency depressurize. Operator allows depressurization until the Minumum Alternate RPV Flooding Prescure f or eight open SRVs is reached (90 psig).
COMMENT:
0030 EVENT: RPV pressure drops below 90 psig. RPV level is below top of active fuel. l
(.
REMARKS: Operator is directed by Att. 5 to commence 4 injection at this pressure using Motor Feedpump or l
I l
Feedwater Booster Pumps (neither are available) or CRDH (which is not sufficient). Operator is then directed to bypass the Shutdown Cooling Isolation signal to the RHR A/B Shutdown Cooling Return to Feedheader Valve E12-F053A/B and to start the associated RHR pump. This will allow the operat~or to throttle makeup to the vessel, and to direct that makeup outside the core shroud.
COMMENT:
s l
0035 EVENT: RPV water level begins to increase after dropping below top of active fuel. Suppression pool temperature increases as result of emergency depressurization. Pool level may be either high (as a result of depressurization or automatic pool dump) or low (as a result of ECCS drawdown).
REMARKS: Operator will follow attachment 5 and begin return to a water level range of 16.5 to 30 inches, while SRVs remain open to continue depressurization.
Operator must maintain this level until all control rods are inserted. Operator must enter PEI-2 to control suppression pool temperature if 95 F is reached.
Operator must enter PEI-6 to control suppression ' pool level if level is out of band. -
COMMENT:
0040 EVENT: Drywell temperature has increased to approximately 270 F. RPV pressure has decreased to zero. Water level indications become erratic (Instructor must verbally present this indication problem to the operator).
REMARKS: The possibility of RPV water level instrumentation reference leg flashing becomes a
() concern. The operator will exit Att. 5 and enter
Attachment 4 (RPV Flooding). This will require the identical injection lineup as was being used in Att. 5, except that injection &must continue to the vessel to maintain RPV pressure just above the Minimue Alternate RPV Flood Pressure of 90 psig. This must be maintained
- until all control rods are inserted.
COMMENT: .
s i
0050 EVENT: Operator is informed that all rods have inserted, if not already inserted by operator at this point.
REMARKS: Operator will proceed per Attachment 4 to increase injection, using normal ECCS flow paths, until RPV pressure is 130 psig above containment pressure.
Injection continues until urywell temperature is reduced below 212 F.
i COMMENT:
0100 EVENT:' NCC inboard isolation valves are returned to service and may be opened.
REMARKS: Operator should immediately restore.drywell cooling per PEI-3 and/or PEI-5. .
COMMENT:
C O
0110 EVENTS Drywell temperature decrease below 212 F.
REMARKS: Operator terminates injection for the time allowed by the Maximum Core Uncovery Time Limit in Att.
4 to observe developement of on-scale RPV level j indication (on Shutdown Range). '
COMMENT:
b i
l 0115 EVENT: On-scale indication observed.
REMARKS: Level indication will allow the operator to exit PEI-1. Operator continues to use PEI-3 and PEI-5 to reduce drywell temperature and pressure. Operator uses PEI-2 to reduce suppression pool temperature and 1 PEI-6 to reduce suppression pool level.
) COMiiENT:
0120 EVENT: End of scenario.
COMMENT:
f O
C SCENARIOS 5 STARTUP IN PROGRESS ,
O- INADVERTENT WCS INITIATION POWER TRANSIENT WITH FUkL DAMAGE MSIV HI RAD ISOLATION ,
, SMALL LEAK IN DRYWELL RCIC INITIATION / STEAM LINE BREAK -
RAD RELEASE OUT OF CONTAIlfrENT PEIs EVALUATEDs 1 (RPV CONTROL) 3 (DRYWELL TEMPERATURE) ,
5 (DRYWELL PRESSURE) 7 (RADIOLOGICAL RELEASE)
INITIAL CONDITIONS: REACTOR STARTUP IN PROGRESS-RPV PRESSURE = 920 PSIG REACTOR POWER = 1 PERCENT BEGINNING OF CYCLE (IC-7)
MOTOR FEED PUPP OOC (M 601)
TIME -
EVENT - REMARKS - EVALUATOR COMNCNT 0000 EVENT: While reactor power is in the intermediate range, HPCS inadvertently initiates (M 848) and injects into the O i- c ==iaa 569) on IRM high flux trips.
r aid a=~ r ci #a r =*=r ==r"- <a HPCS pump then trips on overcurrent (M 517).
REMARKS: HPCS injection is an Unusual Event per EPI-A1 (Emergency Action Levels). Operator should investigate HPCS actuation and subsequent trip, Mile carryiiig out .
the actions of ONI-C71 for a reactor scram.
COMMENTS:
b 0003 EVENT: MSIVs isolate on Steam Line High Radiation (M 566). Subscquently, a small break occurs in the drywell (M 704), causing drywell temper ature and pressure to increase. ,
O REMARKS: Operator has lost all high pressure RPV make-up capacity except for RCIC, which he should anticipate saanually initiating to provide make-up. Operator enters t..
- - ,.m--__.,--~-. - - - . . ._--_,.----w ww, ,,,, g-n -,w--q-wwy--.
.T 2'; ' - T NT ~ UM " ' " <- r ".- - n- -N W t eK- v biffs'; Cc c.h?PEI .1',because of MSIV isolation,'PEI-3 becausca e
O g temperature increase, and PEZ-5 because of drywell of drywe pressure increase.
l COMMENTSs l
l 1
0005 EVENTS Operator initiates RCIC. Within five minutes I several annunciators- (M 1007, 1245, 1256, 1190) are recieved on P601 and P680 indicating a steam line crack j in the steam tunnel. Additional annunciators (M 1574 <
1573, 1568) are received indicating a radiological release problem in the tunnel and turbine building. !
REMARKS: Operator will enter ONI-N11 for steam line break outside containment, and ONI-D17 for rad alarms
' (though no guidance is given for this situation). 3 Operator should consider entry into PEI-7 for rad releas :l control, which implys Alert level conditions exist per l EPI-A1. Operator will associate alarms with RCIC i operation arid snay choose to isolate RCIC. If isolated, operator will be informed that Inboard Isolation Valve ;
t E51--F063 f ailed to shut (Release still in progress). , {
COMMENT: '
j
. \
0010 EVENT: Turbine Building Ventilation Exhaust System is exhausting to outside of. plant (non-fil t ered) . Operator is informed that radiation levels at the site boundary east of the turbine building are 13 ar/hr and increasing (measureecnt taken by team sent out by control room or b*
- a team that "Just happened" to be practicing in the area for the upcoming E-P1,an Exercise).
REMARKS: Operator may concider Emergency O
n grescuru ation to sto, rotease, aithou h he is still ;
f ar away from Gcneral Escrgency levols which require this l
1 l
i Q action. Operator continues to control drywell related PEIs, schile level decreases slowly. The lack of te makeup may also cause the operator to consider Emergency Depressurization (at 16.5 inches) or slow depressurization through SRVs to allow LPCS to insect.
Operator may also consider securing the Turbine Building c Exhaust System, though release would still continue in an a
unmonitored manner through turbine building leakage if this option is used.
COMMENTS:
l 1
0025 EVENT: Operator is informed that E51-F063 has shut after technician troubleshot the valve's MCC. Release rates fall. Motor feed pump becomes available.
O- REMARKS: Operator continues to control RPV and
) containment, while monitoring rad levels in turbine building.
COMMENTS:
4 0030 EVENT: END OF DRILL. ,
COMMENT:
O
SCENARID: MSIV ISOLATION WITH CTUCK OPEN SRV,
(
PEIs EVALUATED: 1 (RPV CONTROL) 2 (SUPPRESSION POOL TEMPERATURE) 6 (SUPPRESSIDN PO(X. LEVEL)
INITIAL CONDITIONS: 100 PERCENT POWER, BEGINNING OF CYCLE (Simulator- Malfunction 707 - Failure to Scram - is active)
TIPE EVENT - REMARKS - EVALUATOR COMMENT 0000 EVENT: A rupture of a weld on the intermediate pressure condenser occurs, resulting in decreasing condenser vacuum. (Simulator malfunction number 717 is inserted).
REMARKS: Operator enters ONI-N62, starting an additional cire water pump. checking SJAE parameters, reducing reactor power by reducing reactor recire flow.
COMMENTS: !
O O '
0003 EVENT: Vacuum continues to decrease in spite of operator actions above. (Simulator malfunction 717 is increased)
REMARKS: Operator should trip the main turbine if vacuum is less than 25 inhg when power is reduced below 35 percent. The following actions will occur as vacuum rapidly decreases: Main turbine trips at 22 inhg, Feed Pump turbines trip at 17 inhg, MSIVs isolate at 9 inhg.
COMMENTS:
(
O
. . .. W: ' 'C 7 N' esu ,
0007 EVENT: MSIVs isolate as vacuum reaches 9 inhg.
C- (Simulator malfunction 717 is increased). For some
(,) non-apparent reason, almost all scram pilot valves do not open, reactor does not scram.
REMARKS: Operator should recognize that scram has not 1 occured and enter PEI-1 for power, pressure, and level control. Operator should attempt manual scram and place <
Mode switch in " shutdown." ALERT should be declared per EPI-A1 (Emergency Action Levels), because of failure to scram.
COMMENTS: W 0008 EVENT: Reactor pressure increases rapidly, and with it reactor power. Several SRVs open to relieve pressure.
Level decreases due to loss of reactor f eed pump
) turbines.
REMARKS: Operator should continue PEI-1 power control section and attempt a manual Alternate Rod Insert (ARI) using the Redundant Reactivity Control System. Operator should cycle SRVs per PEI-1 pressure control section to keep pressure below 1033 psig. Operator should bse .
Motor Driven Feed Pump, HPCS, and/or RCIC per PEI-I level control section to control level 183 to 218 inches. -
COMMENTS:
0008 EVENT: ARI is successfull in causing full reactor scram. However, one SRV has mechanically stuck near full open. (Simulator mal function 0521 - B21-F051D).
( Suppression Pool Temperature rises toward 95 F.
() REMARKS: Operator will not be able to tell that this L
1
SRV is stuck open until reactor pressure falls below the low low setpoint of 926 psig and SRV fails to shut.
Operator should enter ONI-B21-1 (Stuck open SRV), which will direct his entry into PEI-2 (Suppression Pool Temperature Control), if it has not already been entered due to high suppression pool temperature. This will direct him to line up at least one loop of RHR-Suppression Pool Cooling per SOI-E.12. Failure of an SRV to shut is an UNUSUAL EVENT per EPI-A1. .
s COMMENTS:
0015 EVENT: Reactor pressure decreases, suppression pool temperature increases, suppression pool level increases ;
due to stuck open SRV. j REMARKS: Operator should monitor suppression pool temperature verses reactor pressure limitations as given by the Heat Capacity Temperature Limit (HCTL) curve f rom PEI-2. If necessary, the operator should use additional methods to lower RPV pressure to remain below the HCTL per PEI-1 pressure control (using RCIC, RHR Steam Condensing, cpening additional SRVs, etc.), or an additional loop of suppression pool cooling should be lined up to keep below the HCTL. Emergency .
depressurization will be required if HCTL can not be met. Addi ti onall y, operator should attempt to lower suppression pool level below 18.5 feet by entry into PEI-6 (Suppression Pool Level Control). This will require use of the Suppression Pool Cleanup System to reject water to radwaste. Operator must also observe Suppression Pool Load Limit (SPLL) curve of PEI-6. If unable to maintain level below SPLL, emergency depressurization will be required.
COMMENTS: I I
I O
-._ _ . , - , - , , . - - ~ - . , . , , --
0030 EVENTS Reactor pressure decreases at such a rate that ,
the HCTL is not exceeded, although suppression pool l temperature continues to increase. l l
REMARKS: When directed by.PEI-1, .the operator will exit to ONI-C71 (Reactor Scram) for follow-up actions. As pressure decreases toward 135 psig, preparations should be made to place a loop of RHR into Shutdown Cooling Mode per SOI-E12. Operator may allo be following IOI-6 (Cooldown, Main Condenser Not Available) as a guide for 8
the shutdown of balance of plant. Appropriate emergency
- plan notifications with updates should be ongoing per l EPI-A2 (Unusual Event), EPI-A3 (Alert), EPI-A10 7
. (Recovery), and EPI-B1 (Emerge 7cy Notification System). '
COMMENTS:
l n 0040 EVENT: END OF SCENARIO REMARKS: Scenario may be terminated at any point after reactor pressure has decreased to 700 psig and PEI-2 and PEI-6 have been entered.
COMMENTS:
l i
I
, f, o i
SCENARIO: 7 SRYS STUCK OPEN, FAILURE TO SCRAM
{
PEIs EVALUATED: 1 (RPV CONTRM.)
1 ATT. 5 (POWER CONTROL WITH LEVEL) 2 (SUPPRESSION POOL TEMPERATURE) 6 (SUPPRESSION POOL LEVEL)
INITIAL CONDITIONS: 100 PERCENT POWER, BESINNING OF CYCLE ;
TIPE EVENT - REMARKS - EVALUATOR COMMENTS
. 0000 EVENT: Two SRVs open spuriously due to instrument technician error (Simulator malfunction 519, 522).
REMARKS: Operator enters ONI-B21-1 (Inadvertent SRV opening).
COMMENTS O
O 0005 EVENT: Operator unable to shut SRVs (apparent mechanical '
bind). .
REMARKS: Operator lines up both loops of RHR shutdown cooling per SOI-E12-1, as directed by ONI-B21-1.
COMMENT:
0006 EVENT: Suppression Pool temperature approaches 95 F.
REMARKS: Operator enters PEI-2 (Suppression Pool
( Temperature Control), declares Unusual Event per EPI-A1 (Emergency Action Levels). Atterr. pts f ast shutdown per O o"x-822-2 <=r eet-2).
l
l 1
O r .
b EVENT: No rod motion occurs when reactor is manually 0008 scrammed (Simulator malfunction 707).
REMARKS: Operator simultaneously enters PEI-1 sections for POWER, PRESSURE, LEVEL control. Attempts RRCS ARI per PEI-1.
I
/ COMMENT: j 1
1 0
0 0009 EVENT: ARI fails to cause rod motion (Simulator Malfunction 856).
REMARKS: Operator declares Site Area Emergency per '
EPI-A1. Continues PEI-1 POWER, PRESSURE control. Enters PEI-1 Attachment 5 (Power control using level). Recirc Pumps are tripped and main generator is lef t on line as heat sink. Operator attempts repeated resets of scram and ARI fallowed by manual scram and ARI.
COMMENT:
l 0010 EVENT: Supression pool high level alarm obtained.
O acnaaxs= oa r ter mav aaticiaate a a to eater eeI-*
1
~
- e - de-* , m ,
. (Suppression Pool Level Control) and attempt to lower water level using the SPCU systee.
COMMENT:
. i 0
i 0011 EVENT: Scram signal will not reset (Simulator malfunction 707).
REMARKS: Operator will bypass RC&IS low power setpoint, drive rods in. Operator controls water level 183 to 218 inches.
COMMENT:
h O
0015 EVENT: Suppression pool temperature approches 140 F. Ckt Suppression pool high level, RCIC and HPCS suction lineup shifts to suppression pool.
REMARKS: Per PEI-1 POviER control, operator initiates SLC system A and B, and overrides automatic ADS. Per PEI-1 Att. 5, operator begins to terminate f eed to the vessel to lower level and reactor power. Operator bypasses the LOCA trip of the XH-11 and/or XH-12. Operator manually initiates all ECCS and overrides each injection valve shut, and bypasses the MSIV low level isolation trip. Dn RHR initiation, SPCU being used f or suppression pool level control will isolate, and containment air supply will isolate, removing the capability to use the CRDH system to insert rods RHR re-aligns for LPCI mode. Operator will have to manually re-align RHR f or suppression pool cooling, noting that it will be 10 minutes before the RHR heat exchanger bypass valves may be shut. Operator should refer frequently to suppression pool heat capacity and
(- load limits of PEI-2 and PEI-6. Operator must attempt to keep generator on-line until power is below the capability O of the Steam Bypass System, or otherwise a pressure L
s transient will occur. Should a pressure transient occur, the operator must utilize additional SRVs per PEI-1 PRESSURE control to keep pressure below 1033 psig. If pressure reaches the RRCS high pressure trip point of 1063 psig, a feedwater system runback will occur 25 seconds later.
COMMENT: .
s i
P 0020 EVENT: Nonie.
I REMARKS: Notifications and facility activiations per EPI-A3 (Alert) and EP1-B2 (Emergency Notification System) should be in progress.
COMMENT: l 4
0030 EVENT: Power level decreases to below the capacity of the Motor Feed Pump, due to actions of SLC, rod insertion through RCSIS, and lowering water level. Reactor pressure decreases due to action of open SRVs and the EHC/ Steam Bypass Pressure control system.
! REMARKS: Operator must note that the main turbine, acting j as the heat sink, will trip on reverse power when reactor
! power is still about 12 percent, due to the steam that is j bypassed through the SRVs. The operator should take >
actions to manually secure the Main Turbine as generator megawatts decrease below 90 MWe. This will cause a pressure transient until Bypass Valves settle out.
COMMENT:
C
- O
I 0035 ' EVENT: Reactor power approaches the APRM downscale setpoints RPV water level approaches Level 2. R6 actor l pressure decreases rapidly when power level drops below that necessary to make up heat losses through the stuck open SRVs.
REMARKS: Operator must note carefully the consequences of '
reaching Level 2, which causes a BOP isolation and RCIC initiation. RCIC initiation will cause a trip of the Main Turbine (if not already tripped) and the Reactor Feed Pump Turbines.
COMMENT:
i 0040 EVENT: Stuck open SRVs shut (Remove simulator malfunctions $19, 522).
REMARKS: Operator will observe maximum cooldowrf rate gf 100 F/HR and control pressure per PEI-1 PRESSURE control.
COMMENT:
4 t
0040 EVENT: Power level decreases below APRM downscale setpoint.
REMARKS: Per PEI-1 Att. 5, water level should be e maintained at this level using the feed system (alternate
( makeup systems may be used).
COMMENT:
~,
b r.--,--- - - - - . , , - - r -- ,
ym-wn ,y-,--. -- -. --._-__--..-,__,_.-,---,_-w_nn, --,n , _ - - n.- - . , - - - - - . ,-e, --e -
1 h
0045 EVENT: All rods insert automatically (Remove simulator malfunction 707)
REMARKS: Operator exits attachment 5 and enters the level control section of PEI-1. Water level is restored to 183, to 218 inches. Per PEI-1 PRESSURE control, operator continues normal cooldown using Bypass Valves and RHR Shutdown Cooling (when pressure drops below 135 psig).
Maximum allowed cooldown rate is 100 F/HR. Operator exits ,
PEI-1 and enters DNI-C71-1 (Reactor Scram). Operator may reset RHR LPCI initiation signals and restore SPCU and containment air. Also, any BOP isolation that has occured should be reset. As much suppression pool cooling as possible should be placed in operation, and operator should attempt to restore proper suppression pool level using the SPCU system.
I COMMENT:
4 0055 EVENT Suppression pool temperature and level under operator control below heat capacity and load limit.
REMARKS: Operator continues use of PEI-2 and PEI-6 until suppression pool conditions return to normal. Reactor '
plant is being controlled through supplemental actions of ONI-C71-1. Emergency plan restoration and notifications j occur per EPI-A10 (Recovery).
I j COMMENT:
i
(:)
l -
.. . _ , _ _ . _ - _ _ _ _ _ _ , _ , _ _ _ _ ~ . _ _ . _ . , _ , _ _ _ _ , , _ _ . - _
0100 EVENT: END OF DRILL.
REMARKS: None.
COMMENT:
a O
e 6
6
. =
SCENARIO: 8 STUCK OPEN SRYS, FAIL TO SCRAM CRACK IN LPCI-B INJECTION LIFE
'CONTAIN!1ENT PRESSURIZATION FLOODING REQUIRED PEIs EVALUTED: 1 (RPV DONTRm.)
. s 1 ATT. 2 (DEPRESSURIZATION) i
, 1 ATT. 4 (RPV FLOODINS) ,
1 ATT. 5 (LEVEL / POWER CONTROL) 2 (SUPFRESSION POOL TEff.)
3 (DRYWELL TEMPERATURE) 4 (DRYWELL TEPPERATtRE) 5 (DRYWELL/ CONTAINMENT PRESStRE) ,
. . , _ . _ . .._. . . .- . .- .-- --- - - ~ ~ ~ ~ ~ ' ~ ~ ~ " ~ ~
INITIAL CONDITIONS: 100 PERCENT POWER, BOC RHR-B OOS BECAUSE F041B INDICATES INTERMEDIATE POGITION, F042D AND PUMP TAGGED OUT. NDT SCHEDULED FOR ACCESSIBLE PIPING / WELDS UPSTREAM F042B (M 515). "
SPMQ LEVCL INSf, (g-TRAW) oug op S gq,qr ty,
) E.Ri% o W T e 5 5 EttVetG.
TITE EVENT - REMARKS - EVALUATOR COMMENTS ,
0000 EVENT: Two SRVs receive spurious energization of their "A" solenoids and open (M 536, 527).
REMARKS: Operator enters ONI-821-1 (ERV stuck doen). .
COMMENT:
0003 EVENT: Operator pulls fuses for SRVs. Solenoids -
de-enorgize, one SRV chuts (remove M 536) but other SRV remains mechanically stuck open.
, REMARKS: Operator lines up RHR-A suppression pool I
( cooling, anticipates entry into PEI-2, declares Unusual Event per EPI-A1.
COMMENTS:
N ..
g.- , , - - - - , - . _ . - ---_.,-r_,. -- p..--,_-,p,._...,m., , , t,m,, . . _ _ _ - - _ _ . . -,__,,_,,--,...,_o.,,,, _ . _ - _ _ _ _ , , _ - _ - , , . _ , , . - _ -
1 s
0005 EVENTS Suppression pool temperature exceeds 90 F.
REMARKS: Requires entry into PEI-2. Operator will conduct fast shutdown before pool temperature reaches 110 F.
COMMENTS:
1 i
0007 EVENT: No rod eation occurs when reactor is manually scrammed (M 707) or when RRCS-Alternate Rod Insert is attempted (M 856).
REMARKS: Operator declares an ALERT per EPI-A1.
Operator enters PEI-1 and Attachment 5. Reactor recirc pumps are stopped. SLC is initiated before 110 d, and .
power control by lowering water level is begun. Rods are manually driven in. As soon as operator initiates'and overrides injection systems per attachment 5, operator will begin to recover NCC, CRDH, and Instrument Air lost as a result of the Division 1 and 2 ECCS initiation.
This may be accocplished through PEI-1, 3, 4, and/or 5.
P50 statkoaA0 Me6t. vee.# vAi,vss ran, n omsnj, COMMENTS:
I
(
0013 EVENT: Report of steam fors.ation in vicinity of E12-F042B is made by HP Technician who observed condition as he evacuated containment, having been conducting a rad survey in vicinity. Containment pressure begins to rise,
I l
W drywell vacuum breakers open, and drywell pressure begins to increase (M 704, 1073, 1120, 1075, 1121, 1575/3 min, cover erroneous indication)
REMARKS: (The events above are simulated by the instructor, using data sheety showing parameter values to f be used by the plant operator.) Drywell pressure will follow containment pressure due to drywell vaccum breaker i.
system. When drywell pressure reaches 1.68 psig, the :
ECCS initiations and isolations will take place, ;
including the drywell vacuum breaker system isolation.
Drywell pressure increase will then increase only slowly in response to containment pressure increase. This increasing differential pressure will cause the suppression pool level to drop as it is forced into the drywell. Operator continues to combat failure to scram and SRV problems, while anticipating PEl-5 entry.
COMMENT:
4 1e?? EVENT: Suppression Pool Low-Low level reached, SPMU C01%l dumps (M 879). Containment High Temperature Alarms occur at 110 F (M 1081, 1115) AT TW C CO LS . -
REMARKS: 'Drywell is still sealed from containment by the level in the suppression pool, and increases slowly until i th r? level of the suppresion pool reached top row of vents (requires about 4.5 psid between Containment and Drywell af ter SPMU Dump f or the vents to be reached).
COMMENTS:
p!
0025 EVENT: Prsmary containment pressure continues to rise toward the Pressure Suppression Limit of PEl-5. Aver age Containment Temperature is 156 F and only slowly rasang.
REMARKS: Operator will consiqer initiating Containment Spray Mode of RHR-A, which must be accomplished before reaching the limit. Operator will also be following PEI-4, monitoring containment temperature (Limit is 185 F). '
COMMENT:
a 0030 EVENT: Operator initiates containment spray, with only minimal effeet on containment pressure. Top row of vents are reached as Suppression Pool level drops from effects
- of high containment pressure and the initiation of cont.*fament spray.
l .
REMARKS: Drywell pressure will begin to f ollow containment pressure, minus 4.5 psi as pressure equalizes through vents. Suppression pool level will stop dropping.
j COMMENTS:
4 0033 EVENT: Containment pressure continues to increase beyond ,
Pressure Suppression Pressure limit of PEI-5.
REMARKS: Emergency depressurization is required.
Operator performs depressurization af ter .gumping to portion of attachment 5 that terminates all i n.j ec ti on execpt SLC and CRD (step 7).
COMMENTS:
0
O 0040 EVENT: Eight ADS SRVs are opened by operator.
REMARKS: Pressure will be allowed to decrease until the 6
minimum alternate RPV flooding pressure for eight open '
SRVs is reached (90 psig), at which time the operator will recommence injection using the feed system to maintain water level 16.5 to 30 inches.
[
COMMENTS:
I 0045 EVENT: All rods that are out scram in (remove M 707, 836).
O acaaax== oa r ter PEI-1 level control. He will feed quickly using the feed it ** ca at = ma r sura to system to maintain water level 183 to 218 inches.
CONNENT: .
9 909tr EVENT: Containment pressure has continued to increase, I 0050 but at a slower rate. The Containment Design Pressure k, Limit of PEI-5 has been reached.
REMARKS: RPV Flooding (Attachment 4) is required (Note:
Operators still have reliable RPV level indication).
Operators will inject using available systems to maintain
(
level above 400 inches until containment pressure can be .
maintained below the Containment Design Pressure Limit. l COMMENTS:
1
.,v. . , - - , - - - - - , - - - - - - - - - - - ~ - - - - ~ ~ ~ - - - ' ~ ~ ^ ' ' - ' ~ ~ ~ ~ ~ ' ~ ' ' ~ ~ " ' ' ' * ' " "~ ~ ' " * ""
-e '
4449 EVENT: Containment pressure begins rapid decrease.
0000 REMARKS: Attachment 4 will be exited af ter containment pressure is reduced below the Design Pressure Limit.
Before containment pressure drops below zero, containment spray will be terminated.:
COMMENT:
l 1
i I
0120 EVENT: Drywell pressure decreases as leakage between containment and drywell is reduced. Containment and drywell parameters approach normal.
REMARKS: Operator exits PEI-1 and enters ONI-C71 (reactor scram) and 201-13 (Cooldown after Boron Injection). Operator continues execution of containment related PEIs, including the restoration of Suppression Pool Temperature and Level.
COMMENTS:
i 0130 EVENT: END OF SCENARIO.
COMMENT:
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PNPP PLANT EnnenhENCT INSTRUCTION VALIDATION Sf'1DIARIO h CREN NUMBER: DATE: EVALUATOR: SCENARIO: 81 (1/88) - MSIV ISOLATION, RODS OUT, LOCA PEIs EXXBCISED: RPV Control (B13) Emergency Depressurisation (B13 Att. 1) 4 RPV Flooding (313 Att. 4) , Power / Level Control (B13 Att. 5) Containment Temperature Control (D23-1) Drywell/ Containment Pressure Control (D23-2)
- Drywell Temperature Control (D23-3)
Suppression Pool Temperature (E12) Suppression Pool Level (G42) INITIAL CONDITIONS: Full power, BOC; HPCS pump out of commission due to breaker repair since 0800 (M517), expected available in four hours; NCC pump C tripped on overcurrent last shift, maintenance investigating; I&C SVI on MSL rad monitors in progress. TIME EyJNT/Elo! ARKS / EVALUATOR COMMENTS 0000 EVENT: A motor feed pump flow control valve control signal loss alarm occurs (M808). REMARKS: Operator has lost control of the MFP FCVs from P680, and the FCVs fails as are. MFP was not in service. Operator should refer to ONI-C34 and have I&C investigate problem. COMIfENTS: ,
.- : . :m -
0004 EVENT: I&C Tech errors cause a spurious MSIV isolation and scram (M849). Several rods do not fully insert (M421-7). RElfARKS: A rapid loss of RPV water level occurs due to loss of the turbine feedpumps. MFP provides only a small amount of makeup through the failed FCVs. Operator enters PEI-B13 because of MSL isolation, high RPV pressure, and low water level. Operator will enter
Attachment 5 of PEI-B13 for water level control with rods not fully inserted, which directs that only feed, RCIC, or CRDH be used for injection. COMMENTS: 0006 EVENT: RPV water level drops below Level 2 (130 inches). RCIC initiates, Balance of Plant isolations occur. REMARKS: Operator will verify that RCIC is operating properly. Operator will use SRVs to control pressure below 1033 psig, while investigating reason for MSL isolation. QOMMENTS: 0010 EVENT: Suppression Pool temperature and level rise as a
. result of SRV openings. I&C Tech reports the error to Unit Supervisor which caused spurious MSL isolation from MSL radiation monitors. . . - - t-RFMARKS: Operator may make early entry to PEI-E12 for suppression pool temperature control (required when temperature reaches 90 degrees) by placing at least one loop of RHR suppression pool cooling in operation. Upon determination of cause of MSL isolation, operator should take steps to restore vacuum, if it is low, and open MSIVs to enable pressure control to be resumed using the steam bypass valves to the main condenser.
.CQUMENTS:
( (:)
0020 EVENT: RPV water level trends toward normal water level using RCIC. Pressure control may be on the bypass valves. RENARKS: Operator should make efforts per PEI-B13 section 3.1 to insert rods using the CRD system. Only two rods will be movable (Remove M424-5). Scram signals I
- may be reset when RPV water level is above 170 inches.
i If desired, operator may restore turbine driven feed pumps to operation when RPV level returns above the RCIC automatic initiation point of 130 inches and the RCIC initiation signal is manually reset. Any operator , attempt to manually control the motor feedpump FCVs locally will fail due to mechanically stuck valves. COMMENTS: 1 1 0025 EVENT: A small break LOCA occurs in the drywell from a small crack in a reactor recirculation loop (M704). Drywell pressure and temperature begins to increase. REFARKS: Operator should continue to use PEI-B13, with concerns directed toward the newly developed leak. ,
--Operator should take steps to maximise-11rywell cooling i te slow drywell pressure rise.
Shift
- - Supervisor should direct investigation of Drywell
- .- - equipment and floor drain sump fill rates at the leakage detection panels to determine magnitude of the leak, and declare an Unusual Event per EPI-Al (Emergency Action Levels) for pressure boundary leakage.
COMMENTS:
0030 EVENT: Drywell pressure reaches 1.88 psig. Automatic C isolations of NCC and Instrument Air to theAll containment occur as Division 1 and 2 ECCS initiate. low O pressure ECCS start as expected. Any RHR system aligned for suppression pool cooling realign for LPCI mode. Stub busses XH-11 and XH-12 trip, doenergising all I available NCC and CRD pumps.
. REMARKS: Operator enters PEI-D23-2 for drywell pressure ,'
control, including restoration of a stub bus for a ECCS division that is not providing makeup to the vessel, e restarting NCC and CRD, and restoring NCC and instrument t air to the containment and dryvell. Operator should be , using RCIC or Feedwater to maintain RPV water level, so ! any stub bus may be chosen for reenergisation. Operator will also enter PEI-D23-3 for drywell temperature
- control when temperature reaches 135 degrees, and PEI-D23-1 for containment temperature control when temperature reaches 90 degrees. If operator diagosos leak rate to be in excess of 50 gym, the Shift Supervisor should declare an Alert per EPI-A1.
COMMENTS: C O-00W EVENT: AfailureoftheHtIburgeTanklowlevelswi.tch occurs, causing a trip of all.re' actor feed booster pumps and feed pumps (M841, 801)., - - EKMARKS: RPV water level begins to decrease slowly as l
- - -operator uses RCIC to provide maximum makeup. Operator should diagnose the inability to maintain water level above sero inches and proceed to steps in PEI-B13 Attachment 5 which require emergency depressurization and the override of all ECCS injection valves to the shut position. This will prevent uncontrolled injection i within core shroud when rods are still stuck out of the Core.
COMMENTS: O'
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0090 EVENT: Emergency Depressurization in progress. REMARKS: Operator will terminate all injection into ' vessel in order to assure no power and pressure ^ transients occur during depressurization. Operator will
. depressurize using eight ADS SRVs until the minimum alternate RPV reflood pressure of 90 psig is reached. !
Operator should make preparations for using RHR loop A or B to inject when this pressure is reached. This will . require the bypassing of the shutdown cooling interlocks to the R8R return to feedwater valves E12-F053A/B in order to direct injection into RPV downcomer annulus. Operator should choose a division of RHR whose Stub Bus is not energized, or should deenergize the stub bus prior to using the ICCS for makeup. COMMENTS: 1 i l i 1 0100 EVENT: Injection from RHR in progress. Vessel level , t may be.below the bottom of the-Fuel Zone level - tnstruments before injection begins to recover. water
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level. Suppression po'ol te'mperature may hcve reached the boron injection temperature limit of 133 degrees due - to'depressurization.-
- REMARKS:- If water level dropped below -150 inches on the fuel sene~ instruments, then the operator should declare that RPV level cannot be determined, and enter Attachment 4 (RPV flooding). This will lead to nearly '
identical actions as Attachment 5 was requiring. When j water level returns to indicating range and is consistent, and if drywell temperature has not exceeded f the saturation temperature corresponding to present reactor pressure, then the operator may determine that RPV water level indication has been restored. This will place the operator back in Attachment 5. If Suppression Pool temperature has reached 133 degrees and the operators have not been informed the reactor engineering ( has determined that the reactor is shutdown, then Standby Liquid Control should be initiated. If drywell l l () pressure is above 1.68 psig when water level is below
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) 18.25 inches, the Shift Supervisor may declare a Site Area Emergency per EPI-Al based on a technicality of that proceedure under loss of makeup systems.
COMMENTS: s 0110 EVENT: RPV water level i.s rising with consistent : readings betwoon instruments. REMARKS: Operator will restore water level to above zero inches with RHR. COMMENTS:
.. 0115 Event: Operator is informed that all rods have suddenly inserted (Unable to simulate this, operator will be
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handed a message to this effect). , REMARKS: Operator may leave Attachment 5, terminate
- boron injection if it had been initiated, and return to.
main body of PEI-B13 to restore water level 185 to 215 inches, at which point PEI-B13 may be exited for the Off Normal Instructions to further control the plant. Operator will continue in other PEIs as necessary to restore containment, drywell, and suppression pool paramenters. COMMENTS: (:) !
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h PNPP PLANT EWrnmtNCY INSTRUCTION VALIDATION SITNARIO CREN NUMBER: DATE: EVALDATOR: SCENARIO: #2 (1/86) - LOM VACUUM NSL ISOLATION, ' i SRV FAILURE, LOCA , I 5 PEIs EXERCISED: RPV Control (B13) Emergency Depressurization (B13 Att. 1) l j Level Restoration (B13 Att. 3) RPV Flooding (B13 Att. 4) ' Containment Temperature Control (D23-1) Dryvell/ Containment Pressure Control (D23-2) Drywell Temperature Control (D23-3) Suppression Pool Temperature (R12) Suppression Pool Level (G42)
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INITIAL CONDITIONS: Full power, BOC; HPCS pump out of commission for breaker repair since 1100 (M517); NCC pump C tripped on overcurrent last shift, maintenance investigating; Cire Water pump C protective relay calibration in progress, pump tagged out.
. TIME EYDIT/nmBEIS/ EVALUATOR COMMENTS
< 00:01 EVENT: A motor feed pump flow control valve control signal loss alarm occurs (M806).
REMARKS: Operator has lost control of the MFP FCVs from P880, and the FCVs fail as are. MFP was not in service. , 1 Operator should refer to ONI-C34 and have I&C ' investigate problem. Any attempt to open the FCVs . locally will fail when the operator finds the sq.allest FCV to already be full open and the large FCV to be - mechanically stuck shut. If desired, the small FCV may l be open or shut as desired. cot?N: l 00:04 FVrtlT: A complete loss of division 1 DC bus EDIA occurs (M701). EELES: The loss of this bus will place all division 1 ECCS and the division 1 diesel generator out of service. RCIC will be out of service. SRV "A" solenoids and lodic will lose power supply. Operator should refer to
4 ( Tech Spec limitations for ECCS and elactrical power, and refer to ONI-R42-1 (Loss of EDIA). O g CONNENTS:
- I 1
I e i i 00:09 EVENT: Cire Mater Pump A trips on inverse-time , overcurrent (M582). Vacuum begins to drop, though not drastically. REMARKS: Oper ator should refer to ONI-NS2 for loss of , vacuum and begin power reduction to attempt to maintain vacuum with one cire water pump. An automatic recirculation system flow control valve runback may i occur as vacuum drops. l comfatTg: . i O ' 00:10 EYW T: Circ Mater pump B trips (M583) on instantaneous overcurrent due te pump cavitation. Vacuum drops suddenly resulting in a trip of the main turbine and the reactor feedpump turbines, a scram, and a MSL isolation. One outboard MSIV fails to shut (M698). REMARKS: Operator should enter PEI-B13 based on the MSL , isolation. In the first few moments of this transient, ' ' the possibility exits "that the feedpump level 8 trip , will seal-in before level drops as a result of loss of high presure makeup to the vessel. The operator should take action to restore the motor driven feed pump to operation, even though only reduced flow will be available through the stuck FCVs. This will be the only source of high pressure makeup available. Level will (.. trend down slowly until some feed is restored, due to SRV cycling. Operators should note the failure of the i MSIV to shut and attempt to shut it manually.
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I l i O , f i 00:11 EVENT: Upon the first attempt of the pressure relief ! logic to cycle the SRVs using the "B" colonoids, the l ' combined power feeder to the "B" solenoids from ED1B06 i fails when a fuse blows. " ADS B OUT OF SERVICE" annunciator is received (M1252) on P601. . Pressure rises to the safety setpoint of 1165 psig, at which point eight SRVs cycle. - FRFMS: Operators should recognize the loss of the I normal relief function of the SRVs, which will make it impossible to control reactor pressure below 1033 peig as specified by PEI-B13. Control of SRVs at the remote shutdown panel will likewise not be available because of the loss of EDIA bus. Operator should begin the use of l l f' augmented means of pressure control specified in PEI-B13 l using RWCU and RER steam condensing. Operator should O investinate rossibi11ty of re<ainin, a ciro water eum, to restore the main condenser vacuum which will allow l the opening of MSIVs, the use of the steam bypass 1 valves, and the use of feed pump turbines to piovide . I feed. .
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I cot"":NTS: 00:14 EVFUT: Suppression Pool temperature and level rise as a result of SRV openings. FAME: Operator may make early entry to PEI-E12 for suppression pool temperature control (required when temperature reaches 90 degrees) by placing RHR B in (- ei;uppression pool cooling mode. O ca m ns: 1
(O 00:25 IVENT: Pressure control recovered using a modified steam condensing line up, and RNCU; or through restored main condenser. RPV water level trending up towards normal range. Operators have replaced fuse to 'B' solenoids and thus partially restored the relief function to the SRVs (remove N1252, page FX). REMARKS: When pressure control has been restored using e the bypass valves to the main condenser, operators may direct energies toward restoring EDIA. Because the "A" solenoids have no power, any manual cycling of SRVs must t take place et P631. Operator may declare that he has exited PEI-B13. f CONMENTS- . 1 l l
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p- 0000 EVENT: A small break LOCA occurs in the drywell from a small crack in a reactor recirculation loop (M704). 0 Dryvell pressure and temperature begins to increase. EMfAELS: Operator should return to PEI-B13, with concern directed to the newly developed leak. Operator should take steps to maximise drywell cooling to slow pressure rise. Operator may use authority of PSI-B13 to lower pressure while observing a 100 degree per hour - cooldown rate limit, in order to minimise the leak rate. Shift Supervisor should direct investigation of drywell -
- equipment and floor drain sump fill rates at the leakage detection pecels to determine the magnitude of the leak, and declaro aa Onusua.1 Event per EPI-Al (Imargency Action Levels) for pressure boundary leakage.
COE TNTS:
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l C 00:33 FVENT: Drywell pressure reaches 1.68 psig. Automatic Oy ' isolations of NCC and instrument air to the containment and drywell oc:ur as Division 2 ECCS initiates (Div. 1 is out of service from loss of EDIA). RHR B realigns for LPCI mode. Stub bus XH12 trips deenergizing NCC B.
. REMARKS: Operator enters PEI-D23-2 for dryvell pressure a control, including the restoration of NCC to the drywell. Level may be maintained using the feed system. -
If RH3 B or C is chosen for injection, operators must de-energize the XH12 if they had reenergized it. Operator wil also enter PEI-D23-3 for drywell temperature control when temperature reaches 135 degrees, and PEI-D23-1 for containment temperature control when temperature reaches 90 degrees. If operator diagoses leak rate to be in excess of 50 gym, the Shift Supervisor should declare an alert per EPI-A1. COMMENTS:
'i (0
00, 38 EVENT: Operator is unable to unisolate Div. 2 (Inboard) NCC isolation valves due to failure of LOCA bypass circuits. - ERMARKS: Operators should recognize the implications of , a continued drywell heatup on vessel water level instrumentation and refer occasionally to the level
. instrumentation temperature limits in the PEIs.
Investigation of bypass problem should be started. COMMENTS: 00:43 gVENT: A failure of.the Hot Surge Tank low level switch () occurs, causing a trip of al reactor feed booster pumps and feed pumps (H841, 601).
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l _ REMARKS: RPV water level begins to decrease because of
.() the loss of high pressure injection. Operator will eventually diagose the inability to maintain water level above sero inches and proceed to PEI-B13 Attachment 3 for level restoration. Operator may try to lower reactor pressure to the point where LPCI-B or C can
. inject without emergeucy depressurization. However, the i leak rate will be auch that level drops faster that pressure. Upon determining that emergency r
depressurisation will become necessary, the operator should rapidly depressurise to the main condenser if it is available until the MSL isolates at 16.5 inches. ' Emergency depressurization per PEI-B13 Attachment 1 should teke place when zero inches is reached to allow LPCI injection to take place. COMMENTS: 0 00:30 EVENT: Emergency Depressurization in progress. RFMARKS: Since the "A" solenoids for the SRVs are not powered, the opening of the eight SRVs for this function will have to be done by initiating ADS logic, oh by - opening the SRVs from P631. COMMENTS: 00:53 EVENT: LPCI B and C inject into the vessel as pressure drops. Water level drops into the fuel zone before beginning to rise. ( REMARKS: If water level drops below 18.25 inches while dryvell pressure is above 1.68 psig, the Shift Supervisor may declare a Site Area Emergency per EPI-A3. (]) If water level drops below -150 inches or if the drywell I i )
l I temperature is above the temperature corresponding to i RPV saturation pressure as shown in the level instrumentation temperature limit curves, then water level indication is suspect and RPV flooding per PEI-B13 Attachment 4 is required. This will require the ) operators to continue flooding the vessel until level I indications are consistent. If reference lec flashing )
- is suspected then flooding will continue until an RPV !
pressure 112 psid above containment pressure is reached ) for a specified amount of time and drywell temperature control is restored. At this point injection would be ) terminated for the maximum core uncovery time limit to await the return of on scale indication. COMMENTS:
; 01:00 EVENT: Maintenance reports the fix to the hot surge tank low level trip and to the NCC inboard valve bypass circuit.
BEMARKS: Operators should use the feed booster pumps to help reflood and pressurize the vessel. NCC should be restored to the drywell. , COMMENTS: 01:10 EVENT: Drywell temperature and pressure are dropping rapidly. RPV pressure has been held at the appropriate point. l EIQfAPES: Operator should take steps per RPV flood I procedure to verify restoration of water level. Upon f-doing so, operator may leave RPV flooding and return to ( the main body of PEI-B13. When normal water level is restored, PEI-B13 may be exited. Operator will have to O restore proper suppression pool temperature and level. l
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COMMENTS:
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01:15 EVENT: END OF SCENARIO OVERALL COMMERIS: . 1 i 1 k
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O ATTACHMENT C 8 , 1 4 PNPP CONTROL ROOM VALIDATION HEDS I O i i i l O
HED REPORT SHEET HED-613 D DESCRIPTION: Validation - The digital displays for reactor vessel level on 1H13-P680 do not indicate when they are off range and the range selector switch should be changed. t I HUMAN FACTORS REVIEW: Provide indication to the op'erator that the display is off range so he can switch to a different range. , l I H.F. GUIDELINES: l
.MPLEMENTATION: Fix eSAFETY AND OPERABILITY AFSESSMENT: CATEGORY I SAFETY CONSEQUENCE? _No INCREASE ERROR POTENTIAL? No VERIFICATION: CORRECTS HED? NO NEW HED7
- CORRECTION SCHEDULE
- PLANNED COMPLETION Prior To Startup Following First Refuel WORK COMPLETE
REFERENCES:
I l O DW187A/A/393/cp
P HED REPORT SHEET HED-614 D DESCRIPTION: Validation - The upper pool level scales on meters G43-RC22A(B) and recorders G43-R073A(B) are nonstandard with zero referenced to the normal surface level. - t I l HUMAN FACTORS REVIEW: The upper pool scale zero reference will be changed to reference , a spot near the bottom of the pool which will be more meaningful to the operator. 1 H.F. GUIDELINES: IMPLEMENTATION: Fix SAFETY AND OPERABILITY ASSESSMENT: CATEGORY I SAFETY CONSEQUENCE 7 No INCREASE ERROR POTENTIAL? No C'IERIFICATION: CORRECTS HED? NO NEW HED? l, V CORRECTION SCHEDULE: PLANNED COMPLETION Prior To Startup Following First Refuel ; WORK COMPLETE , i
REFERENCES:
i 3
\
O DV187A/A/394/cp i
HED REPORT SHEET HED-615 HED DESCRIPTION: Validation - The P680 SD attempted to restore the feedwater system to maintain vessel level however, he did not realize that the RCIC initiation had tripped the RFPT's since the RFPT trip alarm clears after the discharge valve closes. HLMAN FACTORS REVIEW: The RFPT trip alarm will remain as designed to eliminate the nuisance alarm. However, a RCIC initiation contact will be added to the " Main Turb & Feedpump Trip L8" Annunciator on P680-3A-A8 to alert the operator that RCIC initiation has tripped the RFPT's. The alarm window will be changed to "MN & Feed Turb Trip RCIC/L8". H.F. GUIDELINES: IMPLEMENTATION: Fix SAFETY AND OPERABILITY ASSESSMENT: CATEGORY I SAFETY CONSEQUENCE 7 No INCREASE ERROR POTENTIAL? No s ERIFICATION: CORRECTS HED? NO NEW HED? J CORRECTION SCHEDULE: PLANNED COMPLETION Prior To Startup Following First Refuel WORK COMPLETE
REFERENCES:
DW187A/A/395/cp
l HED REPORT SHEET HED-616 ED DESCRIPTION: Validation - The RCIC manual initiation pushbutton must be held in for two seconds before RCIC will initiate to ensure that the Main and Feedpump Turbines have tripped. Ifowever the RCIC seal in light comes in immediately which causes operator confusion. i HUMAN FACTORS REVIEW: Modify the circuitry so that the operator does not have to hold in the RCIC manual pushbutton for two seconds to start RCIC. [ e H.F. GUIDELINES: . IMPLEMENTATION: Fix SAFETY AND OPERABILITY ASSESSMENT: CATEGORY I No i SAFETY CONSEQUENCE? INCREASE ERROR POTENTIAL? No VERIFICATION: CORRECTS HED? NO NEW HED? (9 CORRECTION SCHEDULE: PLANNED COMPLETION Prior To Startup Following First Refuel VORK COMPLETE [
REFERENCES:
( O DW187A/A/396/cp
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