IR 05000387/1992001
| ML17157B048 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/10/1992 |
| From: | Conte R, Walker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157B047 | List: |
| References | |
| 50-387-92-01OL, 50-387-92-1OL, 50-388-92-01OL, 50-388-92-1OL, NUDOCS 9202210016 | |
| Download: ML17157B048 (15) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
REQUALIFICATIONPROGRAM EVALUATION REPORT NO.
FACILITYDOCKET NOS.
FACILITYLICENSE NOS.
92-01 (OL),
50-387 50-388 NPF-14 NPF-22 LICENSEE:
FACILITY:
EXAMINATIONDATES:
NRC EXAMINERS:
Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Susquehanna Steam Electric Station January 6 - 9, 1992 J. Hanek, INEL M. Riches, PNL CHIEF EXAMINER:
Tracy W lker, Senior Operations Engineer BWR Section, Operations Branch Division of Reactor Safety Date APPROVED BY:
Richard J. Conte, ref, BWR Section Operations Branch, DRS Date 92022100i6 920213 PDR ADOCK 05000387 V
EXE TIVE S MMARY Written and operating examinations were administered to six Reactor Operators (ROs) and six Senior Reactor Operators (SROs).
These operators were divided into three crews.
The examinations were graded concurrently by the NRC and the facility training staff.
As graded
.by the facility and the NRC, four of the six ROs and five of the six SROs examined passed all portions of the examination.
Two ROs and one SRO did not perform satisfactorily on the
.
dynamic simulator evaluation as graded by the NRC and the facility. These operators passed the remaining two portions of the examination.
Two of the three crews that were evaluated performed satisfactorily on the simulator portion of the examination as graded by the NRC and the facility. The remaining crew performed unsatisfactorily on the simulator portion of the examination as graded by the NRC and the facility. The most significant generic weaknesses were observed in the operators'bility to operate the Feedwater system and in sensitivity to reactor pressure vessel (RPV) level control and prevention of RPV cooldown with the reactor not shut down.
The licensee's licensed operator training program is considered to be satisfactory based on the criteria established in section ES-601 of NUREG-1021, Revision 6.
However, a programmatic weakness in the evaluation techniques used by the facility during dynamic simulator evaluations was identified.
The facility's initial evaluations of the unsatisfactory crew and individual operators did not agree with the NRC evaluations.
The initial disagreements appeared to be the result of an inadequate post-scenario review of the Individual Simulator Critical Tasks (ISCTs) and a lack of sensitivity to the significance of RPV level and pressure control with the reactor not shutdown on the part of the facility evaluators.
The licensee took prompt and comprehensive actions to address the individual and programmatic weaknesses.
The licensee suspended the requalification simulator examinations immediately following the NRC administered examinations in order to provide training to the facility evaluators.
The licensee is also performing a root cause analysis to determine the reason for the differences in the initial evaluations between the facility evaluators and the NRC, to look for additional programmatic problems, and to assess whether PP&L has remained current with industry standards.
The examination materials for all portions of the examinations required modifications prior to administration.
The required changes were not significant enough to indicate a lack of quality control; however, the examination materials need improvement to meet the NUREG-1021 standards and to be consistent with industry practices.
While preparing for the examinations, the examination team identified a deficiency in the Emergency Operating Procedures (EOPs) for control of injection into the RPV following rapid depressurization with all rods not inserted.
The licensee determined that the problem will be resolved when they revise their EOPs to follow Revision 4 of the B oiling Water Reactor Owner's Group (BWROG) Emergency Procedure Guidelines (EPGs) in September 1992.
In the interim, the licensee intends to train the operators on RPV level control with all rods not inserted as part of the planned training on sensitivity to parameter control ~
DETAILS 1.0 Introduction The NRC administered requalification examinations to 12 licensed operators (6 ROs and 6 SROs).
Three crews were evaluated on the dynamic simulator. 'he examiners'sed the process and criteria'described in.NUREG-1021, "Operator Licensing
'xaminer Standards,"
Rev. 6.
The Job Performance Measure (JPM) portion of the examination was administered using the Alternative B methodology described in ES-603, "Requalification Walk-through Examination," of NUREG-1021.
An entrance meeting was held with the licensee on November 6, 1991, in the NRC Region I office. The personnel contacted during the examinations are listed in Attachment 1.
The members of the combined NRC/facility examination team and the facility evaluators are also identified in Attachment 1.
- 2.0 R
uglification Examination Results 2.1 Individual Examination Re ul The following is a summary of the individual examination results:
NRC Grading Written Simulator Walk-through Overall RO Pass / Fail 6/0 4/2 6/0 4/2 SRO Pass / Fail 6/0 5/1 6/0 5/1 TOTAL Pass / Fail 12/0 9/3 12/0 9/3.
Facility Grading Written Simulator
'Walk-through Overall RO Pass / Fail 6/0 4/2 6/0 4/2 SRO Pass / Fail 6/0 5/1 6/0 5/1 TOTAL Pass / Fail.
12/0 9/3 12/0 9/3 2.2 ~R The following is a summary of generic strengths noted from the results of the individual requalification examination D namic Sim viator Examination Understanding and interpretation of annunciator and alarm signals 2.3 Generic Weakne es The following is a summary of generic weaknesses noted from the results of the individual requalification examinations.
Written Examinations Abilityto determine APRM/IRM overlap Understanding of the basis for rapid depressurization of the RPV Understanding of ADS logic Abilityto determine HPCI availability Abilityto control RPV level to prevent exceeding RPV heatup and cooldown limits Walk-throu h Ex minations'ailure to review procedures and thoroughly assess plant conditions prior to bypassing an APRM D namic Simulator Examinations Abilityto operate the Feedwater (FW) system (and understanding of FW system operation)
Lack of sensitivity to RPV level control and prevention of RPV cooldown with the reactor not shutdown Timeliness of initiation of suppression pool cooling Knowledge of a procedure change to initiate Standby Liquid Control with reactor power above 5%
3.0 R
ualificati n Pro ram Evaluation Re ults 3.1 Examiner tandards Evaluati n
riteria The facility program for licensed operator requalification training is considered SATISFACTORY in accoidance with the criteria established in ES-601, paragraphs C.2.b. (1)(a-c) and C.2.b. (2)(a-f).
The facility grading was as conservative as the NRC grading on 100% of the pass/fail decision for individuals satisfying the criterion of C.2.b.(1)(a).
Seventy-five percent'(75%) of the operators passed the examination satisfying the criterion of C.2.b. (1)(b).
Two of the three crews evaluated were determined to be satisfactory on the dynamic simulator portion of the examination as evaluated by the NRC satisfying the criterion of C.2.b.(1)(c).
The facility evaluators concurred with the NRC on the one unsatisfactory crew evaluation; therefore, criterion C.2.b.(2)(a) is not applicable.
The facility trains and evaluates operators in all positions permitted by their licenses; therefore, criterion C.2.b. (2)(b) is not applicable.
No facility evaluators were determined to be unsatisfactory; therefore, criterion C.2.b.(2)(c) is not applicable.
However, the facility evaluators initially evaluated all individuals and crews as satisfactory on the dynamic simulator examinations.
Following discussions with the NRC, the facility evaluations were in complete agreemerit with the NRC evaluations.
The initial disagreements appeared to be the result of an inadequate post-scenario review of the Individual Simulator Critical Tasks (ISCTs) and a lack of sensitivity to the significance of RPV level and pressure control with the reactor not shutdown on the part of the facility evaluators.
The weaknesses identified in dynamic simulator evaluation techniques are considered a program weakness.
The facility has administrative controls to preclude an RO or SRO who does not possess an active license from performing licensed duties without satisfying the requirements of 10 CFR 55.53 to restore the license to active status.
There were no indications of deficiencies in this area; therefore, criterion C.2.b. (2)(d) is not applicable.
The examination materials for all portions of the examinations required modifications prior to administration.
One written examination question had to be changed after the examinations were administered.
The required changes were not significant enough to indicate a lack of quality control; therefore, criterion C.2.b. (2)(e) is not considered applicable.
However, the examination
materials need improvement to meet the NUREG-1021 standards and to be consistent with industry practices.
(Sections 3.2, 3.3, and 3.4 describe the examination material weaknesses.)
The facility's failure rate for individuals was identical to the NRC's failure rate; therefore, criterion C.2.b. (2)(f) is not applicable.
3.2 Written Examinations Changes were made to the written examination questions during examination preparation to clarify the question or raise the level of knowledge evaluated by the question.
Several questions were replaced for the same reasons.
One question had to be modified after administration of the examination to accept two correct answers.
The NRC had commented on this question during examination preparation, but had been assured by the facility that there was only one correct answer.
During examination preparation, the NRC,questioned whether a number of questions were appropriate for ROs.
The facility was firm on its expectations for ROs and the questions were used in the examinations.
The ROs performed poorly on several of these questions indicating that the facility's expectations are not being fulfilled.
Some of the questions in the examination bank did not have estimated response times as required by the Examiners Standards.
Most of the estimated response times in the examination bank appeared to be too high.
The licensee provided revised estimated response times for the questions on the proposed exams.
These revised times appeared to be more reasonable; however, the resulting estimated total time was shorter than that specified by the Examiners Standards.
The overall results of the examinations indicated that the examinations discriminated between good and poor operator performance..
However, all operators completed each part of Section A in 50 minutes or less indicating that Section A may have been too short.
(The Examiners Standards specify 60 minutes for each part of Section A.) Time validation of the written examination needs improvement to ensure that examinations can be developed that discriminate, but do not allow time to look up every answer.
3.3 Walk-throu h Examinations The initiating cues and performance standards for reviewing prerequisites and precautions for several Job Performance Measures (JPMs) needed revision to make them realistic with the way the tasks would actually be performed in the plant.
These minor changes were agreed upon by 'the examination team during examination preparation; however, the facility's methodology for performance of JPMs was a concern.
The facility's methodology for performing JPMs is that all tasks are performed using the procedure.
As a result, they do not have
any tasks that are considered to be time critical. This methodology is not consistent with industry standards.
There are a number of tasks that, while a specific time criteria cannot be determined, are expected to be performed from memory in a prompt manner to ensure the safety of the plant.
Initiation of Standby Liquid Control (SLC) and manual initiation of the Automatic Depressurization System (ADS) to rapidly depressurize are examples of such tasks.
For an evaluation tool to be effective, it should be as realistic as possible.
In the case of JPMs, this means the task should be performed as it would be expected to be performed in the plant.
The facility's methodology for performance of JPMs does not provide the optimum evaluation tool for tasks that are expected to be performed promptly without'prior reference to the procedure.
The JPM questions as submitted were not time.validated and did not indicate'hether the use of references was allowed as required by the Examiners Standards.
The examination team agreed on a standard of five minutes per question and allowed use of all normally available reference materials for this examination.
The JPM questions need to be time validated and reviewed to determine the appropriate use of references to meet the requirements of the Examiners Standards.
Additionally, the JPM questions need improvement to test more in-depth knowledge and to avoid direct lookup and areas covered during performance of the task.
3.4-D namic'm lator Examin tions The facility examination bank consisted of 21 scenarios which did not meet the specifications of the Examiners Standards.
The expectation is that the facility will increase its examination bank by five scenarios per year until at least 30 scenarios are developed.
Susquehanna Steam Electric Station (SSES)
developed three scenarios this year and should have had 25 scenarios for this examination.
The facility made a commendable effort to incorporate actual industry events into the scenarios; however, the scenarios were weak in depth of Emergency Operating Procedure (EOP) usage and simultaneous events that require prioritization of actions.
The lice'nsee needs to develop scenarios until all expected or plausible abnormal and emergency situations to which control room operators are expected to respond are covered.
The scenarios proposed by the facility had to be modified to add ISCTs, use EOP contingency procedures, evaluate previously identified weaknesses, and make them consistent with scenarios used by the rest of the industry for requalification examinations.
All scenarios in the bank were reviewed by the facility prior to submittal; however, this review did not appear to be thorough.
The scenarios contained ISCTs that did not meet the criteria in the Examiners Standards, did not contain all the items specified by the Examiners Standards, and required revisions to the performance standards and expected actions.
Facility personnel did identify some of these problems prior to the examination
preparation week.
A thorough review needs to be performed prior to submittal to ensure that the scenarios are up-to-date and meet the guidelines of the Examiners Standards.
The scenarios were not included in the Sample Plan/Test Outline.
As a result, there was some duplication of tasks between the walk-through examinations and the dynamic simulator examinations.
Consideration of the scenarios as part of the Sample Plan/Test Outline willalso facilitate coverage of previously identified weaknesses, industry events, procedure changes, etc..
3.5 Conclusion The licensee's licensed operator training program is considered to be satisfactory, based on the criteria established in NUREG-1021, However, the facility's evaluations on 'the dynamic simulator'ere determined to'be a programmatic weakness.
This weakness was related to the operator performance weaknesses in controlling RPV level and pressure with the reactor not shut down.
Additionally, the examination materials for all portions of the examinations required modifications to meet the Examiner Standards and to be consistent with industry practices.
Stand Li uid ntrol S stem Initiation 4.0 Emer enc eratin Pr cedures The examiner questioned the bases for a recent EOP revision during preparation for the examinations.
The RPV Control EOP was recently revised to require initiation of SLC whenever all rods are not inserted and reactor power is above 5%.
This revision adds a conservative action while still ensuring that SLC is initiated prior to exceeding the Boron Injection Initiation Temperature (BIIT) in the suppression pool in accordance with the Boiling Water Reactor Owners Group (BWROG) Emergency Procedure Guidelines (EPGs).
This change was made to ensure consistent implementation by the operators and to protect against rod pattern changes that could lead to fuel damage.
The licensee determined that the negative consequences of boron injection are not as great as the negative consequences of a Main Steam Isolation Valve (MSIV)
closure due to fuel failure.
They assured the examiner that they will not hesitate to initiate SLC in a real emergency.
4.2.
ontrol of In'ection ources with All Rods Not Inserted During scenario validation, the examination team identified a deficiency in the EOPs for control of injection into the RPV following rapid depressurization with all rods not inserted.
In a situation with all rods not inserted, but with
reactor power below 5% or no challenge to primary containment, the direction for RPV level control is not appropriate for limiting the addition of positive reactivity.
In this situation, ifRPV water level drops below the top of active fuel (TAF),
reactor pressure is above 150 psig, and an injection system is lined up with a pump running, rapid depressurization of the RPV is required by EO-111,
"Level Restoration."
EO-112, "Rapid Depressurization,"
requires prevention of all new injections ifall rods are not inserted.
As a result, all low pressure Emergency Core Cooling System (ECCS) pumps would have to be secured prior to rapid depressurization.
When RPV pressure drops below 150 psig, EO-111 directs the operator to start all pumps which are lined up.
Starting the low pressure ECCS pumps as directed would result in an uncontrolled injection of cold water into the RPV, adding positive reactivity when the reactor may not be shutdown.
Ifall rods are not inserted, reactor power is above 5%, and there is a challenge to primary containment (a safety relief valve (SRV) is open or drywell pressure is above 1.72 psig), entry into EO-113, "Level/Power Control," is required.
When EO-113 is entered, EO-111 is exited.
EO-113 requires rapid depressurization in accordance with EO-112 ifRPV level cannot'e.
maintained above TAF. EO-113 directs the operator to slowly increase injection to restore and maintain level above TAF following rapid depressurization and cautions against a rapid increase in injection.
This direction would preclude an uncontrolled injection of cold water while restoring RPV level above TAF. The addition of positive reactivity would be controlled and power excursions could be anticipated and mitigated.
The conflict in direction for RPV level control in these two situations results from a deviation that SSES has taken from the BWROG EPGs (Rev. 3).
The BWROG EPGs direct termination arid prevention of injection sources prior to rapid depressurization only when boron injection is required.
The conditions for entry into the Level/Power Control contingency are the same as the conditions that require boron injection.
Therefore, the level control guidance of Level/Power Control would be used under the same conditions that injection sources would be terminated and prevented.
Using this guidance, injection would be slowly increased following rapid depressurization.
The licensee reviewed this problem with the level control guidance and determined that the issue willnot be a problem when they revise their EOPs to follow Revision 4 of the BWROG EPGs.
The revised EOPs are scheduled to be implemented in September 1992.
In the interim, the licensee intends to train the operators to start pumps and inject in a controlled manner following rapid depressurization with the reactor not shutdown.
This training is scheduled as part of the training on sensitivity to parameter control to be covered during the first requalification cycle following the examination.0 Exit Meetin and onference Call An exit meeting was held at the conclusion of the examinations on January 9, 1992, in the facility training center.
The personnel in attendance are listed in Attachment 1.
Preliminary requalification program evaluation results were discussed.
The generic strengths and weaknesses of the examinees were presented.
The NRC complimented the licensee on the smooth administration of the examinations and their professional manner in dealing with the disagreements on the simulator examinations.
The licensee concurred with the NRC results and stated that they were already in the process of addressing the individual and programmatic weaknesses.
On January 16, 1992, a conference call was held between the NRC and the licensee to discuss the corrective actions being taken by SSES to address the weaknesses identified during the requalification program evaluation.
The conference call participants are listed in Attachment 1.
The licensee suspended the requalification simulator examinations immediately following the NRC administered examinations in order to provide training to the facility evaluators.
Simulator examinations would recommence at the completion of the training and evaluation of the evaluators by the Plant Superintendent.
The licensee was also performing a root cause analysis to determine the reason for the differences in the initial evaluations between the facility evaluators -and the NRC, to look for additional, programmatic problems, and to assess whether PP&L has remained 'current with industry standards.
Additional items in the facility's action plan include requalification training for the operators, internal and external evaluations of operator performance, training on and implementation of revised EOPs that incorporate Rev. 4 of the BWROG EPGs, and a preparation we'ek for facility evaluators in future examination ATTACHME<NT1 PERSONS CONTACTED Penn lvania P wer and Li h m an H. G. Stanley, Superintendent of.Plant (2), (5)
H. J. Palmer, Jr., Manager of Nuclear Operations (1); (2), (4)
W. Lowthert, Training Manager (2)
A. Fitch, Operations Training Supervisor (1), (2), (3), (4)
T.'R. Markowski, Dayshift Supervisor (2), (4)
J. Radishofski, Operations Outage Supervisor (2), (3),
(4)'.
Bartel, Operations Technical Support Supervisor (4)
T. W. Logsdon, Simulator Instructor (2), (3), (4)
F. Tarselli, Training Instructor (4)
J. Jones, Training Instructor (4)
D. Shaw, Training Instructor Nuclear Re ulato Commission M. Hodges, Director, Division of Reactor Safety (2), (5)
R. Conte, Chief, BWR Section T.'Walker, Senior Operations Engineer (1), (2), (3), (5)
S; Barber, Senior Resident Inspector (5)
D. Mannai, Resident Inspector (5)
B. Westreich, Reactor Engineer (2)
J. Raleigh, Project Manager, NRR (5)
W. Bateman, Executive Director of Operations Office L. Vick, License Examiner, OLB, NRR J. Hanek, Examiner (INEL) (3)
M. Riches, Examiner (PNL) (3)
N TES (1) Attended Entrance Meeting, November 6, 1991 (2) Attended Exit Meeting, January 9, 1992 (3) Member - Combined Facility/NRC Exam Team (4) Facility Evaluator (5) Participated in Conference Call, January 16, 1992
'
ATTACHMENT2 REQUALIFICATIONTEST ITEMS imula or Scenarios 121a:
Loss of RPS/DC - MSIV Isolatiori, illa: 'TWS Following a Turbine Trip Without Bypass Valves 119a:
Startup at 20% Power with Main Transformer Fire and ATWS 115a:
LimitCycle Oscillations After Recirc Runback ob Perf rmance Mea ure 200.070.01:
223.009.02:
200.152.04:
200.014.02:
262.003.06:
206.003.01:
211.005.02:
212.003.54:
218.002.01:
217.003.01:
Initiate SGTS A on Zone I and Zone IIIin accordance with ES-070-001 Start the Containment H2 Recombiner in Manual in accordance with OP-.173-001 Bypass the HPCI High Drywell Pressure Initiation Signal in accordance with ES-152-001 Establish and Maintain Reactor Vessel Level (RCIC Injecting) from the RSDP in accordance with EO-200-009 (Unit 2)
Place the Vital AC UPS Alternate Source In Service Using the Manual Bypass Switch in accordance with OP-157-001 Override an Inadvertent Start of the HPCI System in accordance with OP-152-001 C
Initiate SBLC System in accordance with OP-153-001 Bypass an APRM Channel F Trip Input to the RPS (Faulted)
Perform Manual Operation of the ADS from Panel 1C601 in accordance with OP-183-001 Perform RCIC System Manual Startup, Component by Component, in accordance with OP-150-001
Attachment 2 ATTACHMENT2 REQUALIFICATIONTEST ITEMS Written Examination Part A - Static imulator Part 8-las ro m MF22 OP002/MF22/R 003 OP002/MF22/R 007 OP002/MF22/R 008 OP002/MF22/R 011 OP002/MF22/R 014 OP002/MF22/R 001 OP002/MF22/R 013 (RO Only)
OP002/MF22/R 009 OP002/MF22/R 002 (RO Only)
OP002/MF22/S 002 (SRO Only)
OP002/MF22/S 001 (SRO Only)
~AS1
.
OP002/AS19/R 008 OP002/AS19/R 007 OP002/AS19/R 002 OP002/AS19/R 001 OP002/AS19/R 003 OP002/AS19/R 011 OP002/AS19/R 009 OP002/AS19/R 012 OP002/AS19/R 005 OP002/AS19/R 004 (RO Only)
OP002/AS19/R 013 OP002/AS19/S 002 (SRO Only)
OP002/004/S 001 OP002/200-105/R 001 OP002/200-007/S 0018 OP002/200-021/R 001 OP002/200-086/R 001*
, OP002/200/S 002*
OP002/300-024/R 001 OP002/200-037/S 001 OP002/299/R 001*
OP002/300/R 002*
OP002/215/S 001 OP002/004-002/S 001 OP002/271-002/R 002 OP002/263-006/R 001 OP002/262-013/R 001 OP002/245-005/R 001"'P002/218-003/R 001"'P002/234/S 002 OP002/261-001/R 001 OP002/223/R 001 OP002/215-002/S 001*
OP002/215-017/R 001 OP002/204-007/R 001 OP002/292-001/R 001*
OP002/205/R 001 OP002/200-060/R 001 OP002/200-104/R 001 OP002/200-007/S 001*
OP002/200-100/R 001 OP002/200-086/R 001"'P002/200/S 002*
OP002/300-025/R 001 OP002/300/R 002*
OP002/299/R 001*
OP002/300/R 001 OP002/215-002/S 001*
OP002/285-003/R 001 OP002/271-001/R 001 OP002/300-029/R 001 OP002/262-010/R 001 OP002/245-005/R 001"'P002/218-003/R001*
OP002/295/R 001 OP002/290/R 001 OP002/215-010/S 001 OP002/215/R 005 OP002/204/R 002
'P002/292-001/R001*
OP002/234-013/R 001 OP002/234-016/R 001
ATTACHMENT3 SIMULATIONFACILITYREPORT
, Facility Licensee:
Pennsylvania Power and Light Company Facility Docket No:
50-387 and 50-388 Requalification Examinations Administered on:
January 6 - 9, 1992 This form is used to report observations.
These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations.
ITEM
~H5 N
Failure to Scram/Stuck Rods The capabilities of the simulator for failure to scram events and failure of control rods to insert are very limited. The only options available are a full hydraulic ATWS and.failure of RPS which do not allow any rod insertion until the malfunction is removed, at which time all the rods insert.
There are only two rods that can be stuck so that they do not insert when the reactor is scrammed.
As a result, training and evaluation of operators on mitigation actions for failure of rods to insert is restricted to scenarios in which only two rods stick out which has minimal effect on the plant and scenarios in which the mitigation actions that will insert all the rods at once are the only success paths.
NOTE:
Pennsylvania Power & Light has ordered a new simulator which is expected to be delivered in the summer of 1992.