IR 05000029/1987002
| ML20214D188 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/12/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214D157 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-660, RTR-NUREG-737, TASK-1.C.1, TASK-1.D.2, TASK-2.B.1, TASK-2.F.1, TASK-2.K.3.01, TASK-2.K.3.05, TASK-2.K.3.10, TASK-2.K.3.12, TASK-2.K.3.25, TASK-TM 50-029-87-02, 50-29-87-2, GL-82-12, GL-82-33, IEIN-85-001, IEIN-85-045, IEIN-85-1, IEIN-85-45, IEIN-86-106, NUDOCS 8705210233 | |
| Download: ML20214D188 (72) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-29/87-02 Docket No.
50-29 Licensee No.
DPR-3 Licensee:
Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted:
January 27, 1987 - April 17, 1987 Inspectors:
H. Eichenholz, Senior Resident Inspector K. Gibson, Repident Inspector - Salem D. Wallac e
or Engineer S
Approved By:
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T. Elsasse p ef, Reactor Projects Section 3C
/Date Inspection Summary: Inspection on January 27 - April 17,1987 (Report No.
50-20/87-02)
Areas Inspected: Routine onsite regular and backshift inspection by resident and.
region-based inspectors (222 hours0.00257 days <br />0.0617 hours <br />3.670635e-4 weeks <br />8.4471e-5 months <br />). Areas inspected included review of licensee action on previous findings, operational safety verification reviews, review of radiological controls, reviews of events requiring telephone notification to the NRC, bimonthly safety system walkdown, review of plant events, maintenance obser-vations, review of licensee event reports, surveillance observations, onsite review committee activities, Plant Information Report reviews, review of excore nuclear instrumentation surveillance, licensee response to IE Bulletins, Information Notices, licensee action on NUREG 0660, review of the drug and-alcohol abuse pro-gram, licensee's response to the Surry feedwater line break event, review of the emergency exercise, and new fuel receipt and inspections.
Results: One violation was found involving the failure to provide an adequate re-view prior to implementation of a Temporary Change Request involving the No. 1 main coolant pressure channel (Section 8b).
Areas needing increased licensee attention involve the shift turnover practices (Section 4a(2)); processing temporary proce-dure changes (Section 4a(3)); and providing timely resolution of an NRC-identified issue involving the operability of the safety injection building ventilation system (Section 4a(5)).
Operator response to plant transient and emergency load reduction events (Section 8); fire protection, prevention, and housekeeping practices (Sec-tion 4c); training provided to members of the Plant Operations Review Committee pertaining to the safety review process (Section 12); and the drug and alcohol abuse program (Section 18) were considered notable licensee strengths.
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8705210233 870512
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PDR ADOCK 05000029 G
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-DETAILS 1.
~ Persons Contacted Yankee Nuclear Power Station B. Drawbridge, Assistant Plant Superintendent T.;Henderson, Technical Director N. St. Laurent, Plant Superintendent The inspector'also interviewed other licensee employees during the inspection ~,
including members of.the operations, radiation protection, chemistry, instru-ment and control, maintenance, reactor engineering, security,. training, tech-nical services, and general office staffs.
2.
Summary of Facility Activities
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At the completion of the last resident inspection period on January 26, 1987, the plant was at 100% of rated power.
Plant conditions remained stable until February 18, 1987, when an automatic reactor trip occurred on high main cool-ant system pressure.
This condition resulted from the plant operators de-creasing turbine load in response to a control rod that dropped from it's full out to full in position.
Later this same day, with the plant in Mode 3, an inadvertent safety injection occurred as a direct result of maintenance being performed on the No. 1 main coolant pressure channel instrumentation.
fol-lowing a reactor startup on February 19, 1987'the plant attained 100% of rated power on February 22, 1987. The plant entered Cycle XVIII end of cycle coast
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down operations on March 5, 1987.
On March 18, 1987 an emergency load reduc-
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tion from 92% to 73% of rated power was implemented by the plant operators in response to overheated packing on the inboard side of the No. 1 boiler feedwater pump.
The plant returned to normal ~coastdown operations on March 20, 1987, with the reactor power limited to 78% at the end of the inspection-period.
3.
Licensee Action on Previous Inspection Findings (Closed) Inspector Follow Item 50-29/85-18-04: Strengthen procedural controls associated with maintenance, surveillance and operations activities that are used to insure that containment integrity is being maintained.
The inspector reviewed the actions taken by the licensee to resolve the inspector's concerns in this area, and determined that the following' procedures were revised to address the issue: OP-3117, Rev. 12, Refueling Accidents; OP-5404, Rev. 6, Inspection of the Steam Generator-Secondary Side; OP-5454, Rev. 7, Inspection of the Steam Generator Feed Line Check Valves; and OP-4239, Rev. 8, Setting Vapor Container Integrity and Operability Check of the Vapor Container and Spent Fuel Pool Ventilation Systems.
The inspector had no further questions.
This item is closed.
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(Closed) Unresolved Item 50-29/86-08-02: Review LER 86-07 for root cause and corrective action. LER 86-07, Rev. 1, dated March 19, 1987 was submitted by the licensee and is discussed in Section 10 of this report.
This item is closed.
(Closed) Inspector Follow Item 50-29/80-20-01: Review licensee re-evaluation of block walls in accordance with IE Bulletin 80-11.
During this inspection period Special Inspection 50-29/87-01 was conducted by NRC:RI, which encom-passed review of licensee responses and subsequent analysis and modifications of masonry walls related to IE Bulletin 80-11, Masonry Wall Design.
Although the bulletin remains open, this inspector follow item is considered closed as a result of the current inspection establishing future NRC review require-ments for this bulletin.
4.
Operational Safety Verification Reviews a.
Daily Inspection During routine facility tours, the inspector checked the following items:
shift manning, access control, adherence to procedures and limiting con-ditions for operations (LCOs), instrumentation, recorder traces, protec-tive systems, control rod positions, containment temperature and pressure, control room annunciators, radiation monitors, radiation monitoring, emergency power source operability, control room and shift supervisor log, tagout log, and operating orders. No inadequacies were identified except as noted below.
(1) The inspector observed an anomalous condition on the Safety Para-meter Display System that involved a source range indication while the plant was at full power.
Inspection comments pertaining to this item are contained in Section 9 of this report.
(2) On February 6, 1987, a shift technical advisor (STA) did not report for duty as scheduled, resulting in the on-duty STA working 32 con-secutive hours.
With regard to licensee control of overtime, the inspector held discussion with licensee personnel and reviewed the following documents:
YAEC T.S. 6.2.2g
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NUREG 0737 (Item I.A.1.3)
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l YAEC OP-MEM0 2T-2 Policy for Overtime Work for Plant Personnel
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Who Perform Safety-Related Functions ANSI 18.7 - 1976 " Administrative Controls and Quality Assurance j
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for the Operational Phase of Nuclear Power Plants" t
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RG 1.33 Appendix A " Typical Procedures for Pressurized Water
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Reactors..."
The inspector identified the following concerns and discussed them with plant management: (1) OP-MEMO 2T-2 states that the overtime policy delineated in the memo does not apply.to STAS if the licensee provides them with sleeping accommodations, which is_ consistent with the NRC guidance for this item, as listed above.
However, the lic-ensee has not provided appropriate sleeping accommodations for STAS required to work overtime in excess of the policy. The licensee has committed to make such sleeping accommodations (cots, etc)
available to the S"; and (2) T.S. 6.8.1 and 6.8.2 require that written procedures be established by the licensee and that the pro-cedures be reviewed by the Plant Operations Review Committee (PORC).
Procedures on overtime control as delineated in ANSI 18.7-1976 Sec-tion 5.2 which is incorporated into T.S.6.8.1 by reference, are required to be PORC reviewed as indicated above.
It appears that OP-MEM0 2T-2 has not been reviewed by the PORC, and that this is an isolated oversight.
However, the licensee has committed to re-view OP-MEMO 2T-2 and the other OP-MEM0's to ascertain whether PORC review is applicable and to assure that this condition will not recur. The inspector had no further questions regarding this matter.
(3) During a backshift inspection on February 14, 1987, the inspector determined that a control room operator, who had recently come on
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shift, was not aware that Temporary Change Request (TCR) No. 86-72 had been implemented on December 31, 1986.
[This TCR involved plac-ing the high voltage power supply switch for source range neutron monitoring channel No. 2 in the off position, and was previously discussed in Section 4.a of Inspection Report 50-29/86-19.] This operator had not been on shift for quite some time as he was in a current licensing upgrade class.
Licensee corrective actions con-sisted of (1) counseling all shift personnel via Special Order No.
87-19 about the importance of shift turnover practices, (2) ensuring that the training department disseminates operationally related material to licensed candidates in the upgrade program since they may have to stand watches from time to time, and 3) providing ade-quate turnover time to watchstanders who have not recently stood watch and need to update themselves on the current plant configura-tion and operating conditions.
The licensee's actions were deter-mined by the inspector to be responsive to the NRC concerns on this matter.
(4) On February 19, 1987 at approximately 4:30 p.m. licensee prepara-tions were in progress for a plant startup following the shutdown
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that resulted from dropped control rod No. 2.
The licensee was l
preparing to warm up the No. 2 steam line to test the closing operation of the No. 2 non-return valve (NRV) by using the train
'A' control switch.
The licensee's action was the result of a re-ported observation that the control switch did not close the No.
2 NRV following the plant trip on February 18, 1987.
In an attempt
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to maintain main coolant system temperature, the operations depart-ment had developed a temporary change to procedure OP-2256, Rev.
9, Operation of the Main Steam System.
The inspector reviewed the licensee's intended actions and determined that the temporary change constituted on alteration of the original intent of the procedure.
In the inspector's view, this necessitated a Plant Operations Review Committee review, prior to implementation, as is required by TS 6.8.2.
Middle plant management above the shift supervisor, present in the control room at the time, disagreed with the inspector's determination; however, they cooperated with the inspector's request to hold up further implementation of the revised procedure until clarification could be obtained from NRC:RI management and the plant superintendent (PS).
The inspector immediately discussed this mat-ter with regional management, who concurred with the inspector's initial determination.
Shortly thereafter the inspector held a telephone conversation with the plant superintendent who advised the inspector that he had directed his staff to not implement the revised procedure and that he concurred with the inspector's deter-mination in this matter.
The PS initiated additional corrective action to counsel his staff on the need to provide the proper level of sensitivity when dealing with procedure changes and insuring that the proper reviews are conducted.
The inspector had no further questions regarding this matter.
(5) On April 6, 1987, the inspector noted a control room log entry that indicated at 8:50 a.m. the circuit breaker for the safety injection building vent fan, PRV-1, was tagged open for electrical maintenance.
By 10:45 a.m., the fan was returned to service. Similarly, the re-dundant fan PRV-2 was removed and returned to service at 11:35 a.m.
and 1:50 p.m., respectively.
In both cases there was no evidence in the control room log that the operator considered that they had entered the TS applicable action statement for the safety injection system.
In fact, when the inspector questioned the on-duty shift supervisor as to his views on the applicability of the TSs, he in-fomed the inspector that as long as one fan was operable he did not need to consider the TSs.
Consistent with this position was the existence of Special Order 87-07, in which the operations de-partment provided guidance to the plant operators on the actions they should take when both PR-V-1 and PR-V-2 fans are inoperable.
In Section 7.d of Inspection Report 50-29/86-19, the inspector raised concerns about the apparent need to acknowledge prior NRC concerns relative to single failure eventt., and the impact they i
could have on the building's ventilation.
In that inspection period, the inspector requested that the licensee resolve this is-sue.
At the time the issue was raised the outside ambient tempera-ture was significantly below the 40 degrees fahrenheit that the licensee's engineering staff has indicated that the ventilation system is required to be operable.
The ambient outside temperature was 43 degrees fahrenheit when the fans were removed from service in April 6, 198.
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At the March 11, 1987 meeting of the PORC, procedure OP-4269, Safety Injection Building Ventilation System Operability check, was re-viewed.
The PORC also discussed operability requirements and system configuration for the fans necessary to call ECCS operable.
The PORC requested the following actions that (1) a TS interpretation on safety injection building fan operability as a subsystem of ECCS
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be requested, and (2) a TS revision to include safety injection building fan operability be requested. According to the technical services manager, the issue of the PRV-1 & PRV-2 operability ques-tion was a subset of a much larger issue involving environment 01 qualification (EQ) requirements and their impact of the plant's TSs as a whole.
Accordingly, the PORC at it's February 7,1987 meeting had considered the issue of having an up-to-date EQ cross reference list to be important with the following actions being monitored:
(1) the EQ cross reference list is to be revised by Yankee Nuclear Power Services Division (YNPS) and accepted by the plant; (2) the revised EQ list is to be issued to the TS manuals; and (3) the operators are to be trained in the use of the EQ cross reference
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list.
The plant issued Service Request No. 86-62, on October 28, 1986 to re-evaluate the EQ Master List versus the TS cross reference.
The licensee's actions to address the entire issue of EQ and it's impact on TS operability determinations was appropriate, however, the PORC follow items developed at the March 11, 1987 meeting were
of sufficient concern that action should have preceded independently of the February 7, 1987 actions.
On April 7, 1987, the inspector held a discussion with the assistant plant superintendent (APS),
and requested that interim guidance be issued to the plant operators
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while evaluations on this subject are being conducted.
The opera-tions department issued Special Order No. 87-30 on April 8, 1987 which contained the interim guidance developed by the APS.
Another issue remaining to be resolved on this matter, which is of concern to the NRC, involves the design process that upgraded the safety injection building ventilation system for single failure concerns as it relates to the degree that (1) the design basis was a
translated into procedures instructions or drawings, and (2) the
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extent that TS operability and surveillance requirements were re-solved.
The design, licensing, and PORC responsiveness to an NRC
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identified issue involving the PRV-1 and PRV-2 fans is an unresolved
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item (UNR 50-29/87-02-01).
b.
System Alignment Inspection Operating confirmation was made of selected piping system trains.
Ac-cessible valve positions and status were examined. Power supply and breaker alignments were checked. Visual inspections of major components were performed. Operability of instruments essential to system perform-ance was assessed. The following systems were checked during plant tours and control room panel status observations:
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Charging system (control board status observations)
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Low pressure ana high pressure injection systems
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Motor driven emergeacy feredwater pumps
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Steam driven emergency feedwater pump
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Emergency diesel generator unit
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No unacceptable conditions were observed.
c.
Biweekly and Other Inspections (1) General Facility Observations
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During plant tours, the inspector observed shift turnovers, compared boric acid tank sample analyses and tank levels to Technical Speci-fications requirements, and reviewed the use of radiation work per-mits and radiation protection procedures.
Area radiation levels and air monitor use and operational status were reviewed.
Verifi-
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cation of tagouts indicated the action was properly conducted.
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I inadequacies were identified.
i (2) Fire Protection and Housekeepina No inadequacies were noted regarding licensee housekeeping or fire protection practices.
A strong commitment to proper housekeeping conditions and practices by the plant staff is routinely observed by the inspector.
Additionally, the inspector noted that required hotwork controls were properly implemented as part of repair acti-vities on March 19, 1987 on the No. 1 boiler feedwater pump.
(3) Observations of Physical Security Selected aspects of plant security were, reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures.
Additional details regardin the scope and findings of this review will be described in a sepa g rate special inspection report, NRC Region I Inspection Report No.
50-29/87-08.
5.
Review of Radiological Controls Radiological controls were observed on a routine basis during the reporting period.
Standard industry radiological work practices and conformance to
radiological control procedures and 10 CFR Part 20 requirements were observed.
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Independent surveys of radiological boundaries and random surveys of nonradio-logical areas throughout the facility were taken by the inspector.
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The inspector noted during this inspection period that plant personnel were exhibiting some apparent confusion pertaining to their responsibilities in completing information, e.g., exposure and completion time data on the Radi-ation Work Permits (RWPs).
The inspector brought this condition to the at-tention of radiation protection department supervision, and learned that this was the result of parallel testing of the access control feature of the Health Physics Information System (HIS) and the hand-written backup method. The HIS is a fairly new computer based system that is being utilized by the licensee to upgrade their recordkeeping capabilities associated with personnel exposure controls.
To clarify the intended personnel responsibilities during the testing effort, the radiation protection manager issued a memorandum to all department heads on February 13, 1987 that described the necessary require-ments for recording data during the parallel testing effort.
The licensee's corrective action was fully responsive to the NRC concerns.
The inspector had no further questions on this matter.
On February 6, 1987 the inspector participated in a requalification respirator fit quantitative test.
At this time, the operator of the fitting booth dis-cussed the licensee's respiratory protection program policies, which are con-tair.ed in AP-8422 Rev. 6. Respirator Fitting.
The topics covered by the fit booth operator were (1) ALARA philosophy and engineering controls before res-pirator issue; (2) issue and inspection; (3) donning and removal; (4) quali-tative and quantitative tests; and (5) emergency actions in event of respira-tor malfunction.
The inspector verified that the licensee's actions were in conformance with 10 CFR 20.103(c)(3).
The inspector reviewed AP-8012, Rev.
6, Respiratory Protection Training, and verified that training responsibili-ties assigned to the radiation protection department during respirator fit requalification activity was being conducted on a 24 month basis.
No deficiencies were identified in this area.
6.
Review of Events Requiring Telephone Notification to the NRC The circumstances surrounding the following events, which required NRC noti-fication via the dedicated ENS-line, were reviewed.
A summary of the inspec-tor's review findings follows or is documented elsewhere as noted below:
At 1:01 a.m. on February 18, 1987, the NRC was notified in accordance
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with 50.72 (b)(2)(ii) of an automatic reactor scram that occurred at 12:01 a.m. as a result of high main coolant system pressure that occurred following a control rod drop event.
This event is discussed in Section 8.a of this report.
Although this event is a four hour report, the plant operators treated the notification as a one hour 50.72 non-emergency event.
On February 18,1987 at 11:54 a.m., with the plant in Mode 3, the licen-
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see notified the NRC in accordance with 50.72 (b)(2)(ii) of an inadvert-ent initiation of the ECCS system that resulted from performing mainten-ance on the No. 1 main coolant pressure instrumentation channel.
This event is discussed in Section 8.b of this report.
The inspector had no further questions on this matter.
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7.
Bimonthly Safety System Verification The inspector independently verified the operability of a selected engineered safety features (ESF) system by performing a complete walkdown of the access-ible portions of the system to:
Confirm that the licensee's system lineup procedures match plant drawings
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and the as-built configurations; Identify equipment conditions and items that might degrade performance;
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Inspect interior of cabinets for abnormal conditions;
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Verify instrumentation lineup and calibration; and
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Verify proper valve position, availability for function and position
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indication.
The Emergency Core Cooling system was reviewed. No inadequacies were identi-fled.
8.
Review of Plant Events a.
At 12:01 a.m. on February 18, 1987, control rod No. 2 dropped from full out to full in.
The plant was at 100% of rated power at the time of this event.
No control rod manipulations were in progress at the time.
In-itial operator response consisted of reducing plant load in response to the dropped control rod to restore T~ average to its pre-transient value.
The control room operator action was in accordance with the immediate operator action requirement spect f fed in procedure 09-3118, Rev. 8, Mis-positioned or Dropped Control Rod (s).
Subsequently at 12:03 a.m., the reactor protection system initiated an automatic scram in response to a high main coolant pressure condition.
This condition resulted from the secondary side control room operator overcompensating in the genera-tor load reduction for the power reduction from the dropped rod.
The inspector verified that timely notification of the event was made by the licensee to the NRC via the ENS line.
Plant operators implemented Pro-cedure OP-3000, Rev. 26, Emergency Shutdown From Power, following the automatic talp from power.
The inspector reviewed the completed post trip review report (Scram No.132) for this event, which included the sequence of events recorder and major plant parameter strip recorder charts, and noted that there were no off-normal or unexpected plant re-sponses as a result of this event.
The licensco initiated corrective maintenance for the dropped No. 2 con-trol rod while maintaining the plant in Mode 3.
The cause of the event was attributed to an open circuit in the stationary gripper coil for the
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l No. 2 control rod, with the corrective action consisting of replacing the coil stack for this control rod. The licensee informed the inspector
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that this was a first time occurrence for this type of an event whereby
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the rod dropped from full out to full in.
The licensee's analysis of this event concluded that the plant response to the dropped control rod l
was bounded by both the FSAR analysis and the Core Performance Analysis Report.
Appropriate procedures were verified by the inspector to have
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been implemented for all operational and maintenance activities associ-ated with this event.
No deficiencies were identified by the inspector
during the review of this event.
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b.
At 11:24 a.m. on February 18, 1987, with the plant being maintained in Mode 3 - hot standby, an inadvertent initiation of the safety injection I
and non-essential containment isolation systems occurred.
There was no l
actual injection into the main coolant system.
This event was the result i
of implementing maintenance activity associated with Maintenance Request 86-1251, issued on August 28, 1986, which was in response to the identi-fied inoperability of control room annunciator NC-9, that would not function when the main coolant pressure channel No. 1 normal / test switch
was placed in the test position.
Until repairs could be effected, the
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I&C department supervisors have been annotating the monthly performed procedure OP-4659, Rev. 9, Main Coolant System Pressure Channels Fune-
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i tional Test, with a note to have the I&C technicians to inform the con-trol room personnel that the respective annunciator will not operate when
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main coolant loop the No. 1 instrumentation is placed in the test mode.
i The inspector noted that, due to the importance that the I&C department j
places on the operability of annunciation of test bypass conditions for
safety related functions, this plant shutdown was envisioned to be an
opportune time to perform the repair activity.
The I&C department initiated the Temporary Change Request (TCR) procedure, f
which was to be performed in accordance with procedure AP-0018, Rev. 12,
Temporary Change Control, to control the maintenance activity associated with replacing the test switch MC-TS-100 on the No. 1 main coolant pres-
sure channel.
The event occurred when the technicians lifted the main coolant pressure transmitter signal lead, which simulated a zero pressure.
The technicians performing the work immediately restored the main coolant
i pressure channel to operable status, with control room operators verify-ing that all systems responded properly to the loss of main coolant pres-
sure indication and that the actuation was inadvertent.
Upon performing the proper verifications the plant operators returned the systems to
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normal status at 11:28 a.m., with the NRC notified via the ENS in a j
timely manner at 11:54 a.m.
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The inspector reviewed the TCR, No. 87-41, and noted that it stipulated that the temporary change will not affect an operable system and that j
the applicable TS Section is 3.3.1 although the channel is not required
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during Mode 3.
The two factot contributing to this event were that (1) the TCR, as developed by the I&C department, did not identify the
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TS 3.3.2 Table 3.3-2 operability requirement for the ECCS and contain-i ment isolation systems associated with the subject instrument channel, and (2) the TCR us not properly reviewed by the control room operator and the shift supervisor in that they did not identify that the perform-i
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ance of the TCR would place the plant in the Action Statement of TS 3.3.2, Table 3.3.2.
Recognition by the I&C department and/or the plant operators that the intended lifted lead activity would result in a simu-lated low main coolant pressure condition and system actuation, could have allowed the maintenance to be performed if (1) the channel was con-sidered to have been inoperable within the constraints of the applicable TS Action Statement, and (2) the safety injection auto start / auto cutout switch for Train A was placed in the auto cutout position.
As required by 10 CFR 50.73, the licensee reported the inadvertent actu-ation in Licensee Event Report 50-29/87-04, dated March 20, 1987.
How-ever, due to the significance of the personnel errors that resulted in an unnecessary challenge to plant safety systems, this item is being cited.
Specifically, the failure of the licensee to (1) detect that the implementation of TCR No. 87-41 on February 18, 1987 would affect an operable system, (2) identify the applicable TS that should have been complied with, and (3) specify the use of the safety injection auto /
start-auto cutout switch for Train A as part of the proper implementation of the TCR is considered a violation (VIO 50-29/87-02-02).
Additional concerns that the inspector had pertaining to this event, and has discussed with licensee representatives, involves the participation of control room personnel in the review aspects of the TCR process.
For this portion of the process to be successfully and safely accomplished requires control room personnel to insure that adequate time and atten-
tion is devoted to the TCR review.
c.
On March 18, 1987 at 3:42 p.m., an emergency plant load reduction from 92% to 73% of rated power was implemented by the plant operators in re-j sponse to an overheated packing on the inboard side of the No. 1 boiler feedwater pump (BFP).
Subsequent to the power reduction the pump was i
removed from service to effect repairs.
Because of excellent communica-tion and coordination between the plant operators and station support staff, including the STA, reactor engineering and chemistry departments, the emergency controlled load reduction was accomplished with the use of controls rods only.
With end of cycle main coolant system (MCS) boron essentially zero, boration of the MCS was a last resort.
The inspector observed that the plant operators were performing in a deliberate manner, but without undue haste, in an environment of potential incipient seal failure in which the inboard BFP packing could blow out at any moment and a plant trip ensue.
The inspector verified that appropriate operat-ing procedures were in use at the time in the control room, and that no
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operational limits were exceeded as a result of the event.
The plant
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returned to normal coastdown operations on March 20, 1987.
No inadequacies were identified by the inspector as a result of reviewing this event.
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9.
Maintenance Observations Th1 inspector observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administration and mainten-ante procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualification, radiological controls for worker protection, fire protection, retest requirements and re-portability per Technical Specifications.
The following activi'ies were in-cluded:
Maintenance Request (MR) 86-58, Safety Parameter Display System (SPDS) -
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Source Range Indication Display At Power with Channel De-energized.
MR 86-1251, No. 1 Main Coolant Pressure Channel Panalarm NC-9
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MR 86-1352, Nuclear Instrumentation Power Range Channel 6-Erratic High Spikes and Channel Trips.
i MR 86-1739, Nuclear Instrumentation Intermediate Range Channel 4 errone-
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ous Rod Stop and High Start-Up Rate Scram Alarm f
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MR 87-213, No. 2 Non Return Valve Train A - Failure To Close On Manual Signal MR 87-214, No. 2 Control Rod Dropped
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MR 87-317, No.1 Low Pressure Safety Injection Pump Air Circuit Breaker
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(ACB) - Investigate Closure Problem MR 87-347, No.1 Boiler Feedwater Pump Inboard Packing Failure.
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Based upon a review of licensee activities in this area the inspector noted the following:
Regarding MR 86-58, the inspector noted that an onsite audit of the in-
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stalled SPDS conducted by NRC:NRR identified that the source range neu-tron monitor reads onscale during power operations.
In it's letter FYR 87-016 to NRC:NRR dated February 12, 1987 the licensee responded to the NRC concerns and indicated that source range channel circuitry will be reviewed, with troubleshooting to be performed during the upcoming re-fueling outage that is scheduled to commence on May 2, 1987.
The in-spector had no further questions of the licensee on this matter at this time.
The inspector reviewed the licensee's repair activities associated with
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MR 87-317, which covered the corrective maintenance for the malfunction-ing of the air circuit breaker associated with the No.1 low pressure safety injection (LPSI) pump.
The inspector's review is documented in Section 11 of this report.
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As a result of maintenance activities being performed in response to MR
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86-1251 an inadvertent initiation of the safety injection system occurred.
Event details are discussed in Section 8 of this inspection report.
10.
Review of Licensee Event Reports Licensee Event Reports (LERs) submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implica-tions were indicated, and whether the event warranted onsite followup.
Additionally, on March 18, 1987 the inspector provided the licensee Attachment I, a detailed evaluation of LER quality.made by the NRC Office of Analysis and Evaluation of Operational Data using the basic methodology presented in NUREG/CR-4178, March 1985.
No inadequacies were identified, except as noted below.
The following LERs were reviewed.
Report Report LER No.
Date Date Subject 50-29/86-07 6/18/86 3/19/87 480V A-C Buses Cross-Tie Electrical Loading Rev. 1 Problem 50-29/86-08 6/19/86 7/18/86 No. 1 Main Coolant Pump Suction Valve Stem Failure
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50-29/86-09 6/28/86 7/16/86 Incorrect Overload Devices for Four Motor Operated Valves 50-29/86-10 6/27/86 7/28/86 Potential Loss of Shutdown Cooling 50-29/86-11 10/3/86 10/31/86 Containment Isolation Valves Missed Surveillance 50-29/86-13 10/4/86 11/3/86 Reactor Scram Due to Operator Error 50-29/87-01 1/5/87 2/5/87 Missed Surveillance - Technical Specifica-tion 4.1.2.3.b 50-29/87-02 1/6/87 2/5/87 Safety Injection Building Vent Fans PRV-1 and PRV-2 Inoperable a.
LER 50-29/86-07 and Revision 1: This LER was first reviewed in Inspection Report 50-29/86-08, Section 7, and documented the need for the licensee to submit a revised LER to describe the root cause and appropriate cor-rective actions to preclude recurrence of this type of event.
The in-spector reviewed the revision to the LER and noted that it was fully
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responsive to the NRC request.
The event discussed in the LER represents the licensee's failure to operate the facility in accordance with General Design Criteria-17 however, based upon the documentation provided in the LER the event is being classified as a licensee identified violation in accordance with 10 CFR 2, Appendix C, and a notice of violation will not be issued.
Additionally the inspector noted that long term corrective action con-l sisting of a major electrical modification was identified by the licensee i
to resolve potential overloading conditions on the 480V a-c electrical l
buses during cross-tie operations.
During the upcoming Cycle XVII-XIX refueling outage a new 2400V a-c circuit breaker will be installed that will enable each 480V a-c bus to be fed directly from its own 2400V a-c bus, which will eliminate the need to tie the 480V a-c buses together.
The inspector had no further question pertaining to the LER.
b.
LER 50-29/86-08: Documentation of the licensee's corrtctive actions and inspection findings is contained in Inspection Report 50-29/86-08, Sec-tion 7.
c.
LER 50-29/86-09: Documentation of the licensee's corrective actions and inspection findings is contained in a special Inspection Report 50-29/
86-09.
d.
LER 50-29/86-10: Documentation of the licensee's corrective actions and inspection findings is contained in Inspection Report 50-29/86-08, Sec-tion 7.
e.
LER 50-29/86-13: Documentation of the licensee's corrective actions and inspection findings is contained in Inspection Report 50-29/86-08, Sec-tion 7.
f.
LER 50-29/87-01: This LER involved the licensee's discovery that TS 4.1.2.3.b surveillance requirement was not performed within the required 31 day interval.
This surveillance involved verifying the correct posi-tion of a valve in the boron injection flow path from the boric acid mix tank.
The inspector reviewed the LER and the event details, and noted some difficulty in verifying the accuracy of certain licensee statements.
This difficulty involved verifying exactly which TS surveillance re-quirement was missed.
The inspector's questions were brought to the attention of the technical services manager, who acknowledged the com-ments and concerns, and indicated that, if warranted, an updated LER would be submitted to correct the record.
Notwithstanding these condi-
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tions, the inspector verified that the licensee's proposed corrective actions were both appropriate to the circumstances and implemented as stated in the LER.
As a result of this event, but prior to the issuance of the LER, the inspector reviewed the licensee's practices associated with having the control room operator perform an operability determination for the af-i i
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fected system at the time of detection of a missed surveillance.
Other than the TS 4.0.3 wording that the performance of a surveillance re-quirement within the time interval constitutes compliance with operabil-ity requirements for a Limiting Condition for Operation, the licensee has no explicit guidance for the operators on this matter. As a result, the inspector discussed this issue with the assistant operations manager,
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who issued on January 7, 1987 Special Order (S.0.) 87-04.
This 5.0.
instructed the operators to (1) consider the system incperable upon the identification of a missed surveillance and enter the associated action
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statement and (2) consider the system operable upon satisfactory comple-tion of the surveillance test.
Subsequently the inspector became aware of a March 11, 1983 NRC:NRR position on this subject that specified the need to consider that the applicable TS action statement is entered at the time the surveillance requirement should have been performed rather than at the time it is discovered that tests were not performed.
The inspector held discussions with licensee personnel, including the plant superintendent, about the NRC's position and learned that the plant would adhere to the NRC:NRR position.
This position, if routinely fol-lowed, was viewed by both the inspector and licensee as a potentially safety significant issue that warranted further questioning as to the appropriateness of the prescribed conduct of operation.
Essentially this concern was that as a result of untimely performance of a surveillance test a one hour shutdown period as prescribed in TS 3.0.3, could be in-itiated that increases the possibility of a plant transient which could ultimately require equipment needed to cope with the transient that was involved in the missed surveillance.
Accordingly, the inspector reviewed the situation with NRC:RI management and the NRC:NRR project manager (PM).
At this time, the inspector was informed by the PM that this subject was soon to undergo review before the NRC's Committee to Review Generic Re-quirements.
Based upon the intention of the NRC in the near term to review the current position and issue guidance in the form of a Generic Letter on how to deal with this situation, regional management instructed the inspector to inform the licensee that an interim licensee response to a missed surveillance other than the initiation of a one-hour plant shutdown was acceptable if (1) the previous operating history of the system has not resulted in the loss of confidence in its reliability, and (2) the operability of the affected system is verified in an expedi-
tious but safe manner.
The inspector had no further questions on this item at this time.
This LER remains open pending the licensee's disposition of the correct-ness of the reported information.
g.
LER 50-29/87-02: Documentation of the licensee's corrective actions and
inspection findings is contained in Inspection Report 50-29/86-19, Sec-tion 7.
However, inspector concerns pertaining to the manner in which the safety injection building vent fans are removed from service and viewed by the operators in terms of TS operability is contained in Sec-tion 4 of this inspection report.
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11.
Surveillance Observations The inspector observed tests or parts of tests to assess performance in ac-cordance with approved procedures and LCOs, test results (if completed), re-moval and restoration of equipment, and deficiency review and resolution.
The following tests were reviewed.
OP-4214, Rev. 12, Chemical Shutdown System Operability Check
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OP-7105, Rev. 10, Normal Operation of the Incore Flux Mapping System
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OP-4204, Rev. 34, Test or Special Operation of the Safety Injection Pumps
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ard Determination of ECCS Subsystem Leakage OP-4260, Rev. 3, Turbine Control Valve Exercise
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OP-4207, Rev. 22, Surveillance of the Station Power DC and AC Distribu-
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tion Systems and the Emergency Diesel Generators OP-4201, Rev. 13, Power Range Channel Calibration Heat Balance
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OP-4217, Rev. 12, Charging System Operability Test
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OP-4659, Rev. 9, Main Coolant System Pressure Channels Functional fest
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Based upon a review of the licensee activities in this area, the inspector noted the following:
a.
During the performance of OP-4204 on March 12, 1987, the surveillance identified that the ACB for the No. 1 LPSI pump failed to stay closed (or latched) on the first two attempts.
On the third and forth attempts the ACB was successfully closed.
At 11:15 a.m., this date, the surveil-lance was considered complete.
However, the licensee issued MR 87-317 to investigate the closure problem with the ACB.
At 12:30 p.m. on March 12, 1987 the No. 1 LPSI pump was removed from service for inspection.
The plant operators were utilizing the seventy-two hour time period in Action Statement a of TS 3.5.2.
Maintenance activities were performed in accordance with plant procedures and involved (1) inspection and cleaning of the ACB, and (2) checking the control room located remote hand switch for grounded conditions.
An anti pump relay was found de-fective, but this was not related to the observed malfunction of the ACB.
At approximately 5:30 p.m., following a successful start of the No. 1 LPSI pump, a second attempt to close the ACB resulted in another failure to stay closed.
Additional inspections were conducted by licensee main-tenance personnel, which included cleaning off carbon deposits on the ACB's 52x contacts.
No abnormalities of the ACB operating mechanism were noted.
Subsequently the ACB was operated in the test position seven times successfully by using the control room remote hand switch.
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lowing placement on the bus, the ACB was operated twice with no further problems.
The No.1 LPSI pump was declared operable at 7:10 p.m. on March 12, 1987.
The inspector discussed the event and corrective actions with the plant's maintenance manager, and learned that no additional actions as a result of this event were planned.
The licensee was unable to provide the in-spector with a high level of assurance that they isolated the root cause of the ACB's failure to stay closed.
The inspector reviewed the circum-stances of the occurrence and licensee corrective measures with NRC:RI management.
Based upon NRC concerns, the inspector held a discussion with the assistant plant superintendent to request the licensee's con-sideration of increasing the surveillance test frequency on the No. 1 LPSI pump.
On March 19, 1987, the No. 2 LPSI pump was scheduled for a surveillance test run, and it was envisioned that (nother successful run of the No. 1 LPSI pump would provide an increased level of assurance that the maintenance performed on March 12, 1987 had in fact corrected the equipment problem.
The NRC agreed with the licensee that following a successful operation of the No. 1 LPSI pump ACB it would be unnecessary to perform additional testing on this ACB.
The test of the No. 1 LPSI pump was successfully completed on March 19, 1986.
The inspector had no further comments on this item.
b.
A review of nuclear instrumentation channel calibration and functional test surveillances was conducted during this inspection period and is documented in Section 14 of this inspection report.
No violations were identified.
12. Onsite Review Committee Activities On March 19, 1987 the inspector observed the meeting of the Yankee NPS onsite review committee to ascertain that the provisions of TS 6.5.1 were met.
As a result of reviewing the licensee's activities in this area, no violations were identified.
The licensee had become aware, via their director of industry affairs, of an 5 NRCspecialsafetyinspectionthatwasconductedataRegionIIIfacility.
This inspection focused on the adequacy of that licensee s safety reviews performed pursuant to 10 CFR 50.59, and identified concerns with their imple-mentation of requirements. Subsequently, the technical services manager transmitted the subject NRC inspection report to all PORC members for their review,andinitiatedactiontoscheduleatrainingsessionforthePORC members on the 10 CFR 50.59 process.
The licensee s Manager of Engineering, who is presently involved in industry subcommittees that are dealing with the 10 CFR 50.59 process, conducted the training on January 9, 1987.
The inspec-tor reviewed the documentation associated with this training, and noted that handouts were used that specifically highlighted key requirements of 10 CFR 50.59 and provided generic and plant specific examples of the safety review i
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process.
The inspector viewed the licensee's actions in this area as a post-tive indication of the importance that they place on maintaining a high level of performance for their safety review process.
13.
Plant Information Reports Plant Information Reports (PIRs) prepared by the licensee per AP-0004 were reviewed.
The inspector determined whether the conditions were reportable as defined in the TS and whether the licensee's system of problem identifica-tion and corrective action is being effectively utilized.
The following PIRs were reviewed:
Occurrence Report PIR No.
Date Date Subject 86-01 1/9/86 2/27/86 Partial Loss of Control Air to the Primary Auxiliary and Waste Disposal Buildings 86-02 9/11/86 9/17/86 Failure to Secure Chemical Sampling of the Neutron Shield Tank 86-03 9/23/86 12/23/86 Inadvertent Release of the High Pres-sure Cardox Fire Suppression System for Manhole No. 3 PIR 86-03: This event was described in Inspection Report 50-29/86-08.
In-spector concerns contained in this inspection report were addressed within the corrective action section of the PIR.
The inspector determined that the licensee was responsive to NRC concerns associated with this event.
There were no further questions of the licensee on this matter at this time.
The inspector identified no violations regarding the licensee's actions as-sociated with these events, and noted that the licensee determined the cause of the occurrence and spe:ified appropriate short term and long term correc-tive actions.
14.
Excore Nuclear Instrumentation Surveillance During this inspection period a region-based inspector conducted a safety review and onsite inspection of nuclear instrumentation (NI) channel calibra-tion and functional test surveillances that are related to a proposed change (PC #201) to TS Table 4.3-1.
The safety review was initiated by Region I in response to a request for licensing action review by NRC:NRR.
This item re-suited from an inspection finding that is documented in Inspection Report 50-29/85-24, Section 7.
The change involves clarifying TS surveillances that require a monthly channel functional test and quarterly channel calibration of the power range NI low power setpoint trip function, a function that cannot be performed at power due to the fact that a reactor trip would result.
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The review consisted of discussions with station personnel, procedure walk through by the inspector while accompanied by an Instrumentation and Controls (I&C) technician, and document reviews.
The following documentation were reviewed by the inspector and used to evalu-ate the adequacy of the proposed TS change:
Westinghouse Technical Manual 40-Y-12101B, Chapter 5, Nuclear Instrumen-
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tation.
YNPS Training Manual, Section 31, Excore Nuclear Instrumentation, Revi-
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sion 1, 10/86.
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YNPS Final Safety Analysis Report, Section 214, Nuclear Instrumentation
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and Reactor Protection System, 7/12/85.
Standard Technical Specification For Westinghouse Pressurized Water
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Reactor, Revision 4, 1981.
Procedure OP-4645, Rev. 5, Nuclear Instrumentation Channel Calibration.**
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Procedure OP-4601, Rev. 18, Nuclear Instrumentation Channels Functional
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Test, Revision 18, 10/86. **
Procedure OP-4201, Rev. 13, Power Range Channel Calibration Heat Balance.**
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An ** indicates walk through of procedure was conducted by the inspector while accompanied by an I&C technician, i
In addition, the walk through inspections included an evaluation of certain references, prerequisites, and precautions as well as procedural steps.
Pro-
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l cedure review of operations procedures consisted of a verification of the following attributes:
l Proper review and approval
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Appropriate format
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Conformance to TSs
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Technical accuracy
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Adequacy of prerequisites and precautions
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Necessary detail to ensure adequate procedure implementation.
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The inspector was informed by the Ifconsee that portions of the nuclear in-
l strumentation system (NIS) may be replaced during the 1988 refueling outage.
The TS change roquested under PC #201 may not be appropriate for any new NIS components.
Thorofore, the frequency of surveillance and extent of the func-
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tional test and calibration of those channels that may be replaced during the 1988 outage must be reevaluated prior to the replacement, taking into consi-deration the guidance provided in NUREG-0456 (Standard Technical Specifica-tions), as specified in 10 CFR 50.59, paragraph c.
In licensee Memorandum YRP 103/87, dated February 9, 1987, the Yankee Nuclear Services Division pro-ject engineer described their intended design approach that will be part of Engineering Design Change Request No.86-310, Nuclear Instrumentation Upgrade.
This approach involves testability and calibration capability that will en-compass the full range of each channel's intended operation.
The inspector found the operators and I&C personnel to be very knowledgeable and familiar with station procedures.
Station surveillance procedures were complete, and observed to be conservative with respect to safety and work practices.
Discussions concerning the licensing action review for the pro-posed technical specification change will be addressed separately in the NRC Safety Evaluation Report for the proposed change.
No violations were identified.
15.
Licensee Response to IE Bulletins The licensee's response to the following IE Bulletin (IEB) was reviewed.
This review included: adequacy of the response to IEB requirements, timeliness of the response, completion of identified corrective actions and timeliness of completion.
IE Bulletin No. 85-01, steam binding of emergency feedwater (EFW) pumps.
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This bulletin was issued to inform licensees of serious safety problems which occurred at certain operating facilities concerning the inopera-bility of EFW pumps due to steam binding.
This issue was reviewed in Inspection Reports 50-29/85-11 and 86-08, Sections 14 and 12, respec-tively.
As a result of the inspector's review and comments on this issue, the bulletin remained open pending the licensee's development of proce-dures for monitoring fluid conditions with the EFW system and providing instructions to notify the shif t supervisor upon discovery of of f normal conditions.
The inspector review station procedure AP-2007, Rev. 19, Maintenance of Operations Departmental Logs and the primary and secondary auxiliary operator logsheets.
These documents conform to the bulletin requirements and satisfactorily addressed inspector concerns enumerated during the prior inspection on this issue.
This bulletin is closed.
16.
Information Notices During this inspection period, as requested by the NRC's Office of Inspection and Enforcement (IE), the inspector reviewed the actions taken by the licensee in response to IE Information Notice No. 85-45: Potential Seismic Interaction Involving The Movable In-Core Flux Mapping System Used In Westinghouse Design Plants.
This inspection was conducted in accordance with IE Temporary In-
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l struction 2500/16, Inspection to Determine If A Potential Seismic Interaction Exists Between The Movable In-Core Flux Mapping System and Seal Table at Westinghouse Designed Facilities or Facilities with Similar Designs, issued
on June 13, 1986.
Information Notice 85-45 was received at the site and assigned for licensee review and followup.
Based upon the licensee's review of the information notice, and additional information that they obtained from Westinghouse, the licensee determined that the potential seismic interactions are not applicable to the YNPS design.
The station's design uses reactor vessel top mounted instrumentation.
No unacceptable conditions were identified by the inspector as a result of this inspection.
17.
Licensee Action on NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident
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l The NRC's Region I office has inspection responsibility for selected action l
plan items.
These items have been broken down into numbered descriptions
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(enclosure 1 to NUREG-0737, Clarification of TMI Action Plan Items).
Licensee letters containing commitments to the NRC were used as the basis for accept-
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ability, along with NRC clarification letters and inspector judgment.
The j
following items were reviewed:
I NUREG-0737, II.B.1, Reactor Coolant System (RCS) Vents
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This item was discussed previously in Inspection Reports 50-79/83-15, 84-07, and 85-15. The latter inspection results noted that a TS change would be required to reflect intended system operation with the valves in a de-energized condition. On October 15, 1985 the licensee submitted Proposed Change No.195 to the TSs to reflect their intention to operate the plant with the valves in a de engerized state.
Subsequently, the licensee was requested by NRC:NRR to submit a request for exemption to the 10 CFR 50.44(c)(3)(iii) requirement that the reactor coolant system vents must be remotely operated from the control room.
This exemption was requested by the licensee on October 3, 1986, with NRC:NRR granting the exemption on December 29, 1986.
Subsequently, on January 7, 1987, the NRC issued Amendment No. 102 to the facility operating license, that modified the TSs regarding the reactor coolant system vent system by allowing power to be removed from the valves during normal plant opera-tions.
This item remains open pending verification of corrective actions during a subsequent inspection.
NUREG-0737. I.C.1.3.B. Revise Procedures for Transients & Accidents
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This item required the licensee to reanalyze transients and accidents and prepare Technical Guidelines, with the associated requirements spect-fled in Supplement 1 to NUREG-0737. (Generic Letter No. 82-33).
In it's letter FYR 83-91 to NRC:NRR on October 21, 1983, the licensee specified that it would implement revised emergency operating procedure (EOPs) at
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the start of cycle XVII operation (June,1984).
Subsequently on February
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22 2, 1984, the licensee submitted it's Procedure Generation Package (PGP)
for the E0Ps.
On May 9, 1984, the NRC Staff issued a letter to the lic-ensee that identified a number of problems in the PGP of sufficient mag-nitude that warranted a delay in the implementation of the upgraded E0Ps until the problems were resolved.
A revised schedule to implement the upgraded E0PS at the start of the cycle XIX operation (June, 1987) was committed to by the Itcensee in its July 27, 1984 letter (FYR 84-82).
In response to the licensee's proposed schedule, the NRC issued an im-mediately effective order on July 5,1985 that confirmed the licensee's commitment for the emergency response capability associated with the upgrade of E0Ps.
The licensee then submitted on August 18, 1986 (FYR 86-076) a draft revised PGP that incorporated additional information as a result of a continuing program and in response to previously identified NRC concerns.
As part of this submittal, the licensee recommended to the staff that a meeting be held to discuss the details of their program development and allow themselves an opportunity to respond to questions that might arise from a preliminary staff review.
No meetings occurred in response to the licensee's suggestion.
In accordance with their com-mitment, the inspector verified during the current inspection period that training activities were being conducted by the licensee in preparation for the full implementation of the E0Ps prior to the start of Cycle XIX operation.
During the preparations for operator licensing examination to be con-ducted by NRC:RI personnel that were to be administered at the YNPS on April 28, 1987, a meeting between the licensee NRC:NRR staff, and NRC:RI personnel was held in Region I on April 2, 1987.
It was the intent of this meeting to allow the licensee to present their E0P program.
As a result of this meeting questions were raised concerning advisability of allowing the licensee to implement the E0Ps as planned.
NRC concerns pertaining to the E0Ps is contained in Meeting Summary Report 50-29/87-038.
Current NRC plans call for extending the schedule for implementa-tion of the new E0Ps.
Pending further NRC:NRR staff review and actions, this item remains open.
NUREG-0730. I.D.2.2 and I.D.2.3 Install and Implement an SPDS
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The SPDS hardware installation was completed in September, 1982.
On September 1, 1983, the licensee submitted with its letter FYR 83-82 a safety analysis report which was subsequently reviewed by the NRC, with the staff issuing its safety evaluation report (SER) on December 17, 1984.
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The NRC staff's conclusion was that it was acceptable for the licensee to continue implementation of the SPDS program provided that certain actions were taken, which included the need for submittal of additional information.
This information was submitted by the licensee, in their letters FYR 85-47, and FYR 85-95, dated April 8, and September 3, 1985, respectively.
In April, 1986 an on-site audit of the installed SPDS was conducted by NRC:NRR which resulted in the NRC staff issuing a supple-mental SER on November 18, 1986.
This SER concluded that the SPDS at the YNPS will meet the requirements of NUREG-0737, Supplement 1 when
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(1) written commitment that the relative position, orientation and visual access of the containment isolation status panel with respect to the SPDS station will be maintained, and (2) minor concerns relating to.the human factors engineering of the SPOS are addressed.
These items were ad-dressed in the licensee's letter FYR 87-016, dated February 12, 1987 to NRC:NRR.
This item remains open pending the NRC staff's acceptability of the above licensee response.
In addition, the NRC conducted an Emergency Response Facilities Appraisal at the YNPS during the period March 30 - April 3, 1986 that included an assessment of the SPDS.
The NRC's appraisal is being documented in Inspection Report 50-29/87-05.
NUREG-0737, II.F.1.2.B.2, Accident Monitoring - Implement
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Iodine / Particulate Sampling This item was reviewed during a Post Accident Sampling System Team In-spection 50-29/85-05 conducted on March 4-8, 1985.
This inspection identified that the licensee's system for sampling and analysis of plant l
effluents does not meet the NRC guidance given on NUREG-0737, II.F.1, i
Attachment 2 in regard to (1) providing continuous particulate and iodine sampling, (2) providing isokinetic sampling of particulates and iodines,
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l (3) provide representative sampling per the criteria of ANSI N13.1-1969 and Regulatory Guide 1.97, and (4) providing backup analysis capability.
t Prior to the NRC inspection, the licensee had concluded that the system lacked sufficient suitability and planned on resolving these deficiencies by the installation of a completely redesigned system.
The licensee's Engineering Design Change 84-326, Vent Stack High Level Iodine Sampler, was installed on April 11, 1985.
Subsequently, the licensee used proce-dure OP-9450, Rev. 5, Post Accident Sampling and Analysis, as verifica-tion of correct system operation and a training tool for sampling per-
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sonnel.
In the licensee's letter FYR 85-66, dated June 18, 1985, sub-mitted to NRC:RI the licensee committed to installing the new system by
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August 1985 and have it tested and verified operable by the end of the l
l year.
The inspector verified that procedure OP-9450, Section E, contains the instructions necessary to obtain a primary vent stack effluent sample.
This item remains open pending NRC RI specialist inspector followup of Unresolved Item 50-29/85-05-05.
NUREG-0737, II.F.1.6, Accident Monitoring-Containment Hydrogen
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This item was previously inspected in Inspection 50-29/85-14, Section 15.
The item was maintained in an open status pending the licensee's evaluation relative to the use of the Bendix analyzer as part of its commitment to provide redundant channel capability for the containment monitoring system (CHMS).
The inspector's concern involving this item involved the fact that the sample residues from the Bendix analyzer neither returned to the containment nor a closed system.
Although there are no specific requirements pertaining to this aspect of operation in
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NUREG-0737, Item II.F.1.6, the licensee was requested to provide assur-ance to the inspector that off-site dose at the exclusion area boundary would not exceed regulatory limits when the Bendix analyzer was in operation during post accident conditions.
The inspector reviewed licensee memorandum YRP 1120/86, dated November 28, 1986, which documented the results of analysis conducted by the YNSD Radiological Engineering Group, that, demonstrated the Bendix analyzer can begin operation at 30 minutes post-LOCA without exceeding either GDC 19 limits for control room dose or 10 CFR 100 limits for off-site dose.
Furthermore, this memorandum reflected the completion of an evaluation of the licensee's licensing commitments pertaining to the use of the Bendix analyzer.
A subsequent licensee memorandum YRP 117/86, dated February 7, 1986, provided environmental / performance evaluation of the Bendix analyzer.
The inspector noted that appropriate recommendations were made in this memo to update the Environmental Qualification (EQ)
manual, however a check of the I&C department's instrumentation folder for the analyzer did not identify any EQ requirements. The inspector's concerns were brought to the attention of the maintenance support super-visor and the assistant I&C department supervisor.
They indicated that the appropriate documentation would be added to the file to ensure that future maintenance on the Bendix analyzer would not result EQ operability concerns to either the analyzer or nearby electrical equipment in the switchgear room.
Based upon the licensee's evaluation with respect to the use and design of the Bendix analyzer, this item is considered closed.
NUREG-0737, II.K.3.5 Automatic Trip of Reactor Coolant Pump (RCP)
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The NRC:NRR has refiewed the licensee's analysis and the justification and criteria for manual RCP trip as part of the solution to the small-break loss-of-coolant (LOCA) accident.
On May 6,1986, NRC:NRR issued a SER that cetermined that the information provided by the licensee for the justification of manual RCP trip is acceptable. Furthermore, it was stated that the methods employed to justify manual reactor coolant pump trip are consistent with the guideline and criteria set forth in Generic Letter 83-10.
The pump trip criteria for the YNPS specifies that the offsite powered reactor coolant pumps be manually tripped in a loss of subcooling margin (setpoint of 35 degrees F) coincident with a 5 psi containment over pressure.
The licensee's criteria also addresses non-LOCA events.
In its letter to NRC:NRR FYR 85-90 dated August 23, 1985, which was submitted in response to Generic Letter No. 85-12 Implementation of TMI Action Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps", the licensee proposed an integration between this item and Item I.C.1 for incorporating procedural controls and training plant operators.
As in-dicated in the section on Item I.C.1.3.B, the NRC plans on extending the implementation date for the upgraded E0Ps.
Based upon this intended action, the inspector discussed the implementation plans for incorpora-tion of the RCP trip criteria with the operations department manager on
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April 14, 1987.
The inspector was informed that procedure OP-3051, Rev.
11, Loss of Main Coolant Pressure and/or Safety Injection Initiation, would be revised to include the licensee's proposed RCP trip criteria.
Operator training and procedure implementation would occur prior to start-up for cycle XIX operation.
This item remains open pending verification of the licensee's revision-of OP-3051.
NUREG-0737, II.K.3.1, Automatic PORV Isolation
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As required by this item the licensee was to assess whether an automatic PORV isolation system was required.
On December 1, 1983, in their letter FYR 83-100 to the NRC:NRR,'the licensee adopted the Westinghouse Owners Group conclusion that the concept of an automatic PORV block valve clos-ure system, which closes the PORV isolation valves when lower pressure is sensed subsequent to the PORV failing to close, cannot be warranted on the basis of providing additional protection against a PORV LOCA.
In its December 13, 1983 SER, the NRC Staff concluded that an automatic PORV isolation system is not required at the YNPS.
This item is closed.
NUREG-0737, II.K.3.10, Proposed Anticipatory Trip Modifications
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On June 12, 1980, in their letter WYR 80-66 to NRC:NRR, the licensee confirmed that it has an anticipatory reactor trip on turbine trip which is effective at power levels above 15 MWe, they further stipulated that they have no plans to change this set point.
In their letter of December 15, 1980 to NRC:NRR, the licensee stipulated for this item that the item is not applicable to the YNPS since they have no plans for modification.
Subsequently on September 16, 1981, the NRC staff acknowledged that this item is resolved.
This item is closed.
NUREG-0737, II.k.3.12, Confirm Existence of Anticipatory Reactor
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Trip Upon Turbine Trip Confirmation of the existence of anticipatory reactor trip was provided-in the licensee's letter WYR 80-66 to NRC:NRR dated June 12, 1980.
An anticipatory reactor trip on a turbine trip is included in the original Westinghouse design and is incorporated in the existing YNPS TSs.
- This item is closed.
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NUREG-0737, II.3.3.25.B.2, Effect of Loss of Power on Pump Seals
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As stipulated in the licensee's letter to NRC:NRR on December 15, 1980, this TMI item is not applicable to the YNPS.
The inspector noted that the integrity of the RCPs is assured by the canned rotor design, which precludes the concern of consequences resulting from pump seal failure.
This item is closed.
18.
Drug and Alcohol Abuse Program A review was conducted by the inspector of the licensee's Drug and Alcohol Fitness for Duty Program (the Program).
This review included ascertaining the status of the various elements of the program, determining that policies have been established by licensee management, and that these policies have been incorporated into employee training programs and applicable plant proce-dures.
The licensee distributed their drug and alcohol policy to all employees on September 17, 1986.
This policy specifies the requirement for drug and alco-hol testing, stipulates De prohibitions dealing with drug and alcohol, and provides for the availability and use of an Employee Assistance Program.
The effective date of the program's implementation for permanent Yankee employees was October 1, 1986.
However, the inspector was informed by a licensee rep-resentative that the drug and alcohol testing portion of the program has been in effect for contractor employees for approximately two years.
The licensee has established three policy guidelines that reflect the program, and are contained in the licensee's Personnel Administrative Guideline Book.
The awlicable guidelines are (1) 8.2-Drug and Alcohol Fitness for Duty Policy, (2) S.3 Supervisor Guidelines for Administering the Company Policy on Drug and Alcohol Abuse, and (3) 8.4-Drug and Alcohol Fitness for Duty Testing Pro-cedures.
The administration of the Employee Assistance Program, which is available to all Yankee employees and their families, is provided by a con-tractor service.
This service provides personnel counseling, supervisor training and prevention programs.
The prevention programs involve stress management, burnout, and time management, and are to be implemented at a later time.
Additionally, the licensee's plant management supervisory and senior technical personnel attend fitness for duty and behavior observation training given by a behavior observation professional once per year.
The inspector reviewed plant procedure AP-0501, Rev. 13, General Plant Train-ing Program and its implementation, and verified that new and temporary em-ployees and contractors, as well as existing plant employees, are trained on the licensee's Drug and Alcohol Policy as part of the industrial safety com-ponent of the training program.
During the training, a copy of the Drug and Alcohol Policy is provided, with the person being trained signing that they have reviewed and understand this policy.
Plant procedure AP-0617, Rev. 2, Outside Contractor Access Information, specifies the requirement that con-
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tractors requesting unescorted access must submit written evidence that a drug and alcohol screen were passed.
The procedure contains the minimum and es-sential elements a laboratory conducting drug testing should follow.
No inadequacies were identified.
However, the documents referenced above were sent to the NRC:RI designated coordinator for further review.
19.
Steam, Feed and Condensate System Surveys The inspector was directed by NRC:RI Division of Reactor Projects to perform a followup of IE Information Notice 86-106, Feedwater Line Break.
The fol-lowup was to determine what actions the licensee may have taken as a result of the Surry Power Station feedwater line break.
Region I Temporary Instruc-tion No. RI-87-02, dated January 30, 1987, was used as guidance in performing the information notice followup.
The following inspection results were transmitted to NRC: Region I on February 4,1987:
a.
The licensee was in the process of developing a program to determine whether their extraction steam, heater drain, condensate, feedwater, auxiliary steam, and main steam piping is subject to thinning of the piping wall.
The program development was approximately 80% complete.
b.
The factors utilized by the licensee to select the potentially high sus-pect secondary plant piping for wall thinning were mainly carbon steel fittings and spools and fitting that were less than 10 pipe diameters apart.
However, fluid flow velocity was considered, c.
There were a total of 40 locations selected for inspection during the May, 1987 refueling outage that would involve the feedwater, main and auxiliary steam, condensate, heater drain, and extraction steam piping.
d.
Previous inspections were performed on the extraction steam piping, erosion / corrosion data was collected, and sections of piping were re-placed.
The inspector noted that on January 13, 1987, the licensee's President issued a memorandum to his senior managers that directed them to implement on an immediate basis a vigorous detailed study of the plant's system's as a result of the Surrey event.
On this same day, the licensee's project manager issued Service Request No.87-004 to the Manager of Engineering to plan for and direct the inspection program of secondary plant piping for erosion / corrosion during the upcoming refueling outage.
20. 4nergency Exercise The inspector participated in the review of the licensee's emergency exercise which took place on March 31, 1987.
This review included three major areas:
exercise preparation / review of scenario
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exercise observance
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The details of the inspector's comments and findings were presented to the
.NRC:RI team leader and is described in NRC Inspection Report No. 50-29/87-03.
21.
New Fuel Inspection The inspector reviewed the results of the new fuel receipt inspections per-formed by the licensee to verify that inspections had been performed in ac-cordance with OP 7200, Rev. 8, Receiving, Unloading and Inspecting New Reactor Fuel.
The inspection documentation for fuel shipment received on-site on April 13, 1987 consisting of 36 new assemblies was reviewed.
The inspector witnessed portions of unloading and storage of new fuel including Quality Control and Radiological Control involvement.
Shipping containers were noted to be properly sealed upon arrival. Personnel were knowledgeable in procedures
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governing the handling of new fuel.
The inspector informed the licensee's reactor engineering manager about a minor concern pertaining to inspection personnel activities prior to the com-pletion of radiological surveys of the shipment.
Immediate and effective licensee corrective measures were noted that consisted of issuing a Radiolo-gical Occurrence Report and revising procedure OP-7200 to advise inspection personnel of appropriate precautions.
The licensee's actions were timely and fully responsive to NRC concerns in this area.
The inspector had no additional comments.
22. Management Meetings
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During the inspection period, the following management meetings were conducted or attended by the inspector as noted below:
An onsite meeting was held on February 11, 1987 between NRC RI Division
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of Reactor Safety QA specialists and licensee representatives to discuss the licensee's planned corrective actions as a result of inspection findings documented in Inspection Report 50-29/86-17 The inspector attended a combined entrance meeting conducted by an NRC
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team on March 30, 1987 at the start of Inspection 50-29/87-03, review of the licensee's annual emergency plan exercise, and Inspection 50-29/
87-04, appraisal of the licensee's emergency response facilities.
On April 2, 1987, the inspector attended at the NRC Region I Office a
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licensee presentation on their new emergency operating procedures.
The results of this meeting is documented in Meeting Summary Report 50-29/
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At periodic intervalsiduring the-course of the inspection period, meet-
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-ings were'. held with senior facility management to discuss the inspection scope and.prelimina_ry findings' of the resident inspector.
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AE00 SALP INPUT FOR YANKEE R0WE -
OPERATIONS (LER QUALITY) FOR i
.THE ASSESSMENT PERIOD OF
j February 1, 1985 to September 15, 1986 j
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SUMMARY
An evaluation of the content and quality of a representative sample of the Licensee Event Reports (LERs) submitted by Yankee Rowe during the february 1, 1985 to September 15, 1986 Systematic Assessment of Licensee Performance (SALP) period was performed using a refinement of the basic methodology presented in a report entitled "An Evaluation of Selected Licensee Event Reports Prepared Pursuant to 10 CFR 50.73 (DRAFT)",
NUREG/CR-4178, March 1985. The results of this evaluation indicate that Yankee Rowe LERs have an overall average LER score of 8.5 out of a possible 10 points, compared to a current industry average score of 7.9 for those unit / stations that have been evaluated to date using this methodology.
The principle weakness identified in the Yankee Rowe LERs. in terms of safety significance, involves the requirement to adequately identify failed components in the text. The failure to adequately identify each component that fails prompts concern that possible generic problems may go unnoticed by others in the industry for too long a time period.
Strong points for the Yankee Rowe LERs are the the discussions of the root cause and corrective actions for the events, and the requirement to provide the failure mode, mechanism, and effect of each failed component in the event.
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YANKEE R0WE Introduction In order to evaluate the overall quality of the contents of the Licensee Event Reports (LERs) submitted by Yankee Rowe during the February 1, 1985 to September 15, 1986 Systematic Assessment of Licensee Performance (SALP) assessment period, a representative sample of the unit's LERs was evaluated using a refinement of the basic methodology presented in NUREG/CR-4178.I The sample consists of 11 of the 13 LERs that were on file at the time the evaluation wa.s started (see Appendix A for a list of-the LER numbers in the sample).
It was necessary to start the evaluation before the end of the SALP assessment period because the input was due such a short time after the end of the SALP period.
Therefore, not all of the LERs prepared during the SALP assessment period were available for review.
Methodology The evaluation consists of a detailed review of each selected LER to determine how well the content of its text, abstract, and coded fields meet
the requirements of 10 CFR 50.73(b), NUREG-1022, and Supplements 1
and 2 to NUREG-1022.
The evaluation process for each LER is divided into two parts.
The first part of the evaluation consists of documenting comments specific to the content and presentation of each LER.
The second part consists of determining a score (0-10 points) for the text, abstract, and coded fields of each LER.
The LER specific comments serve two purposes:
(1) they point out what the analysts considered to be the specific deficiencies or observations concerning the information pertaining to the event, and (2) they provide a basis for a count of general deficiencies for the overall sample of LERs
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that was reviewed.
Likewise, the scores serve two purposes:
(1) they serve to illustrate in numerical terms how the analysts perceived the content of the information that was presented, and (2) they provide a basis for determining an overall score for each LER. The overall score for each LER is the result of combining the scores for the text, abstract, and coded fields (i.e., 0.6 x text score + 0.3 x abstract score + 0.1 x coded fields score - overall LER score).
The results of the LER quality evaluation are divided into two categories:
(1) detailed information and (2)-summary information. The detailed information, presented in Appendices A through D, consists of LER sample information (Appendix A), a table of the scores for each sample LER.
(Appendix B), tables of the number of deficiencies and observations for the text, abstract and coded fields (Appendix C), and comment sheets containing narrative statements concerning the contents of each LER (Appendix D).
When referring to these appendices, the reader is cautioned not to try to directly correlate the number of comments on a comment sheet with the LER
scores, as the analysts has flexibility to consider the magnitude of a deficiency when assigning scores.
Discussion of Results A discussion of the analysts' conclusions concerning LER quality is
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presented below. These conclusions are based solely on the results of the-
? evaluation of the contents of the LERs selected for review and as such
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represent the analysts' assessment of the unit's performance (on a scale'of 0 to 10) in submitting LERs that meet the requirements of 10 CFR 50.73(b).
Table 1 presents the average scores for the sample of LERs evaluated for Yankee Rowe. The reader is cautioned that the scores resulting from the methodology used for this evaluation are not directly comparable to the scores contained in NUREG/CR-4178 due to refinements in the methodology.
In order to place the scores provided in Table 1 in perspective, the distribution of the overall average score for all licensees that have been evaluated using the current methodology is provided in Figure 1.
Additional scores are added to Figure 1 each month as other licensees are
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TABLE'1. l SUMMARY OF SCORES FOR YANKEE R0WE Averaae Hiah Low Text 8.2 9.4 5.9 Abstract 9.2 10.0 7.7
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Coded Fields-9.0 10.0 8.0 Overall 8.5 9.6 6.8 a.
See Appendix B for a summary of scores for each LER that was evaluated.
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Figure 1. Distribution of overall average LER scores
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TABLE 2.
LER REQUIREMENT PERCENTAGE SCORES FOR YANKEE R0WE
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TEXT Percentage
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Reauirements f50.73(b)1 - Descriptions Scores ( )#
(2)(ii)(A) - - Plant condition prior to event 91 (11)
(2)(ii)(B) - - Inoperable equipment that contributed b
(2)(ii)(C) - - Date(s) and approximate times 98 (11)
(2)(ii)(D) - - Root cause and intermediate cause(s)
94 (11)
(2)(ii)(E) - - Mode, mechanism, and effect 100 ( 5)
(2)(ii)(F) - - EIIS Codes 27 (11)
(2)(11)(G) - - Secondary function affected b
(2)(11)(H) - - Estimate of unavailability 100 (3)
(2)(ii)(I) - - Method of discover-y 91 (11)
(2)(ii)(J)(1) - Operator actions affecting course
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(2)(ii)(J)(2) - Personnel error (procedural deficiency)
83 (6)
(2)(ii)(K) - - Safety system responses 100 ( 2)
(2)(ii)(L) - - Manufacturer and model no. information 40 (5)
(3)
Assessment of safety consequences 71 (11)
Corrective actions 94 (11)
(4)
(5)
Previous similar event information 41 (11)
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(2)(1) - - - - Text presentation 77 (11)
ABSTRACT Percentage a
Reauirements ISO.73(b)(1)1 - Descriptions Scores ( l
- Major occurrences (Immediate cause and effect 100 (11)
information)
- Description of plant, system, component, and/or 100 ( 2)
personnel responses
- Root cause information 95 (11)
- Corrective Action information 88 (11)
- Abst'ract presentation 83 (11)
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TABLE 2.
(continued)
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CODED FIELDS Percentage
Item Number (s) - Descriptic L,
Scores ( 1 1, 2, and 3 - Facility name (unit no.), docket no. and 100 (11)
page number (s)
4 - - - - - - Title 68 (11)
5, 6, and 7 - Event date, LER No., and report date 100 (11)
8 - - - - - - Other facilities involved 100 (11)
9 and 10 - - Operating mode and power level 97 (11)
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11 - - - - - Reporting requirements 98 (11)
12 - - - - - Licensee contact information 100 (11)
13 - - - - - Coded component failure information 98 (11)
14 and 15 - - Supplemental report information 91 (11)
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Percentage scores are the result of dividing the total points for a requirement by the number of points possible for that requirement.
(Note: Some requirements are not applicable to all LERs; therefore, the number of points possible was adjusted accordingly.) The number in parenthesis is the number of LERs for which the requirement was considered applicable, b.
A percentage score for this requirement is meaningless as it is not possible to determine from the information available to the analyst whether this requirement is applicable to a specific LER.
It is always given 100%
if it is provided and is always considered "not applicable" when it is not.
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evaluated. Table 2 and Appendix Table 8-1 provide a summary of the
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information that is the. basis for the average scores in Table 1.
For example, Yankee Rowe's average score for the text of the LERs that were evaluated is 8.2 out of'a possible 10 points.
From Table 2 it can be seen that the text score actually results from the review and evaluation of 17 different requirements ranging from the discussion of plant operating conditions before the event [10 CFR 50.73(b)(2)(ii)(A)] to text presentation. The percentage scores in the text summary section of Table 2 provide an indication of how well each text requirement was addressed by the licensee for the 11 LERs that were evaluated.
Discussion of Specific Deficiencies A review of the percentage scores presented in Table 2 will quickly point out where the licensee is experiencing the most difficulty in preparing LERs.
For example, requirement percentage scores of less than 75 indicate that the unit probably needs additional guidance concerning these requirements. Scores of 75 or above, but less than 100, indicate that the unit probably understands the basic requirement but has either:
(1) excluded certain less significant information from most of the discussion concerning that requirement or (2) totally failed to address the requirement in one or two of the selected LERs. The unit should review the LER specific comments presented in Appendix D in order to determine why it received less than a perfect score for certain requirements. The text requirements with a score of less than 75 or those with numerous deficiencies are discussed below in their order of importance.
In addition, the primary deficiencies in the abstract and coded fields are discussed.
One of the more important requirements, root cause, was discussed very well. A review of the comments in Appendix D will aid in identifying any-minor problems concerning this requirement.
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'The safety assessments for five of the LERs were found'to be deficient or not included, Requirement 50.73(b)(3). A detailed safety assessment is-required in all LERs and should include information such as:-
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An assessment of the consequences and_ implications of the event
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including specifics as to why'it was concluded that there were
"no safety consequences", if applicable.
It is inadequate to
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without explaining how that conclusion was reached.
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A safety assessment should discuss whether the event could have-occurred under a different set of conditions where the safety
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implications would have been more-severe.
If the conditions during the event are considered the worst probable, the LER should state so,
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Finally, a safety assessment should~name other systems (if any)
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that were available'to. perform the function of the safety systems
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that were unavailable during the event.
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i was not provided in the text of three of the five LERs that involved a j
component failure, Requirement 50.73(b)(2)(ii)(L). Components that fail l
should be identified in the text so that others in the industry can be made
' aware of potential problems.
In addition,-although not specifically
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components whose design contributed to an event. An event at one station i
can often lead to the identification of a generic problem that can be corrected at other plants or stations before they experience a similar
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problem.
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Six of the eleven LERs reviewed failed to mention previous similar
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events or state that there were none, Requirement 50.73(b)(5). Previous similar events should be referenced appropriately (LER number'if possible),
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Eight of the eleven LERs reviewed failed to include the Energy Industry Identification System (EIIS) codes.
Requirement 50.73(b)(2)(ii)(F) requires inclusion of the appropriate EIIS code for each system and component referred to in the text.
The text presentations received an overall score of 77%. This score can be improved upon by the use of a consistent text outline (see NUREG-1022, Supplement No. 2, Appendices C and D).
For example, every text should include outline headings such as:
Event Description, Reportability,.
Cause, Safety Assessment, Corrective Actions, and Similar Occurrences.
If applicable, other headings such as: Background, Time Sequences, Plant and/or System Responses, System Descriptions or Generic Implications can be added. Once a basic outline is adopted by all those responsible for writing LERs, the overall quality of the reports will improve, based simply on the fact that every LER will contain at least some information concerning each requirement applicable to the event.
While there are no specific requirements for an abstract, other than those given in 10 CFR 50.73(b)(1), an abstract should, as a minimum, summarize the following information from the text:
1.
Cause/Effect What happened that made the event reportable.
2.
Responses Major plant, system, and personnel responses as a result of the event.
3.
Root / Intermediate The underlying cause of the Causes event. What caused the component and/or system failure l
or the personnel error.
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Corrective Actions What was done immediately to
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Yankee Rowe had good discussions of item numbers 1, 2, and 3 above.
Item number 4 could use some improvement however.
Scores for this item should improve if the corrective action infornation contained in the text is summarized in the abstract.
The main deficiency in the area of coded fields involves the title.
Item (4). Nine of the titles failed to indicate the root cause and five failed to include the link between the cause and the result (i.e., why the event was required to be reported). While result is considered the most important part of the title, cause information (and link, if necessary)
must be included to make a title complete.
An example of a title that only addresses the result might be " Reactor Scram". This is inadequate in that the cause and link are not provided. A more appropriate title might be
" Inadvertent Relay Actuation During Surveillance Test LOP-1 Causes Reactor Scram".
From this title, the reader knows the cause was either personnel or procedural and surveillance testing was the link between the cause and the result.
Table 3 provides a summary of the major areas that need improvement for the Yankee Rowe LERs.
For more specific information concerning additional deficiencies, the reader should refer to the information presented in Appendices C and D.
General guidance concerning requirements can be found in NUREG-1022, Supplement No. 1 and 2.
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TABLE 3.
AREAS MOST NEEDING IMPROVEMENT FOR YANKEE R0WE LERs
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Areas Comments Safety assessnent infornation All LERs should include a detailed safety assessment. The text should discuss whether or not the event could have been worse had it occurred under different, yet probable circumstances and provide
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information concerning backup systems that were available to mitigate the consequences of the event.
Manufacturer and model number Component identification information information should be included in the text whenever a component fails.or (although not specifically required by the current regulation) is suspected of contributing to the event because of its design.
Previous similar events Previous similar events should be referenced (e.g., by LER Number)
or, as stated in NUREG-1022, Supplement No. 2, if none are identified, the text should so state.
EIIS codes Codes'for each component and system referred to in the text should be provided.
Text presentation An outline format is recommended for the text of all LERs.
Abstract Corrective action information is not being adequately summarized in the abstracts.
Each abstract should contain a good summary of the corrective action information that is discussed in the text.
Coded fields a.
Titles Titles need to be written such that they better describe the event by including the root cause, result, and the link between them.
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REFERENCES
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1.
B. S. Anderson, C. F. Miller, B. M. Valentine, An Evaluation of Selected Licensee Event Reports Prepared Pursuant to 10 CFR 50.73 (DRAFT), NUREG/CR-4178, March 1985.
2.
Office for Analysis and Evaluation of Operational Data, Licensee Event Report System, NUREG-1022, U.S. Nuclear Regulatory Commission, September 1983.
3.
Office for Analysis and Evaluation of Operational Data, Licensee Event Report System, NUREG-1022 Supplement No. 1. U.S. Nuclear Regulatory Commission February 1984.
4.
Office for Analysis and Evaluation of Operational Data, Licensee Event Report System, NUREG-1022 Supplement No. 2. U.S. Nuclear Regulatory Commission, September 1985.
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APPENDIX A LER SAMPLE SELECTION INFORMATION FOR YANKEE R0WE
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TABLE'A'-1.
LER SAMPLE SELECTION FOR' YANKEE R0WE
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Sample Number LER Number Comments
85-001-00
85-003-00
85-004-00
85-006-00
85-007-00
85-008-00
85-009-00-Scram
85-010-00 Scram
86-001-00
86-002-00
86-003-00
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APPENDIX 8 EVALUATION SCORES OF INDIVIDUAL LERS FOR YANKEE R0WE
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TABLE B-1.
EVAlfATION SCORES OF INDIVIDUAL LERa FOR YAIIKEE ROWE
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a LER Sample Number
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2
4
6
8
10
12
14
16 i
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l Text 9.0 7.6 9.2 8.4 7.7 5.9 8.2 8.6 9.4 7.3 8.5
--
--
--
-
--
l l
Abstrac t 9.0 7.7 10.0 9.5 8.4 8.1 9.5 9.8 10.0 8.9 9.9
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-
-
--
--
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Coded
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Fields 9.5 9.0 8.5 8.8 9.3 8.0 8.7 10.0 9.0 9.3 9.2
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--
--
--
-
i Overall 9.1 7.8 9.4 8.8 8.1 6.8 8.6 9.1 9.6 8. 0 -
9.0
--
-
-
--
--
4 LER Sample Number *
18
20
22
24
26
28
30 AVERACE
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Text
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-
-
-
--
--
--
--
--
--
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--
8.2
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Abstract
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--
--
--
--
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-
-
' -
--
--
--
9.2 to e
Coded
-"
Fields
-
--
--
--
-
--
--
--
-
-
-
-
--
--
9.0-t l
overall
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-
--
-
-
--
-
--
--
--
--
--
-
--
8.5 l
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c.
See Appendix A for a list of the corresponding LER numbers..
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APPENDIX C
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DEFICIENCY AND OBSERVATION COUNTS FOR YANKEE R0WE
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TABLE C-1.
TEXT DEFICIENCIES AND OBSERVATIONS FOR YANKEE R0WE
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Number of LERs with Deficiencies and
__
Observations Sub-paragraph Paragraph Description of Deficiencies and Observations __, Totals *
Totals (
)
50.73(b)(2)(11)(Al--Plant operating 2 (11)
conditions before the event were not included or were inadequate.
50.73(b)(2)(ii)(8)--Discussion of the status
-- ( 0)
of the structures, components, or systems that were inoperable at the start of the event and that contributed to the event was not included or was inadequate.
50.73(b)(2)(ii)(C)--Failure to include 1 (11)
sufficient date and/or time information, a.
Date information was insufficient.
O b.
Time information was insufficient.
50.73(b)(2)(11)(D)--The root cause and/or 1 (11)
intermediate failure, system failure, or
'
personnel error was not included or was inadequate, a.
Cause of component failure was not
included or was inadequate b.
Cause of system failure was not
included or was inadequate c.
Cause of personnel error was not
included or was inadequate.
50.73(b)(2)(ii)(E)--The failure mode.
0 ( 5)
mechanism (immediate cause), and/or effect (consequence) for each failed component was not included or was inadequate.
a.
Failure mode was not included or was inadequate b.
Mechanism (immediate cause) was not included or was inadequate c.
Effect (consequence) was not included or was inadequate.
C-1
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TABLE C-1.
(continued)
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Number of LERs with Deficiencies ~and Observations-Sub-paragraph Paragraph Description of Deficiencies and Observations Totals'
Totals-(
-)
50.73(b)(2)(ii)(F)--The Energy Industry 8 (11)
Identification System component function identifier _for each component or system was not included.
50.73(b)(2)(ii)(G)--For a failure of a 0 (0)
component with multiple functions, a list of systems or secondary functions which were also affected was not included or was
,
inadequate.
'
50.73(b)(2)(11)(H)--For a failure that 0 (3)
rendered a train of a safety system inoperable, the estimate of elapsed time from the discovery of the failure until the train was returned to service was not included.
50.73(b)(2)(ii)(I)--The method of discovery 1 (11)
of each component failure, system failure,-
personnel error, or procedural error was not included or was inadequate.
a.
Method of discovery for each
component failure was not included or was inadequate
,
b.
Method of discovery for each system
failure was not included or was inadequate c.
Method of discovery for each
personnel error was not included or was inadequate d.
Method of discovery for each
procedural error was not included or was inadequate.
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t C-2 i
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TABLE C-1.
(continued)
Number of LERs with
'
Deficiencies and Observations
'
Sub-paragraph Paragraph l
Description of Deficiencies and Observations Totals #
Totals (
)
50.73(b)(2)(ii)(J)(1)--Operator actions that
-- ( 0)
affected the course of the event including operator errors and/or procedural deficiencies were not included or were inadequate.
50.73(b)(2)(ii)(J)(2)--The discussion of 2 (6)
each personnel error was not included or was inadequate.
a.
OBSERVATION: A personnel error was
implied by the text, but was not explicitly stated, b.
50.73(b)(2)(ii)(J)(2)(1)--Discussion
as to whether the personnel error was
cognitive or procedural was not included or was inadequate, c.
50.73(b)(2)(ii)(J)(2)(ii)--Discussion
as to whether the personnel error was contrary to an approved procedure, was a direct result of an error in an approved procedure, or was associated with an activity or task that was not covered by an approved procedure was
,
not included or was inadequate.
d.
50.73(b)(2)(ii)(J)(2)(iii)--Discussion
, ~ -
of any unusual characteristics of the work location (e.g., heat, noise) that directly contributed to the personnel error was not included or was inadequate.
e.
50.73(b)(2)(11)(J)(2)(iv)--Discussion
of the type of personnel involved (i.e., contractor personnel, utility licensed operator, utility nonlicensed operator, other utility personnel) was not included or. was inadequate.
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C-3
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' TABLE C-1.
(continued)'
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Number of LERs with Deficiencies and Observations
.Sub-paragraph Paragraph Description of Deficiencies and Observations Totals'
Totals (
)
50.73(b)(2)(ii)(K)--Automatic and/or manual 0 (-2)
,
safety system responses were not included or were inadequate.
50.73(b)(2)(ii)(L)--The manufacturer and/or:
3 gi)
-model number of each failed component was not included or was inadequate.
50.73(b)(31--An assessment of the safety-5 (11)
consequences and implications of the. event was not included.or was inadequate.
a.
OBSERVATION: The availability of
other systems or components capable of mitigating the consequences of the event was not discussed.
If no other systems or components were available, the text should state that none
-
existed.
b.
OBSERVATION: The consequences-
of the event had it occurred under
,
more severe conditions were not discussed.
If the event occurred under what were considered the most
,
severe conditions, the text should so state.
50.73(b)(4)--A discussion of any corrective-1 (11)
actions planned as a result of the event including those to reduce the probability of similar~ events occurring in the future was not included or was inadequate.
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- TABLE C-1.
(continued)
l
,
Number of LERs with Deficiencies and-Observations
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Sub-paragraph Paragraph Description of Deficiencies and Observations Totals *
Totals (
)
'
a.
A discussion of actions required to O
correct the problem (e.g., return the q
component or system to an operational
'
condition or correct the personnel error) was not included or was inadequate.
b.
A discussion of actions required to
,
reduce the probability of recurrence of the problem or similar event (correct the root cause) was not included or was inadequate.
c.
OBSERVATION: A discussion of actions'
required to prevent similar-failures in similar and/or other systems (e.g.,
correct the faulty part in all components with the same manufacturer and model number) was not included or was inadequate.
50.73(b)(5)--Information concerning previous 6 (11)
similar events was not included or was inadequate.
C-5
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TABLE C-1.
(continued)
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Number of LERs with
'
Deficiencies and Observations Sub-paragraph Paragraph Description of Deficiencies and Observations Totals'
Totals (
-)
50.73(b)(2)(1)--Text presentation 2 (11)
inadequacies.
a.
0BSERVATION: A diagram would have
aided in understanding the text discussion.
b.
Text contained undefined acronyms
and/or plant specific designators.
c.
The text contains other specific
deficiencies relating to the readability.
,
a.
The "sub-paragraph total" is a tabulation of specific deficiencies or-observations within certain requirements.
Since an LER can have more than
'.
one deficiency for certain requirements,.(e.g., an LER can be deficient in the area of both date and time information), the sub-paragraph totals do not necessarily add up to.the paragraph total.
,
b.
The " paragraph total" is the number of LERs that have one or more requirement deficiencies or observations. The number in parenthesis is the number of LERs for which the requirement was considered applicable.
,
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C-6 i
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- - - -.. - -, - -.
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TABLE C-2.
ABSTRACT DEFICIENCIES AND OBSERVATIONS FOR YANKEE R0WE
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Number of LERs with Deficiencies and Observations Sub-paragraph Paragraph Description of Deficiencies and Observations Totals'
Totals (
)
A summary of occurrences (immediate cause 0 (11)
and effect) was not included or was inadequate A summary of plant, system, and/or personnel 0 ( 2)
responses was not included or was inadequate, a.
Summary of plant responses was not included or was inadequate.
b.
Summary of system responses was not included or was inadequate, c.
Summary of personnel responses was not included or was inadequate.
A summary of the root cause of the event 1 (11)
was not included or was inadequate.
A summary of the corrective actions taken or 4 (11)
planned as a result of the event was not included or was inadequate.
.
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C-7
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TABLE C-2.
(continued)
.
Number of LERs with Deficiencies and Observations Sub-paragraph Paragraph Description of Deficiencies and Observations Totals #
Totals (
)
Abstract presentation inadequacies 2 (11)
a.
OBSERVATION: The abstract contains
information not included in the text.
The abstract is intended to be a summary of the text, therefore, the text should discuss all information summarized in the abstract.
b.
The abstract was greater than
1400 characters
c.
The abstract contains undefined
acronyms and/or plant specific designators.
!
d.
The abstract contains other specific
l deficiencies (i.e., poor j
summarization, contradictions, etc.)
a.
The "sub-paragraph total" is a tabulation of specific deficiencies or observations within certain requirements.
Since an LER can have more than one deficiency for certain requirements, the sub-paragraph totals do not necessarily add up to the paragraph total.
b.
The " paragraph total" is the number of LERs that have one or more deficiency or observation. The number in parenthesis is the number of LERs i
for which a certain requirement was considered applicable.
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C-8
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TABLE C-3. -CODED FIELDS DEFICIENCIES AND OBSERVATIONS FOR YANKEE R0WE
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Number of LERs with Deficiencies and Observations Sub-paragraph Paragraph'
,
l Description'of Deficiencies and Observations Totals *
Totals (
)
Facility Name 0 (11)
+
[
a.
Unit number was not included or
-incorrect.
i b.
Name was not included or was incorrect.
I c.
Additional unit numbers were included but not required.
,
Docket Number was not included or was 0 (11)
,
l incorrect.
!
Page Number was not included or was 0 (11)
-
incorrect.
j
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Title was left blank or was inadequate 9 (11)
]
a.
Root cause was not-given in title 9-b.
Result (effect) was not given in title
>
i c.
Link was not given in title
Event Date 0 (11)
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a.
Date not included or was incorrect.
b.
Discovery date given instead of event-l date, i
j LER Number was not included or was incorrect 0 (11)
'
l Report Date 0 (11)
.
I a.
Date not included t
b.
OBSERVATION: Report date was not within thirty days of event date (or discovery date if appropriate).
l Other facilities information in field is 0 (11)
.
inconsistent with text and/or abstract.
.
,
Operating Mode was not included or was 1 (11)
inconsistent with text or abstract.
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C-9
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- _ _. _ _.._,_ -- _ _.-- _..._.. _.,._. _
_ _ _ -. _. - _, _ - - _ _, - _ _, - _ -. _, - -.
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TABLE C-3.
(continued)
,,
humber of LERs with Deficiencies and Observations Sub-paragraph Paragraph Description of Deficiencies and Observations Totals'
Iotals (
)
Power level was not' included or was 0 (11)
. inconsistent with text or abstract Reporting Requirements 1 (11)
a.
The reason for checking the "0THER"
requirement was not specified in the abstract and/or text.
b.
OBSERVATION:
It may have been more
appropriate to report the event under a different paragraph.
c.
OBSERVATION:
It may have been
appropriate to report this event under an additional unchecked paragraph.
Licensee Contact 0 (11)
a.
Field left blank b.
Position title was not included c.
Name was not included d.
Phone number was not included.
Coded Component failure Information 1 (11)
a.
One or more component failure O
sub-fields were left blank, b.
Cause, system, and/or component code
is inconsistent with text.
c.
Compnnent failure field contains data
when no component failure occurred.
d.
Component failure occurred but entire
field left blank.
C-10
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TABLE C-3.
(continued)
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Number of LERs with Deficiencies and Observations Sub-paragraph Paragraph
Description of Deficiencies and Observations Totals'
Totals (
)
Supplemental Report 1 (11)
a.
Neither "Yes"/"No" block of the
supplemental report field was checked.
b.
The block checked was inconsistent I
with the text.
Expected submission date information is 0 (11)
inconsistent with the block checked in Item (14).
a.
The "sub-paragraph total" is a tabulation of specific deficiencies or observations within certain requirements.
Since an IER can have more than
'
i one deficiency for certain requirements, the sub-paragraph totals do not necessarily add up to the paragraph total.
-
b.
The " paragraph total" is the number of LERs that have one or more requirement deficiencies or observations. The number in parenthesis is the number of LERs for which a certain requirement was considered applicable.
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APPENDIX D LER CONNENT SHEETS FOR YANKEE R0WE
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TA8LE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
Section Comments-j-
1.
LER Number: 85-001-00
Scores: Text = 9.0 Abstract = 9.0 Coded Fields = 9.5 Overall = 9.1
Text 1.
50.73(b)(2)(11)(F)--The Energy Industry
'
'
Identification System component function
identifier (s) and/or system name of each component or
!
system referred to in the LER is not included.
,
2.
Acronym (s) and/or plant specific designator (s) are undefined. Acronyms such as EDC.'CHGR, and YAEC
>
should be defined.
,
r
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Abstract 1.
50.73(b)(1)--Summary of corrective actions taken or
'
planned as a result of the event is inadequate.
It
is not clear from the abstract that. future cores will be analyzed with the more restrictive axial power
,
profiles.
!
2.
Abstract contains acronym (s) and/or plant specific
!
designator (s) which are undefined. YAEC should be defined.
-
'
3.
OBSERVATION: The abstract contains information not included in the' text.
The abstract is intended to be I
a summary of the text; therefore, the text should discuss all information summarized in-the abstract,
j The last~ sentence of the abstract contains
!
information not discussed in the text.
)
Coded Fields 1.
Item (4)--The use of acronyms in a title should be j
avoided, unless space is a consideration.
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.-.-.,-..-.,,,--.,---..,-,,,.m.,.
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,,, _,,,,.. -,...,....,...,,. -.,,..,. _ - _,,
,,m-
,
-.---,._,nr 7. -.,,._,, -,,,,,, -., -
,,....,,
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TABLE 0-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
,
Section Comments 2.
LER Number:
85-003-00 Scores: Text - 7.6 Abstract - 7.7 Caded Fields - 9.0 Overall - 7.8 Text 1.
50.73(b)(2)(11)(L)--Identification (e.g. manufacturer and model no.) of the failed component (s) discussed in the text is not included.
2.
50.73(b)(31--Discussion of the assessment of the safety consequences and implications of the event is inadequate.
OBSERVATION-The consequences of the event had it occurred under more severe conditions should be discussed. If the event occurred under what are considered the most severe conditions, the text should so state.
3, 50.73(b)(5)--Information concerning previous similar events is not included.
If no previous similar events are known, the text should so state.
Abstract 1.
50.73(b)(1)--Summary of corrective actions taken or planned as a result of the event is inadequate.
The modifications to reduce the possibility of damage due to flow as discussed in the text should be summarized in the abstract.
2.
OBSERVATION: The abstract contains information not included in the text. The abstract is intended to be a summary of the text; therefore, the text should discuss all information summarized in the abstract.
Coded Fields 1.
Item (4)--Title: Root cause is not included.
D-2
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
.
Section'
Comments 3.
LER Number:
85-004-00 Scores: Text - 9.2 Abstract - 10.0 Coded Fields - 8.5 Overall - 9.4 Text 1.
50.73(b)(2)(11)(F1--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
2.
50.73(b)(31--Discussion of the assessment of the safety consequences and implications of the event is very good.
.
3.
50.73(b)(2)(ii)(0)--08SERVATION: The score for this requirement is based on the assumption that the supplemental report will contain all the necessary information.
4.
50.73(b)(51--Information concerning previous similar events is not included.
If no previous similar
events are known, the text should so state.
Abstract 1.
No comments
-
Coded Fields 1.
Item (41--Title:
Root cause and link are not included. A better title might be, "Offsite Testing of Pressurizer Code Safety Valve Identifies Setpoint Greater Than Technical Specification Tolerance -
Setpoint Drift".
2.
Item (91--Operating Mode is not included.
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
-
Section Comments 4.
LER Number: 85-006-00 Scores: Text - 8.4 Abstract - 9.5 Coded Fields - 8.8 Overall - 8.8 Text 1.
50.73(b)(2)(ii)(D)--A discussion as to why the test plug threads were stripped was not included.
2.
50.73(b)(2)(ii)(F)--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
3.
50.7'3(b)(2)('ii)(L)--Identification (e.g. manufacturer and model no.) of the failed component (s) discussed in the text is not included.
Abstract 1.
No comment Coded Fields 1.
Item (4)--Title:
Root cause (failed test plug) and link (surveillance testing) are not included. A more appropriate title might be " Surveillance Testing Reveals an Inoperable Condesate Pump Trip Circuit Due to a Failed Test Plug".
2.
Item (ll)--0BSERVATION:
It appears it would have been appropriate to also report this event under paragraph (s) 50.73(a)(2)(v).
D-4
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
,
Section Comments 5.
LER Number: 85-007-00 Scores: Text - 7.7 Abstract - 8.4 Coded Fields = 9.'s Overall = 8.1 Text 1.
50.73(b)(2)(ii)(F)--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
2.
50.73(b)(2)(ii)(J)(2)--It appears that personnel error is involved in this event, but it is not discussed. The text implies that the solder joints were made up using an inadequate technique, but does not say why.
3.
50.73(b)(31--Discussion of the assessment of the safety consequences and implications of the event is inadequate. Since the automatic scram feature of the RPS may not have been available at $15 MWE, the consequences of this should be discussed.
4.
50.73(b)(4)--01scussion of corrective actions taken or planned is inadequate.
A discussion of actions required to reduce the
,
probability of recurrence (i.e, correction of the root cause) is not included or is inadequate. See text comment 2.
5.
50.73(b)(5)--Information concerning previous similar events is not included.
If no previous similar events are known, the text should so state.
Abstract 1.
50.73(b)(1)--Summary of corrective actions taken or planned as a result of the event is inadequate.
See text comments 2 and 4.
Coded Fields 1.
Item (4)--Title:
Root cause is not included.
0-5
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
,
Section Comments 6.
LER Number:
85-008-00 Scores: Text - 5.9 Abstract - 8.1 Coded Fields - 8.0 Overall - 6.8 Text 1.
Submittal of an LER without a text is acceptable; however, the abstract must then meet all the requirements of a text and still be less than 1400 spaces.
The following comments apply to the abstract that was evaluated as if it were a text.
,
2.
50.73(b)(2)(11)(A)--Discussion of plant operating conditions before the event is inadequate. Mode 2 should be defined in the text (e.g. Startup).
3.
50.73(b)(2)(ii)(C)--Time information for nujor occurrences is inadequate. At what time was the blowdown monitor declared inoperable?
4.
50.73(b)(2)(ii)(01--A supplemental report appears to be needed to describe the cause of the check valve failure to reseat.
Without a commitment to submit a supplemental report, this LER must be considered incomplete.
5.
50.73(b)(2)(ii)(F)--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
6.
50.73(b)(2)(ii)(L)--Identification (e.g. manufacturer and model no.) of the failed component (s) discussed in the text is not included.
7.
50.73(b)(4)--A supplemental report appears to be needed to describe the corrective actions necessary to prevent recurrence. Without a commitment to submit a supplemental report, this LER must be considered incomplete.
8.
50.73(b)(5)--Information concerning previous similar events is not included.
If no previous similar events are known, the text should so state.
9.
Some ideas are not presented clearly (hard to follow). What were the consequences of not complying with T.S. 3.3.3.1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />?
D-6
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c
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
,
Section Comments 6.
LER Number:
(continued)
10.
OBSERVATION: A diagram or figure would aid in understanding the event.
Abstract 1.
The abstract is deficient for the same general reasons discussed in the text (i.e., cause and corrective action details).
Coded Fields 1.
Item (4)--Title: Root cause and link information are not included. A better title might be, "feedwater in Blowdown Line Via Check Valve Back Leakage Causes 810wdown Monitor To Be Inoperable".
2.
Item (14)--The block checked is inconsistent with information in the text. A supplement is needed to provide cause and corrective action information.
.
0-7
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-
-
-
-
-
-
-
- -
- -
-
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE.R0WE (029)
Section Comments 7.
LER Number: 85-009-00 Scores: Text = 8.2 Abstract = 9.5 Coded Fields = 8.7 Overall = 8.6 Text 1.
Submittal of an LER without a text is acceptable; however, the abstract must then meet all the requirements of a text and still be less than 1400 spaces. The following comments apply to the abstract that was evaluated as if it were a text.
2.
50.73(b)(2)(11)(A)--A brief description of the
'
operating numbers should be included.
3.
50.73(b)(2)(ii)(F)--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
4.
50.73(b)(3)--Distussion of the assessment of the safety consequences and implications of the event is inadequate.
It is not enough to.say there were no adverse effects. The safety assessment should give specific reasons as to why there were no adverse effects.
Abstract 1.
No coment Coded Fields 1.
Item (4)--Title:
Root cause (personnel error) and link (maintenance operation) are not included.
2.
Item (13)--Component failure field contains data when
-
no component failure occurred.
.
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TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
.,
Section Comments 8.
LER Number: 85-010-00 Scores: Text - 8.6 Abstract - 9.8 Coded Fields = 10.0 Overall - 9.1 Text 1.
50.73(b)(2)(ii)(F)--The Energy Industry Identification System component function identifier (s) and/or system name of each component or system referred to in the LER is not included.
2.
50.73(b)(3)--Discussion of the assessment of the safety consequences and implications of the event is inadequate.
OBSERVATION: The consequences of the event had it occurred under more severe conditions should be discussed.
If the event occurred under what are considered the most severe conditions, the text should so state.
Abstract 1.
No comment
-
Coded Fields 1.
No comment
.
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,
TA8'L E' D-1.
SPECIFIC LER COMMENTS FOR YANKEE ROWE (029)
,
Section Comments 9.- LER Number: 86-001-00 Scores: Text - 9.4 Abstract - 10.0 Coded Fields - 9.0 Overall - 9.6 Text 1.
Submittal of an LER without a text is acceptable; however, the abstract must then meet all the requirements of a text and still be less than 1400 spaces.
The following comments apply to the abstract that was evaluated as if.it were a text.
'2.
50.73(b)(5)--Information concerning previous similar events is not included.
If no previous similar events are known, the text should so state.
Abstract 1.
No comment Coded Fields 1.
Item (4)--Title:
Root cause and link are not included. A better title might be, " Administrative Control Problems Results In A Change To The Yankee Emergency Plan Without PORC Review - Technical Specifications Violation".
.
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2
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'I l'BLE'D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE (029)
,
Section Comments 10. LER Number: 86-002-00 Scores: Text = 7.3 Abstract = 8.9 Coded Fields = 9.3 Overall = 8.0 Text 1.
Submittal of an LER without a text is acceptable; however, the abstract must then meet all the requirements of a text and still be less than 1400 spaces.
The following comments apply to the abstract that was evaluated as if it were a text.
2.
50.73(b)(2)(ii)(J)(2)--Discussion of the personnel error is inadequate. A discussion as to how the personnel er.ror occurred and the type of personnel involved should be included.
3.
50.73(b)(3)--Discussion of the assessment of the safety consequeates and implications of the event is not included.
Abstract 1.
No comment Coded Fields 1.
Item (4)--Title:
Root cause (personnel error) is not included.
,
.
D-11
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i TABLE D-1.
SPECIFIC LER COMMENTS FOR YANKEE R0WE-(029)
Section Comments
.
11. LER Number:
86-003-00-Scores: Text. 8.5 Abstract. 9.9 Coded Fields - 9.2
. 0vera11 - 9.0 Text 1.
Submittal of an LER without a text is acceptable; however, the abstract must then meet all the requirements of a text and still be less than 1400 spaces.
The following comments apply to the abstract
-
that was evaluated as if it were a text.
2.
50.73(b)(2)(11)(F)--The Energy Industry Identification System component function
,
identifier (s) and/or system name of each component or system referred to in the LER is not included.
3.
50.73(b)(2)(iil(I)--Discussion of the method of discovery of the Technical Specification action statement not being met is not included.
4.
50.73(b)(51--Information concerning previous similar events is not included.
If no previous similar events are known, the text should so state.
Abstract 1.
No comment Coded Fields 1.
Item (4)--Title:
Root cause is not included.
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