ML20117C666
| ML20117C666 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Hatch |
| Issue date: | 07/12/1984 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20114F930 | List:
|
| References | |
| FOIA-84-616 NUDOCS 8505090485 | |
| Download: ML20117C666 (3) | |
Text
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SAFETY EVALUATION FOR OPERATION WITH MISSING PIPE RESTRAINTS FOR HIGH ENERGY LINES OUTSIDE CONTAINMENT - HATCH UNIT NO. 2 i
1 BACKGR0tMD The Hatch thit 2 licensee has reported that eight pipe whip restraints have not been installed in high energy lines outside of containment.
The design calls for these restraints and they are described in the FSAR but they were never installed.
The systems affected include the reactor core isolation cooling (RCIC) system (one restraint), the reactor wa.ter cleanup (RWCU) system (two restraints), the control rod drive (CRD) system return to feedwater (three restraints) and the auxiliary steam line in the reactor building (two restraints).
DIS CUSSION Auxiliary Steam Line This line comes from the auxiliary boiler and is primarily used for plant heatup prior to startup.
Until the pipe whtp restraints are installed, the licensee has connitted to isolate this line in the reactor building.
Therefore this line will no longer be a high energy, line and we do not need to assume its failure and subsequent pipe whip.
On a related note, the licensee has stated that the auxiliary boiler has been inoperab,le for nearly a year.
RCIC and RWCU Systems The RCIC and MCU pipe restraints were designed to protect the outboard containment isolation valves on these systems.
If these lines were postulated to rupture and disable their respective outboard containment isolation valves, reliance can still be placed on the automatic isolation of the inboard containment isolation valve for each system.
The contain-ment isolation valves on the 4-inch RCIC and the 6-inch RWCU systems automatically isolate to limit blowdown on receipt of a signal indicating a failure in the respective system line.
The isolation signals for the RCIC valves consist of high steam line space temperature, high steam line flow, low steam supply pressure, and high turbine exhaust pressure.
The WCU valves will isolate on high differential flow, high differential l
temperature between the inlet and outlet cleanup. room ventilation, high ambient temperature, or high temperature downsteam of the non-regenerative l
heat exchanger.
l The licensee has explored the possibility of the RCIC or RWCU system line rupturing outside contaimeent, thus disabling the outboard containment isolation valve as a result of the postulated pipe whip along with the simultaneous single active failure of the inboard containment isolation valve.
This would resul.t in a loss of coolant accident outside containment through either the 4-inch RCIC system or the 6-inch RWCU system.
Using 85050904s5 841002 FI
-616 PDR
failure rate probabilities from WASH-1400 and considering pipe diameter e
and length of piping, the licensee has -determined that the. probability of the combined pipe failure along with the single active failure of the inboard containment tsolation valve for the RCIC or RWCU systems to be between 10-9 and 10-7 per year.
Although we have not verified these numbers, we agree that the probabflity of a pipe break in the pipe sec4 tions of interest and a failure of both the inboard and outboard isolation valve is sufficiently small to justify interim operation.
We have asked the licensee to consider the consequences of a RCIC or RWCU pipe whip and its affect on the operability of other safety systems.
The licensee states that pipe whip of the RCIC system could damage the safety related main steam isolation valve leakage control (MSIVLC) system.
Rupture of the MSIVLC would allow steam leakage into the secondary contain-ment, thus adding to the steam flow from the failure of the RCIC steam line.
The licensee has stated that this additional offsite release source would not raise the total offsite release to above 10 CFR Part 100 limits. We agree that the additional failure of the MSIVLC would not be significant and would not be outside the plant's design basis accident analysis.
Rupture of the RWCU would not, according to the licensee, result in disabling any safety-related systems.
Contr61 Rod Drive Return Line The pipe whip restraints on the CRD return line are designed to protect safety related cable trays, CRD insert and withdraw lines.
The licensee has stated that the probability of damage to essential cable grays due to a pipe break in the CRD return line is on the order of 10- per year.
The licensee.has examined the consequences of the CRD line disabling safety related cables.
They state that three conduits are ' capable of being damaged.
These conduits provide power to:
1.
Main steam line condensate drain line; 2.
Main steam isolation valve leakage control system; and 3.
Main steam isolation valve (outboard side only).
The licensee states that if any o'r all of conduits are damaged, the plant could still be brought to a safe cold shutdown assuming the worst single active failure for each conduit. We have discussed these postulated events with the licensee and we concur that the loss of any or all of these conduits plus a single active failure would not prevent the plant from achieving a safe cold shutdown condition.. Based on the licensee's evaluation, they have chosen not to install the pipe whip restraints on the CRD line.
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CONCLUSION 4
We have examined the licensee's analysis of the consequences of an unrestrained pipe whip of the affected RCIC, MdCU and CRD lines outside containment. ~We conclude that the licensee has examined the appropriate areas of concern.
The probability of a pipe rupture concurrent with a single active failure of the inboard isolation valve for the RCIC and 24CU are sufficiently small to allow interim operation.
In addition, the probability and consequences of a CRD pipe rupture and resultant cable tray damage is also sufficiently small.
Based on our review of the licensee's submittal and discussions with their staff, ne conclude that a reasonable basis exists to permit plant startup.
Our review is based, in part, on the understanding that the licensee plans to install the pipe whip restraints (with the exception of the CRD system) on or before December 31, 1983.
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j UNITED STATES 3
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33 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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- g April 4,1984 Docket No. 50-289 MEMORANDUM FOR:
E. Conner, Chief, Reactor Projects Section No. 38, PB No. 3, Region I FROM:
James Van Vliet, Project Manager Operating Reactors Branch #4, DL
SUBJECT:
NRR SALP' INPUT FOR TMI-l Enclosed is NRR's SALP input for TMI-l for the period 10/1/82 through 1/31/84. This input has been prepared in accordance with NRC Manual Chapter 0516 criteria.
G James Van Vliet, Project Manager Operating Reactors Branch #4, DL
Enclosure:
As Stated cc:
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\\...+J Facility Name: Three Mile Island, Unit No. 1 Licensee: GPU Nuclear Corporation NRR Project Manager: James A. Van Vliet I.
Introduction This report presents the results of an evaluation of the licensee, GPU Nuclear Corporation in the functional area of licensing activities.
It is intended to provide NRR's input to the SALP review process as described in NRC Manual Chapter 0516. The review covers the period 10/1/82 to 1/31/84.
The basic approach used for this evaluation was to first select a number of licensing issues which involved a significant amount of staff manpower. Comments were then solicited from the staff.
In most cases the staff applied the evaluation criteria for the perform-ance attributes based on their experience with the licensee or its products. Finally, this information was assembled in a matrix which allowed an overall evaluation of the licensee's perfonnance. This evaluation is based on staff input from branches in three NRR divisions.
II.
Sunnary of Results NRC Manual Chapter 0516 specifies that each functional area evaluated will be assigned a performance category based on a composite of a number of attributes. The single final rating is then tempered with judgement as to the significance of the individual elements.
Based on this approach, the performance of GPU Nuclear Corporation in the functional area - Licensing Activities - is rated category 2.
III. Criteria Evaluation criteria, as given in NRC Manual Chapter Appendix 0516 Table 1, were used for this evaluation.
IV.
Perfonnance Analysis The licensee's performance evaluation is based on a consideration of seven attributes as given in the NRC Manual Chapter.
For most of the licensing actions considered in this evaluation, only three or four of the attributes were of significance. Therefore, the composite rating is heavily based on the following attributes:
- Management involvement
- Approach to resolution of technical issues
- Responsiveness
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Of the remaining attributes of:
- Enforcement History
- Reportable Events
- Staffing
- Training only staffing was judged to apply to the licensing activities evaluated.
The evaluation was based on our evaluation of the following licensing activities:
- Response to.NUREG-0737 It' ems
- F. ire Protection Program (Appendix R Requirements)
- Steam Generator Recovery Program
- Pre-Restart License Amendments
- Seismic Qualification of Auxiliary Feedwater
- Licensed Operator Requalification Program Changes
- Inadequate Core Cooling Instrumentation
- Plans for Preventing Exceeding PTS Screening Criterion
- Long Term Review of Containment Purge & Vent
- Effluent Discharge Monitor Relocation
- Raising HPI & LPI Bypass Setpoints Station Distribution Voltage Verification Test
- Post-Accident Shielding Alternate
- Environmental Qualification A.
Management Involvement in Assurino Quality
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Overall rating for th'is attribute is category 2.
All rated a'ctivities were considered category 2, except for the steam generator recovery program and the effluent discharge monitor relocation which were rated category 1 and the environmental qualification program which was ra.ted category 3.
In general, the level of management involve-ment has been appropriate for the significance of the issue.
Prior planning, prioritization of activities and corporate management involve-ment in site activities are evident.
In the case of the steam generator recovery program, an issue of high company priority, safety significance, and public visibility, involvement by the highest levels of GPU management has been readily apparent. The effluent discharge monitor relocation licensing activities seemed to have been well founded and properly presented, thus implying close management involvement.
There is, however, little indication of management involvement in the TMI-1 environmental qualification issue. This conclusion was reached based on review of a number of environmental qualification
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submittals, and one meeting on this subject with GPU Nuclear personnel.
Subsequent to the evaluation period, another meeting ahd a two-day audit of the environmental qualification files were co'ncucted: and the results confirm our conclusion in this report.
There is little evidence of programmatic planning for the TMI-1 environmental quali-fication program. The Corporate Policy on environmental qualification became effective on January 20, 1984 and it is not clear what the previous policy may have been. There is no indication of any management or quality assurance review of the environmental qualification files.
Although the files generally seem to contain the information needed to demonstrate qualification, there is no GPU analysis, other than miscellaneous hand-written notes, describing how the information relates to TMI-1 and why it demonstrates qualification. There is no indication that environmental qualification decision making is being done at the appropriate management level.
More management attention is needed.
B.
Approach to Resolution of Technical Issues from a Safety Standpoint Overall ' rating for this attribute is cateoory 2.
Six issues were rated category 1 and eight issues' were rated category 2.
There were no category 3 ratings.
The licensee's understanding of the issues has been. generally apparent and the proposed resolutions have been generally conservative and sound.
In particular the licensee's approach to resolution cf fire protection (Appendix R requirements) demonstrates a clear understanding of the technical 1ssues; leading to technically sound, thorough approaches for resolution of the issues. The licensee's steam
' generator recovery program has continued to be thorough, well planned, conservative and technically sound. For both of these issues, the licensee has frequently posed questions and requested clarifications from the staff on technical or licensing aspects of the. issues. This has tended to assure continued clarity of the issues to be resolved and mihimized false starts, rework, etc.
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environmental qualification, the cateaory 2 rating is marginal, but inprovement is anticipated as a result of increased management involvement (see above).
C.
Responsiveness to NRC Initiatives Overall rating for this attribute is category 2, with all activities rated category 2.
A noted trend is that the licensee is most responsive to those issues that licensee considers having higher priority (those issues impacting restart).
Isrues to which licensee assigns lesser priorities periodically require submittal schedule extension.
Although it is not an activity listed in the evaluation matrix, the Control of Heavy Loads is one issue for which significant submittal extensions have been necessary.
Licensee responses to NRC initiatives are generally sound and thorough; and acceptable resolutions are generally proposed.
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Enforcement History Not applicable.
E.
Reporting and Analysis of Reportable Events Not applicable.
F.
Staffing (Including Management)
Staffing was only evaluated for two activities, thus there is insufficient basis for a meaningful overall rating of this attribute.
Staffing was rated category 1 for the steam generator recovery program. Consistent with the scope and priority of the steam generator recovery program, the licensee has dedicated ample staffing (including management) of appropriate cualifications.
Staffing was rated category 3 for environmental qualification.
Two engineers are currently assigned to TMI-1 environmental qualification. This level of staffing is significantly smaller than the levels seen at other utilities.
It therefore appears that additional staffing would be appropriate, (see above).
G.
Training Training was not evaluated for any of the activities evaluated.
Thus there is no basis for evaluation.
V.
Conclusions Based on an NRR evaluation of 14 licensing activities during the period October 1,1982 through January 31, 1984, the overall performance rating for GPU Nuclear licensing activities for Three Mile Island Nuclear Station, Unit 1 is category 2.
The overall rating for each evaluated attribute is category.2. No major deficiencies affecting licensing activities became apparent during the evaluation period.
GPU Nuclear should focus on improving its environmental qualification program.
The licensee generally devotes an adequate level of management involvement to licensing activities; the licensee's approach to the resolution of technical issues is generally sound and conservative; and, the licensee is generally responsive to NRC initiatives.
W James A. Van Vliet, Project Manager Operating Reactors Branch #4 Division of Licensing 4
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1141-1 EVALUAll0ll IIAIRIX TG lcu Licensing Manageinelit (Lgproach to ltesponsiveness Enforcement Iteporta51e sea 7t ing IrilidW Urancli Action involvement itesolution of to.HRC llistory Events luv ulv ed ICCI'nical Issues initiatives VIMil lleijidiise to llullEG-0737 Items 2
2 2
N/A ti/A Ho basis ll/A
'ASli Fire PF6tection Clu program (Appendix 2
1 11 requirements) 2 il/A H/A 110 basis H/A 1
i!I.h steam Generator OHAl Recovery Program i
1 2
11/A it/A y
yfg 11/ 5 Pre-Hestart License Amendments 2
2 2
li/A II/A 110 basis ll/A' lilinit seismic uualific5U on of Auxillary Feed-2 2
j water 2
II/A II/A Ito basis H/A
.Etjli Licensed Operator llequalification 2
2 Program Clianges 2
II/A N/A 110 basis N/A till liiillequite Core o
Cooling lustru-2 I
mentation
.2 II/A N/A llo basis H/A i
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_lHi-1 EVALUAllull 11AlltlX (c' ontinued)
E leu uceEsing Hiiiivement Al'i>roacli to itesiiB6siveness I.Elircemeiit iteji rEi5le SEaffliiDFi1EIEi grancis Action involvement itesolution of to littC llistory Events luvulv ed lectinical issues initiatives 1
ORBl4 Pressurized 2
2 Shock 2
Thermal fl/A N/A No basis N/A
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CSB Vent &
Not 2
2 N/A Purge evaluated
, N/A No basis N/A IlETB Effluent 1
1 i
N/A' N/A No basis N/A 4
Discharge ilonitor Relocation RSB Raising IIPI &
Not 1
1 i
i LPI Bypass evaluated N/A N/A No basis N/A Setpoints PSB Station No t 1
Distribution evaluated fl/A N/A No basis N/A 2
i Vol tage Veri fica tion Test RSB Post Not 2
Accident evaluated ft/A N/A fio basis ' 'N/A 2
Shielding Alternate EQB Environmental 3
2 Quali fication 2
ti/A N/A 3
N/A ~
OVERALL 2
2 1
2 N/A it/A No basis f(/A i
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50-277 50-278 8 -3~
Pniladeichia Electric Company ATTh:
Mr. V. Boyer Senior Vice President - fluclear 2301 Market Street Pniladelphia, Pennsylvania 19101 Gentlemen:
Suojects: 1.
Order leadifying License Effective leoedtetely 2.
Notice of Violetten and proposed Impositten of Civil Penalties 0: April 12, 1984, an Enforcement Conference was held by Dr. Thomas E. Iterley, kegiona' Administrator, Region I with you and mesters of year staff a: the NRC Region I Office to review the circasestances associated with cent violations of IRC requirements idlich occurred at the fleesti Bettaa Atomic Station Units 2 and 3.
Tuo of the violattees were idestified by the K daring an NRC inspection conducted Jansory 5-20,134.
of this inspection 1994.
(Reftlemoet ten Supert Ses.
c:as sent to you on February) other violettees.;
tsort identified by 50-277/84-01; 50-278/84-Of.
members of your staff, an' esadseted en January 13 - February 29 of thi uns ihnserend te ~
you on March 19. 1984.
E spostles hs. 5 277/9643s 50-278/84-03.) At the Enforcement Conference,the essess of these violattens and your corrective actions were discussed.
The violations are described in the enclosetts. The first violation,which is lumsdiately, described in the enclosed Order stedifying License Efflective. hetdeun and a involved a change to a plant operating precedere for plant 1g change to the shutdown segesace described in.the FSAR istthest havies performed an adequate evaluation to ensure that the changes 484
' violate technical specifications er result to an enroviewed safety
- As a result of the
~ f'the reactor from changes, rods were scramed indtylesally.
4 1977 te late 1983, effectively typassing the 'se
' of the Red Marth Mi:;isirer (MSI) and the Red Segmenes Centrol Systes (
),iThesesystems ensure asterence to appresed costrel red seguences and more regelred by the technical specifications'to be aperable at the time.
CERTIFIED MAIL nuuun namrr itEQUESTED
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s Philadelphia Electric Company 2
In addition, as described in the enclosed Notice of Violation and Proposed Imposition of Civil Penalties, three other violations occurred which involved failures to adhere to facility technical specification limiting conditions for operation. The first violation involved two occurrences during startup of both Unit 2 and Unit 3 in which the reactor heatup rates exceeded the limits specified in the technical specifications.
The first instance occurred because reactor operator license trainees working in the control room did not properly use recorded data to obtain heatup rates.
In the second instance, the violation occurred because an operator was withdrawing control rods too quickly.
In both instances, adequate supervision and oversight of startup activities was not provided.
In the second violation, an unplanned reactor pressurization, above atmospheric pressure, occurred with the reactor at 110*F. At that temperature, reactor pressurization is prohibited by the technical specifications. This violation was caused by a failure to provide sufficient detail in a procedure regarding checks of valve positioning, thereby resulting in failure to recognize that valves were not properly positioned.
In the third violation, although a control rod was inoperable, as indicated by a slow response time during a reactor scram on November 17, 1983, this condition was not recognized until the rod again exhibited a slow response time during another scram of the reactor on January 14, 1984. Although the scram response times were reviewed in November 1983 by a junior technical assistant, tech-nical assistant, shift supervisor, and supervisory engineer, the slow response time of the particular control rod was not identified.
These violations demonstrate the need for improvements at Peach Bottom to assure that the plant is operated in accordance with the technical specifications.
To emphasize the need for improvements in the process for reviewing changes to the plant and procedures, I am issuing the enclosed Order Modifying License Effective Innediately to require an appraisal of your review process and certain plant procedures. To emphasize the need for improved procedures, improved adherence to procedures, and improved supervisory perfonnance and oversight of plant activities, I am issuing the enclosed Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $30,000 for the violations described in Section I of the Notice. The violations described in Section I of the Notice involve the failure to adhere to technical specification limiting conditions for operation.
Although if considered individually these violations are of low safety significance, collectively they reflect a significant problem with adherence to technical specifications and, accordingly, have been categorized in the aggregate as a Severity Level III problem.
The base civil penalty amount for a Severity Level III violation or problem is $40,000. The civil penalty has been mitigated to $30,000 because of the unusually prompt and extensive corrective actions taken for violation I.C.
Philadelphia Electric Company 3
Section II of the enclosed Notice of Violation contains three examples of failures to follow procedures. The failures to follow procedures concern maintenance and surveillance activities involving the RWM and RSCS.
These examples further illustrate the licensee's problems regarding the inoperability of systems. This violation is classified as Severity Level IV.
A civil penalty is not proposed for this violation.
You are required to respond to the enclosed Order and Notice and you should follow the instructions specified therein when preparing your response.
In your response, you should address the specific actions taken and planned to ensure adequate safety reviews, attention to detail in routine plant operations and testing, and improved supervisory performance and oversight of plant activ-ities.
Your response to this letter and Notice will be used in detennining whether further enforcement action is warranted.
In accordance with Section 2.790 of the NRC's " Rules and Practice," 10 CFR Part 2, a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
The responses directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget, othemise required by the Paperwork Reduction Act of 1980, PL 96-5;1.
Sincerely,
- A Richard C.
oung Director Office of pection and Enforcement
Enclosures:
1.
Order Modifying License Effective Inunediately 2.
Notice of Violation and Proposed Imposition of Civil Penalties cc w/encis:
R. S. Fleischmann, Station Superintendent Troy B. Conner, Jr., Esquire Eugene J. Bradley, Esquire, Assistant General Counsel Raymond L. Hovis, Esquire II, Assistant Attorney General MichaelJ.Scibinico.(PDR)(LPDR)
Public Document Room Local Public Document Room Nuclear Safety Infonnation Center (NSIC)
NRC Resident Inspector Commonwealth of Pennsylvania
o Philadelphia Electric Company bec w/encis:
Region I Docket Room (with concurrences)
Section Chief, DPRP PDR ACRS SECY CA R. DeYoung, IE J. Taylor, IE J. Axelrad, IE P. Farron, IE T. Murley, RI J. Liebennan, ELD V. Stello, DED/ROGR Enforcement Coordinators RI, RII, RIII, RIV, RV F. Ingram, PA G. Messenger, OIA B. Hayes, 01 H. Denton, NRR J. Crooks, AEOD E. Jordan, IE N. Grace IE IE:ES EA File EDO Rdg File DCS RI:ES RI:DPRP RI:DPRP RI:DPRP RI:DPRP RI:RC Holody/geb Blough Tripp Keimig Starostecki Gutierrez 4/11/84
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UNITED STATES NUCLEAR REGULATORY COPHISSION In the Matter Docket Nos. 50-277; 50-278 License Nos. DPR-44; DPR-56 PHILADELPHIA ELECTRIC COMPANY EA 84-39 Peach Bottom Atomic Power Station.
Units 2 and 3
)
ORDER MODIFYING LICENSE EFFECTIVE IMMEDIATELY I
Philadelphia Electric Company (the " licensee") is the holder of Facility Opera-ting License Nos. DPR-44 and DPR-56, issued October 25, 1973 and July 2, 1974 respectively, which authorize the licensee to operate the Peach Bottom Atomic Power Station, Units 2 and 3 (the " facility") located in Delta, Pennsylvania.
II In November 1983, the NRC became aware of the licensee's prai:tice of individually scraming control rods to effect a normal reactor shutdown.
The practice was further reviewed during an NRC inspection conducted January 5-20, 1984, and a violation of NRC requirements was identified.
The violation involved changes to the facility and facility procedures allowing individual scraming of control rods without an adequate safety review, as required by 10 CFR 50.59, to determine if the changes involved a modification to technical specifications or an unreviewed safety question. Specifically, in 1977, plant operating procedure GP-3 used for normal plant shutdowns was changed, and in 1978, plant operating procedure GP-9 was written such that the safety functions of two systems required to be operable by facility technical specifications during plant shutdowns, namely the Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS), were effectively bypassed during plant
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2 shutdowns. This operating mode was different than described in the Final Safety Analysis Report (FSAR) and inconsistent with technical specification operability requirements, and was implemented without prior NRC approval, without a change to the technical specification, and without a documented safety evaluation to indicate that the change did not involve an unreviewed safety question. This change was reviewed by the licensee's Plant Operations Review Comittee (PORC),
but the implications of the change apparently were not recognized by the PORC.
Further, in 1979, a separate shutdown sequence was programed into the RWM, that differed substantially from the startup sequence, without evaluating the change to determine if it involved an unreviewed safety question with respect to the FSAR, Consequently, from 1977 through 1983, the licensee failed to recognize that the method used in shutting down the reactors was contrary to the plant technical specifications and the FSAR.
The RWM and RSCS function to avoid control rod patterns that could result in unacceptable consequences in the event of a control rod drop accident. The licensee's practice of individually scraming control rods effectively bypassed the RWM and RSCS controls and reduced the margin of safety in the event of a rod drop accident.
III This violation demonstrates the need for an assessment at the Peach Bottom Atomic Power Station to determine (1) whether adequate safety reviews have been and are currently being performed when plant and procedure changes are made; and (2) whether inconsistencies exist in other procedures with regard
3 to the FSAR and technical specification requirements, as a result of procedure changes not receiving adequate safety review.
Since such inconsistencies, if any exist, could reduce the level of safety at the facility, I have determined that the actions set forth below are required for the public health, safety, and interest, and therefore, should be imposed by an innediately effective Order.
IV In view of the foregoing, pursuant to Sections 103,161(i),161(o),and182of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR Part 2 and 10 CFR Part 50, IT IS HEREBY ORDERED EFFECTIVE IMMEDIATELY THAT:
Within 60 days of the effective date of this Order, the licensee shall submit to the Regional Administrator, Region I, for review and approval, a plan for an appraisal of: (1) the licensee's process for performing safety evaluations and reviews of procedures pursuant to 10 CFR 50.59 to determine if the process is currently effective, or if improvements are needed; (2) plant and system operating procedures to verify that existing procedures are consistent with technical specifications, technical specification bases, and those sections of the FSAR concerning systems necessary to mitigate Design Basis Accidents, and do not involve unreviewed safety questions; and (3) the licensee's program for ensuring that employees involved in the review and approval of operating procedures remain cognizant of the licensing bases, i
4 The NRC expects that this appraisal will involve a process of screening numerous facility procedures to identify those warranting a detailed review. The appraisal shall be conducted, coordinated, and reviewed, by individuals who are familiar with the application of the Boiling Water Reactor technical specifications.
In addition, the appraisal shall be perfomed in a manner that shall not detract from safe plant operation.
The appraisal plan shall describe:
(1) the qualifications of the appraisal team members, and a discussion of their degree of independence, regarding areas reviewed; (2) the methods of perfoming the appraisal and documenting the results; (3) the schedule for completion of appropriate milestones; and (4) the methods for resolving appraisal findings in a timely manner.
Upon approval of the appraisal plan by the Regional Administrator, Region I, the appraisal plan shall be implemented.
Scheduled milestone completion dates may not be extended without good cause and the concurrence of the Regional Administrator, Region I.
5 The licensee shall direct the appraisal team to submit to the Regional Admin-istrator, Region I, at the time it is submitted to the licensee management, a copy of any report of the appraisal and recomendations resulting from the appraisal. The licensee shall direct the appraisal team to report imediately, upon identification, to the licensee management and the NRC any inconsistencies which could affect the safe operation of the facilities.
In addition, the licensee shall consider the reconsnendations resulting from the appraisal and provide to the Regional Administrator, Region I, an analysis of each such reconmendation and the action to be taken in response to the recomendation.
The licensee shall also provide a schedule for accomplishing these actions.
The Regional Administrator, Region I, may relax or tenninate in writing any of the preceding requirements for good cause.
V The licensee may request a hearing on this Order. A request for hearing shall be submitted to the Director, Office of Inspection and Enforcement, U.S.
Nuclear Regulatory Comission, Washington, D.C.
20555 within 30 days of the date of this Order.
A copy of the request shall also be sent to the Executive Legal Director at the same address and to the Regional Administrator, i
Region I, 631 Park Avenue, King of Prussia PA 19406. ANY REQUEST FOR A HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.
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6 If a hearing is to be held concerning this Order, the Comission will issue an Order designating the time and place of hearing.
If a hearing is held, the issue to be considered at such hearing shall be whether this Order shall be sustained.
FOR THE NUCLEAR REGULATORY COP 911SS10N f
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- 7 Richard C.
oung, W rector Office of Inspection and Enforcement Dated at Bethesda, Maryland this/7 W ay of June 1984 9
NOTICE OF VIOLATION AND PROPOSED IMPOSITION UF CIVIL PENALTIES Philadelphia Electric Company Docket Nos. 50-277; 50-278 Peach Bottom Atomic Power Station License Nos. DPR-44; DPR-56 Units 2 and 3 EA 84-39 During a routine NRC inspection on January 13 - February 29, 1984, the NRC reviewed the circumstances associated with violations of technical specifica-tion limiting conditions for operation which were identified by the licensee and reported to the NRC. These violations involved two examples of excessive reactor vessel heatup rates, reactor pressurization at a temperature at which pressurization is prohibited, and startup and operation of the reactor with an inoperable control rod in that the rod exhibited a slow scram response time.
The occurrence of excessive heatup rates at each unit, plus an inadvertent reactor pressurization, demonstrate a lack of attention to detail and inade-quate supervisory performance and control of plant activities.
The excessive heatup rates occurred in the one instance because trainees did not properly utilize recorded data and supervision did not recognize this failure.
In the other instance, an operator was withdrawing control rods too quickly.
The unplanned pressurization of the reactor occurred because valves were not properly positioned, and the improper positions were not recognized during valve checks. The failure to recognize a slow control rod scram time in November 1983 followed by startup and operation of the reactor until January 1984 with this rod fully withdrawn, is of serious concern because several in-dividuals reviewed the scram response times after the November shutdown, but the slow response time of the one rod was not recognized. As a result, ade-quate shutdown margin was not assured.
Although the individual safety signif-icance of these events was minimal, collectively, these events involved both facilities, various shifts and some experienced operators, and they demonstrate
- 1) inadequate attention to detail during the perfonnance of plant operations;
- 2) inadequate control and supervision of routine plant operations and tests; 3)inadequateprocedures;and(4)failuretoadheretoprocedures.
Toemphasizetheimportanceofproviding(1)(adequateattentiontodetail during the perfonnance of plant activities, 2) adequate procedures, and (3) adequate supervision of plant activities to ensure procedures are followed and parameters are maintained within Technical Specification limits, the Nuclear Regulatory Connission proposes civil penalties in the cumulative amount of $30,000.
In accordance witi the NRC Enforcement Policy 10 CFR Part 2 Appendix C, and pursuant to Section 234 of the Atomic Energy Act of 1954, as amended ("Act"), 42 U.S.C. 2282, PL 96-295 and 10 CFR 2.205, the particular violations and the associated civil penalties are set forth below, y /k
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'2 Notice of Violation t'
I.
VIOLATIONS ASSESSED A CIVIL PENALTY Technical Specification 3.6. A.1 requires that the average rate of A.
change of reactor coolant temperature not exceed 100'F in any one-hour period during nomal heatup or cooldown.
Contrary to the above, 1.
During the heatup of Unit 3 on January 24, 1984 between 9:15 a.m. and 10:15 a.m. and between 9:30 a.m. and 10:30 a.m.,
the average rate of change (average over an hour) of the i
reactor coolant temperature, as indicated on the B recirculation The actual loop temperature recorder, exceeded 100*F per hour.
l temperature changes over the respective one hour periods were l
102*F and 111'F.
l 2.
During heatup of the Unit 2 reactor, on January 31, 1984, between 4:20 a.m. and 5:20 a.m., the reactor coolant temperature, as 7
indicated by the A and B Recirculation Loop temperature traces.
increased 110'F.
Technical Specification 3.6.A.2. Thermal and Pressurization Limits, l
B.
and Figure 3.6.2, prohibit reactor vessel pressurization above l
atmospheric pressure at vessel temperatures below 120'F.
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Contrary to the above, for approximately five minutes at about 5:30 25, 1984, the Unit 3 reactor vessel was pressurized p.m. on January above atmospheric pressure to about 10 psig(, and at the time, the reactor vessel temperature was below 120*F about 110*F).
j Technical Specification 3.3.C.3 specifies that the maximum scram
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C.
time for 90 percent insertion of any operable control rod shall not exceed 7.0 seconds. Technical Specification 3.3.A.2.C specifies that control rods with scram times greater than those specified in Technical Specification 3.3.C.3 shall be considered inoperable.
Contrary to the above, on November 17, 1983, control rod 34-27 had a scram time of greater than 12 seconds, as indicated on a strip l
chart recorder, but this condition was not recognized at that time and the control rod was not considered inoperable until a subsequent reactor scram on January 14, 1984.
These violations have been categorized in the aggregate as a Severity l
l Level !!! problem (Supplement !).
(CivilPenalty-$30,000distributedequallyamongtheviolations).
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II. VIOLATION NOT ASSESSED A CIVIL PENALTY l
Technical Specification 6.8 and Regulatory Guide 1.33 (November 1972) require implementation of written procedures for troubleshooting, for
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control of maintenance, and for surveillance tests.
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Contrary to the above, written procedures, as required above, were not adequately implemented as evidenced by the following examples:
a.
Administrative Procedure A-26, Revision 23, dated June 24, 1983, Procedure for Corrective Maintenance, requires immediate a
l investigation of plant problems and initiation of a Maintenance i
Request Form (MRF) for problems that cannot be corrected within eight hours.
However, problems with testing and operating the RWM and RSCS during a plant shutdown on November 17, 1983, were not sufficiently investi-gated to correct the problem within eight hours, and no MRF was i
initiated, b.
Administrative Procedure A-47, Revision 2, dated April 14, 1980, Procedure for the Generation of Surveillance Tests, requires that surveillance test procedure steps which document completion of 4
Technical Specification related surveillance requirements to be indicated with an asterisk. The test results section shall be
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signed only if all asterisked steps are completed satisfactorily.
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Technical Specification Surveillance Requirement 4.3.B.3a states that t
the group notch mode of RSCS shall be demonstrated to be operable by j
attempting to move a control rod more than one notch in the first pro-l gram group after reaching 50 percent rod density on a reactor startup.
However, ST10.6, Revision 10, dated July 18, 1980, Rod Sequence Control System (RSCS) Function Test, was written and implemented without mak-ing the technical specification requirement an asterisked step.
As a result, completed tests do not contain documentation of the completed l
technical specification surveillance requirement, and they were
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signed off as satisfactory.
C.
Surveillance Test Procedure ST10.5, Revision 11, dated July 18, 1980, RWM Operability Check, requires, in an asterisked step, selection and listing of at least three rods to verify operability of the RWM rod select error function.
l However, on May 28, 1983, ST10.5 was completed and signed off as satisfactory when only one rod was listed as having been used to l
verify the operability of the rod select error function.
l This is a Severity Level IV violation (Supplement I).
I I
i Pursuant to the provisions of 10 CFR 2.201, Philadelphia Electric Company is hereby required to submit to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, and a copy to the Regional Administrator, U.S. Nuclear Regulatory Connission, Region I, 631 Park Avenue, King of Prussia, PA.19406, within 30 days of the date of this Notice, a written statement or explanation, including for each alleged violation:
(1)
. admission or denial of the alleged violation; (2) the reasons for the violation, if admitted; (3) the corrective steps which have been taken and the results achieved; (4) the corrective steps which will be taken to avoid further viola-tions; (5) the date when full compliance will be achieved. Considerations may be given to extending the response time for good cause shown.
Under the author-ity of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
Within the same time as provided for the response required under 10 CFR 2.201, Philadelphia Electric Company may pay the civil penalties in the amount of $30,000 or may protest imposition of the civil penalties, in whole or in part, by a writ-ten answer.
Should Philadelphia Electric Company fail to answer within the time specified, the Director, Office of Inspection and Enforcement, will issue an order imposing the civil penalties proposed above. Should Philadelphia Electric Company elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, such answer may:
(1) deny the violations listed in this Notice j
in whole or in part; (2) demonstrate extenuating circumstances; (3) show error in this Notice; or (4) show other reasons why the penalties should not be imposed.
l In addition to protesting the civil penalties, in whole or in part, such answer may request remission or mitigation of the penalties.
In requesting mitigation j
of the proposed penalties, the five factors contained in Section IV(B) of 10 CFR Part 2, Appendix C should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation by specific reference (e.g., citing page and paragraph numbers) to avoid repeti-tion. Philadelphia Electric Company's attention is directed to the other provi-sions of 10 CFR 2.205, regarding the procedures for imposing civil penalties.
Upon failure to pay any civil penalty due, which has been subsequently deter-mined in accordance with the applicable provisions of 10 CFR 2.205, this matter l
may be referred to the Attorney General, and the penalty unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section j
234c of the Act, 42 U.S.C. 2282.
FOR THE NUCLEAR REGULATORY C0milSSION jf f..-
l Richard C.
Young,-
rector Office of Ins'pectiorvand Enforcement Date at Bethesda, Maryland j
thisff%iay of June 1984 i
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'o UNITED STATES g
8' NUCLEAR REGULATCRY COMMISSION g
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REGION I G
631 PARK AVENUE
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FEB 2 91984 MEMORANDUM FOR:
D.G.Eisenhut, Director,DivisionofLicensing,NRR/
J. fieltemes, Director, Office for Anaysis and Evaluation of Operational Data T. T. Martin, Director, Division of Engineering and Techni-cal Programs, Region I FROM:
R. W. Starostecki, Director, Division of Project and Resident Programs, Region I
SUBJECT:
EXTENSION OF THI-1 SALP ASSESSMENT PERIOD In order to provide a more current assessment of license performance prior to an NRC Commission decision on the restart of TMI-1 (currently projected for June 1984), we are hereby extending the assessment period by four months, from Sep-tember 30, 1983 to January 31, 1984. Consequently, the assessment will cover a sixteen (16) month period.
Please review the SALP input you previously provided, update the material as necessary for the, four additional months and submit the update to E.L. Conner of my staff by March 16, 1984.
If no update of your previous input in required, please advise Mr. Conner (FTS 488-1355) of that fact. We expect to have a SALP Board raeeting on April 2,1984.
Your efforts and cooperation are appre P
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R. W. Starostecki, Director Division of Project and Resident Programs cc:
T. Murley, RI J. Allan, RI DPRP Branch Chiefs DETP Branch Chiefs DPRP Section Chiefs DETP Section Chiefs J. Stolz, NRR J. Van Vliet, NRR C 'Tb@ I)
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GEGULATuPY IfdF0dvATION DISTRIBUTIUM SYSTEM (RIOS)
ACCE3FION.NoR: A405300277 OuC.DATE: 84/05/14 NOTARIZED: NO 00CMET s
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FACIL':50-277 Peach Bottom atomic Power Station, Unit 2, Philadelph 05000277 50-278 Peach Bottom atomic Power Station, unit 3, Ph11adelph 0500u27P AUTH.hAME AUTHyk AFFILIATION STAROSTECKI,R.
division of Project & Hesioent Programs RECIP.NAvt dECIPicNT AFFILIATION DALTROFF,S.L.
Anilaceipnie Electric Co.
SUBJECT:
Forwards IE Insp depts 50-277/84-11 & 50-278/84-11 on OISTRIBUT N
O L: IE01S LOPIES RECEIVED:LTR ENCL SIZE:_
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TITLE: General (50 OktJ-Inso Rect / Notice of Violation Hesponse NOTES:
RECIPIEtiT COPIES RECIPIENT COPIES
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1 GE RS G INTERAAL: AE00 1
1 ELD /HDS4 1
1 IE FILE 01 1
1 IE/00ASIP/ORPB 1
1 IE/ES 1
1 NRR/DSI/RAR 1
1 EXTEHAAL ACn3 2
2 LPDR 1
1 HRC PDR 1
1 NSIC 1
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TOTAL NUMbEk 0F COPIES 4 4UARED: LTTR 14 ENCL 14
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NAY 141984 Docket / License: 50-277/DPR-44,50-278/DPR-56 Philadelphia Electric Company ATTN: Mr. S. L. Daltmff Vice President Electric Production 2301 Market Street Philadelphia, Pennsylvania 19101 Gentlemen:
Subject:
Enforcement Ccnference 50-277/84-11; 50-278/84-11 This refers to the enforcement conference held at our request at the NRC Region I Office, King of Prussia, Pennsylvania, on April 12, 1984 relating to the Peach Bottom Atomic Power Station. The conference was attended by yourself and other members of your staff and by myself and other NRC staff members. We believe that the conference was beneficial and improved mutual understanding of the matters involved.
The items discussed at the enforcement conference are described in the enclosed report. NRC analysis of enforcement considerations for these items is continu-ing.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of this letter. Such application must be consistent with 10 CFR 2.790(b)(1).
Telephone notification of your intent to request withholding, or any request for an extension of the ten day period which you believe necessary, should be made to the Supervisor, Files, Mail and Records, USNRC Region I, at (215) 337-5223.
No reply to this letter is required. Your cooperation with us is appreciated.
Sincerely,
- Original Cisned 33
.we-n _macu4 Richard W. Starostecki, Director pga Apocn oSooo277 Division of Project and Resident N
e Programs
Enclosure:
NRC Region I Enforcement Conference Report 50-277/84-11; 50-278/84-11 "0FFICIAL RECORD COPY" g
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gay I 41984 cc w/ encl:
R. S. Fleischmann, Station Superintendent Troy B. Conner, Jr., Esquire (Without Report)
Eugene J. Bradley, Esquire (Without Report)
Raymond L. Hovis, Esquire (Without Report)
II, Assistant Attorney General (Without Report)
MichaelJ.Scibinico,(PDR)(LPDR)
Public Document Room Local Public Doceent Room Nuclear Safety Infonnation Center (NSIC)
NRC Resident Inspector comonwealth of Pennsylvania bec w/ encl:
Region I Docket Room (with concurrences)
Troy B. Conner, Jr., Esquire Eugene J. Bradley, Esquire R p nd L. Hovis Esquire Michael J. Scibinico, II, Assistant Attorney General Senior Operations Support Office (w/o encis)
Section Chief, DPRP hI n
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8 U. S. NUCLEAR REGULATORY COMMISSION Region I 50-277/84-11 Report No. 50-778/84-11 50-277 Docket No. 50-278 DPR-44 License No. DPR-56 Priority Category c
Licensee:
Philadelohia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name: Peach Bottom Meeting at: USNRC, Region I, King of Prussia, Pennsylvania Meeting conducted:
April 12, 1984 A.
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AnspecLOP dd te 'si gned NRC Personnel:
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[.H. Williams,ResidentInspector d&te ' signed y
date signed Approved by:
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$0!M C. E. TripY, Chief Mats signed Reactor Projects Section 3A Enforcement Conference on April 12, 1984 (Report 50-277/84-11; 50-278/84-11)
Meetino Summary: Special enforcement conference convened to discuss findings of Region 1 Inspections 278/83-32, 277/84-01 & 278/84-01, and 277/84-03
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and 278/84-03, relative to individual rod scraming and several LCO violations.
Senior Philadelphia Electric Company, NRC Region I and I&E management personnel attended this two hour meeting at the Region I office.
l hPDR ADOCK 05000277 G
PDR Region I Form 12-1 (Rev. August 77)
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DETAILS
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1.
Attendees Philadelphia Electric Company V. S. Boyer, Senior Vice President Nuclear Power S. L. Daltroff, Vice President, Electric Production M. J. Cooney, Manager, Nuclear Production s
W. T. Ullrich, Superintendent, Nuclear Generation Division R. S. Fleischmann, Station Superintendent, Peach Bottom G. M. Leitch, Station Superintendent, Limerick R. H. Logue, Superintendent, Nuclear Services J. E. Winzenried. Technical Engineer, Peach Bottom S. R. Roberts, Operations Engineer, Peach Bottom L. F. Rubino, Engineer in Charge, Fuel Management Section J. W. Spencer, Startup Director, Limerick J. M. Corcoran, Limerick, Quality Assurance J. L. Collings, Bechtel, Project Operations, Limerick U.S. Nuclear Regulatory Comission T. E. Murley, Regional Administrator J. M. Allan, Deputy Regional Administrator R. W. Starostecki, Director, DPRP R. R. Keimig, Chief, Projects Branch #3 L. E. Tripp, Chief Section No. 3A A. R. Blough, Sr. Resident Inspector J. H. Williams, Resident Inspector J. M. Gutierrez, Regional Counsel D. S. Holody, Enforcement Coordinator J. A. Axelrad, Director, Enforcement, ISE P. R. Farrow, Enforcement Staff. I&E 2.
Meeting Purpose To discuss events involving individual control rod scramming, reactor heatup rates, reactor vessel pressurization at low temperature, and slow control rod scram times.
3.
Events of Concern Tiie NRC identified the following violations as being of concern.
3.1 Individual rod scraming for nomal shutdown.
(see NRC:RI Reports 278/83-32 and 84-011
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t 3.2 Excessive reactor heat-up rates.
(See NRC:RI Report 84-03.)
3.3 Reactor pressurization at low tenperature.
(See NRC:RI Report 84-03.)
3.4 Inoperable control rod - slow scram time.
(See NRC:RI Report 84-03.)
4.
Cause and Safety Significance 4.1 The NRC stated that the individual rod scramming violation appeared to be caused by inadequate 50.59 review of procedure changes associated with normal plant shutdown. The safety significance is a potential rod drop i
accident which results in fuel damage.
4.2 The NRC stated that the excessive heatup rate violations appeared to be caused by human error, inadequate supervision, and procedural guidance.
These events were of minor safety significance.
4.3 The NRC stated that the reactor pressurization et low temperature viola-tion appeared to be caused by a poorly written procedure and failure to follow the procedure. This event was of minor safety significance.
4.4 The NRC stated that the inoperable control rod violation appeared to be caused by inadequate review of trainee's work. The safety concern over this event relates to shutdown margin requirements.
5.
Licensee Discussions 5.1 Causes Philadelphia Electric Company management acknowledged the events and causes.
In response to questions raised at a January meeting with NRC as to whether the plant was in an analyzed condition for a Rod Drop Accident (RDA) during the November 17, 1983 shutdown utilizing individual rod scrams, the licen-see presented results of a GE study. A copy of the study will be provided to NRC. The study concluded that no safety problem existed when the reac-tor was above 10% power. Below 10% power, with worst case operator error and FSAR analyses techniques, the results were outside the design basis for RDA. Using moderator feedback effects, as discussed in BNL/NUREG-21819, the RDA was within the FSAR bounds. The 10 CFR 50.59 review con-ducted for the 1977 procedure change to allow individual rod scramming was reconstructed from memory. Reviews for procedure changes were not well documented at that time.
5.2 Licensee Initiated Corrective Actions Stopped practice of individually scramming control rods.
Restored RWM to vendor recomended sequence.
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Revised appropriate procedures.
Strengthened procedural controls.
PORC will better document procedure changes in the future.
Senior engineers were requested to review their areas for weaknesses in procedures.
Management issued instructions to operators to maintain heatup rates of 60-800F per hour. Procedures changed to indicate this.
Placed line on chart recorder for operator guidance.
Simulator training modified to emphasize importance of staying with-in heatup limits.
Directed inside supervisor to oversee operators actions more closely.
Procedure on valve lineups for establishing short and long path re-circulation revised.
Initiated an in-depth job task analysis.
Tested all scram solenoid prior to Unit 3 startup after discovering rod with slow scram time. Also tested backup scram valves.
Revised test procedure ST10.9, CRD Scram Insertion Timing.
Tested scram time of each rod during power ascension.
4 Reviewed QA/QC controls on scram pilot valve rebuild kits and replace-ment parts.
Stopped using Loctite 242 on scram pilot valves.
Designed means of acoustically monitoring solenoid valve movement and l
implemented weekly check for solenoid plunger dropout during half-scrarm. The test frequency has been changed to biweekly at this time.
Initiated engineering evaluation of impact of excessive heat rate on reactor vessel.
6.
Conclusion The licensee's corrective actions were discussed in some detail. The NRC thanked the licensee for his input and indicated that NRC review would be facilitated by the licensee-provided information.
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