ML20117C507
| ML20117C507 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Peach Bottom |
| Issue date: | 04/10/1984 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Keimig R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20114F930 | List:
|
| References | |
| FOIA-84-616 NUDOCS 8505090444 | |
| Download: ML20117C507 (11) | |
Text
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l APR I O 1984 MEMORANDUM FOR:
R. R. Keimig, Chief, Projects Branch No. 3, DPRP
[. E. Tripp, Chief, Reactor Projects Section 3A, DPRP i
/offi THRU:
FROM:
A. R. Blough, Senior Resident Inspector, Peach Bottom
SUBJECT:
ENFORCEMENT CONFERENCE WITH PHILADELPHIA' ELECTRIC COMPANY ON APRIL 12, 1984 Peach Bottom Issues 1.
Overview This conference covers the issue of individual rod scramming (Inspection 84-01) and several LCO violations.
Although the LCO violations, taken in-dividually, are of relatively minor significance, they collectively indicate lack of attention to detail in (1) conducting and supervising routine oper-ations and (2) controlling and supervising both trainees and recently quali-fied operators.
2.
Event Description The events, causes, safety significance, and initial licensee corrective ac-tions are summarized in Table 1.
Inspection report details ~ and LER's are attached.
Additional long-term corrective actions may be presented at the enforcement-conference.
3.
Resident Inspector Perceptions 3.1 Individual rod scramming procedures were developed in l'977 due to lic-ensee concern over thermal cycling of CRD collect and feedwater nozzles in full-scram shutdowns.
Inadequate consideration was given to the intent of the Technical Specifications and the intended functioning of the Rod Sequence Control System and the Rod Worth Minimizer during reactor shutdown.
3.2 Extensive corrective actions were taken for the individual rod scram-ming and slow rod scram issues. Corrective action has been less exten-sive for the other violations. This may be due to (1) their relatively minor safety significance, (2) licensee belief that each was primarily an individual personnel error, (3) reduced reportability for these items, and (4) heavy licensee workload due to the INPO audit.
Licensee corrective actions can be further evaluated at the conference.
gh Senior Resident Inspector g 50 g 4 841002 BARFIEL84-616 PDR
Memo for R. R. Keimig 2
APR 10 884
Enclosures:
As Stated cc w/encis:
T. Murley, RI J. Allan, RI R. Starostecki, RI L. Tripp, RI D. Holody, RI J. Williams, RI J. Axelrad, IE J. Gutierrez, RI 0
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a TABLE 1 PEACH BOHOM ENTORCEMENT CONFEREffCE ISSUES Event Causes Safety slantricance Licensee Corrective Action Reduced protection nqninst unnecopt.nble (1) Ceased practice Individual Rod Scrneming (1) Inadequate 50.59 review consequences in nynut. nr rod drop acci- (2) Rostered RWM to vender far Norma l Shutdown of procedurn chanan, (2) Lock of 50.59 review or dont during reactor shutdown.
recommended sequence.
(3) Pnvised,and strengthed Rod Worth Minimizer l
procedurel controls.
sequence charge, (4) Further studies and/or (3) Inappropriate interpre-analyses planned.
tation of Technical j
l Specifications Exc3ssive Heat-Up Rates (1) Personnel error.
Probably minimal. Slightly higher (1) Reinstructed personnel.
Unit 3 on 1/24/84 (2) Inadequate supervision stresses on the vessel than in a normal (2) Discussed at shirt meet-Unit 2 on 1/31/84 o f t ra i nee s.
- hentup--may result in need.to slightly
- ing,
,l (3) Lack of SRO oversight adjust therma l cyclo ana lyses.
(3) Initleted engineering evaluation.
i of heatup.*
7 (4) Inadequate procedural guidance and cautions on heatup.*
Unplcnned Reactor Pres- (1) failure to follow proce-Minimal - Transient promptly termin-(1) Discussed at shift meet-surization (10 psig at du re--Imp rope r va lve ated by operator actions.
Ings.
(2) Reinstructed personnel 110*f) line-up.
J (2) Inadequate supervision.
(3) Poor human factors in p rocedu re. '
Inoptrable Control Rod (1) Insppropriate guidance Shutdown margin requirements were (1) Very extensive--see re-( Slow Sc ram T ime ),
in GE Sil.
probably not met for the case in which port excerpt.
11/17/83-1/14/84 (2) Inadequate review or rod edJacent to the Inoperabie one also t ra i nee wo rk. '
falls to scram.
(3) Inadequate supervisory review of test.'
(4) Inadequate procedural guidance.
- inspector Opinion l
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10.2 On-site Followur
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pqj e, For LERs selected for ensite review (denoted by asterisks above), the inspector verificd that appropriate corrective action was taken or responsibility assigned and that continued operation of the facility was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as ' defined in 10 CFR 50.59.
Report accuracy, compliance with current reporting requirements and applicability to other site systems and components were also reviewed.
This report updated an LER of December 7,1983, which LER 3-83-18/3X-1.
17, 1983.
After reported slow scramming of one control red on November 14, 1984, it was found that an additional a subsequent scram on January control rod, 34-27, had scrammed slowly on both November 17,1983, and January 14, 1984.
As a result of these occurrences, the licensee took the following actions:
e
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16 (1) The pilot air solenoid valves for each of the two control rod i
hydraulic control units (HCU) were sent to the NSSS vendor for exam-ination.
A yellow varnish-like substance was found to have stuck the plunger in the " energized" position on one solenoid valve from each HCU.
Both the licensee and the MSSS vendor detemined, through chemi-cal analyses, that the substance was Loctite 242 adhesive / sealant, which was applied per NSSS vendor recomendations to the solenoid
'i housing cap nut to prevent loosening.
All 370 Unit 3 solenoid valves had been rebuilt during the last refueling outage to replace. limited shelf-life parts.
(2) Prior to Unit 3 startup, the licensee disassembled and inspected 40 additional solenoid valves.
One of the 40 showed minute traces of a yellow substance, but apparently had not stuck.
All scram solenoids were functionally tested prior to startup.
(3) The licensee reviewed all scram time data available for the cur-rent cycle at each unit.
This included reevaluation of the scram time recorder strip charts that are printed out after each scrat.
No other slow scramming rods were identified.
In the case of the January 14 scram, operators had demanded a process computer scan of control rod positions shortly (i.e. within 30 seconds) after the scram.
This scan indicated that, as of the scan time, rod 34-D was the only one that had not moved.
Since the automatic scram time recorders are capable of timing only 58 of the 185 rods during any one scram, the licensee instructed operators to demand control rod scans shortly after each scram, as was done on January 14, in the future.
Also, the licensee
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determined that the failure to identify the slow scram of rod 34-27 during initial review of scram time data from November 17 was due to Rod 34-27 misinterpretation and inadequate review of the strip chart.
was the first of 30 rods printed across one chart.
The continuous line printed by the chart (indicative of the rod remaining at position 48 for the entire twelve second duration of the print-out) was mis-interpreted as an ordinate of the graph.
The LER comitted to a revision of the test procedure, ST10.9, CRD Scram Insertion Timing, to include detailed instructions and sample chart traces for identifi-cation of abnomal scram times. This inspector will review the pro-cedure when revised (278/84-03-04).
- (.4) The licensee tested the backup scram valves prior to startup.
(A rod whose solenoid valve sticks will still scram, after a time delay of typically 25-45 seconds, if backup scram valves actuate as designed to depressurize the entire scram air header.) The back-up scram valves, which are not safety-grade, operated properly and de-pressurized the header in about 25 seconds in the test.
(5) The licensee tested the scram function of each rod prior to startup and scram time tested each rod during power ascension.
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m 17 (6) A licensee task force reviewed QA/QC controls on scram pilot valve This review, which included a rebuild kits and replacement parts.
Visit to the vendor facilities, concluded that the problem with valve sticking did not originate in the vendor (ASCO) facilities or during NSSS vendor (GE) storage and reshipment of parts.
(7) The licensee contacted the manufacturer of Loctite 242, who stated The licensee has suspended that Loctite 242 has a tendency to migrate.
its use on scram pilot valves and is evaluating alternatives.
The NSSS vendor is perfoming similar evaluations and attempting to detemine if
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similar problems have occurred at other plants.
(8) The licensee designed means of acoustically monitoring solenoid positions and implemented a weekly check for solenoid plunger drop-out This testing will be continued until the licensee i
during half-scrams.
l has obtained sufficient data to assure on-going system reliability.
(9) The licensee participated in development of an INPO Significant Event Report, planned for issuance in early March 1984.
The inspector verified selected licensee actions through review of completed tests, observation of in-progress scram time testing, inter-views of personnel, and examination of disassembled solenoid valves.
The inspector forwarded this issue to NRC Region I management for evaluation of generic aspects.
i The inspector reviewed the Technical Specification requirements appli-cable to plant operation between November 17, 1983 and Jaauary 14, 1984, the time period during which rod 34-27 had an excessive scram time.
Technical Specification (TS) 3.3.C.3 states that scram insertion time for 90 percent insertion of any operable control rod shall not exceed 7 seconds.
TS3.3.A.2 requires any inoperable rods to be positioned TS3.3.A.1 requires operability of suf-such that TS3.3.A.1 is met.
ficient control rods such that the core could be made sub-critical in the most reactive condition during the operating cycle with the strong-rods fully est control rod fully withdrawn and all other operable Because the review of data from the November 17 scram failed inserted.
to identify the slow scram time of rod 34-27, the licensee subsequently operated the plant with rod 34-27 fully withdrawn and considered it
- operable, in apparent Violation of Technical Specification 3.3.C.3 (278/84-03-06).
Also, because the rod was not declared inoperable, it was not positioned such that shutdown margin requirements were assured; thus TS3.3.A.1 and 3.3.A.2 were also apparently violated in this event.
i Rod 34-27 is a relatively high worth rod located near the center of the Had its excessive scram time been known, appropriate action during plant operation would have been either (1) to fully insert the rod un-core.
til repaired, or (2) to perform detailed, worst-case shutdown margin analyses if operation with the rod not fully inserted was desired.
gg3fg;g ope h valves a Service Water System g
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were also checked.
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Followup on Events Occurring Durino the Insp7ction On January 24, 1984, Excessive Heatup Rate - Unit 3, January 24. trainee workin 3.2 while startup of Unit 3 was in progress, a taking plant 3.2.1 the supervision of the licensed reactor operator was hi l Spect-heatup data required by paragraph 4.6.A.1 o i
te time the reactor coolant temperature change over fications.
in an hour increase in the coolant temperature of over 100"
'B' recirculation loop indi-interval.
increase of cated 230 F at 9:15 a.m. and 332 F at 10:15 a.m., a
'B' recirculation loop had occur 5ed.
increase indicated 2450F at 9:30 a.m. and 3560F at 10:30 a
'A' recirculation 102 F in one hour.
0 9:30 - 10:30 a.m.
of 111 F in one hour. loop indicated a 1100F increase du i
0 Exceeding the heatup rate is one of two such 0
Technical Specifications allow a 10 period.
lations (278/04-03-01).
reduced and temperature change in any one hour.the excess the heatup proceeded at a much slower rate.The inspector ve licensee's correc-for corrective action.)by the licensee and will con tive action.
idelines for The inspector asked whether the licensee had an control of trainees on-shift.
policy that the qualified individual is fully resThe lified individual actions is solely at the discretion of the qua shift duties.
0 CFR 50 (except for duties designated as " licensed ator trainees ator and reviewed the training schedule of the opert d and 10 CFR 55).
to determine if the trainee had been instruc e.d and cooldowns and specifically been instructe15, 19E1, test procedure STg.12, Revision 5, DecemberIt was d ST9.12 are Vessel Temperatures. instructions as well as instruc scheduled for later in the training program.
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3.2.2 Excessive Heatup Rate, Unit 2, January 31.
On January 31, 1984, while Unit 2 was being started up,0F over a one hour period.another appare the heatup rate exceeded 100 The reactor i
was critical at 4:38 a.m.
Between 4:43 and 4:54 a.m., the operator withdrew control rods 38 notches.
This caused a heatup of 47oF over a 15 minute time period (4:45 to 5:00 a.m.).
At 5:05 a.m., the oper-ator inserted a control rod 3 notches to decrease the heatup rate.
At 5:07 a.m..'he inserted the control rod full in (7 notches).
8e-tween 5:10 and 5:16, the operator inserted additional control rods 22 notches in attempting to decrease the heatup rate.
The A & B recir-culation loop reactor coolant temperatures increased an additional 450F between 5:00 and 5:15 a.m.
The reactor coolant temperature, as indicated by the recirculation loop A & B recorder traces, increased 110 F between 4:20 a.m. and 5:20 a.m.
Technical Specification allows 0
a 100 F reactor coolant temperature change in any one hour.
This is 0
an apparent violation (277/84-03-01).
After this event, the operators were instructed to be more careful in maintaining heatup rates within the technical specification limits.
They were also instructed to use diverse instrumentation as a means of checking heatup rate.
The lic-ensee stated he plans to add a heatup curve on the chart recorder as an operator aid.
In addition, ST9.12 will be required to be initated earlier in the startup. The licensee is perfoming an engineering evaluation of the effects of the heatup rate on the reactor vessel.
The inspector verified the actions taken by the licensee and will con-tinue to follow the licensee's corrective actions.
The inspector ex-amined the chart recorder traces of the event and the rod pull sequences to reconstruct the event and verify the licensee descriptions.
The inspector reviewed vendor training materials and noted that the nega-tive temperature coefficient of reactivity for this event (near end-of-cycle) could be expected to be very small, markedly less than at beginning of cycle.
The inspector noted that licensee procedures do not provide guidance on how to establish the proper heatup rate (e.g., expected amount of rod movement and power levels), and do not provide reminders on the (predictable) manner in which the temperature coefficient of reactivity varies with core life and temperature.
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3.'2. 3 Unplanned Cooldown with Slight Reactor Vessel Pressurization--Unit 3, l
January 25.
About 5:30 p.m., January 25, with the reactor about 135oF in Cold Shutdown, operators were shifting the condensate and feedwater i
system from short path to long path recirculation. When the procedure was perfomed, feedwater injection isolation valves, which were re-l quired to be closed, were left open.
As a result, an injection path to the reactor was established in addition to the recirculation path.
Reactor Water level rose rapidly from about 40 inches to about 110 inches.
Reactor temperature dropped to about 1100F due to the injec-j l
tion of cold feedwater.
At the same time, the reactor pressure rose I
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brier.y (i.e., for about five minutes) to about 10 psig, because the or.e-inch reactor vessel head vent path was inadequate tc maintain the vessel at atmospheric pressure during such a rapid level transient.
The normal vent path, the three-inch main steam line drain, was iso-lated due to previous maintenance.
Technical Specification 3.6.A, Thermal and Pressurization Limits, and its associated temperature-pressure graph, Figure 3.6.2, prohibit reactor pressurization above atmospheric pressure at temgeratures be-low 120 F.
Pressurization to about 10 psig at about 110 F is.there-0
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fore an apparent violation (278/84-03-02).
The inspector reviewed logs and recorder charts and discussed this The licensee stated that the event with licensee station management.
causes appeared to be operator error (failure to follow the procedure)
I and weak supervisory control and communication during the evolution.
t The licensee plans appropriate retraining, counselling or disciplinary action.
Further licensee evaluation of this event is in progress.
The inspector recommended that the licensee also (1) review human factors aspects of the procedure involved, in that the procedure appars unclear with respect to positive verification of the process valves involved, and (2) review procedural guidance with respect to reactor vessel vent path selection.
l,sh 3.2.4 Unit 3 Scram - February 9, About 6:06 p.m., February 9, with the unit
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Backgraund On November 17, 1983, the licensee was shutting down to r2placo a main steam safety-relief valve which was giving acoustic indications of pilot valve seat leakage.
At 10:30 p.m., a turbine high vibration alarm was received.
Reactor shutdown was accelerated through individually scramming rods, which was allowed per PORC approved procedure GP-3.
By 10:34 p.m., about 10 rods had been scrammed and the turbine was taken off-line.
Individual rod scram-ming continued.
About 10:36 p.m., a scram discharge instrument volume (SDIV) high level rod block annunicated; although rod scrams were then suspended, a SDIV high level scram followed shortly.
Licensee analysis indicated that about 25 additional rods had been scrammed in two minutes (from 10:34 p.m.
to 10:36 p.m.) from the scram time test panel.
Since many ir.dividual rod scram switches were left in the " scram" position, scram discharge volume in-leakage from control rod drives was signific. ant ano continuous, exceeding 3
the small SDIV drain capacity.
Thus, an actual SDIV high-level condition resulted.
The licensee promptly issued instructions to limit both the rate of individual rod scrams and the number of switches remaining in the scram position.
The NRC questioned, however, the licensee's justification for individually scramming rods, in light of Rod Worth Minimizer (RWM) and Rod Sequence Control System (RSCS) operability requirements.
In particular, NRC expressed concern that one of the intended functions of the' RWM and RSCS (te restrict insertions of control reds to prespecified sequences to minimize the power excursion and possibility of fuel damage if a control rod drop accident was to occur) was lost by this mode of operation.
Individual rod scramming circumvents RWM and RSCS controls which place constraints upon the Reactor Manual Control Systems (RMCS).
Scramming does not involve the RMCS.
l The licensee stated that the practice had been permitted by licensee proce-dures for several years; therefore, time would be required to research its origin and justification.
Consequently, on December 1, the licensee com-mitted, in response to the senior resident inspector's request, to suspend l
the use of individual rod scrams for purposes other than either testing or ATWS response.
t l
This special inspection, which included rcsident inspector efforts January l
5-20 as well as a January 12 meeting onsite with senior PECo, NRC:IE, and Region I personnel was to review both the November 17, 1983 event and the i
issue of individually scramming rods for normal shutcewns.
3.
Licensee Meeting l
l The licensee and NRC personnel listed in Detail 1 met on January 12, 1984.
l The NRC identified the meeting as part of the information gathering process for the special inspection.
l 3.1 Licensee Presentations
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The insp;cter evaluated the fo11cwing licansee actions relative to 10 CFR 50.59.
4.1.1 Approval of Procedures that Allowed Individual Rod Scrams.
Individual rod scramming was allowed through implementation of Revision 10 to Gp-3, Normal Plant Shutdown, on April 8, 1977, with prior PORC procedure approval.
Subsequent revi-sions of GP-3, as well as GP-9 (Fast Reactor Power Reduc-tion), including those revisions in effect on November 17, also allowed individual rod scramming.
Technical Specifications require the RW and RSCS to be oper-able, below 25 percent power and 21 percent power respec_
tively, during shutdowns.
Final Safety Analysis Report Sec-
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tions J.4.13 and 7.16.3 describe the RW and RSCS control rod sequences and operations during startup and shutdown.
Sec-tion J.4.13 states that the purpose of the RSCS is to prevent the operator from moving an out-of-sequence rod during start-up or shutdown.
Section 7.16.3.3 states that the RW supple-ments the operator through a control rod monitcring routine that enforces adherence to startu;;, shutdown and low power L
level control rod mevement procehes.
Indivicual rod scram-ming is independent of RW anc RSCS controls and therefore renders RW and RSCS incapable of either enforcing a sequence or preventing movement of an out-of-sequence rod.
Thus, a Technical Specification change, with prior Commission appro-val, was necessary for procecural changes that institute in-dividual rod scramming.
Failure to obtain prior Technical Specification changes and Commission approval for procedural changes affecting functioning of the RW and RSCS is an apparent Violation (first of two examples).
4.1.2 RW Program Changes.
As noted in Detail 3.2.1, the licensee had changed one of two RW rod sequences in order to be com-patible with the individual rod scram practice.
Licensee personnel said the change had occurred in about 1979.
No PORC review or safety analysis was performed.
Section 7.16.3.3 of the FSAR states that the operator can select either one of twc permissible sequences.
However, after the 1979 char.ge, only one of the two secuences could be used for startue since the other was designed (onsite) for shutdown only.
FSAR Section J 4.13 indicates that rod secuencing must be strictiy adhered to during shutcown and is basically the reverse of startup.
The shutdown sequence was not the reverse of an allowable startup sequence.
Further, FSAR Section 7.16.3.3 states that the RW sequences stored in the computer memory are based on procedures designed to limit rod worths to acceptable levels as determined by the design basis rod drop accident (RDA).
As noted in Detail 3.2.1, the licensee did not verify compati-bility of his RW program for shutdown with the vendor's RDA analysis referenced in the FSAR.
Thus, the 1979 revision to the RW program changed the system as described in the FSAR, yet no written safety evaluation of this change was made, nor were formal records of the change maintained.
This is an additional example of the Violation noted in Detail 4.1.1 above (277/84-01-01, 278/84-01-01).
4.2 RW and RSCS Status Durine the November 17 Shutdown
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p ter UNITED STATES o
NUCLEAR REGULATORY COMMisslON g
REGION I g-g g
<E S31 PARK AVENUE t
KING OP PRUSSIA, PENNSYLVANIA 19406
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S....l MAY 141984 Docket / License:
50-277/DPR-44,50-278/DPR-56 Philadelphia Electric Company ATTN: Mr. S. L. Daltroff Vice President Electric Production
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2301 Market Street Philadelphia, Pennsylvania 19101 Gentlemen:
Subject:
Enforcement Conference 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 sqyself 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-1 ing.
l 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, Richard W. Starostecki, Director Division of Project and Resident Programs
Enclosure:
NRC Region I Enforcement Conference Report 50-277/84-11; 50-278/84-11
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Philadelphia Electric Company 2
ggy 141984 ccw/ enc 1:
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)
Michael J. Scibinico.(PDR) (LPDR)
Public Document Room Local Public Document Room Nuclear Safety Infonnation Center (NSIC)
NRC Resident Inspector Connonwealth of Pennsylvania bec w/ enc 1:
Region I Docket Room (with concurrences)
Troy B. Conner, Jr., Esquire Eugene J. Bradley, Esquire Raymond L. Hovis. Esquire Michael J. Scibinico, II, Assistant Attorney General Senior Operations Support Office (w/o encis)
Section Chief. DPRPy l
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j U. S. NUCLEAR REGULATORY COMMISSION Region I 50-277/84-11 R:: port No. 50-278/84-11 50-277 Docket No. 50-278 DPR _
License No. DPR-56 Priority Category c
Licensee:
PhiladelDhia 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 f'fM/gV/
NRC Personnel: (-/
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[.H. Williams,ResidentInspector d6 te 'si gned date signed y
Approved by:
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$C C. E. Tripi, Chief tats sianed
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Reactor Projects Section 3A Enforcement Conference on April 12, 1984 (Report 50-277/84-11; 50-278/84-11)
Special enforcement conference convened to discuss findings of Meeting Summary: ions 278/83-32, 277/84-01 & 278/84-01, and 277/84-03 Region 1 Inspect 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.
hQQ' Region I Form 12-1 g
(Rev. August 77)
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e DETAILS l
1.
Attendees r
Philadelphia Electric Company V. S. Boyer, Senior Vice President, Nuclear Power i
S. L. Daltroff, Vice President, Electric Production M. J. Cooney, Manager, Nuclear Production W. T.' Ullrich, Superintendent, Nuclear Generation Division R. S. Fleischmann, Station Superintendent, Peach Bottom j
G. M. Leitch, Station Superintendent, Limerick R. H. Logue, Superintendent, Nuclear Services 4
J. E. Winzenried. Technical Engineer, Peach Bottom S. R. Roberts, Operations Engineer, Peach Bottom 2
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 scraming, reactor heatup rates, reactor vessel pressurization at low temperature, and slow control i
rod scram times.
3.
Events of Concern The NRC identified the following violations as being of concern.
3.1 Individual rod scraming for ncrmal shutdown.
(see NRC:RI Reports 278/83-32 and 84-01).
3 3.2 Excessive reactor heat-up rates.
(SeeNRC:RIReport84-03.)
3.3 Reactor pressurization at low temperature.
(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 l
4.1 The NRC stated that the individual rod scraming 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.
S.
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 i
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 scraming 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 scraming control rods.
Restored RWM to vendor recomended sequence.
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4 Revised appropriate procedures.
Strengthened procedural controls.
PORC will better document procedure changes in the future.
Seni'or 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.
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 implemented weekly check for solenoid plunger dropout during half-scrams. 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 i
the licensee-provided information.
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UNITED STATES g
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JUN 18 M i
Docket Nos. 50-277
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50-278 EA 84-39 Philadelphia Electric Company ATTN: Mr. V. Boyer Senior Vice President - Nuclear 2301 Market Street Philadelphia, Pennsylvania 19101 Gentlemen:
Subjects: 1.
Order Modifying License Effective Immediately 2.
Notice of Violation and Proposed Imposition of Civil Penalties On April 12, 1984, an Enforcement Conference was held by Dr. Thomas E. Murley, Regional Administrator, Region I with you and members of your staff at the NRC Region I Office to review the circumstances associated with apparent violations of NRC requirements which occurred at the Peach Bottom Atomic Power Station, Units 2 and 3.
Two of the violations were identified by the NRC during an NRC inspection conducted January 5-20, 1984. The report of this inspection was sent to you on February)29,1984. (
Reference:
NRC Inspection Report Nos.
50-277/84-01; 50-278/84-01.
Three other violations, which were identified by members of your staff, were reviewed during an NRC inspection conducted on January 13 - February 29, 1984.
The report of this inspection was forwarded to you on March 19, 1984.
(Reference NRC Inspection Report Nos. 50-277/84-03; 50-278/84-03.) At the Enforcement Conference, the causes of these violations and your corrective actions were discussed.
The violations are described in the enclosures.
The first violation, which is described in the enclosed Order Modifying License Effective Immediately, involved a change to a plant operating procedure for plant shutdown and a change to the shutdown sequence described in the FSAR, without having performed l
an adequate evaluation to ensure that the changes did not violate technical l
specifications or result in an unreviewed safety question. As a result of the l
changes, rods were scramed individually, during shutdowns of the reactor from i
1977 to late 1983, effectively bypassing the safety functions of the Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS). These systems ensure adherence to approved control rod sequences and were required by the technical specifications to be operable at the time.
~ CERTIFIED MAIL RETURN RECEIPT REQUESTED c) 9 s
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Phila"delphia 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 Immediately 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 j
and extensive corrective actions taken for violation I.C.
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l 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.
l 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 determining 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 I
Document Room.
The responsgs directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget, otherwise required by the Paperwork Reduction Act of 1980, PL 96-511.
Sincerely, f
Richard C.
oung irector Office of pection and Enforcement
Enclosures:
1.
Order Modifying License Effective Imediately 2.
Notice of Violation and Proposed Imposition of Civil Penalties cc w/encls:
R. S. Fleischmann, Station Superintendent Troy B. Conner, Jr., Esquire Eugene J. Bradley, Esquire Assistant General Counsel Raymond L. Hovis, Esquire Michael J. Scibinico II Assistant Attorney General Public Document Room (PDR) (LPDR) local Public Document Room Nuclear Safety Information Center (NSIC)
NRC Resident Inspector Comonwealth of Pennsylvania
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, AE0D 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 1
.(, JT, IE D IE RI:DRA RI:RA IE:ES ELD E-ad -
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Murley PRFarron JLieberman 5/S4 84 5/ #/84 J/.f//84
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UNITED STATES NUCLEAR REGULATORY COMISSION In the Matter
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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 IMEDIATELY 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 practice of individually scraming control rods to effect a nomal reactor shutdown. The practice was i
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 detennine if the changes involved a modification to technical specifications or an unreviewed safety question._ Specifically, in 1977, plant operating procedure GP-3 used for nomal 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 n
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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 lic.ensee'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 detennine (1} whether adequate safety reviews have been and are currently being perfonned when plant and procedure changes are made; and (2) whether inconsistencies exist in other procedures with regard
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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(1), 161(o), and 182 of 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.
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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 performed in a nenner 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; i-(2) the methods of performing 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 rey not be extended without good cause and the concurrence of the Regional Administrator, Region I.
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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 recommendations resulting from the The licensee shall direct the appraisal team to report immediately, appraisal.
upon identification, to the licensee management and the NRC any inconsistencies In addition, the which could affect the safe operation of the facilities.
licensee shall consider the reconnendations resulting from the appraisal and provide to the Regional Administrator, Region I, an analysis of each such recommendation and the action to'be taken in response to the recommendation.
The licensee shall also provide a schedule for accomplishing these actions.
The Regional Administrator, Region I, may relax or terminate 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 I
be submitted to the Director, Office of Inspection and Enforcement, U.S.
20555 within 30 days of the Nuclear Regulatory Commission, Washington, D.C.
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, 19406. ANY REQUEST FOR A Region I, 631 Park Avenue, King of Prussia, PA HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORD
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