ML19340D897
| ML19340D897 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/24/1980 |
| From: | Ornstein H NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| Shared Package | |
| ML19340D693 | List: |
| References | |
| TASK-AE, TASK-C004, TASK-C4 AEOD-C004, AEOD-C4, NUDOCS 8101050536 | |
| Download: ML19340D897 (33) | |
Text
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O AE00 ACTIONS CONCERNING THE CRYSTAL RIVER 3 LOSS OF NONNUCLEAR INSTRUMENTATION AND INTEGRATED CONTROL SYSTEM POWER ON FEBRUARY 26, 1980 BY THE OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA NOV 2 4 : sac l
PREPARED BY:
H. L. ORNSTEIN NOTE: This report documents results of studies completed to date by the Office for Analysis and Evaluation of Operational Data with regard i
to a particular operating event. The findings and recommendations contained in this report are provided in support of other ongoing NRC activities concerning this event. Since the studies are on;oing, the report is not necessarily final, and the findings and recass?nd-ations do not represent the position or requirements of the restin-sible program office of the Nuclear Regulatory Commission.
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6 TABLE OF CONTENTS PAGE 1.
INTRODUCTION............................
1 2.
EVENT DE SCRIPTION.........................
A 3.
LICENSEE ACTIONS..........................
7 4.
NRR & I&E ACTIONS.........................
9 5.
AE00 ACTIONS SUBSEQUENT TO THE PLANT VISIT ON FEBRUARY 29, 1980..
10 6.
CONCLUSIONS............................
14 TABLES..................................
16 R EF E R EN C E S................................
18 APPENDIX A................................
19 APPENDIX B................................
20 APPENDIX C................................
24
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6 AE00 ACTIONS CONCERNING THE CRYSTAL RIVER 3 LOSS OF NONNUCLEAR INSTRUMENTATION AND INTEGRATED CONTROL SYSTEM POWER ON FEBRUARY 26, 1980 1.
INTRODUCTION On February 26, 1980, Crystal River 3 experienced a plant transient which was initiated by a loss of instrument power. During that event, power to the Non-Nuclear Instrumentation and Integrated Control System was lost, a small loss of coolant was created, and almost all of the cnntrol room instrumentation was lost for a period of about 20 minutes.
During the first a8 hours after the event, AE00 had received only scant informa-tion about the event (primarily from the Phillips Building PA system on the day of the incident, from a Commission briefing held the next day, and from brief discussions with NRR and IE personnel).
Since the information being received by AE00 was neither consistent nor complete, and the AE00 staff believed that the event was highly significant, it was decided l
that one of the AEOD's two oermanently assigned members for the new office should go to the plant to acquire accurate information about the event, help assure that no significant aspects of the event were overlooked, and later determine that adequate follow-up actions are taken to prevent recurrence of similar events.
It was agreed among the Directors of AE00, IE, and NRR that, above all, AE00's presence at the plant was primarily an information gathering mission which was not to perturb the plant personnel and the IE staff who were still in the proc-ess of assisting the licensee who was trying to bring the plant to a safe shut-down condition.
The author lef.t NRC headquarters on Thursday afternoon February 28, and was briefed near the plant by the Director of IE, Region II, at 7:30 a.m.
the next morning. While at the plant during the next few days, the author had numerous discussions with Florida Power Corporation (FPC) personnel, IE Region II staff, NRR staff, and industry representatives (B&W and INP0/
NSAC). The author returned to Washington on Sunday, March 2,1980, and briefed the AEOD Director the followina morning. Conclusions reached at that time were:
As suspected, the event was potentially serious and, as shown by a.
previous LERs, there had been numerous precursor events which did not receive the requisite corrective action to prelude the trarcient.
(At the time of this transient, FPC and the other B&W Licensees were in the process of responding to I&E Bulletin 79-27, which was issued after a similar event had occurred at Oconee 3 on November 10, 1979.)
b.
During the weekend of February 29, 1980, testing of the plant instru-mentation and control circuitry produced numerous surprises. Conse-quently, it was deemed imperative that Florida Power Corporation quickly determine that the root cause of the event. Furthermore, it was felt that the " proof testing" for system and channel dependencies and inter-l actions should be a minimum requirement for Crystal River 3 restart, 1
and that similar testing of the instrumentation and control systems at all other B&W plants should be required immediately.
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c.
Alternate safety grade instrumentation, independent of the NNI/ICS and single inverter failures, should be installed at CR-3 and all other 98W olants in order to provide information to operators in case of a repet-itive event.
(There had been numerous precursors, and since it appeared that total prevention could not be assured, provision to allow the plant to ride through such an event with alternate instrumentation seemed to be the way to 90.)
d.
Contrary to the belief of the onsite IE team, the author believed that the loss of the NNI-X power was caused directly by the newly installed T-Sat meter and tiie NNI-Y circuitry. This contention wcs based upon a discussion with the electrical technician who was at the NNI-Y cabinet trouble shooting the T-Sat meter at the time of the event (see Appendix A).
This opinion was conveyed to the IE response team leader (R. Wessman) on March 1,1980 with the additional thought that a generic problem exists, and that unless the root cause of the transient was found quickly, it might be necessary to have all B&W plants "show cause" to allow centinued operation.
There were questions about the meaning of the self powered neutron e.
detector readings which might be associated with localized boiling in i
the core.
f.
There were questions raised regarding the maintenance and lubrication associated with the CR-3 olant's decay heat removal system and design capability of DHR systems in general.
3
t 2.
EVENT DESCRIPTION On February 26, 1980, the Crystal River Unit 3 nuclear generatino plant located in Citrus County, Florida, experienced an incident involving an electrical malfunction in an instrumentation and control system. This resulted in a rer0 tor and turbine trip; the opening of the pressurizer power operated relief valve (PORV), the pressurizer spray valve, and a code safety valve; decreased feedwater flow to the steam generators; actuation of the engineered safety feactures (ESF) systems; and discharge of approximately 43,000 gallons of primary coolant into the containment building. The pressur-ized water reactor was designed by the Babcock & Wilcox Company (B&W).
A portion of the non-nuclear instrumentation (NNI) +24 volt power supply was lost due to a short to ground. This loss affected automatic plant control systems and about 70% of NHI control board indicators (such as reactor coolant system temperature, pressure, and flow; steam generator pressure and level;
~
and pressurizer level).
It caused the pressurizer PORY and the pressurizer spray valve to open. The failure also caused false control signals to be l
sent to the Integrated Control System (ICS), the most significant of which caused a reduction in feedwater flow to the steam generators. Al so, the false T signal caused the ICS to withdraw the control rods to increase ave power.
The reduction 'in feedwater flow reduced the reactor heat removal rate to below the reactor heat generation rate, which caused reactor coolant system temper-ature and pressure to increase in spite of the open PORY and spray valve. As 4
a result, the reactor tripped on high pressure and then was subsequently partially depressurized. The operators secured the reactor coolant pumps as reoaired by emergency procedures. High Pressure Injection (HPI) auto-matically initiated as a result of Reactor Coolant System (RCS) depres-surization due to loss of coolant inventory through the open PORV and the coolino effects associated with the reactor trip. Shortly after receipt of a high reactor coolant drain tank level alarm, the PORV block valve was closed and, with approximately 70% of NNI inoperable or inaccurate, the operator correctly decided that there was insufficient information available to justify terminating HPI. Therefore, the RCS and pressurizer were filled solid, causing RCS pressure to increase to the point where one safety valve lifted, and flow through the safety valve spilled water into the containment through the Reactor Coolant Drain Tank rupture disk.
Power was restored to the HNI about 20 minutes after the start of the transient. Plant conditions that then existed included the pressurizer filled solid with water, reactor coolant pressure was at 2400 psig, a reactor coolant outlet temperature of 556*F, steam generator "A" was dry, and the core was being cooled by water flow from the high pressure injection system out the open safety valve and by natural circulation through steam generator "B".
Three minutes after the transient began, an operator specifically assigned to purge containment to the atmosohere terminated the purge. Then, because water was still being discharged from the RCS into the reactor building, the reactor building pressure increased to the point (4 psia) where automatic building isolation occurred.
5
After the restoration of power to the instrumentation, the operators throttled high pressure injection to reduce the flow of water through the open safety valve and into the reactor bailding. The operators also re-established the water level in steam generator "A".
About 1-1/2 hours after the beginning of the transient, the operators estab-lished RCS oressure control using nomal makeup and letdown flows. Because RCS temperature and pressure were well under control and the core was being adequately cooled by natural circulation, high pressure coolant injection was shut off. The decision was then made to heat up the pressurizer in preparation for establishing a steam bubble in the pressurizer. About 3-3/4 hours after the incident began, the pressurizer was heated sufficiently and a steam bubble was established in the pressurizer.
About 6-3/4 hours after the beqinning of the incident, forced flow in the RCS was established by starting two reactor coolant pumps.
During the incident, about 43,000 gallons of RCS water was spilled into the containment. The radiation level within containment reached 50 rems per hour early in the event and decayed rapidly. This was the result of short-lived gaseous activity (i.e., xenon-138, krypton-89) which were released from the reactor coolant. Radioactivity released to the environment was within reaula-l tory limits, and the spilled RCS water was reprocessed for in-plant use. No core damage occurred. There was no impact on the general public or plant employees as a result of this incident.
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3.
LICENSEE ACTIONS Immedistely after the transient, in addition to taking action to bring the plant to a.iafe-shutdown condition, the licensee took the initiative to identify the cause of the transient, assess equipment deficiencies and perform an analysis of the transient to determine short and long tarm modifications necessary to assure safe plant operation.
At a meeting in Bethesda on March 4,1980 the licensee informed the NRC that they had found the short that had initiated the transient, provided their secuence of events, and presented the results of their testing and inspection of the NNI-ICS circuitry (including descriptions of equipment deficiencies and a summary of operable instruments following the NNI power failure of 2/26/80). During the meeting Florida Power Corporation also presented their Corrective Action Plan (see Table 11. On March 12 aad March 17, 1980 Florida Power Corporation provided the information which was requested by NRR at the March 4,1980 meetina - namely:
A summary of previous power upset event on NNI-ICS at CR-3.
a.
b.
Copies of information presented at the March 4,1980 meeting.
g c.
A discussion of loss of NNI-ICS power:
l (1) How the operator determines which information is reliable.
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t (2) What information is needed by the operators to bring the plant to cold shutdown (see Table II).
d.
Confirmation that prior to plant restart, the licensee would verify by functional testino that information on the major plant parameters required to bring the plant to cold shutdown are available to the operator upon loss of ICS or either train of NNI.
e.
An updated list of corrective actions that would be taken prior to restart (Items 1 through 12 appearing in Table I plus eleven additional items).
f.
A review of the benefits of the NRC requirements imposed since the TMI-2 accident (Short Term Lessons Learned and Bulletins and Orders) - The assessment concluded that the " requirements were of benefit to plant operations," and that the requirements for improved / additional operator training provided the largest contri-bution for dealing with the 2/26/80 transient.
Subsequent to review of the licensee's short term corrective actions and longer term commitments, the NRC staff provided a Safety Evaluation allowing j
restart of the CR-3 plant. The plant completed its refueling outage on i
August 10, 1980, and subsequently returned to power production.
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4 NRR AND 18E ACTIONS I&E regional tesms arrived at the site a few hours after the transient started. Their mission was to assess plant conditions and evaluate the significance of the event. An NRR Operator Licensing Branch member was in the CR-3 control room giving a licensing examination while the event occurred, and similarly an IAE Performance Appraisal Team was onsite while the transient occurred. NRR sent a senior instrumentation and controls engineer to the site on March 1, 1980.
NRR requested that R4W and s11 the B4W licensees attend a March 4, 1480 meeting to provide information on their histories of non-nuclear instru-mentation problems, the susceotibility of their plant to the CR-3 event and its imolication on the safe operation of their plant.
Via a March 6, 1980 letter, NRR requested each B&W licensee to provide written cnpies of this information and also address the feasibility of performing a test to verify that reliable information is available to the operators subsequent to NNI/ICS oower upsets as well as addressino the aoolicability of each of CR-3's proposed corrective actions tn their plant.
In addition to reviewing the aforementioned licensee submittals, a special task force (B&W Reactor Traisient Response Task Force) was established by NRR on March 12, 1980, to assess the generic aspects of operating experiences of the R&W plants. The task force consisted of 12 members from NRR, IAE, RES 1
and ORNL. The task force assessed the sensitivity of B&W plants to transients 9
l
involving overcooling and undercooling conditions, small break loss of coolant accidents, consequences of malfunctions and failures of the NNI and ICS. The task force published their report (NUREG 0667) in May 1080.
The report contained 22 recommendations. It also included the Probabilistic Analysis Staff's assessment of the risk reduction potential associated with each recommendation.
In meetings held with the B&W licensees, NRR requested that each licensee take actions to assure plant operators have the capability of coping witn various combinations of loss of instrumentation and control functions (these tasks included immediate design review, analysis, and testing at each facility; and immediate correction of electrical and control deficiencies, as well as the upgrading of procedures and operator training). IE representatives witnessed the NNI/ICS testing which was performed at each plant, and also reviewed the revised plant procedures.
On July 18, 1980 NRR issued a Safety Evaluation Report for restarting CR-3.
It noted that the corrective actions taken by the licensee "are sufficient to permit restart of CR-3 and to ensure operation in a safe manner during the interim period until the long term requirements resulting from the Three Mile Island Unit 2 accident and those of OIE Bulletin 79-27 are all implemented."
5.
AE00 ACTIONS SUBSEQUENT TO THE PLANT VISIT ON FERRUARY 29, 1980
[
It was clear from the actions taking place at the plant during the weekend of February 29,1980 (including the scheduling of a meeting in Bethesda with B&W and the B&W licensees on March 4,1980) that Florida Pnwer Corporation, NRR, 10
and IE were taking appropriate follow-up action on a timely basis. The Office for Analysis and Evaluation of Operational Data participated in the Ma. h 4,1980 meeting and in subsequent NRC meetings held with B&W, FPC, and the other B&W licensees. At the March 4, 1980 meeting, each B&W licensee addressed NNI problems at his facility, their plant's susceptibility to an event like the one that Crystal River had just experienced, and their proposed corrective actions.
At some of those meetings, AE00 raised questions about the susceptibility of once through steam generators to overfill due to the lack of safety grade instrumentation and high leve' trips. In an August 4, 1980 memo, AE00 recommended that this item be considered a "New Unresolved Safety Issue."
In an April 24, 1980 meeting in which Florida Power Corporation outlined their proposed actions (near-term and long-term plant modifications, improvements in operator training, etc.), AE00 expanded on the fact that many BAW plants are susceptible to overcooling transients because their atmospheric dump valves open to the half open position upon loss of NNI/ICS power. AE00 has brought this fact to NRR's attention and NRR in turn in-dicated that this problem will be resolved; however, the schedule for resolution has not been established (see References 1 and 2).
While at the Crystal River 3 plant, AE00 probed the circumstances behind the fact that there was a three day gap between the NNI-ICS power loss, and the cooldown of the plant to the level where the RHR system could be used. The problem stemmed from inadequate lubrication of a decay heat 11
removal pump (DCP l-A - one of two redundant pumps). In a March 20, 1980 memo, AE00 documented this information and ccpies were forwarded to NRR end IE along with some preliminary recommendations (see Appendix B). Subseque ntly, AE00 also searched the LER and NPRDS files to determine how widespread the i
problem of inadequate pump bearing lubrication might be.
Initial results indicated that the problem was not widespread and that the bearing lubrication problem at Crystal River 3 may have just been an isolated incident. Pump and motor bearing li,brication will be watched by AE00.
In the course of examin1' the decay heat removal system for Crystal River 3, AE00 noted that the Nucle.ir Services Closed Cycle Cooling System (NSCCCS) was subject to a single passive failure which could have :;ignificant implica-tions. This information was conveyed to NRR by a memorandum (see Reference 3).
AE00 is concerned that short-term improvements in decay heat removal systems are not a high priority item in the TMI Task Action Plan and, therefore, is not receiving the attention which is required. AEOD has bequn a case study of the recent losses of decay heat removal capability at the Davis-Resse Plant
}
during the summer 1980 refueling outage.
AE00 has reviewed the May 1980 draft of the Crystal River 3 IREP report. AE00's review (see Reference 4) pointed out serious omissions in the study, including the fact that the operational experience associated with the loss of NNI-ICS power and inverter failures at CR-3 and other B&W plants were not considered in the IREP report (see Reference 4, Page 15). In its review, AE00 also pointed i
out that in addition to the dependencies noted in the IREP report, the integrity of 12 l
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all of the reactor coolant pump seals, pump bearings, and motor bearings are dependent upon the NSCCCS. AE00's review also noted a disconnect between the IREP study and operating experience in the area of procedural or operator error (see Reference 4, Pages 7,14, and 15) and the incemolete assessment of OTSG overfill and OTSG tube rupture problems.
Subsequent to examinino self powered neutron detector data from the CR-3 events recorder, AE00 questioned the possibility of localized core boilino during the event. Florida Power Corporation reviewed this aspect nf the transient and concluded from comparison with earlier trip data that local core boiling did not take place during the event (see Appendix C).
Presently, AE00 is carefully monitoring the actions being taken by Florida Power and the other B&W licensees in the areas of NN!/ICS power, steam generator overfili, emergency feedwater control, etc., and as additional concerns appear, AE00 will bring them to the attention of the appropriate NRC office.
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6.
CONCLUSIONS The immediate and long-term corrective actions proposed by Florida a.
Power Corporation, as well as those required by NRC of a11 B&W plants regarding the major aspects of the February 26, 1980 event, apoear to be appropriate. (These include innediate confirmatory testing of NNI/
ICS instrumentation and circuitry, provision of alternate sources of information required by the operators to bring the plant to a safe shutdown subsequent to another similar power loss, and improved or additional operator training).
b.
Some items which NRR or IE did not focus attention on immediately after the transient have ben raised by AE00, and NRR or IE are now addressing them, e.g.,
o Steam generator overfill o Opening of atmospheric dump valves upon loss of ICS power o Improved decay heat removal reliability I
It is ton early to determine how these issues will be resnived; however, l
AE00 is following efforts on these subjects closely.
AE00 has reviewed the oilot IREP study of the Crystal River 3 Plant.
c.
AE00 noted t'it in addition to other deficiencies, the report, i
omitted significant operating data (especially that acquired from the February 26, 1980 event). AEOD has conveyed this information to RES/Probabilistic Assessment Staff, who in turn has requested 14
the contractor who performed the study to provide schedule and cost information associated with correcting the deficiencies noted by AE00 and other reviewers.
4.
Even though the February 26, 1980 event at Crystal River 3 resulted in a long f and costly) power outane, it resulted in enhancina reactor safety, with minimal imoact upon the public, the environment and industry. For example, the sub.iect transient succeeded in focusing Commission and industry attention on a previously known problem, and sped uo the schedule for its resolution. (Similar events had occurred at other 84W plants, but none had resulted in spillage of primary coolant into containment.) The transient undoubtedly resulted in speeding up the pace of implementing plant improvements that appeared to be necessary at B&W plants subsequent to previous NNI/ICS power failures.
Table I FLORIDA POWER CORPORATION'S CORRECTIVE ACTION PLAN (MARCH 4,1980)
Short Term 1.
Thorough testing of the NNI system to determine cause of failure.
2.
Modify PORY so that NNI failure closes valve.
3.
Modify pressurizer spray' valve so that valve doesn't open on NNI failure.
4.
Provide positive indication of position of all three relief or safety valves.
5.
Establish procedural control of NNI selector switches.
6.
Train all operators in response to NNI failures.
7.
Move 120V ICS "X" power to vital bus.
8.
Initiate more extensive program for events recorder system.
Q.
Provide operator with redundant indication of main plant parameters.
10 Install indication lights on all panels to know if power is on.
11.
uict access fuses to be designed into cabinets.
- 12. Modify EFW pump circuit to start pumps on any low steam generator level signal.
Long Term 1.
Investigate upgrade of NNI capabilities - total loss of NNI.
2.
Remote shutdown panel to be designed.
3.
Provide backup AC sources to inverters with automatic transfer.
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t Table II FLORIDA POWER CORPORATION'S LISTING OF MAJOR PLANT PARAMETERS NEEDED TO BRING PLANT TO COLD SHUTDOWN (March 12,1980)
Pressurizer Level Pressurizer Temperature Reactor Coolant Loop A Wide Range Cold Leg Temperature Reactor Coolant Loop B Wide Range Cold leg Temperature Reactor Coolant Loop A Wide Range Hot Leg Temperature Reactor Coolant Loop B Wide Range Hot leg Temocrature Emergency Feedwater Flow A and B Reactor Coolant System Wide Range Pressure OTSG A Startup Level OTSG B Startup Level OTSG A Operating Level OTSG B Operating Level OTSG A Pressure OTSG B Pressure Decay Heat Removal Flow (Both Trains)
Core Flood Tank A Level Core Flood Tank B Level Core Flood Tank A Pressure Core Flood Tank B Pressure High Pressure Injection Flow (Both Trains)
Makeup Tank Level BWST A Level BWST B Level Tsat Meters Letdown Flow Decay Heat Suction Temperature Low Range RC Pressure Core Exit T/C Indication Makeup flow a
I 17
REFERENCES 1.
Memo:
C. Michelson to H. Denton, " Lessons Learned from the Crystal River Transient of February 26, 1980 - Correcting Atmospheric Dump Valve Opening Upon Loss of ICS Power," (May 23, 1980).
2.
Memo:
H. Denton to C. Michelson, " Lessons Learned from the Crystal River Transient of February 26, 1980 - Correcting Atmospheric Dump Valve Opening Upon loss of Integrated Control System Power," (July 21, 1980).
3.
Memo:
C. Miche540n to R. Mattson, "NRC Action Plans Developed as a Result of TMI-2 Accident - Draf t 3, Task II E.3 Decay Heat Removal," ( April 24, 1980) 4.
Memo:
H. Ornstein to R. Bernero " Review of IREP Crystal River-3 Safety Study," (June 19,1980).
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APPENDIX B one%
UNITED STATES f
NUCLEAR REGULATORY COMMISSION y
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s W ASHINGTON. O. C. 20$55 oh' t y5 V GWf MAR 2 d 1580 MEMORA::DU:: FOR:
Carl Michelson, Director Office for Analysis and Evaluation of Operational Data FROM:
Harold L. Ornstein Office for Analysis and Evaluation of Operational Data
SUBJECT:
CRYST AL RIVER NUCLEAR POWER PLANT DECAY HEAT CLOSED CYCLE COOLING WATER PUMPS /DCP-1A At:0 DCP-1B I have pulled together the folicwing information:
The NRC re:uires surveillance testing of these pumps every 30 days. Pump OCP-1A tested satisfactorily on 2/6/80. Above and beyond the inservice inspection program recuired ::y :;RC, Florida Power Corporation has an outside consultant (C&S Maintenance Consul uts) conduct a plant-wide overall vibration maintenance program.
On 2/12/2C, one week aftcr DCP-lA passed the NRC required surveillance test, CLS found the pu p motor vibration to have a 1.4 mil displacement (1.0 mil alcrt is requirec, 1.~ mil action is recu1 rod). A C&S report, dated 2/1G/00, noted the increased displacement and indicated that it was not at a point of concern, and advised FF" that the pump should be watched for increases in vibration and/or temperature.
On 2/25/3, at 5:00Ati, Cf.S retested the motor and found it to be out of spec (4.7 mil) and ir. need of ir=ediate repairs. The vibration was attributed to the motor bearing closest tc the coupling. A work request was written uo noting the fact that this is a tech s:s: :riority item (72-hour action status). The fact that DCP-1A was inoperable was nct relayed to the control room perscnnel until af ter the incident, partially due te syste-ine tia and the plant evacuation.
(A meeting between the shift super-r visor and :ne naintenance planners / coordinators to go over the pump repair did not l
take ;1a:s u-til after reactor trip. The question of communication between operations and mair.ts.a.ce personnel regarding declaring vital etuipment out of service needs to te cc:rissed - is this a problem at many other plants vs. just Crystal River vs.
just r.n is;;ticd event at Crystal River?)
w M e J d l b@
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1 Carl Michelson The motor bearing clcsest to the coupling was changed on the morning of 2/27/80; however, excessive vibration was still reccrded (displacement of 1.1 mil). Because of this vibration, the motor bearir.g on the opposite side of the motor was then changed. Changing the seced bearing brought the displacement down to 0.68 mil which was within the alert range, but still outside the normal operating range of about 0.5 mil. Subsequently, FPC broke down the coupling, found that one end of the ccupling was unlubricated, and that it "had bad taeth which were scmewhat cerroded"; mest crobably due to imprecer lubricatien. The coupling was then replaced and the pump / motor were realigned. C&S ran vibratier tests on the pump /
motor en 3/28/80 and readings were within normal ranges (displacement was 0.3 mil).
The redundant decay heat closed cycle cooling system pump, DCP-lB, was examined, and the pump and motor were found to be cut of alignment. There was grease in the couplirg; hcwever, the grease showed signs of degradation (loss of " slickness").
It was concluced that the degracaticn arose from misalignment which caused the heating of the grease and excessive wear.
The coucling coes not have any grease cup or simple provisions for lubrication.
In ceder to crease the coupling, it must be disassembled. FPC (G. Claar - shop super-visor) informed me that FPC was of the impression that the DCP couplings were not permanently lubricated. FPC had fcund that the clant eouipment lubrication list (which was put together with the assistance of Gulf Atomic) had omitted the couplings en CCP-1 A and 18. As a result, FPC was planning to check the couplings during the next refueling cutage for lubrication.
I&E (Ashenden) was informed tnat, as a result of their recent experience with DCP-1 A anc 18, FPC is now planning to imolement a plant-wide program for alignment and coupling lubrication.
The lubricatien problem on the decay heat closec cycle cooling system pumps, the unexpec;ed f ailure, and the resultant celay in going to cold shutdown highlight the vulnerability of nuclear pcwer plants to inadecuate lubrication. The NRC does not ap;: ear o have adecuate visibility of luorication anc maintenance on vital pumps and other critical ecuitment.
Inadequate lubrication and alignment checks and the "impencing cc=cn mode failures" of the decay heat closed cycle cooling water pumps, 3C?-1A an: 1B, at the time of the 2/26/80 Crystal River incident, are viewed as actencial accident precursors which should be adcressed it eoiately.
,4 9 Gi & S Harold L. Ornstein Office for Analysis and Evaluation of Loerational Data Encictures:
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..a Mr. Hal Ornstein Office of Analysis and Evaluation of Operating (AEOD)
United States Nuclear Regulatory Cottenission Washington, D. C.
20555
Subject:
Incore Self-Powered Neutron Detector Information Hal:
Sorry for the delay on getting this information to you. I am afraid the strip chart reproductions are pretty poor but this is the best we can do for now.
For any additional infor: nation on this subject please contact myself or Bill Cross (1-904--795-6486, ext. 10 1).
Sincerely, hI e,
W. E. Kcmper Plant Training Mnnnuer WEK:ub
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I (Page 2 of 2)
MTA SHEET 1 (Cont'd)
, v-CCRE NEUTRON DETECTOR SYSTEM CALIBRATION EQUIPMENT REOUIRED__
DVM No.
REcor.1)Ek #2 mrm'itx trND
!.tvEL Tutt.
80 K0 8;U IA'H 168:
D08 F
t.t.Qt:s t.ru ACTU4L A UJtisT1;D lu.Qtl11323 ACTUAL Atut':""t p ros t *.fies
.s tul'.'G Iro.
No.
45ST.
76 25 4
C-11 31 2
6 C-11 l'
12 19
(,
2-11 13 29 4
K-11 14 16 6
?I-9 15 9
4 c-05 16 9
l 2
c-05 g
4 17 11 2
K-OS f
.t
~
1t 19 2
K-11 I
k.12.1 The symmetrical in-core detector system is operable (computer calculation a.
using at least 75% of the detectors in each core quadrant) and shall bs I
used as the quadrant power tilt monitor.
Yes No b.
The minimum in-core detector system for quadrant power tilt is operable.
p' i
(See Section 2.0.)
Notify the Shift Supervisor that the computer is inoperative as quadrant power tilt tonitor and refer to STS 4.2.4 i
Yes No
- 6. L2. 2 a.
The in-core detector system is operable (computer calculation) and shall be used as axial power imbalance monitor.
L t
Yes No b.
The minimum in-core detector system for axial power imbalance is operable.
i (See Section 2.0.)
Notify the Shift Supervisor that the computer is l
inoperative as axial power imbalance monitor and ref er to STS 4.2.1.
l Yes No I
i I
I t.
t I
'D
- P' }D '"D'S' j ' rurn e ay Dare
- J l.
'd d Ju o
n Page 7 SP-433 Date 2/11/77
}
31 Rev. I 1