ML20024B588
| ML20024B588 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/05/1979 |
| From: | Arnold R, Nodold R GENERAL PUBLIC UTILITIES CORP., METROPOLITAN EDISON CO. |
| To: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| TASK-01, TASK-02, TASK-03, TASK-04, TASK-05, TASK-06, TASK-07, TASK-08, TASK-09, TASK-1, TASK-10, TASK-2, TASK-3, TASK-4, TASK-5, TASK-6, TASK-7, TASK-8, TASK-9, TASK-GB B&W-0707, B&W-707, NUDOCS 8307090353 | |
| Download: ML20024B588 (97) | |
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Middletown. Pennsvivenia 17057 717 ^ ^
7-- 948-8000 December 5, 1979
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Mr. Victor Stallo, Jr.
Director Office of Inspection and Enforcement o
U. S. Nucisar Ragulatory (*n==4 esion 4 a:;
Washington, DC 20555 5 bo
Dear Mr. Stallo:
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Subject:
Dockat No. 50-820 C.c = S O
'l Rasponse to Notice of Violation and
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Notice of Proposed Issuance of E~
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.- ta Tour letter of October 25, 1979 transmitted a Notics of Violation ami a Notica of Proposed Issuance of Civil Psnalties based upon the Office of s
Inspection and Enforcement's investigation of the March 28, 1979, accident at Three Mile Island Unit 2.
Your lactar also addressed some general remarks concerning Metropelitan Edison and its management controls for the operation of the Three Mila Island facilities.
We have carefully considered the information and ecsclusions set forth in your letter and in tha'Noticas enclosed with it.
Ihis consideratier. has been aided by many studies, analyses and reviews which we and others bava l
undertaken since the March 28, 1979 accident. We have sought to forthrightly address each of the charges while recou" Mag that many of the issues turn upon i
incarpretations of complex procedures. Our detailed responses are set forth in two enclosures to this letter: Macropolitan Edison Company's Statement in Reply to Notica of Violation, and Metropolitan Edison Company's Answer to Notics of Proposed Imposition of Civil Penalties. These responses are based upon our present undersa w ag of the accident. Cartainly, our :matual understanding of the accident and its underlying esuses can be expected to improva as st.uitas continua.
Although the specific violations asserted in the Notice generally addressed the Unit 2 operating organization, the cric1cism of the October 25 letter is aimed at Metropolitan Edison's management co:itrols. The problems with m age-ment controls tears not obvious from events prior to March 23, 1979.
~he natura and extent of noncompliances identified during NRC inspections did not imifraca f"M-tal problems with the safe operation of the plant. During the period from 1975 to 1978, operators at Three Mile Island had a failure case on chair l
NRC written and oral exams. half the industry average. NRC performanca evalua-l tions ranked the Three Mile Island facility above the average for comparable plants. Metropolitan Edison does not faal that chara was any significant decline in the Ccapany's performance. What is clear to us is that changes in the approach to management and management controls aust be made by Macropolitan Edison as well as the total nucisar complex, to address the daficiencias which the severs tasting of the accident revealed.
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Metfoocastari Ecmon Comoeny is a Memeer or me Generas P.:tre Untmas System 03070903S3 791205 k
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- hter Stallo, Jr. December 5, 1979 To identi.fy -the nature of the needed changes, we have undertaken exten-sive internal revieve of the Company's corporate structure and staffing.
Major changes have been made or are in progress in our approach to corporate management and controls.
While many, if not all, of the changes which have occurred or are in progress have already been submitted to the NRC for re-view, it is appropriata to mention them again here. The DfI Generation Group was formed to integrata the nuclear management and tae h4-=1 support capabili-time of Macropolitan Edison and the GPU Service Corporation in a single entity.
As a result, the professional technical staff for Three Mile Island has been tripled to a current level in excess of 200 with over 2,600 man-years of nucinar experience.
A primary objective of the group is to ensure safe operations by means which include strict adbarence to NRC regulations, Technical Specifica-tions, and plant procedure. Unit 1 and Unit 2 lina management responsibilities have been separated in recognition of the different status c,f the units and each unit given direct control, to the anziman possible extent, over the resources necessary for the effective and safe conduct of plant activities. A shift techMn1 advisor has been added to the normal shift complement and sub-stantial additional attention will be directed to the operating experience of similar reactors and the nuclear industry as a whole. Improvements in the g
organizational status and staffing of the health physics departments have been s'
achieved (although va recognize that the unique circumstancas of Unit 2 will require further significant improvements). Upgrading of operational quality assurance, specific procedural requirements for the independent verification of operational activities affecting safety, and changes to the review and approval provisions for plant and emergency procedures have all been undertaken to improve management controls.
Operating and emergency procedures are under review and revision as appropriata.
A major revision and expansion in the training programs for the operating organisations has been made and a annagement ersiains program is under development.
We are taking steps to transmit the management commitments through all levels of the Company to assura that all personnel have a high degree of svareness of our comattaant to safe operation. These items have been described in detail in our submittals to the Staff in connection with the Unit I restart.
These changes and our conmiitaant to continued improvement underscore our commitment to correct the inadequacias which have now become clear. N hard lessons taught us by the events surrounding the accident have been comprehended.
h need to significantly upgrade our nuclear program has been recognized.
N Metropolitan Edison, we believe, is suin ding the issues raised by the acci-donc and its aftaraath and is takfag the steps needed to resolve those issues.
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h Company has also conducted extensive reviews of the accident and related issues to ensure we have as completa an understanding as possible of all factors which contributed to the accident. It is our view that the acefdent cannot be traced solely to inappropriata operation action. Rather, it must be ascribed to a auch more complex set of events. We find support of this view in the findings by the Advisory Committaa on Beactor Safeguards, the President's Commission, and recent Commission statements.
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h hay to unders*=~HM the accident, in our opinion, lies not so mach in the procedural violations which are charged, but with more basic causes. Fe
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believe that there are two such causes. First, there was a lack of understanding, and thus, an absence of clear proceduras, to deal with a small break loss of coolant accident from the steen space of the pressurisar.
No guidance was given the operators for a LOCA from the pressuriser, in which pressura decreases while 4
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- Victor Stallo,.7r. December 5, 1979 i
l pressuriser level increases. In fact, training and procedures prepared with-I out knowledge of this detail of system behavior probably inhibited proper operator action. The second basic cause was an exaggerated concern with "taking the plant solid.
Tae " =1 Specifications and procedures prohibited this condition and the reactor simulator on which operators were trained could not staulate solid conditions. While operator training and plant procedures were deficient in a number of areas, these deficiencias arose from the under-lying lack of awareness by Metropolitan Edison, the nuclear industry, and the Commission that thakind of system behavior which the plant asperienced could occur.
h investigation conducted by your Office in the aftermath of the acci-dont was the largest ever undertaken by the Consission. Even though it did not purport to be a full ane. lysis of the causes of the accident, the investi-gation was unprecedented in terms of scope, intensity, duration, and manpower.
It produced a voluminous and detailed report, NURIG 0600. h investigation d4*atamed some violations of plant procedures. He have recognized chase and are taking or have already taken steps to assure that there will be no repeti-tion.
h investigation disclosed areas in which plant procedures were ambigu-ous or incomplete. We are and1fying or have already rewritten those proce-dures to incorporate the lessons learned from the accident and are undertaking a comprehensive review of all plant procedures. h investigation disclosed aspects of the accident where conditions went beyond the bounds of plant pro-cedures and indeed beyond the bounds of previous industry assumptions of sys-tee behavior in accident conditions. And -finally, the investigation also pro-duced some charges which, based upon our further analysis, we believe are not adequataly supported.
h accident was the worst in the history of the causereial nuclear power industry.
Metropolitan Edison, the nuclear industry, d the NEC have been unalterably changed as a reruit. We recognize the f'=d - tal need for these changes and are committed to the implementation of those that apply to us.
We will, of course, keep you informed of further developments, and are coinsitted to working with the h4== ion to assure the operational safety of our nuclear program.
j Very truly yours, a
L C.
Senior Vice President ICA:cib Attachments 1
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e UNITED STATES OF AMERICA IWCLEAR ILMTORY COMMISSION 1
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METROPOLITAN EDISON COMPANT
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Docket No. 50-320 i
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(Three Mile Island Nuclear
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Power Station, Unit 2)
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METEOFOLITAN EDISON COMPANT'S STATEMENT IN REFLY TO NOTICE OF VIOI.ATION In accordance with 10CFR 2.201 and the Notice of Violation of october 25, 1979, Metropolitan Edison thpany provides the following responses to the apparent items of noncompliance identified in the Notice.
1.
Statement of Annarent Noncountiance:
Technical Specification 3/4.7.1, " Turbine Cycle," requires in Section 3.7.1.2, that three independent steam generator energency feedwater pumps and associated flow paths shall be operable during power operations, except: if one emergency feedwater system is inoperable it must be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant must be in Bot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Contrary to the above, for an undetermined period just prior to the reactor trip at approximately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on March 28, 1979, the flow,
paths to both staan generators were made inoperable by feedwater header isolation valve closure. (In addition, on January 3, February 26 and
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March 26, 1979, the flow paths from all three emergency feedvater pumps were simultaneously made inoperable by feedwater header isolation valve closure during the performance of, and in accordance with. an improper surveillance test procedure.)
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Discussion:
Metropolitan Edison.;grees that continued plant operation with the emergency feedwater header isolation valves (EF-712A and 123) in the closed position is an apparent breakdown in controls over the operability of safety related equipment as stated in.WEEG 0600 Section I 2.3.2.
Teile Metropolitan Edison does not believe that controlled isolation of the feedwater header for rou-time testing is in violation of the Technical Specification, we agree that it is undesirable and stape will be taken to undify surveillance test procedures for the emergency feedwater system, and to prMde routine (including some as frequently da each shift) status checks on components important to the safe operation of the plant. Der analysis supports the conclusion of the Eemeny Commission as stated in the Report of the President's Comeission on The Acci-
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dent at Three Mile Island, "The loss of emergency feedwater for 8 minutes had no significant effect on the outcome of the accident. But it did add to the confusion that distracted the operators as they sought to understand the cause of their primary problem".
The emergency feedwater system, like ueny engineered safeguard features, is 4
required to undergo periodic surveillance and testing which in some cases reduce the ability of the system to perform its intended function while in the test condition. Technical Specification 3/4.7.1 recognizes this condition and specifically allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of inoperability of the emergency feedwater system before additional corrective action must be taken. Although the Technical Specification is written in terms of the inoperability of "one emergency feedwater systas", (implying the exis-tence of more than one " system") there is only one emergency feedwater System for TMI-2.
The Safety Evaluation Report for TMI-2 (NUREG-0107, Sept. 1976, pg. 7-5) states, "The emergency feedwater system consists of one turbine drive pup, two motor-driven piasps and associated piping P,
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and valvcs." Th3 Technical Specification therefore allows full isolation of all or part of the emergency feedwater system for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before proceeding to hot shutdown.
, On this basis, the Surveillance Procedure 2303-M14A/3/C/D/E, "E:nergency Feed System valve Lineup Verification and Operability Test and Turbine Driven E.F. Pump Operability Test", Revision 8, is in compliance with Technical Specification 3/4.7.1.
It would appear that this judgement was confirmed by NRC inspection. " Combined Inspectica Report 50-289/
78-23 and 50-320/78-36", dated January 9, 1979, stated "The observations and records review were conducted to verify that startup, power and/or shutdown reactor operation were in conformance with Technical Specifica-tion safety limits, limiting safety system settings, and limiting con-dicions for operation". Among the procedures inspected with accept:.ble results was SP 2303-MlaA/,5/C/D/E Revision 8, and acceptance criteria included requirements from Technical Specification 3.7.1.2.
his surveillance procedure was followed in January, February, and March 1979.
In January and February, as far as we can determine the emergency system was returned to full operation Ismeediately upon completion of the testing. The elapsed time from the March 26 test to the time of dis-(
covery of the closed valve was 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> (and, therefore, also within the Technical Specification limit), if we assume that the valve was closed l
l and not reopened after completion of the test.
The status of these valves following the March 26 test was reviewed with ther people performing the surveillance on that date. Bree people have stated that the valves were reopened following completion of tha surveil-lance procedure. How and when the valves actually became closed following the performance of the surveillance has not been decernined despite ex-censive investigation by many parties.
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, The President's Commissign was also unable to determine the reason for et.ese valves being closed. Their report states:
"A Commission investigation has not identified a specific reason as to why the valves were closed at 8 minutes into the ac cide nt. The most likely explanations are: the valves were never reopened after the March 26 cest; or the valves were re-opened and the control room operators mistakenly closed the valves during the very first part of the accident; or the valves were closed mistakanly from control points outside the control room after the test."
We agree that the inability to provide any other testimony or documenta-tion to support that the valves were opened after the surveillance testing was complaced on March 26 indicates a lack of management controls, and we have revised our procedures and training to correct this deficiency.
To assess the effects of. EF-V-12A/3 being closed on the outcome of the
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accident, a comparative analysis has been made of the TMI-2 accident with and without delay in the initiation of emergency feedwater (ETW). This analysis was performed with the RETRAN Systems Analysis code. After a benchmark against plant data was made.for the first eight minutes of the accident (the EFW delay time), the same case was reanalyzed without the delay. The results are provided in the attached figures. A review of Figures 1 and 2 shows that the depressurization rate is less severe with-out EFW than it is with normal EFW. This is actributed to the depressuri-zation resulting from the additional cooldcvn.
In borh cases, however, t
che primary system would saturate as can be seen from Figure 3 which shows i
that the hot leg saturation occurs at almost six minutes with normal EFW flow. This is approximately the same time that the hot leg saturated daring the accident.
The pressurizer level for each caas is shown in Figures 4 and 5.
Al-though the pressurizer filled in aaout six minutes during the accident as N....
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, a result of high pressure injection flow. Figure 5 shove that the pres-suriser would also fill in about nine minutes for the case with normal EFU as a result of void swell into the pressurizer. High pressure injection included in this case, so the pressurizer would in fact be filled was not In both cases, pressurizer level is rapidly increasing while even sooner.
pressure is not responding similarly.
In both cases, the primary system is voiding and losing mass through the stuck open P017 at the eight minute point.
I ne conclusions of the ab^ove discussions are that system behavior, includ-ing pressuriser level, with, or without an eight minute delay in EFW, is very similar and that subsequena operator actions keyed to pressurizer r
level would be essentially the same whether or not the emergency feedwater had been lose for 8 minutes.
corrective Action:
Metropolitan Edison believes that in the interest of improved plast safety it is important to take all reasonable steps to assure the maximum avail-abih.ity of all safety related systems and systems required for safe shut-down. Toward this goal, all surveillance and test procedures are being reviewed for both unit: 1 and 2.
In particular, the Unit 2 emergency feed-(
water system survei1M.ca and test procedures will be modified to avoid isolating all flow capability from the emergency feedwater system during surveillance and testing so that at least two feed ptmpe and one flow path from the emergency feedvater system will.be operable at all times, his item is not an issue for Unit 1 due to the differences in design of the Unic I system.
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y In addician, a farmal routine shift check of engin'eered safeguards squip-ment, including the status of emergency feedwater pumps and valves has been instituted for those systems necessary for safe operation of T.MI Units 1 and 2.
Die importance of diligent monitoring of the status of safety equipment and the r' ole of the various administrative control systema in assuring proper and safe operation of plant systems is being emphasized in our operator training program.
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- 2.. Statement of Apparent Noncompliance:
Me severity and uniqueness of the accident which occurred at Mree Mile Island resulted in a marked reduction in the normal good health physics practices which are mandated by the NRC Regulations. Under the circum-stances of an accident of this magnitude the NRC recognizes that in the interest of reactor safety a departure from normal health physics prac-tices and standards may sometimes be mandated by the exigencies that exist during such conditions. However, the NRC also believes that the licensee, with the resources available and taking into account the time J
frame available for ' conduct of safety-related functions, could have takan i
additional measures to better control the overall health physics actions
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ad decisions which were anda during the course of the accident. Me l
following items of noncomplimes exemplify unacceptable degradation from health physics practices pertaining to control of access to high radia-tion areas, conduct of radiation surveys, and personnel radiation exposure monitoring.
10 CFR 20.201, "Surveye," requires in Section (b) that each licensee shall make or cause to be made such surveys as may be necessary to comply with the regulations in 10 C7120.
10 C71 20.202, " Personnel Monitoring," requires that the licensee supply p
appropriate personnel monitoring equipment and requires its use for each individual who enters a restricted area and is likely to receive a dose in excess of 25 percent of the applicable value specified in 10 CFR 20.101.
Technical Specification 6.12. "Righ Radiation Ares," requires that each area in which the intensity of radiation is greater than 1000 ares /hr be provided with locked doors to prevent unauthorized entry into the area and that any individual entering the area be equipped with a continuously indicating does rate monitoring devise.
10 CFR 20.103, " Exposure of individuals to concentrations of radisective materials in air in restricted areas," requires in Section (a)(3) that the licensee make suitable seasurements of the con entrations of radio,
active materials in air for detecting and evaluating airborne radioac-tivity in teatricted areas for the purposes of determining compliance with the regulation in 10 CFR 20.103(a)(1).
10 C7120.101, " Exposure of individuals to radiation in restricted areas," requires that no licensee possess, use or transfer licensed material in such a manner as to cause any individual in a restricted area to receive in any period of one calendar quarter a dose in escess of three res to the whole body, or 18 3/4 rem to the hands and forearms, or 71/2 ran to the skin of the whole body.
General Discussion:
Me specific statements of apparent noncom),11ance (2A through 2F) address apparent break downs in good health physics practices. ~ Although
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Metropolitan Edision does not agree with the conclusions of noncompliance in each case, the total assessment clearly indicates areas where i
. improvements in emergency health physics practice are needed. Addi.
tional comments are provided ismediately following each scacement (2A through 2F) for each example of apparent noncampliance.
In evaluating the execution of the post-accident radiation protection pro-gram it must be remembered that this accident was the worst experienced in 1
the commercial nuclear power industry. In addition only five people re-ceived doses exceeding their occupational quarterly dose limits during the period imediately following the accident.
As pointed out by the NRC above, under the circumstances of an accident of this magnitude departure frama normal health physics practices and standards
/
any somocimes be aandated by the exigencies that exist during such condi-tions. In some cases, the extreme levels of radiation did not allow for rapid complete area aspying prior to access. The heavy use of all available radiation instrumentation and contamination of analytical equipment req-quired alternate dose assessment measures. The immediate need for equip-ment operation and surveillance resulted in some violations of controlled access requirements. And the assignment of less experienced personnel no man stations replacing more experienced personnel required elsewhere re-suited in delays in radiation assessments. In all of the above situations, alternate measures were sought and applied in carrying out good health physics practices ar.d constant of fons were made to maintain acceptable health physics performance levels while meetiis the operational demands of the accident.
Summary of Corrective Actions:
Ta further strengthen our program actions including the following are f
being implemented.
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Mantatiment haa picced increased emphasis on observation of good radia-tion protection practices in all aspects of routine daily activity.
Since the issnediate post-accident period there has been a substantial ef-fort to upgrade the entire Health Physics / Radiation Procaccion program.
A:merous contract HP technicians and supervisors have been added te sup-port the station staff. Additional people in technician and supervisory positions have also been added to the permanent plant staff. Further additions to the pernanent staff are planned.
He revisions to the Radiation Emergency Plan has placed significant addi-tional emphasis on In-Plant Health Physics retraining on accident condi-tions. Procedures are being developed to define the specific approt.ch to r
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high airborne activity and the ability to analyse samples with extreme levels of gaseous activity.
Because of the lack of on-site monitoring capabilities, site monitoring devices will be reevaluated and enhanced as necessary. He upgrading of equipment requirements will enhance and thus eliminate the deficiencies in the respiratory program.
Me' addition of long handled tools and portable ' shielding will be completed as well as training of chemistry personnel. Additi,onal sir monitoring equipment has been purchased and is in place. De added d
capabilities for normal operations will provide increased assurance that response during emergencies will be adequate.
Retraining programs for Radiation Protection Personnel will also place additional emphasis on air sampling techniques and respiratory protection during normal operations. He procedures covering the respiratory pro-tection program have been upgraded and are in effect.
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The health physics program will be revised to firmly establish the posi-tive control concept and required training of all appropriate personnel will be undertaken to assure that full compliance with a positive control program is achieved, even under accident circumstances, The Technical Speci!Ications 6.12.2 should be modified to permit the imposition of a positive control entry systen during periods when locked doors are in-practicable, impossible or inconsistent with good health physics practices.
The revised Radiation Protection Plan has been submitted to the NRC in Amendment 7 to the Unit 1 Rastart Report and a similar plan is being-completed for Unit 2.
Revisions to the Emergency Plan have resulted in specific procedures which
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will be written to addre'ss post accident sampling and analysis to insure minimal exposures by personnel involved. The revision to the Emergency Plan includes specific plans for increased Health Physics support during the response to an incident. Included in the plan is the organization to focus on th's documentation and evaluation of individuals who 'are contami-nated. Procedures will be developed and training of personnel will be accomplished to fully Laplement the emergency plan prior to the start up of Unit 1.
The revised Emergency Plan has been submitted to the NRC as part of the J
Unit 1 Restart Report and is currently being reviewed for their acceptance.
The fo11 cuing gives additional comments to be considered in evaluation of the specific claimed noncompliances.
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2A..Statsmient of Aogarant Noncompliance 6
From 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on March 28, 1979 until the afternoon of March 30, 1979, the doors to the auxiliary building were not locked and access was not otherwise controlled even though the building was known to be a high radiation area with radiation levels such Sreater than 1000 erse/hr during this period; 4
piscussion:
Technical Specification 6.12.2. Providas that " locked doors shall be pro-vided to prevent v'* authorized, entry into... areas" in dich the radiation levei exceeds 1000 mesm/ hour. During the period from 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on March 28, 1979 until the afternoon of March 30, 1979, areas within the auxiliary building had radiation levela above 1000 mres/ hour. 2he anzi-11ary building was not locked. However, a program of posi.tive control f
over entry was established as contemplated by 10CFR,20,203.
Under the circumstances of the accident, there were a small amber of cases (two of l
Wich we are aware) in which the positive control program did not result l
l in the level of control desired. The failure to have locked doors coa-seituted a deviation from the Technical Specification, but steps were taken to be consistent with good health physics practices. Although Metropolitan Edison acknowledges that failure to maintain positive coa-trol in any particular instance constitutes a noncompliance with 10CFR20,
(
the overall program as implemented during the March 28 - March 30 period
(
was in conformance with 10CFR20.
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I The locking requirements of Technical Specification 6.12.2 are not written to provide an exception fsr cases in which spacea cannot or should not be locked.
10CFR20 on the other hand allows control of access to "high radi-ation areas" (i.e. areas with radiation levels greater than 100 arem/ hour) based upon " positive control over each individual entry" during periods j
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when /O:ess to the area is required. 10C71 20.203 (c)(2)(iii). It was this anchanism which was used to control access to the auxiliary building during the period in question.
Following declaration of the Site and General Emergency, positive control was established. The first steps were positive control of all individuals on site, ingress and egress control to the station, evacuation of non-essential personnel and individual task assignment briefings by responsible individuals prior to entry for usonitoring or other activities. To the ex-tent possible, the W==
amount of protective and monitoring equipment was provided. Under ene conditions of the accident, " positive centrol" i
was maintained during the period in question. No entry to the auxiliary
(
building was to be made without" appropriate anchorization. As circum-stances allowed, information on known conditions inside the auxiliary building was communicated to personnel prior to entry.
To improve reliability of entry control on March 29, the radiological con-trol point was moved from the ICC and ECS to the main entrance of the auxiliary building. Mis was possible because of the reduced levels of airborne contamination. 2he process of moving control points continued as' radiological conditions permitted.
As the Notice of Violation recognises, departures from normal health 3
physica practices are sometimes mandated by ascident conditions.
Un-anticipated conditions arise where neraal practices, such as locked doors or local control points, can lead to unnecessary exposures to plant per-sonnel and can be contrary not to be good health physics practices. By setting up a positive control program during the accident, the attempt was made to comply with good health physics practices to the nazisme estant while meeting the operational demands of the accident. We believe
i that the magnitude of the accident must be taken into account in evaluating i
the seriousness of the. instances where the controls which were established failed to be fully effective.
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25., Statement of Apparent Noncompliancer Tems the evening of Itarch 28, 1979 until the evening of March 29, 1979, at least two entries into the auxiliary building were made by individuals who were not equipped with a radiation monitoring device which continuous-ly indicated the dose race; Discussion:
Metrepolitan Edison agrees that a violation occurrad in that the individ-uals specified in NUREG 0600 Section II 3.2.4.6 and II 3.2.4.8 did not have radiation monitoring devices which at all times indicated the dose In each case, individuals were making entries into the Auxiliary race.
Building with some awareness of information on dose rates and anticipated exposures, and attempts were made to provide monitoring equipraent from the available equipment.
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s As indicated in Section II 3.2 other individuals who had previously entered the areas were questioned and approximate dose rate information was deter-mined. The approximate total exposure information for previous entries into the Auxiliary Building was also available for guidance to individuals planning as entry. A briefing was generally provided to each individual to insure that he was f' miliar with the specific task to be performed and the a
exact areas to which he was going.
(_
Section 3.2.4.8 indicates that this operator was aware of the area with res pect to proximity to primary system piping and available shielding.
Section 3.2.4.6 states that personnel did carry dose race instrumentation but at ciaes the low range instrument was of f-scale. Every effort was being made to provide each individual with the proper instrumentation.
However the limited number of instruments available during the first hours of the accident prevented this is some cases. The individual was aware of t
the dose rate to which he would be exposed and had estimated the total ex-3 A
posure to perform his task as 500 area. This was based on discussions with i
other personnel who had recently exited from the same area.
The exposures received by personnel making Auxiliary Building entries were not as low as we would have liked to have achieved. However, there were few overezposures relative to the number of entries and the associated radiation levels. The highest exposures due to Auxiliary Building entries were not significantly greater than the occupational limits specified in 10C7120. Also, during the initial days following the accident, evaluations were constantly being made to determine the risk associated with each oper-
~
ations and maintenance function, based on previous dose-rate and exposure
/.
information, areas which had to be entered and time needed to complete the s.
action. These risks were evaluated against the risk to plant personnel and
.the general public due to not performing the function.
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2C. Statement of Apparent Noncompliance:
f No measurements were made of the concentrations of airborne radioactive i
meterials in the Unit 2 auxiliary building for periods during which individuals were azposed freet 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on March 28, 1979 through aidnight March 30, 1979, nor in the Unit I nuclear sample room and primary chemistry laboratory for periods during which individuals were exposed from 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on March 28 through 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on March 30, 1979.
Discussion Technical Specification 6.11 requires that procedures for personnel radia-tion protection be prepared consistent with the requirements of 10CF120.
10CFR20.201(b) requires that each licensee make or cause to be made such surveys as may be necessary for him to comply with the regulations in this
~
part. Section 20.201(a) states that a " survey means an evaluation of tha radiation hasards incident to the production, use, release, disposal, or presence of radioactive materials or other sources of radiation under a specific set of conditions. When appropriate, such evaluation includes a physical survey of the location of asterials and equipment, and measurement of Invels of radiation or. concentrations of radioactive material present."
The general guidance of 10CFt20.201 therefore is applicable and Metropolitan Edison does not believe it was in violation of this regulation.
Energency conditions existed which made pitysical measurements impossible.
Specifically, because the on-site analytical equipment was located in a high background area and the samples that were taken had been gra saturated, the Ge (Li) capabilities were ineffective. Although a physical survey was de-sirable, the ineffective on-site analytical equipment prohibited its comple-tion. It was not felt that a physical survey at that time was nu essary to
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comply wfth 10CFR 20.201(b). Upon the arrival of a mobile laboratory on 2
Merch 28 at the Observation Cancer, two in-plant air samples were analysed t
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for iodine activity using a Ge (Li) detector. Neither indicated above the f
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minimum detectable activity for I-131 (NUREG 0600 Chapter 3 Section 3.2.4.6).
5 This delayed sampling demonstrated that the airborne limits wre not exceeded, f
1 The use of Self-Contained Breaching Apparatus units as demonstrated' in
.:UREG 0600 Chapter 3 sections 3.2.4.4, 3.2.4.5, 3.2.4.7, 3.2.4.8 and 1
3.2.4.9 exemplifies the maxi.aum positive action taken in the evaluation and protection of inhaled contaminates by the personnel involved. The referenced parsgraphs also establish that individuals were given whole body counts as soon as pesetical. Results of whole body count analysis showed that the protection afforded was sufficient to maintain internal exposures within acceptable limits, r ~^
The actions taken during the accident were within the general guidance and the intent of 10CTR20.201 for the circumstances that existed. Efforts were made to establish airborne activity levels, but, since operations which were vital to directing the plant to a safe condition and minimizing the impact on the health and safety of the public were necessary, some actions were taken without the benefit of thorough sampling and analysis.
Repeated efforts to obtain. additional samples and perform analyses would have resulted in added personnel exposure without assurance that the de-termination accurately represented the hasard., The data, although infor-native would not have caused a change in the health physics practices given the equipment available during the emergency.
d.
5
2D. Statement of Aooerent Noncomoliance:
2
~
On March 29, 1979, an Auxiliary Operator was permitted to enter areas of 5
j e auxiliary building where exposure rates of up to 100 R/hr existed.
Rzdiation survey information and appropriate personnel monitoring were f
provided to the operator for this entry. This contributed to the oct operator receiving a whole body dose of 3.170 rens. idhen this dose was added to the operator's previous dose for the quarter, the operator's quarterly whole body dose was 3.870 rems as measured by personnel dosi-astry devices; Discussion:
10cF120.101 (b)(1) states that "during any calendar quarter the total occupational dose to the whole body shall not exceed 3 rens."
It is evident by the indicated reading of the individuals TLD that the 3 ram limit was exceeded and therefore the regulation has been violated.
However, Metropolitan Edison feels that appropriate instrumentation to
('
define radiation levels was provided as well as adequate dosimetry in s
the form of TI.D's.
As indicated in NUEEG 0600 Chapter 3, section 3.2.4.7, a high range self-reading dosimeter was not available. As described in 3.2.4.7 of NUEIG 0600, the individual did not inform Radiation Protection personnel of his intention to make a second entry into the. auxiliary
. building. The individual assumed that his arposure was about I rem based on dose rate and stay tima information. No one was informed that the in-dividual's Low-Range Self-Reading Dosimeter was off-scale.
(
Following the second entry by the individual and upa determination by his supervisor of the off-scale reading on the self-reading dosimeter, the individual was removed from radiation areas and his TLD processed.
The events show the intent to follow sound Health Physics practices and
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provide adequate monitoring during the accident conditions. The events
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t do however show the need for additional high range monitoring equineent i
2 and the desire of personnel to respond to the actions necessary to y
aitigste the consequences of the accidant.
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4 2E.. Streament of Apparent Non-liance:
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on March 29, 1979, a Nuclear Ingineer entered an area of the auxiliary 3
building where the radiation level was greater than that which could be
?
asasured by=his portable survey instrument (21/hr). Failure to perform a survey of the exposure rate in this area contributed to the individual receiving a whole body dose of 3.14 rems for this entry. When this dose was added to the engineer's previous dor.e for the quarter, the engineer's quarterly whole body dose was 4.175 rems as measured by personnel dosi-metry devices; Discussion:
This item is a violation of 10CTR20.101. However, the' following circum-stance must be considered in evainaHng this incident:
{
In the time period (approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) prior to the Nuclear Engineer's entry into the auxiliary building, there was considerable concern for water 1eaking into the building from an unknown source. During the day of March 29 accumulated water had been removed from the basement floor via floor drains to the auxiliary building sump pump (s). By 6 PM the water level had again increased 2 to 3 inches on the 281' elevation.
This water was thought to be a possible contributor to the continuing, econtrolled release of radioactivity to the building and ultimately the environment. Since the water level was continuing to increase, it*was con-sidered vital to identify the source of leakage. An entry team was sent into the anziliary building in an attempt to visually locate sources of 1aakage. Prior to entry the tems, consisting of two engineers, was briefed J
on the known radiation arecs in the suziliary building. Radiation levels known at that time were limited to information obtained from previous en-tries. Information was limited because of the relatively small number of 3,
previous entries and the rapidly changing conditions. The entry team also I
reviewed ths intended locations of the areas to be checked and entry and i
azar routes.
Th2 carry ccan vac preperly dressed as indicated on I&E Transcri f
pt of Interview of " Engineer J" (May 2,1979, Tapes 92-93) i
, p. 13.
Each engineel was provided with a radiation monitor.
One had a high range instrument (0-1000 R/h) and the other had a low range instrument (0-2R/hr). A second high range instrument was not available.
Each was also previdad with self-reading dosimeters.
Together, the two engineers intended to jointly inspect a variety of equip-ment and cubicles, frequently monitoring radiation levels Shortly after entering, the high range monitor failed.
Therefore, the engineers tried to avoid areas with radiation levels beyond the range of th e low range in-strument.
However, in the one area where they were unable to avoid radia tion levels above.*R/hr, they checked their self-readin
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g dosimeters after leaving the area.
Self reading dosimetry was checked after leaving an area in which the low range monitor reached the full scale reading At that time each engineer had received less than 500 mR.
The engineers separated after 10 minutes of the 15-20 minute total entry
, time.
The engineer with the low range instrument returned to the 281 elevation to conduct a further investigation.
The other engineer went to the radweste panel (a lower radiation area) to adjust valvepos itions.
The engineer with the low range instrument continued the tour
, frequently checking the radiation levels and attempting to cover as such of the building as possible.
After finding that the radiation level at the doorn way to the RC Bleed Tank room was 2R/hr, he checked his low range self-reading dosiasters and noted that it was off scale.
I The engineer immediately.
left the auxiliary building.
The high range dosimeter read in excess of I
3R.
Each san was debriefed by a Radiation Protection Foreman for the lo cation of high radiation areas.
The Supervisor of Radiation Protection
and Chemistry instructed the engineer with the,high reading to have his f_
i I
1 TI.D read to confirm the exposure. Me esposure was confirmed and the Y
engineer was, restricted from further activities in controlled areas for l
the remainder of the quarter since his esposure exceeded the 3 Rem / quarter limit of 10CFR20. Separate reports and e-aluations have been submitted to the NEC regarding this matter.
Me entry which resulted in the exposure of the Nuclear Engineer in excess of linics specified in 10CF120 was made with strong' consideration toward exposure control. Se entry was considered to be vital to the limiting of release of radioactivity and =hi=iming the effect on the general public.
~
Proper radiological practices were followed to the degree possible under r
the existing conditions.
his dose was well within the manual guidelines of 10CF120 and far be-low the guidance of the National Council on Radi.ition Protection for emer-gency and accident conditions.
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Streement of Apperent Nonconolirnen:
3 On March 29, 1979, a Chemistry Foreman was permitted to repeatedly enter i
high radiation areas and handle samples of highly radioactive reactor
.I coolant. This contributed to ch'a Foreman receiving a whole body dose' of 4.100 ress. When this dose was added to the Foreman's previous dose for che quarter, the Foreman's quarterly whole body dose was 4.115 rems as
=easured by personnel dosisetry devices; 2G.
Statement of Apearent Noncomoliance:
On March 29, 1979, a Chemistry Foreman and a Radiation Protection Foreman were permitted to handle a highly radioactive reactor coolant sample without adequate personnel monitoring and without first performing a survey of hand and forearm exposure races. Handling of this sample resulted in a calculated dose to the hands and forearms of the Chemistry Foreman of about 147 rems and a calculated dose to the hands and forearms of the Radiation Protection Foreman in the range of 44 to 54 reas; Discussion Items 2F and 2G deal wit,h the same event and therefore are covered 7
together in this discussion. Metropolitan Edison Company agrees that tdtile obtaining a Reactor. coolant System sample on 29 March 1979, ex-posure to the whole body of one individual and exposure to the extram-icies of several individuals exceeded 10CTR20 limits. Also we agree that adequate extremity monitpring was not used by the individuals, however all evaluations of extremity exposure have been complaced and documented.
Metropolitan Idison feels however that the circumstances surrounding the drawing of the sample indicated that serious attention was given to radio-logical practices and that the sample was obtained in a way that minimized a
exposure using available equipment considering the urgency of the sample requirements.
The individuals involved in the sample were knowledgeable both in the 3
sample rystem and radiation protection practices. A plan to obtain the sample was developed by the individuals which included providing respira-tory protection, protective clothing, high range dose rate instruments,
rotation of personnel, checking of dose rates after each step and check-I i
ing of self-reading dosimeters after each step. The sample was not 7
obcained uncir the individuals felt confident that all were aware of the plan and that all reasonable planning had been done. Extremity monitors were not ismediately available. The use of wrist badges which were on-site would not have eliminated the need for the evaluations perfatwd but would have simplified them.
During the drawing of the sample, the plan was followed to the degree pos-sible and dose ratas and exposures were continually checked. Although contact dose rates from the sample were greater than the range of the high range instrument available, dose rates at a few inches were measured
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which provided an indic' cion of the readings on contact.
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Statement of Apparent Noncoisoliance:
T On March 28, 1979 and March 29, 1979, several individuals received skin I.
contamination of the hand and other parts of the body sufficient to cause
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exposure races in the range of 20-100 mR/hr when measured with a hand-held survey instrument and no evaluation of the dose to the skin of these individuals was made.
Discussion:
Metropolitan Edison agrees that in the cases of individuals contaminated on March 28 and March 29, 1979, a timely evaluation of the dose to the skin of these individuals was not performed. For those individuals whose axposure exceeded the limits of 10CF120.101, failure to provide a written report within 30 days is a violation of 10CTR20.405. Evaluations have been made based on available information, including whole body count data, survey data, and personnel interviews. A report (Raf. Mec-Ed letter GQL1094 of 21 August 1979) has been submitted for those individuals in which the skin dose due to the contamination vsa a contributing factor in exceeding their quarterly dose limit. For those individuals whose skin dose was below the 10CTR20 specified limits, evaluations are complete and available.
It should be noted that the exposure rates were messured immediately fol-loving entries into the Auxiliary Building or Nuclear Sample Room.
Initial decontamination efforts occurred within a few hours and in some cases,
~
vithin minutes. The decontamination efforts resulted in a significant re-duction of contamination levels for personnel involved.
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Statement of Apparent Noncompliance:
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Technical Specification 6.5.1, " Plant Operations Review Committee,"
I requires:
in Section 6 5 1 6 4., that the Plant Operations Review Commit-i tee (PORC) review all procedures (and changes thereto) required by Technical Specification 6.8 and any other procedure (or change) determin-ed to affect nuclear safety.
Contrary to the above, inadequate reviews were performed on both Proca-dure Change Request No. 2-78-707, Revision 4 to Surveillance Procedure 2303-M27A/3, and Procedure Change Request No. 2-78-895, Revision 8 to Surveillance Procedure 2303-M14A/3/C/D/E; both were reviewed and approved by the PORC (November 9, 1978 and August 15, 1973 resp'ec tively). Each approved change included a valve lineup which resulted in emergency feedvater header isolation, contrary to Technical Specification 3/4.7.1 requirements.
Discussion:
Metropolitan Edison does not believe that it has violated the cited Technical Specification.,
On August 15, 1978 the Plant Operations Review Committee (PORC) reviewed and approved in writing Procedure Change Request (PCR) No. 2-78-707, finding that this. item did not constitute an unreviewed safety question.
Ca Novembei 9,1978 the PORC reviewed and approved in writing PCR No.
2-78-895, finding that this item did not constitute an unreviewed safety question. These actions demonstrate conformance with Technical Specifi-cation 6.5.1.6.a and 6.5.1.7.b which requires a written determination as t'
to whether or not changes to procedures constitute an unreviewed safety question.
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As discussed in connection with Apparent Noncompliance 1, Metropolitan i
l Edison believes that neither PCR was contrary to the interpretation of l
the Technical Specification 3/4.7.1 requirements. Therefore.the reviews
- i conducted by the PORC were not inadequate. This belief is confirmed by e
the review and acceptance by the NRC of Revision to Surveillance Pro-
~
cedure 2303-M14A/3/C/D/E documented in their Inspection Report letter dated January 9, 1978, " Combined Inspection 50-289-78/23 and 50-320/
78-36.".
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2 Many approved surveillance and test procedures render safety related sys-
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tems inoperable for short periods of time and so long as the inoperable
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period does not exceed the Technical Specification time limit for system inoperability this practice has been considered accepcable by both Metropolitan Edison and the NRC. Examples of similar surveillance and test procedures are:
2303-H2A/B - Decay Heat Removal Pump Functional Test and valve Operability Test 1303-5A2
- RB Hatch Leak Race and Interlock Test 2303-M25A1B - Decay Heat Closed Cooling Water Pumps Tunctibnal and Valve Operability Test.
The reasons for the changes to the surveillance procedures were to take into account unnecessary thermal shock to the emergency feedwater nozzles
/*
and to ob;ain repeatable results for tests required by the ASME Code Section II.
The thermal shock consideration for the emergency feedvater nozzles is significant since the frequency of the costs required by the ASNE Code Sect XI would reduce the number of available thermal cycles associated with normal K7W actuation, thereby reducing the service lif.:
of the nozzle connection due to higher cumulative thermally induced str es ses.
Corrective Action:
Because Metropolitan Edison believes that the subject PORC PCR review was in conformance with the Technical Specifications no specific action is required. We do not believe that these PCR's placed the, emergency feedvater system outside the licensee requirements.
However, as discussed in connection with Apparent Noncompliance 1, Metropolitan Edison believes that in the interest of improved plant safety it is important to take all reasonable steps to assure the
maximum availability of all safety related systems required for safe shut-I.
Toward this goal, all surveillance and test procedures are being
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reviewed for both Units 1 and 2.
In particular, the emergency feedwater system surveillance and test procedures will be modified to avoid isolating all flow capability from the emergency feedveter system during surveillance and testing so that at laaet two feed pumps and one flow path from the emergency feedwater system will be operable at all times.
In addition, a formal routine shift check of engineered safeguards equip-including the status of emergency feedwater papa and valves, will
- meat, be instituted for those systema necessary for safe operation of DfI Units I and 2.
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4.A Statement of Apparent Noncompliance; 5
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Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified i
activities.
A.
Imergency Procedure 2202-1.5, "Pressurizar Systen Failure " Revision 3, requires in Section A.2.3.1 that electromatic relief isolation i
valve RC-E2 be closed if, mong other things, the valve discharge line temperature asceeds the normal 130*T.
Contrary to the above, the electromatic relief valve discharge line temperature had been in the range of 180*-200*T since October of 1978 and isolation ' valve RC-t2 was not closed as of OMO hours on March 28, 1979. Additionally, on March 28, 1979, the diseaarge line temperature of 283*T was noted at 0521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br />, but the isolation valve RC-E2 was not closed until 0619 hours0.00716 days <br />0.172 hours <br />0.00102 weeks <br />2.355295e-4 months <br />, allowing a significant loss of RC inventory.
Discussion:
1.
Operation from October 1978 Metropolitan Edison believes that Beargency Procedure 2202-1.5, " Pres-surizer System Failure", was not violated during the period from October 1978 through March 28, 1979 notwithstanJing the temperatures of the discharge line from the pilot operated (electromatic) relief valve
("PORV").
Althohgh this procedure was understood by the plant staff, it is not clearly written and does not reflect actual plant conditions. It will be changed. Bovever, although Metropolitan Edison is concerned about the issue, there is no indication that this procedure or the history of PotY discharge line temperatures delayed recognition that the PORY had stuck open during the course of the accident.
l Bnergency-Procedure 2202-1.5 describes in each of its sections a poe-sible failure in the pressuriser system, including 1eaking or inopera-i tlw Potv and leaking or inoperative code relief valves. Each section of the procedure sets forth a number of " symptoms" and several immed-l Late and follow-up actions. The crux of the claimed noncompliance f
is the assuspelon that the occurrence of a " symptom" autoestically requires the implementation of the associated iMiate and follow-up 34
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- E actions. 2 1s asesseption is not supported by anything in plant I
i':
procedures or Technical Specifications and was contrary to the under-
.i standing of Metrop:11can Edison personnel at the time of the accident.
A symptom is not a determination that a probism exists. Esther it is a signal that conditiosa should be===i==d to determine whether the problem eziata. S e same sympton may be equally consisteat with several underlying situations. For this connection, it should be noted that the symptoms for leaking 70kT (Procedure 2202-1.5, Section A) are essentially idenciaal to the procedures for leaking code re-lief valves (Frocedure 2202-1.5, Section C). M us, the existance of a symptom for leaking POIY doe,s not maan that thsre g a leaking POIY.
If there is no leaking PotY, then the procedure for leaking PORY is not relevant ad it is not appropriate to apply the immediate and fellow-up actions of that procedure.
As described in Section A of Procedure 2202-1.5, the immediate action for. a leaking POI? is the closwe of the Electromatic Relief Isolation Yalve, RC-Y2.
He claimed noncompliance is that this valve was not closed dwing the October - March period despite the existence of one of k
the symptoms of a leaking PORY, specifically "Ic. lief valve dischar'ge line temperature enceeding 3
the nozmal 130*7. Alarms on computer at 200*T."
More is no dispute that relief valve discharge line temperature enceeded 130*F during the period in question. Se temperature range during this period was generally 170* to 190*7.
Eowever, these temperatures do not i
i appear to have bees the result of a leaking PORT. During the October -
I January period, the reactor coolant drain tank leak rate (which would have reflected leaks past the PCIT) was essentially zero. After the outage which ended on January 31, the reactor coolant drain tank leak rate increased.
i 2
However, this was accompanied by a sharp increase in the discharge line temperatures for the code relief valves.
In the October -
January period, these temperatures had been in the 100* - 115 'T range.
After the outage, the temperatures sharply increased to the 160* - 180*F range. These matters were discussed by the plant staff. Based on tanp-erature readings, a detensination was made that code relief valve 171A vas leaking and Votk Request No. C-i137 (February 9,1979) was prepared far une repair of this salve.
Additional eeidence that the 170* - 190*F temperatures on the POR7 iischarp line did not result he= a laaking PORY can be found by com-parics chase terperaturt.s with plant einditicas. During the cetobst -
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March period, this temperature renge occurrvd wnether Usit I was at p<wer er in het r. hut dovr..
For example, on October 1,1978, uhile the peintry systex vis sc 250*F and 265 pai, ths PCEV distharge line :emo-erature was, 171.1*7.
Om October 29, 1978, with primary tystem tempers-i cure of $56*F ap.d pressurs tc. 2135 pai, the discharge line tempersture l
L was 176.4*F.
Galy when the Unit was in cold shutdown did the discha.rgs line temperature fall below the 170' - 190*F range. For example, un January 18, 1979, with primary system temperature at b
130*F and pres-sure at 0 psi, discharge line temperature was recorded at 80*F.
These w
J values make it clear that discharge line temperatures did not, of them-selves, establish that the PORY was leaking. More likely, the tempera-l tures resulted from the heating of the line by conductivity from the i
pressurizer itself. Because the temperature sensors on the code relief valve discharge lines are located much further from the pressurizer I
than those for the PORY discharge lines, the normal temperatures for the former were not affected to the same degree by conductive heating from the pressurizer.
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l Corrective Action:
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Based upon the above discussion, it appears that the underlying cause for the claimed non-compliance was the statement in Section A.1.1. that the " normal" temperature of the relief valve discharge line was 130*F.
The normal temperature was actually in the 170* - 190*7 range. Once the plant staff determir.ad that this discharge line temperature was not nor:nally below 130*7, the p 9.edura should P.sve besa changed. Hatropolitau Edisarr.'s training program dll thertfors include vtg.t to assure thet i
tasse t' pes of changes are initicted when approp-iate. Ta addition,
- tropolism Mison procedwes will Le clarified :o aske erplicit e):e i
maari.nt act role of "symptoe.s* in thane procedures. And, Me:ropolitaa b
1 toisen will addense at.d recoghise the deficiencies iden:ified in the Tr0ced.tre by the Staff of the P~cesident,'s Commission en *:he Accident at i
fr.ree Mile Island. M Tschul:a1 Staff Analysia Report on Technical l
.Lasvunnaast of Operating, abeomal, and Energer.cy Pro:edurse (Cetoba, 379h pp. 14-17*.
i 2.
reilure to Closa Isolation valve on March 28, 1979 The second aspect of this claimed noncompliance stater that a dis-charge line temperature of 283*F was noted at 0521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br /> in March 28,
(
1979, but that the isolation valve RC-12 was not closed unt'il 0619 A
hours. latile the discharge line temperature at 0521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br /> was 283*7, this value was not known by the Shift Supervisor at that time. Only at 0618 did the on-coming Shift Supervisor observe that the FORY discharge line temperature was significantly higher than the temperatures for the I
code relief valve discharge lines and emise the closing of the POIV isolation valve RC-72. Since Emergency Procedure 2202-1.5 does not matomatically require that the isolation valve be closed on a high temperature reading, failure to close it until 0619 did not violate the 5.
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i procedure. However, an explanation of this aspect of the accident is g
)
7 nonecaeless appropriate.
Following the turbine trip in the early stages of the March 25 acci-dent, the Pot? was. expected to open (see Procedure 2203-2.2, '*;urbine Trip", step 2.A.3).
The operatcrs noted that it did oper. and shortly chareafter received an erronecus indication thk2 it had eined.
At 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br />, toe temperattrres for ths PORY ami code relief valve discharge line were called up.!rr.e cae ca pu er, the reported tsny*
~
oracues me 2t.5*/ (PORT), 2M*7 (relief valvs A) and 275*7 (relief
- alia 1)
- Slae.a the PC27 had juAt liftsd, thEse CSSperaCETes Mt-eut ecusIdered t.nuseal.,
Pur*'inneori, the three temperatures were i
j gecuped rencouably close together.
.e C52.' hours, another sat of discharge line temperctures was called i
i up from the conyucat. h wayot u priectat, a egy of which is at-l tached, printe* the. tageratsre for code rslief veles 1713 twice and physically separated by about an 12.ch the 283*F value for the P017 i
l discharge line. Because of this printing error, it appears that the I
high valve for the POE7 discharge line may have been missed. The i
Shift Supervisor has recalled char, the valves were lower than those 4
taken at 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br /> and that all three readings were si=41me, thus not alerting him to e abnormal situation.
Finally at 0618, the on-coming Shift Supervisor again called up the temperature data from the computer. Perhapa noting that the PORY
'These values were reported to the Shift Supervisor. His later re-
~
collection was that the FORY discharge line temperature was 225* 230*T.
I&E Transcript of April 12, 1979 Interview of W. R. Zeve, p. 29. IEE Transcript of March 30, 1979 Incarview of W. R. Zeve, p. 16.
e.
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k s
discharge line was almost 40*F higher than the code relief valves, he
.,2, i
suggested the closing of the PORY isolation valve. With that closing.
the loss of reactor coolant inventory was stopped.
Corrective Action:
Althoust this asserted noicompliance did not isvolve any violation of pla;2c peccedures, it does poiat up tt. esod f7r i.g rond and expandel trainin6 of the oFerating organizrtion. son cne wei for battar diagnostic capabilities. With the than axis, ting pt ocedaer ad traini.Tg and the availabilie7 of instrueen:ation fr.r tdantifring a PORY failure, the delayed diagnonie of the 70R7 status is est surprising. Je ter training i
l f'.
(
and procadares and plant instrunaccation wdificaticns wi.11 be implee4ated e
to improve the ability of the operating orgsnizacion to diagoose such i
conditions.
1 These actions will be corplated for Unit 1 prior ta re-start of the unit.
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05:31:05 DATA 0098 CtOS PUl#S CUTLET ICR PRESS 164.5 GROUP 4 i
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i 4.3.1 Statement of Aonarent_ Noncomoliance:
I Y
5 3.1 Emergency Procedure 2202-1.3, "I,oss of Reactor Coolant / Reactor Coolant System Tressure," Revision 11, requires in Sections 3.2.2.3, 3.3.6.2 and A.3.2.5:
that high pressure injection is initiated on low RCS pressure (1600 psig), and that the operator verify high pressure injection is operating properly as evidenced by flow in all four legs (250 gpe); that flows be maintained at this rate by throttling as RCS pressure drops; and that high pressure injection not be terminated until ACS pressure can be mai: stained above the reset point (1640 peig) or natil low pressure injection flow is estselished at 3000 gym.
Castrary to the above:
1.
At about 0405 on March 28, 1979, high pressure injection flow was throttled to miniatu conditions even though RC3 pressure was 1ess than 1600.si and falling, and without low pressure injec-5 tion flev established.
1.
At various times throughout the day of March 28, 1979, the high
/
pressure injection system was modi.fied such that the required flow rates were not maintained during continuing lo'w pressure conditions within the RCS following the period when the reactor coolant pumps were stopped and the high pressure injection system was the only mode available for the removal of core decay hast.
Discussion:
Metropolitan Edison recognizes that, in the light of detailed after-the-fact analysis the failure to maintain full Righ Pressure Injection flow in the first few hours of the accident led to severe core damage.
l It is not clear however that this failure was a failure to comply with
[
L.
t one procedure, as described in the statement of apparent noncompliance.
I 3
i This failure was due to a complex interaction of system performance characteristics dictated by design, equipment failures, analytical myopia, procedural inadequacies, technical specification conflicts, the focus on regulation requirements as necessary and sufficient standards in themselves, and training which reinforced many of these inadequa-cies as well as being deficient in the general treatment of accidents outside of predefined events.
4 42 '
.4
=
In order to assess grocedural compliance, it must first be determined
-j 7
which procedures were possibly applicable and then deter: sine which j
i
' procedure was in use.
However it is generally recognized and accepted that specific procedures cannot be written to cover every possible sequence of events or conditions.
Thus analysis of these more complez conditions requires a much broader review of influences beyond the specific guidance of written procedures.
Analysis : uat consider thqr basis for judgements made by the operator in order to uska A determination of non-compliance. The conditions surround-ing the TMI-2 acciden: require that this latter approach be taken.
The conditions that the ' operators perceived inssediately following the turbine trip appeared to reflect.a normal response to a loss of feedwater.
f The immediate actions for the trip were followed. These included verifi-cation of automatic functions, manual start of a second makeup pump to account "for system shrinkage, and isolation of letdown to help reduce the effect of. shrinkage. Shortly thereafter the situation deviated i
from normal. This vaa apparent to the operators when the pressurizer 1evel stopped its decline sooner than expected and then started a rapid i
recovery. The rise in pressurizar level could not be stopped by the operator as the level rose through the normal range and continued to e
fill the pressuriser beyond the Technical Specification limit of 385 inches, until it appeared to be solid or nearly so.
During this same period two other significant events occurred. Bsergency feedwater was discovered to be blocked by the unexplained, closed condition of EF-712A&B, and High Pressure Injection was automatically iniciated as RCS pressure dropped through 1640 psi. The operators took corrective action for each; High Pressure Injection was placed in manual control and significantly cut back to attempt to prevent going solid, and Emergency Feedwater flow 43
g
, was initiated to both steam generators to restore levels to approximately 1
2 30" as described by, the 1,oss of Feedwater Procedure (IP2202-2.2A). It b
should be noted that the possibility of a loss of both main feedwater pumps and failure to achieve emergency feedvater flow is not addressed by procedure, nor is the condition where Reactor Coolant Systes pressure is low (less than 1640 psi) and pressuriser level in high. At this time (approximately 0408), the conditions of the plant were outside of the entirs set of unit cperating procedures. Bis situation required the eparator to make judgements on the best course of action on the basis of his seneral training and experience, technical spee.ification limits, general plant operating limits and orecautions and whatever other guidance, although possibly inconsplete, might be derived from specific amergency procedures.
De response of the operator is cenditioned by the order in which ha receives, recognizes and reacts to conditions in ese plant. At events j
continue to develop, early judgements are modified, discarded or reinforced by the analysis of incoming data versus the expectation of what should have occurred. De expectation of what should occur is a function of training and experience.
4 2e specific events which followed the brief period of " normality" did not fall into easily recognizable, discrete events. De operators were forced to recognize the correct symptoms from several hundred abnormal alaras and indications. Me following brief discussion reflects the key f
procedures which may have been applicable and were used in part, and the I
events which reflect that these procedures may have been applicable to some degree. Steam Supply System Rupture (AP2203-2.3 Rev.5) is identified by the following symptoms:
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7:
"1.1 Rapid decrease in secondary pressure. (Both OTsc's start to blow
+
down)." At 0407:45 both OTSC's reached their minimum pressure of 788 psi on A and 777 psi on B.
This was the approximate time e
the Emergency Feedvater valves were opened.
"1.2 Electrical load reducing rapidly." The turbine generater had already tripped.
~
"1.3 Decrease in pressurizer level,1. C. Pressure, and cold les temp-erature." Pressurizar level had decreased inically then rose un-espectedly; 1.C. Pressure had rapidly decreased below the Safety Injection setpoint of 1640 psi and was still decreasing although more slowly; and following an initial rise during OTsc dryout, the RC cold leg temperature began a steady declina.
"1.4 For a rupture inside the 2eactor Building; Indicscion of increas-ing building pressure and temperature. (Possible higt Radioac-tivity Lavsls on IP-t-227 if a tube leak exists)." At approzi-mately 0415 the Reactor Building pressure started to increase followed shortly by an increase in temperature. His was due to the rupture of the RC Drain Tank but was unknowa at that time.
"1.5 For a rupture outside the Beactor Building; Noise may be heard in Control Room or a report made from personnel outside the control Room." There were no reports of this kind.
"1.6 Decrease in main condenser hotwell level or condensate storsge tank level." The hoewell level had increased dua to a blockage in flow at the condensate polisher outlet but the condensate storage tank was being slowly depleted by the Emergency Feewater System.
2 These are all the symptoms for Steam Supply System Rupture. All of these symptoms were essentially met in the first,15 minutes of the accident.
l During this early period not all of the symptoma continued to persist but
~
the deviations were partially understood by the operator when the combined effects of loss of usin feedwater, reactor trip on high pressure, and sub-sequent dryout of the OTSC's were considered. These operators had been previously exposed to the conditions of a main steam line break from the I
April 23,1978 Main 3 team Belief Yalve failure event on TMI-2, and there-fore the existence of low reactor coolant pressure without a LOCA was. an i
easily conceivable situation. Since this was apparently not a large rupture based on the rate of change of parameters, the operators had no analytical base or training to rely on in datermining whether these
-45
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specific conditions were fully reflective of a small rupture. They were 2
forced to make judgements at this time to assess the next appropriate steps.
This senary is supported by the Shift Supervisor and control Room Operator in interviews with the NRC (I&E Group Interview, 6/28/79, Tape 319, pg. 38-48). Specifically, the control Room Operator and the Shift Supervisor agreed that at about 0415 they thought they had a leak in the steam gener-ator. As the control Roat Operator highlights the fact that the principle distinguishing factor between a LOCA and a steam system rupture is the ab-sence of a Reactor Building Air Sample alarm on EF-t-227, (which was not prssene until about 0615) and,the absence of an alarm on the condensar off gas monitor, 7A-t-748, (which indicates an CTSG tube failure and was re-l cerved about 0700).
The automatic and manual imediate aseions outlined by the Steam Sysean Rupture procedurs' were followed for the conditions as they existed.
Since both OTSC pressure had not reached the feedveter latch setpoint (585 peig) the CTSG that was leaking had yet not been isolated. However about 0510 a significant pressure differential started to exist at
(
between the two OTSG's, with the B OTSC rapidly falling. By 0520 a 150 s
poi differential existed. At 0526 the B OfSG was isolated. Continuing 2
to perform follovup actions in this procedure would have the operator i
initiate High Pressure Injection if pressuriser level dropped below 20",
ECS pressure decreased below 1600 psig or neutron fluz starts to rise.
I Subsequently, when pressuriser level returned to above 100", High Pressure Injection was properly secured. There are no criteria or additional guidance on restoring reactor coolant system pressure above 1600 pai.
Included within these steps,is the requirement to initiate a cooldown with the unaffected steam generator.
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, These steps were all followed, and yet the plant was not adequately pro-5 tacted. The operator did not understand why pressurizer level was per-4 forming anomalously and because of this continued to evaluate other possibilities while dealing with the other events of the moment.
1
~
A similar analysis of the symptoms in Emergency Procedure 2202-1.3 " Loss of Reactor Coolant / Reactor Coolant System Pressure" Rev. II uould not result in the conclusion that this procedure should have been followed.
A I.0CA of significant size is identified in EP 2202-1.3 3 by:
"1.1 Rapid extinuing decessee of reactor coolant pressure.
(1) Lo alaan 2055 pdg.
(2) Lo-Lo-stara 1700 psig.
(3) Safety Injection acucatien a: 1640 peig."
This condition exist;ed until pressure initialy stabilized anar i
1000 pai. Tais was substantially above the accident analysis levels of less than 630 psig for classical large and small I
break analyses.
"1.2 Rapid centinuias decrease of possuriser level.
(1) Lo alarm 200".
(2) Lo-Lo alara 30" (Interlock heater shutsff)".
Fallwing an initial decrease following the trip, pressurizer level ceasinually rose uniti caceeding Technical Specification (385") and procedural (400") limits.
i "1.3 Ri Radiation alarm in Reactor Building."
Although the specific alarm is not noted in this step, it is discussed in the note below step 1.8 as IF-R-227.
This did not alarm until approximately 0615.
"1.4 Reactor Building Ambient Temperature Alarm."
This alarm was received at about 0420 following the RC Drain I.
Tank failure.
"1.5 Hi Reactor Building Sump level."
This was not recognized until approximately 0430 when reported by an auxiliary operator to the control room.
"1.6 El Reactor Building pressure (RC3 or main steam line rupture)."
"1.7 Rapidly decreasing makeup tank leveT."
The makeup tank level increased steadily.
"1.8 Both core flood tank levels and pressures decreasing."
This did not happen in the first eight hours.
t:
T1.s note at the end of the synpcome sections states that an operator may distinguish between LOCA's, OTSC tube ruptures, and steam breaks inside 4
l the reactor building by specifically checking three conditions " unique to the aforementioned accidents." These conditions are:
e
-47 '
.+r.
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.,.,,-.,..--..,,.m..
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"1.
Loss of coolant inside Reactor 3uilding particulate, iodine 5
gas monitor alarm on EP-t-227 " Reactor Building Air Sample."
y "2.
OTSG tube rupture - Gas monitor alare on VA-R-748.
"3.
Steam break inside Reactor Building:
e (1) Low condensate storage tank level alarm - and or low hoewell level alarm.
(2) FW Latch System Acutation."
None of these criteria were fulfilled until approximately 0615.
i ne actual conditions of the accident following 0405 did not clearly fit the symptom of a significant loss of coolant but more closely fit the symptoms of a staan systar tupture. De operators felt they had a steam systra rupture and ec,aplied with that procedure. It was 0619 when the operators become avere that the plant had actually Juffered a LOCA in
{
the pressurizer steam space.. his recognition insediately followed the closure of the 7017 block valve, but the plant conditions had degraded to a point whers all subsequent activities were clearly outside of pec--
cedural guidancs. Bere were oftst. procedures which to a lesser extant could have been considered since some of the symptoms present did exist during this period. However, these procedures would not have been appropriate.
De discussion above highlights the potential for msbiguity or conflicts I
in procedures and the difficulties facing plant operations when the plant conditions in a complex accident no longer fall within existing A
l l
plant procedures. lihan the conditions within the plant deteriorate to a point outside the scope of procedural guidelines, the operators must be required to exercise judgements on the spyropriate course of action.
I Dese judgements must be founded on training which prepares the operator I
to evaluate alternative actions that will lead to a sequence of events which will satisfy the fundamental requirements of decay heat removal, control of radioactive material and protection of the general public, plant staff and equipment.
4g,,
5 5
In the period following the closure of the PCR7 block valve, the full TE 5
extent of the initiating events came into clear focus to the operators.
i Normally this would be the end of the transient phase and would be followed by an orderly transition to cold shutdown following existing procedural, guidelines. The conditions at TMI-2 were not however within conditions defined by procedure. The operators were therefore required ca maka judgnments in order to fulfill the basic steps towards long tava stability. During this period the operator made several attempts to schieve long tema stability of operacious within the scope of his gocedures and training.
i
('
At cartain times the uro of High Pressure Injection fulfilled the P
c.uolius req sirescuts of the core and at 'achet times it rcpslavented core hist rencval via the steam generators. The sre of Vigh Pressurt Injecelon had to be carefully balanced with other plant eerditions and parameters..The operators would not take the plant ei the 2500 psi safety valve setpoint since this presented an additional pocestial failure that in their minds would seriously degrade the plant condi-l tion. Judgement was sound since a valve which failed open would have
{
b ultimately fore'ed the use of the Decay Heat System in the recirculation i
ande from the Reactor Building sump. The continued use of the atmospheric ste m dsmp valves which was removing core heat via the steam generators was perceived to be a significant release path by off-site authorities ad they exerted pressure to terminate the use of this cooling pech.
This was done at approximately 1230. From this point on, a combination i
of High Pressure Injection and core boil off through the FORY block
~
valve was the only effective method of cooling.
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2
~
Analysis of the plant data of that period would show that cooling appeared f
i to be adequate as evidenced by gradually decreasing hot leg temperature in the A stema generator, until the point the A cold leg temperature be-gan to respond. At this time the response in the A cold leg was indica-i tive of some natural circulation type flow resulting from parcial refill-ing of the A loops. Throughout the next period variatiena in the acount and location of High Pressure injection coupled with the use of the 70R7 block valve continued to effectively rencve core heat whila taking the place towardr the more stable condicion of fcrced circulttica.
The performanca of the operatters proved to be ulitimately effe:riva
(,
during the period subreteest to closirg the block value at 0619 vuon no 1
procedural guidance or specific training guidance was available. It is.
I recognized that this process may not have been the optimum nethod to ichieve stable, forced circulatico condities, but there la insufficient analysis available to conclude that tt.a perforsmace of the operators after 0619 was an additional contributor to the final condition of the core.
Correceive Action:
[
The shortcomings in procedures and training identified by this accident are being corrected in both units. A thorough and on going program to review, upgrade and effectively integrace those aspects of design, ana-lysis, operating experience and regulatory requirements essential for a sound nuclear program is underway. Organizational changes necessary j
to support this program have been made on both units, with each organ-i laation tailored to the specific needs of the unit.
~
8 Specific training emphasis on the responsibilities of operators to comply with procedures has been. added. Guidance on actions to be taken when S
5 i
. outside of the procedural envelope has been added. Metropolitan Edison f
is participating in industry progran to improve operator guidelines which
?
sia toward satisfying the basic requirements of reactor safety for all transient conditions. Mose guidelines will be integrated into the training progran and procedures as they become available.
Meeropolitan Edison will continue to use diverse management tools seca as internal and extezzal eveninatiess of operations, independent saie:y re-views and expanded goality assurance activities on a continuing basis to pesvide the==M-=
possible opy3retasity for issues important to safety
~
es be identified asd resolved.
(
Me coa.tinual development wf improved analytic.a1 techniques will be used to perform best esti:sace scalyses of transient and accident sequencer using specific plant data. De results of these analyses will be used as an aid in opentor training. Sese malytical models have been initially henchnacked against actual plant performance. His will allow future predictive analysis to understand operational events not previously ani:icipated in safety analyses. Procedural change will result when pase guidance is shown to be in error.
C.'.
It is recognized that these changes not specific in their implementation a
schedule. Dese changes are not quickly conceived or implemented, but i
rather require deligent preparation, careful integration, and ongoing re-view to assure that the improvements necessary are achieved and maintained.
i Metropolitan Edison is committed to this effort and will continue to meet the high standards demanded.
i Me requirements in training and procedure identified by the NRC in each of their task force activities are being reviewed by the NltC and will be
=
completed prior to the restart of Unit 1.
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I 4.3.2 Statement of Apoarent Noncomaliance:
I 3.2 Emergency Procedure 2202-1.3, " Loss of Reactor Coolant / Reactor I
Coolant System Pressure," Revision 11, requires certain actions to be taken following the automatic initiation of high presssure injec-tion, including in Section 3.3.1, that all ESP aquipment is verified to be in its F.3F position (capable of performing its intended function).
Contrary to the above, durias the period af appearisately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> totil 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> on March 28, 1979, duriss continuing low pressure conditions within the RCS, t.he core Flood Systes was removed from l
i ta E3F poJition (Tendered inoparabla) by closing both can isola-tiran valvse. (U.is portion of the XSF was icact2vsted during a pe siod wha? rudcation of Reactor Coolant Systen pressure was not the irsudiste as al.
This removed from service this safecy feattre during a miod ween it could have been called upoa.. In the course ei tea accident while a.:temptias to depressarias to activata th j
decay hea removal sysses NFC recognissd that it was necessary t3 isolato :ha core flood systee ard encoraged this action. Ela cittrion s?es not apply co i'solacisa dirting this actampt.]
Disszes".y Met::polittu Edison dees.nc bellave that the Core Flood Tank Isolation vsives (CF-VIA and )) were close:f 4:zrig the period from approximately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> until 1300 heurs on March 28,197?. "barefore,thisitemis not a coacompliance. In addition, iJ thy were la fact clesed, no vio-lation of Emergency Procedure 2202-1.3 occurred since the core flood tanks were verified to be in their ESF position after autoestic initia-tion of high pressure injection.
8 A Shift Supervisor testified that he closed the isolation valves. I&E l
)
Transcript of Interview of Shift Supervisor A" (July 11, 1979), p. 8.
However, others in the Control room during this period have stated that i
l the valves in fact were open. I&E Transcript of Interview of Frede.ick, I
i Faust, Scheimann, Zeve and Ross (May 29,1979, Tapes 269 and 270), pp.
l 19, 20, 23,24; I&E Transcript of Interview of Zeve, Schef====, Faust and i
Frederick (June 28,1979, Tape 321), pp. 47. Subsequent events confira
~
that the isolation valves were never shut. At about 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, the reactor coolant system pressure was low enough to permit the discharge
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of some of the core flood tank inventory into the reactor vessel. Since T
chere is no indication that anyone opened the valves after 0500, the
[e clear indication is that they were never closed.
This implies that if Shift Supervisor "A" in fact attempted to close the core flood tank isolation valves, that at:empt did not succeed. The pos-sible explaartics for this hypochesis i.e that the procedures for closing the vahus rnquire that the elacerical braakers (normally locked open) must first be nanually c1csid at the actor control cectars before the 1
valves can be glosed from abs control caon. If Shift Suporvisor "A" 1
tried to operate the valves prior to mant;al closing of the breakers, f-tne valves would act hav.e cicai<4.
Waite the breakers vare closed at g
sa:ue pcist dt.rist the mo-ning of March 28, IIiI Trs.rx:ript of Intervitv of $chtizaca and T.aude.vdich (Matria 30, 1979, Tape 95), p. 21, ch's open status of the valves indicates that Shift Supervisor "A"'s actions, If taken, meurred before the breakers vire clored.
In any case, even if the isolation valves were closed, Emergency Procedure 2202-1.3 would not have been violated. In this procedure, one of the
(.
follow up actions to automatic initiation of engineered safety features is:
" Verify that all E.S.F. equipment is in its EST position, by observing that all equipment status lights Indicate as shown in Table 3-1" Emergency Frocedure 2202-1.2, section 3.1.
One of the indications in Table 5-1 is that the isolation valves be in the open position. However, i
the procedure does not specify that the valves must remain in that under all conditions. At least one other procedure, Operating Procedure 2102-3.2, " Unit cooldown, provides for closing the isolation valves. In T
addition, where plant conditions did not fall within existing procedures, operator judgment must be allowed reasonable discretion.
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Because Metropolitan Edison believes
- hat the core nood tank isolation I
s valves were not closed, no correceive action is appropriace.
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5 4.C Statement of Apparent Noncomoliance:
operating Procedure 2104-6.2, " Emergency Diesels and Auxiliaries," Revi-
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sion 9, establishes the procedures for the control of the emergency diesel generators :
1.
Section 4.10, " Diesel Generator
. Automatic Start Upon Engineered Safety Features Actuation," states in the closing step, 4.10.6, that the unit can be shu:down after the Engineered Safeguards Feature actuation has been cle'ared.
2.
Section 4.6, " Diesel Ceterato.: 1A(13) thutdown to Emergsacy Standby,"
states in the closing step, 4.6.6, to place the diesel generator on st.tzdby in acco-dsnes sich Secent 4.2; and 3.
Section 4.2, when casoleted, establishe. conditions, for automatically starting the diesels upo a actuecian of an Ingineered Safeguards Testure (IS7) including requiremen:s to place the *1w gency Standby /
)(sintanance Exerciee swi:ch in the 5sergency Standby positios and resatting the fuel ra.ks.
f.
Conersry to the above, as abcut 0439 h<rura os March 28, 1979, bo Q the 1A and 13 diwsel gacera or fuel ' racks were sacually ripped, thereby prevent-isg ar. automatic sea.it of the diesel generators upon 157 actuation acJ manual secre frm the centrol until 0749 hours0.00867 days <br />0.208 hours <br />0.00124 weeks <br />2.849945e-4 months <br />.
Discussion:
'llhs shutdovu of the emergency diesel gecerecors at 0430 cy manually trip-ping the fuel racks and the failure to reset the diesels for automatic start violated Operating Procedure 2104-6.2, " Emergency Diesels and Auxiliaries".
[.
At 0402 hours0.00465 days <br />0.112 hours <br />6.646825e-4 weeks <br />1.52961e-4 months <br />, the two emergency diesel generators started with automatic engineered safeguards actuation. Because the diesels were running un-loaded (offsite prover continued to be available), their shutdown was appropriate. See Operating Procedure 2104-4.2, section 2.1.1.4.
A Control Room Operator dispatched an Auriliary Operator to shut down the i
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diesels by manually tripping the fuel racks, the only method by which the diesels can be shut down after an automatic diesel start on engineered 1
i safeguards actuation.
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Since off-site power remained available, the unit was no longer at 4
1 power, further engineered safeguards actions were possible, and additional 4
manual tripping of the fuel racks would require dispatching personnel into areas with airborne contamination, the Control Roca Operator con-complaced that the diesels, once tripped, would be reset in the neintenance azercise positlen. This would prevent automatic set.rting on engiseated safeguards ac.ttsation but alloa diesel start from the control roca in the i
everf. of that off> ice i,oasr ssre lest. RovaveT, the Control Ro.:n Opacator die cet give spar.ific instr **cticas to the Auxiliary tirpesto-r and the i
Auxiliary operator dit art repor: his specific accioca Lack ta the l
Conersl Kocw Operator. As a tesalt, the fuel racks were cod reset and the f
diesels could only have be n restatted by dispatchlag an opot-vtor to the t'uel racks. Only at 0930 bcurs was this situation recessized reset and the fuel racka reret.
,l.'ter 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />, the diesels coet have been 5
sacually started from -he control tooet had there been a loss of off-site power.
Corrective Acticut
's In order to assure that the situation which existed from 0430 to 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br /> is not repeated, the operator accelerated retraining program is addressing the importance of procedural compliance and the Unit 2 licensed operator requalification program will specifically emphasize both the procedural compliance issue and the diesel generator procedures.
Because Operating Procedure 2104-6.2 does not contemplate keeping the i
emergency diesel generators in the maintenance exercise position under the conditions as they existed after 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br /> on March 28, 1979,
~
t Metropolitan-Edison considering amending the procedure to allow the main-tenance exercise position when the reactor is in hot standby condition e.-~ -,-.
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2 and when there are good reasons (such as occupational exposure) to avoid I.
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j Emergency Procedure 2202-2.2 " Loss of Feedvater," Revision 3, requires in Section 2.3.2.d that the operator adjust feed flow to control steam generator levels at 30 inches.
Contrary to the above, from a7Froximately 0532 hours0.00616 days <br />0.148 hours <br />8.796296e-4 weeks <br />2.02426e-4 months <br /> until 0343 hours0.00397 days <br />0.0953 hours <br />5.671296e-4 weeks <br />1.305115e-4 months <br />.
the level in A steam generator decreased to 10 inches (the mini:sua level indication) while the A staan, genera:or level vaa bains est.crslied annually.
Discussion:
h :ropclitan Edison does act hellest that 2a lee 4 is a rencampliacce.
I During the perir.d in guesti;:n, Essertpa:y Procah.rt 2207-2 4 (whi:*t rs=
i
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l quires taintainies steam gsnvirator laval at 30 iuckas us.da: serrain con-dicions) did not apply.
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h the perict' just prior c:s OS?2 here et Marr.h *3,1979, the stems gen-erstor were being conavilaf nanually siree autcrsr:ic ccde had behaved erratically in the eatly stages of the accident. I&E Trwescript of Inter-
-iaw cf Frederick, Faust, Scheimaa4, Zeve sad Ross (May LT,1379, fapes 269 and 270), pp. 42-43. With feedvater being supplied through the emer-ency feedwater system, levels were being kept at essentially a steady At 0514 hours0.00595 days <br />0.143 hours <br />8.498677e-4 weeks <br />1.95577e-4 months <br />, reactor coolant pumps 13 and 23 were tripped due' j
state.
to vibration levels. With the loss of primary side flow in the 3 steam i
generator, heat transfer was lost. 2his caused h rapid reduction in i
staan generator pressure and a pressure differential between the A and 3 l
staan generators. As a result, emergency feedwater flow preferentially went to steam generator 3, and the A steam generator level dropped rapidly.
2he operator took corrective action to restore level slowly.
2he claimed concompliance states that section 2.3 2.d. of Emergency
}
Procedure 2202-2.2A, " Loss of Main Feedwater Flow to 3cch OTSC's",
was violated. 2his manual action procedure did not apply during the period in question. Section 2.3 2.4 states:
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~
"If reactor coolant temperature and pressure cannot be maintained, j
or if feedwater flow cannot be restored, or if the reactor trips, 3
start the emergency feed pumps and ' maintain 30 inches in the steam generators (S/U cange indication)."
However, this provision only applies to " Loss of feedwater due to valves closing". Settion 2.3.2.
That condition aid not occur during the acci-b dent. The othat manual action alternative, "If loss of feedvater is due to loss of Loch feed pumps" (sectinn 2.3.1), was applicable in the early Jeages of the act!,dete. E:vever, the applicable step in section 2.3.1 1
(section 2.E.1.d) is baeid upon auten.atic operacios tf the emergency faa.fwete: valves:
c I
" Verify emergency fc ed sacer valyJe (17-IIA (5)) are in succmatic l
4::d costrol?.ing (7:EG 1evel c 3G Ir.che (S/U range ;:ndication)."
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There was no procedura it, this section 2.3.1 geverning ususi operation.
t As of r.he 0532-0543 hour tini period, the acte approprist$ peccedure was Emergency Procedure 2202-2.25, " Loss of Main Taet.aatar Flow to Ose OfSC", since the condition which led to the 1e.w level f.a A steam generator was the loss of flow to it alone. Under this procedure, which is the only one to anticipate a stema generator boiling dry, the appropriate action is to " establish feed flow using emergency feedvater pump through the emergency feed valves EF-7-11A(B) verv slowly (2 Inches per minute)." Emergency Procedure 2202-2.23, section 3.2 Note. (original emphasis). 'The operator e
complied with this procedure.
Corrective Action:
i Although no noncompliance is involved, Metropolitan Edison's revised and l
augmented training programs will stress the importance of careful control I
of plant functions in the manual mode.
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- 4.E ~ Statement of Apoarent Noncompliance:
k*
4 Three Mile Island Nuclear Station, Administrative Procedure 1004, "Three Mile Island Emergency Plan 1004," Revision 2, dated February 15, 1978:
?
1.
Requires in Section 2.1, that the " Station Superintendent / Senior Unit Superin~t endent, Unit Supt./ Shift Supervisor / Unit Supt.- Technical Support in the Control Room will, after reviewing the emergency condicione, classify the emergency as ona of the following:
l "a.
Persennel or Local Enertency, l
"b.
3 Lee Emergency, and "c.
Caseral EmergeECy
+
He will make this classifiution seemdiag te ths m.di: lea af Cable 1 of whis plan, and initiate.tctiou carordi:.g to the Ewen.r :7 Plan Implazanting Procedures, and acceeding to his oma best jud;wat;"
and
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2.
States in Tabis 1 of Section 2.1 that a Site Evergency exists shan l'
there is a etector building high range gaar.a acsiter alert alarm (Condition No. e).
Cocercry eo the above:
1 1.
Adequet written ;rrocedvras were not establiand. rad upleinmated in i
that Section 2.1 of Procedure *004 for imptemestir.g the Instgency Plan lacked sufficient specificity and fallad u result is a fite l
Emergency being declared at approximately 04JO on March 28, 1979, even though primary system pressure had decreased to the point where safety injection was automatically initiated and a reactor building simp high level alara existed; and Discussion:
The claimed noncompliance is that section 2.1 of Administrative Procedure i
\\
1004 was not adequate, rather than that Metropolitan Edison failed to comply 4
with its procedures. While Metropolitan Edison thus believes that Administra-tive Procedure 1004, rev. 2, section 2.1 was not in noncompliance, Metropolitan
{
Edison recognizes the need for greater specificity in its emergency procedures.
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2he Notice of violation does not assert that Metropolitan Edison failed to comply with its procedures by no't declaring a Site Energency at 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />.
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The on y potentially applicable condition was 5
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" Loss of primary coolant pressure, coincident with high reactor building pressure and/or high reactor building
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sump level."
Administrative Procedure 1004, section 2.1, Table 1 (Site Emergency, Con-dition (c)). Unless there were a " loss of primary coolmat pressure',
condicion (c) wos1d not be operative. As set forti iu 1RDZG 0$00, p. II-1.-1, primary coolsd pretsure had aropped from 2435 psig to 1173 peig, a
- avel below the rsac or ceclaut low press res trip setpoine and the set-point for amargency core cooling syste.e initiation. Ibt 1hift Supervisor determined that prirtsry systea ptusure, dila it had decrsased, had stabilized. De pressure was several hwdred psi above the level Wich t
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(
wdd ut:ur durig a large brsh loss of wtant accident.
In the absesce r
cf a definition af " loss of primary coolan presntre", tire Shift Super-visar inter;tetid the phrase as relating to a. LOCA or other accidents t
(such as Main Sesam Line gresk) which cwid give siviilst symptoms.
Ab-sent a " loss of primary. coolant pressurs' as interpreted by the Shift Supervisor, condition (c) was not satisfied and a Site Emergency was not declared.
Corrective Action:
Metropolitan Edison has totally revised its emergency procedures and has sdaitted them to the Commission. The revised procedures greatly expand the categories and conditions for declaration of emergencies. They have also been made auch more specific in order to avoid ambiguities and uncertainty to L
the greatest extent possible. The revised procedures will be covered in the I
Operator Acclarated 2etraining Program and vill be tested in the emergency drill program.
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1 2.
A wite energency was not declared at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br /> on March 28, 1979, at j
which time Condition "e" of Three Mile Island Emergency Plan 1004 had 5
occurred.
y Discussion:
e Condition (e) for declaring a Site Emergency is " Reactor building high range gasuna monitor alert alarm". Administrative Procedure 2004, section 2.1, Tahta 1.
Se claimed noncompliance is that this condicios occurred at 0635 buts on March 18, 1379. Macropolitan M.sen does not 'oeliave that i
this itca ie
- 4. noncompliancs. Its best infermat, son la that Gudi:ian fa) occurred at 0641 hours0.00742 days <br />0.178 hours <br />0.00106 weeks <br />2.439005e-4 months <br /> and that the Site Emergency us dee.lared sous seven sisucas later. Under the cir:nastancas, tris sapanca was cor areason-l sbis ar.4 did not violate ury procsdwe.
(
the apparent basis for sa 0635' hour figura is a revies by a Inspectica is Caforwaett investigaur of the strip chart tren *he chet recorde (HP-01-1803 for the does monitcr (Ep-R-214). S e krettiCator'a enton64:uc-i tio!. of the chart is preset.ted in hUREG 0600 as Figuria *sI-3-3 (p. II-3-75).
This recenstruction would have the alert alarm (25 a:t/hr) occur at C635 hourt.
Me exact time at which the alert alarm occurred cannot be determined fra the multiprint strip chart. The time annotations on the chart are not Nor can the specified chart speed of eight inchee per hour be used exact.
to arrive at a precise time since that speed varies with the radius of the chart roll, a radius which changes over time. By attempting to duplicate the NUREG 0600 reconstruction, it appears that the 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br /> timing of
.I the alert alarm was based on an assumption that the actual chart speed that time was eight inches per hour.
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1 A more. accurate method of reconstructing the time of the alert alars is 4
ts locate on the strip chart two events for which the timing is precisely 5
known and to interpolate the timing of the event in question. In this case, the reactor trip and the containment done monitor reading of 8 1/ hour (causing the declaration of a general 1sergency) can be both precisely located 6u -he strip chart and identified with a specific time, i.e. reestor trip ac 0s00:37 (17145 0600, Appendix I-A) and 8 R/hr at 0724 (%7RIC 0600, p. II 2-6).
17 iaterpolation, the time at which the dose monitor reaches the alert alare level (25 muhr) wcvid be at shout OM3.
b best inimatics on the rius et which the Site E:=cegsacy was declared
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Indicates L' sac : hip occaned as 0650 bent:r. The Shift Supervia.cr has con-i I
d entently identified 0650 A.: the time of the declaration. IEZ: *'rcuscript of March 30, IM9 Interrisw.2f V. E. Zeve, p. 22; I&E'Iranscript cf April 23, 1373 Isterviezw ef U. E Zeve, p. 57; I&E Transcript of Intenise of W. R. Z.sws (undtted, Espa 173), p. 7.
h time of the declaration was re-corded on a blackboard in the Unit 2 contral roce and subsequently trans-cribed for historical purposes (IEE Esquest item 116.1).
This too shows,
a time of 0650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br /> for the declaration. The only other time record was
(
a notation in the Unit I control room of 0655 hours0.00758 days <br />0.182 hours <br />0.00108 weeks <br />2.492275e-4 months <br /> based on the annuncia-tion over the plant page system. I&E laterviewers also used 0650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br /> i
for the timing of the declaration. I&E Transcript of June 28, 1979 Group Interview (Tape 319), p. 37 (question by Dale Donaldson).
I The Site Emergency was declared when many of the radiation monitors began
~
to alarm and alert together, when "the Christmas tree [went] off" I&E Transcript of June 28, 1979 Group Interview (Tape 319), p. 54.
These
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m alarms and alerts were recollected as occurring between 0645 and 0650 5
hours.
H i President's Consission Interview of Brian Mehler (May 10, h
1979, Tape 5), Transcript, p. 15-16.
The time interval of seven minutes or less from the alert alarm level of the done monitor to the declaration of the Site Emergency was not unreasonable under the circumstances surrounding the accident. Administrative Procedure 1004 does not require that the declaration occur simultaneously with the occurrence of the particular condition. Instead it provides that "after reviewing the amergency conditions", the Shift Supervisor will classify the emergency "according to the condiziens in Table 1 of this plan... and according to his own best judgment." Administrative Procedure 1004, section 2.1.
(
There is no indication that the Shift Supervisor failed to comply with this procedure.
Corrective Action:
As note'd above, Metropolitan Edison has dramatically tevised its amargency procedures and will implement an expanded training and drill program to assure that the operating organization is thoroughly familiar with these
~
procedures and can effectively carry them out.
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Additionally, procedures for the monitoring team response, including j
1 operating procedures for the monitoring equipment, were immediately avail-able to the monitoring tems members. Although the 1978 training, require-ment was not completely fulfilled, the response of the monitoring teams wee sufficient to properly implement the emergency plan and provide radiation monitoring information that did ensure proper assessment of the effect of the incident on the health and safety of the general public.
Due to administrative problems, none of the individuals specifically des-ignated as tapair Party Teas members recalved in 1978 the complete training program as required by the emergency plan. However, all of the individuals are radiation workers and have recsived extensive ergining and have experi-('
ance in radiation protection relative to their normal repair functions. It should be noted that during the lacident on March 28, if 79 as well as the days following, many functions, both operations and saintenance, were per-formed that could be classified as emergency Repair Party functions. As the need for any task was detetuined, individuals most qualified to perform the task were assigned. Qualifications included specific job knowledge, familiarity with the systems, radiation protection knowledge and previous'.
exposure. The intent was also to equalise as much as possible the exposure I
over all qualified personnel. As a result, the individual selected for a
- a particular task was the most qualified individual available although that individual asy not have completed the entire training program.
1 Corrective Action:
A revision to the Radiation Energency Plan has placed significan* emphasis on in plant Health Physica during an accident. Procedures v1.11 be devel-
.~.
oped to clearly define the training requirements of all personnel.
In-plant Health Physics will become a major factor in the training of all
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4.F statement of Apsarue Moncompliance:
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Three Mile Island Nuclear Station Naalth Physics Procedure 1670.9,
" Emergency Training and Emergency Drills," Revision 4, dated I
January 16, 1978:
1.
Identifies in see.sion 3.1, the on-site emergency job categories and requires that training programs for these categories will be conducted om an annual (calendar year) basis; and 2.
Describes in section 3.1.1 through 3.1.9, the training program for all on-site emergency job categories.
Contrary to the above during calendar year 1978, not all individuals having emergency respo,nsibilities were trained in that two Emergency Directors, one Ac ident Assessment individual, eight Radiological Monitoring Team Members, and 37 Repair Party Team Members had not received the specified training. In addition on March 28, 1979, during as emergency, ac least four individuals who were assigned as required members of a Radiological Monitoring Team and seven indivi-duals who were assigned as required members of a Repair Party Team
(,
performed emergency duties for which they were not trained.
Discussion:
Metropolitan Edison agrees that the administration of the emergency training requirements was not complete. Two potential Emergency Directors (both shift foremen), and one potential Accident Assessment individual (a Shift supervisor)'did not receive the training during the year 1978.
It is important to note however that a total of 27 potential Emergency Directors and 29 potential Accident Assessment personnel did receive
(
the training and that each shift complement did at all times have personnel on site who received training in each category. On March 28, 1979, none e
of the three individuals mentioned acted in a tapacity for which he did not have up-to-date, doctamented training.
i of the eight Radiological Monitoring Team members, two did not have documen-ted training in 1978 and six are believed to have received training which A
was not totally in accordance with the training procedure requiremencs.
Training which was not received, due to an administration error, was in areas that are similar to the routine monitoring functions performed
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by Health Physics technicians with which these individuals are familar.
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emergew y personnel. Training in the implementation of the procedures I
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vill be implemented prior to start up of Unit 1.
Mditionally management i
5 has increased the emphasis on all aspects of radiation protection during I
daily operation. Additicnal training of operations and maintenance personnel as well as radiation protection personnel will provide increased
, asst.rsace that response during emergencies will be adequate.
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- , 4.G Statement of App *47%ne Noncompliance:
2 Station Administrative Procedure 1002, " Rules for the Protection of I
l Employees Working on Electrical and llechanical Apparatus," Revision I
14, requires in Section 4.3, 4.4 and 4.5 that on restoration of equip-to service, removed tags will have all required information entered ment thereon and then be suitably stored, and that the shift foreman shall approve equipment operation by signing the original tagging application.
Additionally, Station Corrective Maintenance Procedure 1407-1, Revision l
1 0, specifies in Section 3.0, " Job Ticket (Work Request) Flow," the step-by-step process for initiating, processing, obtaining approvals and ultimate filing of the " Job Package" dich will include, among other things, documentation of corrective action taken (resolution description and certification of satisfactory post maintenance testing) and Station Preventative Maintenance Procedure I-2, " Dielectric Check of Insulation, Motors and Cables," specifies how to make the seasurements and contains data sheets for recording the values measured.
Contrary to the above, when inspected on June 20, 1979, the tagging application could not be found for maintenance performed in January, 1979, on Emergency Feedwater isolation valves (EF-712A,123, 32A, 325, 33A, and 333). No suitable documentation to determine whether the
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maintenance work had been completed, tags removed, acceptance criteria met, or valves approved for operation could be found. The TMI-2 main-tenance log lists this work request as being in an open status as of June 20, 1979.
Discussion:
Metropolitan Edison agrees that this was a violation of procedures in that a tagging application could not be found for this maintenance.
As of July 10, 1979 the TMI-2 maintenance los lists this work request se being in a closed status and associated docuentation is flied in the Maintenance Department.
4 The subject Job Ticket we's closed out by the operations Department on January 25, 1979, signed off by Quality Control on June 12, 1979, and signed by the Supervisor of Maintenance on July 10.
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Due to the delay in closing out this work request, a retrospective re-view was conducted to determine what assurances there were that the
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valves were operable at the time between the January maintenance and March 28th. The electrical supervisor for the company which performed
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acceptable but there are no written records to support this statement.
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As part of the unit startup on January 30th, the valves were cycled with-out difficulty in accordance with Surveillance Procedure 2303-M14(A/3).
Bis was done prior to exceeding a main steam pressure of 850 psi.
Broughout the rensining period the normal surveillance procedures were accomplished with satisfactory results.
De January work request was c1'osed on July 10, 1979. De motors for EF-Y12A/B were satisfactorily tested la accordance with the procedure prescribed in the work request C2555.
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Ievision to station Adminstrative Procedure 1002 dated July 25, 1979 deletes the requirements for storage of tags because it is not felt to be a necessary feature of records maintenance.
Corrective Action:
Me records management ' function at Bree Mile Island is currently being ez-panded to include a significantly larger staff of professional and clerical personnel dedicated to the collection and retention of records. Bis fund--
i tion will be led by a department level manager supported by appropriate l
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supervisory personnel directing separate groups for Unit I and Unit 2.
J Among the improvement in records retention will be computer-sided filing, Improved storage and control, and advanced reproduction methods.
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l De emphasis on records retention and control will be complemented by gen-i
'ersi training that emphasizes the necessity to properly complete each as-g pect of any work, including the documentation. Finally the staff and 1
supervision of the on-site Quality Assurance Department has taken on an expanded role to assure that each area of performance prescribed by sea-
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clon procedures and practices are followed.
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'De overall organizational structure for accomplishing these changes will
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b'e subject to thorough review by the NRC as part of the review of Unit 1 1
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prior to its restart. De Unit 2 portion of these changes will be sep-e'.
arately submitted to the NRC and will reflect the special nature of ac-
. civities in that unit.
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seatmedne of Apparent Noneouroliance:
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Technical Specification 6.8, " Procedures," requires in Section 6.8.2 that 3,
changes to procedures which laplement the Emergency Plan shall be reviewed by the Plant Operati8ns Review Committee and approved by the Unit Super-intendent prior to implementation.
Contrary to the above, a change to Station Health Physics Procedure 1670.7, "Zaergency Assembly, Accountability and Evaluation," was made without the required review and approval. An additional assembly area was designated and the method used to perform accountability was modified by a memorandse dated October 13, 1978, from the Radiation Protection Supervisor to all departments. As a result, on March 28, 1979, in response to an emergency, some licensee personnel followed the approved procedure while others followed the guidance la the October 13, 1978 memorandus, creating some confusion and delaying prompt attainment of full accountability.
Discussion:
On October 13, 1978, the Radiation Protection Supervisor issued a memor-(
andum, " Accountability During Radiation Emergency", which added an addi-clonal assembly point to 'those already specified in Station Health Physius Procedure 1670.7, " Emergency Assembly, Accountability and Evacuation".
This memorandue did not receive formal review by'the Plant Operations Review Committee or approval by the Unit superintendent prior to its in-plementatloa. This was a failure to comply with Technical Specification i
6.8.2, which requires such review and approval for changes to such pro-cedures.
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Notwithstanding the lack of review and approval, the change made by the memorandse did not, to the best of Metropolitan Edison's knowledge, delay prompt attainment of personnel accountability, or cause confusion.
The change made by the October 13 memorandum was to have non-essential I
l personnel outside the security fence report to the North Warehouse. Prior to this change, these people were required to assemble in the North Audi-4 corium, a location within the security fence. Requiring these people to
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pass through security to gat to an assembly point was felt to be an un-necessary, time-consuming step.
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i To11owing the October 13 samorandum, Metropolitan Edison conducted seven radiation enertency drills with 'a scope equivalent to a Site / General Emergency. See'NUREC-0600, pp. II-1,-17, 18. The requirements of the October 13 semorands were carried out in each of these drills. The final 1978 drill, on November 8, 1978, was abserved by NRC. As stated in the Combined Inspection Report Nos. 50-289/78-21 and 50-320/78-34, NRC made " detailed observations" of a amber of emergency drill activities, Including accountability. The inspectors' determinations included the following findings:
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1.
"The licensee's response was generally in accordance with existing procedures"., and 2.
"The response was coordinated, orderly, t
and eimely."
No items of noncompliance were observed.
On March 28, 1978, accountability was achieved within approximately 1-1/2 hours. While this was not as prompt as in some of the drills, it was in fact better than in others. Given the facts that a real accident 4
was in progress, that personnel recall and shift changes were in progress, that plant personnel were facing severe operational demands, and that there are no standards for timeliness, full accountability vichin 1-1/2 hours was timely.
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As to the assertion that the October 13 memorandum introduced any additional
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confusion into the accountability procedure, Metropolitan Edison is aware i
of no information which would support this claim. NUREG-0600 contains no
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" supporting references. De fact that seven drills in late 1978 implemented j
the revised assembly points makes such confusion auch less likely.
3 And interviews of members of the security protection force indicace their awareness that the North Warehouse was a designated assembly point.
See I&E Transcript of Interview of Mr'. William J. Busansky and James y,
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5tacey (May 7, 1979, Tape 161), pp. 5, 8.
Corrective Action:
Accountability requirements are now incorporated in the Energency Plan and its supporting procedures. These have been reviewed by the Plant Operating Review Counittee and approved by the Unit Superintendent., They are also being submitted as part of the information to be reviewed by letc prior to restart of Unit 1.' Training in the use of the procedures t
is a part of the on going and accelerated training prograns in Units 1 and 2.
Additional training for supervisory personnel will be undertaken to assure adherence to the review and approval requirements of the Tech-nical Specifications.
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3:.tement of Anoarent Noncompliance:
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Invironmental Technical Specification 5.7 requires that detailed written I
procedures for instrument calibration be prepared and followed.
t Three Mile Island Nuclear Station Surveillance Procedure 1302-5.24, Revision 3, dated December 19, 1974, specifies the method of calibration and requires that it be performed annually.
Contrary to the above, as of March 29, 1979, eight environmental samplers had not been calibrated since 1974.
Discussion:
Although noncompliance with Surveillance Procedure 1302-5.24 was tech-nically not a violation of any Unit 2 procedures, Metropolitan Edison acknowledges that the procedures should have complied with or withdrawn.
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Environmental Technical Specification 5.5 requires that detailed written procedures be prepared and followed to implement the environmental tech-nical specifications and that these procedures include " instrument cali-beation". A Unit 1, procedure, Surveillance Procedure 1302-5.24
(" Environmental Monitors Calibrator") set forth a procedure to calibrate the continuous air monitors located around the site and required that calibration be performed annually. As noted in NUR.IC 0600, p. II-1-45, there was no Unit 2 procedure equivalent to Surveillance Procedure 1302-l 3.24.
It was however listed as an active procedure for Unit 1.
4 The Surveillagsco Procedure had been prepared in 1974 to address a draft Technical Specification. Although that draft Technical Specification included in the final Technical Specifications, the ' procedure was not was not deleted. The procedure was not, however, followed because of h
difficulties in obtaining useful results and because UCI, the vendor C
i of the monitors, had orally informed Met-Ed that the monitors (par-cicularly their floumeters) were considered to be primary standards and not subject to accurate calibration.
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It should be noted that I&E was aware that these monitors were not being 7
m routinely calibrated. Combined Inspection R port 50-289/78-08 and
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50-320/78-16 (May 31,1978), p.10, included the following discussion:
Discussions with the licensee indicated that at the present time, air sampling units are on a limited pre-ventive maintenance schedule. Ilhes a sampler fails, plant personnel either replace or repair the failed unit. He timeliness of this action is dependent, however, on plant operational status at the time of failure. Me licensee stated that at the present cias the dry gas asters employed with the air samp-1ers are not routinely calibrated. De licensee stated that the preventive maintenance program for the air sampling syrtens and the calibration of the dry gas meters would be re-evaluated. He inspector stated that pending completion of these evaluations, this item is cosidered unresolved.
(289/78-08-03; 320/78-16-02)
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Mus, I&E considered the absence of routine calibration of these monitors to be an unresolved, item rather than a noncompliance.
Corrective Action:
He NEC Staff in October 1979 published a draft Regulatory Guide,
" Calibration and Error Limits of Air Sampling Instruments for Total Volume of Air Sampled" (Division 8, Task OR 905-4). De draft Regula-cory Guide identifies methods acceptable to the Staf for calibrating l
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air sampling instruments such as the continuous air monitors covered by Surveillance Procedure 1302-5.24. Metropolitan Edison will review l
- the procedures described in the draft Ragulatory Guide against the de-l i
stan of the continuous air monitors. If calibration can be performed i
i as described in the draft Regulatory Guide and if it produces accurate, I
reproducible results, Metropolitan Edison will modify the Surveillance Procedure accordingly and include this calibration as part of the station surveillance program. If such calibration cannot be performed, the procedure will be deleted.
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This'ites also points out the need to thoroughly review all Station 7
procedures. If there are other procedures which need not exist, they k
will be removed from the active procedure file so that there can be no confusion as to witich procedures are to be inplemented. At the same time, the ongoing and accelerated training program will emphasize the requirement to comply with all written procedures as well as describe the adminstration controls for addition, modification or deletion of procedures. The Operational Quality Assurance Plan is also being re-vised to improve the effici'ncy and validity of that program in support e
of its role in assuring compliance with station and unit procedures.
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reacement of Apoarent Noncomo11ance:
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Technical Specification 6.2, " Organization," states in Section 6.2.1 and I
6.2.2 that the unit organization and the organization of the corporate technical support staff shall be as shown on Figure 6.2-1.
Contrary to the above, on March 28, 1979, the organization of the unit and corporate technical support staff was different from that specified i
in Figure 4.2-1 in that:
1 A.
A position titled, " Superintendent of Administration and Technical Support" was added to the organization on September 18, 1978 and filled on March 1,1979, such that the " Supervisor, Radiation Protec-tion and Chemistry," reported to this new position rather than directly to the " Station Superintendent / Senior Unit Superintendent;"
and,,
3.
Bare were two " Supervisor of Maintenance" positions, one for each
- unit, rather than one; and C.
A position titled " Superintendent of Maintenance" had been added such that the " Supervisors of Maintenance" report to this new position g
rather than directly to the " Station Superintendent (Station Manager)/
Senior Unit Superintendent;" and D.
De position of " Chemical Supervisor" had been vacant since the j
issuance of the Technical Specifications.
On March 28, 1979 through March 30, 1979, the above organizational discrepancies decreased the effectiveness of the licensee's response to the accident.
l Discussion:
1 Metropolitan Edison agrees the organization in effect at the time of the iaccident was different than that specified h Figure 6.2-1 of the Unit 2 Technical Specifications. However, the impact of these changes had no adverse effect on Met opolitan Edison's performance fo'11owing the ac-cident. He following comments should be considered in evaluating chis apparent noncompliance, j
A.
Although a Superintendent of Administration and Technical Support was added to the organization prior to approval of a Technical Specifi-y cation change, this position was not part of the emerger.cy organ-isation put in place in response to the declaration of the site and E
general emergencies. He Emergency Plan organization was manned as
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i led the Radiological Assessment Effort in the Emergency Control Cancer (the Unic 2 Control Room), reporting directly to the Zmerg-ency Director (Station Manager). The Superintendent of Administra-tion and Technical Support did not have an latervening role.
5/C.Although a position of Superintendent of Maintenance was added to the organization prior to approval o'f a Technical Specification change, the specific individual who filled this positle a had pre-viously held the position of Supervisor of Maintenanc.e for the
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station. Prior to the accident he was promoted to ch new posi-
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tion of Superintendent of Maintenance and two managere it positions were created below him as Supervisor of Maintenance for each unit.
H is change provided improved management attention and capability of the on-site maintenance department structure by not leaving the first level maintenance supervisory attention split between units, and by introducing a middle management position below the Station Manager level who could focus on the station needs and resources In the maintenance area.
As described in NUREG 0600 (pg. II-2-1i) following the emergency declarations, "a Repair Party composed of six maintenance shift workers was formed at the ECS under the control of Maintenance Foreman B. (Inc. 187). A second Repair Party, composed primarily of daylight instrument and control personnel, was formed in the Unit 2 control room under the direction of the Superintendent of Maintenance and the Unit 2 Supervisor of Maintenance. The Superintendent of
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Maintenance and Unit 2 Supervisor of Maintenance were aware that the assigned location for either of them during an emergency was the ECS where they would act as the Repair Party leader. However, to ensure prompt availability of their expertise, and since a Repair 1
Party was already formed at the ECS under the direction of a main-t tenance foreman, they decided to remain in the control room (Inc.
120). Bis decision was reinforced by the Station Manager. On asstsaing the position of Emergency Director, he announced that the Superintendent of Maintenance would be the one in the control room to be in charge of emergency repair functions (Ref. 72 Inc. 120).
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Some time later, the Superintendent of Maintenance directed the Repair Party at the ECS to move to the Unit 2 control room. By 0800, C
all Repair Party personnel were assembled in the Unit 2 control room, I
separated from the ECS."
This organizational alignment placed the control of the Repair Party Teams with the senior maintenance personnel who, by their location and proximity to up-to-date operational and radiological inforastion, could best control and coordinate their activities. Had the organ-Laational-changes prior to the accident not taken place, the response would have been the same.
D.
Although the position of Chemical Supervisor had been vacant, this position was being filled in each unit by a Senior Chemistry Foreman.
Each of these individuals were fully qualified as required by ANSI /
ANS 3.1-1978 and in fact e::ceeded these minimma requirements. Al-though not specifically designated as supervisors, having two quali-fled individuals assigned, one to each unit, provided improved super-vision and control over each unit's chemistry programs. There 'is no evidence that the lack of a specific individual designated as Chemistry Supervisor resulted in inappropriate actions.
the organizational changes identified above were discuased with the NRC at the time of implementation on March 5, 1979. On the same day the office of Inspection and Enforcement was also notified in writing of the changes to be made. It was, agreed verbally that a Technical Specification change would be submitted. These changes had been pre-g pared but not yet submitted at the time 'of the accident.
It has been CPU's experience in the past that organizational changes to Technical Specifications may take as long as 6 months from the time of initial discussion with the NRC to final implementation. A recent E
coutemplated organizational change at Oyster Creek took 4 months from the
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,I time of submittal until final approval by the NRC. The final changes 5
E were essentially the same as these discussed initially with the NRC.
3 The Company believes that it is not in the best interests of public health and safety to delay changes that will strengthen the organiza-tion. Mechanisms similar to those la place for changes to the QA and security plans need to be developed for implementing changes in the organizational structure.
Corrective Action:
Me organizational structure for both Units 1 and 2 has been re-documented with the NRC and we are attempting to provide for some flexibility to trest necessary minor organizational chanties within these specifications.
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The Unit 2 Technical Specifications are under final review and should be issued in the near future by the NRC. The Unit 1 organization has been re-defined in the TMI-I Restart Report Amendment 6, submitted November 28, 1979. This' organisation will be incorporated in the Technical Specifications prior to restart.
l Major changes to address the organizationa.'. deficiencies noted through t
l the many post-accident lavestigations have been discussed with the NRC f
ed implemented with their agreement.
The guergency Plan has also been modified to improve the emergency organ-1 isation. The revised plan has been documented in the TMI-1 Restart Report l
l ad is currently under review by the NRC. Drills will be conducted prior i
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to restart of Unit I to test the effectiveness of the revised emergency plan.
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Statement of Apoarent Noncompliance:
!L Technical Specification 6.4 " Training," requires that a retraining and 2
replacement training program for the unit staff be maintained that meets or exceeds the requirements and recommendations of Section 5.5 of ANSI N18.1-1971.
Contrary to the above, as of March 28, 1979, a retraining program meeting
. or exceeding ANSI N18.1-19,71 recommendations had not been maintained for members of the radiation protection and chemistry staff in that only 2 of
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the 10 topics recommended were included in the program.
i Discussion:
Metropolitan gdison believes that the retraining program established for I
the radiation protection and chemistry staff met the requirements of Technical Specification 6.4 and Section 5.5 of ANSI N18.1-1971. Metro-(
politan gdison therefore disagrees that this item is a noncompliance.
Technical Specification 6.4 commits Metropolitan gdison to a retraining program for the ucit staff which " meets or exceeds the requirements and recommendations of Section 5.5 of ANSI E18.1-1971' and Appendiz 'A' of 10 CFR Part 55."
(The latter regulacios applies only to operators). Section
.I 5.5 of ANSI N18.1-1971 requires a training program "which maintains the proficiency of the operatina orzanisation......" Section 5.5.1 of the ANGI standard s'tates:
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"A retraining program should include:
a 1.
Plant startup and shutdown procedures; 2.
Normal plant operating conditions and procedures; 3.
operational limitations, precautions, and set points; 4.
guergency plans and security procedures; 5.
Abnormal operating procedures; 4.
Emergency shutdown systems; j
7.
Chantes in equipment and operating procedures; 8.
General safety, first aid, and radiation safety; 9.
Alarms and instrumentation signals; and
- 10. operation of selected auxiliary systems important to E
overall plant safety."
t The charged violation asstanes that g members of the operating organi-aation must have retraining in each of these ten areas. There is no
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5 basis in the language of Section 5.5 or elsewhere to our knowledge for I
I this assumption. The retraining program described in Section 5.5.1 is for the entire " operating organization," not for each job category. Many of these areas are relevant to plant operators but not to radiation / chemistry technicians. M e Station procedures clearly reflect this understanding.
See Besith Physics Procedure 1690, described in lRTgIC-0600 at p. II-I-16.
mus, the retraining program for radiation / chemistry technicians did not include items 1-3, 5-7 and 9-10.
ne interpretation of Section 5.5.1 implied by the charged violation vould also be inconsistent with the general training provisions of i
ANSI N18.1-1971. Sus, section 5.1 provides in part that the training program "shall be formulated to provide the required training based on individual employee esperience and intended position." (emphasis added) If retraining in particular subject areas is not appropriate for the " intended position", then it would be illogical to interprec section 5.5.1 to require such training.
It should also be noted that the training program for radiation / chemistry
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technicians as described in the FSAE (13.2.1.5) is consistent with Mec-gd's s
position. Furthermore, I&E combined Inspection Report 50-289/78-09 and 4
50-320/78-18 (dated May 25, 1978) reviewed Met-gd's general employee train-ing and craft and technician training.,(ne latter was specifically ad-dressed to Metropolitan gdison's " program for training and retraining of craft and technician personnel who are available for assignment to either unit.") No items of non-compliance were identified.
f Corrective Action:
a c.
Although Metropolitan Edisoit does noc believe that its retraining progras for radiation protection and chemistry staff was in non-compliance with M !.4'207am._j l --*_
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Section 3.3 of ANSI N18.1-1971, Metropolitan Edison recognizes the need
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for more and better training of these and other components of the operating
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organization. The upgraded training and retraining program is especially necessary in light of the particular 1s. vel of challenge associated with Unit 2.
A improved and expanded training program to address these -
concerns is under development and will be in place prior to the restart of Unit 1.
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9.
Statement of Apparent Noncosoliance:
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Technical Specification 3/4.4.5, " Reactor Coolant System I.eakage " requires in Section 3.4.6.2, that Reactor Coolant System (RCS) leakage be limited to 1 gallon per minute (CPM) of " Unidentified Leakage," and that unless i
rates above this limit are reduced to within the limit within four hours, l
the plant must be placed in " Hot Standby" in the next six hours and in
" Cold Shutdown" in the next thirty hours, i
contrary to the above, from March 22 until March 28, 1979, RCS "Unidenti-fled Leakage" remained above 1 spa, and the plant was not placed in " Cold Shutdown."
Discussion:
Due to an error in the esiculational procedure, computations of unidentified reactor coolant systen leakage for the March 22-28, 1979, period mistakenly produced values less than 1 spa.- As a result, Technical Specification 3.4.6.2 was violated.
Surveillance Procedure 2301-3DI, "RCS Inventory", calculates leak rates in caras of spa at reactor coolant system operating conditions. TCN 2-79-070, issued on March 16, 1979, made a correction to the calculational procedure by correcting changes in the reactor coolant drain tank inventory to reactor coolant system operating conditions. Bovever, the Temporary Change Notice failed to recognise that a similar correction for additions to the sakeup tank was also needed. The result was to remove an off-d setting error. By correcting only one of the two errors, the leak race calculaeion becone inaccurata.
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Correction Action:
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1; the procedure has been thoroughly reviewed and changes have been prepared.
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The proposed procedure change was provided to the NRC (Site Offica) on August 24, 1979. Upon receipt of comments, the procedure change will be completed. A similar review of the Unit 1 procedure has been conducted l
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cedure shall be complete and all changes incorporated prior to the re-start of Unic I.
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i 10.A Statement of Apoarent Noncomo11ance:
10 CFR 20.401, " Records of surveys, radiation monitoring, and disposal,"
requires in Section (a) that each licensee maintain records showing the radiation exposure for all individuals for whom personnel monitoring is required on a Form NRC-5 or equivalent and in section (b) requires that each licensee maintain records of the results of surveys required by 10 CFR 20.201(b).
Contrary to the above:
A.
he results of approximately 500 ground level radiation surveys conducted during March 28-30, 1979 in oifsite areas bordering the Bree Mile Island site were not documented in a manner which permitteid a precise evaluation of the type of radiation (Beca/Canma) which existed in the environs. Pertinent information such as the type of instrumentation used and whether the end window on the probe was open or closed was not recorded.
t Discussion:
10C7120.201(b) states in its entirety, "Each licensee shall make or cause to be made such surveys as may be necessary for him to comply with the regulations in this part." As pointed out in the above statement of apparent noncompliance,10CFR20.401(b) requires that each licensee shall maintain records showing the results of surveys required by 10C7R20.201(b).
Metropolitan Edison believes that although more detailed documentation of survey results is desirable, the documentation of the results of 4
g the 500 ground level radiation surveys conducted during March 28-30, 1979 was not in violation of the regulations.
As required by l'0CFR20.201 (b), monitoring taans were dispatched around the 2MI Site to measure and report the general radiation levels. 2he results of these offsite radiation measurements (surveys) were radioed to C
the EmerEency Control Station (ECS) where they were recorded by emergency 5
i personnel. Discussions with the Supervisor-Radiation Protection and Chemistry who was the ECS Director noted that the individuals receiving the radio reports were using an agreed upon shorthand notation to record radioed survey results. 2 is process allowed them to report the information
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to supervisory personnel while simplifying the recording process and maxi-E I
mising the number of surveys that could be taken. De annotations would simply record location, time and level for all surveys using the standard meters available in the established fashion. Exceptions were specifically noted such as a change in instruments, both open and closed window readings, By understanding this shorthand, it was possible to reconstruct the etc.
full survey information.
2 It is recognized that this method of recording is not sufficiently formal to be understandable without other explanation and was not of the quality desired for historical evaluations of conditions that existed throughout the accident. It is also recognized that the potential for error with the informal shorthand method of recording is greater than with formal recording methods.
However, in spite of these recognized shortcomings, it is not apparent that the method of recordkeeping hampered the real time evaluation of radiological conditions which is a principal-function of surveys as l
described in 10C7R20.201. Dere were no erroneous actions taken as a essult of these survey records.
4 Se post-accident evaluations of offsite exposure primarily relied on information obtained from the existing dispersed TLD array and its supplemental TLD's, and from the numerous air, vacar, animal, and vegeta-tion sampling prograes. R ose exposure estimates were correlated against in plant monitor information. Since the early radiation survey informa-tion was not a significant contributor to these evaluations, the uncer-f tainties of these records had no detrimental Impact' on the ability 2
to perform the necessary evaluations of exposure to the general public.
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Cosreetive Action:
t To enable future reviews of the survey data to be fully understood, an I
explanatory document will be developed that will provide the shorthand method utilised for the 500 offsite radiation measurement taken insedi-
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acely following the TMI-2 accident. This document will be provided as part of the permanent record requirement of 10Cy120.401 for the accident.
To improve the quality and clarity of future offsite survey records, forms will be developed and included in each emergency kit and at other locations as appropriate. These forme will include field survey team
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identifiers, identification of instruments used, person conducting survey, location, reading eine, level of radiation, type of radiation, and a remarks colunn to note iteme such as directionality, shielding or other pertenant facts. These forms will be required to be filled out by the survey team at the time of each measurement, and will be collected and retained for historical purposes as required. Information from these surveys radioed to a central location will be recorded in an abbreviated i
fashion noting the survey team identifier, time of measurement, location, level and type of radiation, and remarks.
Training will emphasize the need to establish survey practices which will a
provide high quality measurements to aid in the response to radiological emergencies. It will also emphasize that producing clear, complete, understandable records of those surveys is essential in order to allow historical evaluations of exposures to people not provided with individual e
dos imetry.
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t 10.5 Statement of Aoparent Noncompliance
,i 10 C71t 20.401, " Records of surveys, radiation monitoring, and dispcsal,"
i requires in Section (a) that each licensee maintain records showing the 1
i radiation exposure for all individuals for whom personnel monitoring is required on a Form NRC-5 or equivalent and in Section (b) requires that each licensee maintain records of the results of surveys required by 10 CFR 20.201(b).
Contrary to the above:
5.
The records of the radiation exposure for at least 5 ladividuals exposed during the period March 1 to 31, 1979 had not been recorded or meintained on a form NRC-5 or equivalent as of July 5,1979.
Further-more, as of July 5,1979 the assessment of their doses had not been completed.
Discussi n:
Metropolitan gdison agrees that a recording error has been made la that of the more than 1000 individuals for whom personnel monitoring was re-
"t quired during March,1979, the records for 5 individuals were not properly malatalsed.
This recording error is a direct result of the heavy work load placed on a limited staff under emergency conditions and is not be-lieved to represent a shortcoming in the maintaining of radiation ex-posure recorda during normal operation.
Within a few days of the accident, the number of individuals for which radiation exposure records was required nearly doubled from approximately e
600 to approximately 1200. At the same time, experienced people contri-buting to the maintenance of radiation exposure records during normal operation were called upon to perform other functions during the emergency response to the accident. Less esperienced people were therefore utilized to maintain a large volume of new radiation arposure records.
a As a
)
result a few records were not properly maintained. No violation of erposure limits or increases in occupational exposure resulted from this error in record keepiss.
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f corrective Action:
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In order to correct for the error in records maintenance for the 5 in-I I
dividuals identified above, and to assure that no other errors exist la the radiation exposure records of all persons for whom personnel monitor-Ing has been required from March 28 to June 30, 1979, a review of exposure records will be completed by January,1980.
i This review will involve examining all radiation exposure records during the time period of the accident and identifying those records for which any sort of abnormality or incompleteness exists. For those records showing abnorma11 ties or incompleteness, a re-evaluation will be completed using telephone and personal interviews. Re-evaluations will also be performed for all individuals identified as having been present is an area with abnormal conditions, such as the auxiliary building, during the period of March 28 to June 30, 1979.
It is not expected that this program can be completed before the end of January 1980 because at least 2000 exposure records will need to be evaluated, requiring the full time effort of six personnel. This program is documented in a Metropolitan Edison letter to the NRC dated October 29, 1979.
\\
In addition to the above records correction program, to assure proper 2
records maintenance la the course of future emergencies, the revised i
site Emergency Plan specifically charges the Emergency Director with the responsibility for assuring that accurate exposure records are maintained.
i This will require appropriate training for any persons assigned temporary b
responsibilities for records maintenance and advance planning to assure 5
the capability to handle the expected increase in records processing in the event of a radiological emergency.
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- 11. St.ntement of Apparent Noncomellance:
7 t
10 CFR 50, Appendiz 3, Criterion X, " Inspection," requires that a program I
for inspection of activities affecting quality shall be established and executed to verify conformance with documented instructions, procedures and drawings for accomplishing the activity.
Three Mile Island Nuclear Station - Unit 2, Final Safety Analysis Report, Chapter 17.2.13.Section X, requires that the inspection program include i
random observation of operations and functional testing by individuals independent of the activity being performed.
Procedure CP 4014,. "QQA Surveillance Program " Revision 0, requires independent observation of activities affecting quality to verify conform-ance with established requirements utilizing both inspection and auditing techniques...for compliance with written procedures and the Technical i
Specifications.
Contrary to the above, as of March 28, 1979, the normal operations sur-veillance testing activities had not been made subject to randon and/or routine laspections by independent methods.
Discussion:
Metropolitan Edison believes that there is no noncompliance in connection with normal operations surveillance testing.
In accordance with CP4014 Rev. O, the ' Metropolitan Edison QC Department scheduled and performed Inspections of TMI-2 operations Technical Specification surveillance testing as documented by the following QC Surveillance Reports prior to March 28, 1979.
t Date Surv. Rep. No.
Surv. Proc. No.
Title 9/78 78-175 2303-M15A/B Control Room Emergency Ventilation Systes 9/78 78-181 2322-Al Waste ces and Unit Vent Dis-charge Functional Test 10/78 78-191 2303-M14C E.F. Sys. Talve Lineup verification and opera-i bility Test; and Turbine 1
Drives E. Feedpump Opera-bility Test 11/78 78-196 2304-Q1 Diesel Fuel Testing
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11/7 8 78-206 2303-M34 Safety Inj. Manual Actua-I tion and Act. Logic Fune.
Test 11/78 78-219 2303-M13 16 Pump & Valve Functional Test
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N2 Date Sury. Rep. No.
Sure. Proc. No.
Title 7
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i 12/78 78-235 2301-M8 Cont ainnent Integrity veri-I fication 1/79 79-04 2612-R4 LIQ Radiation Monitor Cali-bration 2/79 79-12 2302-R23 TW Line Rupture Auto Detec-i tion Calibration Technical Specification surveillance procedures were selected for per-formance inspection at random and witnessed when performed to verify conformance with documented procedures. 2his process ses in accordance with 10C7150, Appendiz 3 Criterion Z.
Corrective Action:
Although this item does n'ot involve noncompliance' with regulations, Technical Specifications or procedures, Metropolitan Edison is planning to expand its program for inspection of surveillance testing activities.
Initial steps to expand this program had begun prior to March 28, 1979.
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Isspecefully subaieeed, g
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R. C.
old Senior Vice President Metropclitan Edison Company December 5, 1979
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UNITED STATES OF AMERICA I
NUCLEAR REGULATORY COMMISSIOPI h
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In the Matter of
)
)
.MPOLITAN EDISON COMPANY )
Docket No. 50-320
)
(Three Mile Island Nuclear
)
Power Station, Unit 2)
)
METROPOLITAN EDISON COMPANT'S ANSWER TO NOTICE OF PROPOSED IMPOSITION OF CIVIL FENALTIES Pursuant to 10CFR2.205(b) ed the Notice of Proposed Imposition of Civil Penalties of October 25, 1979, Metropolitan Edison company provides the follow-ing written newer. The answer to each apparent noncompliance incorporates by reference the statement in reply to that item set forth in Metropolitan Edison Company's Statement in Reply to Notice of Violation (hereafter
" Statement").
Apparent Noncomo11ance 1: Metropolitan Edison believes that this item is not a noncompliance. See Statement, pp.
1 1
Apparent Noncompliance 2A: Metropolitan Edison believes that this item is a concompliance. However, the Stacament, pp.
16, sets forth what Metropolitan Edison believes are extenuatin7gce stances and requests remission or mitigation of the proposed penalty.
Apparent Noncompliance 23: Metropolitan Edison believes that this item is a noncompliance. However, the Statement, pp.
19, sets forth what Metropolitan Edison believes are extenuating circum-stances and requests remission or mitigation of the proposed penalty.
Apparent Noncompliance 2C: Metropolitan Edison believes that this item is not a noncompliance. Sae Statement,"pp.
21 Apparent Noncompliance 2D: Metropolitan Edison believes that this I
item is a noncompliance. Bovever, the Statement, pp.
23, sets forth what Metropolitan Edison believes are extenuating circia-t stances and requesta remission or mitigation of the proposed pennity.
s Apparent Noncompliance 2Z: Metropolitan Edison believes that this item is a noncompliance. However, the Statement, pp.
25, sets
)
forth ebat Metropolitan Edison believes are extenuating circime-
)
i stances and requesta remission or mitigation of the proposed 7
)
penalty.
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Apoarent Noncompliance 27: Metropolitan Edison believes that this I
item is a noncompliance. However, the Stacament, pp.
23, sets forth dat Meeropolitan Edison believes are extenuating ciretam-stances and requests remission or sitigation of the proposed penalty.
Apoarent Noncompliance 20: Metropolitan Edison believes that this item is a no== compliance. However, the Statement, pp.
28, sets forth dat Me:ropolitan Edison believes are extenuating ciretar stances and requests remission or mitigation of the proposed penalty.
Apparent Noncompliance 2H: Metropolitan Edison Company believes that this item is a noncompliance.
Apparent Noncompliance 3: Metropolitan Edison believes that this item is not a noncompliance. See Statement, pp.
31 Apparent Noncompliance 4A: Metropolitan Edison believes that this item is not a noncompliance. See Statement, pp.
34 _.
Apparent Noncompliance 4.3.1:
Metropolitan Edison believes that j
this item is not a noncWianca See Statement, pp.
42 l
t Apparent Noncompliance 4.B.2:
Metropolitaa Edison believes that this
~
item is not a noncompliance. See Statmeest, pp.
52 Apparent Noncompliance 4.C: Metropolitaa Edison Company believes that this item is a noncompliace.
Asparent Noncompliance 4.D: Metropolitan Edison believes that this item is not a noncompliance. See Statement, pp.
58 Apparent Noncompliance 4.E: Metropolitan Edison believes that this ites is not a noncompliance. See Statement, pp.
60 Apparent Noncompliance 4.F: Metropolitan Edison believes that this item is a noncompliance.
Apparent Noncompliance 4.C: Metropolitan Edison believes that this item is a noncompliance.
Apparent Noncompliance 5: Metropolitan Edison believes that this item is a noncompliance. Bowever, the Statement, pp.
71, sets l
forth dat Metropolitan Edison believes are extenuating circum-I stances and requests remission or mitigation of the proposed penalty.
C 1
Apparent Noncompliance 6: Metropolitan Edison believes that this 5.
item is a noncompliance. However, the Statement, pp.
74, sets i
forth dat Metropolitan Edison believes are extenuating circime-stances and requests remission or mitigation of the proposed.
penalty.
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