ML19322C981
| ML19322C981 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/23/1980 |
| From: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| 80-002085002010, 80-2085002010, NUDOCS 8002070098 | |
| Download: ML19322C981 (35) | |
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UNITED STATES o
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W 2 31950 Docket Nos. 50-289
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and 50-320 Metropolitan Edison Company Attn: Mr. R. C. Arnold Senior Vice President 260 Cherry Hill Road Parsippany, New Jersey 07054 Gentlemen:
This is in response to your letter of December 5,1979, which was in response to our letter of October 25, 1979, transmitting a Notice of Violation and a Notice of Proposed Imposition of Civil Penalties in the amount of $155,000.
Your letter had as an enclosure a detailed response to each item of noncompli-ance set forth in the Notice of Violation.
Similarly, Appendix A to this letter contains our evaluation of your response 'and states our conclusion regarding each itea.
In this regard, you are requested to submit supplemental responses as described in Appendix A.
These responses should be in accordance with the instructions contained in Appendix A of the October 25, 1979, Notice of Violation.
Our letter of Octc3er 25,1979, discussed the overall impact of inadequacie.s discovered as a result of the investigation undertaken after the March 28, 1979 accident. These inadequacies were the basis for the statement that your management controls for the operation of the Three Mile Island facilities were inadequate.
Your response to the proposed items of noncompliance provides additional details as to the aspects of these items and the accident.
However, our belief that management controls were inadequate has not changed.
The Metropolitan Edison Company apparently believes that there was generally good perfomance, both prior to and subsequent to the accident, that there were few real items of noncompliance and that these were relatively unimportant, and that the other cited.. items either were not noncompliances or were mere technicalities.
As'has been pointed out by many investigating organizations, there were numerous contributing factors to the accident on March 28.
- Moreover, for at least two hours following the reactor trip, actions could have been taken which would have changed the accident from the " worst in the history of the nuclear power industry" to a relatively minor operational problem.
- Clearly, during the time interval reviewed by the IE Investigation Team, the Investiga-tion Report [NUREG-0600] shows that overall performance was not good, either preceding, during or following the accident.
With regard to the coraitments you have made in your letter of December 5, 1979, and the additional commitments asked for in Appendix A to tHs letter, we wish' to remind you of the difficulties experienced in the recent past, CERTIFIED Mall.
a ;0 0 9.b RETURN RECEIPT REQUESTED u
~JAN 2 31980 Metropolitan Edison Company'
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particu2 rly with regard to your Radiation Safet Program.
It was Metropolitan Edison's failure to meet specific commitments made in July 1979 to upgrade the Three Mile Island Radiation Safety Program which led to the (stablishment on September 26, 1979, of the Special Panel on Three Mile Island Unit 2 Radiation Protection Program.
The response to two proposed items of noncompliance (Items 1 and 3) was based in part upon your belief that an NRC inspector confirmed a Metropolitan Edison judgesent concerning Technical Specification 3/4.7.1.
The inspection report may not be explicit in describing the areas inspected; however, the areas inspected and the inspection findings were discussed with Metropolitan Edison at the completion of this inspection. During the inspection covered by inspection report (50-289/78-23 and 50-320/78-36) the inspector verified that the required surveillance procedures had been completed, and that the Technical Specification requirements concerning the frequency of surveillance and the acceptance criteria as specified in the procedures were satisfied. There was rio attempt during this inspection to evaluate the technical adequacy of the surveillance procedure for Emergency Feedwater (SP 2303-M14A/B/C/0/E).
- Moreover, despite the fact that previous inspections did not identify items of noncom-pliance, this fact does not absolve the licensee of the responsibility for items of nonccmpliance identified in this inspection.
The NRC recognizes the necessity of allowing reasonable operational discretion in these instances where plant conditions do not fall within existing procedures.
However, the significance of isolating designed safety features, or removing those systems inccrporated into the plant design specifically to protect the plant during accicent conditions, cannot be overemphasized.
The fact that plant conditiens are cutside those normally encountered or expected requires careful assessment before deliberately removing safety fe tures since the unusual conditions thenselves say increase the probability that the disabled safety feature will be needed.
In several places in Appendix A you are requested to submit additional information to cocplete your response to the Notice of Violation and the Notice of Proposed Imposition of Civil Penalties.
We are aware that informa-tion has been and continues to be supplied by you to various NRC offices as a part of the ongoing activi. ties at Three Mile Island.
When submitting the additional information you may include by reference any information previously provided to any NRC organization component.
We have reviewed your response to the items of noncompliance cited.
After careful consideration, we conclude that the items of noncompliance did occur as cited in the Notice of Violatica, with the exception of items 4.0, 4.E.2, and 11 which were not found to exist as cited.
Therefore, your enforcement histor/ will be corrected. The Civil Penalties for withdrawn items are remitted.
Since the proposed Civil Penalty of S155,000 was much less than the cum'ulative Civil Penalty of $717,000 because the Atomic Energy Act limits the total Civil Penalty for any 30-day period to 525,000, the mitigation has no effect on the dollar amount of the i: posed Civil Penalty.
Accordingly, we hereby serve the enclosed Order on Metropolitan Edison Co=pany, imposing Civil Penalties in the amount of one hundred fifty-five thousand dollars (5155,000).
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J N 2 319S0 Metropolitan Edison Company'.-
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5-In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Docucent Room.
Sincerely, O
,s"% f d'..'.
Victor Stelloi, Jr' Director Office of Inspection and Enforcement
Enclosure:
1.
Appendix A 2.
Order Impesing Civil Monetary Penalties i
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THis DOCUMWT COMS POOR QUAUTY.PAGES
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Acoendix A
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For each item of noncompliance and associated Civil Penalty identified in the Notice of Violation (dated October 25, 1979) the original item of noncompliance is restated and the Office of Inspection and Enforcement's evaluation and conclusion regarding the licensee's response to each item and proposed imposi-tion of Civil Penalty is presented.
ITEM 1 Statement of Noncemoliance Technical Specification 3/4.7.1, " Turbine Cycle," requires in Section 3.7.1.2, that three independent steam generator emergency 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 Hot 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 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on March 28, 1979, the flow paths to both steam generators were made incperable by feedwater header isolation valve clo s ure.
(In addition, on January 3, February 26 and March 26, 1979, the flow paths from all three emergency feedwater pumps were simultaneously made inoperable by feedsater header isolation valve closure during the performance of, and in accordance with, an improper surveillance test procedure.)
This violation contributed to an accident. (Civil Penalty $5,000)
Evaluation of Licensee Resoonse The licensee denies this is an ites of noncompliance and bases that denial on the assertion that there is only one emergency feedwater system for Unit 2.
Metropolitan Edison further asserts that the Technical Specifications (TS) permit the emergency feedwater system to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and thus no noncompliance existed immediately prior to the accident or during previous
'urveillance tests.
This assertion that inoperability of all emergency feed-ater capability wuld or.should be acceptable for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is wholly inconsistent with the minimum equip::ent assumptions used in the analysis of accidents contained in Chapter 15 of the FSAR.
While the terminolosh used in the TMI Unit 2 FSAR describes one system, as i
noted by the licensee, the term system" in the TS cannot be construed to support the licensee's assertion. Metropolitan Edison's interpretation of this term is not supported by the analysis assumptions used in the FSAR or the Safety Evaluation Report Analyses. To place this systein in a condition contrary I
to these assumstions violates the operability of this system; operability is defined in TS 1.6.
The licensee further asserts that support for their position is to be found in an IE Inspection Report.
This inspection report (50-289/78-23 and 50-320/78-36) states that the inspector verified that the required surveillance procedures had been completed, and that the Technical Specifications requirements concerning
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Appendix A.
the frequency of surveillance and the ac"ceptance criteria as specified in the procedures were satisfied. There was no attempt during this inspection to evaluate the technical adequacy of the surveillance procedure for Emergency Feedwater (SP 2303-K14A/8/C/0/E).
The licensee's response indicates that its analyse; support the conclusions of the President's Commission on.the Accident at Three Mile Island regarding the effect of the closed EFW valves on the outcome of the accident. That conclusion is that the closed valves had no significant effect on the outcome of the accident. This cenclusion is consistent with the evaluation provided in Sectica I-4.2.3 of the Investigation Report.
In that all of these conclusions and, evaluations Concur that the Closed EFW valves misled the operators into drawing erroneous early conclusions, the item of noncompliance is appropriately classified as one which contributed to an accident.
Conclusion The iten as stated is an item of n:ncocpliance.
The information provided by the licensee does not provide a basis for modification of the enforcement acti or..
In view cf. Metropolitan Edison's interpretation of TS 3/4.7.1 and of our ccaclusions concerning this item, a supplemental response is requested which specifies:
(1) each procedure reviewed for Units 1 and 2 which isolates or defeats part or all of any system whose operation is required by the TS or by the accident analysis contained in the FSAR; and (2) the method by which the operability requirements will be satisfied during the conduct of each procedure identified in (1).
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Appendix A 3-
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ITEM 2 Statecent of Noncaroliance The severity and uniqueness of the accident which occurred at Three Mile Island resulted in a sarked reduction in the normal good health physics prac-tices which are mandated by the NRC Regulations.
Under the circumstances of an accident of this magnitude, the NRC recognizes that in the interest of reactor safety a departure from normal health physics practices and standards may sometimes be candated by the exigencies that exist during such conditions.
However, the hRC also believes that the licensee, with the resources available and,taking into account the tioe frame available for conduct of safety related functions, could have taken additional measures to better control the overall health physics actions and decisions which were made during the course of the accident. The following items of noncompliance exemplify unacceptable degrada-tion from health physics practices pertaining to control of access to high radiation areas, condect cf radiation surveys, and personnel radiation exposure monitorinc.
10 CFR 20.201, " Surveys," requires in Section (b) that each licensee shall make er cause to be made such surveys as may be necessary to comply with the regulations in 10 CFR 20.
10 CFR 20.202, " Personnel Monitoring," requires that the licensee supply appropriate personnel sonitoring e;uiptent and requires its use for each individual whc enters a restricted area and is likely to receive a dose in excess of 25 percent cf tha applicable value specified in 10 CFR 20.101.
Technical Specificatica 6.12, "Higa Radiation Area," requires that each area in which the inter.sity of radiation is greater than 1000 mrem /hr be provided with locked doors to prevent unauthorized entry into the area and that any individual entarirg the area be equipped with a continuously indicating dose rate ;tonitoring device.
10 CFR 20.103, "Exacstre cf individuals to concentrations of radioactive materials in air in restricted areas," requires in Section (a)(3) that the licensee sake suitable ceasurements of the concentrations of radioactive materials in air for datecting and evaluating airborne radioactivity in restricted areas for the purpcses cf determining compliance with the regula-tion in 10 CFR 20.103(a)(1).
10 CFR 20.101, " 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 incividual in a restricted area to receive in any period of one calendar quarter a dose in excess of three rem to the whole body, or 18 3/4 ren to the hands and forearms, or 7 rem to the skin of the whole body; i
Appendix A.-
Contrary to the above:
A.
From 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> en 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 much greater than 1000 mrem /hr during this period; B.
From the evening of March 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 continuously indicated the dose rate;
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C.
No measurements were made of the concentrations of airborne radioactive saterials in the Unit 2 auxiliary building for periods during which indi-viduals were exposed from 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on March 28, 1979, through midnight, March 30,1979, nor in the Unit 1 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; D.
On March 29, 1979, an Auxiliary Operator was permitted to enter areas of the auxiliary building where exposure rates of up to 100 R/hr existed.
Radiation survey information and appropriate personnel monitoring were not provided to the cperator for this entry.
This contributed to the Operator receiving a whole body dose of 3.170 rems.
When this dose was adced to the operator's previous dose for the quarter, the operator's quarterly whcle body dose was 3.870 rems as measured by personnel dosicetry devices; E.
On March 29, 1979, a Nuclear Engineer entered an area of the auxiliary builcir.g where the radiation level was greater than that which could be ceasured by his portable survey instrument (2R/hr).
Failure to perform a survey of the exposure rate in this area contributed to the individual receiving a dole body dose of 3.14 rems for this entry.
When this dose was added to the engineer's previous dose for the quarter, the engineer's cuarterly whole body dose was 4.175 rems as measured by personnel dosimetry devices; F.
On March 29, 1979, a Chemist-y Foreman was permitted to repeatedly enter high radiation _ areas and hana'a samples of highly radioactive reactor coolant. This centributed to the Foreman receiving a whole body dose of 4.100 rens.
When this dose was added to the Foreman's previous dose for the quarter, the Foreman's quarterly whole body dose was 4.115 rems as ceasured by persennel dosimetry devices; G.
.On March 29, 1979, a Chemistry Foreman and a Radiation Protection Foreman were persitted to handle a highly radioactive reactor coolant sample without adequate personnel monitoring and without first performing a survey of had and forears exposure rates.
Handling of this sample resulted in a calculated dose to the hands and forearms of the Chemistry Foresan 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 rems; and
Appendix A '
H.
On March 28, and March 29,19V9, several individuals received skin contamination of the hand and other parts of the body sufficient to cause exposure rates 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.
Each day constitutes a separate violation, [ March 28 (A, B, C and H), March 29 (A, B, C, D, E, F, G, and H), March 30 (A and C)]; a civil penalty of $5,000 is imposed for each.
(Cumalative Civil Penalty S15,000)
Evaluation of Licensee Resconse A.
The response to example 2. A adnits noncompliance but argues that the
'overall access control program was reasonable under the circumstances and was in conformance with 10 CFR 20.
10 CFR 20.203(c)(2)(iii) requires positive control over each individual entry.
NUREG 0600 (pages II-3-34, 25, 54, 57, 70 and 71) establishes that such control was not exercised.
The NRC continues to believe that, with the resources available, addi-tional measures could and should have been taken to better control access to high radiation areas. The ccm:itnant for corrective action does not state specific changes to be cade to the health physics program to improve access centrol nor does it state the date when full compliance will be achieved.
B.
The respense to example 2.8 admits noncompliance but requests remission er mitigation of the proposed penalty since the number of instruments available was insufficient to meet demand.
The response also states that each individual entering the Auxiliary Building had "some awareness of information en dose rates" based on previous surveys and the number of individuals overexposed was icw and exposures were not significantly above limits. The fact that an insufficient number of instruments was available does not relieve the licensee of responsibility for providing such instruments to individuals entering high radiation areas as required by technical specifications. The fact that more than half of the licensee's survey instruments were out of service for maintenance or calibration endoubtecly contributed to this problem.
Informing individuals of previous survey results does not provide protection eq,.ivalent to equipping them with a monitoring device as required by Technical Specifications and does not provide adequate protection when radiation levels are as high and variable as they. vere during the period in question.
The NRC does not believe that an9 of the overerposures which occurred at TMI were justified; and certainly does not accept the statement that there were "few overexpo-sures" as justification for not providing monitoring devices to individuals entering high radiation areas.
The response stated that " site monitoring devices will be reevaluated and enhanced as necessary" but did not describe
. specific steps to be taken nor the date when full compliance will be achieved.
C.
The response to example 2.C denies noncompliance.
The denial is based on the licensee's belief that analysis of air samples was impossible due to the loss of counting room facilities, that urgent need for access in some cases justified entries without air samples, that collection of air l
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Appendix A.
sar.:ples would have caused unnecessary radiation exposure, and that the evaluation performed met the survey requirements of 10 CFR 20.201(b).
Analysis of air samples was possible and should have been done since the concentration of radioactive materials in the air was not known.
Such analysis could have been performed initially using the licensee's portable instrumentation and later by HRC and licensee contractor mobile labs which arrived onsite March 28 and 29 respectively.
Air samples could have been collected and analyzed without delay of vital entries into the Auxiliary Building and without receipt of excessive radiation exposure.
Although 10 CFR 20.201(b) requires surveys to include physical measurements of concentrations of radioactive material only when such measurements are
. appropriate, such measurements were appropriate in this case and should have. been made. The commitment for corrective action states that additional air monitoring equipment is in place,'but provides no information regarding the amount of equipment, performance capability, or intended use.
The response also states that retraining programs will place additional emphasis on air sampling techniques but the techniques to be emphasized are not describec and no information is provided regarding results achieved due to corrective steps taken.
The date when full compliance vill be achieved is not specified.
D.
The response to example 2.0 admits noncompliance but requests remission er mitigation of the proposed penalty based on the licensee's intent to follow sound health physics practices.
The circumstances related to the everexposure cited in example 2.0 exemplify lack of sound health physics practices.
For example, the overexposed individual was not briefed on radiological conditions prior to entering the building, he did not carry a high-range dosimeter, access controls were ineffective for preventing his reentry, ever. though he was contaminated, and he cade a reentry even though his lcw-range self reading dosimeter was offscale.
Other examples are cescribec in Section 3.2.4.7 of the Investigation Report.
Although the licensee suggests otherwise, appropriate instrumentation was not provided since the individual did not have a high-range dosimeter and r.ade a re-entry even though his low-range pocket dosiceter was offscale.
Although there was no doubt of the intent on the part of the individual and management to follow sound health physics principles, the individual had not been provided an understanding of health physics principles and canagement controls were not sufficiently effective to protect him.
The response states, that certain actions are being taken which could correct this problem such as revisions to Emergency Plan implementing procedures and changes in retraining programs, but the specific steps which have been taken and results achieved, the steps to be taken, and the date when full compliance will be achieved are not stated.
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.The response to example 2.E admits noncompliance but requests remission or mitigation of the proposed penalty based on the licensee's belief that the entry was vital to public safety and that proper radiological practices were followed to the degree possible. The NRC agrees that the entry was justified but does not agree that proper radiological practices were followed to the cegree possible.
The two engineers should have promptly exited the auxiliary building when their only high-range survey instrument
' Appendix A failed. Instead, they continued on even though their low-range instrument was frequently " pegged" [ radiation levels exceeded the instruments' capa-bilities].
Although identifying the source of leakage was important, the probles had been recognized for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> preceding the entry and the additional delay which would have resulted from exit to replace the failed instrument would not have affected public health or safety. More effective training of radiation workers and radiation chemistry technicians is essential to preventing recurrence of this problem, but the response does not describe specific steps to be taken in this regard r,or does it specify the date when full compliance is to be achieved.
F-G.' The response to examples 2.F and 2.G admits noncompliance but requests sitigatien or remission based on the licensee's belief that measures taken to minimize exposure were reasonable under the circumstances.
Alth ugh some planning was done and protective measures were taken which reduced exposure, the planning was not sufficient to anticipate the high cose rates ericour.tered nor to identify the need for extremity monitoring.
In acdition, the dose received by the Chemistry Foreman during a previous sampling operatica was not taken into account when planning the sampling in q'.-estion.
The NRC believes that the overexposures resulting from this sampling were unjustified and could have been prevented by more effective preplanning. The response states that special handling, tools, shielding, and training of chemistry personnel will be provided; however, this c: tait.2ent la:ks specificity and fails to address the more general area cf preplanning f:r all radiological work.
No date is specified for full compliance.
H.
ine respcase to iter.s 2.H admits noncompliance and states that dose evaluaticas have been completed and reports made to the NRC as required.
Ea specific corre:tive steps were specified for assuring more prompt e/aluatica of personnel contar.ination in the future.
Concitsion The items as stated are items of n ncompliance.
The information provided in the licensee's response does not provide justification for withdrawing any of the exacples of noncorpliance cited, nor does it provide justification for l
remission or citigation of the proposed penalty.
Commitments provided for corrective action arh inco:plete as discussed.
A supplemental response is requested which specifies _in greater detail: (1) the corrective steps which have been takea and results achieved; (2) corrective steps which will be taken to avoid further ite=s of noncompliance; and, (3) the date when full compliance will be achieved.
This supplemental information is requested for each example listed.
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Appendix A.-
ITEM 3 Statecent of Noncocoliance Technical Specification 6.5.1, " Plant Operations Review Committee," requires in Section 6.5.1.5.a, that the Plant Operations Review Committee (PORC) review all procedures (and changes thereto) required by Technical Specification 6.8 and any other procedure (or change) determined to affect nuclear safety.
Contrary to the above, inadequate reviews were performed on both Procedure Change Request No. 2-78-707, Revision 4 to Surveillance Procedure 2303-M27A/B, and.Drocedure Char.ge Request No. 2-78-895, Revision 8 to Surveillance Procedure 2303-M14A/B/C/0/E; both were reviewed and approved by the PORC (November 9, 1978 and August 15, IS78 respectively).
Each approved change included a valve lineup which resulted in energency feedwater header isolation, contrary to Technical Specificatien 3/".7.1 requirements.
Each of these inadecuate reviews constitutes a separate violation which contributed tc an accident; a civil penalty of $5,000 is imposed for each.
(Cumulati.e Civil Penalty 510,000)
Evaluation of Licensee Resconse The licensee cenies this item of noncompliance on the basis that the PORC revie-ed the procedure in question and tnat, as discussed in its response to Ite= 1, the procedure was act contrary to the requirements of TS 3/4.7.1.
While the ? ORC reviewed each procedure, this review failed to identify. the safety significance of changes to the surveillance procedures.
Based on IE's evaluation of the acncompliance cited in Item 1 and on a review of the opera-bility requirements of the Emergency Feedwater System, the PORC review was inadecuate.
The FORC members should have recognized that implementation of Surveillar.ce Procedure 2303-M27A/B or 2303-M14A/B/C/D/E would result in emergancy feecsater header isolation, contrary to technical specifications.
The licensee asserts that changes to the surveillance procedures were made to take into acccunt unnecessary ther al shock to the emergency feedwater nozzles and tc obtain repeatable results for tests required by the ASME Code.
While Metropolitan Edisen's motives to reduce thermal shock to these nozzles and obtain repeataale test results may have merit, this does not absolve the licensee of the responsibility to conduct operations in accordance with regulatory requiresents.
Conclusion The ites, as stated, is an iten of noncompliance.
The information provided by the licensee does not provide a basis for codification of the enforceme'nt action.
The licensee should address in a supplemental response the actions to be taken to assure PORC members hava the necessary technical expertise to demonstrate a
Appendix A 9-clear understanding of the implications of TS requirements and system operability requirements as stated in the TS and FSAR.
The specific further examples of similar test procedures contained in the response of the licensea should be included in the review of procedures planned by the licensee.
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The licensee should also address an appropriate target date for the completion of these reviews.
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l Appendix A.-
ITEM 4.A Statement of Noncoenliance Technical Specification 5.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
Emergency Procedure 2202-1.5, " Pressurizer System Failure," Revision 3, requires in Section A.2.8.1 that electromatic relief isolation valve RC-R2 be j
closed if, among other things, the valve discharge line temperature exceeds the non=al 130*F.
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Contrary to the above, the electromatic relief valve discharge line temperature had been in the range of 180 -200 F since October of 1978 and isolation valve RC-R2 was not closed as of 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on March 28, 1979.
Additionally, on March 28,1979, the discharge 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 />, but the isolation valve RC-R2 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.
Each cay the plant operated in noncompliance with this procedure constitutes a separate violation, a civil penalty of $5,000 is imposed for each.
(Cumulative Civil Penalty 5630,000)
Evaluation of Licensee Resconse
[RC-R2 should read RC-V2 in the Statement of Ncncompliance.
This typograpical error also appeared in the October 25, 1979 letter.]
The licensee denies this item of n:ncompliance.
The basis for this denial is their assertien that the existence of cne or more " symptoms" as listed in an energency procedure dces not call for implementation of the associated immediate and followup acticas. The licensee also asserts that such implementation would be contrary tc the understanding of Metropolitan Edison personnel at the time of the accident.
Interviews and discussions with plant personnel during the course of the investigation did not demonstrate a generally accepted understanding by the THI staff that sycotoms do not require implementation of the emergency proce-dures or that all symptoms must exist before any actions are taken.
The identification of a single sycptom is, as noted in the licensee's response, a signal that conditions shculd be examined to determine whether a problem exists.
It is this examination which allows the ' operator to implement all of the appropriate procedures to insure that plant safety is maintained, and that license requirements are not violated.
The fundamental method of determining whether the PORV was leaking is the only immediate action stated in the Emer-gency Procedure: shut RC-V2.
The position stated by the licensee that 'no action is required after the identification of a " symptom" or an abnormal condition is not consistent with operator training nor is it consistent with a conservative apprcach to nuclear safet:r.
Licensees are required under emer-gency procedures i:plementing TS to insure that abnormal conditions will tp identified, evaluated, and as appropriate, corrected.
Appendix A In any event, of the 4 symptoms listed in EmergebEy Procedure 2202-1.5, 3 were identified by plant operators prior to the accident, NUREG 0600, Section 1.2.4.
These 4 symptoms were identified, and deliberate operator actions were taken based upon existing pressurizer system conditions.
However, Energency Procedure 2202-1.5 was not followed.
1.
Symptom 1 of a leaking PORV is a high valve discharge line tecperature.
The licensee admits that the relief valve discharge temperature exceeded the 130 F normal temperature during' the October 1978 to March 1979 period.
The licensee asserts that the PORY was not leaking during the October-January pericd, and that the high temperature was caused by a leaking code relief valve (RVIA). However, even if this determination was correct,
, the' licensee failed to follow the Emergency Procedure in that the high temperatures were not placed on the Analog Trend Recorder.
Metropolitan Edison further asserts that these high temperature readings were due to plant design (conductive heating and temperature sensor location).* The licensee is responsible for insuring that all procedures are consister.t with plant design in order to assure safe operation.
The licensee's assertion leads to the untenable position that plant procedures could not be fcliowed due to plant design.
2.
Symptom 2 of a leaking ?0RV (RC-R2) is RC drain tank pressure above normal.
This syrptem also existed prior to the accident. The operators were operatir3 the RC3T transfer pump continuously to maintain the RCDT temperature (and pressure) at ambient conditions with apparent valve leakage into :he tank.
This continuous operation of the drain pump was another indicatica that a problem existed.
3.
Spnptom 3 of a leakin; ?ORV (RC-R2) is RC System makeup flow above nornal for the varia:le letdown flow and RC pump teal in-leakage conditions.
This symptom was indicated by the frequent transfer of reactor coolant between the R:DT and the.Make-Up Tank.
This third condition also existed prior to the accident.
4.
Symptom 4 of a leaking ?ORV (RC-R2) is boric acid concentration continually increasing in the pressurizer. While not identified in NUREG 0500, in order to ecualize boron concentration, pressurizer water was being recirculated through the spray valve.
Even though all of these symptoms exis:ed sicultaneously and were identified prior to the accident, the proper procddure (Ecergency Procedure 2202-1.5) was not followed.
Prior to the accident the FORV discharge tecperature was approximately 180 F.
This condition excaeds the norcal (130 F) by approximately 50 F.
The operators expected to see this tamperature above normal without taking action as specified
- Considering the pressurizer temperature necessary to maintain the reactor coolant pressure conditions the licensee uses to support its conductive heating theory, we conclude that conductive heating is an unlikely explanation of the PORV discharge line temperatures for the period from Octooer 1978 to March 1979.
Appendix A in Emergency Procedure 2202-1.5. "As outlined in Appendix I-A of NUREG 0600, at about +3 seconds into the accident sequence the reactor coolant drain tank pressure began to increase.
The PORV opened at approximately +6 seconds, and at approximately +13 seconds should have shut.
At +30 seconds the reactor pressure decreased to the low pressure-trip setpoint (1940 psig) and the PORY discharge temperature reached 239 F.
This temperature was not placed on the trend recorder; an action which would have helped in identifying an open PORV.
At +14 minutes, the reactor coolant drain tank [RCDT] rupture disc blew out and the reactor building pressure increased.
Despite these conditions, all of which indicated an open PORV, no action was taken to shut RC-V2 until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 18 minutes into tne accident.
The licensee also asserts that there is no indication that this procedure or the history of pilot-cperated (electro =atic) relief valve [PORV) discharge line temperature delayed recognition that the PORV had stuck open during the course of the accident. Shutting the relief isolation valve early in the accident could have prevented the accident entirely, reducing it to an operational transient.
There is a clear indication that recognition of an open PORV was delayed in pa-t by the past history of the discharge line temperature in that the Ecergency ?rocedure had not been implenented.
Much of the response of the licensee addressed these cany valid technical reasons which should have procoted a review and revisien to the applicable emergency procedure to make it appro-priate o the existing plant conditions.
Those revisions were not made, and therefore, the precedure was ignored rather than implemented.
As adcressed in t.5e investigation report, it was recognized that there is a certain cc:=onality between a leaking PORV and a leaking safety valve (pp i
I-15,5); hcwever, the appropriate diagnostic actions to differentiate between symptcas includine the use of the analog trend recorders were also not initiated.
The licensee, in its response, refers to the findings of the President's Coccission on the Accident at three Mile Island, specifically, the Technical Staff Analysis Report on Technical Assessment of Operating, Abnormal, ar.d Ecergency Procedures (October 1979).
A review of the referenced document (page 15) shows that the Presidential Commission also concluded that the symptcos cascribec above require closure of the PORV isolation valve.
Conclusion i
The item, as stated, is an item of noncompliance.
The information presented j
by the licensec dcas not provide a basis for modification of the enforcement l
action.
l The corrective actions preposed by the licensee to prevent recurrence of similar conditions lack the specificity to permit evaluation.
It is understood that the specific revisions to the PORV as regards position indication and leakage detercination will be part of the review of the restart proposal for Unit 1 and, at some later date, Unit 2.
However, the licensee should address in a suppleroer.tal response those steps being taken to assure that changed plant operating conditions will be factored promptly into emergency and operating procedures to assure that such procedures remain appropriate for l
staff use.
Additionally, the actions required upon identification of " symptoms" should be included in this response.
c
Appendix A.-
ITEM 4.8 Statement of Moncom31iance Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
B.1. Emergency Procedure 2202-1.3, " Loss of Reactor Coolant / Reactor Coolant System Pressure," Revision 11, requires in Sections B.2.2.3, B.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
, cperating prcperly as evidenced by flow in all four legs (250 gpm); that flows be maintained at this rate by throttling as RCS pressure drops; and that.high pressure injection not be terminated until RCS pressure can be saintained above the reset point (1640 psig) or until low pressure injec-tion flow is established at 3000 gpm.
Contrary to the above:
1.
At about 0405 on March 28, 1979, high pressure injection flow was thrcttled to minimum conditions even though RCS pressure was less than 1600 psi and failing, and without icw pressure injection flow established.
2.
At varicus times thr: ugh:ut the day of March 28, 1979, the high pressure injection system was modified such that the required flow rates were not maintained during continuing low pressure conditions within the RCS following the period when the reactor coolant pumps were stcpped and the high pressure injection system was the only
=cde available for tae removal of core decay heat.
B.2. Emergency Prccedure 2202-1.3, " Loss of Reactor Coolant / Reactor Coolant System Pressure," Revision 11, requires certain actions to be taken foilcaing the automatic initiation of high pressure injection, including in Section 3.3.1, that all ESF equipment is verified to be in its ESF position (capable of performing its intended function).
Contrary to the above, during the period of approximately 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> until 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on March 28, 1979, during continuing low pressure condi-tions within thb RCS, the Core Flood System was removed from its ESF position (rendered inoperable) by closing both tank isolation valves.
[This portion of the ESF vas inactivated during a period when reduction of Reactor Coolant System pressure was not the immediate goal.
This removed from service this safety feature during a period when it could
.have been called upon.
In the course of the accident while attempting to depressurire to activate the decay heat removal system NRC recogni2ed that it was necessary to isolate the core flood system and encouraged this action.
inis citation does not apply to isolation during this attecpt.]
I This violation contributed to an accident.
(Civil Penalty $5,000) l l
r
Appendix A Evaluation of Licensee Response The licensee denies this item of noncompliance, and responded to each example separately.
The licensee's denial of noncompliance Item 4.B.1 is based upon two pre =ises.
The first is that procedural compliance can only be ascertained after deter-mining which procedures were applicable and were in use.
The second is that ambiguities and inaccuracies in procedures due to the limits of accident analysis pre <!icticns sufficiently confused the interpretation of the situation so as to reascnably justify operator actions.
The ' natter of procedural compliance must be limited to those procedures applicable to an event.
However, the identification of sytptocs applicable to different procedures should be followed by evaluation of these symptoms.
It is in:uccent upon the licensee to insure that these two steps are properly completed.
These steps (identification and evaluation) were not properly completed.
The N?.C reccgr.izes chat there r.ay ce ambiguities and inaccuracies in procedures due te licits of accicant analysis predictions.
It is precisely such concerns over the inability to develop perfect procedures which have produced industry and regulatory recuire:ents on the scope and detail of the training and retraining prcgrans fcr operating personnel.
The Investigation Report identified a number cf procecures being icpienented almost simultaneously by the operating staff, and as neced in Section I-2 of that report, numerous instances of appro;riate c mpletior of procadural requir'ements.
The licensee response indicates a number of such procedural compliance examples and these parallel the findi.gs sucmarized in the Investigatica Report.
Tne central issue in this example of nonco:pliance is that the facility experienced a loss of coolant accident, and the operator action to limit HPI flow -as not in accorcance with TMI Unit 2 Emergency Procedure 2202-1. 3, " Loss of Reactor Cociant/ Reactor Coolant System Pressure." This procedure (in Secti:n B) lists eight symptoms indicative of a leak or rupture of sufficient size such thet the Engineered Safety Features System, including high pressure safety injection, are automatically initiated.
Such an automatic initiation did occur at the beginning of the TMI accident.
Four of these symptoms existed prior to the time that the reactor fuel became uncovered.
These were:
1.
Rapid, ccntir.uing decrease of reactor coolant pressure; 2.
High reactor building ambient temperature; 3.
High reactor building sump level; and 4.
High reactor building pressure.
The fcur listed sympto:s that did not exist were:
~
1.
Rapidly decreasing make-to tank level; l
2.
F.apid decrease of pressurizer level; i
i l
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Appendix A '
3.
High radiatien in the reactor building; and 4
Decreasing c:re flood tank level and pressure which would not be expected to occur, as the minimum pressure experienced during the early phases of the accident was 660 psi, and the core flood tanks begin to inject at 600 l
psig.
Thus the evidence available to the operators, in concert with their training in PW?. technolocy, was indicative of a reactor coolant loss.
The Ecergency Pro:edure repeatedly states the necessity of maintaining both pressurizer level and RCS pressure above the 1640 psig safety injection initi-atio'n point.
Ites A.3.2.5, for example, specifically cautions that, if the level cannot te maintained above 200 inches and pressure cannot be maintained above 1640 psig, the plant has suffered a cajor rupture and recuires operatio; in accordance with the section of the Procedure (Part 3) applicable to this condition (em;hasis added).
This section requires establishing an HPI flow of 250 g;m to ea:h of the four reactor coolant legs (125 gpm if one HPI pump fails to start).
Contrary to this reouirenent, although the pressure remained below 1640 after tae first 3 cinutes of the accident, the net addition rate to the RCS was reduced tc an average :f a:out 25 gpm during most of the first 3\\
hours.
The licensee's re;iy to pr: posed i:en of noncompliance 4.B.2 states that the Core Ficoc Tark isslation valvas (CF-VI A and B) were not shut during the perio: cited (3500 hou s until 1303 hours0.0151 days <br />0.362 hours <br />0.00215 weeks <br />4.957915e-4 months <br />).
This response further states that the elec rical breakers (n:rmally locked open) must be manually shut before the valves can be shut fros the ccatrol roca.
The Investigation Report estab-lished (Sectica 4.5, page I-4-28, and Interviews 95 and 198) that the valves were shut at appretima eiy 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />.
This finding is based on testimony of an operations Operator and a shift supervisor. The operations operator stated that he broke the lockJ cff of the breakers, and then shut the breakers.
This action allovec the Cors Flood Systaa isolation valves to be shut from the contrc1 rcom.
The shift supervisor stated that he shut these isolation valves from the Control F.:oc.
Conclusion Item 4.3.1, as stated, is an ite= cf noncompliance.
The corrective actions proposed by the licehsee appear adequate to preclude recurrence.
These proce-dural reviews and icprovements will be subject to review during evaluation of the restar; proposal for Unit I and, at a later date, Unit 2.
Item 4.S.2, as stated, is an item of acncorpliance.
Thd licensee should addre.ss in a supp1Ecental response those measures to be taken to insure that the operability recuiriments of Engineered Safety Features are met during all phases of operation.
The information providad by the licensee for Items 4.3.1 and 4.B.2 does not provide a basis fcr sodification of the enforcement action.
1
Appendix A 16 -
-w.
ITEM 4.C Statement of Noncccaliance
. Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
Operating Procedure 2104-6.2, " Emergency Diesels and Auxiliaries," Revision 9, establishes the procecures 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 shutdewn after the Engineered Safeguards Feature actuation has been cleared.
2.
Section 4.5, " Diesel Generator 1A(13) Shutdown to Emergency Standby,"
states in the cicsing step, 4.6.6, to place the diesel generator on standby in accordance with Section 4.2; and 3.
Section 4.2, when co pleted, establishes conditions for automatically starting the diesels upon actuation of an Engineered Safeguards Feature (ESF) including requirements to place the " Emergency Standby / Maintenance Exercise" switch in the Energency Standby position and resetting the fuel racks.
Contrary to the above, at about 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br /> on March 28, 1979, both the 1A and 13 diesel generatcr feel racks were manually tripped, thereby preventing an automatic start of the diesel generators upon ESF actuation and manual start from the cor.troi room until OS?9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.
This violation hac the potential to contribute to an accident.
(Civil Penalty S4,000)
Evaluation of Licensee Resconse The licensee a&r.its that the item as described above is an item of noncompliance.
Conclusion The itec as stated is an admitted item of non.ompliance.
The licensee has not requested mitigation of the Civil Penalty for this item.
l
Appendix A ITEM 4.0 Statement of Noncomoliance Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
Emergency Procedure 2202-2.2, " Loss of Feedwater," Revision 3, requires in Section 2.B.2.d that the operator adjust feed flow to control steam generator levels at 30 inches.
Cont ~rary.to the above, from approximately 0532 hours0.00616 days <br />0.148 hours <br />8.796296e-4 weeks <br />2.02426e-4 months <br /> until 0543 hours0.00628 days <br />0.151 hours <br />8.978175e-4 weeks <br />2.066115e-4 months <br />, the level in A stemn cenerator decreased to 10 inches (the minimum level indica-tion) while the A steam generator level was being controlled manually.
This is an infraction.
(Civil Penalty $3,000)
Evaluation of Licensee Resconse The licensee cenies this is an ites of noncompliance and bases that denial on the assertion that during the time frame indicated in the item, the referenced procecure did nct apply.
The circu: stances associated with the nonccmpliance were reviewed and ccmpared to the recuirements of Emergency Procecure 2202-2. 2, " Loss of Feedwater. "
The assertion in the licensee response that Section 2202-2.28, " Loss of Main Feedwater Flow to Jna OTSG" would be the appropriate procecure is not supported since it addresses a loss of main feedwater to a single generator while the Unit is in operation. The conditicn at the time was the case when no main feedwater was available (both oumps had tripped at the start of the accident).
Therefore, Section 2202-2.2A, " Loss of Main Feedwater Flow to Both OTSG's,"
remains the appropriate procedure.
The licensee correctly points cut that the procedural requirement referenced in the item of noncompliance is in that portion of the procedure which is applicable to the case when the loss of main feedwater is due to the feedwater valves closing.
H] wever, there is clear indication in both Sections 2.3.1 and 2.3.2 of the goal of~r.aintaining a 30-inch level in the generators when they are being fed (r.anually or automatically) by the emergency feedwater system.
The investigation team considered the failure to maintain the level as the item of noncompliance, not the rate at which level was recovered once it was lost.
Since these actions were involved just at the time of the shutdown of the second pair of reacter coolant pumps, it inc,ludes the period of preparation to use the natural circulation mode of cooling, which is controlled by Operating Procecura 2102-3.3, " Decay Heat Recoval Via OTSG." Since the operator actions involved in the noncompliance included dealing with a. transient situation, and
Appendix A
~
18 -
coving between two sets of procedural controls, the review of this item provides a basis for codification of the proposed enforcement action.
Conclusion A review of the circumstances and actions involved with this item shows that the licensee failed to maintain the steam generators at the desired level.
However, this review showed that this item was not a noncompliance. We are concerned that the licensee failed to maintain a heat sink to provide a means to cool the core. The licensee is requested to address in a supplemental respor.se the actions to be taken, including procedural improvements, to estab-lish, the reouired steam generator water level in all modes of feedwater or ecergency.feec.<ater addition.
Item 4.D is withdrawn and the Civil Penalty of 53,000 is remitted.
e, k
l l
f
Appendix A '
ITEM 4. E Statement of Noncemoliance Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
E.
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
. Superintendent, Unit Supt / Shift Supervisor / Unit Supt - Technical Support in the Control Room vill, after reviewing the emergency conditiens, classify the emergency as one of the following:
"a.
Personnel er Local Emergency, "b.
Si;a Evergency, and "c.
Ger.eral Emerg-ency "He will make this classification according to the condition of Tab'.e 1 of this plan, and initiate actions according to the Ecergency Plan Implementing Procedures, and according to his own best judgement;" and 2.
States in Table 1 of Section 2.1 that a Site Emergency exists when there is a reactor building high range gamma monitor alert alarm (Cor.dition No. e).
Contrary to the above:
1.
Acecuate written procedures were not established and impiamented in that Section 2.1 of ?rocedure 1004 for implementing the Emergency Plan lacked sufficient specificity and failed to result in a Site Emergency being declared at approximately 0430 on March 28, 1979, even thcugh primary system pressure had decreased to the point where safety infection was automatically initiated and a reactor building sump high level alara existed; and 2.
A site emerg-ency 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 which tice Condition "e" of Three Mile Island Emergency Plan 1004 had occurred.
This is an infraction.
(Civil Penalty $4,000)
Evaluation of Licensee Resconse The licensee's response to Item 4.E denied noncompliance.
Regarding 4.E.1, the licensee admits that greater specificity is needed in emergency plan i
+
Appendix A.
u-implementing procedures but implies the procedures were adequate to meet regulatory requirenents.
The NRC continues to believe that the procedure did not clearly identify those factors required to declare a site emergency.
As a result, the licensee failed to declare an emergency in a timely manner.
Technical Specificaticn 6.8 requires that procedures be established covering identified activities. The mere existence of an inddequate procedure does not fulfill this requirement.
Regarding 4.E.2, the licensee argues that the dome monitor alert alarm occurred at 0643 hours0.00744 days <br />0.179 hours <br />0.00106 weeks <br />2.446615e-4 months <br /> instead of 0635 as stated in the Investigation Report and that a site emergency was declared at 0650 instead of 0655.
Since it is understandable that. different involved individuals recall the time as being a few minutes different in one direction or the other, and since the time differences are so-snall, the NRC has decided to withdraw this portion of the item of noncompliance.
The cc=aitment for corrective action is acceptable except the date for implemen-tation of the revised training drill program is not specified.
Conclusion Item 4.E.1, as stated, is an item of ncncompliance.
Item 4.E.2 is withdrawn.
The Civil Penalty is partially remittec in the amount of $2,000.00.
The correctiva action specified is incomplete in that the date full compliance is to be achieved is act specified.
A su; placental response is requested to provide this information.
Appendix A '
~~
, = -
ITEM 4.F Statenent of Noncemoliance Three Mile Island Nuclear Station Health Physics Procedure 1670.9, " Emergency Training and Emergency Drills," Revision 4, dated January 16, 1978:
1.
Identifies in Section 3.1, the on-site emergency job categories and requires that training prograns for these categories will be conducted on an annual (calendar year) basis; and 2.
, Cescribes in Section 3.1.1 through 3.1.9, the training prcgram for aT1 en-site emergency job categories.
Contrary to the above, during calendar year 1978, not all individuals having emergency res;onsibilities were trained in that two Emergency Directors, one Accident Assessmer.t ir.dividual, eight Radiological Monitoring Team Members, and 37 Repair Party Team Members had not received the specified training.
In addition en March 28, 1979, during an emergency, at least four individuals who were assigned as recuired members of a Radiological Monitoring Team and seven individuais who were assigned as reouired r. embers of a Repair Party Tea:
performed e ergency dcties for which they were not t ained.
This is an infraction.
(Civil Penalty 54,000)
Evaluation cf Licensee Resconse Although the licensee in its response to item 4.F admits noncocpliance and agrees to pay the Civil Penalty, the licensee seemingly minimizes the signif-icance of inccmplete energency training by emphasizing the amount of training which was performed and implying that the incomplete portion did not have a significan; acverse affect on performance. The NRC believes that many of the problams associated wi:h the licensee's health physics performance folicwing the accicent could have been prevented by more effective training in this area. The commitment for corrective action is acceptable.
Conclusion The iten as stated is.an item of n ncompliance. The licensee has not requested mitigation of the Cisil Penalty for this item.
1 4
e
Appendix A '
ITEM 4.G Statecent of Noncomoliance Technical Specificatien 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified activities.
Station Administrative Procedure 1002, " Rules for the Protection of Employees Working on Electrical and Mechanical Apparatus," Revision 14, requires in Secticas 4.3, 4.4, and 4.5 that on restoration of equipment to service, removed tags, will have all recuired information entered thereon and then be suitably stored, and that the shift forecan shall approve equipment operation by signing the original tagging application. Additionally, Station Cerrective Maintenance Procedure 1407-1, Revision 0, specifies in Section 5.0, "Jcb Ticket (Work Request) Flow," the step-by-step process for initiating, processing, obtaining approvals, and ultimate filing of the " Job Package" which will include, among other thir.gs, documentation of corrective action taken (resolution description and cercification of satisfactory post maintenance testir.g) and Station Preventative Maintenar.:e Procedure E-2, "Cielectric Check of Insulation, Motors anc Cables," specifies how to make -he measurements and contains data sheets for recordiac -he values measured.
Contrary to the above, when inspected en June 20, 1979, the tagging application could not be found for caintenance performsd in January 1979, en Ecergency Feedwater isolaticn valves (5F-V12A,123, 32A, 328, 33A, and 333). Nc suitable documentation to determine whe:her the maintenance work had been completed, tags recoved, acceptance criteria :et, or valves approved for cperation could be found.
The TMI-2 aintenance lag lists this work request as being in an open states as of June 20, 1979.
This is a deficiency.
(Civil Penalty S2,000)
Evaluation of Licensee Resconsa The licensee admits that this is an item of noncompliance, and the corrective actions proposed and in force appear adequate pending site followup.
Conclusion J
The item is an admitted item cf noncompliance.
The information provided by the' licensee does not provide a basis, nor a request, for modification of this 4
enforcement action.
Appendix A..-
ITEM 5
~
Statacent of Noncomoliance Technical Specification 6.8, " Procedures" requires in Section 6.8.2 that changes to procedures which implement the Emergency Plan shall be reviewed by the Plant Operatiens Review Committee and approved by the Unit Superintendent prior to implementation.
Contrary to the above, a change to Station Health Physics Procedure 1670.7,
"&nergency Assembly, Accountability and Evaluation," was made without the required review ar.d approval. An additional assembly area was designated and the method used te perform accountability was modified by a memorandum dated October 13, 1378, fro: the Radiation Protection Supervisor to all departments.
As a result, en March 28, 1979, in response to an emergency, some licensee personnel followed the approved procedure while others followed the guidance in the October 13, 1973 memorancum, creating some confusion and de. laying proepc attainment of full accountability.
This is an infractica.
TCivil Penalty 5",000)
Evaluation of Lics,see Resconse The response adcits acacompliance tut requests remission or mitigation based on the licensse's belief that this iten did not delay prompt attainment of personnel acc:ur.taoility cr cause confusier.
The investigators concluded, based on -hree in:erviews with site security personnel, that delay and confusion did result fr:a this imprcper procedure change.
See page II-1-21 of the Investigation Repcrt. Regardless of this, the Civil Penalty was based primarily on the f act that.:recedure 1670.7 was chan;ed withcut the required review and approval cf the P'.ar.t Operaticas Review Ccr.r.ittee and not on whether delay and confusion resulted.
The commitments for ccrrective action are acceptable.
Conclesion The itec as written is an admitted iten of noncompliance.
The licensee's response does not contain'information that would serve as a basis for modifica-tion of the proposed enforcement action.
- ~
h, A
Apper. dix A '
s ITEM 6 State _-ent of Noncomoliance Environcental Technical Specification 5.7 requires that detailed written
~
procadures for instrument calibration be prepared and followed.
Three Mile Island Nuclear Station Surveillance Procedure 1302-5.24, Revision 3, dated Cecember 19, 1974, specifies the cethod of calibration and requires that it be perforted annually.
Contrary to the.above, as of March 29, 1979, eight environmental samplers had not been calibrated since 1974.
This is an infraction.
(Civil Penalty S4,000)
Evaluation of Licensee Resconse The response admits noncompliance but requests remission or mitigation of the proposed penalty since the procedure followed applied only to Unit 1, since a vendor had advised the licensee that calibration was unnecessary, and since NRC had previously classified the matter as an unresolved item in a May 1978 inspec-ion report. The fact that the procedure in question is a Unit 1 procecure is irrelevant since it applied to instrumentation common to Units 1 and 2.
Regardless of statements made by vendors, NRC considers that calibration of envircr ental air samplers is needed and is required at TMI by Environmental Technical Specification 5.7.
Upgrading an unresolved item to an item of noncocpliance is censistent with NRC enforcement policy and is not considered by NRC as evidence for mitigation. The corrective action commitment is not acceptable because it does not provide a ccomitment for instrument calibration and dcas n:t specify the date by which full compliance will be achieved.
Concitsi on The itam as stated is an admitted item of noncompliance.
The information proviced by the licensee does 'not provide a basis for modification of this enforcarecnt action.
The licensee is requested to submit a supplemental response addressing the areas described in the above evaluation.
i i
Appendix A.-
ITEM 7 Statecent of Noncc:aliance Technical Specificatica 6.2, " Organization," states in Section 6.2.1 and 6.2.2 that the unit organization and the organization of the corporate technical support staff shall b' 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 in Figure 6.2-1 in that:
A.
A positica titled, " Superintendent of Administration and Technical Support" was added to the organization on September 18, 1978 and filled en March 1,1979, such that the " Supervisor, Radiation Protection and Chemistry," reported to this position rather than directly to the
' Station Superintendent / Senior Unit Superintendent," and 3.
Thers were two "S;pervisor o' Maintenance" positions, one for each unit, rather than ene; and C.
' positica titlec "Su;erintendent of Maintenance" had been addad such that the " Supervisors of Maintenance" report to this new position rather
- han directly to the " Station Superintendent (Station Manager / Senior Unit Superintancer.:;" anc D.
Ine ; sitien of " hemical Supervisor" had been vacant since the issuance cf the Techni:al Specifications.
On Mar:h 23, 1979 thrc;gh March 30, 1979, the above organizatica discrepancies decreased the effe:tiveness of the licensee's response to the accident.
This is an infraction.
(Civil Penalty $3,000)
Evaluttion of Licensae Res:ense The rasponse admits noncompliance but request's remission or mitigation of the proposed penalty based on the licensee's belief that the organizational changes did nc: acversely affect its response to the accident and on its belief that the cited changes webe discussed with NRC on March 5,1979.
Although it appeared to the investigators that differences between the actual organ;zation and the organization assumad by emergency plan implementing procedures did reduce effectiveness of the licensee's response to the accident, the NRC recognizes that this conclusion is somewhat subjective and acknowledges that 4
these organizatior,al differences may not have had a significant effect on response.
In view of the above, the.53000 penalty proposed in the original Nctice of Violatica for this item of noncompliance was selected from the bottoa of the nonetary scale ($3000-4000) generally followed in the assessment
~
j l
of Civil Penalties for infractions by power reactor licensees. The more icportant ccncern here is the licensee's failure to obtain approval of new Technical Specificaticas prior to caking its organizational changes.
NRC i
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Appendix A.-
4-Region I' does not recall discussing with the licensee the organizational changes cited in this item of nonccmpilance.
In any event, the licensee's organizational changes were contrary to the licensee's existing Technical Specificaticns and should not have been made prior to obtaining an amendment to these Technical Spe:ifications. As the licensee is surely aware, the Commission's regu'.stions specifically provide that changes to Technical Specifications shall be cade through the fon=al amendment process, not through methods of the licensee's own choosing.
See 10 CFR 50.59(c).
The corrective actions proposed and underway appear adequate pending NRC completion of its revi ew.
Conclusion The item Is stated is an admitted item of noncompliance.
The information proviced by the li:enste d:es not crovide a basis for modification of this enforcement action.
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Appendix A.-
e..
ITEM E State ent of Noncemoliance Technical Specificatica 6.4, " Training," requires that a retraining and repla:ecent training program for the unit staff be maintained that meets or exceeds the requirements and recomcendations 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 N15.1-1971 recommendations had not been maintained for members of the radiation protection and chemirtry staff in that only 2 of the 10 topics recommended were included in the program.
This is aIn infraction.
(Civil Penalty $4,000)
Evaluatier. of Licensee Resconse The response cenies' this item of noncocpliance based on the licensee's belief that eniy :wo af the ten training areas specified in Section 5.5 of ANSI N13.1-19 71 appliec to r, embers of the radiation protection and chemistry staff and based en its belief that applying all ten areas to all members of the "c era-in; organi:stion" is contrary to the intent of the ANSI.
The NRC agrees tha: a;:licability cf some cf the ten training areas is somewhat limited for sc.e rambers cf the cperating organization, but believes that the radiation prote::icn an: chenistry staff should receive seme training in each of the ten areas.
- r extmple:
Area #1 specifies training in " Plant startup and shutdown proce:;ris;" 5;ch :recedures may require te:hnicians to take radiation measure-cents tr.d coolant ramples curing startup, but no such training was provided to the te:hnicians.
Area !2 speciffes : raining in " Normal plant operating conditions and procedures;"
c:viously radiation / chemistry technicians do not need the degree of training in this area trat is required for cperators; however, a general understanding of syste s fun:ticas is essential for caintaining effective radiological contr:1 cvar c;erations and maintenance activities.
Similar statements could be made fer the re:aining eight areas.
The licensee in its response notes that the IE combined Inspection Report 50-239/73-09 and 50-320/78-18 reviewed the general employee, craft and tech-nician training pregiam and identified no items of noncomliance.
The fact that n: itams of n:ncocpliance were found during an NRC inspection does not, and has never, been interpreted to mean that no items of nor. compliance existed.
The NR! believes that inadequate training of the radiation protection and chemi.stry staff was a major contributor to problems identified during the IE investigation and that substantial upgrading of retraining in this area:is e s sential. The cocaitments for corrective action in this area lacked sufficient specifici ty.
A supplemental response is needed which describes in detail the scope and extent of training to be provided.
Appendix A -
28 -
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Conclusion
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The item as stated is an item of noncompliance.
The information provided by the licensee does not provide a basis for codification of this enforcement action.
A supplemental response is requested to provide more specific training cc,Tcitments as discussed above.
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Appendix A.-
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ITEM 9 Statetent of Nor.cccoliance Technical Specification 3/4.4.6, " Reactor Coolant System Leakage," requires in Sectica 3.4.5.2, that Reactor Coolant System (RCS) leakage be limited to 1 gallo per mi.ute (GPM) of " Unidentified Leakage," and that unless rates above this iisit are recuced to within the limit within four hours, the plant must be placed in " Hot Standby" in the next six hours and in " Cold Shutdown" in the next :nirty h:urs.
Centrary to the above, from March 22, until March 28, 1979, RCS " Unidentified Leakage" remained above 1 gpm, and the plant was not placed in " Cold Shutdown."
Each cay constitutas a separate infraction; a civil penaltv of 53,000 is icposed fer each.
(Cumulative Civil Penalty $21,000)
Evaluttien of Licensee Resconse The li:ensee admits the item of noncompliance. The corrective actions proposed a..d U-te r-sy a: pea aciquate.
Cenc l.-s i on The i:sc is ar. a::.itted item of noncompliance.
The licensee has not requested mitiga:icn of the Civi; Penalty for this item.
Appendix A '
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- . e ITEM 10 Statenent of Nonccmoliance 10 CFR 20.401, " Records of surveys, radiation monitoring, and disposal,"
requires in Secticn (a) that each licensee maintain records shcaing the radia-tion exposure for all individuals for whom personnel monitcring is required on a For: NRC-5 or ecuivalent and in Section (b) requires that each licensee maintain records cf the results of surveys required by 10 CFR 20.201(b).
Contrary to the above:
A.
The results of approximately 500 ground level radiation surveys conducted during March 28-30, 1979 in offsite araas bordering the Three Mile Island site were not documented in a manner which permitted a precise evaluation of the type of radiation (Beta /Ga.:ma) which existed in the environs.
Fertinent information such as the type of instrumentation used and wnether the end window on the probe was open or closed was not recorded.
3.
The records cf the radiation exposure for at least 5 individuais exposed curirg the period March 1 to 31,1979 had not been recorded or maintained
- n a forc NRC-5, or equivalent, as of July 5,1979.
Furthermore, as of
- .uly 5,1979 the assessment of their doses had not been cc:pleted.
This is an ir. fraction.
(Civil Penalty S4,000)
Evalustion of Licensee Resconse A.
Ine respense denies that example 10A is noncompliance based on the licensee's belief that the absence of adequate records did not hamper the real time evaluation of radiological conditions.
The licensee admits that the surveys were required by 10 CFR 20.201(b).
NRC also believes the surveys were required and thus records of these surveys were also required.
Further, the NRC believes that the inadequate survey records hacpered the real time as well as the nistorical evaluation of radio-logical conditions..Although the licensee states that it was possible to reconstruct the full survey informatica from the original radioed survey results, the NRC investigation determined that the survey records were inadequate fcr the reasons stated in N'JREG-0600, page II-3-97.
The
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commitment for dorrective action is acceptable except that the date when full compliance will be achieved is not specified.
3.
The response ad=its that example 108 is noncompliance and requests citigation or remission nf the proposed penalty ba' sed on the unusually
.,large nur.ber of records generated and heavy demands on the individuals processing these records.
NRC recognizes that maintaining accurate l
records was difficult under the circucstances; but this difficulty is not justification for the failure of the licensee to identify and assess the cases of individuals who were known to have significant exposures.
The l
commitment for corrective action is acceptable.
Appendix A.-
Conclesion The iter., as stated, is an item of noncompliance.
The information provided by the licensas d:es not provide a basis for modification of this enforcement action.
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Appendix A '
i
- i. e ITEM 11 State:ent of Noncocoliance 10 CFR 50, Appendix B, Criterion X, " Inspection," requires that a program for inq3ection 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, Chaptar 17.2.15,Section X, requires that the inspection program include rando: observation of operations and functional testing by individuals independent of the activity being performed.
Procedure GP 4014, "OQA Surveillance Program," Revision 0, ~ requires independent observation of activities affecting quality to verify conformance with estab-lished recuirements utilizing both inspection and auditing techniques...for compliance with written precedures and the Technical Specifications.
Centrary to the above, as of March 28, 1979, the normal operations serveillance testing activities had not been made subject to random and/or routine inspec-tions by independent cathods.
This is an infraction.
(Civil Penalty S3,000)
Evaluation of Licensee Res:ense The irforcatica providad by the licensee is sufficient to justify withdrawing this itec as cited.
Ccnclusion This itaa of nonccrpliance is withcrawn; the associated Civil Penalty is remit ed. Metrepciitan Edison stated in its response that it is planning to expand its prcgrar for inspection of surveillance testing activities.
In view of this, a supplecental response is requested which addresses the specific.
require =ents, and cethods iof implecenting these requirements, concerning the inspection of activities as they are performed.
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