ML19317H396
| ML19317H396 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/19/1980 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19317H395 | List: |
| References | |
| TLL-158, NUDOCS 8006040112 | |
| Download: ML19317H396 (23) | |
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Metropolitan Edison Company
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Middletown, Pennsylvania 17057 717 9444041 Writu's Direct Dial Number May 19,1980 TLL 158 Of fice of Inspection and Enforcement Attn:
V. Stello, Director U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Three Mile Island Nuclear Station, Unit II (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Responses to Items of Noncompliance on March 6,1980, Metropolitan Ediron representatives met with NRC Inspection and Enforcement personnel to discuss the items that were originally published in Investigative Report 50-320/79-10. The purpose of the meeting was to agree on acceptable corrective actions for and responses to the itema of non-compliance. The attached sheets contain our plans for achieving compliance.
In accordance with our agreement with your representatives, this subtait.tal will deal only with Unit II.
Corrective actions for Unit I will be completed on a schedule commensurate with the restart of that unit.
Sincerely, 6'
R. C. Arnold Sr. Vice President RCA:CFM/lh cc:
J. T. Collins B. H. Grier 1y
- 6040-t Metropo'. tan Ed: son Companns a Member v + 3+ t a Fuc': U: t es System
Atttchmint 1 TLL - 158 CORRECTIVE ACTIONS FOR ITEMS OF NONCOMPLIANCE ITEM 1 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 inoperable by feedwater header isolation valve closure.
(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 feedwater header isolation valve closure during the performance of, and in accordance with, an improper surveillance test procedure.)
The item as stated is an item of noncompliance.
The information provided by the licensee does not provide a basis for modification of the enforcement action. In view of Metropolitan Edison's interpretation of TS 3/4.7.1 and of our conclusions 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).
RESPONSE
A review of Recovery Mode Surveillance procedures will be performed as requested by.he NRC. A list of all the procedures reviewed will be provided along with explanation of problems found and the method by which component / system a
operability is assured when they are removed from service to perform the surveillance (i.e. instrument calibrations).
The target date for completion of this review is June 30, 1980.
ITES 2 The severity and uniqueness of the accident which occurred at Three Mile Island resulted in a marked reduction in the normal good health physics practices 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 rhysics practices and standards may sometimes be mandated by the exigencies that exist during such conditions.
However, the NRC also believes that the licensee, with the resourcec available and taking into account the r*cse frame available fo* conduct of safety-related f unctions, could have tek;n 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 degradation from health physics practices pertaining to control of access to high radiation areas, conduct of radiation surveys, and personnel radiation exposure monitoring.
Atttchment 1 TLL - 158 10 CFR 20.201, " Survey," 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 CFR 20.
10 CFR 20.202, " Personnel Monitoring," requires that the licensee supply 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 "High Radiation Area," requires that each area in which the intensity of radiation is greater than 1000 mrem /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 dose rate monitoring device.
10CFR 20.103, " Exposure of individuals to concentrations of radioactive materials in air in restricted areas," requires in Section (a)(3) that the licensee makes suitable ~ measurements of the concentrations of radioactive materials in air for detecting and evaluating airborne radioactivity in restricted areas for the purposes of determining compliance with the regulation 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 aanner as to cause any individual 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 rem to the hands and forearms, or 7 1/2 rem to the skin of the whole body.
Contrary to the above:
A.
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 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 continu-ously indicated the dose rate; C.
No measurements were made of the concentrations of airborne radioactive materials, in the Unit 2 auxiliary building for periods during which individuals 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 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; D.
On March 29, 1979, an Auxiliary Operator was pernitted to enter areas i
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 operator for this entry. This contributed to the operator receiving a whole body dose of 3 170 rems. When this dose 9
6 6-Attschtsnt 1 TLL - 158 was added to the operator's previous dose for the quarter, the operator's quarterly whole body dose sas 3.870 rems as measured by personnel dosimetry devices; E.
On March 29, 1979, a Nuclear Engineer entered an area of the auxiliary building where the radiation level was greater than that which could be measured 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 whole 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 quarterly whole body dose was 4.175 rems as measured by per-sonnel dosimetry devicea; F.
On March 29, 1979, a Chemistry Foreman was permitted to repeatedly enter high radiation areas and handle samples of highly radioactive reactor coolant. This contributed to the Foreman receiving a whole body dose of 4.100 rems. When this dose was added to tbr loreman's previous dose for the quarter, the Foreman's quarterly whole dy dose was 4.115 rems as measured by personnel dosimetry devices; G.
On March 29, 1979, a Chemistry Foreman and a Radiation Protection Foreman were permitted to handle a highly radioacive reactor coolant sample without adequate personnel monitoring and without first performing a survey of hand and forearm exposure rates. 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 rems; and H.
On March 29, 1979, 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 instru-ment and no evaluation of the dose to the skin of these individuals was made.
The items as stated are items of noncompliance. The information provided in the licensee's response does not provide justification for withdrawing any of the examples of noncompliance cited, nor does it provide justificaticn f or remission or mitigation. of the proposed penalty. Commitments provided for corrective action are incomplete as discussed below:
A.
The commitment for corrective action does not state specific changes to be made to the health physics program to improve access control nor does it state the date when full compliance will be achieved.
RESPONSE
Immediately af ter the accident, a large contractor supplied radiological control technician staff was recruited to maintain access control at entrances to areas where high radiation levels could be encountered.
In addition, any area that was identified as having radiation levels greater than 1.0 Rem /hr was barricaded and/or locked. Keys to these high radiation areas were maintained by radio-logical control foreman. Any entry into these areas was escorted by radio-3-
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TLL - 158 logical control personnel who unlocked the area and performed radiation surveys prior to and during the entry.
This practice continues to date, in accordance with Procedure AP 1050 " Control of High Radiation Areas".
Additionally, lock changes have been made to assure that each door has a unique lock and key. AP 1050 will be revised and incor-porated into the Radiological Controls Procedure Manual for Unit 2.
This revision is currently scheduled for implementation by August 31, 1980.
In November 1979, TMI Unit 2 management initiated changes in the Radiological Control Program in an effort to achieve a strong, ef fective program. The steps initiated and planned are outlined in the Management Plan for TMI 2 Radiological Control Program presented to the NRC in February 1980.
Each action item addressed in this plan was assigned a completion date. Implementation of all changes addressed in this plan is expected by December 31, 1980.
B.
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.
RESPONSE
Emergency _ monitoring equipnent, dedicated to use for emergency situations, will be placed at strategic locations, such as: designated emergency high radiation area control points and at the dosimetry building. This equipment will be maintained in a state of readiness at all times. A limited number of high range equipment (dose rate instruments and self reading dosimeters) will be in place by June 30, 1980.
C.
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 described and no information is provided regarding results achieved due to corrective steps taken. The date when full compliance will be achieved is not specified.
RESPONSE
In addition to continuous monitoring instruments capable of measuring particulate, Iodine, and gaseous levels installed since the accident, there are 26 fixed filter continuous monitors for particulate activity in use within the Unit 2 complex. There are, in addition, 28 portable particulate sampling devices in use within the Unit 2 complex for the purpose of performing grab samples in specific situations. The following summary reflects the quantities and char-acteristics of all air monitoring equipment in use at TMI Unit 2; grab samplers are not addressed.
Attcchmsnt 1 TLL - 158 RANGE INSTRUMENT QUANTITY PERMANENT MOVEABLE PARTICULATE IODINE NOBLE GAS NMC 5
1.0E6 cpm 1.0E6 cpm 1.0E6 cpm 5
Eberline 4
1.0E6 cpm 1.0E6 cpm 1.0E6 cpm 4
Victoreen 13 1.0E6 cpm 1.0E6 cpm 1.0E6 cpm 11 2
Victoreen 4
1.0E6 cpm 4
Eberline 26 100,000 cpm 26 AMS-3 Training in the use of this equipment is included in the qualification program for all radiological control technicians and their foremen as mentioned in the enclosed response to items 2E and 8.
The deficient conditions identified by this item are considered to be corrected by the above described actions.
D.
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.
RESPONSE
Corrective actions described in 2.A 2.B and 2.E are considered to be applicable to this item and adequate to prevent its recurrence.
E.
More ef fective 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 nor does it specify the date when full compliance is to be achieved.
RESPONSE
The actions stated in 2.A and 2.B above are considered applicable to this item and will aid in preventing its recurrence. The Emergency Plan will also be modified to include exposure guidelines in emergency situations. These criteria will apply the guidelines of the NCRP and 10 CRF 20 to the specific phases and situations that may be encountered in an emergency. All personnel qualified by the RWP training and Radiological Control Technician training programs will be instructed in these l
. criteria commencing June 1, 1980. Standards for Radiological Contcol Training for Radiological Control rechnicians and their foremen were developed in December 1979 and training for the current radiological control technicians was initiated ij.
in January 1980 with an expected completion date of July 1, 1980 for all currently employed radiological control technicians. This training program is detailed in
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Attcchmsnt 1 TLL - 158 the Management Plan. for TMI 2 Radiological' Control Program. The implementing procedure. for this training program is expected to be issued by June 30, 1980.
The training program consists of classroom training followed by a written examination,' oral _ examinations which assess the individuals ability to identify
.and respond to unusual / emergency situations, spill drills which measure the individual's ability to react to staged unusual conditions, and practical f actor training which determines the individual's ability to perform required oper-ational tasks. Retraining will be conducted on an annual basis.
The Radiological Safety Training Program for all personnel employed at TMI
, Unit 2 has been revised and is currently in progress.
In addition to classroom instructions, emergency response and practical factor training provides opera-tional training in the radiological considerations applicable to the individual's craft lines / functions. The implementing procedure defining this program has been' developed and-is currently in the review and approval circuit. Similiarly, retraining will be conducted on an annual basis.
F-G. The response states that special handling, tools, shielding, and training of chemistry personnel will be provided; however, this commitment lacks specifics and fails to address the more general area of preplanning for all radiological work. No date is specified for full compliance.
RESPONSE
Radiological (ALARA) Engineers currently review work requests meeting the criteria established by the Radiological Control Department for tasks in areas or on systems having radiological implications.
ALARA engineers are on call on a twenty-four hour basis to perform these reviews. The purpose of these reviews is to assess the radiological conditions and determine the most ef fective manner to perform the task while maintaining personnel exposures as low as reasonably achievable. Techniques considered in reducing exposures are;
[
flushing operations, shielding, special tooling, and the use of containment systems.
In addition to operational techniques, the need for mock-up training and/or working briefings prior to the performance of the task are considered.
Mock-up training, worker briefings and operational techniques were utilized on major evolutions already conducted at TMI Unit 2, such as the reactor building
-sump sampling (401-penetration) operation and currently on the reactor building re-entry program. Procedures documenting the above practices and requirements are currently being developed and are expected ' to be implemented by July 1,1980.
i A program has'been' implemented to review existing chemistry procedures to '
determine their adequacy' from a technica1' and ALARA view point and 'to provide additional training for chemistry personnel.
The construction of a new Temporary Sampling System is nearing completion within Unit-2 that will replace the need to take Unit-2 samples in the Unit-1 l
Chemistry Laboratory..This system is designed for high activity samples through l
. the use of: shielding, valve' handwheel extensions, compact piping and sink arrangement. Additionally, the system' design and operating procedures have
- undergone detailed ALARA reviews.
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TLL - 158 1
When construction is complete and startup testing is in progress, the chemists will receive formal training on the system that includes ALARA considerations
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with the ultimate objective of further reducing dose rates during sampling evaluations. : Training is also being provided to Radiological Controls personnel in. handling the, types of samples that will be taken at this sink.
i H.-
The response to' item 2.H admits-noncompliance and states that dose evaluations have been completed and ' reports made to the NRC as required. No specific corrective steps were specified for assuring more prompt evaluation of personnel contamination in the future.
i
RESPONSE
Instrumentation necessary for rapid evaluation of-personnel contamination is currently available and radiological control technicians have been trained in l
its usage. Documentation of skin contamination is being accomplished in accor-dance with an existing procedure HPP 1612 " Monitoring for Personnel Contamination".
l A new procedure describing the evaluation, handling, and documentation of skin contaimination situations is currently being developed and is scheduled i
for implementation prior to September 1,1980. Thumbrules for rapid evaluations (for reaction purposes only) have been developed as field use tools for techni-cians during unusual / emergency situations.
These thumbrules have been introduced e
to_the technicians during emergency response training sessions.
Formal eval-uations of personnel exposures resulting f rom skin contaminations are and will be performed by professional and Technical individuals within the Radiological Technical Support-and Radiological Support Services groups of the Radiological Control Department.
The 'above described conditions are considered to be adequate corrective actions for the cited deficient conditions.
1 ITEM 3 Technical -Specification 6.5.1, " Plant' Operations Review Committee," requires F
in Section 6.5.1.6.a. that the Plant Operations Review Committee (PORC) review all procedures (and changes thereto) required by Technical Specification v.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, 4
4 and Procedure Chnnge -Request No. 2-78-895, Revision 8 to Surveillance Procedure j
2303-M14A/B/C/D/E; both were reviewed-and approved by the PORC (November 9, i
1978 and, August 15,'1978 respectively). Each approved change included a valve lineup which resulted in emergency feedwater header isolation, contrary to Technical Specification 3/4.7.1 requirements.
i The information provided j
The item, as stated, is an' item of non:ompliance.
i by the' licensee does not provide a basis for modification of the enforcement action.
7'-
Attrehaant 1 TLL - 158 The licensee should address in a supplemental response the actions to be taken to assure PORC members have the necessary technical expertise to demonstrate a clear understanding of the implications of TS requirements and system oper-ability requirements as stated in the TS and FSAR. The specific further examples of similar test procedures contained in the response of the licensee should be included in the review or procedures planned by the licensee.
The licensee should also address an appropriate target date for the completion of these reviews.
RESPONSE
The Unit 2 PORC has undergone major changes since the March 28, 1979 accident.
The major changes are as follows:
(1) A Supervisor-Technical Specification Compliance position has been estab-lished. A primary responsibility of this position is' that' of serving as full time PORC Chairman. Day to day involvement with Technical Specifi-cation matters creates an inherent "cch. Spec. expertise.
(2) The new Unit 2 Recovery Tech. Specs. implement revised PORC membership requirements.
Specifically, the Plant Operations Review Committee shall be composed of the:
(a) Chairman - who shall have an academic degree in engineering or physical science field and a minimum of five years of applicable experience.
(b) 1 Member - who shall meet or excead the qualifications of Regulatory Guide 1.8, September 1975.
(c) 7 Members - who shall meet or exceed the qualifications of Section 4.4 of ANSI N18.1 - 1971.
The PORC memberste y has been significantly changed to include a broader
' spectrum of expet;ise and background.
(3) The PORC review philosophy has evolved to emphasize the broader safety questions, Tech. Spec. compliance, CFR compliance, etc., concerns. This approach has. iready resulted in a marked improvement in PORC reviews.
(4) The Mode 7 Recovery Tech. Specs. have been recently issued by the NRC.
To' assure a clear understanding of the implications of Tech. Spec. require-ments and system operability requirements all PORC members have been provided a copy of - the Tech. Specs. They will be required to document their review and understanding. Further, as Tech. Spec. changes occur PORC is
. required to review the changes prior to submittal to the NRC. Therefore, there is an inherent mechanism for keeping abreast of changes to the Tech. Specs. This documentation is now complete.
Attschzant 1 TLL - 158 ITEM 4.A Technical Specification 6.8, " 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.B.1 that electromatic _ relief isolation valve RC-R2 be closed if, among other things, the valve discharge line temperature exceeds the normal 1300F.
Contrary to the above, the electromatic relief valve discharge line temperature had been in the range of 1800-2000F 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 2830F 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.
The item, as stated, is an item of noncompliance. The information presented by the licensee does not provide a basis for modification of the enforcement action.
The corrective actions proposed 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 determination 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 supple-mental 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 staff use.
Additionally, the actions required upon identification of " symptoms" should be included in this response.
RESPONSE
Management has recently issued a policy statement regarding compliance with Operation and Maintenance procedures. Additionally, a series of meetings are being conducted with Operations Department personnel to address the need to comply with procedures and personnel responsibilities for identifying when procedures require revisions.
Instructions were issued to all appi cable personnel enforcing management's i
position on February 15, 1980.
ITEM 4.B Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that i
procedures be established, implemented and maintained covering identified activities.
t l 1
f' Attcchnsnt 1 TLL - 158 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 (1500 psig), and that the operator verify high pressure injection is operating properly 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 maintained above the renet point (1540 psig) or until low pressure injection flow is established at 3000 gpm.
Contrary to the above:
1.
At about 0405 on March 28, 1979, high pressure injection flow was throttled to minimum conditions even though RCS pressure was less than 1600 psi and falling, and without low pressure injection flow established.
2.
At various times throughout the day of March 28, 1979, the high pressure injection system was modified such that the required flow rates Smre not maintained during continuing low 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 heat.
B.2 Emergency Procedure 2202-1,3, " Loss of Reactor Coolant / Reactor Coolant System Pressure," Revision 11, requires certain actions to be taken following the automatic initiation of high pressure injection, including in Section B.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 con-ditions within the RCS, the Core Flood System was removed from its ESF position (rendered inoperable) by closing both tank isolation valves. [This
, portion of the ESF was 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 depressurize to activate the decay heat removal system NRC recognized that it was necessary to isolate the core flood system and encouraged this action. This citation does not apply to isolation durf ag this attempt.]
Item 4.B.1, as stated, is an item of noncompliance. The corrective actions proposed by the licensee appear adequate to preclude recurrence.
These procedure reviews and improvement will be subject to review during evaluation of the restart proposal for Unit 1 and, at a later date, Unit 2.
Attcchment 1 TLL - 158 Item 4.B.2, as stated, is an item of tsncompliance. The licensee should address in a supplemental response those measures to be taken to insure that.the operability requirements of Engineered Safety Features are met during all phases of operation.
The information provided by the licensee for Items 4.B.1 and 4.3.2 does not provide a basis for modification of the enforcement action.
RESPONSE
The proposed corrective action for Item 4.B.1 of our December 5,1979 submittal was considered acceptable.
The Core Flood Valves which are the subject of this infraction are not required to be operable in the current Technical Specification for Unit II.
Administrative Procedure No. 1012 requires a shift ES checklist to be completed to reflect current plant status of ES ccmponents.
This checklist is not required to be filled out currently on Unit II because of current plant status, 1250F (T Average).
However, Unit 11 currently is utilizing a procedure, "Shif t & Daily Checks,"
Procedure 4301-S1, which is used to verify that those systems / components required by the current Technical Specifications are in the necessary State of Readiness.
Additionally, Operations Department personnel have been instructed to comply with approved procedures.
The action required by this Item is considered complete.
ITEM 4.C Technical Specification 6.8 " Procedures," requires in Section 6.8.1 that procedures be established, implemented and maintained covering identified act ivities.
Operating Procedure 2104-6.2, " Emergency Diesels and Auxiliaries, " Revision 9, established 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 shutdown after the Engineered Safeguards Feature actuation has been cleared.
2.
Section 4.6, " Diesel Generator 1A(1B) Shutdown to Emergency Standby,"
states in the closing step, 4.6.6, to place the diesel generator on standby in accordance with Section 4.2; and
.3.
Section 4.2, when completed, established conditions for automatically starting the diesels upon actuation of an Engineered Safeguards Feature
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(ESF) including requirements to place the " Emergency Standby / Maintenance Exercise" switch in the Emergency Standby position and resetting the fuel racks.
Attcchasnt 1 TLL - 158 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 28,19 79, both,1A and 1B diesel generator fuel racks were manually tripped, thereby prevqnting an automatic start of the diesel generators upon ESF actuation and manual start from the control room until 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br />.
i
RESPONSE
In a letter dated December 5,1979, Metropoitan Edison proposed certain cor-rective actica for this item. In the March 6, 1980 meeting with the NRC, NRC did not require any additional action.
ITEM 4D 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 aenerator levels at 30 inches.
Contrary to the above, fron 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 steam generator decreased to 10 inches (the minimum level indication) while the A steam generator level was being controlled manually.
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 thir. 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 response the actions to be taken, including procedural improvements, to estab-lish the required steam generator water level in all modes of feedwater or emergency feedwater addition.
RESPONSE
Procedures establishing required Steam Generator Water Level for the current operational mode are in place.
ITEM 4.E Technical Specification 6.8, " Procedures," requires in Section 6.8.1 that procedures be established, impicmented and maintained covering identified activities.
e
c Attcchnont 1 TLL - 158 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 /Shif t Supervisor / Unit Supt!- Technical Support in the Control Room will, af ter revie41ng the emergency conditions, classify the emergency as one of the following:
a.
" Personnel or Local Emergency",
b.
" Site Emergency", and c.
" General Emergency" "He will make this classification accord?ng to the condition of Table 1 of this plan, and initiate actions according to the Emergency 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 ben thcre is a reactor building high range gamma monitor alert al
.m (Condition No. e).
Contrary to the above:
1.
Adequate written procedures were not established and implemented in that Section 2.1 of Procedure 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 though primary system pressure had decreased to the point where safety injection was automatically initiated and a reactor building sump high level alarm existed; and 2.
A site emergency was not dec1cred 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 time Condition "e" of Three Mile Island Emergency Plan 1004 had occurred.
Item 4.E.1, an stated, is an item of noncompliance.
Item 4.E.2 is withdrawn.
The Civil Penalty is partially remitted in.he amount of $2,000.00. The corrective action specified is incomplete in that the date full compliance is to be achieved is not specified. A supplemental response is requested to provide this information.
RESPONSE
Unit 2 is revising its Emergency Plans to incorporate revised criteria such as those contained in NUREG-0654. The revised plan will be implemented con-sistent with the implementation of the Unit 1 plan which has already been submitted to the NRC.
Attcchmsnt 1 TLL - 158 Drills are now being conducted on a quarterly basis in Unit 2.
When the new plans and implementing procedures are approved for use, drills and gmergency preparedness will be conducted in accordance with them.
i It is expected that the new plans will be phased in during the month of July 198a. A drill is being planned in concert with the NRC, PEMA, BRP and others.
The drill is scheduled for July 16, 1980 and the entire emergency management organization will be exercised.
Notification and emergency action level criteria per NUREG-0610 (also 0654) have been implemented as of February 22, 1980 as an interim measure until the new plan is in effect. This interim instruction also impic=ents the new notification criteria per 10CFR 50 72.
Unit 2 is now considered to be in full compliance.
ITEM 4.F 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 programs for these categories will be conducted on an annual (calendar year) basis; and 2.
Describes in Section 3.1.1 through 3.1.9, the training program f or all on-site emergency job categories.
Contrary to the above, during calendar year 1978, not all individuals having emergency responsibilities were trained in that two Emergency Directors, one Accident 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 an emergency, at least four individuals who were assigned as required members of a Radiological Monitoring Team and
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seven individuals who were assigned as required members of a Repair Party Team performed emergency duties for which they were not trained.
Although the licensee in its response to item 4.F admits noncompliance and agrees to pay the Civil Penalty, the licensee seemingly minimizes the signifi-cance of incomplete emergency training by emphasizing the amount of training which was performed and implying that the incomplete portion did not have a significant adverse af fect on performance. The NRC believes that many of the problems associated with the licensee's health physics performance following the accident could have been prevented by more effective training in this area. The commitment for. corrective action is acceptable.
RESPONSE
The proposed corrective action of Item 4.F of our December 5,1979 letter which was considered acceptable, is currently in progress..
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a Atttchment 1 TLL - 158 ITEM 4.C Tech,nical Specification 6.8, " Procedures," requires in Section 6.8.I that proc.edures be established, implemented and maintained covering identified activities.
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Station Administrative Procedure 1002, " Rules for the Protection of Employees Working on Electrical and Mechanical Apparatus," Revision 14, requires in Sections 4.3, 4.4, and 4.5 that on restoration of equipment to service, removed tags will have all required information entered thereon and then be suitably stored, and that the shif t foreman shall approve equipment operation by signing the original tagging application. Additionally, Station Corrective Maintenance Procedure 1407-1, Revision 0, specifies in Section 5.0, " Job 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 things, documentation of corrective action taken (resolution description and certification of satisfactory post maintenance testing) and Station Preven-tative Maintenance Procedure E-2, " Dielectric Check of Insulation, Meters and Cables," specifies how to make the measurements 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-V12A,12B, 32A, 32B, 33A, and 33B). No suitable documentation to determine whether the maintenance work had been completed, tags removed, acceptance criteria met, or valves approved for operation could be found. The TMI-2 maintenance log lists this work request as being in an open status as of June 20, 1979.
The licensee admits that this is an item of noncompliance, and the corrective actions proposed and in force appear adequate pending site followup.
RESPONSE
The ' proposed corrective actions of Item 4.G of our December 5,1979 letter was considered acceptable.
ITEM 5 Technical Spec'ification 6.8, " Procedures" requires in Section 6.8.2 that changes to procedures which implement the Emergency Plan shall be reviewed by. the Plant Operations Review Committee and approved by the Unit Superintendent prior to Laplementation.
Contrary to the above, a change to Station Health Physics Procedure 1670.7,
" Emergency Assembly, Accountability and Evaluation," was made without the required review and approval. An -3ditional assembly area was designated and. the method used to perform accountability was modified by a memorandum dated October 13, 1978, from the Radiation Protection Supervisor to all depart-ments. As a result, on March 28, 1979, in response to an emergency, some licensee -
Attcchmrnt 1 TLL - 158 personnel fciloved the approved procedure while others tollowed the guidance in the October 13, 1978 memorandum, creating some confusion and delaying prompt attainment of full accountability.
The response admits noncompliance but requests remission or mitigation based on the licensee's belief that this item did not delay prompt attainpent of personnel accountability or c:use confusion. The investigators concluded, based on three interviews with site security personnel, that delay and confysion did not result from this improper procedure change.
See page 11-1-21 of the InvesC gation Report. Regardless of this, the Civil Penalty was based primarily on the fact that Procedure 1670.7 was changed without the required review and approval of the Plant Operations Review Committee and not on whether delay and confusion resulted. The commitments for corrective action are acceptable.
RESPONSE
The proposed corrective actions of Item 5 of our December 5, 1979 letter was considered acceptable.
ITEM 6 Environmental Technical Specification 5.7 requires that detailed written procedures for instrument calibration be prepared and followed.
Three Mile Island Nuclear Station Surveillance Procedure 1302-5.24, Revis ion 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.
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 inspection report. The fact that the procedure in question is a Unit 1 procedure is irrelevant since it applied to instrumentation common to Units 1 and 2.
Regardless of statements made by vendors, NRC considers that calibration of environmental air samplers is needed and is required at TM1 by Environmental Technical Specification 5.7.
Upgrading an unresolved item to an item of noncompliance is consistent 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 commitment for instrument calibration and does not specify the date by which full compliance will be achieved.
The item as stated is an admitted item of noncompliance. The information provided by the licensee does not provide a basis for modification of this enforcement action. Th licensee is requested to submit a supplemental response addressing the areas described in tue above evaluation.
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Attrchmmnt 1 TLL - 158 Resp 0NSE Calibration of the of f-site continuous air samplers is done on an annual basis during the first week of February. It is performed by the Instrumentation and Control Group.
The procedure followed is Surveillance Calibration Procedure 1302-5.24; this procedure is currently being revised and is now in the review process.
The procedure is expected to be approved by June 1,1980. Important points to note about the calibration procedure are:
(1) The flowmeter is not a precise instrument and errors of 20% are possible in the field. As a result this portion of the sampler is not calibrated but replaced.
(2) The flow totalizer is essentially a home gas meter.
It is calibrated against a standard meter which comes from the manufacturer with specifi-cations of calibration.
(3) The timer is calibrated against a stopwatch.
ITEM 7 Technical Specification 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 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 in Figure 6.2-1 in that:
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 Protection and Chemistry,"
reported to this position rather than directly to the " Station Superin-tendent/ Senior Unit Superintendent," and B.
There 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 rather than directly to the " Station Superintendent (Station Manager / Senior Unit Superintendent);" and D.
The position of " Chemical Supervisor" had been vacant since the issuance of the Technical Specifications.
On March 28, 1979 through March 30, 1979, the above organization discrepancies decreased the effectiveness of the licensee's response to the accident.
Attechm2nt 1 TLL - 158 The response admits noncompliance but requests remission or mitigation of the proposed penalty based on the licensee's belief that the organizational changes did not adversely affect its response to the accident and op its belief that the cited changes were discussed with NRC on March 5, 1979 Although it appeared to the investigators that differences between the actual organization and _the organization assumed by emergency plan implementing procedures did reduce ef fectiveness of the licensee's response to the accident, the NRC recognizes that this conclusion is somewhat subjective and acknowledges that these organizational differences may not have had a significant ef fect on response. In view of the above, the $3000 penalty proposed in the original Notice of Violation for this item of noncompliance was selected from the bottom of the monetary scale ($3000-4000) generally followed in the assessment of Civil Penalties for inf ractions by power reactor licensees. The more important concern here is the licensec's f ailure to obtain approval of new Technical Specifications prior to making its organization changes. NRC Region I does not recall discussing with the licensee the organizational changes cited in this item of noncompliance. In any event, the licensee's organizational changes were contrary to the licensee's existing Technical Specifications and should not have been made prior to obtaining an amendment to these Technical Specifications. As the licensee is surely aware, the Commission's regulations specifically provide that changes to Technical Specifications shall be made through che formal amendment process, not through methods of the licensee's own choosing. See 10CFR 50.59(c). The corrective actions proposed and under-way appear adequate pending NRC completion of its review.
RESPONSE
The proposed corrective actions of Item 7 of our December 5,1979 letter were considered acceptable.
ITEM 8 Technical Specification 6.4, " Training," requires that a retraining and replace-ment training program for the unit staf f 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 N18.1-1971 recommendations had not been maintained for members of the radiation protection and chemistry staff in that only 2 of the 10 topics recommended were included in the program.
The item as stated is an item of noncompliance. The information provided by the licensee does not provide a basis for modification of this enforcement action.
A supplemental response is requested to provide more specific training commit-ments as discussed above.
RESPONSE
The ten items in Section 5.5 of ASSI N18.1-1971 referenced in this item are Attcchmant i TLL - 158 specific to Operator Replacement and Requalification Training. This is further amplified in the 1978 edition of the standard.
Radiological Control HP Technicians The Radiological Control Technicians qualification program, described in the response to Item 2E of this attachment, is an annual recurring program. Training on the ten items listed in Section 5.5 of ANSI N18.1-1971 will be presented as applicable to conditions present at TMI Unit 2 and only in the detail necessary to the performance of the Radiological Control Technicians' duties and respon-sibilities as an intergral part of this training.
Chemistry Technicians A Chemistry Technician Program for Units 1 and 2 is being developed that is job related. It will contain a program for newly-hired technicians and incum-bent Technicians / Foremen similar to the HP Programs.
This program will contain lectures in basic theory, systems and procedures including instrumentation necessary to prepare and maintain chemistry personnel proficient in their assigned job.
The ten Items in Section 5.5 ANSI N18.1-1971 will be addressed to give the Technicians a general understanding of the plant operations with specific emphasis on his rol. in the overall plant evolution.
This program will be available to commence on or before June 1,1980.
ITEM 9 Technical Specification 3/4.4.6, " Reactor Coolant System Leakage," requires in Section 3.4.6.2, that Reactor Coolant Sfstem (RCS) leakage be limited to 1 gallon per minute (GPM) of " Unidentified Leakage," and that unless rates above this limit are reduced 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 thirty hours.
(_n' rary to the above, from March 22, until March 28, 1979, RCS " Unidentified Ler /. age" remained above 1 gpm and the plant was not placed in " Cold Shutdown."
The licensee admits the item of noncompliance. The corrective actions proposed and underway appear adequate.
RESPONSE
The proposed corrective actions of Item 9 of our December 5,1979 letter was considered acceptable.
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Attschment 1
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TLL - 158 ITEM 10 10CER 20.4 1, " Records of surveys, radiation monitoring, and disposal," requires in Section (a) that each licensee maintain records showing the radiation expo-sure 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.
The results of approximately 500 ground level radiation surveys conducted during March 28-30, 1979 in offsite areas bordering the Three Mile Island site were not documented in a manner which permitted a precise evaluation of the type of radiation (Beta / Gamma) which existed in the environs.
Pertinent information such as the type of instrumentation used and whether the end windew oa the probe was open or closed vas not recorded.
B.
The records of the radiation exposure for at least 5 individuals exposed during the period March 1 to 31, 1979 had not been recorded or maintained on a form NRC-5, or equivalent, as of July 5, 1979. Further, as of July 5,1979 the assessment of their doses had not been completed.
10-A The response 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 hampered the real time as well as the historical evaluation of radiological conditions.
Although the licensee states that it was possible to reconstruct the full survey information from the original radioed survey results, the NRC investigation determined that the survey records were inadequate for the reasons stated in NUREG-0600, page 11-3-97.
The commitment for corrective action is acceptable except that the date when full compliance will be achieved is not specified.
RESPONSE
The proposed corrective action for Item 10A (the development of survey forms to improve the quality and clarity of future off-site survey records and including these forms in emergency kits and other locations as appropriate) will be completed by July 1, 1980.
10-B The response admits that example 10B is noncompliance and requests miti-gation or remission of the proposed penalty based on the unusually large number of records generated and heavy demands on the individuals processing these records. NRC recognizes that maintaining accurate records was Attcchssnt 1 TLL - 158 difficult under the circumstances; but this difficulty is not justification for the failure of the licensee to identify and assess the doses of indi-viduale who were known to have significant exposures. The commitment for corrective action is acceptable.
I REESPONSE The proposed corrective action of Item 10.B of our December 5,1979 letter was considered acceptable.
ITEM 11 10 CFR 50, Appendix B, Criterion X, " Inspection," requires that a program 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 2.15.Section X, requires that the inspection program include random observation of operations and functional testing by individu,.ls indepen-dent of the activity being performed.
Procedure GP 4014, "QA Surveillance Program," Revision 0, requires independent observation of activities af fecting quality to verify cor.formance with estab-lished requirements utilizing both inspection and auditing techniques.
.for compliance with written procedures and the Technical Specifications.
Contrary to the above, as of March 22, 1979, the nornal operations surveillance testing acti71 ties had not been madr subject to random and/or routine inspections by independent methods.
This item of noncompliance is withdrawn; the associated Civil Penalty is remitted. Metropolitan Edison stated in i.s response that it is planning to expand its program for inspection of surveillance testing activities.
In view of this, a supplemental response is requested which addresses the specific inapection of activities as they are perfor.rud.
RESPONSE
The Operational Quality Assurance Plan for Three Mile Island Nuclear Station, Revision Eight and the Three Mile Island Unit Two Recovery Quality Assurance Plan, Revision Zero are the two documents which will describe the specific
' requirements to be used by the Licensee in its program for inspection and monitoring of surveillance testing activities.
Both of these plans presently are in final management review cycle and are subject to regulatory acceptance.
Both these plans contain a description of an Operations Quality Assurance Monitoring Program and independent group which will have primary responsiblity to perform reviews and monitoring of surveillance testing activities. These monitorings and reviews are in addition to the Quality Control inspection
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umsu ueuas n TLL - 158 witness and hold points which were being performed prior to the Unit Two accident on both units. Monitors will be qualified in accordance with a documented Q. A. Department procedure that insures that they are kno$iledgeable in the activities they are monitoring to the extent that they can rhadily verify conformance or compliance of the activity being performed. Use of SRO capable individuals and experienced technical personnel is emphasized in the program. Monitoring reports will be distributed to supervisory or managerial personnel that have responsibility for the performance of the activity and nonconformance documents will be issued when nonconformances are identified so that appropriate corrective ' action, including that to prevent recurrence, can be taken by management. Records of the program will be maintained to provide adequate confirmation of the program.
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