EA-97-533 - Three Mile Island 1 (GPU Nuclear Corp.)

January 27, 1998

EA 97-533

Mr. James Langenbach
Vice President and Director, TMI
GPU Nuclear Corporation
Three Mile Island Nuclear Station
Post Office Box 480
Middletown, Pennsylvania 17057-0191

(NRC Integrated Inspection Report No. 50-289/97-09)

Dear Mr. Langenbach:

This refers to the inspection conducted between September 7, 1997, and November 1, 1997, at the Three Mile Island Nuclear Station in Middletown, Pennsylvania, the findings of which were discussed with members of your staff during an exit meeting on November 13, 1997. During the inspection, apparent violations were identified related to your activities during the 12th refueling outage. The inspection report addressing these issues was previously forwarded to you on December 2, 1997. On December 22, 1997, a predecisional enforcement conference (conference) was conducted with you and members of your staff, to discuss the violations, their causes, and your corrective actions.

Based on the information developed during the inspection and the information that you provided during the conference, three violations of NRC requirements are being cited and are described in the enclosed Notice of Violation (Notice). The circumstances surrounding the violations are described in detail in the subject inspection report. The violations involve: (1) inadequate post-maintenance testing following replacement of the pressurizer power operated relief valve (PORV), in October 1995, that resulted in failure to detect that the PORV actuation circuit was miswired rendering the PORV inoperable; (2) failure to follow procedures when filling the reactor coolant system (RCS) on October 5, 1997, that resulted in an uncontrolled spill of water from the control rod drive mechanism (CRDM) vents; and (3) inadequacies in the procedure for the control of radioactive (hot) particles that resulted in a worker receiving a significant skin exposure on October 4, 1997.

The most significant violation involved the inoperable PORV. During rewiring of the PORV actuation solenoid, following replacement of the PORV during the 11R refueling outage in October, 1995, the terminal connections on the solenoid were not clearly marked. Nonetheless, neither the technician who landed the leads, nor the technician that independently verified the wiring, stopped and positively determine the correct terminal locations. Instead, both technicians made incorrect assumptions as to the terminal locations. As a result, the PORV was miswired and would not have opened in response to a manual or automatic actuation signal.

The failure to perform adequate post-maintenance testing following replacement of the pressurizer PORV resulted in this condition not being identified. Specifically, following the incorrect wiring of its actuation solenoid, no test was performed to ensure that the PORV would open in response to an automatic or manual actuation signal. This failure constitutes a violation of the Technical Specification (TS) requirement to perform in-service testing. At the conference, you indicated that the failure to perform the post maintenance test (PMT) was due to procedural and work scheduling inadequacies. Specifically, no PMT checkoff was provided in the PORV replacement and inspection procedures, and there was incomplete guidance in the job order package to direct the performance of the PMT.

The inability to open the PORV would have prevented it from performing its pressure relief function either during power operations or during low temperature conditions during heatup and cooldown. Even though the pressurizer safety valves (the primary pressure relief system), were available to provide overpressure protection during power operations, and administrative controls were in place to provide low temperature overpressure protection, the diversity provided by the PORV for these functions was not available for the entire operating cycle, a period of 23 months. Additionally, the PORV would not have been available to provide a bleed path for high pressure injection (HPI) cooling or to depressurize the RCS to establish long term decay heat removal following a steam generator tube rupture. The unavailability of the PORV for pressure relief or HPI cooling had potential consequences in that it resulted, as determined by your own calculations, in a 16% increase in the TMI core damage frequency, if an event occurred needing the PORV to be opened. This was preventable if requirements for post-maintenance testing had been met. Therefore, the violation has been categorized at Severity Level III in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $50,000 is considered for the Severity Level III violation that occurred prior to November 12, 1995. Because your facility has been the subject of escalated enforcement actions within the last 2 years, (1) the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Credit was warranted for identification because your staff identified, during the 12R refueling outage, that the PORV had been miswired and that no PMT had been performed following the 11R refueling outage. Credit was also warranted for corrective actions because your actions were considered both prompt and comprehensive. Those actions included: (1) communication of management expectations for self-checking, independent verification, and performance of post-maintenance testing; (2) planned revisions to the PORV maintenance procedure to clarify the PMT requirements, and to the job order program to include the vendor manual wiring diagrams in the job order package; (3) review of other work packages to ensure that all required PMTs had been performed; and (4) plans to perform a process study to identify and correct weaknesses in the PMT program.

Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized, after consultation with the Director, Office of Enforcement, not to propose a civil penalty in this case. However, significant violations in the future could result in a civil penalty.

With respect to the overfill of the RCS, the shift supervisor (SS), who was supervising the fill and vent of the RCS in October, 1997, believed that there was not enough water available in the reactor coolant bleed tank (RCBT) to complete the intended evolution. Although the SS appropriately consulted his supervisor and was told that there was sufficient water available in the RCBT, the SS, still believing that there was insufficient water available, used an inappropriate procedure to fill the RCS from the borated water storage tank (BWST). Other control room operators did not question the SS's decision. Your staff failed to adhere to the limitations provided in the decay heat removal (DHR) system operating procedure (OP) when they used the DHR pumps to provide makeup to the RCS directly from the BWST during the fill and vent of the RCS. Additionally, they failed to follow the RCS fill and vent procedure when they failed to terminate the RCS fill at the required point. As a result, borated water spilled onto the reactor vessel head and control rod drive (CRD) components, potentially degrading those components and creating a radiological condition warranting remediation. While this violation is classified at Severity Level IV given the significance of the occurrence, it raises concerns regarding the questioning attitude of the staff and management's expectations for adherence to procedures. At the conference, you indicated that the problem was that the SS failed to comply with normal work practices specified in your conduct of operations administrative procedure (AP), rather than a failure to adhere to the RCS fill and vent procedure or the DHR system operating procedure. The NRC is concerned that plant management may not be providing a high standard for procedure adherence and may be providing operators with the impression that it is acceptable to use procedures that were not specifically prepared to support an activity.

Finally, with respect to the inadequate hot particle control procedure, an emergent hot particle area was discovered during surveys of newly exposed surfaces upon raising the reactor vessel head seal plate following work in the fuel transfer canal in October, 1997. Upon discovery of these conditions, the radiation control technician (RCT) assigned to the job elected to proceed without consulting supervision. Although the area was subsequently decontaminated, the surveys that were performed following the decontamination were not adequate to verify the removal of the hot particles. Additionally, a hot particle control area was not formally established. Your radiological protection (RP) procedure RP for hot particle controls was inconsistent with 10 CFR 20.1501 in that it did not provide sufficient direction to assure that adequate surveys were performed and that adequate hot particle controls were established. This constituted a violation of Technical Specification requirements for the radiation protection program which require that procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20. As a result of the inadequate surveys and lack of sufficient hot particle controls, a worker received a calculated dose of approximately 14 rem to the skin. While the violation is classified at Severity Level IV, the NRC is concerned that, when it was determined that hot particles were present, an evaluation, to determine the quantities and magnitude of the hot particle contamination, was not performed. Consequently, an appropriate interval for personnel frisking for hot particles was not established. Without these controls, there was a potential for skin exposures even more significant than the exposure that occurred.

With respect to the apparent violation involving the failure to follow procedures for the once-through steam generator (OTSG) locked high radiation area that was discussed at the conference, the NRC concluded that the worker that left the OTSG manway area unattended with the high radiation area door unlocked failed to adhere to the requirements of your locked high radiation area AP. However, based on the information provided at the conference and during subsequent telephone conversations with Mr. Etheridge of your staff, the NRC concluded that the potential for inadvertent entry into the high radiation area was low. The manway opening was continuously monitored at a remote location with a video camera and the individual monitoring the opening by camera was in direct communication with personnel in the close proximity of the unlocked manway. Therefore, because it was licensee-identified; was corrected immediately; and was not repetitive within the last two years, the violation of the locked high radiation area AP will not be cited in accordance with Section VII.B.1 of the Enforcement Policy.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response will be placed in the NRC Public Document Room (PDR).


Hubert J. Miller
Regional Administrator

Docket No. 50-289
License No. DPR-50

Enclosure: Notice of Violation

cc w/encl:
J. Fornicola, Director, Nuclear Safety Review
M. Ross, Director, Operations and Maintenance
D. Smith, PDMS Manager
TMI-Alert (TMIA)
M. Laggart, Manager, TMI Regulatory Affairs
E. Blake, Shaw, Pittman, Potts and Trowbridge (Legal Counsel for GPUN)
Commonwealth of Pennsylvania


GPU Nuclear Corporation
Three Mile Island Nuclear Station
Docket No. 50-289
License No. DPR-50
EA 97-533

During an NRC inspection conducted between September 7 and November 1, 1997, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:


Technical Specification 4.2.2 requires that in service testing (IST) of ASME Code Class 1, Class 2, and Class 3 valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code). The ASME Code and OMa-1988, Part 10, paragraph 3.4, requires that, when a valve or its control system has been replaced, an in-service test must be performed prior to returning the valve to service.

Contrary to the above, on October 31, 1995, the pressurizer power operated relief valve (PORV), a Class 1 valve, was returned to service without performing an IST to verify proper valve operation after the PORV was replaced. As a result, a wiring error, that prevented the PORV from opening in response to an automatic or manual signal, was not detected. Consequently, the PORV was inoperable for the operating cycle from October, 1995, until September, 1997. (01013)

This is a Severity Level III violation (Supplement I).


Technical Specification (TS) 6.8.1 requires, in part, that Written procedures be implemented covering the applicable procedures recommended in Appendix `A' of Regulatory Guide 1.33, Revision 2, February 1978. Regulatory Guide 1.33, Appendix `A', section 3.0 recommends, in part, instructions for filling and venting the reactor coolant system (RCS) and for operation of decay heat removal systems.

Operating procedure (OP) 1103-2, "Fill and Vent of the Reactor Coolant System," section 3.1.2, step 17.c, requires, in part, that when the level at the center control rod drive mechanism (CRDM) is observed at one to two feet below the top, terminate the RCS fill and hold level.

OP 1104-4, "Decay Heat Removal System," section II of Enclosure 2, "Make Up to the RCS Directly from the BWST," provides a caution that make up to the RCS directly from the borated water storage tank (BWST) must be carefully monitored since large volumes of water can be transferred very rapidly. Step 1 of section II states, in part, that controlling the level in the RCS using this method is not considered to be, nor should it be, used as a major RCS fill and vent method.

Contrary to the above, on October 15, 1997, the licensee failed to properly implement operating procedures 1103-2 and 1104-4 while filling and venting the RCS following a refueling outage. Specifically, while filling the RCS from the reactor coolant bleed tank (RCBT) in accordance with OP 1103-2, make up to the RCS was established directly from the BWST, contrary to the instructions in Enclosure 2 of OP 1104-4. The additional makeup caused a prompt rise in pressurizer level. Even though the operators observed the level increase in the control room and terminated the RCS fill from the RCBT, the makeup from the BWST was not immediately terminated due to communications difficulties. Consequently, approximately 50 gallons of RCS water overflowed out of the CRDM vents onto the reactor vessel head area. (02014)

This is a Severity Level IV violation (Supplement I).


Technical Specification 6.11, Radiation Protection Program, requires that procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

10 CFR 20.1501 requires that each licensee shall make or cause to be made, surveys that may be necessary for the licensee to comply with the regulations in 10 CFR 20 and are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive material, and the potential radiological hazards that could be present.

Contrary to the above, as of October 4, 1997, the licensee's hot particle control procedure, Procedure 6610-ADM-4110.04, was inconsistent with 10 CFR 20.1501, in that it did not cause surveys to be made to assure compliance with 10 CFR 20.1201(a)(2)(ii), which limits radiation exposure to the skin. Specifically, the procedure did not provide sufficient direction to assure that surveys to verify elimination of hot particles following decontamination of newly exposed surfaces upon raising the reactor vessel seal plate, were adequate to evaluate the potential radiation hazards. As a result, the hot particles were not sufficiently removed such that the area did not require hot particle controls, nor were hot particle controls established. Consequently, due to a hot particle, a radiation worker received a skin exposure of approximately 14 rem, 30% of the 10 CFR 20.1201 annual limit of 50 rem. (03014)

This is a Severity Level IV violation (Supplement I).

Pursuant to the provisions of 10 CFR 2.201, GPU Nuclear Corporation is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region I, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at King of Prussia, Pennsylvania
this 27 th day of January 1998

1. e.g., A Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $210,000 was issued on October 8, 1997 (EAs 97-070, 97-117, 97-127, and 97-256), for numerous violations related to several areas of plant performance including engineering design controls, classification and environmental qualification of components, corrective actions, and emergency preparedness.

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