United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 86-75: Incorrect Maintenance Procedure on Traversing Incore Probe Lines

                                                            SSINS No.: 6835 
                                                            IN 86-75       

                                UNITED STATES
                           WASHINGTON, D.C. 20555

                               August 21, 1986

                                   TRAVERSING INCORE PROBE LINES 


All nuclear power reactor facilities holding an operating license or a 
construction permit. 


This notice is to inform licensees of a potential generic problem concerning
maintenance procedures on traversing incore probe (TIP) systems having TIP 
process lines that penetrate the containment. The problem arises from an 
incorrect FSAR statement that the TIP lines did not have to satisfy 10 CFR 
50, Appendix A, General Design Criterion (GDC) 56. It is expected that 
recipients will review the information for applicability to their facilities
and consider actions, if appropriate, to preclude similar problems from 
occurring at their facilities. However, suggestions contained in this 
information notice do not constitute NRC requirements; therefore, no 
specific action or written response is required. 

Description of Circumstances: 

On August 9, 1985 maintenance was performed on a Duane Arnold Energy Center 
(DAEC) TIP isolation ball valve. The valve was not opening properly, thereby
preventing the TIP from being inserted. The licensee reviewed the design and
the FSAR which states in a note to Table 3.1-1: "Since the TIP system lines 
do not communicate freely with the containment atmosphere and since they do 
not comprise a portion of the reactor coolant pressure boundary [10 CFR 50, 
Appendix A, General Design Criteria] GDC 55 and 56 are not directly 
applicable to this specific class of lines." Following this review, the 
licensee decided that the isolation ball valve could be changed out with the
plant on-line because of the closed system configuration. The indexer was 
positioned on a blanked off line to provide additional assurance against 
leakage if a TIP tube failed in the reactor vessel. The isolation ball valve
was removed and replaced (see Figure 1). 

GDC 55 addresses process lines that are part of the reactor coolant pressure
boundary and penetrate containment. GDC 56 addresses lines that connect 
directly to the containment atmosphere. GDC 57 addresses lines of closed 
systems that penetrate containment and are neither part of the reactor 
coolant pressure boundary nor connected directly to the containment 


                                                           IN 86-75       
                                                           August 21, 1986 
                                                           Page 2 of 3    

In January 1986, the replacement ball valve appeared to be sticking and 
would not fully open. While the licensee's Operations Review Committee was 
reviewing the design before performing maintenance, it was concluded that 
the FSAR basis for the previous action was not correct. Inside the primary 
containment, the indexing mechanism (a revolver-like mechanism used for 
selecting a guide tube leading into the reactor vessel) is mounted inside a 
housing equipped with a pressure relief valve which communicates with 
primary containment atmosphere. This pressure relief valve is unique in that 
it will open on a negative or positive pressure inside the housing. The 
indexing mechanism (indexer) has a slip fitting with the guide tubes leading 
to the reactor vessel. If a loss of coolant accident occurred and the 
primary containment became pressurized, the pressure relief valve would open 
to reduce the pressure differential across the indexer housing and the 
containment atmosphere would leak past the slip fitting into the TIP line. 
Under these conditions GDC 56 would apply rather than GDC 57. With the ball 
valve removed for repair or replacement, primary containment integrity would
be compromised. 

The licensee is conducting a review to determine the necessary revisions to 
the Technical Specifications and FSAR. The subsequent required maintenance 
on the system was conducted such that primary containment integrity was not 
compromised and post-maintenance testing included Type B and C leak testing.


NRC has concerns that incorrect interpretation of the design basis for TIP 
systems can lead to hazardous maintenance practices. The possibility exists 
for this type of maintenance error to occur on a generic basis. The FSARs 
for each of several additional BWR reactors with MARK I containments were 
reviewed. In those cases where the FSAR contained a section in Part 3 on 
meeting individual general design criteria of 10 CFR 50, wording is included
that is essentially identical to that in the FSAR of the Duane Arnold Energy
Center, as quoted above. 

Plants which have committed to using ANSI/ANS 3.2-1976 (or 1982), "American 
National Standard Administrative Controls and Quality Assurance for the 
Operational Phase of Nuclear Power Plants," have committed to the 
requirement of therein which states: "Maintenance shall be scheduled 
and planned so as not to compromise the safety of the plant." The finding of 
the DAEC Operations Review Committee is consistent with the NRC licensing 
practice in defining these systems as open to the containment atmosphere, 
thus, opening the system to the atmosphere outside of containment during 
plant operation constitutes a violation of containment integrity and 
compromises the safety of the plant in violation of the above commitment. 

In addition, if a leak from the reactor coolant system occurred or existed 
in any of the TIP tubes of an indexing mechanism, opening the external line 
from that mechanism for maintenance while the plant is in operation would 
result in a non-isolable LOCA. Primary coolant released into the work area 
would result in a hazardous steam environment and radiation exposure to 
maintenance personnel in the area. The consequences of a coolant-to-tube 
leak are potentially 

                                                           IN 86-75       
                                                           August 21, 1986 
                                                           Page 3 of 3    

significant unless the system is depressurized prior to performing repairs. 
An accident situation similar to what might occur is described in 
Information Notice 84-55 Seal Table Leaks at PWRs. 

No specific action or written response is required by this information 
notice. If you have any questions about this matter, please contact the 
Regional Administrator of the appropriate regional office or this office. 

                                   Edward L. Jordan Director 
                                   Division of Emergency Preparedness 
                                     and Engineering Response 
                                   Office of Inspection and Enforcement 

Technical Contact:  William F. Anderson, IE

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