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| {{#Wiki_filter:SSINS No.: 6835IN 86-70UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF INSPECTION AND ENFORCEMENTWASHINGTON, D.C. 20555August 18, 1986IE INFORMATION NOTICE NO. 86-70: POTENTIAL FAILURE OF ALL EMERGENCYDIESEL GENERATORS | | {{#Wiki_filter:SSINS No.: 6835 IN 86-70 UNITED STATES NUCLEAR REGULATORY |
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| | COMMISSION |
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| | OFFICE OF INSPECTION |
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| | ===AND ENFORCEMENT=== |
| | WASHINGTON, D.C. 20555 August 18, 1986 IE INFORMATION |
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| | NOTICE NO. 86-70: POTENTIAL |
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| | ===FAILURE OF ALL EMERGENCY DIESEL GENERATORS=== |
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| ==Addressees== | | ==Addressees== |
| :All nuclear power reactor facilities holding an operating license or aconstruction permit. | | : |
| | All nuclear power reactor facilities |
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| | holding an operating |
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| | license or a construction |
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| | permit. |
|
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|
| ==Purpose== | | ==Purpose== |
| :This notice informs recipients of multiple design deficiencies in a vital acpower system that could potentially result in the loss of all emergency dieselgenerators (EDGs). It is expected that recipients will review the informationfor applicability to their facilities and consider actions, if appropriate, topreclude a similar problem occurring at their facilities. However, suggestionscontained in this information notice do not constitute NRC requirements; there-fore, no specific action or written response is required. | | : This notice informs recipients |
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| | of multiple design deficiencies |
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| | in a vital ac power system that could potentially |
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| | result in the loss of all emergency |
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| | diesel generators (EDGs). It is expected that recipients |
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| | will review the information |
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| | for applicability |
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| | to their facilities |
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| | and consider actions, if appropriate, to preclude a similar problem occurring |
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| | at their facilities. |
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| | ===However, suggestions=== |
| | contained |
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| | in this information |
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| | notice do not constitute |
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| | NRC requirements; |
| | there-fore, no specific action or written response is required.Description |
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| | of Circumstances: |
| | On June 10, 1986, in accordance |
| | |
| | with 10 CFR Part 21, Florida Power & Light (FP&L) reported the results of a failure modes and effects analysis (FMEA).This analysis identified |
| | |
| | multiple design deficiencies |
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| | in the Turkey Point Units 3 and 4 vital ac power system that could potentially |
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| | result in the loss of both EDGs due to overloading. |
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| | In addition, one deficiency |
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| | was identified |
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| | that could result in the loss of all ac power to Unit 4. Turkey Point Units 3 and 4 share two EDGs. Simplified |
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| | diagrams of part of the Turkey Point 3 & 4 electrical |
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| | distribution |
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| | system are provided as Figures 1 and 2.One design deficiency |
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| | involved a "swing bus." As shown on Figure 1, 480 volt MCC "D" is normally supplied from EDG "B" but will shift its power supply to EDG "A" if needed. MCC "D" also feeds a non-safety |
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| | stub bus through a single circuit breaker. This breaker is normally opened by the 48 load sequencer |
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| | to isolate the non-safety |
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| | loads from MCC "D" when the 48 4160V bus is loaded on EDG "B". A single failure in that circuit breaker would leave the non-safety |
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| | loads connected |
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| | to MCC "D". During safety injection |
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| | for one unit, the addi-tional non-safety |
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| | loads could overload EDG "B", possibly resulting |
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| | in a trip of this EDG. MCC "D" would then transfer its power supply and could overload EDG "A" as well.8608130578 IN 86-70 August 18, 1986 In addition to the above design deficiency, five deficiencies |
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| | in the EDG load evaluation |
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| | were found, each of which could contribute |
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| | to an overloading |
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| | of both EDGs. Both EDGs could be overloaded |
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| | in the event of a safety injection |
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| | on one unit and a loss of offsite power to both units. In some cases this could occur without a single failure. These deficiencies |
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| | are as follows: 0 The post-TMI Emergency |
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| | Operating |
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| | Procedures |
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| | require an additional |
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| | charging pump and high head safety injection |
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| | pump in the 1-30 minute and 30-60 minute EDG load interval, respectively. |
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| | o In the event of a single failure of one EDG, actual testing of component cooling water and intake cooling water showed that pump loads (KW) were higher than expected. |
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| | Under this condition, the cooling water pumps provide flow to parallel hydraulic |
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| | trains. Since the flow resistance |
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| | is lower, the pump flow rate is greater and the pump motor horsepower |
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| | would increase.This in turn may increase the load on the EDG to an unacceptable |
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| | value.o Recent environmental |
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| | qualification |
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| | evaluations |
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| | per 10 CFR 50.49 assumed that the emergency |
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| | containment |
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| | coolers operate longer than previously |
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| | assumed.0 Load center transformer |
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| | losses had not been accounted |
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| | for in prior EDG load evaluations. |
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| | o Some nonsafety |
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| | loads that are auto-start |
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| | enabled when the safety injection signal is reset could result in EDG overloading. |
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| | These loads would not always immediately |
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| | start, but would auto-start |
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| | when their start setpoints were reached after the reset.In addition to the overloading |
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| | deficiencies |
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| | above, a design deficiency. |
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| | was found that could result in the loss of one EDG, and at the same time could result in the loss of all ac power to Unit 4 as mentioned |
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| | earlier. In the event of a single failure of the 4A battery or the 4A load sequencer |
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| | subsequent |
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| | to a loss of off-site power, the 4A 4160V bus shown in Figure 2 would be prevented |
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| | from automatically |
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| | loading onto the "A" EDG. It would also prevent the transfer of the 4A MCC swing bus from 4A load center to 4B load center. The "B" EDG has a self-contained |
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| | supply of fuel oil that lasts about one hour. After that time, fuel must be gravity supplied by the day tank via a solenoid operated valve.However, this valve is powered from MCC 4A. Therefore, the above mentioned single failure would also cause the loss of the "B" EDG and the loss of all ac power to Unit 4. The Unit 3 3B 4160V bus would also be deenergized, along with three of the four high head safety injection |
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| | pumps and the control room air-conditioning. |
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| | Discussion: |
| | The single failure vulnerability |
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| | of the non-safety |
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| | stub bus discussed |
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| | above may have been evaluated |
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| | during the licensing |
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| | review and the loads on the EDGs found to be low enough so that failure of the stub bus breaker would not overload the diesel. Subsequent |
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| | to initial operation, however, additional |
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| | loads have been |
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| | IN 86-70 August 18, 1986 added on the diesels. Prior to November, 1985, there was the potential |
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| | for loading the EDGs to 3544 kw in the 1-30 minute period and 3451 in the 30-60 minute period following |
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| | a LOCA in one unit with a loss of offsite power in both units. These load levels exceed the highest FSAR value (1/2 hour exceptional |
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| | rating) of 3050 Kw by a substantial |
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| | margin. With the EDGs overloaded, the potential |
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| | exists that safety-related |
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| | loads would not start or would not operate at design ratings. Under these conditions, the EDG engines could stall and damage to the generator |
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| | and/or engine could occur. There are no test data available |
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| | to demonstrate |
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| | that the EDGs could accept the above load levels.As an interim solution to the above design deficiencies, FP&L disabled the automatic |
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| | transfer between EDGs and established |
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| | a dedicated |
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| | operator to take corrective |
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| | action (i.e., manipulate |
| | |
| | appropriate |
| | |
| | breakers) |
| | as necessary. |
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| | The licensee has now modified EDG loads, loading sequences |
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| | and plant procedures. |
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| | This example illustrates |
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| | the importance |
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| | of reviewing |
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| | EDG load limits under all possible operating |
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| | configurations |
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| | for a range of design basis accident condi-tions. FP&L, through an FMEA, system testing, and evaluation |
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| | of EDG load limits, found design deficiencies |
| | |
| | that may have prevented |
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| | one or both EDG's from performing |
| | |
| | their required function.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 |
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| | of the appropriate |
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| | regional office or this office.Jordan, Director Divisi of Emergency |
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| | ===Preparedness=== |
| | and Engineering |
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| | Response Office of Inspection |
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| | and Enforcement |
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| | Technical |
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| | Contacts: |
| | Henry Bailey, IE (301) 492-9006 Joseph Giitter, IE (301) 492-9001 Attachments: |
| | 1. Figures 1 and 2, Turkey Point -Units 3 & 4*2. List of Recently Issued IE Information |
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| | Notices |
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| | Attachment |
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| | 1 IN 86-70 August 18, 1986 FIGURE 1 TURKEY POINT -UNITS 3 E 4 EDG "B" I)LOAD CENTERS 480V 4C 3D 480V* h N.C.N.C.480V MCC D NON-SAFETY |
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| | SAFETY UNIT 3 UNIT 4'Simplified |
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| | Diagram of Portion of Electrical |
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| | Distribution |
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| | System. |
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| | Attachment |
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| | 1 IN 86-70 August 18, 1986 FIGURE 2 TURKEY POINT -UNITS If 3 & 4*EDG "'A"/-\ 4160V 3A 3B 4160V EDG 'B" 4160V 4A 4B 4160V I)I 3S Si Cl 38 Si 4A Si 4B Si LOAD CENTERS 480V 4A 4D 480V N.O.N.O.480V NON-SAFETY |
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| | SAFETY"B1 EDG AUXILIARIES |
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| | 'Simplified |
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| | Diagram of Portion of Electrical |
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| | Distribution |
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| | System. |
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| | Attachment |
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| | 2 IN 86-70 August 18, 1986 LIST OF RECENTLY ISSUED IE INFORMATION |
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| | NOTICES Information |
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| | Date of Notice No. Subject Issue Issued to 86-69 86-68 86-67 86-66 86-65 Scram Solenoid Pilot Valve 8/18/86 (SSPV) Rebuild Kit Problems Stuck Control Rod 8/15/86 Portable Moisture/Density |
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| | 8/15/86 Gauges: Recent Incidents -And Common Violations |
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| | Of Require-ments For Use, Transportation, And Storage Potential |
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| | For Failure Of 8/15/86 Replacement |
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| | AC Coils Supplied By The Westinghouse |
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| | Electric Corporation |
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| | ===For Use In Class lE Motor Starters And Contractors=== |
| | Malfunctions |
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| | Of ITT Barton 8/14/86 Model 580 Series Switches During Requalification |
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| | Testing Deficiencies |
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| | In Upgrade 8/14/86 Programs For Plant Emergency Operating |
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| | Procedures |
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| | Loss Of Safety Injection |
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| | 8/6/86 Capability |
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| | Potential |
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| | Problems In West- 7/31/86 inghouse Molded Case Circuit Breakers Equipped With A Shunt Trip Failure Of Auxiliary |
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| | Feed- 7/28/86 water Manual Isolated Valve All BWR facilities |
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| | holding an OL or CP All BWR holding All NRC licensees authorized |
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| | to possess, use, transport, and store sealed sources All power reactor facilities |
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| | holding an OL or CP All power reactor facilities |
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| | holding an OL or CP All power reactor facilities |
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| | holding an OL or CP All PWR facilities |
|
| |
|
| ==Description of Circumstances==
| | holding an OL or CP All power reactor facilities |
| :On June 10, 1986, in accordance with 10 CFR Part 21, Florida Power & Light(FP&L) reported the results of a failure modes and effects analysis (FMEA).This analysis identified multiple design deficiencies in the Turkey PointUnits 3 and 4 vital ac power system that could potentially result in the lossof both EDGs due to overloading. In addition, one deficiency was identifiedthat could result in the loss of all ac power to Unit 4. Turkey Point Units 3and 4 share two EDGs. Simplified diagrams of part of the Turkey Point 3 & 4electrical distribution system are provided as Figures 1 and 2.One design deficiency involved a "swing bus." As shown on Figure 1, 480 voltMCC "D" is normally supplied from EDG "B" but will shift its power supply toEDG "A" if needed. MCC "D" also feeds a non-safety stub bus through a singlecircuit breaker. This breaker is normally opened by the 48 load sequencer toisolate the non-safety loads from MCC "D" when the 48 4160V bus is loaded onEDG "B". A single failure in that circuit breaker would leave the non-safetyloads connected to MCC "D". During safety injection for one unit, the addi-tional non-safety loads could overload EDG "B", possibly resulting in a tripof this EDG. MCC "D" would then transfer its power supply and could overloadEDG "A" as well.8608130578 IN 86-70August 18, 1986 In addition to the above design deficiency, five deficiencies in the EDG loadevaluation were found, each of which could contribute to an overloading of bothEDGs. Both EDGs could be overloaded in the event of a safety injection on oneunit and a loss of offsite power to both units. In some cases this could occurwithout a single failure. These deficiencies are as follows:0 The post-TMI Emergency Operating Procedures require an additional chargingpump and high head safety injection pump in the 1-30 minute and 30-60minute EDG load interval, respectively.o In the event of a single failure of one EDG, actual testing of componentcooling water and intake cooling water showed that pump loads (KW) werehigher than expected. Under this condition, the cooling water pumps provideflow to parallel hydraulic trains. Since the flow resistance is lower,the pump flow rate is greater and the pump motor horsepower would increase.This in turn may increase the load on the EDG to an unacceptable value.o Recent environmental qualification evaluations per 10 CFR 50.49 assumedthat the emergency containment coolers operate longer than previouslyassumed.0 Load center transformer losses had not been accounted for in prior EDGload evaluations.o Some nonsafety loads that are auto-start enabled when the safety injectionsignal is reset could result in EDG overloading. These loads would notalways immediately start, but would auto-start when their start setpointswere reached after the reset.In addition to the overloading deficiencies above, a design deficiency. was foundthat could result in the loss of one EDG, and at the same time could result inthe loss of all ac power to Unit 4 as mentioned earlier. In the event of asingle failure of the 4A battery or the 4A load sequencer subsequent to a lossof off-site power, the 4A 4160V bus shown in Figure 2 would be prevented fromautomatically loading onto the "A" EDG. It would also prevent the transfer ofthe 4A MCC swing bus from 4A load center to 4B load center. The "B" EDG has aself-contained supply of fuel oil that lasts about one hour. After that time,fuel must be gravity supplied by the day tank via a solenoid operated valve.However, this valve is powered from MCC 4A. Therefore, the above mentionedsingle failure would also cause the loss of the "B" EDG and the loss of all acpower to Unit 4. The Unit 3 3B 4160V bus would also be deenergized, along withthree of the four high head safety injection pumps and the control roomair-conditioning.Discussion:The single failure vulnerability of the non-safety stub bus discussed above mayhave been evaluated during the licensing review and the loads on the EDGs foundto be low enough so that failure of the stub bus breaker would not overload thediesel. Subsequent to initial operation, however, additional loads have been
| |
|
| |
|
| IN 86-70August 18, 1986 added on the diesels. Prior to November, 1985, there was the potential forloading the EDGs to 3544 kw in the 1-30 minute period and 3451 in the 30-60minute period following a LOCA in one unit with a loss of offsite power in bothunits. These load levels exceed the highest FSAR value (1/2 hour exceptionalrating) of 3050 Kw by a substantial margin. With the EDGs overloaded, thepotential exists that safety-related loads would not start or would not operateat design ratings. Under these conditions, the EDG engines could stall anddamage to the generator and/or engine could occur. There are no test dataavailable to demonstrate that the EDGs could accept the above load levels.As an interim solution to the above design deficiencies, FP&L disabled theautomatic transfer between EDGs and established a dedicated operator to takecorrective action (i.e., manipulate appropriate breakers) as necessary. Thelicensee has now modified EDG loads, loading sequences and plant procedures.This example illustrates the importance of reviewing EDG load limits under allpossible operating configurations for a range of design basis accident condi-tions. FP&L, through an FMEA, system testing, and evaluation of EDG load limits,found design deficiencies that may have prevented one or both EDG's fromperforming their required function.No specific action or written response is required by this information notice.If you have any questions about this matter, please contact the RegionalAdministrator of the appropriate regional office or this office.Jordan, DirectorDivisi of Emergency Preparednessand Engineering ResponseOffice of Inspection and EnforcementTechnical Contacts: Henry Bailey, IE(301) 492-9006Joseph Giitter, IE(301) 492-9001Attachments:1. Figures 1 and 2, Turkey Point -Units 3 & 4*2. List of Recently Issued IE Information Notices
| | holding an OL or CP All power reactor facilities |
|
| |
|
| Attachment 1IN 86-70August 18, 1986 FIGURE 1TURKEY POINT -UNITS 3 E 4EDG "B"I)LOAD CENTERS480V 4C3D 480V* hN.C.N.C.480VMCC DNON-SAFETYSAFETYUNIT 3UNIT 4'Simplified Diagram of Portion of Electrical Distribution System.
| | holding a CP facilities |
|
| |
|
| Attachment 1IN 86-70August 18, 1986 FIGURE 2TURKEY POINT -UNITSIf3 & 4*EDG "'A"/-\ 4160V 3A3B 4160VEDG 'B"4160V 4A4B 4160VI)I3SSiCl38Si4ASi4BSiLOAD CENTERS480V 4A4D 480VN.O.N.O.480VNON-SAFETYSAFETY"B1 EDGAUXILIARIES'Simplified Diagram of Portion of Electrical Distribution System.
| | an OL or CP 86-64 86-63 86-62 86-61 OL = OperatinG |
|
| |
|
| Attachment 2IN 86-70August 18, 1986LIST OF RECENTLY ISSUEDIE INFORMATION NOTICESInformation Date ofNotice No. Subject Issue Issued to86-6986-6886-6786-6686-65Scram Solenoid Pilot Valve 8/18/86(SSPV) Rebuild Kit ProblemsStuck Control Rod 8/15/86Portable Moisture/Density 8/15/86Gauges: Recent Incidents -AndCommon Violations Of Require-ments For Use, Transportation,And StoragePotential For Failure Of 8/15/86Replacement AC Coils SuppliedBy The Westinghouse ElectricCorporation For Use In ClasslE Motor Starters AndContractorsMalfunctions Of ITT Barton 8/14/86Model 580 Series SwitchesDuring Requalification TestingDeficiencies In Upgrade 8/14/86Programs For Plant EmergencyOperating ProceduresLoss Of Safety Injection 8/6/86CapabilityPotential Problems In West- 7/31/86inghouse Molded Case CircuitBreakers Equipped With AShunt TripFailure Of Auxiliary Feed- 7/28/86water Manual Isolated ValveAll BWR facilitiesholding an OL or CPAll BWRholdingAll NRC licenseesauthorized to possess,use, transport, andstore sealed sourcesAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll PWR facilitiesholding an OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdinga CPfacilitiesan OL or CP86-6486-6386-6286-61OL = OperatinG LicenseCP = Construction Permit
| | License CP = Construction |
|
| |
|
| }} | | Permit}} |
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| {{Information notice-Nav}} | | {{Information notice-Nav}} |
Potential Failure of All Emergency Diesel GeneratorsML031250102 |
Person / Time |
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Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill |
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Issue date: |
08/18/1986 |
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From: |
Jordan E L NRC/IE |
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To: |
|
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References |
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IN-86-070, NUDOCS 8608130578 |
Download: ML031250102 (6) |
|
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Category:NRC Information Notice
MONTHYEARInformation Notice 2020-02, Flex Diesel Generator Operational Challenges2020-09-15015 September 2020 Flex Diesel Generator Operational Challenges ML20225A0322020-09-0303 September 2020 NRC Choice Letter to NAC International with Attached Safety Inspection Report, IR 0721015/2020201, February 24-27, 2020 and July 22, 2020, Inspection of NAC International in Norcross, Georgia Information Notice 2012-09, PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs.2019-11-30030 November 2019 PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs. 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Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Contain2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Con2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notic2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2012-13, Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool Information Notice 2012-13, Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool Information Notice 2012-11, Age Related Capacitor Degradation2012-07-23023 July 2012 Age Related Capacitor Degradation ML12031A0132012-02-0606 February 2012 U.S. Nuclear Regulatory Commission Investigation Report No. 2-2010-058, Cpn International, Inc Information Notice 2011-19, Licensee Event Reports Containing Information Pertaining to Defects to Basic Components2011-09-26026 September 2011 Licensee Event Reports Containing Information Pertaining to Defects to Basic Components Information Notice 2011-15, Steel Containment Degradation and Associated License Renewal Aging Management Issues2011-08-0101 August 2011 Steel Containment Degradation and Associated License Renewal Aging Management Issues Information Notice 2011-17, Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping2011-07-26026 July 2011 Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping Information Notice 2011-13, Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-13, Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13)2011-06-29029 June 2011 Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13) Information Notice 2011-13, OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-04, IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-262010-12-21021 December 2010 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 ML13066A1872009-12-16016 December 2009 Draft NRC Information Notice 2009-xx - Underestimate of Dam Failure Frequency Used in Probabilistic Risk Assessments ML1007804482009-11-23023 November 2009 Email from Peter Bamford, NRR to Pamela Cowan, Exelon on TMI Contamination Control Event Information Notice 2009-11, NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-112009-07-0707 July 2009 NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-11 Information Notice 2009-10, Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10)2009-07-0707 July 2009 Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10) Information Notice 2009-09, Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify2009-06-19019 June 2009 Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify Information Notice 2008-12, Reactor Trip Due to Off-Site Power Fluctuation2008-07-0707 July 2008 Reactor Trip Due to Off-Site Power Fluctuation Information Notice 2008-11, Service Water System Degradation at Brunswicksteam Electric Plant Unit 12008-06-18018 June 2008 Service Water System Degradation at Brunswicksteam Electric Plant Unit 1 Information Notice 2008-04, Counterfeit Parts Supplied to Nuclear Power Plants2008-04-0707 April 2008 Counterfeit Parts Supplied to Nuclear Power Plants Information Notice 1991-09, Counterfeiting of Crane Valves2007-09-25025 September 2007 Counterfeiting of Crane Valves Information Notice 2007-28, Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls2007-09-19019 September 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls Information Notice 2007-29, Temporary Scaffolding Affects Operability of Safety-Related Equipment2007-09-17017 September 2007 Temporary Scaffolding Affects Operability of Safety-Related Equipment Information Notice 2007-14, Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station2007-03-30030 March 2007 Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station Information Notice 2007-06, Potential Common Cause Vulnerabilities in Essential Service Water Systems2007-02-0909 February 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Information Notice 2007-05, Vertical Deep Draft Pump Shaft and Coupling Failures2007-02-0909 February 2007 Vertical Deep Draft Pump Shaft and Coupling Failures Information Notice 2006-31, Inadequate Fault Interrupting Rating of Breakers2006-12-26026 December 2006 Inadequate Fault Interrupting Rating of Breakers Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear Information Notice 2006-13, E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination2006-07-13013 July 2006 E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Zimmer]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Skagit]] OR [[:Marble Hill]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Zimmer]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Skagit]] OR [[:Marble Hill]] </code>. |
SSINS No.: 6835 IN 86-70 UNITED STATES NUCLEAR REGULATORY
COMMISSION
OFFICE OF INSPECTION
AND ENFORCEMENT
WASHINGTON, D.C. 20555 August 18, 1986 IE INFORMATION
NOTICE NO. 86-70: POTENTIAL
Addressees
All nuclear power reactor facilities
holding an operating
license or a construction
permit.
Purpose
- This notice informs recipients
of multiple design deficiencies
in a vital ac power system that could potentially
result in the loss of all emergency
diesel generators (EDGs). It is expected that recipients
will review the information
for applicability
to their facilities
and consider actions, if appropriate, to preclude a similar problem occurring
at their facilities.
However, suggestions
contained
in this information
notice do not constitute
NRC requirements;
there-fore, no specific action or written response is required.Description
of Circumstances:
On June 10, 1986, in accordance
with 10 CFR Part 21, Florida Power & Light (FP&L) reported the results of a failure modes and effects analysis (FMEA).This analysis identified
multiple design deficiencies
in the Turkey Point Units 3 and 4 vital ac power system that could potentially
result in the loss of both EDGs due to overloading.
In addition, one deficiency
was identified
that could result in the loss of all ac power to Unit 4. Turkey Point Units 3 and 4 share two EDGs. Simplified
diagrams of part of the Turkey Point 3 & 4 electrical
distribution
system are provided as Figures 1 and 2.One design deficiency
involved a "swing bus." As shown on Figure 1, 480 volt MCC "D" is normally supplied from EDG "B" but will shift its power supply to EDG "A" if needed. MCC "D" also feeds a non-safety
stub bus through a single circuit breaker. This breaker is normally opened by the 48 load sequencer
to isolate the non-safety
loads from MCC "D" when the 48 4160V bus is loaded on EDG "B". A single failure in that circuit breaker would leave the non-safety
loads connected
to MCC "D". During safety injection
for one unit, the addi-tional non-safety
loads could overload EDG "B", possibly resulting
in a trip of this EDG. MCC "D" would then transfer its power supply and could overload EDG "A" as well.8608130578 IN 86-70 August 18, 1986 In addition to the above design deficiency, five deficiencies
in the EDG load evaluation
were found, each of which could contribute
to an overloading
of both EDGs. Both EDGs could be overloaded
in the event of a safety injection
on one unit and a loss of offsite power to both units. In some cases this could occur without a single failure. These deficiencies
are as follows: 0 The post-TMI Emergency
Operating
Procedures
require an additional
charging pump and high head safety injection
pump in the 1-30 minute and 30-60 minute EDG load interval, respectively.
o In the event of a single failure of one EDG, actual testing of component cooling water and intake cooling water showed that pump loads (KW) were higher than expected.
Under this condition, the cooling water pumps provide flow to parallel hydraulic
trains. Since the flow resistance
is lower, the pump flow rate is greater and the pump motor horsepower
would increase.This in turn may increase the load on the EDG to an unacceptable
value.o Recent environmental
qualification
evaluations
per 10 CFR 50.49 assumed that the emergency
containment
coolers operate longer than previously
assumed.0 Load center transformer
losses had not been accounted
for in prior EDG load evaluations.
o Some nonsafety
loads that are auto-start
enabled when the safety injection signal is reset could result in EDG overloading.
These loads would not always immediately
start, but would auto-start
when their start setpoints were reached after the reset.In addition to the overloading
deficiencies
above, a design deficiency.
was found that could result in the loss of one EDG, and at the same time could result in the loss of all ac power to Unit 4 as mentioned
earlier. In the event of a single failure of the 4A battery or the 4A load sequencer
subsequent
to a loss of off-site power, the 4A 4160V bus shown in Figure 2 would be prevented
from automatically
loading onto the "A" EDG. It would also prevent the transfer of the 4A MCC swing bus from 4A load center to 4B load center. The "B" EDG has a self-contained
supply of fuel oil that lasts about one hour. After that time, fuel must be gravity supplied by the day tank via a solenoid operated valve.However, this valve is powered from MCC 4A. Therefore, the above mentioned single failure would also cause the loss of the "B" EDG and the loss of all ac power to Unit 4. The Unit 3 3B 4160V bus would also be deenergized, along with three of the four high head safety injection
pumps and the control room air-conditioning.
Discussion:
The single failure vulnerability
of the non-safety
stub bus discussed
above may have been evaluated
during the licensing
review and the loads on the EDGs found to be low enough so that failure of the stub bus breaker would not overload the diesel. Subsequent
to initial operation, however, additional
loads have been
IN 86-70 August 18, 1986 added on the diesels. Prior to November, 1985, there was the potential
for loading the EDGs to 3544 kw in the 1-30 minute period and 3451 in the 30-60 minute period following
a LOCA in one unit with a loss of offsite power in both units. These load levels exceed the highest FSAR value (1/2 hour exceptional
rating) of 3050 Kw by a substantial
margin. With the EDGs overloaded, the potential
exists that safety-related
loads would not start or would not operate at design ratings. Under these conditions, the EDG engines could stall and damage to the generator
and/or engine could occur. There are no test data available
to demonstrate
that the EDGs could accept the above load levels.As an interim solution to the above design deficiencies, FP&L disabled the automatic
transfer between EDGs and established
a dedicated
operator to take corrective
action (i.e., manipulate
appropriate
breakers)
as necessary.
The licensee has now modified EDG loads, loading sequences
and plant procedures.
This example illustrates
the importance
of reviewing
EDG load limits under all possible operating
configurations
for a range of design basis accident condi-tions. FP&L, through an FMEA, system testing, and evaluation
of EDG load limits, found design deficiencies
that may have prevented
one or both EDG's from performing
their required function.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.Jordan, Director Divisi of Emergency
Preparedness
and Engineering
Response Office of Inspection
and Enforcement
Technical
Contacts:
Henry Bailey, IE (301) 492-9006 Joseph Giitter, IE (301) 492-9001 Attachments:
1. Figures 1 and 2, Turkey Point -Units 3 & 4*2. List of Recently Issued IE Information
Notices
Attachment
1 IN 86-70 August 18, 1986 FIGURE 1 TURKEY POINT -UNITS 3 E 4 EDG "B" I)LOAD CENTERS 480V 4C 3D 480V* h N.C.N.C.480V MCC D NON-SAFETY
SAFETY UNIT 3 UNIT 4'Simplified
Diagram of Portion of Electrical
Distribution
System.
Attachment
1 IN 86-70 August 18, 1986 FIGURE 2 TURKEY POINT -UNITS If 3 & 4*EDG "'A"/-\ 4160V 3A 3B 4160V EDG 'B" 4160V 4A 4B 4160V I)I 3S Si Cl 38 Si 4A Si 4B Si LOAD CENTERS 480V 4A 4D 480V N.O.N.O.480V NON-SAFETY
SAFETY"B1 EDG AUXILIARIES
'Simplified
Diagram of Portion of Electrical
Distribution
System.
Attachment
2 IN 86-70 August 18, 1986 LIST OF RECENTLY ISSUED IE INFORMATION
NOTICES Information
Date of Notice No. Subject Issue Issued to 86-69 86-68 86-67 86-66 86-65 Scram Solenoid Pilot Valve 8/18/86 (SSPV) Rebuild Kit Problems Stuck Control Rod 8/15/86 Portable Moisture/Density
8/15/86 Gauges: Recent Incidents -And Common Violations
Of Require-ments For Use, Transportation, And Storage Potential
For Failure Of 8/15/86 Replacement
AC Coils Supplied By The Westinghouse
Electric Corporation
For Use In Class lE Motor Starters And Contractors
Malfunctions
Of ITT Barton 8/14/86 Model 580 Series Switches During Requalification
Testing Deficiencies
In Upgrade 8/14/86 Programs For Plant Emergency Operating
Procedures
Loss Of Safety Injection
8/6/86 Capability
Potential
Problems In West- 7/31/86 inghouse Molded Case Circuit Breakers Equipped With A Shunt Trip Failure Of Auxiliary
Feed- 7/28/86 water Manual Isolated Valve All BWR facilities
holding an OL or CP All BWR holding All NRC licensees authorized
to possess, use, transport, and store sealed sources All power reactor facilities
holding an OL or CP All power reactor facilities
holding an OL or CP All power reactor facilities
holding an OL or CP All PWR facilities
holding an OL or CP All power reactor facilities
holding an OL or CP All power reactor facilities
holding a CP facilities
an OL or CP 86-64 86-63 86-62 86-61 OL = OperatinG
License CP = Construction
Permit
|
---|
|
list | - Information Notice 1986-01, Failure of Main Feedwater Check Valves Causes Loss of Feedwater System Integrity and Water-Hammer Damage (6 January 1986)
- Information Notice 1986-02, Failure of Valve Operator Motor During Environmental Qualification Testing (6 January 1986)
- Information Notice 1986-03, Potential Deficiencies in Enviromental Qualification of Limitorque Motor Valve Operator Wiring (14 January 1986)
- Information Notice 1986-04, Transient Due to Loss of Power to Intergrated Control System at a Pressurized Water Reactor Designed by Babcock & Wilcox (31 January 1986)
- Information Notice 1986-05, Main Steam Safety Valve Test Failures and Ring Setting Adjustments (31 January 1986)
- Information Notice 1986-06, Failure of Lifting Rig Attachment, While Lifting Upper Guide Structure at St. Lucie Unit 1 (3 February 1986, Topic: Control of Heavy Loads)
- Information Notice 1986-06, Failure of Lifting Rig Attachment, while Lifting Upper Guide Structure at St. Lucie Unit 1 (3 February 1986)
- Information Notice 1986-07, Lack of Detailed Instruction & Inadequate Observance of Precautions During Maintenance & Testing of Diesel Generator Woodward Governors (3 February 1986)
- Information Notice 1986-08, Licensee Event Report (LER) Format Modification (3 February 1986)
- Information Notice 1986-09, Failure of Check & Stop Check Valves Subjected to Low Flow Conditions (3 February 1986)
- Information Notice 1986-10, Degradation of Reactor Coolant System Pressure Boundary Resulting From Boric Acid Corrosion. (5 January 1995, Topic: Boric Acid, Stress corrosion cracking)
- Information Notice 1986-10, Feedwater Line Break (10 November 1988, Topic: Coatings, Anchor Darling)
- Information Notice 1986-10, Degradation of Reactor Coolant System Pressure Boundary Resulting from Boric Acid Corrosion. (5 January 1995, Topic: Boric Acid, Stress corrosion cracking)
- Information Notice 1986-11, Anomalous Behavior of Recirculation Loop Flow in Jet Pump BWR Plants (31 December 1986)
- Information Notice 1986-13, Standby Liquid Control System Squib Valves Failure to Fire (21 February 1986, Topic: Squib)
- Information Notice 1986-13, Standby Liquid Control Squib Valves Failure to Fire (5 August 1986, Topic: Squib)
- Information Notice 1986-14, Overspeed Trips of AFW, HPCI & RCIC Turbines (26 August 1991, Topic: Fire Barrier)
- Information Notice 1986-14, Overspeed Trips of Afw, HPCI & RCIC Turbines (26 August 1991)
- Information Notice 1986-15, Loss of Offsite Power Caused by Problems in Fiber Optics Systems (10 March 1986, Topic: Squib)
- Information Notice 1986-16, Failures to Identify Containment Leakage Due to Inadequate Local Testing of BWR Vacuum Relief System Valves (11 March 1986, Topic: Squib)
- Information Notice 1986-17, Update of Failure of Automatic Sprinkler System Valves to Operate (24 March 1986, Topic: Squib)
- Information Notice 1986-18, NRC On-Scene Response During a Major Emergency (26 March 1986, Topic: Squib, Backfit)
- Information Notice 1986-19, Reactor Coolant Pump Shaft Failure at Crystal River (21 March 1986, Topic: Squib)
- Information Notice 1986-20, Low-Level Radioactive Waste Scaling Factors, 10 CFR Part 61 (28 March 1986, Topic: Squib)
- Information Notice 1986-21, Recognition of American Society of Mechanical Engineers Accreditation Program for N Stamp Holders (31 March 1986, Topic: Squib)
- Information Notice 1986-22, Underresponse of Radiation Survey Instrument to High Radiation Fields (31 March 1986, Topic: High Radiation Area, Squib)
- Information Notice 1986-23, Excessive Skin Exposures Due to Contamination with Hot Particles (9 April 1986)
- Information Notice 1986-23, Excessive Skin Exposures due to Contamination with Hot Particles (9 April 1986)
- Information Notice 1986-24, Respirator Users Notice: Increased Inspection Frequency for Certain Self-Contained Breathing Apparatus Air Cylinders (11 April 1986, Topic: Hydrostatic)
- Information Notice 1986-25, Traceability and Material Control of Material and Equipment, Particularly Fasteners (11 April 1986)
- Information Notice 1986-26, Potential Problems in Generators Manufactured by Electrical Products Incorporated (17 April 1986)
- Information Notice 1986-27, Access Control at Nuclear Facilities (21 April 1986, Topic: Contraband)
- Information Notice 1986-28, Telephone Numbers to the NRC Operations Center and Regional Offices (24 April 1986)
- Information Notice 1986-29, Effects of Changing Valve Moter-Operator Switch Settings (25 April 1986)
- Information Notice 1986-30, Design Limitations of Gaseous Effluent Monitoring Systems (29 April 1986)
- Information Notice 1986-31, Unauthorized Transfer and Loss of Control of Industrial Nuclear Gauges (14 July 1986)
- Information Notice 1986-32, Request for Collection of Licensee Radioactivity Measurements Attributed to Chernobyl Nuclear Plant Accident (2 May 1986, Topic: Chernobyl)
- Information Notice 1986-33, Information for Licensee Regarding the Chernobyl Nuclear Plant Accident (6 May 1986, Topic: Chernobyl)
- Information Notice 1986-34, Improper Assembly, Material Selection, & Test of Valves & Their Actuators (13 May 1986)
- Information Notice 1986-34, Improper Assembly, Material Selection, & Test of Valves & their Actuators (13 May 1986)
- Information Notice 1986-35, Fire in Compressible Material at Dresden Unit 3 (15 May 1986, Topic: Chernobyl)
- Information Notice 1986-36, Change in NRC Practice Regarding Issuance of Confirming Letters to Principal Contractors (16 May 1986, Topic: Chernobyl)
- Information Notice 1986-37, Degradation of Station Batteries (16 May 1986, Topic: Chernobyl)
- Information Notice 1986-38, Deficient Operator Actions Following Dual Function Valve Failures (20 May 1986, Topic: Chernobyl)
- Information Notice 1986-39, Failures of RHR Pump Motors and Pump Internals (20 May 1986, Topic: Chernobyl)
- Information Notice 1986-40, Degraded Ability to Isolate the Reactor Coolant System from Low-Pressure Coolant Systems in Bwrs (5 June 1986, Topic: Chernobyl)
- Information Notice 1986-40, Degraded Ability to Isolate the Reactor Coolant System from Low-Pressure Coolant Systems in BWRs (5 June 1986, Topic: Chernobyl)
- Information Notice 1986-41, Evaluation of Questionable Exposure Readings of Licensee Personnel Dosimeters (9 June 1986, Topic: Chernobyl)
- Information Notice 1986-42, Improper Maintenance of Radiation Monitoring Systems (9 June 1986, Topic: Temporary Modification, Chernobyl)
- Information Notice 1986-43, Problems with Silver Zeolite Sampling of Airborne Radioiodine (10 June 1986, Topic: Chernobyl)
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