ML20209E702
| ML20209E702 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/27/1986 |
| From: | Bernthal F NRC COMMISSION (OCM) |
| To: | Dingell J HOUSE OF REP., ENERGY & COMMERCE |
| Shared Package | |
| ML20209E709 | List: |
| References | |
| 860827, NUDOCS 8609110325 | |
| Download: ML20209E702 (16) | |
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, F f (o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20565 a j
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CHAIRMAN August 27, 1986 The Honorable John D. Dingell, Chairman Comittee on Energy and Comerce United States House of Representatives Washington, D. C. 20515
Dear Mr. Chairman:
Your letter of July 10, 1986 requested information concerning the Comission's decision to allow the restart of Fermi-2.
One of the areas for which you requested information pertained to the Mark I '
containment. The Comission is aware that the Monroe County Comissioners asked you to look into the safety of Boiling Water Reactors with Mark I con-tainments, recognizing that Fermi-2 is a plant of that type and there have been publicly expressed concerns regarding that specific type of containment.
Therefore, the Enclosure 3 to this letter. is also being forwarded directly to the Monroe County Comissioners with an offer for the staff to meet with them on this subject if they wish.
The additional information you requested is also provided in the attached enclosures. If the Comission can be of further assistance, please do not hesitate to let me know.
Sincerely, b $ _
Frederick M. Bernthal i Acting Chairman
Enclosures:
- 1. Response to request on restart decision
- 2. Response to request on investigations
- 3. Response to request on Mark I containment cc: Rep. Norman F. Lent 8609110325 860827 l PDR COMMS NRCC CORRESPONDENCE PDR i
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ENCLOSURE 1
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Enclosure 1 Request: Now that the Commission has acted on this matter, I request that you provide me with a full report on the Commission's decisions regarding restart of Fermi-2 and the assessment of civil and
. criminal penalties with regard to the earlier problems at the facility. The report should indicate the nature of the problems identified by the Consnission, and how these have been addressed by the licensee in a manner sufficient to assure the Commission that in the future the plant will be operated safely and in compliance with all NRC regulations.
Response: Shortly after the initial operation of the Fermi-2 facility, numerous management, operational, and regulatory problems surfaced.
The nature of these problems were:
Early inadequate management involvement in and oversight of reactor operating activities. -
Failure by management to successfully accomplish the transition from a construction to an operating organization.
Failure by management to ensure that actions taken to com-pensate for a lack of operating experience were effectively implemented.
Ineffective engineering support.
Ineffective implementation of the security program.
Numerous hardware problems.
The NRC, through a series of inspections, management meetings and enforcement actions has identified problems and required appropriate changes to improve the utility's operations. Enforcement actions have included Notices of Violations; the issuance of a letter under the provisions of 10 CFR 50.54(f) to determine if the license should be suspended, modified, or revoked; the issuance of Civil Penalties; and the issuance of an Order modifying the license.
The Fermi-2 reactor has remained shutdown since October 1985. Since that time, significant management changes have occurred, new opera-tional procedures have been developed, management controls have been enhanced, training has been improved, engineering issues have received extensive review, a security improvement plan is in place and equipment problems have been resolved.
The Commission believes these changes are sufficient to provide reasonable assurance that the plant can be safely restarted.
Augmented inspection efforts are planned during reactor restart to assure that the changes which have been initiated are fully implemented.
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Detroit Edison Company's Fermi-2 received its low power operating license (not to exceed 5% power) on March 20, 1985. The Comisison met and approved Femi's full power license on July 10, 1985, and the license was subsequently issued on July 15, 1985. On July 15, 1985, just after issuance of the license, the NRC staff learned that operator " rod pull" errors which occurred in the control room on July 1-2, 1986, were subsequently determined to have resulted in the plant achieving nuclear criticality, a fact unknown to the Comission when it approved the full power license. Consequently, on July 16, 1985, Region III issued a Confimatory Action Letter (CAL) to confirm utility corrective actions including a provision that any operation of the facility above 5% power required Regional Administrator concurrence (Attachment 1 to Enclosure 1). The utility has not yet requested such concurrence because of other problems which subsequently developed.
During the time frame from mid-July to early October 1985, the company also experienced a number of other operational errors.
Management meetings were held to discuss corrective actions initiated by the utility to improve regulatory perfomance. Based '
on improvement initiated by the licensee, low power operation was allowed to continue. Subsequently, on October 10, 1985, the licensee fomally submitted a Reactor Operations Improvement Program -
(ROIP) which detailed steps taken and planned to improve overall control and conduct of operations (Attachment 2 to Enclosure 1).
The licensee initiated numerous changes to assure better control of activities in the control room. Physical work stations were relocated, individuals' roles were clarified, individuals were trained in their responsibilities, and simulator training was upgraded to better represent actual plant procedures.
The utility shut the reactor down in October 1985 prior to exceeding the 5% power level, primarily because of cracking identified in a non-safety-related bypass line. It also had to complete modifications for environmental qualification of certain equipment prior to operations after November 30, 1985. After the shutdown, additional hardware concerns began to surface on other plant equipment such as the emergency diesel generators. These problems were resolved by the licensee as they occurred and corrective actions were approved by the NRC.
In addition to the equipment problems, concerns developed with control of maintenance activities and the increase in security violations, and questions were raised on the adequacy of engineering analyses for modifications. Based on the staff's analysis of the problems, a letter was issued on December 24, 1985 under the provisions of 10 CFR 50.54(f), which addressed the above problems and required the licensee to respond (Attachment 3 to Enclosure 1). The i letter requested the licensee to addrcis three specific areas: 1) the adequacy of management, management structures and systems that contributed to the perfomance of Femi-2, specifically in the areas of operations, maintenance, engineering, and security; 2) actions l
planned to ensure readiness of the facility to support restart and l
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power escalation after testing at each power ascension plateau; and
- 3) actions to be taken to improve general regulatory and operational performance during and after the startup testing phase.
The licensee responded to the 10 CFR 50.54(f) letter on January 29, 1986 (Attachment 4 to Enclosure 1). The key elements of the licensee's action to correct the management problems are as follows:
- 1. Increased direct involvement by the Chief Executive Officer (CEO), who took personal charge of the restart efforts.
- 2. The creation of an Independent Overview Committee (IOC) comprised of experienced managers from the nuclear industry, who reported directly to the CEO. The comittee was to make an early review of the management issues and was to be retained to review the restart of the plant and power escalation activities.
- 3. The organization was modified so the Quality Assurance (QA) program reported directly to the company president until such time as an experienced senior vice president could be hired. '
- 4. The licensee also made a comitment, subsequently fulfilled, to hire a senior vice president who was experienced in the nuclear industry.
- 5. The vice president for nuclear operations was provided with an advisor who was experienced in nuclear operations.
- 6. The Company comitted to actively recruit for a new vice president for engineering and for a new security director experienced in the nuclear business.
The IOC perfonned an early and candid evaluation of management adequacy (Attachment 5 to Enclosure 1) and its recommendations led to some of the above corrective actions. The IOC also met with Region III on June 3, 1986, at which time it informed the NRC that it had no major problems with a Fenni restart. The IOC also informed the Comission of this view during a briefing conducted on July 7, 1986, related to Fermi 2 restart. The licensee addressed the initial IOC concerns in a letter to Region III on July 2,1986, in which it transmitted a " Restart Report" (Attachment 6 to Enclosure 1).
In addition to the ROIP, which mainly addressed operations, the licensee comitted in its January 29, 1986 response to develop a broader Nuclear Operations Improvement Plan (NOIP) as well (Attachment 7 to Enclosure 1). This addresses planning, account-ability, attitude, comunications, teamwork, and training for the entire operation. The plan was fully implemented on July 1, 1986, and will be monitored by the NRC.
To correct concerns in the maintenance area, the licensee modified the work order process to more clearly state the post-maintenace testing requirements. Additional documentation requirements were added which must be met before the shift operating authority can 1-3
e accept a component or system for service. Instrument and repair technicians were provided additional training and on-the-job instructions on proper maintenance techniques. Additional items are being monitored by management to track the maintenance backlog, and action levels have been defined to identify when additional management actions will be initiated.
In the engineering area, the licensee reviewed past modification packages to verify the accuracy of earlier results and to correct any identified problem areas. Changes were made in design control procedures to assure future compliance. Outside engineering firms also were retained to review the engineering changes and calculations. Additional audits were conducted of engineering design packages and associated documentation and procedures have been changed. A few hardware changes had to be made as a result of these reverifications. The licensee presented its findings to Region III on June 26, 1986, and a copy of its presentation is attached (Attachment 8 to Enclosure 1). NRC inspections have found these changes to be acceptable (Inspection Report 50-341/86012 will '
be issued shortly describing these issues).
l With regard to the concerns on security, the licensee developed a Security Improvement Plan (Attachment 9 to Enclosure 1) designed to
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achieve a greater degree of procedural adherence, improve monitoring-l systems, increase management effectiveness and clearly define security
- responsibilities. The hiring of a top level manager is still being
! pursued and the NRC is closely monitoring the licensee's resolution of this issue.
During the time since July 1985, the staff had held numerous meetings and enforcement conferences with the utility to understand and resolve the issues that have been raised. Continuing inspection i activities have been ongoing to verify the licensee's actions with regard to its proposed changes and corrective actions. The results of the inspection findings have resulted in the issuance of certain escalated actions and may result in additional actions of this nature. A civil penalty in the amount of $50,000 was issued on May 20, 1986 (Attachment 10 to Enclosure 1), based on a lack of management effectiveness in the security area as evidenced by numerous violations of the NRC approved security plan. As mentioned above, the licensee has developed a security Performance Improvement Plan which addresses the major issues required to correct deficiencies in the security area. Nevertheless, the licensee has continued to have some problems in this area, and the actions initiated have not yet been fully effective in this area. Augmented inspection in this area will continue.
Another proposed Civil Penalty was issued on July 3, 1986 in the amount of $300,000 (Attachment 11 to Enclosure 1). It was based on personnel errors and management weaknesses associated with the premature criticality of July 1-2, 1985. The licensee fomally responded and paid the penalty in full; all of the issues identified have been addressed in the licensee's response to the 10 CFR 50.54(f) letter, and it is the staff's view that the programatic and structural changes which the licensee has initiated 1-4
will be effective in addressing the concerns that existed. The i proposed civil penalty was accompanied by an immediately effective Order Modifying the License which required that 1) the licensee demonstrate that the Nuclear Shift Supervisor involved in the incident'has been retrained and reexamined before being allowed to resume control room responsibilities, and 2) that management implement a control room audit program.
An additional civil penalty will be proposed for numerous other technical specification and license violations that occurred in the middle of 1985. These multiple violations will be consolidated and focus on problems with the licensee's control over work activities.
It is the staff's view that the licensee's broad managment changes and its ROIP and NOIP program have resolved the problems which caused these violations to occur.
In early October 1985 the licensee recognized that the delays being
! experienced would prevent completion of the startup test program
! befort. the arrival of the date (December 21,1985) at which '
containment inerting would be required pursuant to the Comission's regulations. An exemption and license amendment were requested which would pemit postponement of the inerting of the Fermi-2 primary containment until either the completion of the startup test i program or until the reactor has operated for 120 effective full power days, whichever is earlier. The staff granted the requests, issuing the exemption and confoming license amendment by letter 2 dated July 30, 1986, including the staff's safety evaluation
- (Attachment 12 to Enclosure 1).
Questions of potential wrongdoing have been raised in connection with some of the violations, and are being addressed by NRC Office of 1
Investigation. One such investigation was conducted on the premature criticality event and the licensee's timeliness of informing the NRC of that activity. The report was given to the Justice Department which formally has declined to pursue any criminal sanctions against the company. Other OI investigations are ongoing that relate to the engineering area.
In sumary, the Femi facility has exhibited a higher than normal degree of technical and regulatory problems since the receipt of its full power license. The staff has continually assessed the problems as they have arisen, has met with licensee management, and has taken aggressive actions to restrict plant operation in order to identify problems and concerns, and to assure licensee corrective actions. It is the staff's view that the corrective action programs initiated by the licensee have substantially addressed all of the ma.for problem areas identified. The staff is continuing to resolve c.ertain specific engineering and hardware issues and will not allow facility restart until these particular issues have been addressed.
Implementation of the developed program will also be assessed by the staff through an Augmented Inspection Program prior to and during the initial weeks after the facility is restarted.
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ENCLOSURE 2 l
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s Enclosure 2 Response to Request on Investigations Request: In addition, the report should identify and indicate the status of any on-going investigations or examinations by the Commission into current problems at the Fermi-2 reactor, including the schedule for concluding all such investigations or examinations.
Response: As of July 15, 1986, the Office of Investigations has three on-going investigations related to the Fermi-2 facility. Two of the investiga-l tions involve potential reporting violations pertaining to representa-tions that (1) specific seismic qualifications of design changes, and (2) analysis of structural embed plates had been completed when they had not. The investigations have been requested to determine whether or not the reports were intentionally misrepresentative. Investiga-tive effort, including interviews of pertinent individuals, has '
aliaady been conducted, however, it is anticipated that all investi-gative activity and the writing of the report will not be completed -
before late September 1986. The third investigation concerning a
- potential safeguards violation concerns a possible false statement by
, a licensee employee. The potential violation involves the utilization of an " unsecured" computer to store safeguards / security data sub-sequent to being advised by the NRC safeguards inspector that the computer could not be used for storing safeguards information because it did not have a secure system. The investigation was requested to
, resolve the allegation that a statement was made by the licensee employee that he was aware that the computer system was not secure.
Both the investigation and report should be complete by late October 1986.
DIA has no ongoing investigations which we believe fall within the definition of " current problems at Fermi-II". There is, however,
- one related ongoing OIA investigation concerning EDO and staff actions with respect to the OI Fermi report issued on February 21, i 1986. All field work has been completed in that investigation and a <
- report is in preparation.
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ENCLOSURE 3 i
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Enclosure 3 Response to Request on Mark I Containment Request: As you may know, the County Comissioners of Monroe County in Michigan recently adopted a resolution calling for a congressional inquiry into the safety and reliability of the Mark I reactor.
Please include in your report relevant information regarding the i safety of these reactors, in particular the containment system, and fdent4fy problems at other Mark I reactors which may indicate potential problems at Ferni-II.
Response: As the name indicates, a boiling water reactor (BWR) is a reactor in which the water fed to the reactor core boils in the reactor
- vessel and then passes as steam directly out to the turbine-generator where its energy is converted to electricity. The exhausted steam, after condensation, is returned to the reactor as i.
feedwater. Figure 1 to enclosure 3 shows a simple schematic of a BWR '
plant. The reactor is enclosed in a special containment structure.
The feedwater enters and the steam leaves this containment structure i through multiple, large diameter pipes equipped with redundant valves 4
which can be closed in an emergency. In the pressure suppression containment which is used in all large U.S. BWRs, a very large quantity of water, up to one million gallons, is stored in a special compartment of the containment called the suppression pool. Many auxiliary and emergency cooling systems are provided to pump cooling water into the reactor and to cool the containment atmosphere and its suppression pool. If a pipe breaks by accident, the containment closes to isolate the reactor in the containment and many cooling i
l systems are called into play to cool the reactor and the suppression pool, removing the stored energy and heat generated by radioactive decay. ,
Thus, the BWR is an open system removing large quantities of energy to nearby equipment which, in emergencies, converts to a closed system, basically relying on external cooling of the containment to remove the bottled-up energy. The most comon type of pressure
! suppression containment in the U.S. is the Mark I type shown in i Figure 2 to enclosure 3, which is used in the 24 U.S. BWRs listed in j Table 1 to enclosure 3. The reactor is contained in the drywell i portion of the containment, shaped like an electric light bulb
! standing upside down. The suppression pool partially fills a toroidal shell around the base of the " bulb" and a series of ducts is installed to guide steam and other releases into the suppression pool which quenches the steam and also absorbs much of the radioactive material (exceptgases),
i " Severe accidents" is the tenn most comonly used to describe accidents in which the reactor core is severely damaged. As happened at Three Mile Island, prolonged loss of core cooling can allow the heat of l radioactive decay in the core to build up to the point that the fuel I I
begins to disintegrate, the rirconium metal cladding melts or reacts j with residual steam to fann combustible hydrogen, and even the i ceramic uranium oxide fuel pellets can melt. A great deal of atten-tion is being given to understanding the behavior of reactors and l their containments in severe accidents, especially since the Three 1 l l
l o l Mile Island accident. The objectives are to ensure that the likeli- '
hood of core melt accidents is very low and that, should one occur, there is substantial assurance that the containment will mitigate its consequences.
The severe accident behavior of a BWR with a Mark I containment, the Peach Bottom plant, was assessed in the Reactor Safety Study 1 (WASH-1400 or NUREG-75/014) which was published in 1975. That study indicated a relatively low overall risk for the BWR, principally due to its ability to prevent core melt. The containment was estimated to provide very little mitigation of core melt consequences because the buildup of pressure under accident conditions would be a direct
- cause of containment failure unless adequate cooling was preserved.
Consistent with operating procedures in place in 1975, the Study assumed little effort by the reactor operators which might effectively preserve the containment's integrity.
The situation, more than ten years later, is different and still changing for the better. It is recognized today that molten core material melting into the ground through the thick containment base '
is not the principal threat; rather, it is an atmospheric release of radioactive material which is the principal threat. The principal j factors which can cause containment failure with atmospheric release <
! are hydrogen ignition, gas overpressure buildup to rupture, and direct attack of the drywell by core melt debris. The general l
situation for each of these is sumarized as follows:
Hydrogen Ignition Recognizing that combustible hydrogen can be generated and released in severe accidents, all Mark I containments now are purged and i filled with inert nitrogen gas during operation so that even if hydrogen gas is formed it has insufficient oxygen available to l support combustion. Remaining questions in this area relate to
- how long the containment may be without this inert atmosphere in
! order to permit inspections, and how air might leak in or hydrogen leak out to nearby rooms under accident conditions.
Overpressure Failure Careful analysis indicates that a typical Mark I containment can
' withstand pressures of more than twice the design pressure without rupture. Nevertheless, severe accidents in the extreme can generate
! such pressures and cause containment rupture. Overpressure damage
- control procedures have been developed for pressure suppression
' containments and are already in place for operator use. With these i procedures,the containment remains closed for most accident con-
! ditions; but, if overpressure failure threatens, large vent valves l above the suppression pool chamber are opened so that the excess i pressure is released gradually by bubbling the releases through the pool, forming a filtered vent containment system. With this path
! assured, virtually nothing but the noble gases are released. The i radioactive noble gases Jose a modest exposure threat offsite only in the area very close to tie plant. A number of questions are being pursued in this area. All the plants have suitably large l
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s vent valves and ducts but they vary one to another in the ability to open these valves under accident conditions. The valves are designed
- for highly reliable closure, not opening. Consideration is being
- given to modifying valve controls. In addition, the vent ductwork downstream of the valves may warrant modification. In most plants it
- is fairly light gauge ductwork and might be breached in accident 1 venting. If so, consideration is being given to the effects of 1 j secondary release of radioactive gas, hydrogen, and perhaps steam j into the reactor building.
! Direct Attack I The core melt debris, since it has melted through the reactor vessel I into the drywell may, by direct radiation of he.at, cause failure of connections in the drywell shell; or the debris, if sufficiently
- fluid, may flow out to the wall and melt through the steel. The Mark I containments have one or more spray systems in the drywell which i
are able to spray water along the walls and onto the floor of the drywell inhibiting direct attack. Concerns in this area are in three '
general areas: core debris modeling, shell and concrete attack
.j modeling, and spray reliability. In the first area, it is recognized that a molten reactor core, to melt through the bottom of a BWR, must dissolve a very large amount of inert metal in the lower reactor j vessel, probably diluting the core melt. The key question is whether 1 the melt would come out moving sluggishly like Hawaiian volcano lava
! or as a hot free flowing liquid. The latter is the more threatening l condition.
i i If core melt debris reaches the concrete floor and steel shell of the i wall, it is important to understand that the path to the outside that i might be opened bypasses the beneficial scrubbing of radioactive j material passing through the pool.
As noted earlier all these plants have drywell spray systems, but
- they are designed as a secondary mode of operation for a reactor ,
i safety system. Strong consideration is being given to enabling <
hookup of these systems to fire protection systems so that spray -
l capability is almost always available, f
, Substantially different emergency operating procedures and training I were put in place at all reactors after the Three Mile Island i accident; further improvements in these procedures are still being made. For the Mark I containments both industry and NRC studies are being used to identify the best combined strategy for procedures and perhaps some changes in equipment such as alternate vent paths, or
, improved valve cperability. The Mark I studies are being given highest priority by the NRC staff and the industry. The expectation
! is that, with modest improvements of this type, one can achieve substantial assurance of core melt consequences mitigation by a Mark I containment.
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TARE 1 BOILING WATER RIACTORS WITH MARK I CONTAINMENTS LICENSED OPERATING PLANT POWER LICENSE '
LEVEL DATE COUNTY STATE UTILITY NAME 3293 12/20/73 LIMESTONE COUNTY At TVA BROWNS FERRY 1 TVA 3293 08/02/74 LIMESTONE COUNTY AL BROWNS FERRY 2 TVA 3293 08/18/76 LIMESTONE COUNTY AL BROWNS FERRY 3 CAROLINA POWER & LIGHT 2436 11/12/76 BRUNSWICK COUNTY NC BRuhSWICK 1 CAROLINA POWER & LIGHT 2436 12/27/74 BRUNSWICK COUNTY NC BRUNSWICK 2 NEBRASKA PUBLIC POWER DISTRICT 2381 01/18/74 NENEHA COUNTY NE COOPER 2527 12/22/69 GRUNDY COUNTY IL COMMONWEALTH EDISON DRESDEN 2 2527 03/02/71 GRUNDY COUNTY IL COMMONWEALTH EDISON DRESDEN 3 Y' 1658 02/22/74 LINN COUN1Y IA IOWA ELECTRIC POWER & LIGHT DUANE ARNOLD 07/15/85 MONR0E COUNTY MI DETROIT EDISON FERMI 2 3292 OSWEGO COUNTY NY POWER AUTHORITY OF STATE OF NY FITZPATRICK 2436 10/17/74 APPLING COUNTY GA GEORGIA POWER HATCH 1 2436 10/13/74 APPLING COUNTY GA GEORGIA POWER HATCH 2 2436 06/13/78 04/11/86 SALEM COUNTY NJ PUBLIC SERVICE ELECTRIC & GAS HOPE CREEK 1 3293 NEW LONDON CT NORTHEAST NUCLEAR ENERGY MILLSTONE 1 2011 10/16/70 WRIGHT COUNTY MN NORTHERN STATES POWER MONTICELLO 1670 01/19/71 OSWEGO COUNTY NY NIAGARA MOHAWK POWER NINE MILE POINT 1 1850 08/22/69 OCEAN COUNTY NJ GPU NUCLEAR CORP OYSTER CREEK 1 1930 08/01/69 3293 12/14/73 YORK COUNTY PA PHILADELPHIA ELECTRIC PEACH BOTTOM 2 3293 07/02/74 YORK COUNTY PA PHILADELPHIA ELECTRIC PEACH BOTTOM 3 PLYMOUTH COUNTY MA BOSTON EDISON PILGRIM 1998 06/08/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON QUAD CITIES 1 2S11 12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON QUAD CITIES 2 2511 12/14/72 WINDHAM COUNTY VT VERMONT YANKEE NUCLEAR POWER VERMONT YANKEE 1593 02/02/73 e *
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