ML19262C654
| ML19262C654 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/24/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| IEB-79-08, IEB-79-8, NUDOCS 8002150231 | |
| Download: ML19262C654 (20) | |
Text
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January 24, 1980
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Docket No. 50-219 Mr. I. R. Finfrock, Jr.
Vice President - Generation Jersey Central Power & Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
Dear Mr. Finfrock:
SUBJECT:
NRC STAFF EVALUATION OF JERSEY CENTRAL POWER & LIGHT COMPANY RESPONSES TO IE BULLETIN 79-08 FOR OYSTER CREEK NUCLEAR GENERATING STATION We have completed our review of the infomation that you provided in your letter dated April 25, 1979 in response to IE Bulletin 79-08 for the Oyster Creek Nuclear Generating Station. We have also cor 'eted our review of the supplemental information that you provided in your setter of August 9, 1979.
We have concluded that you have taken the appropriate actions to meet the requirements of each of the eleven action items identified in IE Bulletin 79-08. A copy of our evaluation is enclosed.
As you know, NRC staff review of the Three Mile Island, Unit 2 (TMI-2) accident is continuing and other corrective actions may be required at a later date.
For example, the Bulletins and Orders Task Force is conduct-ing a generic review of operating boiling water reactor plants. Specific requirements for your facility that result from this and other TMI-2 investigations will be addressed to you in separate correspondence.
Sincerely, M
Dennis L. Zieman, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
NRC Staff Evaluation
cc w/ enclosure:
See next page 8002150 1 -
Mr. I. R. Finfrock, J r. January 24, 1980 CC G. F. Trowbridge, Esquire Gene Fishcr Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 GPU Service Corporation ATTN: Mr. E. G. Wallace Mark L. First Licensing Manager Deputy Attorney General 260 Cherry Hill Road State of New Jersey Parsippany, New Jersey 07054 Department of Law and Public Safety Environmental Protection Section Anthony Z. Roisman 36 West State Street Natural Resources Defense Council Trenton, New Jersey 08625 917 15th Street, N. W.
Washington, D. C.
20006 Joseph T. Carroll, J r.
Plant Superintendent Oyster Creek Nuclear Generating Steven P. Russo, Esquire Station 248 Washington Street P. O. Box 388 P. O. Box 1060 Forked River, New Jersey 08731 Toms River, New Jersey 08753 Joseph W. Ferraro, Jr., Esquire Director Technical Assessment Deputy Attorney General Division State of New Jersey Office of Radiation Programs Department of Law and Public Safety (AW-459) 1100 Raymond Boulevard U. S. Environmental Protection Newark, New Jersey 07012 Agency Crystal Mall #2 Ocean County Library Arlington, Virginia 20460 Brick Township Branch 401 Chambers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region II Office Mayor ATTN: EIS COORDINATOR Lacey Township 26 Federal Plaza P. O. Box 475 New York, New York 10007 Forked River, New Jersey 08731 Robert M. Lazo, Esq., Chaiman Comi ssioner Atomic Safety and Licensing Board Department of Public Utilities U. S. Nuclear Regulatory Comission State of New Jersey Washington, D. C.
20555 101 Comerce Street Newark, New Jersey 07102
EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-08 JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STAf?ON DOCKET NO. 50-219
Introduction By letter dated April 14, 1979, we transmitted Office of Inspection and Enforcement (IE)Bulletin 79-08 to Jersey Central Power & Light Company (JCP&L or the licensee).
IE Bulletin 79-08 specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occurred at Three Mile Island, Unit 2 (TMI-2) on March 28, 1979.
By letter dated April 25, 1979, JCP&L provided responses to Action Items 1 through 10 of IE Bulletin 79-08 for the Oyster Creek Nuclear Generating Station (0yster Creek).
The NRC staff review of the JCP&L responses led to the issuai.c? of requests fer additional information regarding the JCP&L responses to certain action itees of IE Bulletin 79-08.
These requests were contained in a letter dated July i'0, 1979.
By letter dated August 9, 1979, JCP&L responded to the staff's reques'.s for additional information and provided the response to Action Item 11 of IE Bulletin 79-08.
The JCP&L responses to IE Bulletir 79-08 provided the basis for our evaluation presented below.
In addition, the actions taken by the licensee in response
'to the bulletin and subsequent NRC requests were verified by onsite inspections by IE inspectors.
Evaluation Each of the 11 action items requested by IE Bulletin 79-08 is repea'ted below followed by our criteria for evaluating the response, a summary of the licensee's response and our evaluation of the response.
1.
Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the THI-2 March 28,1979 accident included in Enclosure 1 to IE Bulletin 79-05A.
a.
This review should be d'rected toward understanding:
(1) the extreme seriousness ar,d consequences of the simultaneous blocking of both trains of a safety system at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; and (3) the necessity to
. systematically analyze plant conditions and parameters and take appropriate corrective action.
b.
Operational personnel should be instructed to (1) not override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a of this bulletin); and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available.
c.
All licensed operators and plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be documented in plant records.
The licensee's response was evaluated to determine that (1) the scope of review was adequate, (2) operational personnel were properly instructed and (3) personnel participation in the review was documented in plant records.
The licensee's response dated April 25, 1979 described the review sessions which emphasized the lessons learned from the Three Mile Island incident and instructed operators not to override automatic action of engineered safety features unless their continued operation will result in unsafe plant conditions and to confirm plant parameter indications.
The licensee's supple-mental response dated August 9, 1979 confirmed that all but two licensed operators, plant management, and supervisors with operational responsibilities completed review sessions by April 25, 1979.
The two licensed operators participated in a review that was completed by May 31, 1979.
We conclude that the licensee's scope of review, instructions to operating personnel and documented participuka satisfy the intent of IE Bulletin 79-08, Item 1.
2.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate containment isolation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
The licensee's response was evaluated to verify that containment isolation initiation design and procedures had been reviewed to assure that (1) manual or automatic initiation of containment isolation occurs on automatic initiation of safety injection and (2) all lines (including those designed to transfer radioactive gases or liquids) whose isolation does not degrade cooling capability or needed safety features were addressed.
The licensee's April 25, 1979 response noted that the primary containment isolation signals are the same as those which initiate core spray, i.e.,
low-low reactor water level or high drywell pressure.
Several systems were identified which penetrate containment and do not isolate on primary containment isolation.
In its supplemental response dated August 9, 1979, the licensee described modifications that will meet the criteria and provided justification for lines that do not isolate.
The reactor water cleanup system and the shutdown cooling system automatically isolate on low-low reactor water level, however, modifications to also isolate on high drywell pressure are being reviewed and will be made during the January 1980 outage.
The main steam drain line is isolated on low-low reactor water level, main steam line high radiation, low main steam line pressure or steam line break.
During normal operation, the valves are closed and the drain lines do not connect to the drywell so isolation on high drywell pressure is considered unnecessary.
The isolation condenser vent lines are outside of primary containment and closure of these lines on high drywell pressure could possibly render the system inoperable.
Therefore, they will not be modified.
The instrument air line will be modified to incorporate an isolation valve outside primary containment during the 1981 refueling outage.
In the interim, emergency procedures contain instructions to isolate this line from the drywell.
The reactor building closed ccoling water system is a closed system and can be isolated using controls available in the control room.
Automatic isolation would limit the use of this system during transients when needed.
Therefore, instructions were incorporated in the emergency procedures to isolate this system during pertinent emergencies.
The torus vacuum breaker lines protect the suppression chamber against exceeding external design pressure.
The check valves and the air-operated butterfly valves are normally closed.
The butterfly valves open when the chamber pressure is 0.5 psi less than reactor building pressure.
A high drywell pressure isolation signal to these valves will be added during the 1981 refueling outage.
The containment ventilation exhaust lines valves can only be opened durirg an isolation signal by daliberate use of the key-lock bypass switch as long as the mode switch is out of the run mode.
Emergency procedures direct the operator to the proper method of containment purging during an emergency.
We conclude that the licensee's review of containment isolation initiation design and procedures satisfy the intent of IE Bulletin 79-08, Item 2.
3.
Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e.g., RCIC) that are used when the main feedwater system is not operable.
For any manual action necessary, describe in summary form the procedure by which this action is taken in a timely sense.
The licensee's response was reviewed to assure that (1) it described the automatic and manual actions necessary for the proper functioning of the auxiliary heat removal systems when the main feedwater system is not operable and (2) the procedures for any necessary manual actions were described in summary form.
When the main feedwater system is not operable, heat removal during isolation is accomplished by automatic initiation of the isolation condenser on high
reactor pressure or low-low reactor level.
Makeno water is provided through manual action in the control room following precedures and using water level instrumentation.
For reactor temperatures below 350 degrees Fahrenheit, the reactor shutdown cooling system may be manually initiated for continuous heat removal.
We co'iclude that the licensee's procedural summary of automatic / manual actions neces3ary for the proper functioning of auxiliary heat removal systems used when tne main feedwater system is inoperable satisfies the intent of IE Bulletin 79-08, Item 3.
4.
Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems.
Describe other reuundant instrumentation which the operator might have to give the same information regarding plant status.
Instruct operators to utilize other available information to initiate safety systems.
The licensee's response was evaluated to determine that (1) all uses and types of vessel level indication for both automatic and manual initiation of safety systems were addressed, (2) it addressed other instrumentation available to the operator to determine changes in reactor coolant inventory and (3) operators were instructed to utilize other available information to initiate safety systems.
The licensee's April 25, 1979 response lists the types and uses of water level indication for automatic and manual initiation of safety systems. These redundant and diverse systems provide water level monitoring under all conditions or reactor operation.
These level indicators include:
(1 )
Four Yarway Type 4316 differential pressure indicating switches.
calibrated to compensate for conditions of reference leg temperature (275 degrees Fahrenheit), weight of steam and density of water at operating pressure are utilized to provide:
(a) Local indication on two instrument racks in the reactor building.
(b) Retransmitted analog signals to two remote indicators on panel 5F in the control room.
(c) Low level scram contact actuation.
(d) High level turbine trip contact actuation.
(e) High/ low water level annunciator in control room.
(2) Four Yarway Type 4316 differential pressure indicating switches (calibrated identical to (1) above) are utilized to provide:
(a) Local indication on two instrument racks in the reactor building.
(b) Low-low level contact actuation for initiating the following safety systems:
1 Core Spray Containment Spray Isolation Condenser Actuation Recirculation Pump Trip (c) Low-low level contact actuation for initiating reactor isolation, primary containment isolation, and seccndary containment isolation with standby gas treatment system initiation.
(d) Low-low water level annunciator in control room.
(3) Four Barton Type 278 differential pressure indicating switches are utilized to provide:
(a) Uncompensated local level indication on two instrument racks in reactor building.
(b) Reactor vessel low-low-low level contact actuation for initiating the automatic depressurization system. The setpoint of the switch is compensated for the density of water in the reference leg and variable leg taps.
(4) Two GE/MAC Type 553 differential pressure electronic transmitters, in conjunction with two steam pressure transmitters, generate two analog level signals, automatically compensated for density of water in vessel which provide:
(a) Two compensated narrow range level indications in control room on panel SF.
(b) One selected and one standby compensated level sigrial for:
Three element or single element feedwater control Reactor level recorder on panel 5F (5) One GE/MAC Type 553 differential pressure electronic transmitter provides uncompensated wide range level indication on panel SF in the control room.
Additional instrumentation which the operator can use to determine changes in the reactor coolant inventory or other abnormal conditions are:
Drywell F gh pressure Suppression pool high temperature Safety relief valve high temperature Feedwater - steam flow rata mismatch Containment dewpoint and temperature Decreasing reactor pressure Suppression pool water level increasing Identified and unidentified equipment and floor sump leak rates Equipment area temperature Area radiation monitors (outside containment)
The operators have been instructed to utilize other available information to initiate safety systems.
The instructions were part of the review provided in response to Item 1 of IE Bulletin 79-08 and were completed as specified in that item.
We conclude that the licensee's description of the uses and types of reactor vessel level / inventory instrumentation and instructions to operators regarding the use of this information satisfies the intent of IE Bulletin 79-08, Item 4.
5.
Review the actions directed by the oper2 ting procedures and training instructions to ensure that:
a.
Operators do not override automatic actians of engineered safety features, unless continued operation of engineescd safety features will result in unsafe plant conditions (e.g.,
vessel integrity).
b.
Operators are provided ' additional information and instructions to not rely upon vessei level indication alone for manual actions, but to also examine other plant parameter indications in evalating plant conditions.
The licensee's response was evaluated to determine that (1) it addressed the matter of operators improperly overriding the automatic actions of engineered safety features, (2) it addressed providing operators with additional informa-tion and instructions to not rely upon vessel level indication alone for manual actions and (3) tnat the review included operating procedures and training instructions.
The licensee, in its April 25, 1979 response, stated that the review of all applicable emergency procedures and training instructions involving operation of engineered safety features would be accomplished by April 25, 1979.
Any procedures which require additional instructions in the precautions of overriding automatic action of the engineered safety features, unless continued operation will result in unsafe plant conditions would be changed by mecns of change requests.
Supplemental information was provided concerning other instrumentation which could be helpful in evaluating plant conditions.
The licensee confirmed that tha review was completed in its August 9, 1979 submittal.
We conclude that the licensee's review of operating procedures and training instructions satisfies the intent of IE Bulletin 79-08, Item 5.
6.
Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
Also review related procedures, such as those for maintenance, testing, plant and system start-up, and supervisory periodic (e.g.,
daily / shift checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their prcper positions during all operational modes.
. The licensee's response was evaluated to assure that (1) safety-related valve positioning requirements were reviewed for correctness, (2) safety-related valves were verified to be in the correct position and (3) positive centrols were in existence to maintain proper valve position during normal operation as well as during surveillance testing and maintenance.
The licensee's response dated April 25, 1979 described the review of safety-related valve positions, positioning requirements and controls to ensure proper operation of the engineered safety features.
The licensee's supplemental response dated August 9, 1979 indicated that the review of Engineered Safety Features and cafety-related procedures (including valve check-offs) and maintenance, testing, plant and system startup, and supervisory periodic surveillance procedures (including pnst-maintenance / surveillance valve position verification) would be completed by December 31, 1979.
The licensee has confirmed by telephone that the review described in the August 9,1979 re-sponse was completed by December 31, 1979.
We conclude that the licensee's review of safety-related valve positioning requirements, valve positions and positive controls to maintain proper valve positions satisfies the intent of IE Bulletin 79-08, Item 6.
7.
Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists, and b.
Whether such systems are isolated by the containment isolation signal.
c.
The basis on which continued operability of the above features is assured.
. The licensee's response was evalu ted to determine that (1) it addressed all systems designed to transfer potentially radioactive gases and liquids out of primary containment, (2) inadvertent releases do not occur on resetting engineered safety features instrumentation, (3) it addressed the existence of interlocks, (4) the systems are isolated on the containment isolation signal, (5) the basis for continued operability of the features was addressed and (6) a review of the procedures was performed.
In its April 25, 1979 response, the licensee reported that with the exception of high main steam line radiation, no interlocks exist that are based on high radiation conditions to prevent transfer of radioactive gases / fluids from the containment.
In its supplemental response dated August 9, 1979, the licensee stated that inadvertent transfer of gases or liquids out of containment is precluded by incorporation in the emergency procedures of steps to not reset prim &ry containment isolation until samples can be taken of the liquids and atmosphere within containment.
The emergency procedures were modified by June 1, 1979.
The licensee's supplemental response also identifed the systems designed to transfer radioactive gases or liquids outside of containment as:
Drywell Equipment Orain Tank System Drywell Floor Drain Sump Drywell Purge Isolation Valves Drywell Hitrogen Relief Vent Valves Torus Vent Valves Drywell and Torus Oxygen Sample Lines Cleanup System (letdown)
. All but - e cleanup system isolate on primary containment.
The cleanup system letdown line could be used to transfer reactor coolant to either the main condensers or the station radwaste facility. The procedure that directs tha control room operators. to use the letdown system was reviewed and changed before September 15, 1979 to provide instructions in cases of potentially high coolant activity.
The proposed modification identified in Item 2 will effectively isolate this transfer pathway on those signals that will isolate the reactor water cleanup system.
Valves serving as the isolation barrier for primary containment are leak rate tested during each refueling outage to assure valve integrity. Additionally, the setpoint of the instrumentation utilized to initiate primary containment isolation is verified monthly.
Furthermore a functional primary containment isolation test is performed at each refueling outage to verify closure of all automatic isolation valves by the initiation signal.
Applicable emergency procedures concerning systems designed to transfer potentially radioactive gases and liquids out of the primary containment were reviewed by April 25, 1979.
We conclude that the licensee's review of systems designed to transfer radioactive gases and liquids out of primary containment to assure that undesired pumping, venting, or other release of radioactive liquids and gases will not occur satisfies the intent of IE Bulletin 79-08, Item 7.
8.
Review and modify as necessary your maintenance and test procedures to ensure that they require:
a.
Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service.
b.
Verification of the operability of safety-related systems when they are returnec to service following maintenance or testing.
Explicit notification of involved reactor operational personnel c.
whenever a safety-related system is removed from and returned to service.
. The licensee's response was evaluated to determine that operability of redundant safety-related systems is verified prior to the removal of any safety-related system from service. Where operability verification appeared only to rely on previous surveillance testing within Technical Specification intervals, we asked that operability be further verified by at least a visual check of the system status to the extent practicable, prior to removing the redundant equipment from service. The response was also evaluated to assure provisions were adequate to verify operability of safety-related systems when they are returned to service following maintenance or testing. We also checked to see that all involved reactor operational personnel in the oncoming shift are explicitly notified during shift turnover about the status of systems removed from or returned to service since their previous shift.
The licensee's response dated April 25, 1979 indicated that administrative controls exist that require verification of the operatility of redundant safety-related systems prior to removal of a safety-related system from service and instructions to veri *y system operability before return to service.
The licensee's supplemenM1 response dated August 9,1979 stated that the administrative procedures were chdoged to incorporate a form which is used to
-control the activities required to remove a system from and place a system in service.
This form provides the documentation of: (a) the tagging and switching done in removing a piece of equipment from service, (b) the testing required prior to removing equipment from service and (c) the testing required after placing the equipment back in service.
The licensee's April 25, 1979 response indicated that procedures existed that required that information about systems out of service be incluced in plant turnover.
In its supplemental respcase dated August 9, 1979, the licensee noted that Administrative Procedure 106 was revised to include a form to document the status of all safety systems and a portion of balance of plant systems. The form must be completed by the group shift supervisor, group operating supervisor and control room operator prior to leaving shift and must be read and signed by the oncoming personnel.
The review and modification of procedures to provide the necessary control for retest of systems prior to the need for operability has been completed.
We conclude that the licensee's review and modification of maintenance, test and administrative procedures to assure the availability of safety-related systems and operational personnel knowledge of system status satisfies the intent of IE Bulletin 79-08, Item 8.
9.
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour ci the time the reactor is not in a controlled or expected condition of opration.
Further, at that time an open continuous communication channel shall be established and maintained with NRC.
The licensee's response was evaluated to determine that (1) prompt reporting procedures required or were to be modified to require that the NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation and (2) procedures required or were to be modified to require the establishment and maintenance of an open continuous communication channel with the NRC following such events.
A special notification procedure has been prepared which identifies the senior person present (group shift supervisor on up to station manager) as the individual responsible for notifying the Region 1 of the Office of Inspection and Enforcement if the reactor is not in a controlied or expected condition of operation.
To facilitate open and continuous communication with the Region 1 office, two new outside telephone lines have been activated at the station.
One line terminates in a speaker phone in the control room and the other terminates in a speaker phone in the conference room adjacent to the station manager's office.
We conclude that the licensee's response satisfies the intent of IE Bulletin 79-08, Item 9.
10.
Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident
that would either emain inside the primary system or be released to the containment.
The licensee's response was evaluated to determine if it described the means or systems available to remove hydrogen from the primary system as well as the treatment and control of hydrogen in the containment.
The licensee, in its April 25, 1979 response, stated that the operating modes and procedures that address controlling significant amounts of hydrogen would be reviewed, and additional guidance, if required, would be provided to deal with hydrogen gas.
During reactor isolation, the top of the reactor vessel can be vented through the electromatic relief valves from the steam lines to the suppression pool.
The primary containment is inerted with nitrogen to less than five percent oxygen during normal operation.
The handling of the buildup of hydrogen in the primary containment is described in Amendment 66 to the Oyster Creek Facility Description and Safety Analysis Report, dated March 6, 1972.
.We conclude that the licensee's response satisfies the intent of IE Bulletin 79-08, Item 10.
11.
Propose changes, as required, to those technical specifications which must be modified as a result of your implementing the items above.
The licensee's response was evaluated to determine that a review of tht:
Technical Specifications had been made to determine if any changes were required as a result of implementing Items 1 though 10 of IE Bulletin 79-08.
The licensee reported in its letter dated August 9, 1979 that its review has shown that no changes to the Technical Specifications are required.
We conclude that the licensee's response satisfies the intent of IE Bulletin 79-08, Item 11.
16 -
o a
Conclusion Based on our review of the information provided by the licensee to date, we conclude that the licensee has correctly interpreted IE Bulletin 79-08.
The actions taken demonstrate the licensee's understanding of the concerns arising from the TMI-2 accident in reviewing their implementation on Oyster Creek operations, and provide added assurance for the protection of the public health and safety during the operation of the Oyster Creek Nuclear Generating Station.
References 1.
IE Bulletin 79-05, dated April 1,1979.
2.
IE Bulleti,79-05A, dated April 5, 1979.
3.
IE Bulletin 79-08, dated April 14, 1979.
4.
JCP&L letter, I. Finfrock to B. Grier, dated April 25, 1979.
5.
NRC Staff letter, D. Ziemann to I. Finfrock dated July 20, 1979.
6.
JCP&L letter, I. Finfrock to D. Ziemann, dated August 9, 1979,