ML19211C590
| ML19211C590 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/31/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Pilant J NEBRASKA PUBLIC POWER DISTRICT |
| References | |
| IEB-79-08, IEB-79-8, TAC-13206, TAC-42014, NUDOCS 8001140018 | |
| Download: ML19211C590 (16) | |
Text
- a - ~#o TERA
~
UNITED STATES g
8 NUCLEAR REGULATORY COMMISSION o
s
-a wAsmNGTON, D. C. 20555
- \\*****/
December 31, 1979 Docket No. 50-298 Mr. J. M. Pilant, Director Licansing & Quality Assurance Nebraska Public Power District P. O. Box 499 Columbus, Nebraska 68601
Dear Mr. Pilant:
SUBJECT:
NRC STAFF EVALUATION OF NEBRASKA PUBLIC POWER DISTRICT RESPONSES TO IE BULLETIN 79-08 FOR COOPER NUCLEAR POWER STATION We have completed our review of the information that you provided in your letters dated April 25 and July 5,1979 in response to IE Bulletin 79-08 for the Cooper Nuclear Power Station. We have also completed our review of the supplemental information that you provided in your letter of August 7, 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.
Si ncerel Thomas A. Ippolito, hief Operating Ret : tors Branch #3 Division of rperating Reactors
Enclosure:
NRC Staff Evaluation
cc w/ enclosure:
1744 006 See next page 8001140 6 f
Mr. J. M. Pilant Nebraska Public Power District 2-cc:
Mr. G. D. Watson, General Counsel Nebraska Public Power District P. O. Box 499 Columbus, Nebraska 68601 Mr. Arthur C. Gehr. Attorney Snell & Wilmer 3100 Valley Center Phoenix, Arizona 85073 Cooper Nuclear Station ATTN: Mr. L. Lessor Station Superintendent P. O. Box 98 Brownville, Nebraska 68321 Auburn Public Library 118 - 15th Street Auburn, Nebraska 68305 1744 007
EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-08 NEBRASKA PUBLIC POWER DISTRICT COOPER N'JCLEAR STATION DOCKET NO. 50-298 1744 008
Introduction By letter dated April 14, 1979, we transmitted Office of Inspection and Enforcement (IE)Bulletin 79-08 to Nebraska Public Power District (HPPD or the licensee).
IE Eulletin 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 (THI-2) on March 28, 1979.
By letter dated April 25, 1979, NPPD provided responses to Action Items 1 through 10 of IE Bulletin 79-08 for the Cooper Nuclear Station.
NPPD supplemented this response by a letter dated July 5, 1979 to provide the response to Action Item 11 of IE Bulletin 79-08.
The NRC staff review of the NPPD responses led to the issuance of requests for additional information regarding the NPPD responses to certain action items of IE Bulletin 79-08.
These requests were contained in a letter dated July 20, 1979.
By letter dated August 7, 1979, NPPD responded to the staff's requests for additional information.
The NPPD responses to IE Bulletin 79-08 provided the basis for our evaluation presented below.
In addition, the actions taken by the licensee in response to the bulletin requirements and subsequent NRC requests were verified through onsite inspections by IE inspectors.
Evaluation Each of the 11 action items requested by IE Bulletin 79-08 is repeated 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 TMI-2 March 28, 1979 accident included in Enclosure 1 to IE Bulletin 79-05A.
a.
This review should be directed toward understanding:
(1) the extreme seriousness and 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 1744 009
. - 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 rec ods.
The licensee's response dated April 25, 1979 stated that the description of the accident circumstances and chronology had been reviewed, and that operational personnel had been instructed.
The instruction satisfactorily addressed the overriding of automatically actuated engineered safety features, and the need for considering confirmatory signals in operational decisions.
The licensee's supplemental response dated August 7, 1979 confirmed that personnel participation in the required reviews had been documented in the plant records.
We conclude that the licensee's scope of review, instructions to operating personnel and documented participation satisfies 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 featuru or cooling capability, upon automatic initiation of safety injection.
1744 010
. 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 a review of the primary containment isolation design and procedures had been completed.
This review verified that manual valves are required to be closed, and that for automatic valves a safety injection signal will automatically initiate containment isolation of all valves whose isolation does not degrade needed safety features or cooling capability.
In its supplemental response dated August 7, 1979, the licensee confirmed that the review included all lines penetrating containment.
No changes to design or procedures were reported by the licensee.
We conclude that the licensee's review of containment isolation initiation design and procedures satisfies 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 unse.
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 whte the main feedwater system is not operable and (2) the procedures for any necessary manual actions were described in summary form.
The licensee, in its response dated April 25, 1979, stated that the high pressure coolant injection (HPCI) system and reactor core isolation cooling (RCIC) system function automatically if the main feedwater system is not 1744 011
~
_4-available. We acknowledge the capability of these systems to provide the required heat removal action.
The HPCI system initiates automatically on hip drywell pressure or low reactor vessel w.. level.
The RCIC system initiates automatically on low reactor vessel level.
In the automatic mode, the HPCI and RCIC systems will initiate at low reactor level and continue to pump water until the high reactor level trip point is reached or until the reactor vessel is depressurized, at which point the core spray and low pressure coolant injection systems could function automatically.
Following a high level trip, the HPCI system will restart automatically at the low level setpoint if steam is available.
The RCIC system must be reset locally or from the control room before it will restart.
The control room operator can take manual control of one or both systems and maintain a constant reactor water level.
By letter dated August 7, 1979, the licensee stated that the manual actions related to restart and continued RCIC operation are addressed in written procedures.
We conclude that the licensee's procedural summary of automatic / manual actions necessary for the proper functioning of auxiliary heat removal systems used when the 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 redundant 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.
1744 012
, The licensee's April 25, 1979 response stated that the emergency core cooling systems (ECCS's) are initiated automatically by level indicating switches arranged in a one-out-of-two-twice logic.
Separate from the 16 ECCS initiation switches, which indicate locally, are ten independent level indications in the control room.
The licensee augmented its response in its letter dated August 7, 1979 and described the range, location and function of these level indications.
The licensee also pointed out that additional instrumentation, which the operator can use to determine changes in the reactor coolant inventory, are the following:
Drywell equipment and floor drain sump flow recorders.
Drywell equipment sump temperature indicator.
Mismatch between reactor feedwater flow and steam flow recorders and indicators.
The suppression pool water level which is indicated, recorded and alarmed.
Three primary containment and one wetwell pressure indications.
Primary containment internal temperature (detected by 38 temperature elements of which four are used for wetwell pool temperature).
Drywell process radiation monitor which monitors particulate, gaseous and iodine activities, plus provides the capability of a grab sample.
Main steam line high radiation and main steam line high flow alarms.
Reactor water cleanup high flow alarm.
High area temperature (steem leak detection) alarms.
The operators have been instructed to utilize all available information to initiate safety systems. This is documented in the plant training records.
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.
1744 013
. 5.
Review the actions diiacted by the operating procedures and training instructions to ensure that:
Operators do not override automatic actions of engineered a.
safety features, unless continued operation of engineered safety features will result in unsafe plant conditions (e.g.,
vessel integrity).
b.
Operators are provided additional information and instructions to not rely upon vessel 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) that the review included operating procedures and training instructions.
The licensee, in its April 25, 1979 response, stated that operators had been instructed to not override automatic actions of engineered safety features.
In addition, plant operating personnel have been instructed to examine other plant parameters in evaluating plant conditions.
The licensee also reported that the review addressed operating procedures ano training programs.
We conclude that the licensee's review of operating procedures ar.d training instructions satisfies the intent of IE Bulletin 79-08, Item 5.
6.
Review all safety-related valve positions, posit;oning 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 p.ocedures, such as those for maintenance, testing, plant and system start-up, and supervisory periodic (e.g.,
daily / shift checks) ~ veillance to ensure that such valves are returned to their correct pos...o" following necessary manipulatiors and are maintained in their p-positions during all operational modes.
1744 014
. 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 controls were in existence to maintain proper valve position during normal operation as well as during surveillance testing and maintenance.
The licensee's response dated August 7, 1979 described the review of safety-related valve positioning requirements. The licensee also stated that safety-related valves were verified to be in their correct positions following the April 1979 outage.
By letter dated April 25, 1979, the licensee described the procedural co..trols which it considered adequate to maintain proper valve position during operation, testing, and maintenance.
The positions of safety-related valvas are controlled with either control room position indication or valve position seals.
We conclude that the licensee's review of safety-related valve positicaing 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 pr'aary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
In particular, ensure tnat 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 fedtures is assured.
The licensee's response was evaluated to determine that (1) it addressed all systems designed to transfer potantially radioactive gases and liquids out of 1744 015
3 8-primary containment, (2) inadvertent releases do ccur on resetting engineered safety features instrumentation, (3) it addressed the existence of interlocks, (4) the systems are isolated on the co..
inment isolation signal, (5) the basis for continued operability of the features was addressed, and (6) a review of the procedures was perforned.
In its April 25, 1979 response, th licensee reported that there are three main systems which could transfer potentially radioactive gases and liquids out of primary containment.
They are the floor drain, equipment drain and ventilation systems.
The drain systems isolate on low reactor vessel level before the ECCS's initiate and on high drywell precsure concurrent with ECCS initiation.
During reactor operation, the ventilation system is isolated from primary containment.
Secondary containment isolates on the two conditions described above and on high radiation levels in the ventilation exhaust lines.
The standby gas treatment system also initiates on any of the three conditions described above, maintains a negative pressure in the secondary containment and filters the exhaust.
There are also drains from various auxiliary systems which are connected to the reactor coolant system.
However, the drain valves on these systems isolate on low reactor water level or high drywell pressure.
The isolation signals " seal in" and must be cleared by subsequent onerator
- a. tion.
Tha isolation valve timing, operability, logic and leak rates are tested periodically as set forth in the Technical Specifications. On this basis, continued operability of these isolation features is assured.
By letter dated August 7, 1979, the licensee reported that procedures caution the operator to ensure that the problem that caused the containment isolation has been rectified prior to resetting the instrumentation.
By a telephone conversation on October 11, 1979, the licensee advised us that it had reviewed its procedures and updated the:a as necessary to provide additional guidance concerning parameters which should be checked prior to resetting isolation instrumentation.
During the October 11, 1979 telephone conversation, the licensee also confirmed that its procedures protected against the inadvertent transfer of residual radioactive material following reset.
1744 016
9-We conclude that the licensee's review of systems designed to transfer radio-active 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, af 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 returned to service following maintenance or testing.
c.
Explicit notification of involved reactor operational personnel 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 pervious 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 tvaluated 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 operability of redundant safety-related systems was verified through reliance on station rules and practices and other administrative procedures.
The licensee's supplemental response dated August 7,1979 amplified that the Technical Specifications require that redundant systems be tested immediately and periodically thereafter when a system is made or found inoperable. When a system is to be made inoperable for maintenance, station procedures require
~
1744 017
. that the redundant systems be verified operable with an actual system test prior to removal of that systec.
This generally involves a visual inspection of the equipment being tested, '.owever, the control room operator doing the testing has enough information immediately available to him to determine system operability.
The licensee in its April 25, 1979 letter stated that its review had insured verification of the operability of safety-related systems following their return to service.
With respect to operator notification of equipment status, the licensee stated in its August 7, 1979 letter that shift turnover procedures for both operators and supervisors require full equipment status reviews.
The provisions of the procedures which assura this were described.
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 EI Bulletin 79-08, Item 8.
9.
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
Further, at that time an open continuous communicati.1 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 ene hmir of the time the reactor is not in a controlled or expected condition of operation and (2) procedures required or were to be modifed to require the establishment and maintenance of an open continuous communication channel with the NRC following such events.
In its response dated April 25, 1979, the licensee reported that it had reviewed its prompt reporting procedure for NRC notification.
The licensee stated that its procedure for "on call personnel" provides an acceptable 1744 018
, method of assuring that a r2sponsible person is available to establish continuous communication with the NRC, if required, within the one-hour time period.
In its supplemental response dated August 7,1979 the licensee clarified the circumstances under which notification would occur by stating that it will notify the NRC whenever it is determined that the reactor is not in a controlled or expected condition of operation.
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 remain 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 it had reviewed its operating modes and procedures that address controlling significant amounts of hydrogen.
By telephone conversation on November 6,1979, the licensee amplified that during normal operation, the reactor pressure vessel dome is filled with steam which flows to the turbine.
During reactor isolation, the dome is automatically vented through the safety-relief valves to the suppression pool.
In addition, the reactor vessel head has a vent line with a valve remotely operated from the contrc,i room.
However, this valve is not normally opened without first depressurizing the system.
In the event of significant hydrogen release to the primary containment, the nitrogen inerted containment maintains hydrogen below flammability concentration.
1744 019
. 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 the 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 July 5,1979 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.
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 Cooper Nuclear Station operations, and provide added asserance for the protection of the public health and safety during the operation of Cooper Nuclear Station.
References 1.
IE Bulletin 79-05, dated April 1, 1979.
2.
IE Bulletin 79-05A, dated Apail 5, 1979.
3.
IE Bulletin 79-08, dated April 14, 1979.
1744 020
~
_ 4.
NPPD letter, J. Pilant to K. Seyfrit, dated April 25, 1979.
5.
NPPD letter, J. Pilant to K. Seyfrit, dated July 5, 1979.
6.
NRC staff letter, T. Ippolito to J. Pilant, dated July 20, 1979.
7.
NPPD letter, J. Pilant to T. Ippolito, dated August 7, 1979.
1744 021