ML19316A922
| ML19316A922 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/05/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Berry G POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| References | |
| IEB-79-08, IEB-79-8, NUDOCS 8005280058 | |
| Download: ML19316A922 (23) | |
Text
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UNITED STATES
[
g NUCLEAR REGULATORY COMMIS$10N t
WASHINGTON, D. C. 20555
\\*****/
MAY 5 r3r Docket No.
50-333 Mr. George T. Berry General Manager and Chief Engineer Power Authority of the State of New York
Dear Mr. Berry:
SUBJECT:
NRC STAFF EVALUATION OF POWER AUTHORITY OF THE STATE OF NEW YORK RESPONSES TO IE BULLETIN 79-08 FOR JAMES A. ;FITZPATRICK NUCLEAR POWER PLANT We have completed our review of the information that you provided in your letters responding to IE Bulletin 79-08 for the James A. FitzPatrick Nuclear Power Plant. 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.
Specific requirements for your facility that result from this review and other TMI-2 investigations will be addressed to you ir, separate correspondence.
Sincerely, s
Thoma
. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing
Enclosure:
NRC Staff Evaluation
cc w/ enclosure:
See next page 8005280066 b
Mr. George T. Berry Power Authority of the State
.amANgy 5-<W of New York 2
cc:
Mr. Charles M. Pratt Assistant General Counsel Power Authority ~of the State of New York l
10 Columbus Circle New York, New York 10019 Mr. Peter W. Lyon Manager - Nuclear Operations Power Authority of the State of New York 10 -Columbus' Circle New York, New York 10019 Mr. J. D. Leonard, Jr.
Resident Manager
+-
James A. FitzPatrick Nuclear
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Power Plant P. O. Box 41 Lycoming, New York 13093 Director, Technical Development i
Programs State of New York Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Oswego County Office Building 46 E. Bridge Street Oswego, New York 13126 George M. Wilverding, Licensing Supervisor Power Authority of the State of New York 10 Columbus Circle New York, New York 10019 I
l
)
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EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-08 POWER AUTHORITY OF THE STATE OF NEW YORK' JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333
Introduction By letter dated April 14, 1979, we transmitted Office of Inspection and Enforcement (IE)Bulletin 79-08 to the Power Authority of the State of New York (PASNY 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 Nuclear Plant, Unit 2 (TMI-2) on March 28, 1979.
By letter dated April 25, 1979, PASNY provided responses to Action Items 1 through 10 of IE Bulletin 79-08 for the James A. FitzPatrick V -lear Power Plant (FitzPatrick).
PASNY supplemented this response by a letter dated May 18, 1979 to provide the response to Action Item 11 of IE Bulletin 79-08.
The NRC staff review of the PASNY responses ied to the issuance of requests foi.Jditional information regarding the PASNY 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 8, 1979, PASNY responded to the staff's requests for additional information.
The PASNY' 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 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 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 i
2 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 indicated that the pertinent FitzPatrick special procedures had been reviewed and that the description of the circumstances described in Enclosure 1 to IE Bulletin 79-05 would be reviewed by operational personnel and plant management; i.e., a review which covered Subsections 1.a and 1.b would be initiated.
By its letter dated August 8,1979, th. licensee indicated that the reviews specified in Action Item 1 of IE Bulletin 79-08 had been completed and that IE Region I inspectors had confirmed the completion.
We conclude that the licensee's scope of review, instructions to operating personnel and documented participation satisfies tbc. intent of IE Bulletin 79-06, 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 cepability, upon automatic initiation of safety injection.
3 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.
By its letter dated April 25, 1979, PASNY indicated that the requested review of the primary containment isolation design had been completed.
This review verified that 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 April 25, 1979 letter, the license indicated that the FitzPatrick Technical Specifications had been reviewed and that modification was necessary to reflect a needed change to the residual heat removal (RHR) / low pressure coolant injection (LPCI) system control logic.
To this end, PASNY submitted a Class II license amendment request which we have reviewed and approved as documented in license Amendment Number 48 dated January 23, 1980.
In addition to containment isolation valves which receive automatic closure signals, the licensee stated in its April 25, 1979 response that the non-automatic isolation valves had been reviewed and found to meet the criteria for piping isolation class designations.
In a suppental response dated August 8,1979, PASNY indicated that there are no procedures which require isolation action.
However, the plant special procedures require that operators follow the automatic isolation and, should a valve not operate, the valve should be operated manually.
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 l
4 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 tht (1) it described the automatic and manual actior.s 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.
By letter dated April 25, 1979 PASNY indicated that the reactor core isolation cooling (RCIC) system is the specific auxiliary cooling system in the FitzPatrick design. This system is powered by a reactor steam driven turbine.
The RCIC system turbine starts automatically on receipt of a reactor water level signal of -38 inches.
The following actions occur:
(1) Turbine steam supply opens, (2). Cooling water valve opens (to oil cooler),
(3) Pump outboard isolation opens, and (4) Steam line drains close.
The RCIC system turbine will remain on the line pumping until a level of
+58 inches in the reactor vessel is achieved.
At this point, it will trip and will not start again unless manually reset at the turbine.
The RCIC system turbine can be started manually by lining up as follows:
(1) Shut steam line drains, (2) Open pump outboard isolation valve,
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(3) Open cooling water to oil cooler valve, and (4) Open turbine steam supply valve.
The high pressure coolant injection (HPCI) system, although an emergency core cooling system instead of an auxiliary cooling system, would initiate on loss of feedwater.
The HPCI system starts on either a reactor water level of -38 inches or a high drywell pressure signal of 2.7 psig.
Upon receipt of either signal, the following events occur automatically:
(1) Main steam supply valves open, (2) Pump inboard discharge valve open, (3) Auxiliary oil pump starts, (4) Gland seal condenser blower starts, (5) Steam line drains close, and (6) Standby gas treatment starts.
As the oil pressure increases, the tubrine stop valves open and the turbine control valve begi.as to open.
The turbine accelerates and, when the discharge pressure of the turbine is greater than the reactor pressure, the pump injects water into the reactor vessel.
The pump will continue to inject until a
+58 inch reactor water level signal is received.
At this point, the turbine will trip, but it will restart again if a -38 inch reactor water level signal is received.
All this is accomplished without any operator action.
1 The HPCI system, aithough fully automatic, can be initiated manually whenever necessary provided that the reactor pressure is greater than 150 psig.
Manual initation is accomplished by:
l
6 (1) Opening main steam supply valves, (2) Opening pump inboerd discharge valve, (3) Starting gland seal condenser blower, (4) Starting auxiliary oil pump, (5) Closing steam line drains, and (6) Starting standby gas treatment system.
The special procedures / emergency operating procedures for loss of feedwater call for verifying automatic station response as part of the immediate operator action.
The automatic station response lists the HPCI and RCIC systems starting at -38 inches reactor vessel level.
In addition, under immediate action, the operator is directed to start the HPCI and RCIC systems if they have not started.
We conclude that the licensee's procedural summary of automatic / manual actions
- iacessary 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 infor-mation 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) opera-tors were instructed to utilize other available information to initiate safety
- systems, i
7 l
In its letter dated April 25, 1979, the licensee indicated that there are seven separate reactor vessel water level indicators and two recorders in the control room covering four ranges of level.
(1) Three of the level indicators and one recorder, monitor reactor level in the range of zero to +60 inches.
The recorder provides high and low level alarms.
(2) Two indicators monitor reactor level in the range of -150 to
+60 inches and are from lines using the temperature compensated reference columns.
(3) One indicator and one recorder monitor in the. range of
-100 to +200 inches inside the core shroud.
(4), One indicator monitors water level in the zero to +400 inches and uses a separate reference column of water.
There are' eighteen local vessel level indicating switches in the reactor building.
Twelve of these derive their level measurements from the temperature compensating columns.
Six derive their level measurements from non-compensating columns.
(1) Four instruments, initiate reactor scram and primary containment isolation (except main steam isolation valve isolation) at
+12.5 inches of water decreasing, and RCIC/HPCI system turbine trips at
+58.0 inches of water increasing.
(2) Two instruments initiate automatic depressurization system permissive at (+) 12.5 inches of water decreasing.
(3) Four instruments initiate recirculation pump trips, and Group 1 primary containment isolation at -38 inchas of water decreasing.
8 (4) Four inctruments initiate the RCIC/HPCI systems and the standby gas treatment system at -38 inches of water decreasing, and the core spray and LPCI (with high drywell pressure) systems and the emergency diesel generator and automatic depressurization permissive at
-146.5 inches of water decreasing.
(5) Two instruments initiate containment spray permissive at zero inches of water decreasing.
(6) Two indicating level transmitters.
There are a total of 27 instruments that are used in the above systems for reactor vessel level indication and safety system initiations.
(1) Fourteen are Yarway, Model 4419 primary liquid level indicators of the self-operating type (requires no electrical power).
The sensor is a diaphragm, differential pressure type, which senses the differential pressure developed between a static head reference and actual vessel level at the variable leg.
The diaphragm acts upon a permanent horseshoe magnet which, in turn, moves a spiral armature and indicating pointer.
(2) Three are Yarway, Model 4418 indicators (as described in above paragraph) with secondary indicators which are. electronic and receive their input signal from a linear voltage differential transformer whose slug (movable core) is connected by linkage to the primary indicator pointer.
NOTE:
Switches used in the Yarway primary liquid level indicators are magnetic and work off of magnets mounted on the spiral armature.
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(3) Four are GEMAC, Type 555 level transmitters of the power
- i operating type (requires electrical power).
The sensor is a i
diaphram, differential pressure type which senses the differential
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pressure developed between a static head reference and actual vessel I
level as the variable leg.
The diaphram acts upon a transducer which converts the mechanical motion to an electrical signal which-is proportional to vessel level. The signal output of the transmitter drives a GEMAC, Type 180 level indicator.
l (4) Four are Barton, Model 288 level indicating switches of the self-operating type. The sensor is a diaphram, differential pressure type which senses the differential pressure developed between a static head reference and actual vessel level',the variable leg.
(5) Two are GEMAC level indicators / recorders, Type 520 and 530 of the power operating type. The input signal is received from a GEMAC Type 555 level transmitter.
Instructions to operators regarding use of other available iniormation are approved in the responses to Action Items 1 and 5 of IE Bulletin 79-08.
1 In addition to the aforementioned, the licensee indicated in its letter dated August 8,1979 that, in view of the fact that there are seven separate reactor vessel water level indicators in.the control room for the
}
reactor operators which have been tested under accident conditions this multiplex of redundancy should be fully utilized by operators in making I
judgments regarding reactor water level under normal and abnormal conditions.
l There also exist drywell pressure indications, containment sump level indications, suppression pool temperature indications, suppression pool level i
indications and indications on main steam isolatiori valves, safety / relief valves, reactor water cleanup valves, high pressure coolant injection steam line valves, high' pressure coolant injection valves,- cmf low pressure coolant j
injection valves, all of which may very well be monitored by the operator to.
determine gross changes in reactor vessel inventory.
PASNY encourages
10 corroboration of indication.
Specifically, for small changes in the reactor coolant inventory, i.e., less than five gallons per minute, the containment sump indication systems are utilized to determine whether or not the criteria of the Technical Specifications on minor leakage are being met.
- However, none of these instrument systems can be used to adequately determine the vessel level; for this, the operator relies on the seven redundant level systems.
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 operating procedures and training instructions to ensure that:
a.
Operators do not override automatic actions of engineered 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 cf operators improperly overriding the automatic actions of engineered j
safety features, (2) it addressed providing operators with additional informa-tion and instructions to not rely upon vessel level indication alone for 1
manual actions and (3) that the review included openting procedures and training instructions.
By its letters dated April 25, 1979 and August 8, 1978, PASNY indicated that a review of the following procedures had been accomplished:
(1) standard procedures, (2) special procedures (emergency operating procedures),
(3) standing orders, (4) special orders, and (5) training instructions.
This j
review had been conducted with respect to the avoidance of operator override of engineered safety features.
Problems were not uncovered other than the
11 need to revise training programs which the licensee has indicated have been completed.
Operators will be requalified based on these revised training procedures.
PASNY indicated that shutdown of the systems or subsystems should not be performed until the need for the system or sub-system has passed.
Special procedures (emergency operating procedures) require the operator to verify the automatic initiation of(emergency core cooling systems or sub-systems and to operate the systems or sub-systems as required to maintain core coverage.
In addition, the licensee's standing orders state that:
" Believing your indicator means to investigate and corroborate alarms and instrument readings immediately.
All alarms and 'off-normal' instrument readings should be investigated thoroughly and immediately." By corro-boration, it is meant that other indications should be checked to verify conditions.
It should be assumed that an alarm or indicator is valid until investigation proves otherwise.
We conclude that the licensee's review of operating procedures and training instructions satisfies the intent of IE Bulletin 79-08, Item 5.
6.
Revire 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 proper 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 controls were in existence to maintain proper valve position during normal operation as well as during surveillance testing and maintenance.
12 By its letters datsi April 25, 1979 and August 8, 1979, PASNY indica'ted that a review of safety system valve lineups was accomplished in January and February 1979 as a result of a commitment to the NRC.
The lineups were reviewed for correctness and the manner in which they were being performed.
A valve lineup is required after an outage on any system that has been disturbed and on all safety systems.
Surveillance tests to declare safety systems operable are also performed.
Each surveillance test has a " return to normal" section with a signoff to ensure the system is put back in a standby mode af ter a test is performed.
In addition, the licensee indicated that positive administrative controls are implemented to assure that systems requiring retest are in fact retested prior to the need for their operability.
Retest is required on safety-related systems.
The only exception allowed is if plant conditions require a delay in the retest.
If plant conditions prevent immediate retest, a tag is placed on the system in the control room to insure the operator realizes that a retest is required before declaring the system operable.
In addition, the Technical Specificat. ion limitation on systems which are inoperable and require testing of the redundant systems is followed during these periods.
Documented position checks on the valves in safety-related systems are (and have been) made during valve lineups prior to startups and after maintenance actions are completed that required tagging out of the valves in an abnormal condition.
Valve lineup sheets have been reviewed by IE regional inspectors for accuracy and proper usage.
The valve positions and valve operators, if applicable, are operationally checked by the surveillance test required by the Technical Specifications to insure proper operability of these systems.
In addition to the normal surveillance performed for system operability, there exists a " locked valve surveillance" that is performed monthly which checks the position and locking device on certain critical safety system valves.
13 As a result of a site visit to review the licensee's above response, an immediate action letter was forwarded to PASNY on September 13, 1979.
The discrepancies noted therein were corrected by the licensee as documented in its response of September 26, 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 af 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 evaluated 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.
By its letters dated April 25, 1979 and August 8, 1979, PASNY indicated that i
gases are vented from the drywell through the drywell inerting, containment I
air dilution and purge system.
The four isolation valves on this system are isolated by the primary containment isolation system (PCIS).
These valves
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14 prevent venting of gas from the drywell or torus.
Manual override of the isolation of these valves is available to allow containment venting.
No automatic interlock is present to isolate these valves on a high radiation signal under any operating mode.
Procedure F-SP-2, " Post-LOCA Venting of Containment and Operation of the Main Steam Leakage Collection System,"
requires that the containment atmosphere be analyzed and that the activity be within acceptable limits before any venting can take place.
Surveillance tests are performed routinely to verify component operation and system operability of the PCIS.
. Liquids are normally transferred from the primary containment through the drywell floor drain sump, drywell equipment sump and the residual heat removal system discharge to radwaste for the purpose of lowering torus water level.
Valves in each of these systems are isolated by the PCIS.
The isolation valve outside containment in each of these systems must be opened manually at the valve switch provided that the isolating signal has cleared and the PCIS has been reset by operating the two major switches.
This allows the transfer of liquids out of the drywell or torus only by positive operator action.
There is no interlock to isolate these systems when high radiation exists.
For the reactor coolant sample lines from the reactor recirculation loops (3/8 inch stainless steel tubing), the valves isolate with the isolation signal and will not reopen when the isolation signal clears.
However, they will reopen by themselves when the master PCIS reset switches are operated and the isolatea signal has cleared.
Procedure SP-1, " Loss of Coolant," requires that in the event that leaks are indicated within the drywell, samples ec the sump effluent be taken to determine the source of the leak.
Surveillance tests are performed routinely to verify the component operation and system operability of the PCIS.
In addition, the licensee indicated that the procedures would be modified to prevent the inadvertent transfer of radioactive liquids through the reactor coolant sample line by September 1, 1979.
By a subsequent telephone
15 conversation, the licensee advised us that the procedure modifications had been completed.
We unclude that the licensee's review of systems designed to transfer cadioactive 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:
Verification, by test or inspection, of the operability of redundant a.
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 previous surveillance testing within Technical Spec.ification 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 previou's shift.
By its letter dated April 25, 1979 the licensee indicated that on April 6, 1979, prior to the issue of IE Bulletin 79-08, an extensive review was conductcd of maintenance, test and tagging procedures because of known events
l 16 i
j which had occurred at TMI-2.
The maintenance procedures which would be used i
for maintenance on inoperable equipment references the specific section of the i
i Technical Specifications for that particular equipment.
This has been the
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historical method used at FitzPatrick to insure that operability tests on not only redundant safety system trains, but also on additional safety-related i
systems, are performed prior to removal of the system in question from service.
Control of maintenance on safety-related systems requires that operability tests be completed on systems prior to placing those systems back in service.
PASNY has also written a standing order which requires that a tag be placed in j
the control room on a conspicuous operating mechanism for that system should an operability test not be completed as soon as the system has been declared l
out of a maintenance status by the maintenance forces.,The PASNY response j
also indicated that either a positive check is made thai. a system undergoing j
surveillance testing has been returned to normal, or (in the case of instru-i mentation procedures) that the procedures have a step-by-step return to normal line up within the body of the procedure itself.
l During the aforementioned review, it was also noted that Tagging i
Procedure'10.1.2 " Equipment and Personnel Protective Tagging," defines
" released" for the tag holder as indication of his permission to remove the i
tags and that the system or component may be operated, and to the controller, who is normally the nuclear control operator, that the tags have been removed and that the protective boundaries and/or controls are positioned in accordance with the special instruction section of the protective tagging j
record.
I l
Explicit notification of reactor operating personnel is accomplished by the following methods:
(1) The shift supervisor has to give written permission via a " work tracking form" and the nuclear control room operator fills out the
protective tagging record."
_ _,. ~.. _. _... _ - - _
p 4
8' 17 (2) The equipment being taken out of service or returned to service is logged in the shift supervisor!s log.
i (3) The control room has an equipment status board that lists the safety-related equipment, the surveillance tests that must be performed if the equipment is made or found inoperable and the frequency these i
surveillance tests must be performed.
The board is mounted on the wall and is easily visible.
The shift supervisor keeps the board up to date.
i (4) Surveillance Test F-ST-400, " Daily Surveillance Report," has a column with a signoff for limiting condition for operation action items.
If a safety-related component is out ' f service, it is o
listed on this surveillance report to ensure that the surveillance required on the redundant equipment is performed.
Over and above these administrative controls, maintenance on safety-related equipment and, as requested by the resident manager, on other important 7
equipment', is checked on a surveillance basis by the quality control group in order to verify that the administrative controls are being employed.
In addition the aforementioned, PASNY in its letter dated August 8, 1979 committed to make a visual check of system status to the extent practicable prior to removing redundant equipment from service and to concommittantly rely on the prior operability. verification within the current Technical-Specification surve'111ance interval.
Incoming. reactor operating personnel are required to review appropriate logs j
^
and status boards and discuss the status of systems, including safety-related.
systems.
This procedure is set forth in Operations Department Standing Order No. 4, " Shift Relief and Log Keeping."
Specifically, this standing order states:
"In addition to reading of the logs, the incoming operator shall be appraised of any testing,' unusual operations, etc., that are in progress."
~
"The verbal exchange of information between shifts must include the status of f
y e
eE<-
..e
.-..,-v-
18 all safety-related equipment and conditions as set forth in Technical Specifications."
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 of the time the reactor is not in a controlled or expected condition of operation.
Further, at that time an open continuous communication channel shall be established and maintained with NRC.
The actions specified in Item 9 of IE Bulletin 79-08 have been incorporated in the requirements of Section 50.72 of 10 CFR Part 50.
Since all licensees, including PASNY, are subject to these requirements, we conclude that'the licensee will meet 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.
I By its letter dated April 25, 1979, PASNY indicated that FitzPatrick utilizes a nitrogen inerted containment.
Procedure F-SP-2, " Post-LOCA Venting of Containment and Operation of the Main Steam Leakage Collection System," deals with venting of hydrogen gases from the containment.
This procedure has been reviewed by the licensee and has been determined to be adequate.
It provides for radiological analysis and evaluation prior to venting.
Hydrogen accumulation in the primary system at the top reactor vessel would be vented to (1) containment
19 via the head vent and (2) to the main condenser and the recombiner/off gas treatment systems.
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 through 10 of IE Bulletin 79-08.
By letter dated May 18, 1979, the licensee indicated that it currently had under development an amendment request which would correct a number of the instrumentation tables contained in the Technical Specifications and'that this submittal would include the discrepancies that were identified as part of its review of Bulletin 79-08.
This request was submitted by letter dated September. 13, 1979.
We have reviewed and approved the licensee's request as documented in license Amendment Number 48 dated January 23, 1980.
4 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 FitzPatrick operations, and provide added assurance for the protection of the public health and safety during the operation of the James A. FitzPatrick Nuclear Power Plant.
20 References 1.
IE Bulletin 79-05, dated April 1, 1979.
2.
IE Bulletin 79-05A, dated April 5, 1979.
3.
IE Bulletin 79-08, dated April 14, 1979.
4.
PASNY letter, J. D. Leonard to B. H. Grier, dated April 25, 1979.
5.
PASNY letter, P. J. Early to H. R. Denton, dated May 18, 1979.
6.
NRC letter, T. A. Ippolito to G. T. Berry, dated July 20, 1979.
7.
PASNY letter, J. D. Leonard to H. R. Denton, dated August 8, 19,79.
8.
NRC letter, B. Grier to J. D. Leonard, dated September 13, 1979.
9.
PASNY letter, J. Leonard to B. Grier, dated September 26, 1979.
10.
PASNY letter, P. J. Early to T. A. Ippolito, dated September 13, 1979.
11.
NRC letter, T. A. Ippolito to G. T. Berry, dated January 23, 1980.
.