IR 05000220/1989012
| ML17056A002 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/17/1989 |
| From: | Anderson C, Paolino R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17056A001 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-220-89-12, NUDOCS 8906020043 | |
| Download: ML17056A002 (38) | |
Text
U ~ S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-220/89-12 Docket No.
S0-220 License No.
DPR-63 Category C
Licensee:
Nia ara Mohawk Power Cor oration 301 Plainfield Road S racuse New York 13212 Facility Name:
Nine Mile Point Unit
Inspection At:
Salina Meadows Cor orate Office Inspection Conducted:
March 26-31 1989 Inspectors:
R. J.
aolino, Senior Reactor Engineer PSS/DRS/EB d
e Other Partici ants and Contributors to this Re ort:
Approved by:
date C. J. Anderson, Chief, Plant Systems Section, DRS/EB R.
K. Mathew, Reactor Engineer, PSS/EB C.
E. Sisco, Operations Engineer/Examiner, DRS/OB J.
C.
S ewa lectrical Engineer, NRR/SICB C. J.
nderson, Chief, Plant Systems Section, DRS/EB Ins ection Summar
Ins ection of March 26-31 1989 Ins ection Re ort No.
50-220/89-12 Areas Ins ected:
A special announced inspection was performed to determine the extent and safety significance of the deficiencies identified in the November 14-18, 1988 Regulatory Guide 1.97 inspection and the need for corrective actions prior to Nine Mile Unit 1 restart.
In addition, alternatives to the R.G.
1.97 monitoring instruments were reviewed for their usefullness in imple-menting the Emergency Operating Procedures (EOPs).
Results:
No violations were identified.
However, the NRC identified eight restart issues that need to be addressed prior'o restarting the Unit 1 facility.
A summary of these items is presented in the following table.
These items were transmitted to the licensee in an NRC letter dated April 21, 1989, 3906020043 3905i7 PDR ADOCK 05000220
Unresolved Items
~Para ra h
Item Number 1.
Evaluation of R.G.
1.97 cable separation deficiencies.
2.
Evaluation of the R.G.
1.97 isolation deficiencies.
4.2 89-12-01 89-12-05 3.
Review of the R.G.
1.97 instrument circuit loading and the adequacy of installed fuses.
4.5 89-12-03 4.
Review of alternatives to the Category
R.G.
1.97 instruments for which deficiencies exist.
7.3 89-12-04 5.
Completion of the failure modes and effects analysis for the APRM isolation deficiencies.
5.1 89-12-02 6.
Identification of R.G.
1.97 instrument power sources and providing instrument power source information at the site in a form useful to the control room operators.
7.
Evaluation of the safety significance of the Reactor Pressure Vessel (RPV)
common tap for the fuel zone water level instrument.
9.0 8.0 89-12-06 89-12-07 8.
Documenting and docketing Nine Nile 1 R.G.
1.97 10.0 restart activities.
89-12-08
Details 1.0 Persons Contacted 1. 1 Nia ara Mohawk Power Cor oration NMPC J.
Benson, Senior Engineer F. Constance, Electrical Engineer (IEC)
"D. Disc, Vice President Engineering/Quality Assurance (Retired)
R.
Eastham, Assistant Manager, MODS D. Goodney, Project Engineer
- G ~ Montgomery, Technical Staff (NMP2)
C. Terry, Vice President Nuclear Engineering/Quality Assurance
"S.
Wilczek, Jr.,
Manager Nuclear Technology G. Wilson, Attorney 1.2 NMPC Contract Personnel J. Betlack, MPR Associates T. Dixon, MPR Associates J.
Redmond, Compis J. Schilder, OEI M. Wetterhahn, Conner
& Wetterhahn Associates E. York, Compis 1,3 Public Service Commission P.
Eddy, Staff 1.4 U.S. Nuclear Re viator Commission R. Benedict, Senior Project Manager, NRR/HQ J. Johnson, Chief, Reactor Projects Section 2C, PB2/DRP S. Newberry, Chief, SIPC Branch, NRR/HQ
- Denotes personnel not present at exit meeting on March 31, 1989.
2.0
~Per ose The purpose of this inspection was to follow up on the results of the Regulatory Guide 1.97 inspection of November 14-18, 1988 for Unit 1 to determine the extent of specific R.G.
1.97 deviations and their impact on assuring acceptable emergency response capability during and following the course of an accident.
The specific R.G. deviations reviewed related primarily to separation (physical and electrical)
and isolation.
The need for R.G.
1.97 corrective actions prior to Nine Mile 1 restart was reviewed.
Alternatives to the R.G.
1.97 monitoring instruments were reviewed for their usefulness in implementing the Emergency Operations Procedure.0
~Back round The November 14-18, 1988 inspection (Report No. 220/88-34)
was performed to verify that instrumentation systems for assessing plant conditions during and following the course of an accident were installed in accordance with generic letter no. 82-33 "Requirements for Emergency Response Capability" (Supplement 1 to NUREG-0737).
The Generic Letter, issued on Oecember 17, 1982, specifies those requirements regarding emergency response capabilities that have been approved by the NRC for implementa-tion.
The supplement discusses the application of Regulatory Guide 1.97, Revision 2 (R.G.
1.97) to the emergency response facilities, including the control room, technical support center and the emergency response facility at nuclear power plants.
The results of the November 14-18, 1988 inspection identified numerous deviations from R.G. 1.97.
The deviations identified relate to the following areas:
~Redundanc A number of deviations regarding electrical and physical separation were identified with control room R.G.
1.97 instrumentation.
~
Interfaces - A number of circuits were identified which either lacked 1E/non-1E isolation devices or contained inadequate isolation devices.
~
Ta in 5 Markin R.G.
1.97 instrumentation in the control was not appropriately.
marked.
~
Ois la 5 Recordin
- Recording of instrumentation readout for one channel was not provided for certain instruments.
~
E ui ment uglification - Three items of electrical equipment were identified as not in compliance with 10 CFR 50.49 requirements for qualification in a harsh environment.
Following the November 14-18, 1988 inspection, meetings were held between the NRC and the licensee (Oecember 20, 1988 and January 21, 1989) to discuss the R.G.
1.97 deviations.
Ouring these meetings, licensee personnel stated that the referenced problem areas for the plant variables specified, satisfied the design criteria in effect for Nine Mile 1 when the plant was licensed.
The licensee discussed their plans to address deficiencies related to:
tagging and marking; display and recording; and equipment qualification.
The staff concluded that their plans to address these deficiencies appeared adequate.
However, the staff was concerned that the licensee's efforts to address the redundancy and interface issues prior to restart of Nine Mile Unit 1 were inadequat.0 R.G.
1.97 Problem Areas 4 '
~Sco e
This report provides the results of a NRC inspection performed at the Nine Mile Point Unit 1 facility during the period March 26-31, 1989.
The inspection represents the efforts of a team of five members review-ing licensee activities related to NRC findings identified in the R.G.
1.97 inspection (Report No. 220/88-34) of November 14-18, 1988.
This inspection focused on the availability of instrumentation given that deviations from R.G.
1.97 criteria exist in areas of separation, redundancy, independence, single failure and isolation.
Ouring the weeks preceding the inspection, the licensee conducted a review of the above areas for the RG 1.97 Category 1 variables.
The scope of their review was to identify deviations from the R.G.
1.97 criteria for the criteria specified above.
Significant progress was made in establishing the extent of the deviations from the R.G.
1.97 criteria.
However, the licensee's efforts to assess the availability of the Category 1 instruments, considering the noted deviations, was still in progress.
. The inspection findings relative to the criteria listed above are discussed below.
4.2
~Se aration Cable separation for redundant channels is accomplished by routing cables in separate channel trays or conduits.
This is a method by which separation of power, control and instrumentation cables can be maintained so that no single component failure will cause failure of redundant systems.
The licensee's procedure EOG-1300 "Cable Routing Criteria" was reviewed to determine the licensee'
requirements for cable routing.
Cables for safety related 2 channel systems are channelized by power supply using channelized trays and penetrations.
Cables for safety related 4 channel systems are routed in color coded trays as shown in Attachment I.
Type 4 channelization (as defined in Attachment I) is used for instrumentation loops.
The licensee completed a color coding cable routing sketch for key Category 1 Instrumentation loops except for the Neutron Monitoring System (APRM, IRM Part of SRM) and the Valve Position Monitoring System.
The team verified the accuracy of the cable routing sketch and the cable routing problem areas shown in Attachment II.
Most of the separation problems were found in the control complex.
However, a few problems were noted in the reactor buildin The following separation problems were identified:
1.
Source Ran e Monitorin Instrumentation SRM SRM redundant loops RGOl B and D are routed through common cable trays.
Since the licensee review of other loops is not complete, further deficiencies may be identified in the future.
2.
RPV Mater Level Hi h/Lo Lo/Lo Computer/Annunicator cables to redundant channels are routed through common trays 12 CAJ, 12 CAT and 12 CT.
3.
Containment Ox en Monitors Computer cables for redundant channels are routed through common trays
CX and
CU.
4.
Containment Area Hi h Radiation Computer cables for redundant channels are routed through common trays
CU and 12 CT.
5.
RCS Pressure Indicator circuits PI 36-31A/32A for redundant channels are routed through common trays
CAN and
CAH.
6.
Dr well Pressure Narrow Ran e
This variable is not powered from an independent 1E power source in accordance with R.G.
1.97 'owever, indicator circuits for redundant channels are routed through common trays
CAM and
CR.
In addition to the above, other separations deviations were noted.
In some cable trays, cables from two channels were mixed together and non-safety 'circuits were routed through channelized trays.
Discussions with the licensee revealed that the separation deviations in the cable routing for Category 1 instrumentation are still under evaluation.
The issue of cable separation is unresolved pending the licensee's evaluation of the impact of the R.G.
1.97 separation deviations on the availability of the RE G. 1.97 instruments and implementation of the licensee's plans to address the deviations (89-12-01).
4.3 Redundant Channels The following parameters have only a single channel.
.
'ide range reactor vessel water level 4 Torus pressure
'rywell atmosphere temperature
'rywell water level It was noted by the inspectors that drywell water level was recently identified by the licensee as a parameter important to the performance of their EOPs.
The licensee categorized this parameter and other parameters important to the EOPs as Category 1.
The licensee discussed with the inspection team their plans to install modifications at the next refueling outage to provide a
redundant channel for the following parameters:
torus pressure, drywell atmosphere temperature and drywell water level.
The licensee's plans for future modifications to address redundancy deficiencies is discussed further in Section 10.0.
During this outage, the licensee is providing a second transmitter for the wide range reactor vessel water level.
This modification is to meet the Safety Parameter Display System (SPOS)
design basis.'owever, this modification does not satisfy the R.G.
1.97 redundancy criteria.
The licensee needs to address this deviation from the R.G.
1.97 redundancy criteria.
Isolation The following parameters either have computer inputs, annunciator points or a redundant channel of the same parameter wired without isolation.
Neutron Flux APRM Wide Range Reactor Vessel water level Wide Range Orywell Pressure Torus Pressure Orywell Atmospheric Temperature
~ Torus water level
~ Containment Area High Rad Drywell water level
'CS pressure I
Failure of the computer inputs to the APRM could potentially disable the APRM trip function as well as the indication.
The licensee stated that a failure modes and effects analysis is underway for the APRM isolation deficiency.
This deficiency is discussed further in Section 5. 1 of this repor The above isolation deviations from the RG 1.97 criteria constitute an unresolved item pending the licensee's evaluation of the impact of the deviations on the availability of the RG 1.97 instruments and
'mplementation of their plans to address the deviations (89-12-05).
4.5 Power Su l
The inspector.
reviewed the licensee's calculations to determine the current and power used by each circuit on the Reactor Protection System Bus 11.
Six calculations (RPS ll-Fuse-1,2,5, 10 and 12) at random were chosen to verify the adequacy of the fuse sizing to supply RPS Bus 11 Power to the R.G.
1.97 instrumentation loops.
No discrepancies were found.
The inspector observed that no analysis was available to establish the adequacy of fuses to support the existing loads on the RPS Bus 12 instrumentation loops.
The inspector selected circuits 7 and 12 to review the loading, since these circuits feed the only channel available for monitoring the RPV water level (wide range), drywell'emperature and drywell water level.
Discussions with the licensee's staff revealed that the calculations for the above circuits are being performed and will be available for review at a later date.
This item is unresolved pending the licensee's completion of the calculations and any necessary corrective actions (89-12-03).
5.0 Evaluation of the R.G.
1.97 Cate or 1 Variables An evaluation of the Category 1 parameters against the acceptance criteria is provided below.
This section identifies specific deficiencies that were identified for each parameter.
These deficiencies are specific examples of the general problem areas previously identified in section 4.0 of this report.
5.1 Neutron Flux The APRM, IRM and SRM meet the Category 1 criteria for redundancy
.
and power supply independence.
The licensee identified one potentially significant problem, which the inspectors confirmed.
The APRM has numerous (60-70) computer points which are wired via a
dropping resi stor into the APRM circuit without an electrical isolator.
A postulated fault in the Non-1E computer could potentially disable the R.G.
1.97 instrumentation, cause an inadvertent scram or in the worst case prevent a scram from occurring when the scram was called for.
The licensee stated that a failure modes and effects analysis will be conducted.
This item is unresolved pending NRC review of the licensee's evalua-tion of this issue and the licensee's corrective action.
(89-12-02)
5.2 RPV Water Level The Hi/Lo, Lo-Lo (LT 36-03A/0) instruments are redundant from the sensors to the display module.
These instruments are powered from
~
t ~
~
I the redundant (RPS 11 and RPS 12) power supplies.
The Non-1E annunciator and computer points are electrically isolated by multiple relays which are acceptable for this purposely 5,3 The wide range (LT 36-33)
RPV water level is not redundant.
Other deficiencies were observed for this instrument circuit.
Several components in this circuit are powered by RPS 11.
Other components in this circuit are powered by RPS 12.
A loss of either power supply would cause a loss of this instrument circuit.
Another deficiency with this instrument circuit is the inadequate electrical isolation.
The non 1E computer is connected directly to the 1E sensor portion of the circuit through a dropping resistor.
The fuel zone water level has acceptable electrical redundancy and isolation.
The annunciator inputs are isolated with an alarm relay, which is acceptable.
Dr well Pressure The narrow range drywell pressure (PT 201.2-105/106)
has redundant sensors and display.
Currently the drawings indicate that both channels are powered via RPS 12.
The licensee issued Oesign Change Request LG021 would make PT 201.2-105 powered from RPS 11.
Field verification of the actual installation was still needed at the time of the inspection.
The wide range drywell pressure has redundant sensors, power supplies and displays.
There is no isolation between the computer and this circuit.
This variable is discussed further in Section 6. 1..
5.4 RCS Pressure The RCS pressure has redundant sensors (PT 36-31, 36-32),
power supplies and display.
The computer inputs are isolated.
The redundant RCS pressure channels are interconnected through a
common selector switch in control console E.
A single fai lure of thi s switch could lead to the loss of the redundant pressure channel.
This system is interconnected with the Wide Range RPV water level.
These interconnections are such that a failure in either the RCS pressure channel or the RPV water level could disabl,e both instrument loops.
There is no apparent isolation between the RCS pressure circuit and the non-lE feedwater system.
.This is an unresolved item from the previous inspection.
(88-34-01).
5.5 Torus Pressure The Torus pressure is a single sensor (PT 201.2-07)
insirument loop.
The computer input from this circuit is not isolated.
The annunciator is isolated.
The licensee has stated that this parameter is not required for the Nine Mile 1 Emergency Operating Procedures, since drywell pressure monitoring instruments are available.
5.6 Dr well Atmosh eric Tem erature This system provides three sensors at three elevations.
This system is provided power by RPS 12 only and, therefore, is not a redundant system.
The computer points are wired to the transmitters and are not isolated.
This is an unresolved item from the previous inspection.
(88-34-02)
5.7 Torus Water Level The Torus water level (LT 58-05/58-06)
has redundant sensors, power and display.
The computer input is wired directly to the transmitter and meter without isolation.
The annunciator input is isolated.
5.8 Torus Water Tem erature The Torus water temperature has redundant systems and is isolated from the annunciators via coil to contact relays.
This variable is discussed further in Section 6.2.
5.9 Containment Area Hi h Radiation The containm'ent area Hi Radiation parameter has redundant sensors and displays.
The recorder is powered by a Non-lE supply.
The licensee indicated that they are reviewing the need for an upgrade to the, recorder power supply to 1E.
The computer inputs are, wired directly to the General Atomics Radiation Mon,itor which has not been shown to be an acceptable isolator.
This variable is further discussed in Section '6.3.
5.10 Containment H dro en and Ox en Concentration The containment hydrogen and oxygen instrumentation loops are redundant and isolated.
5.11 Dr well Water Level The drywell level instrument is not redundant.
The computer input is not isolated.
The drywell water level is obtained by a pressure transmitter (PT 201.2-14) which is compensated by a differential pressure transmitter (PT 201.2-13) which measures the pressure of the
drywell for 0-4 psig.
For pressures above 4 psig or temperatures above room temperature; the level is not compensated and is, therefore, of limited accuracy.
The inspector requested that the licensee evaluate the accuracy of this instrumentation to assure that it meets the operators needs for post accident monitoring in accordance with R.G. 1.97:
5. 12 Dr well Sum Level and Dr well Drain Level The inspector noted the apparent removal of these parameters from the Category 1,
EOP list.
The licensee was requested to review their bases for deletion of these parameters.
6.0 New Modifications A review of new modifications involving R.G.
1.97 instrumentation for important parameters was performed to assess the extent to which this instrumentation is subject to. separation and isolation deviations.
The modification packages reviewed are discussed below.
6.1 Modification No. N1-80-37 Dr well Pressure/Torus Water Level)
Instrumentation was added to monitor the drywell pressure (wide rage)
and Torus water level.
The instrumentation is safety related and meets the design and qualification provisions of R.G.
1.97 with the exception of the isolation of non-safety circuits, physical separation of drywell pressure indicators in control room panel L and cable separation for Torus water level.
This instrument only provides a
monitoring function.
Signals from this instrumentatjon are not used to activate any of the safety systems'.2 Modification No. Nl-80-74 Torus Water Tem erature)
Additional temperature sensors were added to quickly and accurately alert the operators to changes in the torus water bulk temperature.
The instrumentation has two redundant channels with 12 dual water temperature sensors per channel.
The temperature sensors, loop components, indicators and recorder circuits meet the separation, isolation and redundancy criteria of R.G.
1.97.
6.3 Modification No. 80-12'ontainment Radiation Monitorin )
This modification consists of an installation of two radiation detection and indication units with recording capability for increased range monitoring.
Both of the monitoring systems are physically separated and powered by redundant class 1E power supplies except for the recorder.
The two pen recorder used for this para-meter is powered from a non-safety power circuit.
The computer
points for both channels are routed through common trays and are not separated or isolated.
Documentation of the design criteria utilized for the above instrumenta-tion indicates that redundant power supply, separation or isolation criteria may have been applied in some cases but these criteria were not necessarily applied in other cases.
No consistent design practice was evident for the separation/isolation criteria for the instrumentation modifications.
The specific deficiencies noted above are examples of the general problem areas previously identified in section 4.0 of this report.
7.0 Alternative Post Accident Monitorin Information 7. 1 A
endix R.
Redundant Remote Shutdown Panels A review was performed of information available at the redundant safe shutdown panels to determine the potential usefulness of this informatioq as an alternative to the primary R.G.
1.97 instruments in the control room.
The review also considered the availability of the information to the operators and the design criterion for these instruments.
'wo remote redundant shutdown panels are located outside-of the control room, each of which is capable of independently controlling a hot shutdown of the unit in the event that the control room is inaccessible.
The monitoring instruments provided at each of the remote panels are redundant to and independent of the indicators that traverse the control complex.
The location of the panels is such that they are readily accessible.
One of the panels is located on the same elevation as the control room.
The other panel is located two floors below the control room.
Both panels can be accessed within a few minutes from the control room.
Several methods of communication can be used between the control room and the remote panels including the Gatronix system and portable phones.
The following monitoring instruments are provided at each remote
'shutdown panel:
Reactor Temperature Orywell Temperature
"
Torus Temperature
'mergency Condenser Water Level Reactor Pressure
"
Drywell Pressure
Reactor Coolant Level In addition, each panel has an "All Rods In" indicating light along with a Rotor-Generator trip switch to verify that the scram function has been accomplished.
Those instruments above marked with an asterisk (") provide information similar to the important R.G.
1.97 instruments in the control roo The primary considerations in the review of the above instruments as an alternate to the R.G; 1.97 instruments were separation, isolation and instrument range.
It was determined that the instrumentation at the remote panels does not have electrical and physical separation deficiencies or electrical isolation deficiencies.
The instrument ranges are the same for the Reactor Vessel Pressure, the Drywell Atmosphere Temperature, the Torus Mater Temperature and the Reactor Pressure Vessel (RPV) Water Level at the remote panels as for the corresponding R.G.
1.97 instruments in the control room.
For the RPV Mater Level instruments at the remote panels, the R.G.
1.97 instrument, Hi/Lo, Lo-Lo 0-100 inch instrument range is covered.
However, the Wide Range and the Fuel Zone range for the RPV water level are not covered at the remote panels.
It also was noted that the Drywell Atmosphere Temperature monitors at the remote panels reflect the temperature for a single elevation of the drywell.
Whereas, the R.G.
1.97 Drywell Atmosphere Temperature monitor provides the temperature at three different elevations of the drywell. It was determined that the instruments at the remote panels, were generally installed to acceptable qualification specifications.
7.2 Other Post Accident Monitorin Information Available In addition to monitoring information available at the remote shutdown panels, other potential alternative sources of post accident monitoring information were identified by the licensee.
One of these alternatives is the instruments in the East and West instrument rooms.
The instrument rooms are located in the Reactor Building.
The instrumentation available in these areas is not subject to the separation and isolation deficiencies that have been encountered in the R.G.
1.97 instruments in the control room.
The licensee identified several other sources of alternative post accident monitoring information.
However, the licensee's review of this information was still in progress.
7.3 Conclusion A number of alternatives to the primary R.G.
1.97 monitoring instruments exist for providing important post accident monitoring information.
The licensee's evaluation of these alternatives was still.in progress at the end of the inspection.
The licensee's completion of their study to identify alternatives to the R.G.
1.97 monitoring instruments; their evaluation of the usefulness of these alternatives for implementing the EOPs; their providing operator guidance and training as to when and how the alternative instruments would be used are collectively an unresolved item (89-12-04).
8.0 Sin le Ta Issue The NRC inspection. team noted that the Fuel Zone Level instrumentation system is connected at a
common point.
This point is at the lower vessel tap which provides above core plate pressure at the reference leg for the fuel zone level indicating system.
The instrument tap also provides a
reference pressure to each Core Spray header that provides an alarm if the header is not intact outside of the core shroud.
The team reviewed the Fuel Zone Mater Level modification package, major
.
order ¹1843, modification package
¹Nl-80-038 and accompanying Safety Evaluation Report dated April 4, 1980.
From the documentation reviewed, the team concluded a postulated failure of the common reference leg in the fuel zone level indicating system was not considered in the design, construction, or operation of the system.
A failure of the common reference leg would render all fuel zone level indications.inaccurate, as well as possible annunciation that each core spray header is not intact outside of the core shroud.
A critical step of the Emergency Operating Procedures directs in part operator actions based on a correct determination of actual Reactor Water Level.
The team expressed the concern that if this indicated reactor water level is incorrect, the operator may take action( s) that are inappropriate to mitigate the consequences of an accident.
Tge inap-propriate operator action(s),
combined with the indications of a loss of Core Spray header integrity would require operator action(s) that are not necessary and may be inappropriate.
Adequate core cooling is assured at the facility by:
9.0
~
Core submergence
~
~
Steam Cooling A failure of the common reference leg of the fuel zone level indicating system would provide inaccurate indication of fuel zone level.
Adequate core cooling could not be verified by core submergence.
The failure could provide erroneous indication of core spray header integrity.
Adequate core cooling may not be assured by core spray.
This item is unresolved pending NRC review of the licensee's evaluation and resulting corrective actions.
(89-12-07)
I Confi uration Control The NRC team made an onsite inspection of the R.G.
1.97 instrumentation systems.
From this inspection, the team concluded the operating staff, operations management, Instrument and Control groups are not provided with certain drawings of the instrumentation systems that would enable the
various station groups to operate, trouble shoot and repair the instrumen-tation system.
The team inspected each electrical distribution panel and
~ found that the panels do not contain any information indicating the power supply to instrumentation systems.
This item is unresolved pending NRC review of the licensee's corrective action.
(89-12-06)
The NRC team inspected alternate instrumentation outside the control room.
From discussions held with the plant staff, the team concluded the alternate instrumentation system is not prestaged to be used in the event the R.G.
1.97 control room instrumentation becomes unavailable:
In addition, the team concluded that the plant staff is not trained in the use of the alternate instrumentation systems.
See Section 7.0 for a further discussion of the team's evaluation of alternative post accident monitoring instrumentation.
10.0 Conclusion At the conclusion of the inspection, the licensee discussed their hazards analyses relating to the deviations from R.G.
1.97 Category 1 parameters.
For each of the Category 1 parameters they-noted:
the deviations from R.G.
1.97 criteria; the potential impact of these deviations on the availability of the'.G.
~ 97 instruments and the importance of these monitoring instruments for performing EOP actions.
In addition, the licensee discussed alternatives to the R.G.
1.97 instruments that would be available to support EOP actions.
The inspectors requested that the licensee document their R.G.
1.97 hazards analyses.
In addition, the licensee was requested to document the R.G.
1.97 studies they hag recently performed to support this inspection (89-12)
and to document their current plans for making modifications to address devia-tions from R.G.
1 ~ 97.
This item is unresolved pending the licensee completing documentation of these items and submittal of these documents to the NRC (89-12-08).
The licensee was requested to give particular emphasis on discussions of their plans for modifications that address a current lack of redundancy for important parameters (torus pressures drywell atmosphere temperature, and drywell water level).
11.0 Following the inspection on April 21, 1989, the NRC sent a letter to the licensee identifying RG 1.97 restart issues for Nine Mile Point Unit 1.
Unresolved Items I
Unresolved items are matters which require additional information to determine whether they are acceptable items or violations.
Unresolved items are discussed in details, paragraphs 4.2, 4.4, 4.5, 5. 1, 7.3, 8.0, 9.0 and 10:0.
12.0 The inspectors met with licensee corporate personnel and licensee contract personnel at the conclusion of the inspection on March 31, 1989 at the corporate office, The inspectors summarized the scope of the inspection and the inspection findings at that tim NINE MILE 1 CABLE ROUTING CHANNELIZATION CRITERIA (from procedure EDG-1300)
ATTACHMENT I Channel izati on T
e Color Code Cross. Reference 11/1 12/2 11/2 12/1 Type
System or Loop/
Green Power Supply Yellow Blue Red Type
Power Supply/
System or Loop Green Yellow Red Blue Type
System or Loop/
Green Primary or Secondary Function Red Blue Yellow Type
~Power Supply/
Primary or Secondary Function Green Blue Red Yel low Non'hannelized Tray Computer Tray 11/1 11/2 12/1 12/2 White Orange Channel Channel Channel Channel 11, Subchannel
11, Subchannel
12, Subchannel
12, Subchannel
PARAMETER Drywell Water Level ( PT201. 2-13)
Containment Hydrogen Concentration (201.2-217)
Containment Hydrogen Concentration (201"2"330)
RPV Water Level Hi/Lo Lo/Lo (LT 36"03D)
(LT 36-03A)
NINE MILE UNIT 1 CATEGORY
PARAMETER CABLE ROUTING PROBLEMS ATTACHMENT II POWER SUPPLY RPS12 RPS12 RPS12 RPS12 RPS11 CABLE ROUTING PROBLEM AREA (TRAYS/PENETRATIONS/
CONDUITS/BARRIERS)
(From Reactor Building to Aux Control Room)
261N4B15, 13RC, 281-N4-A14, llCA1, 11TV, 11TK1, WL-S-L-A10, 11TJ, 1 lAB, 11TF, 1 lAE, 12CQ (RSP 11 Channel)
(From Turbine Buil'ding to Control Complex)
F.S. S. L 136, I131, W. L.S. L. C20, A10, 12CBQ1 (Non Channelized)
12TH (RPS 11, Contains Computer Cables)
( From Turbine Building to Control Room)
W.L.SL C33, C11, FS. SL 119 74, 41, 12CAS, 126 (Non Channelized)
11TJ (Partly RPSll)
(East Wall Reactor Building to Auxilary Control Room)
12BA4, 12CAB, 12TF, 12CT (RPS11)
12CAJ, 12CAT (RPSI)
Trays for'omputer/
Annunciator Input Same as above for Computer/Annunciator Input
Attachment II.
Wide Range (LT 36"33)
RCS Pressure Transmitter (PT36-31)
(PT36"32)
Orywel1 Atmospheric Temperature TE-201-36A TE-201" SOA TE-201-51A Orywell Pressure (Narrow Range)
PT-201. 2-105 PT201. 2-106 RPS11/12 RPS11 RPS12 RPS12 RPS12 RPS12 (Reactor Building El.
281 to Control Complex)
13RC, 281-N4-B15, 261-N4-N15, 11CA1, 11TV, llTK1, WLSLA10, 11AB (RPS11)
12CF, 12CR, 12CAL, 12CS, 12CX (RPS12 and Non Channelized for Computer Signal)
12CAM (RPS12)
12CL, 12CF (Non Channelized)
(Control Complex)
12CF, 12CBM2, 12CS, 12CX (Non Channelized)
12CAM/RCR, 11AA (RPS12)
(From Reactor Building to Control Complex)
13RN, 281-L12-A7, 120A3, WL.SLC35, 12CAB, 12CAK, 11AB, 12CP, 12CAH (RPS11)
12CF, 12CBM2 (Non Channelized)
WLSLA16, llTAF1, 11TK, 12CAZ (Computer Tray)
llAO (Computer Cable in RPS12)
11TF (RPS11)
11AF (Non Channel)
(From Reactor Building to Control Complex)
13RC, 281N4A14, 261N4B15, 11CA, 11TV, 11TK1, WLSLA10, 11TJ, 11AB, 11TF, 11AE (RPS11)
12CQ, 12B4, 12TF, WLSLC35, 12AB, 12CL (RPS11)
12CF (Non Channelized)
t>
~
~
'
V ( I r
C 1~
Attachment II Torus Water Level, (58-05)
Torus Pressure ( PT201. 2-07)
Neutron Flux (SRM)
(RG01B)
(RG010)
RPS12 RPS12 RPS11 RPS12 (From East Wall Reactor Building to Control Complex)
llTAG, WLSLA14, llRJ (Non Channelized)
11GE6 (RPS11)
12CAM, 12CAL (Computer Cables Routed through RPS12)
(From East Wall of Reactor Building to Control Complex)
1lRJ, 11TAG (Non Channelized)
11GJ3, 11AE, 11TF, 12CO (RPS11)
11AO (RPS12, Comuter Cable Routed)
11RK, 11TH, 11AWLSLC17, 12CAK, 12CAL, 12CAI, 12CBM1 (RPS12)
Same as Above (Redundant Circuits are not separated)