ML20100A628
| ML20100A628 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/30/1984 |
| From: | Corbin McNeil POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, JPN-84-77, NUDOCS 8412040019 | |
| Download: ML20100A628 (42) | |
Text
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123 Mdn Street White PIsins. New Wak 10001 914 681.6200
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& Authority November 30,1984 JPN-84-77 Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Domenic B.
Vassallo Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A.
FitzPatrick Nuclear Power Plant Docket No. 50-333 Supplement No. I to NUREG-0737 (Generic Letter 82-33)
Regulatory Guide 1.97, Revision 2 Implementation Report
References:
1.
NRC Generic Letter No. 82-33 to all Licennees of Operating Reactors dated December 17, 1982 (includes Supplement No. 1 to NUREG-0737).
2.
PASNY letter, J.P.
Bayne to D.B.
Vassallc, dated April 15, 1983 (JPN-83-33) regarding the same subject.
3.
NRC letter, D.B.
Vassallo to J.P.
Bayne, dated June 12, 1984 regarding Order confirming licensee commitments on emergency response capability.
Dear Sira Generic letter No. 82-33 (Reference 1) transmitted to the Authority Supplement No. 1 to NUREG-0737 which replaced the corresponding requirements for five NUREG-0737 items.
Via Reference 2,
the Authority submitted a schedule for completing each of the basic requirements and a description of our plans for the phased implementation and integration of these emergency response activities.
The NRC subsequently confirmed these commitments with an Order dated June 12, 1984 (Reference 3).
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In accordance with the Authority's commitments, attached is a document entitled " Response to Regulatory Guide 1.97, Revision 2 for James A.
FitzPatrick Nuclear Power Plant" (Attachment No.
1).
As outlined in Section 6.2 of Supplement No. 1 to NUREG-0737,.this report describes how the Authority plans to implement the guidance of Regulatory Guide 1.97 Rev.
2.
A table (included in Section 6 of the attached report) summarizes, for each of five variable types; instrument range, environmental qualification, seismic qualification, quality assurance classification, redundance, power supply, and display location.
Devirtions from the Regulatory Guide and the supporting justification or alternatives are presented in Sections 3.
Section 4 describes those modifications that will be implemented at FitzPatrick.
Schedules for implementation are included as Attachment No.
2.
These schedules (with the exception of the dates included in this letter) are for information only and should not be construed as formal Authority commitments.
The Authority has scheduled the implementation of Regulatory Guide 1.97 (Revision 2) type A variables for thirty days after the end of the 1986 (Reload 7/ Cycle 8) refueling outage or December 31, 1986, whichever is later.
Remaining variables (types B,
C, D and E) will be implemented thirty days after the end of the 1987 (Reload 8/ Cycle 9) refueling outage or December 31, 1987, whichever is later.
Specifically excluded from these schedules is the installation of unique identification labels for existing instruments that meet Regulatory Guide 1.97 (Revision 2) critorial these labels will be installed as part of the FitzPatrick Detailed Control Room Design Review (DCRDR) program.
This schedule is necessary because of the long lead time required for some new or replacement equipment and the numerous modifications (Analog Trip Transmitter System, SPDS, new Emergency Operation Facility, torus program, ADS check valve replacement, ADS piping analysis, fire protection, etc.)
currently in progress.
By providing a schedule linked to refueling outages, the Authority has considered the need for acheduled refueling outages to complete equipment installation.
By linking the schedule to a fixed date, the Authority has also considered the possibility that anticipated mid-cycle outages might be cancelled resulting in a refueling outage taking place sooner than anticipated.
If this were to occur, and the schedule were based solely on the refueling outage, unavoidable schedule extensions may be necessary.
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. If you require any additional information, please contact Mr.
J.A.
Gray, Jr.-my staff.
Very truly yours,.
y C.
A.
McNeill, J r.V Senior Vice President Nuclear Generation cc Office of the Resident Inspector U.
S.
Nuclear Regulatory Commission P.O.
Box 136-Lycoming, New York 13093 i
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ATTACHMENT NO. 1 to JPN NEW YORK POWER AUTHORITY JAMES A.
FIT 2 PATRICK NUCLEAR POWER PLANT Regulatory Guide 1.97, Revision 2,
Implementation 4
at the James A.
FitzPatrick Nuclear Power Plant 4
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TABLE OF CONTENTS 1.0 Introduction 2.0 Evaluation Method 3.0 JAF Deviations and Justifications 4.0 Proposed Additions and Changes to JAF Post Accident Monitoring Instrumentation-5.0 Summary 6.0 Position Summary Tables 7.0 References S
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1.0 Introduction This report provides an assessment of the extent to which the New York Power Authority's James A.
Fitzpatrick (JAF) Nuclear Power Plant meets the recommendations of Revision 2 to Regulatory Guide 1.97 (Reference 2).
The information provided herein meets the requirements of Supplement.1 to NUREG-0737, Section 6.0 (Reference 2), with the exception that sensor locations are not identified in the report.
Section 2.0 of this report describes the evaluation methods used in preparing this assessment.
Section 3.0 lists by variable type, specific exceptions to Regulatory Guide 1.97 Revision 2 recommendations, and justification as to why certain exceptions are acceptable.
Section 4.0 describes proposed modifications to increase the extent of JAF's conformance to the Regulatory Guide.
Section 5.0 summarizes the results of this assessment.
Section 6.0 (Position Summary Tables) summarizes the design criteria of Regulatory Guide 1.97 and compares JAF to Regulatory Guide 1.97 Rev.
2.
In most cases, JAF meets the recommendations of the guide.
Where exceptions to the guide are considered acceptable, technical justification is provided based upon plant specific designs, operation, or alternatives recommended by industry groups.
This is discussed further in Section 3.0.
2.0 Evaluation Method and Criteria 2.1 Regulatory Guide 1.97 defines five types of variables (designated A through E) to be monitored following a postulated accident at a nuclear power plant.
Type A variables are plant. specific and were determined by reviewing I
existing FitzPatrick Emergency Operating Procedures, BWROG Generic Emergency Procedure Guidelines, and draft FitzPatrick i
Emergency Operating Procedures.
Variable types B,
C, D and E l
are generic and listed in Regulatory Guide 1.97 Table No.
1.
Plant documantation was reviewed to determine and evaluate the instruments which measure each Regulatory Guide j
variable.
These documents included system flow diagrams; instrument loop drawings; wiring diagrams; plant modification l
packages; FSAR and Technical Specificationar purchase specifications; and, documents relating to radiation or environs monitoring.
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2.2 Variables listed in Regulatory Guide 1.97 Table No. 1 under multiple " type" headings have been assigned more than one-design category, based upon their specific function.
Instrument conformance to the Regulatory Guide was based upon the most stringent design category.
2.3 Reference documents and information used in this evaluation are decribed in greater detail in Sections 2.4, 2.5 and~2.6.
The.renults of this evaluation are. summarized in the. Position Summary Tables, Section 6.0.
2.4 In"May of 1983, the Authority submitted a document entitled,
" Response to 10 CPR 50.49 (Reference 13)."
The purpose of submittal was to specifically meet the requirements stated in subsection (g) of 10 CFR 50.49.
The submittal identified safety-related electrical equipment which is located in a harsh environment and is. required to mitigate or monitor an accident.
It identified equipment not yet qualified.
Such equipment was to be either qualified for interim operation, i
relocated to a less demanding environment, or replaced.
This submittal takes credit for commitments to qualify equipment by the January 1985 refueling outage.
In general, instrumentation designated as Category 1 and 2 and located in i
a mild environment as a minimum meets the environmental qualification criteria of IEEE-323-1971, "IEEE Standard for Qualifying Class lE Equipment.for Nuclear Generating S t e.t i o n s ", (Reference 8).
This was-the plant operating
(
' license design basis for safety related, QA Category I equipment.
Category 1a2 instruments that have been installed as part of recent modification packages were qualified to NUREG-0588 (Reference 10) and IEEE-323-1974, "IEEE Standard for Qualifying Class lE Equipment for Nuclear f
Generating Stations", (Reference 9).
2.5 Instrumentation installed under the original plant licensing requirements was specified to meet the seismic qualification requirements of IEEE-344-1971, " Seismic Qualification of Class lE Equipment for Nuclear Generating Stations,".
Category 1 and 2 instrumentation which has been installed recently as part of plant modification packages was qualified to NUREG-0588 (Reference 10) and IEEE-344-1975,'" Seismic Qualification of Class lE Equipment for Nuclear Generating Stations," (Reference 12).
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2.6 Power sources for post accident instrumentation will meet, or exceed, the requirements of Regulatory Guide 1.97, Revision 2.
2.6.1 Category I instrumentation is powered from one of the following sources:
- Class 1E 120VAC Emergency Euce
- Class 1E 120VAC Safeguard Buses
- Class lE 120VAC Reactor Protection System Buses e
Note:
The above power sources are part of the onsite Class lE power distribution systems which are backed by the Emergency Diesel Generators.
The Class 1E power systems meet the requirements of Regulatory Guide 1.32, Rev. 2 " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants" (Reference 14).
2.6.2 Category 2 instrumentation is powered from one of the following sources:
- Non-safety 120VAC Normal Buses
- Non-safety 120VAC Common Bus
- Class 1E 120VAC Emergency Buses Class lE 120VAC Safeguard Buses
- Class lE 120VAC Reactor Protection System Buses 120VAC Uninterruptible Power Supply (UPS) Bus The non-safety 120VAC Normal Buses (which are part of the normal AC power distribution system) supply the Turbine Building and Radwaste Building Exhaust Monitors (Noble Gases).
The non-safety 120VAC Common Bus (which is also part of the normal AC power distribution system, has its feeder automatically transferred between two non-safety notor control centers (MCC) when one feeder source is lost.
This common bus supplies the Reactor Building Area Radiation Monitors and the RHR Heat Exchanger Outlet Temperature Monitor.
4
The Class lE 120VAC Emergency Buses supply the Refueling Floor Exhaust Monitors (Noble Gases).
The Class 1E 120VAC Safeguard Buses supply the RHR flow, and Core Spray flow instrumentation.
The Class lE 120VAC Reactor Protection System Buses supply the Reactor Building (Lower Level) Vent Monitor (Noble Gases).
The 120VAC Uninterruptibic Power Supply (UPS) Bus receives its power from a dual motor UPS motor generator (M-G) set.
The AC motor is powered from a Division "B"
Class lE MCC. Upon loss of AC power, feed to the UPS M-G set is from the DC motor which receivec its supply from the Division "A"
125VDC battery.
The UPS Bus supplies the Stack Gas Monitors (Noble Gases), Hi-Range Turbine Building and Hi-Range Radwaste Building Exhaust Monitors (Noble Gases), SRV Position instrumentation, RCIC flow instrumentation, HECI flow instrumentation, and SLCS level instrumentation.
2.6.3 Category 3 instrumentation is powered from one of the following sources:
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- Non-safety 120VAC Normal Buses
- Class lE 120VAC Emergency Bus B3 120VAC Uninterruptible Power Supply Bus The majority of Category 3 instrumentttion is powered by non-safety 120VAC Normal Buses.
The Class lE 120VAC Emergency Bus B3 serves as the emergency source of Sample Panel 278SC-PNLl.
i The 120VAC UPS supplies the Main Feedwater flow instrumentation.
i 2.6.4 A detailed discussion of the 120VAC power distribution systems can be found in FSAR Chapter 8.
3.0 JAF Exceptions and Justifications 3.1 In this section, a number of variables are identified as not fully meeting the requirements of Regulatory Guide 1.97, Rev.
2.
In each of these cases, the Power Authority believes that these exceptions are justified.
In general, justification is based on one or more of the following factors:
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- The instrumentation used to monitor the variables
-provides the required information but not in the exact form recommended by Regulatory Guide 1.97, (e.g.,
several instruments with ove-lapping ranges fullfill the function of a single wide range instrument).
Other qualified indications are available to provide the required information (as discussed in Section 3.2).
4
- Industry groups such as the BWR Owners Group (BWROG) and NUTAC have evaluated the requirements of the Regulatory Guide and have concluded that there are acceptable alternatives to the requirements.
In cases where the Power _ Authority agrees with these it.dustry positions, they have been referenced in the detailed discussions that follow.
- The requirements of the Regulatory Guide do not accurately reflect the importance of certain instrumentation for the FitzPatrick plant.
- The information provided by the instrumentation is not required.
(e.g.,
it is unnecessary to monitor radiation levels in areas which are in direct contact with primary containment where penetrations and hatches are located since access to these areas is not required for any reason following an accident).
- Revision 3 of Regulatory Guide 1.97 provides an alternative requirement to Revision 2 which the Power Authority finds more acceptable.
3.2 Suppression Pool Water Temperature (A-3, D-6)
The installed instrumentation consists of 16 temperature sensors distrJLuted around the suppression pool which display on a data log g e, r in the control room.
Individual area, as well as bulk pool, temperature measurement can be indicated.
The temperature sensors are fully qualified in accordance with Category 1 design criteria.
The system is single channel and powered from a single power source, 120\\AC safety related Bus-Bl.
Suppression pool temperature sensing therefore does not meet ringle failure criteria.
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l The existing instruments will meet Category 2 design criteria, once the data-logger is upgraded to QA Category I.~
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addition to the above described instrumentation, there are redundant channels of pool temperature measurement with continuous indication in the control room.
These instrument loops will.neet Category 1 design criteria once their power source is upgraded to redundant lE supplies.
1 This combined measurement capability assures that the. operator will be provided the pool temperature information required to make decisions regarding preplanned. manual actions assuming a.
single failure.
With a minor upgrade of power. sources,.the existing instrumentations will meet the objective of Regulatory Guide 1.97.
1 i-3.3 Neutron' Flux (B-1) t The instrumentation presently installed for the source range d
5 neutron monitoring ( S RM ) system consists of three groups of monitors with four channels in each group (one group of;8RM's and two groups of Intermediate Range Monitors (IRM).
The source range monitors have a range.of 0.1 to 106 counts per second.
In the power range, neutron flux is monitored by:
j fit d in-core ion chambers which are arranged in a uniform j
pattern throughout the core.
These chambers cover a range of approximately 1% to 125% of rated power with a linear scale..
I The average power range level is measured by six average power-range monitors (APRM).
Each monitor measures bulk power in the core by averaging signals from 14 or 17 LPRM signals.
The source range and power range neutron monitoring instruments are continuovisly displayed (indicator and strip I
chart recorder) in the coattol room.
The SRM and IRM detectors and monitors are povered'from the 24 VDC Buses l
(station batteries), and the strip chart recorders are powered from the 120 VAC UPS Bus.
The power range detectors and-i monitors are' powered from the 120VAC, RPS Bus A and B, and the strip chart recorders are powered from,the-120VAC UPS Bus.
t The source range neutron monitoring instruments are not environmentally qualified, but are seismically qualified to original plant licensing requirements.
As previously stated (Reference 13), the Authority considers that there is no equipment in this system requiring qualification per 10 CFR 50.49 based on the system's design basis.
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k It is the NUTAC position that a rigorous Category 1 design
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criteria is not justified when the purpose and use of the measurement is analyzed.
In addition,
- a. qualification upgrade t -
(environmental and electrical) is not needed since boron sampling and CRD position indication provide adequate backup.
' ' TAC has also taken exception to the lower end of the Regulatory Guide 1.97 specified range based on an approach to criticality scenario.
In addition, diversity is provided by
-the overlapping regions of the source range detectors and also by che boron concentration data from the post accident samling system.
3.4 Drywell Sump Level (C-6) (B-8)
The instrumentation which measures drywell sump level consists of narrow range (15-45 inches), single monitor channel drywell equipment drain sump and drywell floor drain sump.
Both sumps are located inside the drywell (containment).
Only one of these sumps has a seismically and environmentally qualified level transmitter.
Each sump level is indicated in the Control Room on a strip chart recorder.
Each recorder is powered from an emergency bus.
The recorders are QA Category II.
4 Two additional drywell-level measurement channels provide overlapping, wide range, (0-100 feet), monitoring.
These instruments meet Category 1 design criteria.
The BWR Owners Group has taken the position that this variable should be implemented as Category 3.
The Authority believes the above instrumentu are adequate for the purpose of drywell sump level monitoring.
The following discussion supports this conclusion.
The BWR Mark I drywell has two drain sumps.
One drain is the equipment drain sump, which collects identified leakage; the other is the floor drain sump, which collects unidentified l
leakage.
l Although the level of the drain sumps can be a dire ct indication of breach of the Reactor Coolant System Pressure Boundary, other instrumentation required by Regulatory Guide i
1.97 will detect leakage in the drywell before sump level instrumentation.
In the case of the small line break, drywell pressure will increase before a noticable increase in suup level.
In addition, drywell temperature and primary containment radiation will increase, both monitored by l
Regulatory Guide 1.97 instruments.
Large line breaks would I
flood sumps rapidly and render narrow range measurement useless.
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The drywell'aump level signal neither automatically ini-
.tiates safety-related systems nor_ alerts the operator to the _
need to take safety related actions.
Regulatory Guide 1.97 requires instrumentation to function during and after an 4
accident.
.The.drywell sump systems are deliberately isolated by an accident signal to establish containmentLintegritys therefore, the drywell' sump level is not a useful' accident monitoring variable.
3.5 Radiation Exposure Rate (C-14, E-2, E-3)
[
The installed area radiation monitors (ARM) have a range of
.10-1 to 103 MR/HR.
These units are not presently environmentally qualified.
We believe the existing system is adequate since following an accident, there is no requirement to-enter areas monitored by ARMS.
The present four decade system is adequate for-monitoring plant areas.under normal operating conditions.
In addition, in an accident situation, local radiological surveys for beta and gamma dose rates as well as airborne activity samples would be' required prior to entry, and would provide the necessary exposure rate information.
This above position is supported by an evaluation which was performed to comply with a requirement of Emergency Preparedness Appraisal 50-33/83-03, Appendix.B, Item 11 (Reference 18).
In addition, it is the BWR Owner's Group l
position that using radiation exposure rate monitors to detect primary. containment breach is neither feasable nor-necessary.
The Authority endorses this postion.
3.6 Standby Liquid Control System Flow (D-17)
'ndication of the Standby Liquid Control i
There is'no direct i
j System (SLC) flow provided to the operator in the Control 4
Room.
The SLC system is manually initiated.and the.SLC system pump discharge header' pressure is indicated in the Control
[
Room and indicates.SLC systam pump operation.
In addition, l.
the operator can verify the proper functioning of the SLC system by monitoring the following on the control room panels l
Loss of continuity to squib valve.annuciator Loss of amber squib valve ready lights r
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A Pump has red light-indicates running SLC pump. discharge pressure is greater than reac-tor pressure Decreasing SLC tank level Reactivity change in tne reactor as measured by
- Neutron Monitoring 1:1strumentation We believe that the above list of indirect indications provide adequate information to verify SLC system operation.
The BWR Owners Group position supports our conclusion that monitoring the SLC system can be adequately done by measuring other variables associated with the system.
3.7 High Radioactivity Liquid Tank Level (D-23)
Level indication is not provided in the Control Room for the high radioactivity liquid tank.
The JAF plant is designed so that the pump discharge from the drywell floor and equipment drains sump pump is automatically isolated by an accident signal.
Furthermore, before radioactive liquid waste can be forwarded.to radwaste, Control Room operators must verify with radwaste operators that there is sufficient capacity in the radwaste systems to accomodate the wastes.
Plant design and administrative controls preclude the need to indicate this level in the Control Room.
The NUTAC R.G.
1.97 Implementation Guideline (Reference 7) supports our position that the d i s p.'.a y o f this variable in the Radwaste Control Room is adequate, and indication in the Control Room is not required.
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3.8 Trend Recording It is the BWROG and NUTAC position that dedicated recorders
.are necessary only where trend information is immediately required for operator use.
The current value (indicated) of the variables is normally used by the operator for decision making purposes.
3.9 Laboratory Analysis (E14, C2, B3)
The analytical procedures at JAF provide adequate range, sensitivity, and accuracy to evaluate post-accident status with the exception of dissolved gases due to sample system j
inadequacyi (refer to Section 4.17).
Data from post-accident l
sample analysis will be provided to the operator by the radiation chemistry lab as required.
Presently, there is no on-line data link between the Chemistry Lab and Control Room.
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4.0. Proposed Modifications
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4.1 A. number of exceptions to the recommendations of Regulatory Guide 1.97 Rev. 2 have been identified in this assessment.
The following modifications are proposed for future installation.
These additions and changes will comply with the requirement of Regulatory Guide 1.97 Rev.
2.
4.2 RPV Water Level jA-2. B-4)
The range of existing RPV Water level instruments will be extended to meet thw recommendations of the Regulatory Guide.
This will be accomplished by recalibrating one existing instrumentation channel and adding a second channel with trend recording.
4.3 Drywell Temperature (A-7, D-7)
The Category 1 der.ign criteria will be met by adding continuous indication in the Control Room of the two existing temperature loops in the area of the reactor and adding a trend recorder to one channel.
Drywell. cooler inlet and outlet' air temperature sensors will be replaced with fully qualified units.
The power supply to.each loop will be upgraded to Class lE redundant sources, and a trend recor der added to one channel.
4.4 Suppression Chamber Pressure (A-8)
Fully qualified Class IE pressure transmitters will be installed.
Continuous indication will be installed in the Control Room with one channel trend recorded.
Power for these loops will be upgraded to redundant Class 1E power sources.
4.5 Containment oxygen Concentration (A-ll, C-12)
The existing analyzer is not qualified.
A new fully qualified-system will be installed to meet the Category 1 design criteria.
4.6 Primary Containment Isolation Valve Position (B-10)
The instrumentation presently installed for the indication of containment isolation valve position consists of indicating lights on Control Room and Relay Room panels.
Position indication for those containment isolation valves which have their indicating lights on panels in the Relay Room will also be provided in the Control Room.
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4.7 Drywell/Suppresion Chamber Spray Flow (D-3, D-8)
The existing instrumentation meets all Category 2-design criteria.
However, the range of RHR System indicated flow is 25,000 GPM per loop, and the operator is required to throttle the combined flow of both RHR Loops.
In order to display flow measurement with acceptable resolution, new flow measurement loops will be added in the branch lines to the drywell and suppression-chamber spray headers.
The range will envelope the requirementLof 110% design flow with a more appropriate margin.
The flow instrumentation will meet all Category 2 design-requirements.
4.8 MSIV Leakaqe' Control System Pressure (D-9)
MSIV Leakage Control System Pressure is not indicated in the Control Room.
Two pressure measurement loops will be installed (MSLCS Train A & B).
These loops will provide on-demand indication in the control. Room Emergency and Plant Information Computer (EPIC).
The instrumentation will meet all Category 2 design criteria.
4.9 RHR Heat Exchanger Outlet Temperature (D-20)
The existing temperature sensors will be replaced with fully qualified sensors.
The temperature measurement will be indicated on-demand on the Control Room EPIC console.
4.10 Cooling Water Temperature to ESF System Components (D-21A)
The existing temperature sensor will be replaced with.one which is fully qualified.
The temperature measurement will be indicated on-demand in the Control Room Emergency and Plant i
Information Computer (EPIC) console.
4.11 RHR Service Water Temperature (D-21B)~
l The existing temperature sensor will be replaced with one which is' fully qualified.
The temperature measurement will be indicated on-demand on the Control Room EPIC console.
i 4.12 Reactor Building Flood Level (Potential D Variable)
Present plant instrumentation does not provide for measurement of flooding in the reactor building.
In the review of the
_ plant emergency operating procedures, the need to neasure this l
parameter was identified.
Reactor building flood level has been classified as a type D variable, because it provides information to indicate the operation of individual safety systems and other systems.
It also providea information to the operator for making decisions in using the individual systems which mitigate the consequences of a accident.
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4 Therefore, a level measurement loop will be installed in the lower elevation of the Reactor Building.
Instrumentation will be qualified.in accordance with Category 2 design criteria.
The level measurement will be indicated on-demand in the Control Room EPIC console.
4.13. Vent Exhaust Flow Rate (E-4)
Presently, only local measurement of vent exhaust flow is a
installed.
Instrumentation-to measure this parameter will be installed in the Reactor Building, Turbine Building, Radwaste Building exhaustivents, and the plant stack.
The flow indication will be displayed on the Control Room EPIC console.
4.14 Secondary Containment Purge - Vent Flow Rate (E-4.2)
The existing flow transmitter will either be qualified by analysis or replaced with a fully qualified unit.
Indication is presently available in the Control Room.
4.15 Emergency Ventilation Damper Position Emergency Ventilation Dampers not presently having position indication will be equipped with position switches and indication will be provided in Control Room.
Those Dampers not presently equipped with qualified switches will be upgrade as part of,the on going Equipment Qualification Program.
4.16 Status of Standby Power and other Energy Sources Important to Safety.
i Bus voltages which require Control Room indication will be added to the Plant Emergency Computer (IIPIC) Input List and displayed on demand in the Control Room.
4.17 Primary Coolant and Sump Accident Sample l
New equipment will be added to the existing Sample System to l
provide capability for measurement of dissolved gases.
5.0 Summary Since the issuance of Regulatory Guide 1.97 and NUREG-0737, JAF has completed a number of modifications to comply with these documents.
We believe these modifications in addition to modifications proposed in Section 4.0 of this report will bring JAF to an adequate level of conformance with the requirements of Regulatory Guide 1.97, Rev.
2.
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This report. provide the documentation required by-Supplement
.No. l..to.NU REG-07 3 7 (Section 6.2).
Variables A through E of Regulatory Guide 1.97 will be displayed on demand on the JAF Emergency.and Plant Information Computer, which will have display consoles in the Technical Support Center (TSC) and in the Emergency Operations Facility (EOF).
6.0 Position Summary Tables The tables in this section present the information required by Regulatory Guide 1.97 for the James A.
FitzPatrick Nuclear Power Plant.
They compare the plant with the requirements of the guide and thus demonstrate the plants degree of conformance.
1 Where modifications have been proposed, they are identified in the comments column.
Where exception has been taken to a requirement of Regulatory Guide 1.97, this has also been noted in the comments column.
Where the word " EPIC" appears under the Technical Support Center (TSC) or Emergency Operativas Facility (EOF) column, it means that that parameter is displayed in the TSC or EOF.
The Emergency and Plant Information Computer (EPIC) is an integrated system that combines SPDS functions, plant process j
computer functions (including NSSS and BOP) and supplementary operator aids.
Each Regulatory Guide 1.97 variable implemented at FitzPatrick wil) be avaliable'via the EPIC system in both the Technical Support Center (TSC) and 3
Emergency Operations Facility (EOF), with the exception of meteorological variables which will be provided by a separate computer system.
Refer to the Authority's "SPDS Implementation Plan" for further information on these systems.
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SE S.
POER EDr110L ftOGt I1194 VARIABIE CJC.
Oh IU$CE QJAL.
GIAL.
IIEDQ UWrf SUPPLY INDICATICH TSC EF (X30 e ffS R.G.
1 I
0-110% design Cat. 1 Cat. 1 Yes IE Continuous &
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NYPA 1
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+250 peig Recorder (4 ch.)
EPIC HPIC (0-5204)
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l.97 design flow 1 ch. remrded A6 ItHR Systemt Flow Meets the intant of R.G.1.97 i
NYPA 1
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R.G 1
I 40*F to 440*F Cat. 1 Cat. 1 Yes IE Chmtinuous &
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I 40*F to 440*F 10 0 No No IE Indicator (2 ch.)
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SEIS.
POMR CON!ROL 800M I1138 VARIABIE CAT.
QUAL.
Rl!DED10mper SUPPLY INDICATION TSC IDF Q M EBffS R.G.
1 I
Bottas of Core Cat. 1 Cat. 1 Yes IE Continuous &
1.97 to Main Stems 1 ch. recorde1 Line a
B4 Coolant Invel in Inadequate Range Reactor NYPA 1
I
-100 in. to Cat. 1 Cat. 1
.Yes IE Indicator (2 ch.) EPIC EPE Refer to
+224.5 in.
Recorder (1 ch.)
Para. 4.2 j
85 BMt Core Isot recparied per h - - - - 91ev Stgip. I to IRJRIG-0737 R.G.
1 I
15 psia to Cat. 1 Cat. 1 Yes IE Continuous &
2 1.97 1500 poig 1 ch. recorded 86 RCS Pressure Quplies NYPA 1
I 0-1500 psig Cat. 1 Cat. 1 Yes IE Indicator (2 ch.1 EPIC EPIC Recorder (2 ch.)
4 i
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1 I
0-110% design Cat.1 Cat. 1 Yes IE Continuous &
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Sheet 7 of 20 J.A. FITZPATRIOC IEJCIEAR P0bER PIAfrF ltIEUIATGtY GJIDE 1.97, REVISIOf 2 ASSESSperr POSITIO4 SLMERY TABIES
(
DETitLDerr BfV.
SEIS.
POMR CXNDOL RXM I194 VARIABIE CRF.
RANGE QJAL.
GIAI REDLSDert SL;PPIX D8DICATIO4 TSC KF CGeerfS R.G.
3 III 10 u C1/ga to Cat. 3 Cat. 3 neo Isai-IE Continuous or 1.97 10 Ci/gn on demand C2 Analysis of Primary Chaplies Coolant (N Spwtrism)
IfYPA 3
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C3 IMt(bre geot rerpired per 1hermoccmples Sw. I to aftNt!G-0737 R.G.
1 I
15 psia to Cat. 1 Cat. 1 Yes IE Continuous &
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I 0-1500 poig Cat.1 Cat. 1 Yes IE Indicator (2 ch.1 EPE Rearder (2 ch.)
R.G.
3 1-105 R/hr Cat. 3 Cat. 3 too lean-IE Continuous or 1.97 on damerid C5 Primary Omtainment Caplies Area Radiation NYPA 1
1 1-108 R/hr.
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Sheet 9 of 20 J.A. FIT 2 PATRICK IGOCIEAR POBE!R P!JWfr P%IMORY GJIDE 1.97, REVISIos 2 MiSESSeeft POSITION SUIMARY 1Nt!25 DISTRLae!Nr Inev.
SEIS.
Potem Colf!ROL 500st I'11!M VARIAEi2 CAT.
QA RANGE QJAL.
QUAL.
RI!DG O Wff SUPPIX INDICATI06 TT IEF COGEnftS R.G.
1 I
10 psia to 4 x Cat. 1 Cat. 1 Yes IE Continuous or C10 Primary Containment C g lies Pressure t
NYPA 1
1
-5 psig to Cat. 1 Cat.1 Yes IE Indicator (4 ch.) EPE EPE
+250 psig Recorder (4 ch.)
l R.G.
1 I
0-30% R2 Cat. 1 Cat. 1 Yes IE Continuous &
Cll Containment and 1.97 1 ch. recorded Drywell flydrogen Complies Ocmcentratica NYPA 1
1 0-30% H2 Cat. 1 Cat. 1 Yes IE Indicator (2 ch.) EFE EPE Refer to Para. 3.8
{
liefer to R.G.
1 I
0-104 02 Cat. 1 Cat. 1 Yes IE Ccntintamus &
Para 4.5 C12 Containumnt and 1.97 1 ch. recorded Qualifiel t
Drywell oxygen analyzers Concentration (for to be inerted containments)
NYPA 1
I 0-10% O2 No leo Yes IE Recorder (2 ch.)
EPE EPE pau R.G.
3:
III 10-6 to 10-2 Cat. 3 Cat. 3 No lean-IE Recordal l
Cl3 Containment Effluent 1.97 u Ci/cc Radioactivity -
Complies I
Noble ca m (Frose i
identified re1====
NYPA 3
I&
10-1 to 106 eps Cat. 3 Cat. 3 No 10cm-IE Indicator (4 ch.)
EPIC EPE points incitaling II Recorder (4 ch.)
1 stanchy Gas Treatment 10-1 to 107 Systesa Vent) mR/hr r
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Sheet 11 of 20 J.A. FITZPATRICK NUCIEAR PubER PIhff REGUIATORY GUIDE 1.97, REVISIOi 2 ASSESSME3rf POSITIGE SRMARY TABIES DISTRR G rf B8V.
SEIS.
POER (IN!ROL XKM 111M VARIAR[3 CAT.
Q4 RANGE GIAL.
GIAL.
REDO OANT SUPPIX DIDICATIOl TSC 50F (X30erfS i
0-1109 Cat. 2 Cat. 2 10 0 tecn-IE Ctmtinuous or R.G.
2 D3 Sgpression Chamber 1.97 design flow on demand Spray Flow (Rim Iow range flow Systemi Flow indication to be NYPA 2
II 0-250%
Cat. 2 Cat. 2 Yes IE Irdicator (2 ch.) EPE EPE installed design flow Refer to Para. 4.7 R.G.
2 0-1104 Cat. 2 Cat. 2 No liart-IE Omtinuous or 1.97 design flow on demarut D4 Drywell Pressure Complies NYPA 2
1
-5 peig to Cat.1 Cat. 1 Yes IE Indicator (4 cI:Jl EPE EPE
+250 peig Recorder (4 ch.)
BTS Suetion tc Cat. 2 Cat. 2 No Ikm-IE Cantinuoas or R.G.
2 1.97 Ncrual +5 ft.
on demand DS S g pression Pool Otmplies Nater Invel NYPA 2
II 0 - 30 ft.
Cat. 1 Cat. 1 Yes IE Indicator (2 ch.1 EPIC EPE Remeder (2 ch.)
30*F to 230*F Cat. 2 Cat. 2 too tecm-IE Cantinuous or R.G.
2 1.97 on demand D6 Suppression Pool Requires Nater Temperature Onalification NYPA 2
II 30*F to 230*F No tio No UPS On demand EPIC EPE Refer to data logger Para. 3.2 i
t
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5
Sheet 14 of 20 J.A. FIT 2PATREK NUCMAR POER PIANT RICUIMORY GUIDE 1.97, REVISION 2 ASSESSM3rf POSITION SMIARY TABIES IMi' HOE!NF ENV.
SEIS.
POMR Cmr!ROL 300M I'IDI VARIAB M CAT.
RANGE QUAL.
GIAL.
REDGDer?
SUPPIX INDICATION T9C EDF C30GBr!S R.G.
2 0-110%
Cat. 2 Cat. 2 No Ncm-IE Ctatinuous or 1.97 design flow on demand D15 Core Spray complies l
Systen Flow NYPA 2
II 0-151%
Cat. 1 Cat. 2 Yes.
IE Indicator (2 ch.1 EPE EPE design flow R.G.
2 0-1104 Cat. 2 Cat. 2 No Ncm-IE Ocntinuous or 1.97 design flow on demand (RHR Sys Flow) of R.G. 1.97 a
NYPA 2
II 0-1084 Cat. 1 Cat. 2 Yes IE Indicatcr (2 ch.) EPE EPE design flow I ch. recorded R.G.
2 0-110%
Cat. 2 Cat. 2 No Non-IE Continuous cc 1.97 design flow on denand D17 SICS Flow Refer to Para. 3.6 NYPA None None l
R.G.
2 Bottaa to top Cat. 2 Cat. 2 No Non-IE Contintmus or 1.97 on demand D18 SIIS Storage Ccuplies -
Tank Invel amiting Jmei NYPA 3
I 0-1004 No Cat. 2 No UPS Indicator EPIC EPE Resolution i
j 4
1 1
j i
i i
{i!
- i!J!
02 f
t s
o n
rd n
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- j lJ
Sheet 17 of 20 J.A. FITZPATRECK IUCIEAR POMER P!Rfr REUIATORY GJIDE 1.97, REVISION 2 ASS!!S9eff POSITIOl SUNmRY TABIES DISTRLBeff ENV.
SEIS.
POMR CIN!ROL LOOM I1190 VARIABIE CRT.
(R RANGE GJAL.
QUAL.
REDGEDMir SUPPIX INDICATIGI TSC K)F CGe efri R.G.
2 II Plant Specific Cat. 2 Cat. 2 No Non-IE Cbntinuous or D25 Status of Stane y 1.97 on demand Some Ims voltages Power and Other regaire con *aol i
Energy Sources room indicatim.
Inportant to Safety NYPA 2
II various Cat. 2 Cat. 2 YesAeo IE/
Indicators EPIC EPE Refer to Non-IE Para. 4.16 a
R.G.
1 1
1-107 R/hr.
Cat. 1 Cat. 1 No IS Cbntinuous and El Primary Cbntalrument 1.97 1 ch. recorded Refer to Area Rm11ation - High Yes IE Indicator (2 ch.)
EPIC EPE Para. 3.8 Range NYPA 1
I 1-108 R/hr.
Cat. 1 Cat. 1 R.G.
2 10-1-104 mRAr. Cat. 2 Cat. 2 No Non-IE Recorded E2 Reactor Building or 1.97 Refer to Secxmdary Cbntainment Para. 3.5 Area Radiation NYPA 3
III 10-1-103 mR/hr< Cat. 3 Cat. 3 No Non-IE Indicator &
Recorder R.G.
2 10-1-104 Cat. 2 Cat. 2 No Non-IE Records!
Refer to 1.97 mR/hr Para. 3.5 E3 Radiation P ire Rate NYPA 1
1 i
- k
{
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I
Sheet 19 of 20 J.A. FITZPA1RKK IOCIEAR poler PIAprF RIEUIA10RY GUIDE 1.97, REVISION 2 ASSESSPE!NF POSITION SLD94ARY TABIES INS 1RM!NF I!NV.
SEIS.
POER CINDOL IOOM IT1!M VARIABIE CAT.
OpL RANGE QJAL.
01AL.
RI!DGOMrF SUPP!X INDICATION T3C IDF C00EDfrS R.G.
3 10-3 to 104 Cat. 3 Cat. 3 iM Non-IE On demand 1.97 rads /hr.
E8 Plant and Environs Complies Radiation (portable instr.)
NYPA 3
III Instnaments Cat. 3 Cat. 3 No Non-IE Portable can envelope aquipment range recpairements
_m R.G.
3 Italtidiennel Cat. 3 Cat. 3 No Non-IE On demand 1.97 gersna-ray E9 Plant ard Bwirons spectrometer Radioactivity complies (portable instr.)
NYPA 3
III Multi &annel Cat. 3 Cat. 3 No Non-IE Oisite & Offsite rJ - -ray Imboratories spectrometer 0 - 360' Cat. 3 Cat. 3 No Nat-IE ascarded R.G.
3 E10 Wind Direction 1.97 C W lies Met.
Met.
NYPA 3
III O - 360*
Cat. 3 Cat. 3 No Nmi-IE Rearder (3 ch.)
Sys.
Sys.
R.G.
3 0 - 67 aph Cat. 3 Cat. 3 No Non-IE Rearded Ell Wind Speed 1.97 Complies Met.
Met.
NYPA 3
III 0 - 100 mph Cat. 3 Cat. 3 No Ncri-IE Recorder (3 ch.)
Sys.
Sys.
Sheet 20 of 20 J.A. FITEPA!RKK 10CIEAR IOE!R PIANT i
RIKUIATORY alIDE 1.97, REVISICN 2 ASSESSPENF POSITION SIDM4RY TABIES INSTIUEDIT ENV.
SEIS.
IOE!R (Dr1BOL IOOpt I1198 VARIAIEE CRF.
Ok RANGE QJAL.
QUAL.
ItEDIADWIT SUPPIX INDICATION TSC I!DF CGOENIS
-9'F to 18'F Cat. 3 Cat. 3 No 10an-IE Itecorded R.G.
3 E12 Estimatica of 1.97 Complies l
Atmosphere Stability stet.
Met.
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NYPA 3
III -9'r to 18'F Cat. 3 Cat. 3 10 0 Man-IE Decorder Sys.
Sys.
R.G.
3 Grab Sample Cat. 3 Cac. 3 No lim -IE On demand Dissolved H2 s 02 E13 Primary Coolant and 1.97 m pability in Simp Accident Sagle progress.
]
NYPA 3
III Grab Sample Cat. 3 Cat. 3 No tecn-IE Sample Analysis Refer M Para. 4.17 J
J Grab Sample Cat. 3 Cat. 3 No Ntn-IE On demand
]
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g.p 3
Accident Sample of R.G.1.97.
j NYPA 3
III Grab Sa@ le Cat. 3 Cat. 3 No Non-IE Sample Analysis Refer to Para 3.9 i
i i.
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7.0 Roforcncan
.l.
NRC Generic Letter No. 83-33 dated December 17, 1982 includes Supplement No. 1 to NUREG-0737, " Requirements for Emergency Response Capability."
2.
Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Asscc.3 Plant and Environs Conditions During and Following asi Accident," Revision 2, December 1980.
3.
lOCRF50, Appendix A,
" Domestic Licensing of Production and UtElization Facilities" General Design Criteris 13, " Instrumentation and Control."
4.
10CFR50, Appendix.A, " Domestic Licensing of Production and Utilization Facilities" General Design Criteria 19, " Control Room."
5.
" Domestic Licensing of Production i
and Utilization Facilities " General Design Criteria 64, " Monitoring Radioactivity Releases."
1980, " Criteria for Accident Monitoring
- 6. ANSI /ANS-4.5 Functions in Light-Water-Cooled Reactors."
7.
INPO 83-049 (NUTAC) " Regulatory Guide 1.97 (Accident Monitoring Instrumentation) Implementation Guideline."
8.
IEEE 191., "IEEE Standard for Qualifying Class 323 lE Equipment for Nuclear Power Generating Stations."
1974, "IEEE Standard for Qualifying Class 323 9.
IEEE 1E Equipment for Nuclear Power Generating Stations."
I 0588, " Interim Staff Position on environmental
- 10. NUREG Qualification of Safety-Related Electrical Equipment Raview" dated August 1979.
- 11. IEEE-344-1971 " Seismic Qualification offclass lE Equipment for Nuclear Generating Stations, l
Recommended Practices for."
- 12. IEEE-344-1975 "Seismit Qualification of Class lE Equipment for Nuclear Generating Stations, Recommended Practices for."
3
- 13. PASNY letter, dated May 20, 1983, regarding environmental qualification of electrical equipment 10 CFR 50.49, (JPN-84-45).
l l
15 -
l
~
14 Regulatory Guide 1.32, " Criteria for safety - Related Electric Power Systems for Nuclear Power Plants, Revision 2.
- 15. BWR Owners Group Report, " Position on NRC Regulatory Guide 1.97, Revision 2,"-dated May 1982.
- 16. USNRC Regulatory Guide 1.97, Draft 1, Revision 3, dated December 3,
1982.
- 17. USHRC Inspection Report No. 83-03 regarding Emergency Preparedness Appraisal.
O A. _.
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT ATTAHCMENT - 2 P
SCHEDULE FOR IMPLEMENTATIONS j
OF VARIOUS MODIFICATIONS TO MEET REGULATORY GUIDE 1.97, REV. 2 I
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1 of 2 SCHEDULE FOR INSTALIATION OF REGULATORY GUIDE 1.97 VARIABLES AT JAMES A. FITZPATRICK NUCLEAR POWER PLANT f
REGULATORY GUIDE 1.97, REV. 2 INSTALLATION COMPLETION ITEM MEASURED VARIALBE TYPE CAT.
DATE 1.
Coolant Level in A,B 1
30 days after 1986 Refueling Reactor Vessel Outage (Reload 7/ Cycle 8) or January 31, 1984, whichever is later.
2.
Suppression Pool A,D.
1,2 Same as Item No. 1.
Water Temper" ture a
3.
Drywell Atmosphere A,D 1,2 Same as Item No. 1.
Temperature 4.
Suppression Chamber A,
1 Same As Item No. 1.
Pressure 5.
Containment and A,C 1
30 days after 1985 Reft eling Drywell Hydrogen Outage (Reload 6/ Cycle 7) or Concentration December 31, 1985, whichever is later.
6.
Containment and A,C 1
Same as Item 5.
Drywell Oxygen Concentration (for inerted' containment plants).
7.
Neutron Flux B
1 30 days after 1987 Refueling Outage (Reload 8/ Cycle 9) or December 31, 1987, whichever is later.
l.
8.
1 Same as Item 7.
l Isolation Valve Position (excluding check valves) l 9.
Radioactivity C
1 Same as Item 1.
Concentration or l
Radiation Level in L'
Circulating Primary Coolant t
10.
Containment Effluent C
3 Same as Item 5.
l Radioactivity -
Noble Gases (from identified release points including Standby Gas Treatment System Vent) 9 7
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i 2 of 2 SCHEDULE FOR INSTALLATION OF REGULATORY GUIDE 1.97 VARIABLES AT JAMES A. FITZPATRICK NUCLEAR POWER PLANT-REGULATORY GUIDE 1.97, REV. 2 INSTALLATION COMPLETION-ITEM.
MEASURED VARIALBE TYPE CAT.
DATE 11.
Suppression Chamber D
2 Same as Item 7.
Spray Flow 12.
Main Steamline Same as Item 1.
Isolation Valves' Leakage Control-System Pressure 13.
Primary System Safety D
2 Same as Item 5.
Relief Valve Position, Including ADS or Flow Thraugh or Pressure in Valve Lines 14.
SLCS Flow D
2 Schedule to be determined as required by ATWS Ruling.
15.
SLCS Storage Tank D
2 Same as Item 14.
l 16.
RHR Heat Exchanger D
2 Same as Item 1.
Outlet Temperature 17.
Cooling Water D
2 Same as Item 7.
Temperature to ESC System Components 18.
Cooling Water Flow D
2 Same as Item 7.
to ESF System Components 19.
High Radioactivity D
3 Same as Item 1.
Liquid Tank Level I
20.
Emergency Ventilation D
2 Same as Item 1.
Damper Position 21.
Vent Flow Rate E
2 Same as Item 7.
22.
Status of Standby D
2 Same as Item 7.
Power & Other Energy Sources Important to Safety 23.
Primary Coolant and E
3 Same as Item 7.
Sump Accident Sample l
m.
. - - _ _ _.., -,. _ _ _,. _ - _ _ _. ~.. -
.