ML20128L828
| ML20128L828 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/09/1983 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, JPN-83-92, NUDOCS 8507250125 | |
| Download: ML20128L828 (31) | |
Text
--
123 Main Street g,
Whete Plains, New York 10601 914 681.6240 g
g J. Phillip Bayne Executrve Vice President g
Nuclear Generation November 9, 1983 JPN-83-92 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 Attention:
D.
B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Generic Implications of Salem ATWS Events (Generic Letter 83-28)
References:
1.
G.
Eisenhut to all Licensees, dated July 8, 1983.
2.
NYPA letter, J.
P.
Bayne to D.
G. Eisenhut, dated September 6, 1983 (JPN-83-80).
3.
NRC letter, D.
B.
Vassallo to J.
P.
- Bayne, dated October 19, 1983.
Dear Sir:
Reference 1 requested information, plans and schedules relating to the generic inplications of the March, 1983 Salem ATWS events.
Reference 2 provided a preliminary schedule for submitting the requested information.
In addition, Reference 2 requested an extension for those items for which a detailed response cannot be provided by November 7, 1983.
Reference 3 denied the extension and requested that infor-mation be submitted by November 5, 1983.
The attach-ment to this letter provides the Authority's response to the extent practical at this time.
Specifically, the attachment includes:
1.
A description of the current (as of November, 1983) programs and status for each item.
2.
A description of plans for changes to the current programs and/or procedures for each item.
8507250125 831109 6
PDR ADOCK 05000333 0 l; P
In a number of cases, the Authority cannot provide definative schedules for the completion of efforts to meet the require-ments of Reference 1.
The Authority is participating in both the BWROG's (Boiling Water Reactor Owners Group) and NUTAC's (Nuclear Utilities Task Action Committee) generic efforts to address portions of Reference 1.
We expect to be able to provide our plans and implementation schedules shortly after these groups finalize their efforts.
The information, plans and schedules in this submittal are provided in accordance with the information available at this time and the reviews completed to date.
The Authority reserves the right to amend this submittal if necessary.
If, as a result of this review, a revision is required, it will be submitted promptly.
If you have any questions, please contact Mr.
J.
A. Gray, Jr.
of my staf f.
Very tru1.y yours,
\\
3 L
Bayne 3
t E Ex cu ive Vice president Nu 1 r Generation State of New York County of Westchester Subscribed and Sworn to before me this
/ day of November 1983
/
JEANNE LA (UNA hCTARY PUBLIC. STATE OF NEW YORK n
NO, 60-4214103 OUAllFlf0 IN WESTCHtSitt COUNTY j
L g w + j)i s i< n TLtw EAPl.%$ MARCH 30di 19........
{/
Notary Public
NEW YORK POWER AUTHORITY JAMES'A. FITZPATRICK NUCLEAR POWER PLANT RESPONSE TO GENERIC LETTER 83-28
, to JPN-83-92 Dated November 9, 1983 This document provides a description of the current procedures or programs for each item identified in NRC Generic Letter 83-28.
For those items which are not currently addressed in procedures or
. programs, a description of the Authority's position or plans is provided.
The current best estimated date for completion is also provided, wherever sufficient information exists to project a date.
b
_1_
m ITEM 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)
A.
CURRENT POST-TRIP REVIEW PROGRAM
- 1. There is no formal program in place which addresses all the elements outlined in the generic letter for post-trip review and restart.
- 2. The following actions have always been taken when an unexplained shutdown has taken place.
a.
Knowledgeable plant personnel conduct a thorough review f
~
of the transient.
b.
In cases where the cause is not clear, an intensive analysis is initiated to determine the cause.
Documentation exists to support the analysis.
c.
There are only a few transients where the exact cause was not identified with a good level of confidence.
d.
Restart authorization from the Plant Manager or his designated alternate is required.
This requirement is a sign-off step in F-OP-65, Startup and Shutdown Procedure.
D.
PLANNED CIIANGES TO Tile POST-TRIP REVIEW PROGRAM
- 1. A new procedure has been drafted for post-trip review.
This new procedure will be implemented by January 6, 1984 and I
will address Items 1.1.1 through 1.1.6 of Generic Letter 83-28.
- 2. The BWROG is addressing Item 1.1.1 (restart criteria), Item 1.1.3 (qualification and training) and Item 1.1.5 (methods / criteria for event comparison).
Funding for these activities was approved by the UWROG at the October 26/27, 1983 meeting and General Electric is expected to complete the work by February 29, 1984.
Allowing for possible revision of the recently drafted procedure, the submittal of. _ _ _ _ _ _ _ _
m c dateiled report containing the_ final program doacription and procedure will be made by March 31, 1984.
ITEM 1.2 POST-TRIP REVIEW - DATA AND INFORMATION CAPABILITY A.
CURRENT DATA AND INFORMATION CAPABILITY
- 1. Capability for Assessing Sequence of Events The sequence of events function is provided primarily by the plant process computer Sequence of Events log.
The Sequence of Events log is initiated (printed) by a change in state of any of the digital (on-off) inputs selected for the sequence of events function.
Currently 161 points are utilized of which 45 are designated as spares.
The 116 points used at this time are listed below:
POINT ID DESCRIPTION D000 RFP A SUCTION PRESSURE D001 RFP B SUCTION PRESSURE D002 RFP A DISCHARGE PRESS
{
D003 RFP B DISCHARGE PRESS
{
D013 ll5KV SOUTH BUSS UV RLY l
D014 115KV NORTH BUSS UV RLY D015 ll5KV BRKR 10012 OPEN l
D016 ll5KV BRKR 10012 CLOSED D017 115KV BRKR 10022 OPEN D018 115KV BRKR 10022 CLOSED D019 345KV BRKR 10042 OPEN l
D020 345KV BRKR 10042 CLOSED l
D021 BREAKER FAILURE 10012 D022 BREAKER FAILURE 10022 D023 BREAKER FAILURE 10042 D024 BREAKER FAILURE 10052 D025 345KV NMPT LINE #10 D026 345KV EDIC LINE #1 D027 345 BUSS D028 115KV LGHT HSE HLL LINE D029 115KV LINE TO NINE MILE D030 345KV BRKR 10052 OPEN D031 RESERV STATION TFR T2 UV D032 RESERV STATION TFR T3 UV D033 345KV BRKR 10052 CLOSED D034 MOIST SEP A HI LVL TRP D035 MOIST SEP B HI LVL TRP D036 SONIC DETECTOR RV-2-71A D037 SONIC DETECTOR RV-2-71B D038 SONIC DETECTOR RV-2-71C D039 SONIC DETECTOR RV-2-71D
. l
D040 SONIC DETECTOR RV-2-71E D041 SONIC DETECTOR RV-2-71F D042 SONIC DETECTOR RV-2-71G D043 SONIC DETECTOR RV-2-71H D044 SONIC DETECTOR RV-2-71J D045 SONIC DETECTOR RV-2-71K D046 SONIC DETECTOR RV-2-71L D052 TBN EHC PANEL 24 VDC PWR D054 MAIN TURBINE TRIP D055 TBNE BACK UP OVERSPD TRIP D056 TBNE LOSS OF 125 VDC TRP D057 TURBINE MANUAL TRIP D058 TBNE EXH HOOD HI T TRIP D059 LOW COND VACUUM A TRIP D060 LOW TBNE BRG OIL PRESS D062 TBNE TRIP HI VIBRATION D063 TBNE TRIP-LOSS STAT CLNT D064 TBNE TRP-TilRST BRG WEAR D065 TBNE TRP-SHAFT PMP PRES D066 TBN TRP-EMERG TRP FLUID D067 TBN TRP-HYD FLUID PRESS D068 TBN TRP-LOSS SPEED FDBK D500 SDIV Al W LEVEL SW SCRAM D501 SDIV B1 W LEVEL SW SCRAM D502 SDIV A2 W ANLG TRP SCRAM D503 SDIV B2 W ANLG TRP SCRAM D504 MAIN STEAM LINE CHNL Al l
D505 MAIN STEAM LINE CHNL B1 D506 MAIN STEAM LINE CIINL A2 D507 MAIN STEAM LINF. CHNL B2 D508 CONTMT HIGH PRESS CH Al D509 CONTMT HIGII PRESS Cil B1 D510 CONTMT IIIGli PRESS Cil A2 D511 CONTMT llIGli PRESS CH B2 D512 REACTOR CliNL Al III PRESS D513 REACTOR CifNL B1 HI PRESS D514 REACTOR CHNL A2 III PRESS D515 REACTOR CilNL B2 HI PRESS D516 REACTOR LO WTR LVL CII Al D517 REACTOR LO WTR LVL Cil B1 D518 REACTOR LO WTR LVL CH A2 D519 REACTOR LO WTR LVL Cl! B2 D520 MSL A-1 ilIGli RADIATION D521 MSL B-1 HIGli RADI/sTION D522 MSL A-2 IIIGli RADIATION D523 MSL B-2 IIIGli RADIATION D524 NEUT MON SYSTEM CliNL Al D525 NEUT MON SYSTEM Cl!NL A2 D526 NEUT MON SYSTEM CilNL B1 D527 NEUT MON SYSTEM CHNL B2 D528 SDIV Al E LEVEL SW SCRAM D529 SDIV B1 E LEVEL SW SCRAM D530 MANUAL SCRAM CilANNEL A D531 MANUAL SCRAM CilANNEL B D532 REACTOR SCRAM CIIANNEL A D533 REACTOR SCRAM CilANNEL B D534 BOTil SCRAM CliANNELS A& B D535 SDIV A2 E ANLG TRP SCRAM D536 SDIV B2 E ANLG TRP SCRAM D538 TSV FAST CLOSURE CHNL Al D539 TSV FAST CLOSURE CHNL B1 D540 TSV FAST CLOSURE CHNL A2
~D541 TSV FAST CLOSURE CHNL B2 D542 TCV FAST CLOSURE CHNL Al D543 TCV FAST CLOSURE CHNL B1 D544 TCV FAST CLOSURE CHNL A2 D545 TCV FAST CLOSURE CHNL B2 D546 APRM CHNL A UPSCALE LVL 1
D547 APRM CHNL B UPSCALE LVL D548 APRM CHNL C UPSCALE LVL D549 APRM CHNL D UPSCALE LVL D550 APRM CHNL E UPSCALE LVL D551 APRM CHNL F UPSCALE LVL D554 IRM CHNL A UPSCALE LVL D555 IRM CHNL B UPSCALE LVL D556 IRM CHNL C UPSCALE LVL D557 IRM CHNL D UPSCALE LVL D558 IRM CHNL E UPSCALE LVL D559 IRM CHNL F UPSCALE LVL D560 IRM CHNL G UPSCALE LVL D561 IRM CHNL H UPSCALE LVL D562 WEST SDIV NOT DRAINED D563 EAST SDIV NOT DRAINED D564 WEST SDIV ROD BLOCK D565 EAST SDIV ROD BLOCK Time discrimination between events is one millisecond; i.e.,
for two events occurring within less than one millisecond of each other, the correct sequence cannot be guaranteed.
The Sequence of Events is printed on a typewriter in the control room.
A facsimile of the actual printout of data point D532 which was generated as a result of part of a routine surveillance test is shown below.
TIME MILLISECONDS POINT ID DESCRIPTION STATUS (hour, (on-off minutes trip-reset and seconds) etc.)
093247 840 SEQ D532 REACTOR SCRAM CHANNEL A TRIP 093250 719 SEQ D532 REACTOR SCEAM CHANNEL A RSET The process computer does not retain (store) Sequence of Events function data except as part of its providing data printout.
Once - -
tha dato 10 provid:d to tha printer it in no longer availablo within the computer.
The Sequence of Events printout hard copy is retained as part of the records retention program.
Power for the i
process computer and printer for the Sequence of Events is from the Un-interruptable Power Supply (UPS) system which is described in FSAR Section 8.9.
2.
Capability of Assessing the Time History of Analog Variables Equipment used to assess the time history of analog variables consists of the plant process computer Post Trip Log function and a number of strip chart recorders.
a.
The Post Trip Log contains 20 selected plant inputs and is automatically initiated upon occurrence of pre-defined plant trips.
It can also be initiated upon operator demand.
b.
Strip chart recorders (which are part of the numerous indicators and recorders provided for routine startup, operation, and plant shutdown) also provide data and information which may be useful for analysis of unscheduled shutdown and/or the functioning of safety-related equipment.
Parameters monitored for assessing the time history of analog variables are listed below for both the Post Trip Log and those strip chart recorders that may also be utilized...
r POST TRIP LOG POINTS B032 APRM A FLUX LEVEL
%PWR B033 APRM B FLUX LEVEL
%PUR B034 APRM C FLUX LEVEL
%PWR B035 APRM D FLUX LEVEL
%PWR B036 APRM E FLUX LEVEL
%PWR B037 APRM F FLUX LEVEL
%PWR B044 TOTAL CORE FLOW M#/HR B045 CORE DIFFERENTIAL PRESS B047 FDWTR LOOP A FLOW M#/HR B048 FDWTR LOOP B FLOW M#/HR B053 REACTOR WATER LEVEL INCH B054 TOTAL STEAM FLOW M#/HR B057 REACTOR PRESSURE PSIG B062 RX FW INLET Al TEMP DEGF F204 MAIN STEAM HEADER PRESS M017 DRYWELL PRESS
( ABSOLUTE)
M019 TOR WTR LVL (-72/+72)
INCH M020 TOR WTR-AT (NRM LMT=95)
T040 TURBINE SPEED RPM STRIP CffART RECORDERS RECORDER ID DESCRIPTION 06-LR/PR-97 REACTOR PRESSURE O to 1200 psig REACTOR WATER LEVEL 164.5 to 224.5 inches above Top of Active Fuel (TAF) 06-FR-96 REACTOR STEAM FLOW 0 to 12 (x106) lbs/hr.
FEEDWATER FLOW 0 to 12 (x106) lbs/hr.
06-PR/FR-98 REACTOR PRESSURE 800 to 1100 psig TURBINE STEAM FLOW 0 to 100% of rated i
l 02-3-FR/PR-95 CORE DP O to 25 paid j
CORE FLOW 0 to 90 (x106) lbs/hr.
07-PR-46 A, B, C, & D APRM A, B, C, D, E, & F 0 to 125% Power IRM A, B, C, D, E, F, G&ll 0 to 125 10-FR-143 LPCI LOOP A FLOW 0 to 25 (x103) gpm LPCI LOOP B FLOW 0 to 25 (x103).gpm 02-3-LR-98 REACTOR WATER LEVEL
-100 to +200 (FUEL ZONE) inches below (-)
or abcVe (+) TAF
~7-
02-TR-165 RECIRCULATION LOOP A TEMP O to 600*F RECIRCULATION LOOP B TEMP O to 600*F 3) 02-FR-163 RECIRCULATION LOOP A FLOW 0 to 70 (x10 RECIRCULATION LOOP B FLOW 0 to 70 (x103 ) gpm gpm i
94-TS-VP TUBINE BYPASS VALVE POSITION O to 100%
07-R-45 SRM B or D 10-1 to 106 CPS SRM A or C 10-1 to 106 CPS Post Trip Log data it continuously stored and updated at 2 second intervals in a portion of the process computer memory and remains in the memory for 2 minutes.
Thus the memory contains the most recent 60 data bits for each post trip log parameter prior to a plant trip.
Upon occurrence of a pre-selected plant trip condition the plant process computer program causes the data which was stored for 2 minutes prior to the trip (and new data at two second intervals for the same 20 parameters for a 2 minute period after the trip) to printed out as the Post Trip Log.
The Post Trip Log therefore contains data for 2 minutes prior to the trip and data for 2 minutes after the trip at 2 second intervals totaling 120 data entries for each of the 20 post Trip Log data points.
Selection of the parameters currently used was based on the recommendation of the NSSS vendor during initial startup of the plant, and no substanital change to the 20 parameters monitored has taken place since that time.
Sample rate for the Post Trip Log parameters is limited by the capability of the plant process computer and its program.
Each of the 20 parameters monitored is limited to j
120 data points.
Selection of the 2 second sample rate is considered optimum, given the limitations of the currentJy installed equipment. -
ra
)
38-Strip chart ricordar data which cay also ba'uced is recorded continuously.
Chart paper speed is normally'one (1) inch per hour
.and each chart is generally date/ time stamped on a daily basis for reference, Neutron monitoring recorders ( 07-PR-46 A, B, C, &D, R-45 )
and the. reactor water level recorder (06-LR/PR-97) may'also be operated with a chart paper speed of one-(1) inch per minute.
Ilowever this feature is normally used by the operator only during plant startup and scheduled shutdown.
Post Trip Log Format is shown below:
Time PT.ID1 PT ID2 Pt ID20 XXXXXX.
XXXX XXXX XXXX VALUE1 VALUE2 VALUE O 1XXXXX 1XXXXX 1
Time 120 VALUE VALUE2 VALUE20 1
XXXXXX 1XXXX 1XXXX 1XXXX Strip chart recorder format is typical of most strip chart recorders used in the industry today, i.e.,
a strip chart 6 inches wide with lines and numerals indicating the scale, with red and black pen traces on the chart.
Power sources for the Post Trip Log and the Strip Chart recorders listed above are as follows:
i,
PROCESS COMPUTER UN-INTERUPTABLE POWER (POST TRIP LOG AND SUPPLY (UPS) SYSTEM ASSOCIATED PRINTER)
(FSAR SECTION 8.9)
Strip Chart Recorders Recorder ID 06-LR/PR-97 UPS 06-FR-96 UPS 06-PR/FR-98 UPS 02-3-FR/PR-95 UPS 07-PR-46A, B, C, & D UPS 10-FR-143 Loop A - Safeguard Control &
Instrument BUS Al (71-ESS-A1)
(FSAR) Figure 8.9-1)
Loop B ESS-B1 (FSAR Figuro 8.9.1) 02-3-LR-98 71-ESS-B1 (FSAR Figure 8.9-1) 02-TR-165 Common Control & Instrument BUS 9 (71-AC-9) (FSAR Figure 8.9-1) 02-FR-163 71-AC-9 (FSAR Figure 8.9-1) 94-TS-VP 71-AC-9 (FSAR Figure 8.9-1) 07-R-45 UPS B.
PLANNED CHANGES TO DATA AND INFORf1ATION CAPABILITY The Authority is in the process of reviewing bids on a now computer system including an SPDS, which will ultimately result i
in improvement of the data and information capabilitics at the plant.
Detailed specifications and proposed delivery / installation datos for this now equipment will not be finalized until negotiations with the successful bidder are complete.
The completion dato for the now system will be provided in accordance with our NUREG-0737 Supp. I commitments.
The Authority considers the currently installed Sequence of Event system and strip chart recorders to be adequate for post-trip review in the interim period.,.
r Itca'2.1; EQUIPMENT ~ CLASSIFICATION AND. VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS).
A. CURRENT EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS).
The~ component Quality Assurance Category List'in existence at the' James A.
FitzPatrick Nuclear Power Plant has been reviewed.
The review included. verification that all components in System 5-Reactor Protection System (RPS) are presently classified as QA Category I except for the RPS Motor Generator Sets which are classified as QA Category II.
Components in the RPS System are protected from Motor Generator Set malfunctions, such as over-voltage, under-voltage, and under-frequency conditions, by electrical protection assemblies which are classified Category I.
The QA Category Classification of other systems, such as Reactor Vessel instrumentation (System 02-3), Neutron l.
Monitoring (System 07) and Process Radiation Monitoring l
(System 17) which comprise part of the " Reactor Trip l-Function", has been preliminarily reviewed.
All or'part of these systems are classified as QA Category I,. indicating that those portions of the systems which are associated with the " Reactor Trip Function" are properly classified.
As discussed below (under Planned Changes) additional reviews will be completed to assure that all Reactor
r-Y
' Trip Function componsnts are classified QA Catetory I.
L' The' Authority has performed an initial, preliminary review of the documents, procedures and information handling systems used in the plant to control safety-related activities, including maintenance work requests (work orders), parts replacement and plant modifications.
The documents, procedures and information handling systems
-concerned are controlled under the QA Program, or are-identified as safety related and require review by the Plant 4
Operations Review Committee. This provides assurance that maintenance, parts replacement and modification work is properly classified as QA Category I when required.
The Authority will provide a schedule for a complete indepth review of these items by March 31, 1984.
l~
l A formal Operating Experience Review Program is in effect and includes review of, and response to, General Electric DWR Service Information Letters (SILs).
DWR SILs are used to document recommended changes in equipment and procedures, as well as convey information concerning unique operating conditions and experiences at BWR plants.
The review and implementation of SILs is recorded and fed back to the i
General Electric Company using a standardized SIL Status Response form.
Periodically a SIL Index is issued, assuring that all applicable information has been received.
b -
B.
PLANNED CHANGES TO EQUIPMENT CLASSIFICATION AND
' VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)
The BWROG is' addressing /I, tem 2.1-Equipment Classification and Vendor Interface for Reactor Trip System Components.
A BWROG committee has been formed to specifically address Generic Letter 83-28. The committee has sub-divided Item 2.1 into f
L-several subtasks for ease of handling and in order to allow r <
l generic response to the maximum extent practical.
The subtasks and the status of each are provided below:
Subtask 1
' Reactor Trip System Equipment List i
The BWROG has discussed several options (or alternatives) within this subtask, which have varying degrees of generic T:
application to the individual.BWROG members.
This is due to the difference in plant design such as product line ( BWR 2 i
. through BWR 6) and containment type - (Mark I, II and III).
At
-this. time a review to achieve the most effective use of
'available resources continues.
The' Authority expects a
' decision on'the BWROG course of action to be finalized by February 29, 1984.
Therefore the Authority will update the c
status'of this l' tem by March 31, 1984.
The Authority considers the review of the Component Quality Assurance Categoty List discussed above, to be adequate in the interim.
e I
Subtenk 2 - Updating Opsrntions And Maintenanco i
Manuals For All Safety Related Equipment (RTS L
And Other) l I
f.
The BWROG has discussed this subtask with General Electric.
Some generic benefit can be realized by the BWROG members by l
f grouping the BWR plants into several classifications.
Disc sion as to the best course of action continues.
The Authority expects the BWROG course of action to be finalized by February 29, 1984. Therefore, the Authority will update the status of this item by March 31, 1984.
l Subtask 3 - Description Of General Electric Information Programs The BWROG approved funding during the October 26-27, 1983 meeting for this subtask.
General Electric is expected to splete the description of how information concerning operation and maintenance is obtained from their suppliers, with emphasis in those areas intended to provide support to the plant throughout its lifetime.
General Electric will also provide a description of the SIL system and other communication channels which provide plants with information not encompassed by the SIL system.
This work is expected to be complete prior to February 29, 1984. Therefore, the Authority will provide additional information on this subtask by March 31, 1984.
. \\
Suhtank 4 - Procnduro For Safety Ralated Equipm:nt Identification The BWROG approved funding for this subtask during the October 26-27, 1983 meeting.
General Electric will provide BWROG members with a description of the procedures and the process involved in classification of equipment and components, as safety related and non-safety related, at various levels of equipment complexity.
Work is expected to be complete prior to February 29, 1984. As a result of this effort, changes may be made to the Component Quality Assurance Category List described above.
The Authority will provide additional information on this subtask by March 31, 1984.
a Subtask 5 - Evaluation Of Scope Of Vendor Interface Program (VIP)
The BWROG has deferred action on this substask until the Nuclear Utility Task Action Committee (NUTAC), which is addressing Item 2.2.2, has substantially completed its recommendations.
Due to uncertainty with respect to the ultimate BWROG action on this subtask and the current status of the NUTAC work on Item 2.2.2, the Authority cannot estimate a schedule for Reactor Trip System components vendor interface program until March 31, 1984.
However, the Authority notes that the General Electric information programs discussed under Subtask 3 provide vendor interface for most of the equipment provided within the NSSS vendor scope of supply.
l
.Itca 2.2 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAM FOR ALL SAFETY RELATED COMPONENTS)
A.
Current Equipment Classification and Vendor Inter-face Program for All Safety Related Components.
Equipment Classification
- 1.. During the original classification of components at JAFNPP, the criteria for identifying components as safety related within systems classified as safety related was as follows:
QA Category I Plant systems, or portions of systems, structures, and equipment whose failure or malfunction would cause a release of radioactivity that would endanger public safety.
This category also includes equipment which is vital to a safe shutdown of the plant and the removal of decay and sensible heat, or equipment which is necessary to mitigate consequences to the public of a postulated accident.
The above definition should be interpreted to mean those structures, systems, and components that:
- a. Are necessary to assure the integrity of the reactor coolant pressure boundary.
- b. Are necessary to assure the capability to' shutdown
.the reactor and maintain it in a safe shutdown condition.
l 4
~
1
- c. Ars necoccary to ccouro the capability to provent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of 10 CFR 100.
- d. Contain or may contain radioactive material and whose failure would result in conservatively calculated potential off-site doses which are more than 0.5 rem to the whole body or its equivalent to any part of the body.
These same criteria are presently used for identifying components as safety related and are currently included in the Engineering Design Procedures in use at JAFNPP.
2.
The JAFNPP presently has a Component Assurance Category List which identifies the safety-related components within the systems and is issued and controlled by the Quality Assurance Department at JAFNPP.
.The.'.ist was developed as part of a contract with a consultant.
The consultant had a resident staff at JAFNPP and followed the criteria given in
- 1. above.
In addition to the stated criteria, the consultant'used the following data:
JAFNPP FSAR Plant Drawings provided by the A/E System Descriptions provided by the A/E Instrument Lists provided by the A/E Technical Funuals provided by the NSSS vendor Vendor Manuals and Instructions which were provided by the equipment vendors The consultant 's on-site staf f performed walk-throughs of the plant and verified installation, name plate data, j
rating =, and other information, ao part of the development i
of-the list.' Typical entries on the list are as follows:
Type Category (technical)
Component Description Component Number Data Reference Remarks
_ Quality Assurance Category The lists were compliled and cross checked by the consultant.
They were then transmitted to the Site Quality Assurance Department for review and concurrence.
Members
.of the. Quality Assurance staff reviewed the lists for completeness and accuracy.
Comments were returned to the consultants, as necessary, and when all comments were resolved, the Quality Assurance Department concurred with the lists.
Revisions and/or changes to the list are controlled in accordance with approved plant procedures.
The-list was originally developed in 1978.
Only minor revisions / additions have been made to the list since that time.
As a result, components installed as a result of plant m'odifications which were completed since 1978, are not included.
To overcome these omissions, Quality Assurance personnel perform a review of Plant Modification records to ascertain the correct category when performing the reviews outlined in 3.1 and 3.2 below.
V 3.
A controlled copy of the Component Quality Assurance t-l Category List is issued to the Superintendent of each major plant department which requires the information.
The Quality Assurance Category of plant components is thus readily available to personnel requiring the information.-
In addition, Plant Administrative Procedures require that the Plant Operations Review Committee review Administrative Procedures and documents affecting nuclear plant safety, or impacting on_the environment.
Procedures requiring PORC review are identified by an asterisk ( *) a f ter the title,
.thereby alerting personnel as to which procedures involve safety related considerations.
The following procedures are in effect at JAFNPP which describe the controls and requirements which apply to
. safety related activities:
Administrative Procedures Work Activity Control Procedures Rules of Practice Quality Assurance Program Quality Assurance Procedures In addition, 'he following departments maintain controlled Departmental Procedures which govern the conduct of safety related work:
Operations Instrument & Control Radiation & Environmental Services Maintenance Technical Services l.
Quality Assurance Training In conjunction with a recent commitment to the NRC, the work procedures in use.at JAFNPP for safety related activities are in an on-going process of revision to include Quality Control and j
Radiation Protection Hold Points where required.
In addition, the following controls are used at JAFNPP:
a.
A Work Request Event Deficienty Form (WRED) must be filled out to initiate corrective maintenance.
The WRED is routed through Quality Control which reviews and verifies the Quality Assurance Category of the involved component.
The WRED is also marked to denote if Quality Control Inspection is required.
In addition, for corrective maintenance performed on Safety Related (Category I) components, or other work requiring QC inspection, the use of a Work Tracking Form is required.
This form is used to properly pre-plan. track, control and document corrective maintenance and provides for sign-offs by the department performing the activity, Quality Control personnel and the Operations Department.
b.
The procurement of all materials is initiated by a Purchase Requisition.
The Purchase Requisition must be routed to Quality Control which verifies the Quality Assurance Category of the material, denotes if QC Receiving Inspection is required,1and specifies the required documentation, test reports, etc., which must be included on the requisition.
The requisition is then routed to Quality Assurance which checks and verifies the QC information, includes any further requirements to be r
imposed on the vendor and denotes the method used to qualify.the vendor.
When the Purchase Order for Category I material is typed, the P.O.,
with a copy of the requisition, is routed to Quality Assurance to verify that the requirements of the requisition have been correctly
' entered on the Purchase Order.
c.-
The following documents are required to have a Quality Assurance review and sign-off prior to implementation:
Administrative Procedures Engineering Design Procedures Modification Control Forms
. Modification Documentation Tracking Forms Modification QA/QC and Design Requirements
. Modification Installation Procedures Pre-Operational Tests and Test Results 4.
The management controls utilized to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed, are as follows:
- a. The Plant Operation Review Committee reviews:
- 1) Plant procedures and changes thereto, which affect nuclear safety.
- 2) Proposed tests and experiments that affect nuclear safety.
- 3) Proposed changes or modification to plant systems that affect nuclear safety.
b.
The plant dnpartments utilize an internal departmantal review.
c.
The Quality Assurance Department implements an audit program to provide a comprehensive, independent evaluation of quality related procedures and
-activities to assure that they are in compliance with the Authority's established program requirements.
d.'
The QA & R Department, under the direction of the Safety Review Committee Chairman and the Executive Vice President-Nuc, lear Generation, coordinates efforts to schedule an INPO or a Joint Utility Management Audit Group audit.
(The Authority is a participant in a group of utilities for the purposes of performing independent assessments of QA activities.)
The scope of the INPO or Joint Utility Management Audit Group includes, as a minimum, the activities performed by the Authority's QA and R Department.
In addition, areas outside of the scope may be assigned as directed byEthe SRC Chairman or the Executive Vice President-Nuclear Generation.
The total-audit program covers the 18 criteria of Appenlix B to 10CFR50, within at least a 24 month period.
Vendor Interface The review of General Electric Service Information Letters (SILs) discussed above in the response of Item 2.1 applies equally to systems and components outside of the Reactor Trip System, where those systems or components were within the NSSS Vendor scope of supply.
In addition, Authority review of industry and INPO generated operating experience
~~
documsnts, cuch as Licensee Event Reports and Significant Operating Experience Reports (SOER), is currently being upgraded and will' provide a greater measure of vendor interface for safety related systems and components outside of the NSSS vendor scope of supply.
B.
PLANNED CHANGES TO EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE FOR ALL SAFETY RELATED EQUIPMENT Equipment Classification The BNROG activities discussed in the responses to Item 2.1 above are integrated to some extent with the requirements for Item 2.'2.
Specifically Subtask 4 (procedure for safety
)
related component identification) is equally applicable to Item 2.~1 and 2.2.
As noted above in the response to Item 2.1, the BWROG expects General Electric to complete the work related to safety-related component identification prior to February 29, 1984.
In addition, the Authority is in the process of converting a number of component identification, parts purchtsing, warehousing and other I
functions to a computer based system.
It is expected that the current Component Quality Assurance Category List will be computerized by mid-1985.
Accordingly, a schedule for changes and additions associated with modifications and/or as a result of the BWROG activities discussed above, will be provided by March 31, 1984.
Vendor Interface The Authority is a member of the Nuclear Utility Task Action Committee ~(NUTAC) formed to address Item 2.2.2.
The Authority will review the results or recommendations of the j
l i
NUTAC following its scheduled completion of work in 1
i February, 1984.
Accordingly, the Authority will provide j
I 7 _.
e additional information with ronp:ct to vendor interface by March 31, 1984.
Item 3.1 POST MAINTENANCE TESTING (REACTOR TRIP SYSTEM l
COMPONENTS) l' Item 3.2 POST MAINTENANCE TESTING (ALL-OTHER SAFETY-RELATED COMPONENTS)
A.
CURRENT POST MAINTENANCE TESTING (REACTOR TRIP SYSTEM AND ALL OTHER SAFETY-RELATED COMPONENTS) r 1.
Post Maintenance Testing is addressed for safety-related maintenance at the FitzPatrick plant.
For safety-related work, Work Activity Control Procedure (WACP) 10.1.1 (Procedure for Control of Maintenance) requires testing of equipment following maintenance and, requires that Quality Control be notified of the impending Post Maintenance Test.
Operations Department Standing Order 18 and WACP 10.1.1 provide instructions and guidance to the Shift Supervisor and Quality Control Personnel with respect to test performance, inspection and acceptance, assuring that all safety related equipment is operable t
when it is returned to service.
The requirements of l
Items 3.1.1 and 3.2.1 are fully met.
2.
No formal program exists to verify that vendor and engineering recommendations relating to test guidance is included in test and maintenance procedures at this time. L
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3.
A rovicw of Technical Specification was conducted to i
determine if.any post-maintenance test requirements degraded safety.
The review did not identify any i-l~
cases of obvious safety degradation.
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s B.
PLANNED CHANGES TO POST MAINTENANCE TESTING (REACTOR TRIP SYSTEM AND ALL OTHER SAFETY-RELATED COMPONENTS)
'l.
As noted.above, the Authority considers the f
requirements of Items 3.1.1 and 3.2.1 to be fully l
[
met.
Therefore, no changes are planned.
I l
l[
2.
The Authority plans to implement a program which will incorporate both current and future vendor and/or engineering recommendations, as appropriate, into Test Performance, Maintenance, Post-Maintenance Tests or Technical Specifications.
A major revision of maintenance procedures directed to improving the quality and scope of the procedures, as well as addition of Quality Control and ALARA hold points where appropriate, has commenced.
This program, in response to an NRC inspection, is currently scheduled for completion in mid-1985.
. Incorporation of current and future vendor and/or engineering rebommendations, as well as other potential changes resulting from vendor interface programs, is being considered.
In view of the magnitude of the programs, and the potential impact on
r a
th6 mid 1985 completion commitment, additional evaluation is required before a firm commitment with H
respect to Items 3.1.2 and 3.2.2 can be made.
The F
Authority expects to complete the evaluation'of the impact within the next few months and will provide additional information by March 31, 1984.
3.
As indicated above the Authority is not aware of any post-maintenance requirement in the Technical Specifications which degrades safety.
The Authority notes, however, that BWROG activities associated with improvements in Technical Specifications may at some future date identify recommendated changes to tests required by Technical Specifications.
These potential changes may involve surveillance frequency, allowed out-of-service interval or recommended post-maintenance tests, based on probabilistic risk assessment techniques which consider all or some of the same considerations noted in Item 4.5.3.
ITEMS 4.1, 4.2, 4.3,
& 4.4 REACTOR TRIP SYSTEM RELIABILITY.
The James A.
FitzPatrick Power Planc is a Boiling Water Reactor designed by General Electric.
Therefore, Items 4.1, 4.2, 4.3 and 4.4 are not applicable.
n e
Item 4.5 ' REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)-
1 A.
CURRENT REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING) 1.
Reactor Trip System Reliability is demonstrated by completion of the tests and calibrations required by Technical Specification Table 4.1-1; in conjunction with Technical Specification required control rod scram time testing which periodically demonstrates function and reliability of the entire Reactor Trip System.
Backup scram valves are not required to be tested by Technical Specifications and the Authority
~
does not consider any test necessary.
2.
The Reactor Trip System installed at the James A.
FitzPatrick Nuclear Power Plant is designed to permit on-line testing and such tests are routinely performed to meet the requirements of Technical Specifications g
1.
as indicated in the response to Item 4.5.1 above, except'for the backup scram valves.
The design of the backup scram does not permit a qualitative on-line test.
The BUROG is addressing Item 4.5.2 and funding was approved at the October 26-27, 1983 meeting for General Electric to develop detailed justification for the position that on-line testing of backup scram is not required. _
E-
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3.
The Authority has performed a preliminary review of Technical Specifications to determine if the test intervals specified are consistent with achieving high reliability.
This preliminary review did not address the considerations listed in Item 4.5.3.1 through 4.5.3.5 due to the short time available to evaluate those considerations and due to the activities of the BWROG Technical Specifications Improvements Committee (TSIC).
B.
PLANNED CHANGES TO REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING) 1.
No changes to testing of the Reactor Trip System are planned.
As noted above, the backup scram design at James A.
FitzPatrick Nuclear Power Plant does not permit performance of a qualitative periodic on-line test of the backup scram feature, and the Authority does not consider such a test necessary.
The BWROG position on this matter is expected to provide detailed justification to that end and will be provided by March 31, 1984, following completion.of the General Electric work.
2.
On going work by the BWROG Technical Specifications Improvement Committee (TSIC) is addressing the general subject of surveillance frequency determination.
This work is intended to consider a number of uncertainties, t
f:
^
e.-
hu2cn crrorlprobabilition, component " wear-out" and considerations similar those listed in Item 4.5.3.
The BWROG TSIC recently became aware that the Electric Power Research Institute (EPRI) and the NRC are also h
conducting.or sponsoring work in the same general L'
f area.
As a result, the BWROG TSIC recently I
recommended to the BWROG that funding for development I-and approval of the techniques be defferred until
, additional evaluation of the related EPRI and/or NRC work can be conducted.
j While the Authority is not aware of the date that the NRC sponsored work is expected to be complete it is known that the EPRI work is expected to be complete in the fall of 1984.
Accordingly the Authority cannot l
Provide any meaningful schedule of possible future work =in this area until at leastm the fall of 1984.
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