ML20125D318
| ML20125D318 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/03/1985 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, NED-85-401, NUDOCS 8506120308 | |
| Download: ML20125D318 (42) | |
Text
.-
Georgia Power Ccmpany 333 Pidment Avenue Atlanta, Georgra 30308 Telephone 404 5266526 Marknq Address:
Post Offcc Box 4545 Atlanta, Georgia 30302 GeorgiaIbwer L T. Gucwa tre sout%rn etectrc system Manager Nuclear Engineering
' and Chet Nuclear Eng<neer 0622y June 3, 1985 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 fEC 00CKETS 50-321, 50-366 OPERATING LICENSES OPR-57, NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1, 2 RESPONSE TO REQUEST FOR A00ITIONAL INFORMATION FOLLOWING STAFF REVIEW OF RESPDN5th TO GEPERIC LETTER 83-28 (SALEM REQUIREENTS)
Gentlemen:
Your letter dated April 4, 1985, requested additional information to supplement our previous responses dated November 7,1983 and February 29, 1984.
The requests for information are restated below, followed by the Georgia Power Company response.
REQUEST:
Item 2.1 (part 1) - Incomplete Licensee needs to confirm that the review of the RTS classification program is complete and that it verifles that the RTS components are classified as safety related and identified as such on all documentation and in information handling systems.
RESPONSEt We do not have a system designated "RTS" nor do we have a special classification program for systems required to trip the reactor.
Our safety classification program is described later and encompasses all plant equipment.
Plant documentation for procurement, maintenance, and other plant y
1
1 l
l GeorgiaPower d Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Two activities is not normally identified as " safety related".
Rather, the All designation is "Q", or "NON-Q" to identify quality requirements.
safety related components are considered "Q"
- However, "Q"
does not For necessarily mean that the component or activity is safety related.
example, the main turbine is designated "Q" even though most parts are not considered safety related.
Our review of the Reactor Protection System (RPS), and other systems which input signals to the RPS to trip the reactor, is complete.
We have identified one case, source range neutron monitors (SRMs), where All other some equipment may have inappropriate safety classification.
equipment needed to trip the reactor was properly classified and properly identified as "Q"
on plant documentation.
The SRMs are further review of safety classification and licensing currently under Appropriate classification, based on the classification system basis.
described later, will be applied when the review is complete.
REQUEST:
Item 2.1 (part 2) - Incomplete Licensee needs to submit a description of their program to establish and maintain an interface between all vendors of components needed to perform reactor trip (RTS) and the licensee.
Information submitted shall describe how the program assures that vendor technical information and controlled throughout the plant life; how is kept current, complete, ility the division of responsib between the licensee and vendors who provide test and maintenance services is handled; and verify that lists of vendor technical information and the information itself is available at the reactor site for audit.
RESPONSE
reviewed the concept of establishing an interface with each We have We have vendor which supplies components needed to trip the reactor.
concluded that the administrative burden on CPC and on the various vendors is not justified in every case.
Existing interfaces with the architect / engineer (A/E) and the nuclear steam supply system (NSSS) vendor, along with the Vendor Technical Information Program (VETIP) described in our February submittal provide odequate information to maintain the reliability of the components needed to trip the reactor.
hh. 1
r GeorgiaPower d Director of Nuclear Reactor Regulation Attention:
Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Three The majority of equipment required to trip the reactor was supplied by the General Electric Company (GE).
GPC maintains a very active interface with GE through participation in the BWR Owners Group (BWROG) and through the various technical and commercial interfaces described in our response dated February 29, 1984 A GE operations engineer is assigned at Plant Hatch to help administer the various interfaces. A GE service representative is stationed in Atlanta, and actively interfaces with both the corporate and site GPC staff.
Control of vendor technical information is governed by procedures which provide detailed instruction for receipt, periodic review, revision, and periodic audit of controlled copies.
The document control department has the lead responsibility for maintenance of vendor information.
Each department using vendor information has procedulal requirements governing that use.
All manuals and manual revisions are reviewed and approved by the site Engineering Department or referred to an off-site design group for the appropriate level of review.
thder a new nuclear procurement program implemented in January, 1985, GPC procurement documents and contracts require that vendors submit any changes to technical and quality program requirements to CPC for review and approval prior to shipment. Also, any vendor exceptions to purchase orders must be submitted for prior approval.
The focal point for resolution of changes and deviations is the procurement review section (PRS), which has representatives located at both the plant site and the corporate headquarters.
The PRS is responsible for obtaining approval from the responsible GPC or A/E design organization.
The division of responsibility between GPC and vendors who provide test and maintenance services is handled through the procurement process.
Each requisition for services is submitted to the PRS for approval.
The PRS reviews documents such as inquiries, proposals, contracts and so forth, to ensure that the scope of services is adequately defined, responsibilitics are established, and that sufficient controls are provided prior to contracting with that vendor.
The PRS controls and coordinates the quallrication of vendors at the corporate headquarters Icvel, and maintains a channel of communication to, and obtains input from, the plant site to properly evaluate vendor performance.
GPC performs periodic audits of the QA program of any vendor which provides safety related components or services.
Only quallfled and approved vendors are considered for procurement of safety related components and services.
,wi r,
7-GeorgiaPower d 01 rector of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Four Control copies of vendor manuals are available at the site for audit.
Vendor manuals were examined during an NRC Inspection and Enforcement (I&E) inspection during the week of November 12, 1984.
The inspection results are documented in I&E Inspection Report 84-46.
REQUEST:
Item 2.2.1 - Incomplete Licensee needs to provide a more detailed description of the " Equipment Location Index" to allow better evaluation of compliance with sub-item 2.2.1.2.
The response to sub-items 2.2.1.3 and 2.2.1.4 needs to present more detailed descriptions of the processes and procedures used to classify activities and the management controls used to verify the operation of the information system.
RESPONSE
The Equipment Location Index (ELI) is a computerized list of plant equipment developed by the plant designers to provide a convenient cross-reference of equipment information.
ELI component codes and an example page from the the document are provided as Attachment 1 to this letter.
The information needs for plant operation are frequently different from those of the designer. For this reason, the ELI has not been completely satisfactory for plant activity and material control.
Information requirements for the various plant activities are different and therefore not well supported by one relatively inflexible document.
For example, some information requirements for the procurement activity (ie.
Industry standards, vendor shop and receipt inspection reauirements, etc.)
are different from equipment information requirements for maintenance activities (ie. setpoints, vendor manual information, etc.).
Safety classification is not explicitly stated in the ELI.
The ELI has used the des 10 nation "Q"
to indicate that there are quality requirements, and "N" if there are no specific requirements. All safety related related equipment is "Q"
but a lar0e amount of non-sofety related equipment also has the "Q" designation to indicate a pipinq standard, industry code or other requirement has been specified for that component.
Frequently, a determination of safety classification and related licensing requirements requires consultation with an off-site
'*'"E
GeorgiaIbwer b Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Five design group.
Prior to the issuance of Generic Letter 83-28, GPC had programs proposed or underway to upgrade the procurement program and to improve the availability, control, and use of equipment information.
The programs include a new procurement system described in the response to Item 2.1 (part 2), a large data-base management computer system (described in our February submittal), and an upgrade of the ELI as described below.
The ELI upgrade is based on a detailed fur.ctional analysis of each plant system.
System evaluation documents (SEDs) have been prepared for each of the six systems which function to trip the reactor.
SEDs will be prepared for all important plant systems, with the remaining safety related systems as first priority.
SEDs contain the following information:
o List of drawings, manuals, and other sources of information used to prepare the SED.
o A list of FSAR commitments and 10 CFR 50, Appendix A criteria which apply to each system.
o Detailed description of all modes of system operation related to plant safety.
o A list of functional ties to support systems and other safety related systems, o
A description of safety analyses and design considerations which must be met for plant licensing.
Following development and review of the SED, major compon&ts within that system are listed on Component Evaluation Sheets.is then classified as Saf Each com
, and, if safety related, further designated as to whether the safety function is Active (A) or Passive (P).
For example, a pump may be safety related because its casing is part of a pressure boundary, but pumping function and pumping function failure modes are not safety related.
Under the current system, all parts of the puma includin0 the coupling, motor, power supply and controls are classifted as "Q" and therefore treated as safety related.
Under the new system, the pump would be (S)(P) while the coupling, motor, controls 7..
F A
GeorgiaPower d Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Six and power supply could be classified (N).
Similar examples exist for valves and instrumentation where unnecessary and inconvenient quality requirements can safely be eliminated or modified.
As a result, procurement, maintenance, and operation can be greatly simplified.
Plant safety is enhanced because resources can he focused on components whose function-is in fact safety related.
Quality and regulatory requirements for non-safety related equipment can also be derived from the component evaluation sheets.
An example of a requirement for a non-safety related system is fire protection which is (N) but has quality and regulatory requirements including some criteria similar to 10 CFR 50 requirements.
A draft copy of the SED and component evaluation sheets for the Nuclear Boller (B21) system is provided as Attachment 2.
The B21 system was selected for the example because it provides a good cross section of electrical and mechanical components typical of those found in other plant systems.
Oraft SEDs are complete and under review for all systems whose function is required to trip-the reactor, and are in progress for all other safety related systems.
Once completed, the SED will become a controlled document which will be revised when plant configuration is changed.
The ELI will be kept up to date based on the SED.
Administrative procedures will be used to control the use of the ELI and the procurement system.
We believe that the upgraded ELI will support safety and regulatory requirements, while enhancing the efficiency of plant operation.
If requested, we would be willing to meet with members of the NRC staff to present more detail on this and the procurement upgrade programs.
RE@ JEST:
Item 2.2.2 - Incomplete Licensee needs to present his evaluation of the NUTAC program and describe how it will be implemented at Hatch 1,2.
This program needs to be supplemented because it fails to address the establishment and maintenance of an interface between the licensee and all vendors of carety related equipment to assure that vendor technical information is kept current, complete, and is incorporated as appropriate into plant procedures and maintenance instructions.
The response should also room
l GeorgiaPower d Director of Nuclear Reactor Regulation Attention:
Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Seven address concerns about division of responsibility between licensee and their vendors who provide maintenance or testing services to assure that needed control is maintained over procedures and maintenance instructions.
RESPONSE
GPC participated in the NUTAC along with virtually every other US utility operating or building a nuclear power plant.
The VETIP program described in our February,1984 response is considered an effective way to implement the requirements of Section 2.2.2, without the administrative burden of an interface with every vendor, many of whom are exiting the nuclear business because of excessive requirements placed upon them by their nuclear customers.
GPC has implemented the VETIP program in cooperation with INPO and the other nuclear utilities.
Procedures to ensure that vendor technical information is properly controlled are described in the response to section 2.1 (part 2).
We understand there are ongoing discussions among INPO, the leadership of the NUTAC, and NRC management, concerning the adequacy of the NUTAC approach. We request that acceptance of the GPC position on the subject be addressed within the context of those discussions.
REQUEST:
Item 3.1.3, 3.2.3 - Incomplete Licensee needs to state if he has found any post-maintenance testing requirements for either RTS components or other safety-related equipment that may degrade safety.
If any such are identified, the applicant shall describe actions to be taken including submitting needed Technical Specification changes.
It seems reasonable that the actions called for in these items could be completed in the very near future.
RESPONSE
While not as a result of NRC initiatives, we submitted a Technical Specification revision dated November 7,
1984, to correct certain surveillance requirements we found were degrading the reliability of the emergency diesel generators. We are awaiting approval by the NRC.
We share the concern that certain testing requirements abuse equipment 700775
GeorgiaPower d Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Eight and may be detrimental to reliability.
Due to the high degree of commonality among most BWR Technical Specifications, we have pursued further improvements to the Technical Specifications in cooperation with other BWR utilities in the BWR Owners Group (BWROG).
The improvements being considered in the BWROG program include testing requirements and surveillance intervals.
Due to constraints inherent in the operation of the BWROG, we cannot expect rapid results. However, we believe that the slight delay is more than justified by expected quality and thoroughness of the results, further standardization of the Technical Specifications, involvement of the NSSS vendor, and efficient utilization of our resrurces.
We cannot control the completion dates for a BWROG activity, but we expect to submit Technical Specification revisions based on the generic activity within six months.
REQUEST:
The staff finds that modifications are not required to permit on-line testing of the backup scram valves.
However, the staff concludes that testing of the backup scram valves (including initiating circuitry) at a refueling outage frequency, in lieu of on line testing, is appropriate and should be included in the Technical Specification survel]1ance requirements. The licensee should address this conclusion.
Licensee needs to describe and present the results of their review of existing or proposed on-line testing intervals, the response shall consider the concerns of sub-items 4.5.3.1 through 4.5.3.5 of the generic letter, show how the selected intervals result in high reactor trip system availability, and. present any resulting Technical Specification changes for staff review.
The staff has just received the BWR Owners Group response to item 4.5.3 (NEDC-30844).
If the licensee intends to formally endorse the Owners Group response, the licensee should delay his plant-specific response to Item 4.5.3 until after the staff completes their review of the Owners Group response.
RESPONSE
In response to the conclusion contained in the first paragraph above, GPC does not believe it is necessary to revise the Technical Specification requirements to include the backup scram valves and initiating circuitry.
Initiating circuitry is the same as that which 700775
GeorgiaPowerd Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing June 3, 1985 Page Nine controls the scram
- valves, and is tested per the Technical Specifications as described in our response dated February 29, 1984.
The backup scram valves themselves are not within the licensing basis for Plant Hatch and, therefore, are not included in the Technical Specifications.
Extraneous additions to the Technical Specifications are counter to the NRC and industry goal of simplified and user-friendly Technical Specifications.
In response the the remainder of this item, GPC endorses the BWROG response, and defers a specific response until after tEC review of the generic response.
Please contact this office if you have any further questions or comments.
Very truly yours, fCM-=
L. T. Gucwa PLS/
Attachments as stated xc:
H. C. Nix, Jr.
NRC Regional Administrator Senior Resident Inspector i
s 700??S
A T TA C. H MENT. J.
. - (~')
LEGEND v
EQUIPMENT LOCATION INDEX CODES CONTENTS Part A Explanation of ELI Computer Sheet Columns 1.0 First Column
-Eauipment Location Number 2.0 Second Column Description 3.0 Third Column Vendor Code 4.0 Fourth Column Specification Number 5.0 Fifth Column Requisition Number 6.0 Sixth Column Purchase Order Number 7.0 Seventh Column Six Character Equipment Code 8.0 Column 8-12 Drawing Numbers 9.0 Thirteenth Column Location of Eauipment 10.0 Fourteenth Column Vendor's Instruction Manual S/SX Number Part B Explanation of Codes Used in the ELI 1.0
-Six Character Eaufpment Code
~
1.1 First Character Safety Classification
.rg 1.2 Second Character System Quality Group Classification
\\W 1.3 Third Character Seismic Classification 1.4 Fourth Character Environmental Condition (IEB 79-018) 1.5 Fifth Character Supplier Quality Program 1.6 Sixth Character Surveillance Inspection Recommendation 2.0 Miscellaneous Codes 3.0 Eauipment Drawing Number Code 4.0 Purchase Order Number Code 5.0 New Equipment Associated with PDCR/DCR Response Packages 6.0 Eauipment Location Number 6.1 General Grouping of Systems by Prefix Letter 6.2 6.2.1.
Specific Sqtem Number Safety Related Systems List 6.2.2 Partially Safety Related Systems List 6.2.3 Non-Safety Related Systems Treated As Safety Related (List) 6.2.4 Non-Safety Related Systems List 6.3 Type of Eauip.nent - Hardware Identification Letter 6.3.1 Mechanical Hardware 6.3.2 Instrumentation Hardware 6.3.3 Electrical Hardware 6.3.4 Structures 6.3.5 Special Group 6.4 Item Number 6.4.1-Numbers Assigned 6.4.2 A, 8, C, Etc. Following item Number 7.0 Zone Codes (Included for Reference but deleted from ELI) c Appendix A 9
Vendor Codes - Vendor Alpha & Mfg. Code Alpha Appendix 8 System Quality Group Classification A, B, C, & D Prior to July 1, 1971 (FSAR Table 3.2-2 Sheet 1) after July 1, 1971 (FSAR Table 3.2-2 Sheet 2)
Appendix C Structures, Systems & Component Classifications (FSAR Table 3.2-1)
Appendix 0 Seismic Classification of Structures, Components, & Systems Appendix E Environmental Qualification of Class IE Equipment Appendix F Procurement Supplier Quality Surveillance Inspection Recommendations 4
CEPORT DOTE ' '03/25/85 PACE
.108' BECHTEL' POWER CORPORATION (GPD)
EDWIN I HATCH NUCLEAR PLANT UNIT 2 JOB NO. 6511-20 SORT BY MPL NO.
EQUIPMENT LOCATION INDEX EQUIPMENT V
REQ.
P.O.NO LAY-ELEC LOCATION LOCATION N
SPEC NO.
NO.
DISC & EQUIP P&ID OUT L/0 ELEM CCNN ELE COL INSTR.
NUMBER DESCRIPTION D
SS -
-SS UNIT # CODE H-2 H-2 H-2 H-2 H-2 CO MANUAL H2tP 409 VET PUMP INSTR RACK GE
+ 6019T IA0164 QGSH52
+ 6098 13ORFR15
+
H2tP 410 dET PUMP INSTR RACK GE
+ 6019T IAOt64 QGSH52
+ GO98 13ORFR22
+
H21P 414A HPCI INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RGR24
+
H21P 414B HPCI INSTRUMENT RACK GE
+ 6019T IAOt64 QGSH52
+ 6096 087RGR24
+
H21P 415A MAIN STM FLOW INSTR RACK GE
+ 6019T IA0164 QGSH52
+ 6098 13ORFR15
+
H21P 4158 MAIN STM FLOW INSTR RACK GE
+ 6019T IA0164 QGSH52
+ 6098 13ORFR15
+
H21P 417A RCIC INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RAR14
+
H2tP 417B RCIC INSTRUMENT RACK GE
+ 6019T IAOt64 QGSH52
+ 6096 087RAR14
+
H21P 418A RHR INSTRUMENT RACK GE
+ 6019T IAOtG4 QGSH52
+- 6096 087RLR14
+
H21P 4188 RHR INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RLR14
+
H21P 419 CS INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RLR24
+
H21P 42tA RHR INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RLR24
+
H21P 4218 RHR INSTRUMENT RACK GE
+ 6019T IA0164 QGSH52
+ 6096 087RLR24
+
H21P 425A RHR INSTRUMENT RACK GE
+ 6019T IAOt64 QGSH52
+ 6098 13ORFR22
+
H21P 4258 RHR INSTRUMENT RACK GE
+ 6019T. IA0164 QGSH52
+ 6098 130RFR22
+
H21P 434 HPCI INSTRUMENT RACK GE
+ 6019T IAOt64 QGSH52
+ 6096 087RGR24
+
H21P 435 RCIC INSTRUMENT RACK GE
+ 6019T IA0164 QGSHS2
+ 6098 13ORFR16
+
H2tP 437 RCIC INSTRUMENT RACK GE
+ 6019T IAOt64 QGSH52
+ 6096 087RAR14
+
H21P 501A FIRE PROTECTION RELAY PNL RL 2102-209 45842 12060G QF
+ 1278 3070 3782 3788 1647GT20 X29323 H2tP 5018 FIRE PROTECTION RELAY PNL RL 2102-209 45842 120606-QF
+ 1276 3070
- 3798 130THT20 32360 H21P 501C FIRE PROTECTION RELAY PNL RL 2102-209 45842 I20606 QF
+ 1275 3070
+ 3798 112TATt8 32361 H21P 501D FIRE PROTECTION RELAY PNL RL 2102-209 45842 120606 QF
+ 1279 3070
- 3787 ft2THT13 32362 H21P 501E FIRE PROTECTION RELAY PNL RL 2102-209 45842 I20606 QF
+ 1276 3070
- 3799 130 TAT 18 32363 H2tP 501F FIRE PROTECTION RELAY PNL RL 2102-209 45842 12O606 QF
+ 6417 7442 3785 3785 1328D803 32364 H21P 501G FIRE PROTECTION RELAY PNL RL 2102-209 45842 I20606 QF
+ 6417 7284 3782 3785 13ORER16 32365 H21P SOtH FIRE PROTECTION RELAY PNL RL 2102-209 45842 120606 QF
+ 3072 3072 3780 3786 YARD 32366 H2tP 501d FIRE PROTECTION RELAY PANEL RL 2102-209 45842 I20606 QF
+ 3186 3116 3780 3786 YARO 32366 L45E OO2A MOTOR GENERATOR SET HOIST YT 2102-180 78002 M21368 N X
+
+
+
+
+ 158RJR15 30857 L45E 0028 MOTOR GENERATOR SET HOIST YT 2102-180 78002 M21368 N X
+
+
+
+
+ 158RFR15 30857 L45E 003 MAIN STEAM ISO VALVE CRANE YT 2102-180- 78002 M21368 N X
+
+
+
+
+ BV FIELD 30856 L45E OO4A RDWSTE CENTRIFUGE TROLLEY PK 2102-180 78029 M21370 N X
+
+
+
+
+ 1648D804
+
L45E 004B RDWSTE CENTRIFUGE TROLLEY PK 2102-180 78029 M21370 N X
+
+
+
+
+ 16480804
+
L45E 005 HPCI PUMP AND TURBINE HOIST PK 2102-180 78028 M21371 N X
+ 7271 7271 7024 7988 087RGR24 30270 L45E 006 CRD PUMP HOIST
. YT 2102-180 78002 M21368 N X
+ 7285 7285 7024 7988 087RAR24 30860 L45E OO7A FUEL POOL COOLING HX TROLLEY YT 2102-180 78002 M21368 N X
+
+
+
+-
+ 185RFR22 30858 L45E 0078 FUEL POOL COOLING HX TROLLEY Y1 2102-180 78002 M21368 N X
+
+
+
+
+ 185RFR22 30858 L45E OO8A RWCU NON-REG HX TROLLEY IM 2102-180 78001 M21369 N X
+
+
+
+
+ 158RFR23 30149 L45E 0088 RWCU NON-REG HX TROLLEY IM 2102-180 78001 M21369 N X
+
+
+
+
+ 158RFR23 30149 L45E OO9A RWCU REGEN HX TROLLEY PK 2102-180 78029 M21370 N X
+
+
+
+
+ 158RFR23
+
O O
O
4
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AT TAC HM ENT 2-0 SYSTEM EVALU$ TION DOCUMENT FOR BEORGIA POWER COMPANY FOR EDWIN I. HATCH UNIT 2 O
FOR THE NUCLEAR BOILER (2B21)
SYSTEM PROJECT ENGINEER - BECHTEL PROJECT ENGINEER -
SOUTHERN COMPANY SERVICES _
REV. 8 DATE:
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NUCLEAR BOILER SYSTEM PAGE 1 OF 13 SYSTEM EVALUATION WORKSHEET I.
REFERENCE DOCUMENTS H-26SSS Rev 16 Nuclear Boiler System P&ID Sheet 1 H-26SS1 Rev 12 Nuclear Boiler System P&ID Sheet 2 H-26199 Rev 1
-Nuclear Boiler System P&ID Sheet 3 H-26384 Rev 1 Post Accident Reactor Coolant and Containment Atmos Sampling Sys P&ID H-21838 Rev 16 Cnds & F.W. Bypass Drns Sys P&ID.
H-27458 to H-27464 Nuclear Steam Supply Shutoff System (2A71) Elem Diag H-27465 Nuclear Boiler Process Inst (2B21A)
Elem Diag H-27466 to H-27469 Steam Leak Detection Sys (2B21B)
Elem Diag H-27478 to H-27473
. Auto Depress Sys (2B21C) Elem Diag H-24481 to H-24435 ATTS (2A78) Elem Diag 4
H-27484 Auto Depress Sys (2B21C) Elem Diag (Sheet 5 of 5)
S-25213 Nuclear Boiler Design Spec O
S-25192 Nuclear Boiler Design Spec Data Sheet GEK-45889 Nuclear Boiler Operating and Maintenance Instructions FSAR (Rev 2 7/84)
Chapters 5, 5A, 7 &,
15 Sections 3.2, 3.6, 6.2, 6.3, 18.3, 15 18CFR58 Appendix A General Design Criteria No.
1, 4,
18, 12, 13, 14, 15, 28, 21, 22, 23, 24, 25, 27, 28, 29, 38, 31, 33, 34, 35, 36, 38, 54, 55
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I NUCLEAR BOILER SYSTEM PAGE 2 OF 13 II.
MODES OF SYSTEM OPERATION The nuclear boiler system is physically composed of several components, systems and subsystems eshich are required to support reactor operation.
The functional modes are divided along subsystem lines.
The functions of each mode (and therefore the components of each subsystem) are safety related as indicated in Table 1, and further discussed in Section III.
The general safety design criteria wohich are applicable to components in one or more modes are indicated in Table 2, and further discussed in Section III.
Electrical design considerations and electrical components are covered under the support systems listed in Section IV.
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NUCLEAR BOILER SYSTEM PAGE 3 OF 13 III.
DESCRIPTION OF EACH MODE A.
Primary Steam System Modes 1.
Normal Operation a.
The function of this mode is to conduct steam from the reactor vessel through the primary containmant to the turbine generator.
This mode is not actively safety related.
Four steam lines are utilized between the reactor and a header upstream of the turbine stop valve.
The flow restrictors function to provide flow measurement to the feedwater control system.
Drain lines and a restricting orifice permit continuous draining of the steam line low points.
The inside and outside steam line drains are utilized to equalize pressure across the steam line isolation valves prior to restart following a steam line isolation.
b.
Safety Design Considerations The safety related modes must be testable during normal operation to the maximum extent, practical.
2.
Nuclear Steam Supply Shutoff (NSSS)
The function of this mode is to prevent a.
uncontrolled steam release to the environs.
The primary steam isolation valves function in conjunction with the steam line flow restrictors to prevent core damage and excessive release of radioactivity to the l
environs under assumed conditions of a
(
primary steam line break outside the primary containment.
The steam line isolation valves are closed on any of the following signals:
(1)
RPV low water level 1 (2B21) l (2)
Main steam line high radiation (2D11)
(3)
Main steam line high steam flow (2B21)
(4)
Low turbine inlet pressure (bypassed when reactor mode switch is not in run)
-O (2C71)
(5)
Main steam line pipe chase high temperature (2B21) r
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7 NUCLEAR BOILER SYSTEM PAGE 4 0F 13 III.
DESCRIPTION OF EACH MODE A.
2.
a.
(6)
Turbine building high temperature (2U61)
(7)
Low condenser vacuum (2C71) 1 Each main steam line isolation valve is equipped with two independent position switches for operation into the reactor protection system scram trip circuit when the valves close.
An accumulator, located close to each isolation valve, provides pneumatic pressure for the purpose of assisting in valve closure j
- m. hen both pilots are de-energized or in the event of fmilure of pneumatic supply pressure to the valve operator system.
b.
Safety Design Considerations 4
The closing speed of the main steam isolation valves is slower than that of the turbine stop valves as assumed in the transient analysis.
3.
Overpressure Protection 4'
a.
The function of this mode is to limit the pressure in the nuclear steam generation system.
i The relief valves are the " dual purpose" type
- self actuating and pilot operated.
The valves are self actuating at the set relieving pressure.
The valves permit remote manual or automatic opening at pressures below the setpoint.
Each relief valve discharge is piped to the suppression pool I
with the constant diameter discharge line terminating below the pool water level to permit the steam to condense in the pool when a relief valve operates.
Each relief valve discharge line has a vacuum relief valve to prevent drawing water up into the line due to steam condensation following termination of relief valve operation.
4 Information on primary system response to various l
plant component malfunctions is contained in the plant Transient Analysis Report (FSAR Chapter 15).
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NUCLEAR BOILER SYSTEM PAGE 5 0F 13 III.
DESCRIPTION OF EACH MODE A.
3.
b.
Safety Design Considerations The steam lines and relief valves are designed to accommodate the unbalance forces due to blowing reactions.
The vacuum breakers are qualified for opening and closing loads predicted by the limiting physical configuration.
The analysis of the structural response of the suppression pool to the safety relief valve operation is presented in the Plant Unique Analysis Report.
on the Torus Long Term Program.
See the ADS, ASC and LLS modes for other safety functions of the SRV.
4.
Alternate Shutdoesn Cooling 9 ode a.
The function of this mode is to provide an alternate long term cooling capability in case the normal shutdown cooling path of RHR is not available.
In this mode the RHR or CS pumps take suction i
from the suppression pool and discharge to the RPV.
The fluid exits the RPV through one -
or more SRVs back to the suppression pool.
b.
Safety Design Considerations The main steam lines are designed for thermal, pressure, dead weight and dynamic loads while water filled but seismic loads are not included in the analysis.
The SRV discharge lines are designed for the anticipated trane.ient loads as an upset condition.
The SRVs have been tested for i
operability with water upstream.
5.
Automatic Depressurization Relief valves used for automatic primary system depressurization under assumed loss-of-coolant accident conditions automatically open and remain i
open below their preset closing pressure under j
such conditions.
The remain open signal is based on simultaneous signals from high drywell i
pressure and low reactor water.
By remaining open, the relief valves reduce the reactor pressure to the point where the RHR and/or the core spray systems can adequately cool the core.
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4 NUCLEAR BOILER SYSTEM PAGE 6 0F 13 III.
DESCRIPTION OF EACH MODE i
A.
6.
Low Low Set Logic, manual depressurization, and SRV monitoring.
a.
The function of the low low set (LLS) logic system is to mitigate the effects of postulated thrust loads on the safety relief valve discharge lines and the effects of postulated high frequency loads on the torus 3
shell caused by subsequent actuations of the I
SRVs during a small or intermediate break loss of coolant accident (LDCA).
The LLS design involves four non-ADS SRVs.
The LLS control logic operates the four valves through arming and actuation.
The arming function requires concurrent signals of any SRV opening and a high reactor vessel pressure exceeding scram setpoint.
The LLS system consists of SRV open-close monitors, nuclear boiler pressure i
instrumentation, and a cabinet housing LLS g
logic relays, solenoid valves, and pneumatic supply.
)
b.
The LLS system is designed to:
l i
(1)
Remain operable in event of loss-of-offsito power (LOSP).
5 (2)
Perform its design function assuming the worst postulated single failure.
(3)
Assure that no single failure can cause I
more than one SRV to remain open.
l (4)
Be testable during normal plant
?
operation.
i B.
Feedwater Mode' The feedwater portion of the nuclear boiler system consists of the feedwater piping from the two check valves outside the drywell to the feedwater nozzles l
on the reactor vessel.
The vessel internal piping and feedwater sparger are considered part of the reactor assembly.
The function of the feedwater mode is the convey
'feedwater through the containment to the reactor while it is pressurized during normal operation, operating transients anci accident conditions.
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v NUCLEAR BOILER SYSTEM PAGE 7 OF 13 III.
DESCRIPTION OF EACH MODE B.
Feedwater Mode Under plant conditions when the normal feedwater system is unavailable, there are three sources of feedwater available to the reactor while it is pressurized.
Reactor Core Isolation Cooling System, the High Pressure Coolant In.iection System, and the Control Rod Drive System supply pumps.-
C.
Nuclear Boiler Instrumentation Instrumentation provides process information to the operator and interlock signals to other systems in the nuclear power plant.
i 1.
RPV Water Level Instrumentation monitors water level above the core.
Differential pressure type level i
instrumentation monitors water level below the I
top of the core under post accident conditions.
The functions of the level instrumentation are listed below.
The trips at Levels B, 3, 2, 1,
and 9 are provided by the nuclear boiler system-via the Analog Transmitter Trip System (ATTS).
a.
Level 8
}
This level protects turbine against the occurrence of gross carryover of moisture and and trips the RCIC and HPCI turbines'in the event these systems have been put into operation. The setpoint is near the top of the steam separators.
t b.
Level 3 This level is near the bottom of the separator skirt.
I The react.cr is scrammed and the primary system isolation valves (except the main steam line isolation valves, the main steam line drain valves, and the reactor water cleanup system valves) are closed.
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NUCLEAR BOILER SYSTEM PAGE 8 OF 13 III.
DESCRIPTION OF EACH MODE C.
1.
c.
Level 2 I
The setpoint is high enough so that for complete loss of feedwater flow, the RCIC and HPCI system flow will be suf ficient to prevent initiation of systems at Level 1.
The HPCI and RCIC systems are initiated and the reactor water cleanup system isolation valves are closed at RPV nater Level 2.
Also the Recirculation pumps are tripped, the Standby Gas Treatment system is started and the secondary containment is isolated.
d.
Level 1 This level is high enough to provide time for LPCI and core spray systems to function in the event of the large break. It is approximately 12 to 18 inches above top of active fuel.
C)
At level 1 the RHR, Core Spray,. ADS and standby diesels are initiated.
The MSIVs, the MSL drains and the reactor water sample' isolation valves are also closed.
e.
Level 8 This level is approximately two-thirds of the core height.
After the core has been flooded to this height, one core flood pump cr a core spray pump is more than adequate to ma'.ntain level.
An interlock is provided to the RHR system (Ell) to allow LPCI flow to be diverted for use in the containment spray system.
Levels 4 through 7 are associated with the feedwater control system (C32).
They provide high and low alarms and control range and are selected to minimize moisture carry-over in the steam and steam carry-under in the recirculation flow.
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NUCLEAR BOILER SYSTEM PAGE 9 OF 13 III.
DESCRIPTION OF EACH MODE C.
2.
Nuclear Boiler System Pressure To prevent'escessive depressurization of the a.
i nuclear boiler because of malfunction of the turbine control system, four pressure switches are connected to the header that cross ties the f our steam lines upstream of the turbine stop valves, and automatically close the primary steam isolation valves when system pressure decreases to the switch setpoint.
For information as to how these pressure switches are electrically connected, and how they are l
interlocked to permit reactor startup, ref er to the Reactor Protection System Evaluation.
4 b.
The nuclear boiler pressure instrumentation provides signals via the ATTS system to the following systems.
I i
(1)
Reactor Protection. System (RPS)
(2)
Primary Containment Isolation System (PCI3)
(3)
Core Spray (5)
Recirculation Pump Discharge Valves
~
(6)
Post Accident Monitoring System (7)
Low Low Set Logic-(LLS)
(8)
Safety Parameter Display System (non-safety) (SPDS).
f 3.
Core Flow Instrumentation a.
The core flow instrumentation includes the following:
(1)
Control room readout of the total core i
flow rate and the total discharge flow from each group of jet pumps which is i
driven by an individual drive loop.
Flows are measured using a dif f user j
entrance-to-core-supply plenum pressure differential measurement which.is l
indicated in the control room, for each l
Jet pump unit.
l (2)
Control room readout of fluid temperature in the recirculation pump suction, feedwater, and steam lines (event recall log from the computer).
)
(3)
Control room readout of the total l
feedwater flow rate and cleanup flow
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NUCLEAR BOILER SYSTEM PAGE 10 OF 13 III.
DESCRIPTION OF EACH MODE C.
3.
a.
(4)
Control room readout of flow in each jet pump drive loop (recirculation loop) using the flow nozzles provided.
(5)
Control room readout of the discharge flow from specially calibrated jet pumps.
The diffusers on these jet pump units contain special pressure taps for calibration using prototype test performance maps.
(6)
Locally accessible transducers, and pressure sensing taps for making detailed performance measurements and calibrations during reactor operation for the preceding control room readout equipment.
(7)
The local dif ferential pressure indicators, which are connected to each jet pump branch supply line and to the vessel annulus, may be used to obtain relative supply branch pressure
()
differentials.
b.
The core flow instrumentation is not actively safety related.
4.
RPV temperature monitoring The reactor temperature monitoring subsystem measures the temperature at various points of the reactor vessel in order to map its temperature gradient during startup and shutdown operations.
The reactor temperature monitoring subsystem comprises 18 thermocouples, two junction boxes, two recorders and interconnecting cables.
During startup and shutdown, the temperature recorder is monitored to determine that the temperature rate of change at any of the monitored points and the differential temp erature between the vessels' flange and shell sections does not exceed the predetermined limits given in the vessel specifications.
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NUCLEAR BOILER SYSTEM PAGE 110F 13 v
1 III.
DESCRIPTION OF EACH MODE C.
5.
Steam Leak Detection The leak detection instrumentation is provided to detect and in some cases to, isolate leakage before the results become unacceptable.
A sensor is considered safety related if it is interlocked with an isolation system and non-safety related if it only provides indication and/or annunciation.
The leak detection sensors which are associated with the nuclear boiler mystem are as follows:
Main steam line leak detection temperature transmitters and temperature trip units monitor leakage from the main steam lines and automatically close the primary steam isolation valves on high temperature in the main steam pipe chase.
Temperature elements, connected to a multipoint recorder are provided to detect O
safety and relief valve leakage during reactor operation.
Relief valve temperatur,e elements are mounted, using a thermowell, in the relief valve discharge piping several feet from the valve body.
These elements are also used by the operators as a backup SRV j
actuation indicator.
A line connects to the reactor vessel in the annulus between the two concentric metallic head seals to permit detection of leakage through the inner seal to a collection chamber installed between two solenoid operated valves.
The upstream valve is normally open and the downstream valve normally closed.
When the level switch within the collection chamber actuates an alarm, the contents of the collection chamber can be discharged to the drywell equipment sump and the rate of leakage into the i
f collection chamber measured by timing the i
period until the level alarm is reactivatwd.
A high pressure alarm is provided to aid in determining the relative amount of seal leakage.
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NUCLEAR BOILER SYSTEM PAGE 12 OF 13 III.
DESCRIPTION OF EACH NODE D.
RPV Bottom Draining A drain line is. connected to the bottom head of'the reactor vessel between control rod drive thimbles to permit flushing accumulated crud from the bottom of the reactor vessel to the clean radwaste collection point during plant shutdown.
The drain is piped to the reactor water cleanup system to permit a small amount'of continuous flow to keep the drain line flushed out.
The temperature of the drain line is monitored to determine the water temperature in the bottom head region.
The recirculation system is operated in a manner as to prevent large temperature differences between the water and the bottom head metal.
E.
RPV Head Venting A vent connection is provided on the top head of the reactor vessel to permit venting the reactor vessel during filling for hydrostatic test and to permit remote venting of noncondensibles which may have accumulated in the vessel head space during reactor O
cooldown after the steam lines have been flooded. The vent line is connected within the drywell to one of the main steam lines to remove noticondensible disassociated gases which accumulate in the vessel head space during reactor operation.
F.
Reactor Water Sampling A reactor water sample line is connected to the reactor recirculation system for use in the event that the reactor water cleanup system is out of service.
When the reactor water cleanup system is in operation, samples of reactor water for laboratory analysis are drawn through sample lines connected to it.
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h NUCLEAR BOILER SYSTEM PAGE 13 0F 13 IV.
SUPPORT SYSTEM FOR EACH MODE A portion of the following systems are required to support the scram mode of the Nuclear Boiler System.
For detailed,information pertaining to what portion of the below system are functionally safety related, the system evaluation sheet for each system should be consulted.
,i A.
125/259 VDC Station Battery and Distribution System -
J 2R42A.
B.
Reactor Protection System Power Supply - 2C71B C.
Emerg. Station Service 128/288 VAC system - 2R2SN D.
Analog Transmitter Trip System - 2A7s E.
Nuclear Steam Shut off System - 2A71 F.
Drywell Pneumatic System - 2P78 i O I
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I 1
PROVIDE THE CAPABILITY I
I I
I PROVIDE THE CAPABILITY !
TO PREVENT OR MITIGATE I
I MODES OF SYSTEM t
FORM PART OF tTO SHUTDOWN THE REACTOR I THE CONSEQUENCES OF I
I OPERATION ITHE REACTOR COOLANTI AND MAINTAIN IT IN A I
ACCIDENTS WHICH COULD I
I I PRESSURE BOUNDARY ISAFE SHUTDOWN CONDITION I RESULT IN POTENTIAL I
1 I
I I OFF-SITE EXPOSURE EXCEED *G I
I I
I 10CFR19e LIMITS I
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IA. Primary Steam Sys Modes:
I 1
1 I
-I I
I i
1 1.
Normal Operation Yes No I
No I
I 1
-1
- 2. Nucir Stm Supply ShutoffI Yes t
Yes t
Yes I.
1 I
I I-t I
3.
Overpressure Relief I
Yes t
Yes t
Yes I
1 1
I I
I I
- 4. Alt Shutdown Cooling 1
Yes t
Yes t
Yes I
I I
-t I
I 5.
Auto Depressurization Yes Yes t
Ye.s i
I I-I t
I
- 6. Low Low Set Logic Manual!
I t
i 2
1 depressurization and I
Yes t
Yes t
Yes I
I SRV monitoring I
I I
I I
I I
I ID. Feedwater System I
Yes t
Yes I
Yes I
l--
I I
I
-I IC. Nuclear Boiler Pressure I
I I
I I
-8 I
I I
1.
RPV water level Yes t
Yes Yes I
I I --
t I
I
- 2. Nuclear Boiler Pressure I Yes t
Yes t
Yes I
I I
I
- 3. Core Flow Measurement t
Yes 1
No I
No I
t
-I
-t I
I I
- 4. RPV Temp Monitoring I,
No I
No I
No I
1
--t I
I t
- 5. Steam Leak Detection I
No I
No 1
Yes I
-1 I
I 1
ID. RPV Bottom Draining I
Yes No I
No 8 '
I I
I I
t IE. RPV Head Ventinq Yes No I
No I
~ O O
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PROVIDE TE CAPABILITY I
1 1
1 PROVIDE TE CAPABILITY I TO PREVENT OR MITIGATE I
I MODES OF SYSTEM t
FORM PART OF ITO SHUTDOWN TE REACTOR I TE CONSEQUENCES OF i
1 OPERATION ITE REACTOR COOLANTI AM MAINTAIN IT IN A 8
ACCIDENTS NHICH COULD I
I
! PRESSURE BOUNDARY ISAFE SHUTDOWN CONDITION I RESULT IN POTENTIAL t
t I
I OFF-SITE EXPOSLRE EXCEED *G I
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TABLE 2
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PORTIONS OF COMP WITHIN I
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COMPONENT IALL PIPING AND I PIPING IS ITHE DESIGN ARE ITHIS MODE FALL I I
OPERATION I INSTALLATIONS I COMPONENTS
! EVALUATED TO I
REQUIRED TO I
UNDER THE t
I SHOULD BE I
SHOULD BE I
DETERMINE I
MEET THE IAUSPICES OF THEI I
I EVALUATED FOR I SEISMIC 8 EFFECT OF A i 18CFR58 APP R tREQUIREMENTS OFI I
I SEISMIC II/I I
CATEGORY I I
DESIGN I
19CFR50.49 8
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- 1. Normal Operation Yes I
Yes t
Yes Later I
Yes I.
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1 I
I; I
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Yes Yes I
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I I
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Yes t
Yes Later I
Yes I;~
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O. Alt Shutdown Cooling!
Yes t
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Yes I
t I
I I
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Auto Depressurizatin!
No 8
No I
Yes Later i
Yes I
t
-t I
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I*
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Yes Yes I
Yes I
Later I
Yes I
and SRV monitoring I
I I
I.l 8
I I
I I
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- 10. Feedwater System I
Yes I
Yes I
Yes Later
.I Yes I
1 I
I I
-t IC. Nuclear Boiler Pressure!
1
-I
-1
-I t
1.
RPV water level Yes t
Yes Yes I
Later 8
Yses I
I I-I I
I I
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Yes I
Yes Later Yes I
8 1
I I
I
- 3. C re Flow Measuremnt!
Yes t
Ys.s No I
Later I
No I
9 I
I I
I I
I O. RPV Temp Monitoring I No No I
No I
Later I
No i
I I
I I
I I
I I
- 5. Steam Leak Detection!
Yes Yes Yes I
Later Yes I
I I
I I
i i
1.
ID. RPV Bottost Draining I
Yes Yes No I
Later I
No I'
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OPERATION I INSTALLATIONS 8 COMPONENTS I EVALUATED TO I
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19CFR59.49 8
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I 54 i FM74, B IEscoes Fles Check Val w llest. Line 15 A I N-26001 IPrieary Centaissent I
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I 35 I FMen, B It is. BIshe Valve Illaistenance 15 P I N-2W91 IPressere Boundary Only i
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IH I Ff57 IEscoes Fles Check Valve!!sst. Line 15 f.
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I 65 IFf504-N,E-N,P-Ull is. Olete Valve Imaintenance I5 P I F2Hel IPressere Boundary Only i
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I 66 IFf594-5,K-5,P-lilEscoes Fles Check Valveliest. Line 15 A I F26001 IPrieary Centaisesnt i
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I 67 i FSH 11 is. Blobe Valve Isaintenance 1 5 P I N-26ffl IPressere Isaadary Only i
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i I 74 I F06M - B 11 is. Blabe Valve INaintenance 15 P I F26000 IPressure Boundary Only I
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I 75 i Ff7N - B IEscess Fles Check IFles Limiter i 5 A I 11-26800 IPrimary Centaineont I
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I 78 i Ff73A - B IEscess Fles Check IFles Lleiter 15 A I ll-26000 IPriearyCantaineant i
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I 79 I Ff744 - B 1314 in. Siehe Valve IFeeduster Brain 15 P l N-26fff IPresswa Boundary Only I
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I 84 i FN24, B 1314 in. Elabe Valve IFeedsater Brain 15 P I F26fff IPressure Boundary Only i
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I B5 i FN3A, B 1314 in. Globe Valve IFeedseter Check Valve Brain i5 P i F26000 IPressure Boundary Only i
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I 86 i FN4A,B 13/4 in. Elabe Valve IFeedsater Check Valve Brain 15 P 1 F26000 IPressure Boundary Only i
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I 87 i Ff054,8 1314 in. Elsbe Valve IBrain iN P I F 26800 IPressure Boundary Only i
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I 94 I Ff92A,B 1314 in. Globe Valve liest. Line Brain 15 P I 11-26001 IPressure Boundary Baly I
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I I 91 i FH54 - B 1314 in. 51obe Valve IInst. Line Brain 15 P 1 U 26ffl irresswa Boundary Baly 1
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I 99 I FH7 13/4 in. Blabe Valve linst. Lise Brain 15 P 1 N-26ffl IPresswe Boundary Bely i
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I IN I Ff90 1314 in. Blabe Valve IInst. Line Brain 15 P l 16-26801 IPressure Boundary Baly i
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i 101 IFft9dHl,K-il,P-ill3/4 in. Blobe Valve liest. Line Brain 15 P l 16-26801 IPressure Boundary Bely i
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i 102 i Fiffe,I It in. Globe Valve IFeeduster Check Valve 15 P l11-26000 IPressure Boundary Inly I
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