ML20098A787
| ML20098A787 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/06/1984 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20098A764 | List: |
| References | |
| TDR-598, NUDOCS 8409250243 | |
| Download: ML20098A787 (44) | |
Text
,
TDR NO.
598 REVISION NO. O g
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I PAGE OF TECHNICAL DATA REPORT '
ITY NO.
PROJECT-TMI-l DEPARTMENT /SECTION E&D/rnNSill TING ENG.
EQUIPMENT QUALIFICATION FOR SBLOCA RADIATION ENVIRONMENTS RELEASE DATE REVISION DATE l DOCUMENT TITLE:
METHODOLOGY AND LIST OF ELECTRICAL COMPONENTS REQUIRING RADIATION QUALIFICATION FOR SBLOCA MITIGATION DATE DATE APPROVAL (S) SIGNATpE CRIGigTOR SIGNATURE
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DISTRIBUTION ABSTRACT:
Statement of Problem:
GPUN had been requested by the NRC (Reference No.1) to R. F. Wilson review and affirm the completeness of the list H. D. Hukill (Reference No. 3) of electrical components which are D. K. Croneberger required to mitigate postulated small break T. G. Broughton loss-of-coolant accidents (SBLOCA) with a loss of main R. J. Toole feedwater and which are. located in a harsh radiological J. J. Colitz environment. The list of such components are to be in G. R. Capodanno compliance with 10CFR50.49(b)(1) and (b)(2) utilizing M. O. Sanford the applicable guidance contained in NRC's May 25, 1984 R. J. Chisholm letter (Reference No. 2), for a SBl.0CA with postulated N. G. Trikouros radiation levels associated with large break LOCAs in it. J. Ross accordance with the DOR Guidelines.
L. C. Lanese G. R. Braulke The purpose of this report is to document the R. T. G1aviano methodology and basis for the affirmation of J. S. Wetmore completeness of the electrical component list which L. W. Harding meets the forestated requirements.
Common type electrical components such as power, control and instrument cable, terminal blocks, electrical penetration assenblies, and heat shrink tubing will be qualified and need not be addressed by this report.
Summary of Key Results The list of electrical components submitted by Reference No. 3 is complete with the exception of the following:
1 - New instrumentation which was installed, subsequent to the referenced submittal, for the purpose of segregating the indicated plant parameters from the 8409250243 840907 ICS/NNI systems.
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g TDR No. 598 Page la 2 - lieplacement of a component and assigning it a new identifying number (PS-745 A, B, C).
3 - The reactor incore themocouples were recognized to be required for SR.0CA mitigation as a result of the development of the B&W AT0G document.
4 - The pressurizer spray valves wnich may be required for long term boron precipitation control.
Appendix "A" provides a complete list of electrical components,which are required for SBLOCA mitigation and are located in a postulated harsh radiological environment.
Oxnponents requiring qualification for these conditions are identified.
Conclusion This report providetthe methodology for electrical component identification and basis for affirining the comaleteness of the list of electrical components to be eva"uated for their radiation qualification in accordance with the stated requirements.
liscocunendations 1 - It is recocunended that this report be reviewed by TitI-1 operations in order to ensure'that the operators recognize the qualifications and ifmitations of the components which may be called upon to mitigate a postulated SBLOCA.
2 - Specific procedural guidance must be incorporated to justify not qualifying tiV tank level indication (fiV-14-DPT) HPI flow indication itV-23-DPT 1, 2, 3, 4) and seal injection flow indication (11V-42-DPT). Refer to specific instruments ( Appendix " A" and Appendix "C" for guidance.
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n TDR No. 598 hge 2 s
Table of Contents
1.0 INTRODUCTION
2.0 ET)CDS 2.1 Definition of Scope and Assumptions 2.2 Systems Identification and $10CA Scenario 2.3 Reactivity Control 2.4 Containment Integrity and Radiation Release Control 2.4.1 Containment Isolation 2.4.2 Prevention of, Radioactive Release to the Environment 2.4.3 containment Heat Removal 2.5 Reactor Core Cooling and Heat Removal Electrical Systems Interaction 2.6 3.0 RESULTS
4.0 CONCLUSION
S 5.0 RECOMIENDATIONS
6.0 REFERENCES
APPENDICES Appendix No.
Title A
Assessment of Electrical Components Requiring Radiation 4a11fication for SBLOCA Hitigation B
Abnormal Transient Procedure Logic for a SK0CA C
Astification for not Requiring Radiation @alifying MU-23-DPT-1, 2. 3 & 4 or MU-42-DPT
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TDR No. 598
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Page 3
1.0 INTRODUCTION
The UCS contention regarding the radiation qualification levels to which electrical equipment located in containment and auxiliary building whose operation is necessary to mitigate small break loss of. coolant accidents (SBLOCA) and loss of main feedwater transients was ruled to be a valid contention by the Atomic Safety and Licensing Appeals Board (ASLAB).
The contention essentially stated that such electrical equipment must be qualified for radiation levels associated with large break LOCAs in acconiance with the Division of Operating Reactors (D0R) guidelines.
As a result of the ASLAB ruling, the NRC Comissioners directed the NRC staff to certify that the THI-1 electrical equipment which is required to mitigate such an event is qualified with respect to radiation in accordance with the D0R guidelines. GPUN has been requested (Ref. No.1) to review the list of equipment falling within the scope of the NRC Comission's onier, using the applicable guidance from the Commissions (Ref. No. 2) May 25,1984 letter to GPUN and prepare for an audit of the environmental qualification files for such comnonents.
Tnis report provides the basis and list of electrical equipment falling within the scope of the NRC Comission's onfer, utilizing the applicable guidance of the May 25,1984 letter.
Appendix "A" of this report is a list of electrical components'which are required to mitigate such an event and are located in areas which are postulated to provide a harsh radiological environment. The list identifies the components requiring qualification for harsh radiological environments and the degree of radiological exposure for which such qualification is required.
ort in conjunction with the remarks in Appendix " A" The body of this rep logy,used to identify components that must be provide the methodo environmentally qualified.
2.0 ETH00 The methodology utilized to generate the list of safety related electrical equipment as defined in paragraph (b) (1) of 10CFR50.49 which is required to remain functional during or following a SBLOCA with a loss of main feedwater, to mitigate the event, included a review of the following THI-I documents:
1 - Final Safety Analysis Report (FSAR) 2 - Technical Specifications & Manuals 3 - Normal Operating & Abnornal Transient Procedures
TDR No. 598 Page 4 4 - System Flow Diagrams 5 - Piping Drawings 6 - Electrical Distribution & Elementary Wiring Diagrams The foregoing documents identify the systems, the system operating modes, auxiliary support systems, and components which are required to perfonn the following mitigating functions following a SBLOCA with a loss of main feedwater.
1 - Emergency Reactor Shutdown (Reactor Reactivity Control) 2 - Containment (Reactor Building) Isolation 3 - Reactor Core Cooling and Heat Removal 4 - Containment (Reactor Building) Heat Removal 5 - Prevention of Radioactive Material Release to the Environment Each of these functions is discussed individually and in greater detail in subsequent sections.
2.1 Definition of Scope and Assumptions A [ mall break loss of coolant accident (SBLOCA) shall be defined as a break within the reactor coolant system (RCS) boundary which does not exceed the flow area of the largest RCS connecting pipe e core flood (DF) nozzle on the cross sectional area, namely pbreak is the largest break which reactor vessel. This 0.44 f t does not exhibit large break LOCA symptoms (i.e., continuous depressurization to LPI injection).
The control room operator is assumed to respond to the event in accordance with the Abnomal Transient Procedures (ATPs) and operator training. Subsequent to the initial period of the event, the resources of the emergency response organization are available to provide any required assistance.
The radiological environment which has been utilized for this evaluation is presented in TDR-282 and TDR-121 (Ref. No.'s 4 & 5) with equipment locations identified by the same area designations.
As a result of the postulated event and the defined limits for a SBLOCA, the reactor building (RB) pressure does not reach 30 psig. Thus, the reactor building spray system is not initiated, nor is a 30 psig RB pressure containment isolation signal generated.
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TDR No. 598 Page 5 The defined SBLOCA event does not result in core fuel failure or conditions which would result in hydrogen generation in excess of 45 by volume. 1hus, the postulated radiological conditions are conditions for environmental qualification of electrical l
components which would be required to mitigate such an event while such environmental conditions existed.. Therefore, it is assumed that in acconlance with the criteria for containment isolation, the radiological conditions would not permit the operator to re-open containment isolation valves once they have i
been closed and that the hydrogen recombiner system will remain isolated during the event.
i The radiologicaiconditions in the Auxiliary Building are assumed to be the same as the nomai operating conditions until the borated water storage tank (BWST) has been depleted of its inventory and the decay heat removal pump suction is switched to i
the reactor building sump. The plant ATPs require specific j
operator actions to align system valves and system operating modes in preparation for R3 sump recirculation.
Concurrently, electrical components within the auxiliary building which will not be required to chan;e state (i.e., motor operated valves)
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after RB sump recin:ula41on has been initiated will be properly I
aligned prior to commencing such an operating mode.
t after RB The radiological environment in 'the Auxiliary Building,itted by circulation has connanced, is based on radioactivity em 4
the recin:ulated fluid within the piping systems.
It is assumed that the integrity of these piping systdms is maintained during t
this event with no rnfological releases within the Auxiliary i
Building.
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The postulated loss of main feedwater either concurrently or l
independently of a SBLOCA does not affect the methodology and i
list of mitigating systeWcomponents requiring electrical l
component qualification since the mitigating EFW for either event is addressed by the SE0CA mitigating systems. Thus, the event 4
will henceforth be identified only as a SBLOCA.
i Implicit in this evaluation, is the assumption that there may also exist a concurrent loss-of-offsite-power (LOOP) and that redundancy of safeterelated components be provided for possibic single failure of such components.
l Appendix "A" of this report does not list all safety related j
auxiliary equipment which are required to perform a safety function during or following a $10CA, but which are clearly identifiable by inspection of the plant general arrangement i
drwings to be located in a non-harsh environment. These itarms 1
are contained in the auxiliary support systens identified in 2
Section 2.2.
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4 TDR No. 598 Page 6 2.2 Systems Identification and SBLOCA Scenario The systems required for mitigation of a SEOCA were identified by a review of the plant Abnormal Wansient Procedures (ATPs) in conjunction with an understanding of the symptoms associated with a SBLOCA and a review of the containment isolation system.
The 510CA cannot be defined by a single scenario.
It can, however, be defined by associated symptoms. The TMI-1 ATOG (Raference No 8) described both the plant response for various LOCAs and the mitigating systems. The ATPs are likewise designed to address correct operator actions in a prioritized fashion to the plant response symptoms rather than specific scenarios.
Since the ATPs address symptoms and responses for all postulated events, extractions were made from these procedures of the steps appropriate for the SLOCA. These steps are presented in Appendix "B" as a procedural logic for SBLOCA mitigation.
This logic provides for the identification of the general systems and some specific components which are required in order to perform the SEOCA mitilaEing functions identified in Section 2.1 The supporn auxiliary systems, the system interfaces, the required instrumentation and electrical auxiliary devices are identified by a review of the specific system flow diagrams, component specifications, manuals and drawings and the elementary wiring diagrams.
The following systems provide for the performance of one or more of the required mitigating functions:
1 - Reactor Protection System 2 - Ptaclear Instrumentation (Neutron Monitors) 3 - Engineered Safeguanf s Actuation System 4 - Containment Isolation System 5 - Reactor Coolant System Instrumentation 6 - High Pressure Injection (HPI) System 7 - Core Flood System 8 - Low Pressure Injection System 9 - Emergency Feedwater and Atmospheric Steam Dunp Systems 10 - Reactor Building Dnergercy Cooling System 11 - Vital and ES Electr c Power Systems The cooling water support systems are generally located in noreharsh radiological areas.
Components within these cooling water systems which are not clearly identifiable to be either l
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TDR No. 598 Page 7 located in a non-harsh environment and/or to be non-electrical in nature were. listed in Appendix "A".
All other such auxiliary support systems including the river water supply systems are located in noreharsh environments.
This includes all of the electrical components in the following systems:
l 1 - Decay Heat Closed Cooling Water System 2 - itselear Services Closed Cooling Water System 3 - Intemediate Closed Cooling Water System 4 - Auxiliary Building HVAC System j
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5 - Control Building HVAC System 6 - Vital and ES Power and Control Systems and Relays These mitigating and mitigating support systems encompass all of the systems required to attigate a SBLOCA.
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2.3 Reactivity Control Reactivity control is obtained by the insertion of the reactor control rods and injection of borated water.
Verification of the reactor control rod insertion and reactor shutdown is accomplished by the incore intemediate range power monitoring system.
The reactor control rods are inserted by'the Reactor Protection System (RPS) as a result of abnomal RCS pressure and/or temperature conditions or directly as a result of a loss-of effsite-power (LOOP). A SBLOCA would cause abnomal pressure / temperature conditions that would actuate the RPS to trip the reactor. A reactor trip signal removes power from the control rod drive mechanism, as would a LO@, and would allow these rods to fall to the reactor shutdown position.
The punose of these trip signals is to prevent the fuel from experiencing a departure from nucleate boiling (Die) transient.
Therefore core fuel failure would not yet have occurred and thus, the function of the RPS would have been complete prior to its components having been subjected to a harsh environment.
Therefore, the RPS components meet the criteria of 10CFR50.49 for exception and do not need to be environmentally qualified.
Likewise, the incore intemediate range power monitoring system perfoms its function prior to being subjected to a harsh a
l environment and is excepted for any'ronmental qualification.
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V I
i TDR No. 598 i
Page 8 Boron injection during a 58LOCA is accomplished by the LPI and/or HPI systems which inject water from the borated water storage tank into the RCS. The LPI and HPI systems are discussed and addressed in Section 2.5.
2.4 Containment Integrity and Radiation Release Control 2.4.1 Containment Isolation "
The containment isolation system has been desiped on the basis of diverse and redundant isolation si@a"s. The 1
initiating signals which are anticipatory of postulated fuel failure are the reactor trip and 1600 psig RCS i
pressure isolation signals. Most of the containment isolation valves receive one of these isolation closure p
i signals. These valves would then be closed prior to being subjected to the harsh postulated radiation from a SE.0CA. These valves would remain closed during the remainder of the ev.pnt unless a failure of their electrical components would result in re-opening the j
valve.
In order to satisfy the single failure criteria, these isolation valves may not be allowed to re-open due i
to such failure.
Since motor operated valve control l
systems are located in nort harsh environments, motor operated valves will fail in an as-is position due to l
radiation. Solenoid operated valves are assumed to fail
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to a de-energized state.
Since some containment isolation 7
j valves are air operated and theit air supply is controlled by a solenoid valve, the failure mode of the isolation 4
valve was examined in response to an assumed solenoid i
valve failure.
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All of the powered containment isolation valves contain direct position indication in the control room.
Failure i
of the isolation valve position indication af ter the j
valves have been initiated and indicated closed on the containment isolation status board is deemed to be j
acceptable since their function has been completed and at J
the given level of radioactivity in the reactor building
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would not have been re-opened e' ther by the operator or i
due to the failure of the valve operator.
I The intermediate closed cooling water (IC) system and t
I nuclear services closed cooling water (NS) system containment isolation valves would not necessarily receive t
isolation signals which are anticipatory of postulated high radiation levels.
These valves would be initiated to close if their respective piping systen inside of the i
TDR No. 598 Page 9 containment did not remain intact as a result of the SR.0CA. Their isolation signal is generated by a concurrent 1600 psig RCS pressure signal and a low surge tank level signal. Their respective surge tank inventory may not be immediately depleted to the low level set
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point, however, their isolation signal would be initiated within two (2) hours of..the SBLOCA event. Thus, they must be functional at that time and must, therefore, be radiologically qualified.
I The reactor coolant pumps (RCP) seal return ' containment 1 solation. valves MJ-V-25 and 26 do not receive an isolation signal which is anticipatory of postulated high i
radiation levels. These valves would be isolated by the 1-operator in response to a high radiation monitor alam on the seal return piping in the Auxiliary Building. The RCP i
seal return water remains relative 13 clean while the seal injection water source is from the BWST.
If seal injection is not avaliable due to the failure of the supplying HPI pump or the switch to RB sump recirculation j
has been made, then the radiation monitor will alam I
before the monitor is outranged and fails due to radiation. Since the inside containment isolation valve will have been subjected,to a harsh postulated radiological environment and the outside valve will have i
been exposed to the effects of the contaminated fluid, i
j these valves must still be quali(ied to function and l
provide their respective closed position indication.
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All containment isolation valves, including their position l
indicating devices, must as a minimum, be qualif ted for their nomal harsh radiological environment.
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Specific remarks regarding the qualification needs for j
aach containment isolation valve are provided in Appendix "A".
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2.4.2 Prevention of Radioactive Release to the Environnent l
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Prevention of radioactive material release is provided by i
l properly maintaining containment integrity.
Containment i
i intepity is provided oy ensuring that the reactor building pressure and temperature limits are not exceeded 1
and containment iso *ation is implemented.
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Minuent isolation has been addressed in Section 2.4.1 and is accomplished by the containment isolatfon system I
and operator guidelines, designed to prevent releases, for l
re-opening isolation valves, j
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ge The containment heat removal function to mitigate the energy release within the reactor building and prevent a
- ise of pressure / temperature within the building is addressed in Section 2.4.3.
The energy release due to the largest SBLOCA only result in RB pressures less than 30 psig and temperatures below the RB design temperature.
The pressure response of the Tft-1 reactor building was evaluated using the CONTDFT computer code and blowdown dets for a.0.5f tZ LOCA as presented in Table 6.2-22 if the TMI-2 FSAR. This break size is at the low and of the large break LOCA spectrum. This analysis results in a peak prqssure of 31 psig.
The pressure response for the 0.44 f t' core flood line break will be less than 30 psig because: 1) the core pressure level at TMI-1 is only 2535 MW(t); 2) the break area is 125 smaller than the analyzed break (hence break flow will be reduced); and 3) the CF line break results in RCS pressure being held up, so that the stored energy in the RCS is releaged over a much longer period of time than the 0.5 f t' LOCA.
Hydrogen release, in excess of 4% by volume, has not been postulated since a mitigated SBLOCA will not result in fuel failure.
Thus, containment integrity is maintained and provides the required prevention of radioactive material release to the environment.
2.4.3 Containment Heat removal The reactor building emergency cooling system is initiated by the anticipatory 1600 psig RCS pressure ES signal.
With the exception of the RB Fans located inside of containment, as listed in the miscellaneous section of Appendix "A", all other electrical components required to operate and control this system to rovide the required RB cooling function are located in eitter the Intemediate Building or the Intake Structure Building which do not contain a harsh radioactive environment. Thus, the only components which require qualification for the postulated harsh radiological environment are the RB fan motors.
2.5 Reactor Core Cooling and Heat Removat The SR.0CA consists of a break spectrum as described by the TMI-1 AT0G (Reference No. 8), which produce definable symptoms.
The symptom which is common to the entire spectrum is the loss of
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6 TDR No. 598 Page 11 l
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reactor coolant inventory. Since the reactor coolant is the primary medium for reactor heat transfer, sufficient inventory must be maintained in the reactor core to absorb the core decay heat and to provide for a means of removing this heat from the 4
reactor core region.
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The systems which provide makeup water injection to the RCS water inventory arc the high pressure injection (Hp!) system, core system and the "ow pressure injection (LPI system.
flood (CF) tion systems in< tially take borated water Won the i
Deseinjec i
BWST for their injection supply. The berated water also performs l
the secondary function of reducing the core reactivity.
Once the BWST inventory is depleted, these injection systems are re-aligned to draw water, which has exited the break, from the RB L
sump. The HP! pumps do not draw their water directly from the RB sump at this point. but rather receive their supply from the LPI pump discharge. This mode of operation is called 4he " piggy j
back" mode.
I The means available for~dicay heat removal and core cooling are j
dependent on the SOLOCA symptoms. Therefore, the system that provides this function will also depend on the symptoms of the SBLOCA.
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The TMI-1 ATOG (Reference No. 8) described both the plant
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response for various SOLOCAs and the mitigating systems for core heat and RC5 heat removal. These guidelines also address mitigating actions beyond the single fatlure criteria.
Such additional actions have not been addressed by this report on the i
l basis that they exceed the stated requirements.
l In order to identify the electrical components required to i
perfom or prevent the perfomance of the core cooling and heat i
removal function, we have reviewed the ATPs, operating procedures, system flow diarams, training material and elementary wiring diaFans for components that are called upon to function, system interfaces and control logic that maybe in a harsh radiological environment and could interfere with the i
perfomance of the function if it failed in any given state.
The components that were thus identified and evaluated are shown in l
Appendix "A".
The instrumentation which would be required by the operator to i
perfom the core cooling and RCS heat removal were also i
identified, including tRe RC5 and 0T5G instrumentation which i
is independent of the ICS/NNI systems and added subsequent to the i
submittal of Reference No. 3.
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0 TDR No. 598 Page 12 Instrumentation was reviewed by considering whether sufficient redundancy existed in indication so that failure of non-qualified instrumentation would not mislead the operator and that for the scenarios described the operator would have the required indication s.
The review of the systems operating modes and identified components included the application of the concept of failure as a result of the event and single failure of safety related components independently of the event.
2.6 Electrical Systems Interactions The electrical systems interactions for failure of electrical components which have not been qualified for harsh radiological environments, on electrical components which are required for SBLOCA mitigation has been accomplished. His been accomplished by a review of the THI-1 electrical wiring diagrams, (Reference No.10 & 11) vilh the following results:
1 - All M. O. valves identified in Appendix " A" have dedicated over current protection devices and contain no interfacing devices that could impact their operation.
2 All solenoid valves are individually fused and have no other
,. electrical devices other than limit switches that could impact their operation.
Rese limit switches are identified as part of the valve and are radiologically qualified.
3 - The pumps identified do not contain any electrical devices other than those identified and qualified which could interface with the operation of these pumps.
s 4 - The identified air handling fan motors do not contain any electrical devices which, if not qualified, could interfere with their operation.
5 - Instrumentation The effect of non-qualified instrumentation failure on safety-related instrumentation power supp11es has been evaluated by GAI (Reference No.11). This evaluation
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s concluded that their failure could not result in a failure of the safety-related instrumentation power supplies or affect the operation of components required for SBLOCA mitigation.
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Certain instrumentation identified in Appendix "A", requiring radiation qualification receive their power from ICS/NNI N
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TDR No. 598 Page 13 power. letile the ICS/NNI power is not safety-related, it does have diesel backed power and failure of an ICS/NNI power supply is indicated in the control room.
Each such radiation qualified instrument has its power supply identified on the control room panel, therefore, the operator will be aware of an ICS/NNI power supply failure and could then rely on other qualified instrumentation for the required indication. Where such instrumentation provides an automatic control function, the means for qualified manual control and indication are available in the control room. -
3.0 RESULTS The results of this evaluation are presented in Appendix "A", wherein each electrical component which is located within a harsh radiological environment and which is required to mitigate a SBLOCA has been identified and indicated to require qualification for the harsh environment to which it may be subjected.
The reactor incore themocouples, which had not been included on the list submitted by Reference No. 3, provide verification of operation within the Pressurized Themai shock (PTS) limits and verification of adequate core cooling and subcooling margin. Recognition of. this item is as a result of development of the B&W ATOG document since the Reference No. 3 submittal.
4.0 G)NCLUSIONS This report provides a complete list and the basis for affirming the completeness of the list of items which require qualification to radiation levels associated with large break LOCAs in order to mitigate the spectrum of the SBLOCA in compliance with 10CFR50.49(b)(1) and (Reference No. 2) prepared in accordance with the applicable NRC (b)(2), has been guidance and is responsive to the NRC (Reference No.1) request.
5.0 RECOMENDATIONS 1 - It is recommended that this report be reviewed by TMI-1 operations in order to ensure that the operators recognize the qualifications and limitations of the components which may be called upon to mitigate a postulated SBLOCA.
2 - Specific procedural guidance must be incorporated to justify not qualifying HU tank level indication (MU14DPT).-HPI flow indication i
MU23DPT 1, 2, 3, 4) and seal. injection flow indication (MV-42-DPT). Refer to specific instruments (Appendix "A" and Appendix "C" for guidance.
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l TDR No. 598 Page 14
6.0 REFERENCES
1 - NRC (D. G. Eisenhaut) letter to GPUN (H. D. Hukill) dated August 8, 1984 2 - NRC (D. G. Eisenhaut) letter to GPUN (H. D. Hukill) dated May 25, 1984 3 - GPUN (H. D. Hukill) letter, LIL-161 dated May 18,1981, to NRC Office of NRR (J. F. Stolz) 4 - GPUN Technical Data Report, TDR-282, Rev. 4 "mI-1 Qualified Equipment Location's and Environments" 5 - GPUN Technical Data Report, TDR-121, Rev. 4 " Design Review of Plant Shielding and Radiation Qualification for Post Accident Operations 7
Outside Containment 6 - GPUN Technical Data Report..TDR-083 Rev. 3 " Evaluation of Containment Isolation Signals" 7 - GPUN Calc. No. C-1101-220-5450-013 Revision 0 "TMI-1 Estimate of the Potential for Nitrogen Gas Entrainment in the RCS during CF Discharge" 8 - B&W Technical Document No. 74-1124158-00, "TMI-l Abnormal Transient Operating Guidelines," Part II, Vol. 2, Appendix F.
9 - GPUN Technical Data Report, TDR-269. Revision 0 " Nuclear Services Closed Cooling Water System Pump Runout Capability Test" 10 - EDS Nuclear Report No. 02-0370-1060 Vol. V. Environmental qualification of Class lE Electrical Equipment Report 11 - Gilbert / Commonwealth (GAI) letter (No. GAI/EI-ICS/11077) dated 9/10/84 to GPUN 1
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APPEM)IX A AT THI-I ASSESSENT (F EttunICAL CDWOENTS REQUIRIE RADIATIIM (RJALIFICATION FDR SW.0CA MITIGATION o
NOTES:
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Ridisloaical Environment Code l
2.
Radiation Quali fication Heautrement Code H-15 = Hersh normal operating environment dose.
Y Component needs to be qualified for its identiried radiological H -1 Hersh operating environment, normel operating dose plus SBLBCA environment.
postulated dose.
N =
Component le exempted from H -2
- Hersh environment only afler component completed its operating, radiological qualification if it's function but its failure due to radiation could result in located in a mild environment or unocceptable consequences.
meets the H-4 environment code criteria.
H -3 Component may be required to operate within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter the zweetne trip in a herah environment after which the component function le not required, and its failure would not result in 3.
Location unacceptable conseque see.
l 3,
RB = Reacte-Building H -4
- Hersh environment only after component completed its operating furetion or component function le not required in a harsh Auu._ Bldg. = Auxiliary Building eroe i
environment, and its failure would not result in unacceptable identified in TDR-282 and l
consequences.
TD R-121.
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H
= Component le located in a mild or low radiological environment.
Int. Bldg. = Intermediate Building l
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1 i
j
Pag) 1 WSTEN: KACTOR EXILDING ISG.ATION (RBI)
RADIATION QUAL REQ'D RADIOLOGICAL HFONENT FUICTION LOC AT ION V/N ENVIRON ENT E MARKS H-V-1 B&C RBI Valve (s) for RB Purge Exhaust & Supply RB V
H-l*
Motor operated valves, normally closed.
W 1&2 & LS/
RBI Valve (s) for RB Purge Supply Int era Bldg N
H SV is normally de-energized to close AH-V-ID.
failure of solenoid due to radiation keeps AH-V-IA 1-V-ID closed in its RBI state.
V l&2 & LS/
RBI Valve (s) for RB Purge Exhaust Aux Bldg V
H-l*
SV is normally de-energized to close AH-V-1A.
H-V-1A (A-13)
Failure of solenoid due to radiation keeps AH-V-10 closed in its RBI state.
DL-V-303 RBI Valve (s)for RCDT Liquid to R System RB V
H-l*
Motor operated valve, normally closed.
V & LS/
RBI Velve(s) for RCDT Liquid to tOL System Aux Bldg V
H-l
- SV is de-energized to close tOL-304 E -304 EL-V-304 (A-13) is normally closed. Failure of solenoid due to radiation keeps lOL-304 closed in its RBI state.
IL-V-534 RBI Velve(s) for RB Sump to Aux Bldg Susp Aux Bldg V
H-l
- SV is de-energized to close ICL-534. E -534 is (A-1)
I, normally closed. Failure of solenoid due to
- aolation keeps IEL-534 closed in its RBI state.
IL-V-5?,5 RBI Velve(s) for Sump to Aux Bldg Sump Aux Bldg V
H-l*
SV is de-energized to close tOL-535. tOL-535 is (A-1) normally closed. Failure of solenoid due to radiation keeps 40L-535 closed in its RBI state.
C-V-2 RBI Valve (s) for ICCW Return RB
- V H-3 If IC system line inside RB breaks, then line break isolation signal will not be initiated until IC surge tank drops. Valve may thus be required to operate in the post accident radiological environment.
a
YSTDt: REACTOR BUILDING IS1ATION (RBI) (continued)
Pag) 2 RADIATION QUAL REQ'D RADIE 0GICAL DWONENT FUPCTION LOCA TION V/N ENVIRON KNT REMARKS V & L S/
RBI Velve(s) for ICCW Return Aux Bldg Y
H-2 If valve isolation la required due to IC system B-V-3 (Solenoid Valve & Limit Switches)
(A-9) line break, the valve will have only been subjected to mild radiation within the time before its closure. SV is energized to close IC-V-3.
Failure of SV would fail IC-V-3 open. If IC system line inside RB breaks, then line break isolation sicpal will not be initiated until IC surge tank drops. Valve may thus be required to operate in the poet occident radiological environment. Therefore, SV & LS need to be radiologically qualifled.
V & L S/
RBI Valve (s) for ICCW Supply to CRDH &
Outside of N
N If IC eystem line inalde RB breaks, then line 3-V-4 & 6 Other IC S@ plied Services Aux Bldg break isolation signal will not be initiated -
Solenoid Valve & Limit Switches area (A-13) until IC surge tank drops. Valve may thus be
' required to operate in the post accident radio-logical environment. Valve is located in a mild radiological environment for the entire pcst LOCA accident period. SV is energized to close the respective IC-V-4 & I C-V-6 valves.
e 8
B-V-2 A &
RBI Valve (s) for Norm Cooling Water Int Bldg N
H Hotor operated, normally open, RBI valves. Close B-V-7 Return for RB Coolers on RBI 1600f RCS pressure.
V & L S/
RBI Valve (s) for RB Air Sample Int Bldg N
N SV is normally de-energized to close SV-1, 2,3&4 CWV-1/2/3/4 Failure of solenoid due to radiation keepe these valves in their RBI state.
Valves are not required to be re-opened unless RM-A-2 containment iodine and gas radiation monitor is to be used. Even if a containment air sample is taken, tp integrated radiological dose is less the 1 x 10 rede, and valves are therefore exompt from qualification.
DG-V-3 RBI Valve (s) for RCDT Vent to WDG System RB Y
H-l
- Hotor operated valve, normally open, closes on RBI' isolation signal. This valve does not need to be re-opened for SBLOCA mitigation.
V & L S/
RBI Valve (s) for RCDT Vent to WDG System Aux Bldg Y
H-l*
This solenoid operated valve is de-energized to DG-V-4
( A-13 )
close the valve. Failure of the solenoid due to radiation keeps WDG-V-4 in its closed RBI state.
The magnetic valve position switches are only required to confirm the initial valve closing.
O
iYSTEN: IEACTOR CJILih% ISG. ATION (RBI) (continued)
Pegg3 RADIATION QUAL REQ'D RADIOLOGICAL IHFON ENT FU ETION LOCA TION Y/N ENVIRON KNT REMARKS iS-V-3 5 RBI Valve (s) for NS Cooling Water RB Y
H-3 Motor operated valve. If NS system line inside Return from RCP Hotors RB breake, then line break isolation sipal will not be initiated until NS surge tank level drape.
Valve may thus be required to operate in the post accident radiological environment.
IS-V-15 &
RBI Val've(s) for NS cooling water supply Aux Bldg V
H-3 Motor operated RBI valves. Since a SILOCA will not.
IS-V-4 and return for RCP Motors
( A-13 )
yield a 30 psig RBI sipal, valves may be initi-ated to close on a line break isolation' signal. A line break sipal, if received, would occur while the Aux Bldg is still not contaminated. Therefore if valves are required to function, they will do so in a normal radiological environment while the
, HPI water source is the BWST. Since the valves are located in a normally high radiation area, they need to be qualified to function after receiving normal integrated radiation dose.
iv & L S/
RBI Valve (s) for Normal RCS Hakeup Aux Bldg ya H-1 SV is energized to close MJ-V-18. Failure of SV U-V-18
( A-7 )
resulte in MJ-V-18 re-opening. MJ check valves prevent reverse flow from RCS if MJ-V-18 re-opened and MJ or HPI pump is not operating.
Failure of MJ-V-18 in the open position may result in HPI pump runout. Therefore, MJ-V-18 solenoid volve needs to be qualified for the SS.0CA postulated radiological environment. LS is not required after valve is confirmed closed feJM **mM de-**ewgres & sgr., &md 45 by 1600 # RCS pressure isolatien sicpal EV & L S/
RBI Valve (s) for RCS Letdown Containment Aux Bldg Y
H-1 SV is energized to close MU-V-3.
SV failure due H-V-3 Isolation Valve (A-7) to radiation would de-energize and fail MJ-V-3 open. If MJ-V-2A or B failed to close due to "LO(P" and one failed D/G, then MJ-V-3 failing open is unacceptable. Therefore, it needs radiation qualification. The LS should also be qualified even though they are only required to confirm the initial valve closing.
l i
I
l SYSTEM: KACTOR BUILDING ISG.ATION (RBI) (continued)
Pag 3 4 RADIATION QUAL EQ'D RADIOLOGICAL C00FONENT FUPCiION LEA TION Y/N ENVIRON ENT REMARKS
H-l*
Motor operated RBI valves. RCS letdown le not Isolation Valve Clr. Room) required to mitigste a SEOCA. RCS letdown may only be initiated af ter a LOCA if RCS fluid is not significantly contaminated due to core fuel fail-ure. If fuel reilure did not result due to the transient, then the valve operator would not have been exposed to the RB radiation levels and there-fore may be exempt from radiation qualification.
l SV & LS/
RBI Velve(s) for Reclaimed Water Aux Bldg Y
H-l*
SV is de-energized to close CA-V-189. Failure of i CA-V-18 9
( A-7 )
colonoid due to radiation closes CA-V-189 to its RBI state. LS are only required to confirm initial valve clowing.
e i
i CA-V-1,3,13 RBI Valve (s) for RCS Sample RB Y
H-l*
Motor operated valves are normally closed and are (PZR Wtr & Stm SP, Letdown) not required to opera to for SEOCA mitigation.
j!
1
! SV & L S/
RBI Valve (s) for RCS Sample Aux Bldg Y.
H-l*
SV le de-energized to close CA-V-2. CA-V-2 is lCA-V-2 (A-13) normally closed. Failure of solenoid due to l'
radiation cloose CA-V-2 to its RBI state. LS are i
l only required to confirm initial valve closing.
i CA-V-4 A,(B RBI Valve (s) for OT SG Sample RB y
H-1
- Motor operated valves are normally closed and are not required to operate for SBLOCA mitigation.
i
! SV & LS/
RBI Velve(s) for OTSC Sample Turb Bldg N
H SV & LS are not located in a radiological area and would thus not fail due to radiation.
i CA-V-5A,58 e
l i SV & L S/
RBI Valve (s) for CF Tk Mekeep Aux Bldg Y
H-l*
SV is de-energized to close CF-V-19 AaB.
! E-V-19 A&B (A-13)
CF-V-19 A&B are normally closed. Failure of l
solenoid due to radiation keepe CF-V-19 A&B in their required RBI stats. LS are only required to confirm initial valvo closing.
1 I
i 7-V-2 A&B RBI Valve (s) for CF ik Bleed & Sample RB Y
H-l*
Motor operated valves are normally closed valves and are not required to operate for SEOCA mitigation.
i l
! SV & L S/
RBI Valve (s) for CF Tk Bleed & Sample Aux Bldg Y
H-l*
SV is de-energized to close CF-V-20 A&B.
(A-13)
CF-V-20 A&B ere normally closed. Failure of jCF-V-20 AaB solenoid due to radiation keepe CF-V-20 A&B in their required RBI st ate. LS are only required j
to confiru initial valve closing.
j 1
}
4 I
1 e
n l
iVSTEM: KACTOR BlILDING ISG.ATION (RBI) (continued)
Pago 5 RADIATIENI QUAL REQ'D RADIOLOGICAL DWONENT FU ETION LOCA TION Y/N ENVIRON ENT REMARKS El-V-2 5 RCP real Return Isolation Inside RB RB Y
H-1 Motor operated valve, normally open, closes only on 30# RB pressure RBI sipal. If seal return were to become highly activated, then this velve would need to be manutally closed. This may occur efter going to susp recirc mode. Therefore, this valve must be radiation qualified.
i IV&LS/
RCP Seal Return Isolation Outside RB Aux Bldg. (A-13) Y H-1 SV le normally energized to ke Mi-V-26 open.
B-V-26 (Ses remarks for valve Kl-V-25.
l e
I j
f' 4
i h
4 d
I 6
h t
rSTDt: LOW PESSJE DUECTION FUIC TION (LPI) /
Pega 6 PUW SUCTION FROM RB SUW FOR LPI MDDE OR HPI PIGGYBACK MODE RADIATION QUAL REQ'D RADIOLOGICAL NFONENT FUICT10N LEA TION Y/N E NVIRON ENT REMARKS 1-P-1 A&B Decay Heat Pump - Provide LPI or Aux Bldg Y
H-1 The Di pumpe are required to be operated for the Stpply Suction to HPI Pumps from RB Susp (A-1) entire SILOCA event and must therefore be qualified for the SBLOCA postulated radiological environment.
l i-V-6 A&B RB Sump Suction Valve for OH-P-1 Aux Bldg Y
H-l*
Motor operated valves need to be capable of (A-1) opening when DH recirculation from the RB sump is initiated. Once recirculation from the sump is initiated, the velves do not need to be operated.
Thus, volves need only be qualified for the normal operation radiation environment integrated dose.
i I
4-V-5 A&B DH Pump Suction from BWST Aux Bldg N
N Motor operated, nomally open, valves which need
( A-8, A-10) to be capable of being closed when the DH pumpe are switched to RB susp suction. Once they have l
been closed, they do not need to be re-opened.
/
/
Thus, they need to be qualified for their normal plant operation radiation environment integrated I,
dose only. Since these valves are located in a normally mild radiological environment, they are exempted from radiation qualification.
k 5-V-2 A&B OH Pump Suction from NaOH Tank Aux Bldg N
M Motor operated, normally closed, valves. These
( A-8, A-10) valves are activated to open by a 4 psig RB pressure switch as an anticipatory step to RB oprey actuation. Tim SELOCA does not result in a 30 poig RB pressure which would actuate the RB apray pumps. Since the RC does not initially depressurize to the point where the DH pumps can inject into the RCS, the Di pumps will operate on their minimum recirculation flow and will not draw Namt into the DH system piping.
S b
iYSTEM: LOW PESSJE INJECTION FUEIION (LPI) /
Pag) 7 PUW SLETION FROM R8 SUW FOR LPI H00E OR HPI PICGYBACK H0DE (continued)
RADI ATIIM QUAL REQ'D RADIOLOGICAL X)WONENT FU ETION LEA 1 ION Y/N ENVIRON KNT REMARKS Y
H-1 These motor operated valves will open H-V-4 AaB LPI Hode Injection Valves Aux Bldg )
automatically by either a 1600 peig RCS pressure
( A-9, A-7 or by a 4 poig RB pressure signal. Since velves DH-V-19A&B, used for DH flow throttling, are located in an unoccessible aree efter having operated in a R8 sump recirculation mods, valves DH-V-4A&B need to ha utilized for his throttling -
process. There fore, they need to be qualified for the SBLOCA postulated radiation.
H-V-1 DH Dropline Isolation et "B" Hot Leg RB (Inside Y
H-1 DH-V-1, 2 & 3 are normally closed, motor operated "D" Ring) manually controlled valves. These valves are utilized for normal operation DH removal. They H-V-2 DH Dropline Isolation et Containment RB Y
H-i
These valves need to be M-V-3 DH Dropline Isolation Outside RB Aux Bldg Y
H-1 opened to provide a flow path to prevent baron
( A-7 )
precipitation in the core region. Valves [H-V-1
& 2 have persionive interlocks which are
.e controlled by qualified instruments RC-3A-PT3 and RC-3A-PT4.
These valves provide for one of the redundant paths, in conjunction with RC-V-3 & 4 for this function.
Therefore, these valves need to be qualified for SBLOCA postulated radiation levels.
?
E-V-2 & C5 Inlet & Bypeas Valves for DC System Aux B81dg WA WA Not applicable for electrical qualificatiro.
M flow through DHR Cooler (A-1)
Pneumatic (manually) regulated valves and local hydraulic flow indication (FI-26 & 27). These valves are utilized to control the amount of heat removed by the DH coolers.
0
q
'SID4: HIGH PESSJE INXCTION FUNCTION (HPI)
Pagg 8 RADIATION QUAL REQ'D RADIG.0GICAL pr0NENT FU ETION LOCA TION Y/N ENVIRONENT REMARKS i-745 A8&C Pressure switches, trip MJ/WI Aux Bldg V
H-1 These switches insure that Itse oil pressure le pump on Low oil pressure (A4) aufricient to meintain HPI/make-tp pump operation. Failure of these switchee under SS.0CA conditions could result in MJ/WI pump trip. These pressure switches are located in a radiologically harsh environment and therefore need to be radiologically qualified, i-479 A,8&C Start & Stop Aux. oil Pump (MJ-P-2)
Aux Bldg.
N H-4 These preneure switches provide alare indication of low oil preneure in the MJ/WI pump and motor i448 A,8&C Start & stop Aux. gear oil pump (MJ-P-4)
(A4) and in the M4/WI gear oil systems. They also provide start sipale for the oil systems auxiliary oil pumps. The failure of these i-478 A,8&C Mein oil pump low pressure alarm switches would not prevent the starting or
. operating capability of the MA/HPI pumpe. Sirce the auxiliary oil pumpe MJ-P-2 & 4 are also not required for the MA/WI pump operation and failure of these switches does not af fect the HU/HPI pump operation, the switches can be exempted from Se1.0CA radiation qualification.
D' J-P-1 A.POC Makety (MJ & WI) Pumpe Aux Bldg (A4)
Y H-1 Make-w/WI pumpe required for accident mitigation.
4-P-2 A,B&C Auxiliary 011 Pump for Respective VJ-P-1 Aux Bldg (A4)
N H-4 These pumpe are normally operated in conjunction with and as a backup to the main oil pumps to insure sufficient lubrication for the make up pumpe motor and pump bearings. The failure of this auxiliary oil pump would jeopardize the eveilability of the HPI pump only if the mein oil pump had failed. Failure of the auxiliary oil pump would not affect the operation of the main oil pump or the El/HPI pump. Therefore, those pumpe een be exempted from SE.0CA radiation qualification although they are located in a normally harsh environeent and would also be subjected to the SBLOCA postulated radioactivity.
4
YSTD4: HIGH PESSJE INJECTION FUNCTION (HPI) (continued)
Pag) 'l RADIATION QUAL REQ'D RADIOLOGICAL DWONENT FU PCTION LOCA TION Y/N ENVIRONKNT REMARKS U-P-3 A,8&C Main Oil Pump for Respective MJ-P-1 Aux Bldg (A-6)
Y H-1 These pumpo nperate continuously during make-tp pump operation to supply the pump motors and bearings with sufficient lubricating oil.
Failures of the main oil pumpe would jeopardize WI operation. These pumps are located in a radiologically harsh envirorunent during normal operation and during SBLOCA events with the postulated radioactivity.
U-P-4 A,B&C Gear Oil Pump for Respective MJ-P-1 Aux Bldg N
H-4 These pumps provide a backup pumping capability
( A-6 )
for the gear shaf t driven gear oil pump. These pumps are started by a low gear oil pressure switch. However, failure of the pressure switch
, or this penp does not prevent the start of the MJ/WI pump. As the MJ/WI pump starts, the gear driven oil pump will supply the required gear oil. Thus, these auxiliary gear oil pumpe era not required to stpport or adversely affect the MJ/WI pumps if they fail during the SEOCA 8
event. Therefore, these pumps can be exempted from SEOCA radiation qualification.
U-V-12 Makety Pump Isolation Valve Normal Aux Bldg N
H-4 This motor operated valve is normally open. If Suction from MJ Tank (A-5) the valve operator fails due to the SEOCA postulated radiation, it will fail as-la.$ince procedure 0.P.1104-2 limits the MJ tank gas pressure to prevent gas entrainment to the WI pumps, this valve may be left open or closed during a SBLOCA. Failure of this valve due to radiation will not af fact the ability to mitigate a SBLOCA. Therefore, this volve does not need to be qualified for the SBLOCA postulated radiation.
8 i
c SYSTEM: HIGH PESSUE INJECTION FUNCTION (HPI) (continued)
Page 10 RADIATION QUAL REQ'D RADIOLOGICAL C00f0N DIT FU ETION LOCA TION Y/N ENVIRONENT REMARKS E V-10 HPI Injection Valvea Aux Bldg V
H-1 Upon ES sival, these valves open to permit HPI A,B,C&D
( A-7, A-13) injection flow to the RCS. HPI flow is throttled in accordance with the procedural criteria, by modulating the position of theses valves.
Failure of these valves in the opened or closed position is undesirable. Therefore, these valves need to be qualified for the SBLOCA postulated radiation environment.
EV-14 A&B HPI Mode, NJ Pump Suction Isolation Aux Bldg V
H-1 Upon ES sipal, both valves open to allow HPI from BWST (A-8) pumps to take suction directly from the BWST.
For " piggy back" mode of operation, these valves remain open until LPI pump operation le verified. Thereafter, the valves may be closed.
Failure of the valves in either the open or closed position is undesirable.
EV-13 MJ tank vent and hydrogen & nitrogen Aux Bldg N
H-4 These solenoid valves are normally closed and and sipply valves.
(A-3) de-energized. Failure of the solenoid due to EV-2 7&28 (A-4a) radiation would keep these valves in their I,
normally closed state. These valves will not be required to functiosa for SBLOCA mitigation.
Therefore, they need not be qualified for exposure to radiation.
0 b
I
rSTEM: HICH PESSJK INJECTION FtMCTION (HPI) (continued)
Page 11 RADIATION QUAL REQ'D RADIOLOGICAL Dlf0NENT FU ETIGN L(EA TION Y/N E NVIRON ENT RENARKS I
H-V-36 & 37 Common Recirc Isola tion for HJ Pumps Aux Bldg N
H-4 During normal operating conditions, these velves bedc to MJ Tank (A-5) are open. An ES actuation of the HPI puge close these valves in order to attain the maximum ES injection flow. Per OP-Il01-1 " Plant Limite &
Precautions," these valves are required to be opened when HPI flow is throttled below,
80 gpe/ pump in order to provide a path for recirculation. For a SBLOCA in which the bro remain above the 80 gpm/ pump.p flow will always is m-isolatable, the HPI pum Thus, these valves do not need to be re-opened from their closed ES allspoent. For a SBLOCA in whid the break is isolated some time during the event, the operator will turn off the llPI pumps, utilize the natural circulation cooling to control RCS pressure and
' temperature and may periodically restert on HPI pump to provide RCS makeg for the RCS volume shrinkage. Thus, the se valves do not need to be re-opened from their closed ES elignment.
Therefore, these valves do not need to be a
qualified for the SBLOCA postulated radiological environment. Also, these valves are located in e 8
normally mild radiological environment.
Therefore, these valves may be exempted from radiological qualification.
H-V-7 A&B M1 Pump HPI Mxle Suction from RB Sump Aux Bldg Y
H-1 Under conditions of normel operation, these
( A-8 )
valu e are closed. For purposes of SE.0CA mitigation, it may be necessary to line up WI and LPI for the " piggy back" mode of operation.
In this event, DH-V-7 A&B will be required to open per ATP 1210-6. It may be desired to reclose these valves after having operated in a throttled HPI " piggy beck" mode. Therefore, these valves need to be qualified for the SBLOCA postulated radiatien.
S 9
f i ;
SYSTEM: HIGH PESSJE INJECTION FUNCTION (HPI) (continued)
Peg) 12 4
RADIATION QUAL REQ'D RADIE0GICAL C00f0NENT FU ETION LOCA T ION Y/N E NVIRON ENT REMARKS 3
- DC-P-1 A&8 Decay Closed CW Pump - Supplies Cooling Aux Bldg N
M The decay host clneed cooling system le en ES Water for MJ-Pump ( A&C)
(305' Elev system. During normal operation, puupo IA and 18 NW Corn) Outside are in stene y.
Upon en ES elpel, Area (A-13)
(1600, 500 peig RCS pressurize or 4 poig R8 pressure), both pumpe are started. These pumps opply motor and bearing cooling water for the make tp pumpo IA and IC, decay heet pumpe 1A and 18 and building sprey pumpe 1A and 18. Failure of the pumps is unacceptable. These pumps are located in a mild radiological environment during
}
normel and SEDCA conditione and are therefore i
exempted from radiation qualification.
1 Aux Bldg WA M
The nuclear services closed cooling water system NS-P-1 A, B&C Nuclear Services Closed CW Pump (-)
is en ES system. Two pumpe are normally on ES Sapplies Cig Water for MJ-Pump 8 (305' Elev NW Co rn) Outelde sten 6y. For emergency operation, only one pump Area (A-13) is required for opers tion. Under a lose of i
offeito power, two (2) pumps which have been 4
selected to etert on ES actuation would be eterted wie loeding block 3 from the diesel
{
f generator. The pumpe provide cooling water to i,
equipment that is vital for mitigating e SEOCA.
j i
These pumpe are located in a mild radiological I
environment during normel and 58 LEA conditions, i
and are therefore exempted from radiation j
qualification.
1 1
I NS-V-56. A&8 Cooling Water Control Velve for Aux Bldg.
N N
These are non-electric operated and controlled AH-E-l> A&B (305' Elev.
devices. NS-V-56 A&B era pneumatic flow l
NW Gorn) Outelde regulatlog velves which are located on the NS Ares 13 cooling water inlet lines to the pug cubicle air j
coolers for nuclear services and decay heet closed cooling systems. Both valves are j
controlled by thermootste located in the pump cahicles. Since the velves, controle and i
thermostete are loceted in a mild radiological I
environment during normel and SEOCA conditions, I
and are exempted from radiation quellfication.
~
f AH-E-15 AAB NS & DC Cooling Systeme Pump Area Aux 81dg.
N M
These ES eir handling units which provide cooling
(
(305' Elev for the NS and DC pump cubicles. Fan air flow ie NW Co rn) Outelde controlled by means of temperature switches and l
Aree 13 edjustable despero located in the ducting on the discharge sides of the rene. Since the fene end i
control devices are located in a mild i
radiological eres during normal and 58LOCA conditions, they are exempted from radiation l
qualification.
1
e A
i STD4: EERGDCY FEEDWATER & ATHOSPHERIC DTDP VALVES (Em & ADV)
Page 13 RADIATION QUAL REQ'D RADIl0GICAL of0NENT FU ETION LOCA TION Y/N ENVIRON ENT REMARKS F-P-1 Turbine Driven EFW Pump Int Bldg N
H All End system components are located in non-radiologically contaminated areas.
-P-2 A&B Hotor Driven EFW Pumps & Ntors Int Bldg N
H WV-2 MB Steam Stpply to Turbine Driven EniP Int Bldg
$4 H
L-V-13 A&B WV-4 MB HS Atmospheric Dump Valves Int Bldg N
H
-V-1 A&B EFWP Suction & Discharge Cross Connects Int Bldg N
H
' V-2 MB
' V-4 & 5 EFW Suction f cm River Water Pumps Int Bldg
.N H
RR-P-1 A&B a
8' H
WV-10 MB EFW Suction from Condensate Storage Tanks Cond Stg Tank N
-V-30 A&B EFW Flow Regulating Valves Int Bldg N
H l-E-2S M B EF Pump Area Ventilation Unit &
Int Bldg N
H
=
i-V-55 A&B Cooling Supply Valve Limit Switches e
its: GThe Licensee has not taken credit for the EFW turbine and associated components for $10CA events.
9 4
I
SYSTD4: MISC C00f0NENTS Page 14 NADIATIENd IRJAL llEQ'D RADIS.0GICAL C00FONENT FU8CTIGN LOCA TION Y/N ENVIRONIE NT REttutKS -
E-V-1 AaB CF Tk Isolation & Vent Velves R8 N
H-4 These ere notor opere ted wolves whose electric -
breakere are opened (de-energized) when the E-V-3 AAB reactor le critical. The CF system is designed to inject boreted weter inta the seocter core during a LOCA. Such en injection, if E-V-1 could not isolete and E-V-3 could not vont the CF tank, would result in some time deley in the ability to depreneurire the KS below 600 peig and 275'F. CF injection will not result in nitrogen injection into the KS (Reference No. 7 calc) Therefore, these volves are exegt from radiation qualification.
NS-V-3 2 NS System Sgply for ICP Seel Return, Aux Bldg N
H-4 This motor operated velve wee originally intended Weste Cee Coop & Mioc Weete Evap Cooling
( A-8 )
to provide seene of isoleting uneesentis! cooling Isolation velve functione in order to prevent the runout flow with one NSCCW pump. A -d:n-M test. reported in TOR 269 (Reference 9) indicated that this volve could be left open. Therefore, this velvo le not
.e required to be opereted, e'
1A & 18
' Breaker Cabinet for ES Componente Aux Bldg N
M.
Theen breaker cabinete contain the breakers which ES Valvoo CC
( A-15 )
provide the electric power for ES velves. These cabinete are located in areas which have e mild IC Breaker Cabinet for ES Componente fHS 2 81' elev N
M radiological environment during normal opeteti v; ES Valvoo CC (TDR-121 conditione and as a result of SILOCA postulated AreeXIII) radiation. They are therefore exempted from required radiation qualification.
L 4
E-V-3 Preseuirzer Spray Block Velves R8 Y
H-1 These valves are required to provide a escend (on top of PZR) flow path to prevent boron precipitation in the
/
(ineide D-rings) core. These are required to meet the single failure and redundency criterie in conjunction E-V-4 Auxiliary Spray Block Velve R8 Y
H-1 with DH-V-1, 2 & 3.
Therefore, these are 346' elev.
required to be radiation qualified for SELOCA.
outside D-rings SV & L S/
RCP Seal Injection Isolation velve Aux Bldg Y
H-1 SV le de-energized to close EV-20. This volve EV-20 (A-13 )
requires menuel actuation to close or re-open the velve. Siste a failure of the SV due to radiatius would close the velve and RCP seal injection is desired to prevent pump esel damage, the SV & LS need to be quellfied.
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8
VSTEM: MISC CDW ONENTS (continued)
Page 15 RADI ATIIM QUAL REQ'D RADIOLOGICAL 00FONDIT FU ETION LEA TION Y/N E NVIRON ENT REMMtKS H-E-1 A,B&C Rx Bldg Recirculation Unit Fan Motore R8 Y
H-1 These R8 fene, recirculate the RB eir during normal and poet accident conditione. They are required to operate during a SBLOCA in order to remove the heet rolessed to the R8 no a result of the SEOCA. They muet, therefore, be quellfied for the SEOCA postulated radiation in order to be able to provide the required air circulati;,.i.
i j
5-V-52 & 53 Nuclear Services Cooling water Int Bldg N
M The RB rene required motor and bearing cooling Bhc Stpply & Return for Cooli Water to whidi le provided thru theos valves. Sirce those 3
R8 f an Motore (AH-E-1 A,
)
velves are located in a mild normal and poet LOCA erwironment, they are exempted from radiation quellfication.
R-V-3 A,B&C River Water Supply & Return Int Bldg N
M These valves must open when river water le being R-V-4 A,B,0&D to Cooling Colle In RB Recirc utilized for R8 cooling with the Emergency R8 Unite (AH-E-1 A,8&C )
cooling colle. 1hese valves are initiated to open by a 4 poig R8 preneurs ES actuation or C-7 Pressure Controller & Pressure Int Bldg N
a M
menuel initiation. Since these valves are R-V-4 Control Velves (Norme] & Bypees) to located in mild radiological environment during J-V-5 Heintain R8 Emergency Cooling System normel operation and during a SEEA, they are Pressure above 60 peig exempted from radiation qualification.
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SYSTDt / INSTRtKNTATION Pege 16 RADIATIWi QUAL EQ'D RADIE0GICAL EOWONENT FU ETION LEA TION Y/N ENVIRON ENT RENARKS Reacter Coolant System E-4A-TE 2/3 RCS Narrow Range TH DT RTDs (IFS)
RB Y
H-1 h These instrumente are not required for core.
E-48-TE 2/3 cooling or maintaining the RCS pressure boundary.
The required ES temperature indication is
-4A-TE 1/4 RCS THOT RIDS (indication)
RB Y
H-1 provided by the Incore T/Co. 'These instrumente 1E 1/4 (RTDs) provide very useful indicatione 'and e
)
benefit if qualified for SEOCA radiation.
-5A-TE 2/4 RCS Wide range fra D RTDs RB V
H-1 TE 2/4 (Indication & RCP and PORV interlocke)
-5A-TE 1/3 RCS Narrow Range TCED RIDS (indication) TE 1/3 4-959 & 961 ACS Wide Range TCED RTde (indication)
RB Y
Incore (T/C) Thermocouples RB Y
H-1 These instrumente are requireds to indicate that the core is being cooled; used to maintain the RCS within the PIS & NOT limits; indicate the core aih-cooling merging verify the effect of LPI cooling. Therefore, these instrumente need to be qualified for the SBLOCA postulated radiation.
-3A-P! 1/2 RCS Narrow Range Pressure Transmittere RB Y
H-l*
These instrumente are required to perform the FS P i 1/2 RPS (indication) function of tripping the reactor. After they have performed their KS function, their
)
indication is no longer required. Failure of these instrumente could result in opening the PORV. Such a failure would increase the site of the SBLOCA to a large SEOCA which is still within the emelyzed SBLOCA case. Thus, such e
failure le acceptable and these instrumente can be exempted from qualification for the SEOCA
?
postulated radiation. They must, however, be
]
qualified for their normal operating radiation i
dose.
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IYSTDI / INSTRtBENTATION (continued)
Pega 17 l
RADIATION QUAL REQ'D RADIOLOGICAL Dif0NENT FUICTION LOCA TION V/N ENVIRON DENT REMMtKS E-3A-PT 3/4 RCS Wide Range Pressure (ESAS)
R8 Y
H-1 RCS (ESAS) preneurs is required for initiation of E-38-P T 3 Transmittore (indication) the 1600 peig ES sipal for reactor trip and RBI. After the RPS and R81 functions have been N-949 RCS Pressure (B loop)) for SCM & indication RB Y
H-1 completed, RCS pressure is required for SEOCA RCS Pressure (A loop for SCM only mitigation. The RCS pressure needs to be
'T-963 provided by either the RC-3 Pie or free.the new
'E-958 & 960 RCS Wide Range TH0y RTDs for SCM &
RB Y
H-1 PT-949 & 963 pawesure treneeltters. Thus, either indication set of preneure transmittore needs to be quellflod for the SBLOCA postulated radiation levels. If only one set of pressure transmitters lo qualified, then the operator will have to evaluate whis instrument is providing erronnous indication.
These instrumente are required for the SC monitor
- alarm function, which was added by ATOG, to onoure that the operator trip the RC pumps during e SEOCA.
E-1 Pressurizer Level Treneeltters RB Y
H-1 The pawesurizer level indication is required in
.T 1,2& 3 order to eistablish if the HP1 throttling criterie
.T-777 is being met (i.e., SCM is within acceptable range and pressurizer _ level is greater then zero). The: pressurizer level is normally temperature componented by RC-2-TE 1/2. If the RC-2 TE 1/2 are not qualified for the 58LOCA postulated radiation and falle either hisp or low, any indicated level would only have a larger but minor indication error. Theret'ure, pressurizer (non-temperature componeeted) level tronomittore RC-1-LT 1,2 & 3 and LT-777 need to be qualified for SEOCA postulated radiation.
E-2-T E 1/2 Pressurizer Temperature RTDs RB N
H-4 See RC-1-LT 1, 2 & 3 for justification for i
qualification exemption.
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l SYSTEN / INSTRtBENTATION (continued)
Page 18 RADIATION QUAL REQ'D RADIOLOGICAL i(Def0 MENT FU ETION LOCA TION Y/N ENVIRON ENT RENARKS i Reacter Protection System 1
l PS-672, 673, R8 4 peig Pressure Switches for (IPS)
Aux Bldg Y
H-1*
These pressure switches provide the reactor trip 674,675 (A-13 )
elpel if the reactor building pressure exceede i
j 4 peig. Once the reactor is tripped by those owitches or any other reactor trip ei 1, then these preneurs switsee have compt their function. This funcLion would be completed i
before the postulated herah radiological environment existed in the Auxiliary Building.,
l Thus, failure of these switmos efter the reactor trip is verified would not effect the mitigation of a SEOCA. Therefore, these owildes need only j
be qualified for the normal plant operation i
radiation dose to escure their proper functioning to generate the 4 peig RB pressure reactor trip 4
si p al.
i
] Reacter Buildina Isolation NI-344 Intermediate Range Nuclear Instr.
RB N/A ' ',
N/A Only required to veri fy that the reactor le I
tripped. IE8 79-018, Stppl 2, enewer to question
- 12 states that qualification is not required for those instrumente.
l Reacter Buildina Isolation a
! PT-282, RB Pressure Trenamittore for Aux Bldg Y
H-1 These preneure transmittore provide the 4 peig R8
! 285, 288 (4 pelg) ESAS Initiation (A-13 )
pressure R81 isolation actuation. They are e
required to provide the initial RBI eipal for j
58LOCA which yield a 4 peig R8 pressure.
l
! PT-981 & 982 R8 Pressure Transmittore (C.R. indication) Aux Bldg Y
H-1 These pressure transmittore provide the C. R.
{ A&8
( A-13 )
with RB pressure indication in order to nonitor for a degradation of tio event.
I l PS-283, 284, RB 30 pelg Proesure Switches for ESAS Aux Bldg N
H-1 R8 preneure will not read 30 peig, and R8 aprey 5 286, 287, (A-13 )
ectuation is not required. To get en inedvertent i 289, 290 actuation, two PS on one BS pump must feil i
simultaneously. This le beyond failure acenerlo l
if, after a PS fails, the operator pute the BS l
pump in " pull to lock," en inadvertent actuation i
will not occur. Therefore, these pressure owitches are not required to mitigste a SEOCA.
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SYSTDI./ INSTRtKNTATION (continued)
Pega 19 RADIATION QUAL REQ'D RADIE0GICAL GNFONENT FUICTIGN LOCATION Y/N ENVIRON ENT REMARKS Orc Through Steam Generator SP-6A-PT 1/2 OTSG A & B Preneure Transmitters RB Y
H-1 The OTSG pressure indication le required for OTSG SP-6B-PT 1/2 preneurs control. Since the two OTSGe operate independently of sech other, redundant pressure PT-950 & 951 OTSG A & B Pressure Trenomittere RB Y
H-1 Indication le required for each DTSG. Timrefore, (C.R. indication) the SP-6 A & 8 pressure tronemittore need to be i
quellflod for the SBLOCA postuleted radiation.
The PT-950 & 951 preneure trenemittere are new instrumente, ir* pendent of the ICS system, are qualified for the SE OCA postulated radiation environment.
SP-1A-LT-445 OTSG Startup Range Level Instr RB Y
H-1 OTSG 1evel indication is required for OTSG 1evel SP-18-L T-4& 5 control with the EFW eystem in order to maintain oufficient level for core cooling. Normally, the SP-1A-LT-24 3 DTSG Operating Range Level RB Y
H-1 OTSG 1evel is controlled by the etert-tp range
- SP-1B-L T-2& 3 level and operating range level for forced circulation and natural circulation LT-775, 776 &
OTSG A&B Full Range Level Indication RB Y
H-1 respectively. Theref ore, these level tronomittere need to be quslified for tie SE(EA
- 788, 789 g
8, postulated radiation. New full range level I
s i
transmittore have been added and can also be
)
utilfred to control OTSG level by providing tempereture compensetion bened on OTSG presours in accordance with procedure (F-1102-II. These i
new level trenomittore are qualified for the s
58LOCA postulated tedietion.
l r
Emergency Feedwater PT-65, 71 & 72 EfW Pumps Discharge Pressure Transmittere Int Bldg N
H All EFW eystem componente are located in i
non-radiologically contaminated erees and are therefore exempted from SEDCA radiation qualification.
OPT-77 9, ETW Flow Transmittere Int Bldg N
M All EFW oystem componente are located in 782,788,791 non-rediologically contaminated areas and are j
therefore exempted f rom SK0CA radiation qualification.
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sYSTD4 / INSTRtK:'TATION (continued)
Paga 20 RADIATION QUAL REQ'D RADIOLOGICAL
[DM'ON ENi FU ETION LOCA TION Y/N ENVIRONIENT REMARKS Makete & Purificatinn / High Pressure Injection 41-14-LT MJ Tank Level Indicator Aux Bldg N
H-1 Redundant MJ tank level indication is provided so
.T-778 M1 Tank Level Indicator
( A-3)
N H-1 that the operator will have indication of high and low level gon which he would isolate any remaining inlet flow., "
If operator determinee
. that indication has f ailed, he would isolate the tank from the system. Therefore, a failure would not adversely af fect the mitigation of a SEOCA.
Therefore, these instruments need not be radiation qualified.
MJ-23-DPT HPI Flow Indication dP Transmitters Aux Bldg N
H-4
' The flow indication is only required for a SEOCA 1,2,3&4
( A-7, A-13) whids has depressurized the RCS to less than 1000 peig and only one HPI pug is operable. The MJ-4 2-DPT RCP Seal Injection Flow Transmitter Aux Bldg N
H-4 flow indication is ut111 red by the operator to
( A-5 )
check the total HPI pump flow and verify that it i
is less then the 550 gpa pump runout flow. If the pump flow is greater than 550 gpe, then the 8
pump flow must be throttled. The HPI system cavitating venturis will limit the flow to en acceptable pump flow, with seal injection flow less 10 gpe. If the operator cannot verify the total pump flow to be less than 550 gpm due to indication of flow indicator failures, then the seal injection flow will be isolated in order to protect the operating HPI pump. Therefore, failure of these flow indicators will require the operator to apply conservative actions, including isolation of a sesi injection with MJ-V-20.
Thus, these flow indicators need not be qualified for the SROCA postulated radiation. See Appendix "C" for evaluation logic.
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1Hf-l ABNORMAL TRANSEENT PROCEDURE LfCIC FOR A SB LOCA r
t00r MI RCS Fr30cer3
(
Initicting La RCS Fressure i
Trip Signals V;rtable F/T J
- Ra Trip Contalrusent Isolation (RTI)
Reactor Trip Turbine Trip Procedure 1210-I i l Initiate HPl. Man Letdoun Emerspncy or 13 Ra Tripped --> No --> Tri p lCO2 & ILO2 Borate
{ Yes Refer to 1202-2 13 4160V Fwr --w ho % Is at least one D/C avall % Yes Station Blackout Asall j
l0 Tes y
Refer to 1202-2A No Station Riackout with y
Io P2R level --> No --* Ini tiate HPI--
Loss of both D.C.'s
> 20" O
Tes 3
i r-j I3 RC L Press ---> No e Verify HPl/LPI & RRI Ini ti ated-
> 1600 psig l Yes i,
1 hen go to 1280-2
,e to CB Press --* No Verify HI'I/l(BI Initiated Loss of SC Margin
(. 4 pelg Yes r;
Co to:
Io Sol ) 25'F ---*Wo --> l. Ini tia te HPI 1210-3 Escessive Cooling
- 2. Trip acts or Yes
- 1. Vertfy EFW 1210-4 Lack of Primary
- 4. Rai se OTSC's to 95%
Neat Transfer e
Verify Post Trip Condi tions --* No Co to 1210-5 Ara Normal?
015C Tube Leak / Rupture Yes i
Vertfy RM-A-5 L and M era normal ? --> No 1
Co to follou up actions then use carnal Fit. Op.
(su. 1102-10, II)
I 1
4 s
a.uss un.auutu m u......
1217:-2 il V;rtfy Insed. Actions q,
Isolste Possible Sources of Leak if Use Normal Plant OP Restart RCP c
Yes 4
If 505 is regained Procedures r
(es. 1102-10,81)
I No Go to 1280-7
--Yes Are core flood Tks Lrg Brk LOCA emptying?
1210-4 Lack of Primary To Secondary lleat Transf e r 1.
Attempt to establish Co to 1210-4
- ----Yes 4
Is there lack of mala or amer PW.
Yes Lack of P/S '
Prt /See Heat Xfer?
r Then Ht Transfer h
2.
If RCS pressure h
No reaches PORV seapoint Yes 1.
If neither main nor eser r
Then 1.
Initiate HP1
(
PW can be established 2.
Open PORV Block l
To to 121v 1
- No ;
If there is F/S HE e
sfer & sol is not 1.
Open PORV No 4
Go t o 1210,d 1280-9 l
being recovered I ' Yes i to 1210-7 e-Yes - Are C 1ES emptying Ce to 1210-6 irge Brk SB t.0CA LOCS 1. No If RCS pressure reaches Yes
- Open FORV to reduce FORV setpoint RCS pressure to 100 psig above Sec. press.
No i
i if Pal /See he Xfer - Yes 7
Close FORV Co to 1210-1 reestablished l'
No Go to 1210-9 Are RCPS Avall --> No
Yes is cooldown rate --> N o --* Use RCP Bamps --> Has F/S Hg Xfer
- Yes 100* P/h r been reestablished?
Yes No co to 1210-1 in Co to 1210-6 Ce to 1210-9 SS LOCA I
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,SVSTEM / INSTRLDENTATION (cantinued)
Peg 3 21, 2 RADIATIWd is.
r QUAL REQ'D ROIOLOGICAL
' "d #
lG10FONENT FilETION LOCATION Y/N C*iVIRON ENT RENARKS J
i.,
. y
- Doc y Homt I;emoveP/ t_ow Pressure In.loction r
,DH 1-DP T 1&2 DH Flow Indication,
DH Pump' N
.H-4' The DH pump flow indication le required in orderH I
l Veult
~
to belarce LPI flow thru both core flood lines f
for the postulated eingle LPI pump failure with DPT-812, 803
~0H Flow Indication Aux Bldg Y.
H-1 the SEOCA located at one of the core ficod -
i (on Remote Shutdown. Pane 1) f' (A-1) nozzles on the reactor vessel. DH 1 OPT 1 & 2 provide the control room indication which would be used unless these instrumente failed. If DH 1-DPT 1 & 2 showed no change as the DH flow '
j were being throttled, then the operator would check DPT-802 & 803 to confirm the flow and i
provide him with sufficient information to belance LPI flows. Therefore, only one pair of the instrumente needs to be qualified for SEOCA postulated radiation.
l LS-116 -
H-4 The R8 Sump level indication le not required for A,8,C,D &E SEOCA mitigation. The R8 sump will contain
- end/cr i
sufficient inventory to provide the needs of the
H-4 LPI pumps once the BWST hee been depleted. The 8
.i
- 805 level instrumente are not interlocked with the j
- 80 6 LPI pumpe or the RB sump isolation valves
- 807 DH-V-6A&B.: If it le deelred to provide and 1
inject more BWST water than the initial 8WST i
inventory, then en approximation of the R8 water
}
level can easily be made to establish a limit on the amount of additional water that may be i
injected before the RB flood level for instrumentation is reached. Such action could be taken to delay the time et which recirculation j
from the RB susy is initiated in order to reduce potential radiation levela in the Auxiliary
- Building, i
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1210-6 l
Se:ll Brylt LOCA Castdown h
HPl Initiated V;ri fy All RCPS Off i
s
_EFW Controlling OTSC at 95Z Varlfy Rx Trip contatsusent Isola ti on h
I f RCS Pressure
- Yes --*- Vert fy & Reac ti vate 4 1600 psig 1600 psis &
No 4
4 psig RBI l
t i f 01SC l a Aval l.
---* Ye s ---> 1210-10
& P/S NT XFER Does Not Emlet Bump RCPS No l'
If P/S HT XFER Cr.n not Be Reestablished --> Yes + Open PottV ---> Unill P/S HT XFER Established
& Cooldown 4.100*F/HR or LPI Operation V2rl fy R5 Emer. Cooling 1'
M:Intain RCP Services i,
If RCS la Solid &
l' Go to 1210-9 Subcooled & --> Yes P/S HT XFER is Avall.
to Reestablish PZR Bubble No l
1E If Sol i s Regained --> Yes --* Close CFV 1 A/B Een RCS Press.4 700 psig No t
Moni tor 566T Level it Prior to Colng on Susp Rect re.
Cpen DHV64 & MUVl98
{ Open DHV7A l
Is LP! Flow
- Open DHV6A men BWST Level
> Established?
- N o ----*- Go to Piggyback <
Close DHVSA & BSV2A -P Reaches 3ft.
(0 pen MUV14A)
Yes
- Then do same for ' B' it No Go To LPI I Open DH-V-6 Close DH-V-5 z
Susp Rect re j'Close BS-V-2 l'
Mode Stop EFW Een LPI or DHR In Service f
Throttle DHVl9A/B or DH-V-4A/B as Necess.
I f DC-V-J & 65 a re inaccessible Use DR-VI-A/B to Control Cooldown Rate i
1210-7 f
y l
j l4rg-) Brrkj i
lLOCA Crsldoifn
/
If Vertfy HPI, LPI, CF Vertfy Yotal Rg1 & Cooling 3,
t Verify 1000 Citt LPI Plow / Leg; Open DHV 38A/S as Necess.
34 0 qi
- Co to 121&6 Is RCS Pressure
> LPI Pump Shutof f
.2 Continue vith LB LOCA Procedure Since Break is Larger than 0.5 f t.
1210-9 HP1 Cooling - Recovery Frae Solid Operations E
1.
Start 2 HPI Pumps 2.
Verify HP! Flow Per Enclosure I f SQt i s Regai ned ---> Yes --* 1h r ot t le HPI & S ta rt One RCP l'
No 1!
Attempt to Regain P/S HT XFER Refer to 1210-4 ii
No If sot gegained & RCS Press % Yes ---> Hay Close CFVI A/O 700 pets No If OTSC Heat Removal Capabilty --* Yes ---* Recover Fran Solid Plt Ops
- Using Procedure 1210-9 Sect. 9 Exists Na
' I e Co to 1210-6 I f sot is not Restored or --* Yes St LOCA A Known RCi Break Exists
O APPENDIX "C" a
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6 S
I h
I
{
h a-j l
l
~.,............ - -. - -... -.. - -. -..
e JUSTIFICATION FOR NOT REQUIRING RADI ATION OUALIPTING OF MU23-DPT I, 2, 3. 4 er MU42-DPT s
IS Rf'I NO WITHOUT MU-V-16's OPEN, MU PIMP RUNOUT CAN NOT OCCUR ACTUATED?
a TES i i IS ONLY ONE NO WITH T1D OR MORE MU PUNPS OPERATING, RUNOUT CAN NOT MU PIMP OPERATINCf OCCUR BECAUSE OF 1HE CAVITATING VEN1URIS.
TES I f J
IS RG PRSSSURE NO RUNOUT CAN NOT OCCUR UI1H RG PRESSURE > 1000 PSIC (1000 PSICT DURINC PICCY BACK OPERATION OR SUCTION PROM 1HE BWST TES i t IS SEAL IIUECTION NO
~l WITM MU-V-20 CLOSED 1HE CAVITATING pN1URIS LIMIT THE MU PLMP 10 4 550 CPM PLOW INDICA 10N CLOSE MU-V-20 (MU-42-DFT)
AVAILABLE i
l YES 1 1 IS NPI PLOW INDICA-TION AVAILASLE FOR NO IF SEAL INJECTION PLOW 80 0 THE OPERATING MU
> 80 CPM, CAN MUV32 SE IP MU-V-32 IS INOPERABLE, CLOSE MU-V-20 PtMPT (MU-2 3-DPT THROTTLED 10 bio CPM 7 I/2 or 3/4 188 it
.j W11H SEAL INJECTION PLOW 6 to CPM 1DTAL MU PUMP PLOW WILL BE < 360 CPM 1EST DATA, TP-655/I SHOW THIS ACCEPTABLE YES
' r i
1 1HROTTLE MU-V-16's AS REQUIRED 10 MAINTAIN MU PIMP PLOW (.550 CfM 1
1 i