DCL-93-114, License Amend Request 93-01 to Licenses DPR-80 & DPR-82, Changing TS 3/4.3.3.5, Remote Shutdown Instrumentation, to Add Remote Shutdown Control Functions & Increase Allowed Outage Time for Inoperable Function from 7 to 30 Days
| ML16342B979 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/07/1993 |
| From: | Rueger G PACIFIC GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML16342A106 | List: |
| References | |
| DCL-93-114, NUDOCS 9305200318 | |
| Download: ML16342B979 (44) | |
Text
PG
~ Letter No. DCL-93-114 ENCLOSURE UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
)
In the Hatter of
)
PACIFIC GAS AND ELECTRIC COMPANY )
)
Diablo Canyon Power Plant
)
Units 1 and 2
)
Docket No. 50-275 Facility Operating License No.
DPR-80 Docket No. 50-323 Facility Operating License No.
DPR-82 License Amendment Request No. 93-01 Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) hereby applies to amend its Diablo Canyon Power Plant (DCPP) Facility Operating License Nos.
The proposed changes amend the Technical Specifications (TS)
(Appendix A of the Licenses) regarding TS 3/4.3.3.5, "Remote Shutdown Instrumentation,"
and associated Basis.
Information on the proposed changes is provided in Attachments A and B.
These changes have been reviewed and are considered not to involve a
significant hazards consideration as defined in 10 CFR 50.92 and not to require an environmental assessment in accordance with 10 CFR 51.22(b).
Further, there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes.
Sincerely, Grego y H. Rueger Subscribed and sworn to before me this 7th day of Hay 1993.
Attorneys for Pacific Gas and Electric Company Howard V. Golub Christopher J.
Warner Mildred J.
illiams, Notary Public QISJINEI)mllllllll)!Ill)fallBSC(iUat15'is4,"aJCCRliS MiLDRED j. VllLI.IAMS NOTARY PUBLiC - CALIFORNIA""
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PG i Letter No. DCL-93-114 ATTACHMENT A REVISION OF TECHNICAL SPECIFICATION 3/4.3.3.5 REVISE TECHNICAL SPECIFICATION FOR REMOTE SHUTDOWN INSTRUMENTATION A.
DESCRIPTION OF AMENDMENT REQUEST This license amendment request (LAR) proposes to change Technical Specification (TS) 3/4.3.3.5, "Remote Shutdown Instrumentation," to include additional control functions required to establish and maintain Mode 3 (Hot Standby) from outside of the control room in accordance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 19 and the Westinghouse Standard Technical Specifications (STS) located in NUREG-1431.
The proposed changes are as follows:
TS 3.3.3.5 is revised as follows:
a.
The TS title is changed from "Remote Shutdown Instrumentation" to "Remote Shutdown Instrumentation and Controls."
b, The list of remote shutdown instrumentation in TS Table 3.3-9 is revised to include the following remote shutdown control functions:
auxiliary feedwater (AFW) flow control, charging pump control, component cooling water (CCW) pump control, auxiliary saltwater (ASW) pump control, and emergency diesel generator (EDG) control.
c.
The list of remote shutdown instrumentation in TS Table 3.3-9 is revised to include reactor coolant system (RCS)
Loop 1 hot and cold leg temperature indicators.
d.
Emergency borate flow indication is deleted from the list of instrumentation in TS Table 3.3-9.
e.
Editorial changes are made throughout the TS to reflect the inclusion of the control functions required for remote shutdown.
2.
3.
Surveillance Requirement 4.3.3.5 is renumbered to 4.3.3.5.
1 and a new Surveillance Requirement 4.3.3.5.2 is added to verify that each required control circuit and transfer switch is capable of performing the intended function at least once every 18 months.
Action statement (a) is revised to increase the Allowed Outage Time (AOT) from 7 days to 30 days.
Action Statement (c) is added to clarify that separate condition entry is allowed for each function listed in Table 3.3-9.
6100 S/85K
4.
The associated TS Basis is expanded to be consistent with NUREG-1431.
Changes to the TS and Basis are noted in the marked-up copy of the applicable TS (Attachment B).
B.
BACKGROUND The remote shutdown instrumentation and controls provide the control room operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room.
This capability is necessary in the event that the control room must be evacuated.
For TS 3.3.3.5, a safe shutdown condition is defined as Mode 3.
With the'nit in Mode 3, the AFW system and the steam generator (SG) safety valves can be used to remove core decay heat and meet all safety requirements.
The long term supply of water for the AFW system allows extended operation in Mode 3 from outside the control room until such a
time that either control is transferred back to the control room or a cooldown is initiated.
In addition to being available in the control room, the primary instrumentation and control functions required to establish and maintain Mode 3 are located at the hot shutdown panel (HSP), with the exception of the:
(1) reactor trip indication, which is located at the reactor trip switchgear (2)
EDG local start control, which is located at each EDG control panel (3)
RCS Loop 1 hot and cold leg indicators, which are located at the dedicated shutdown panel The criteria governing the design and specific system requirements for remote shutdown instrumentation and controls are contained in GDC 19.
Operability of the remote shutdown instrumentation assures that there is sufficient information available on selected unit parameters to place and maintain the unit in a safe shutdown condition.
In accordance with the current Diablo Canyon Power Plant (DCPP)
TS 3.3.3.5 Basis, the HSP is designed to maintain the reactor in Mode 3.
The specific instrument channels which are required to be operable per the current DCPP TS 3.3.3.5 (Table 3.3-9) are as follows:
(1)
(2)
(3)
(4)
(5)
(6)
(7)
Reactor Trip Breaker Indication Pressurizer Pressure Indication Pressurizer Level Indication SG Pressure Indication SG Wide Range Water Level Indication Condensate Storage Tank Water Level Indication AFW Flow Indication 6100S/85K
(8)
Emergency Borate Flow Indication (9)
Charging Flow Indication All of the above instrumentation, except for the reactor trip breaker indication, is located at the HSP.
Reactor trip breaker indication is displayed at the reactor trip breaker.
In addition to the indicators located at the HSP, the HSP provides for remote control of the following functions.
As discussed later in the Safety Evaluation Section (Section D) of this LAR, not all of these functions are required to establish and maintain Mode 3.
Currently, these functions are not included in TS 3.3.3.5.
(1)
AFW Flow Control (pumps and valves)
(2)
Charging Flow Control (pumps and valves)
(3)
Emergency Borate Flow Control (pumps and valves)
(4)
CCW Pumps (5)
ASW Pumps (6)
Containment Fan Coolers (7)
Pressurizer Power Operated Relief Valves (PORVs) (close only)
(8) 10% Atmospheric Steam Dump Valves (ADVs) (open and close)
(9)
Pressurizer Heaters (10)
Letdown Orifice Isolation Valves In summary, the current TS 3.3.3.5 controls the instrumentation located at the HSP that is required to monitor operation in Mode 3 from a location outside the control room.
C.
JUSTIFICATION On June 1,
1992, interim guidance was provided to NRR Division Directors in a memorandum titled, "Screening of License Amendment Requests and Processing of Generic Technical Specification Changes."
The memorandum addressed the screening of LARs and processing of generic TS changes, and allows for the partial implementation of the revised STS located in NUREG-1431.
PGSE is proposing to use this guidance to request a license amendment to implement the TS changes of the remote shutdown instrumentation and controls'pecific justification for these changes is listed below.
Remote Shutdown Instrumentation and Control Functions PG&E is proposing the addition of the following remote shutdown instrumentation and control functions to the list of instrumentation presented in TS Table 3.3-9:
(1)
AFW Flow Control (2)
Charging Pump Control (3)
CCW Pump Control (4)
ASW Pump Control (5)
EDG Local Start Control (6)
RCS Loop 1 Hot and Cold Leg Temperature Indication 6100S/85K
Section 7 of the DCPP Units 1
and 2 Updated Final Safety Analysis Report (UFSAR) includes the above control functions as part of the remote safe shutdown systems.
TS 3.3.4, "Remote Shutdown System," of the STS located in NUREG-1431 contains surveillance requirements for remote shutdown control functions along with the instrumentation monitoring functions.
As such, the addition of the above remote shutdown control functions and the associated surveillance requirements is consistent with the description of the remote safe shutdown system presented in the UFSAR and with TS 3.3.4 of the STS located in NUREG-1431.
Although emergency borate flow indication is currently included in the TS, PGKE has determined that emergency borate flow is not required to maintain and establish Node 3.
As such, PGRE proposes to delete emergency borate flow indication from TS Table 3.3-9 and TS Table 4.3-6.
The deletion of the emergency borate flow control is consistent with TS 3.3.4 of the STS located in NUREG-1431.
Increased Allowed Outa e Time Increasing the AOT could avoid unnecessary plant transients and plant shutdowns if operability cannot be restored within 7 days, but could be restored within 30 days.
PG&E is proposing to revise the allowed outage time in DCPP TS 3.3.3.5 from 7 days to 30 days.
The allowed outage time of 30 days is currently licensed at the North Anna and D.
C.
Cook plants, and included in TS 3.3.4 of the STS located in NUREG-1431.
The equipment associated with remote shutdown can generally be repaired within the 30 days and can be repaired during operation without significant risk of spurious trip.
Se grate Entr into Action Statement An action statement is added to clarify that separate entry into Action Statement (a) of TS 3.3.3.5 is allowed for each function (instrumentation or control) listed in Table 3.3-9.
This added statement is consistent with TS 3.3.4 of the STS located in NUREG-1431.
Enhanced Bases This revision to TS 3.3.3.5 incorporates the enhanced bases of the STS located in NUREG-1431.
A principal objective of the enhanced bases is to provide a comprehensive explanation of the safety significance of TS with respect to the accident analyses performed for the facility.
The enhanced bases also provide a complete background for the specification, contain a brief description of the system, and establish a baseline for future specification changes.
D.
SAFE'TY EVALUATION Remote Shutdown Control and Instrumentation The TS remote shutdown instrumentation and control functions provide the ability to establish and maintain operation in Mode 3 from outside the control room in the event that the control room must be evacuated.
The equipment added to the TS is currently included in the DCPP surveillance 6100 S/85K
0 l
test program.
The surveillance test program for this equipment includes the starting of the equipment from the HSP.
Inclusion of this equipment in the TS provides additional restrictions to assure that it is available to establish and maintain the unit in Mode 3.
In order to establish and maintain Mode 3 from outside the control
- room, the reactor must be tripped, decay heat must be removed, and RCS temperature,
- pressure, and inventory must be controlled.
Additionally, systems required to support equipment performing these functions must be operable.
The following provides discussion of the minimum functions required to establish and maintain Mode 3 from outside the control room until a cooldown is initiated or control is transferred back to the control room.
Reactor Trip Core subcriticality is achieved by tripping the reactor.
The reactor can be tripped from outside the control room by opening the reactor trip breakers at the reactor trip switchgear.
Reactor trip indication is provided from outside the control room by the reactor trip breaker position.
The insertion of the control rods during a reactor trip provides the negative reactivity needed to establish and maintain Mode 3 until such time that either control is transferred back to the control room or a cooldown is initiated.
Reactor trip breaker position indication is currently included in TS 3.3.3.5.
Decay Heat Removal via the AFW System and the SG Safety Valves Heat removal from the RCS is accomplished by transferring heat to the secondary plant through the SGs.
The decay heat is then removed from the SGs via boiling and steam release through the SG code safety valves.
Indication of the secondary side heat sink is provided by SG pressure indication (one per SG),
SG wide range water level indication (one per SG),
and AFW flow indication (one per SG) located at the HSP.
The HSP also provides indication of the condensate storage tank level to allow monitoring of water available to supply the suction of the AFW pumps for extended operation in Mode 3.
These functions are currently located in TS 3.3.3.5.
In order to assure that SG level remains within its expected
Upon initiation of a reactor trip, SG level will decrease due to shrink and the trip of the main feedwater pumps.
The AFW pumps will supply feedwater to the SGs to compensate for the loss of main feedwater.
After the level in the SGs recovers, the feedwater supply to the SGs must be controlled to prevent the SG from overfilling and overcooling the RCS, which could result in a safety injection.
The feedwater flow can be controlled from the HSP using the AFW level control valves or by starting and stopping AFW pumps.
The addition of AFW pump and level control valve 6100 S/85K
controls to TS 3.3.3.5 is consistent with TS 3.3.4 of the STS located in NUREG-1431.
In order to monitor the rate of heat removal from the core during all plant conditions, including a loss of offsite power, indications of RCS hot and cold leg temperatures are required.
Loop 1
RCS hot and cold leg temperature indication is available at the dedicated shutdown panel.
The addition of these indicators to TS 3.3.3.5 is consistent with the STS located in NUREG-1431.
RCS Pressure Control Indication of RCS pressure is provided by the pressurizer pressure indication located at the HSP.
This indication is currently required by TS 3.3.3.5.
RCS overpressure protection is provided by the pressurizer code safety valves.
Although pressurizer heaters would assist in controlling RCS pressure, they are not required to maintain pressure control of the RCS.
RCS Inventory Control via Charging Flow Indication of RCS inventory is provided by the pressurizer level indication located at the HSP.
Level control of the RCS is necessary to prevent the loss of level in the pressurizer and the subsequent loss of pressure control of the RCS, to prevent the RCS from achieving a solid water condition where pressure would no longer be readily controllable, and to prevent the core from being uncovered due to low level.
This indication is currently included in TS 3.3.3.5.
The HSP contains controls to start and stop each centrifugal charging pump (CCP).
The charging pumps not only supply water to the RCS for pressurizer level control, but also provide water to the reactor coolant pump (RCP) seals.
By starting and stopping the
- CCPs, pressurizer level can be controlled.
During any time when the CCPs are shut off, RCP seal degradation would be prevented by reactor coolant flowing past the thermal barrier heat exchanger, which is cooled by CCW flow, and out of the RCP seals, This would also remove water injected into the RCS that may have caused an increase in pressurizer level.
The addition of charging pump controls to TS 3.3.3.5 is consistent with TS 3.3.4 of the STS located in NUREG-1431.
Safety Support Systems In order for the above equipment to perform its intended safety function, it must have power and be cooled.
Heat removal can be accomplished via the CCW and ASW systems.
The CCW system removes heat from the lube oil and seals of the engineered safety feature (ESF) pumps.
The ASW system removes heat from the CCW system and rejects it to the ultimate heat sink.
Both the CCW pumps and the ASW pumps can be started from the HSP.
Inclusion of the CCW and ASW pumps is an additional restriction not in the STS.
Although the CCW and ASW pumps are normally in operation and are designed 6100S/85K
to auto start, inclusion of the pump controls at the HSP assures that the pumps are available in the event that they don't start automatically, and emphasizes the importance of the function of the pumps.
To assure that power is available to ESF equipment, EDGs are available to supply power in the event that offsite power is unavailable.
Although the EDG should auto-start during a loss of offsite power, the addition of the starting control function to TS 3.3.3.5 provides additional assurance that power will be available to the ESF equipment required to establish and maintain Mode 3.
Inclusion of the EDGs is an additional restriction not included in the
Although the HSP also contains controls to manipulate the emergency borate flow, 10% ADVs, the containment fan coolers, pressurizer
- heaters, PORVs, and letdown orifice isolation valves, the remote control of these functions is not added to TS 3.3.3.5.
UFSAR Section 7.4 identifies those accidents which would result in the most severe consequences during a remote shutdown.
Based on a review of the accidents identified in UFSAR Section 7.4, the remote control of the emergency borate flow, 10% ADVs, containment fan coolers, pressurizer
- heaters, PORVs, and letdown orifice isolation valves is an operational convenience and not required to mitigate the consequences of an accident.
Consequently, these functions are not required to be included in TS 3.3.3.5.
The above evaluation has shown that with the minimum equipment previously discussed, the reactor can be maintained in a safe condition.
Additional equipment is provided at the HSP, but is not required to be available to establish and maintain Node 3.
Increased Allowed Outa e Time The TS 3.3.3.5 Remote Shutdown Instrumentation and Control functions were evaluated in accordance with the NRC Interim Policy Statement on Technical Specification Improvement.
In accordance with the evaluation, TS 3.3.3.5 was identified as a candidate for removal from the TS and relocation to other plant documents.
However, this TS has not been proposed for relocation.
Since the evaluation determined that the TS was a candidate for relocation, the continued inclusion of remote shutdown instrumentation and controls in the TS with a 30 day AOT, instead of a 7 day AOT, is conservative as compared to the evaluation performed in accordance with the NRC Interim Policy Statement for Exclusion.
In addition, the probability of GDC 19 scenarios requiring remote shutdown capability is considered low.
Control of several components located at the HSP that are required for safe shutdown can be accomplished at other locations (e.g.,
at the 4
kV switchgear, a motor control center, a local control panel, or manually at a valve).
Alternate instrumentation for monitoring selected parameters required for safe shutdown is included at the dedicated shutdown panel.
These parameters include:
(1)
RCS wide range pressure; (2) cold calibrated pressurizer level; and (3) cold calibrated narrow 6100S/85K
range SG level. Therefore, other means exist to monitor and control reactor conditions besides the indications and control functions located at the HSP.
The challenges and risks to plant systems caused by a plant shutdown as a result of the current 7-day AOT in TS 3.3.3.5 are considered to be more significant than continued operation beyond 7 days with a channel inoperable.
This is supported by the above discussion.
The above evaluation has shown that an allowed outage time of 30 days for remote shutdown monitoring and control functions is acceptable.
This allowed outage time is consistent with NUREG-1431
- and, as such, has been evaluated and found to be acceptable by the NRC.
Also, the probability of an event that would require evacuation of the control room is considered low.
Se grate Entr into Action Statement and Enhanced Bases The proposed Action Statement (c) provides clarification allowing separate entry in Action Statement (a) for each instrument and control function listed in Table 3.3-9.
A principal objective of the enhanced Bases is to provide a
comprehensive explanation of the safety significance of TS with respect to the accident analyses performed for the facility.
The enhanced bases also provide a complete background for the specification, contain a
brief description of the system, and establish a baseline for future specification changes.
The emphasis of the enhanced bases is to explain why the requirements of the TS are important to safety, and how the TS assures that the initial conditions assumed during accident conditions exist.
The addition of both the Action Statement (c) and the Enhanced Bases are administrative in nature and do not impact the safety of the plant.
Conclusion E.
Based on the above consideration, PG&E believes that there is reasonable assurance that the health and safety of the public will not be affected by the proposed TS changes.
NO SIGNIFICANT HAZARDS PGSE has evaluated the no significant hazards considerations involved with the proposed amendment, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a facility licensed under paragraph 50.22 or a testing facility involves no significant hazards consideration, 6100S/85K
if operation of the facility in accordance with the proposed amendment would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)
Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
The following evaluation is provided for the no significant hazards consideration.
Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes to TS 3.3.3.5 do not alter the plant configuration or operation.
The inclusion of remote shutdown control functions constitute additional restrictions over the remote shutdown system.
Since the remote shutdown instrumentation and controls are not part of the primary success path to mitigate a design-basis accident or transient, which either assumes the failure of or challenge to the integrity of a fission product
- barrier, and since the probability of an event that would require evacuation of the control room is low, the 30-day AOT is acceptable.
Allowance of separate entry into the Action Statement for each instrument and control function listed in Table 3.3-9 and the incorporation of enhanced bases are administrative in nature and do not impact the safety of the plant.
Therefore, the proposed changes to TS 3.3.3.5 do not have a significant affect on the probability or consequences of any previously evaluated accident.
2.
3.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes to TS 3.3.3.5 do not require physical alteration to any plant system or change the method by which any safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of safety?
The proposed changes to TS 3.3.3.5 will not change any assumptions, initial conditions, or results of any accident analysis.
Consequently, the changes do not involve a significant reduction in a margin of safety.
6100S/85K
I t
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION G.
Based on the above safety evaluation, PGEE concludes that the changes associated with this LAR satisfy the no significant hazards consideration standards of 10 CFR 50.92(c) and, accordingly, a no significant hazards finding is justified.
ENVIRONMENTAL EVALUATION PG&E has evaluated the proposed changes and determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
an environmental assessment of the proposed change is not required.
6100S/85K
PG Letter No. DCL-93-114 ATTACHMENT B MARKED-UP TECHNICAL SPECIFICATIONS Remove Pa e
V1 3/4 3-47 3/4 3-48 3/4 3-49 B 3/4 3-3 Insert Pa e
V1 3/4 3-47 3/4 3-48 3/4 3-49 B 3/4 3-3 B 3/4 3-3a to 3-3d 6100S/85K
- ll-
NDEX LIHITIHG CONDITIONS FOR OPERATION AND'URVEILLANCE RE UIREHENTS SECTION 3/4. 3 IHSTRUMENTATIOH cont5nued TABLE 3.3 4 ENGINEERED SAFFTY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SKTPOINTS.
~
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oooo oo 3/4 3 23 TABLE 3.3-5 EHGIHEERED SAFETY FEATVRES RESPONSE TIMES
~
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3/4 3 28 TABLE 4.3 2 EHGIHEERED SAFETY FEATURES ACTVATION SYSTEN INSTRUMENTATION SURVEILLANCE REQUIREHENTS.
~ ~ ~
o
~ ~
~ ~
~
3/4 3 32 3/4-3. 3 HOHITORING IHSTRUMENTATION
~
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Radfat5on Mon5tor5ng for Plant Operat5ons.........
3/4 3-36 TABLE 3.3 6
RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ oo ~ o ~ ~
o ~ ~ ~ oo ~ ~ ~ ~ oo ~ ~ ~ ~ oooo'/4 3 37 TABLE 4.3 3 RADIATION HQNITORIHG IHSTRUMEHTATIOH FOR PLANT OPERATIONS SURVEIL,LANCE REQUIREMENTS.... ~ ~ ~ ~. ~ ~ ~
. ~
3/4 3 39 Movable Incore Detectorso
~ ~ oooooo ~ ~ ~ ~ ~ ~ ~ o ~ oo ~ ~
o ~ ~ ~ oo ~ ooo ~
3/4 3 40 Seismic Instrumentat50no boffo ~ ~ ~ ~ ~ oo ~ ~ ~ ~
~ o ~ ~ ~ o ~ ~ ~ ~ ooo ~ ~ ~ o 3/4 3 41 TABLE 3.8-7 SEISMIC MONITORING INSTRUMENTATION...........~. ~. ~ ~ ~.
3/4 3 42 TABLE 4.3 4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS. ~ ~ ~
~
~ ~ ~
~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~..
3/4 3 43 Meteorological Instrumentat5on.....;...................
3/4 3-44
. TABLE 3.3-8 HETEOROLOGICAL MONITORING INSTRUMENTATION.~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~
3/4 3 45 TABLE 4. 3-5 HETEOROLOGICAL HONITORIHG INSTRUMENTATION SURVEILLANCE REQVIREMENTS
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ oooooo ~ ~ ~ ~ ~ oo ~ oo
~
~ ~ ooooo IS/4 3 46 CAn Ck CO~+rO15 Remote Shutdown Instrumentat5on............
~.............
TABLE 3.3-9 RENOTE SHUTOOWN NONITORING INSTRUHENTATION~~
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3/S 3 RO AIUl3
~DMWQ~L,Q TABLE 4.3-6 REMOTE SHUTDOMH HONITORIHG INSTRUMENTATION SLlRVEILLAHCE REQVIREMENTSi ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ oooo ~ ~ ~ ~ ~ ~ ~
~ os 3/4 3 49 ACC5dent Hon5 tor5 ng Ins trumentat5ono
~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~
3/4 3 50 TABLE 3-3 10 ACCIDENT MONITORING IHSTRVMENTATION.
~ ~. ~ ~ ~. ~.. ~..
3/4 3-52 TABLE 4. 3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
~ ~
~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~
~ ~ ~
DIABLO CANYON - UNITS 1 ls 2 V5 930520032i 930507 PDR ADOCK 05000275, P
PDR 3/4 3 53
INSTRUMENTATIO REMOTE SHUTDOWN INSTRUMENTATION sA,ug Coe~oCS LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitorin strumentatio Table 3:3-9 shall be OPFRABL.
e dye s
d APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
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t t e n mbe of OPE BL re te hut own oni ri c
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The provisions of Specification 3.0.4 are.not applicable.
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SURYEILLANCE RE UIREMENTS 4.3.3.5'ach remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of'he CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-6.
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DIABLO CANYON - UNITS 1 8I 2
. 3/4 3"47
A 1S
, INSERT A With less than the minimum required Function(s) of Table 3.3-9 operable, restore the inoperable Function(s) to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I 2
TABLE 3.3-9 EMOTE SHUTDOWN MONITOR INSTRUMENTATION PtND CONT'0 OL5 INSTRUMENT c,c emoL l=0 re cZ lad READO r cev~ol LOCATION R E@,v ARE.D Qg gage OP cH~NlV~
1.
Reactor. Trip.Breaker Indication
- 2. Pressurizer Pressure
- 3. Pressurizer Level 4.
Steam Generator Pressure 5.
Steam Generator Wide Range Water Level 6.
Condensate Storage Tank Water Level
- 7. Auxiliary Feedwater Flow er e
or e
. Io Charging Flow Reactor Trip Breaker.
Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel o
Sh do P
1 Hot Shutdown Panel 1/trip breaker 1/stm.
gen.
1/stm.
gen.
1/stm.
gen.
9 pCS Loop I Temperature Indication A
Dedicated Shutdown Panel Hot and Cold Leg Temper atur e Indication 10.
Auxiliary Feedwater Flow Control AFW Pump, and Associated'Valves Transfer Switches ll.
Charging Flow Control Centrifugal Charging Pump Transfer Switch 12.
Component Cooling Mater Control Component Cooling Water Pump Transfer Switch 13.
Auxiliary Saltwater Control Auxiliary Saltwater Pump Transfer Switch 14.
Emergency Diesel Generator Control EDG Start Hot Shutdown Panel 4 kV Switchgear Hot Shutdown Panel 4 kV Switchgear Hot Shutdown Panel 4 kV Switchgear Hot Shutdown Panel 4 kV Switchgear EDG Local Control Panel any 2 of 3 AFW pumps 2 of 2 pumps any 2 of 3 CCW pumps 2 of 2 pumps 3 of 3 EDGs DIABLO CANYON " UNITS 1 & 2 3/4 3-48
.a
INSTROMENT TABLE 4. 3-6 MOTE SHUTOOWN MONITORING INST ENTATION
.SURVEILLANCE RE UIREMENTS CHANNEL CHECK CHANNEL CALIBRATION 1.
Reactor Trip Breaker Indication 2.
Pressurizer Pressure 3:
Pressurizer Level N. A.
N. A.
4.
Steam Generator Wide Range Water Level M
5.
Steam Generator Pressure 6.
Condensate Storage Tank Water Level 7.
Auxiliary Feedwater Flow E er ncy o at Fl g
P Charging Flow M//
M R
- g. CSLOO ICR I
%~Pa WC'+MVe. ~~i ~+>><
CIABLD CANYON - UNITS 1
8, 2
3/4 3-49
INSTRUMENTATION BASES 3/4. 3.3. 3 SEISMIC INSTRUMENTATION The OPERABILITY of the sefimfc fnstrumentatfon ensures that sufficient capability fs avaflablc to promptly determine the Nagnftude of a seismic event and evaluate the rcsponsc of those features important to safety.
This capabflfty fs required to permit comparison of the Neasurcd response to that used in thc desfgn basis for the facfTfty to deterifne ff plant shutdown fs required put suant ta Appendix h of 10 CFR Part 100.
The fnstrumentatfon fs consistent <<0th the recaenendatfons of Regulato~ Guide 1.12, "Instrumentation for Earthquakes.".
3/4.3.3. 4 METEOROLOGICAL IHSTRUMEHTATIOH The OPERABILITY of the metearologfcal instrumentation ensures that sufffcfent metcarolagfcal data fs avaflablc for estimating potential radiation doses to the public as a result of routine or accidental rclcase of radioactive materials to the atmasphere.
This capability fs required to evaluate the need for initiating protective measures ta protect the health and safety of the public and fs cansfstent with the rccammendatfons of Regulatory Guide 1.23, "Onsfte Metearolagf cal Pt ograms," February 1972.
3/4. 3. 3. 5 REMOTE SHUTDOWN INSTRUMENTATION Th OPER LITY o the mote s daw fns enta n cns es tha uffi ent pabilf fs a flable a p ft s utdawn nd maf enance f HOT ST BY o the f lfty om loc ion outs e of e cont roam.
This pabil y fs quire n the en contr@
roa abftab fty fs ost a fs consi ent w
h Gen al Desi n Cr teria'9 o
0 CFR art 50.
3/4. 3. 3. 6 ACCIDENT MONITORING INSTRUMENTATION
)~SEW~ 8 The OPERABILITY of the accident monitoring fnstrumentatfon ensures that sufffcfent information fs available on selcctcd plant parameters to monitor and assess these variables following an accfdent.
The normal.plant fnstrumcnt channels specified arc suitable for use as post-accident instruments.
This capability fs consistent with the rccammendatfons of Regulatory Guide 1.97, Revfsfon 3, "Instrumentatfan for Lfght-Mater Cooled Nuclear Power Plants to Assess Plant Canditfans During and Following an Accfdent,4 May 1983, and NUREG-0737, "Clarfffcatfon of 1MI Action Plan Requirements,"
November 1980.
3/4.3.3.7 CHLOR NE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection System ensures that sufficient capability fs avaflablc to promptly detect and initiate protectfve action fn the event of an accidental chlorine release.
This capability fs requfred to protect control roam personnel and fs consistent with the recommendatfons of Regulatory Guide 1.95, "Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release,"
February 1975.
DIABLO CANYON - UNITS 1 5 2 B 3/4 3-3.
E
, 2 Ae
INSERT B
BACKGROUND The Remote Shutdown Instrumentation and Controls provide the control room operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room.
This capability is necessary to protect against the possibility that the control room becomes inaccessible.
A safe shutdown condition is defined as NODE 3.
With the unit in MODE 3, the Auxiliary Feedwater (AFW) System and the steam generator (SG) safety valves can be used to remove core decay heat and meet all safety requirements.
The long term supply of water for the AFW System allows extended operation in MODE 3 from outside the control room until such a time that either control is transferred back to the control room or a
cooldown is initiated.
In the event that the control room becomes inaccessible, the operators can establish control at the remote shutdown panel (hot shutdown panel),
and place and maintain the unit in MODE 3.
Not all controls and necessary transfer switches are located at the hot shutdown panel.
Some controls and transfer switches will have to be operated locally at the switchgear, motor control panels, or other local stations.
The unit automatically reaches MODE 3 following a unit shutdown and can be maintained safely in MODE 3 for an extended period of time.
The OPERABILITY of the remote shutdown control and instrumentation functions ensures there is sufficient information available on selected unit parameters to place and maintain the unit in MODE 3 should the control room become inaccessible.
APPLICABLE SAFETY ANALYSES The Remote Shutdown Instrumentation and Controls provides equipment at appropriate locations outside the control room with a capability to promptly shut down and maintain the unit in a safe condition in MODE 3.
The criteria governing the design and specific system requirements of the Remote Shutdown Instrumentation and Controls are located in 10 CFR 50, Appendix A, GDC 19.
LCO The Remote Shutdown Instrumentation and Controls LCO provides the OPERABILITY requirements of the instrumentation and controls necessary to place and maintain the unit in MODE 3 from a location other than the control room.
The instrumentation and controls required are listed in Table 3.3-9 in the accompanying LCO.
The controls, instrumentation, and transfer switches are required for:
~
Reactor trip indication;
~
RCS pressure control;
~
Decay heat removal via the AFW System and the SG safety valves;
~
RCS inventory control via charging flow; and
~
Safety support systems for the above Functions, including auxiliary saltwater, component cooling water, and emergency diesel generators.
A remote shutdown Function is OPERABLE if all required instrument and control channels for that function listed in Table 3.3-9 are OPERABLE.
The remote shutdown instrument and control circuits covered by this LCO do not need to be energized to be considered OPERABLE.
This LCO is intended to ensure the instruments and control circuits will be OPERABLE if unit conditions require that a remote shutdown be performed.
APPLICABILITY The Remote Shutdown Instrumentation and Controls LCO is applicable in MODES I, 2, and 3.
This is required so that the unit can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room until either control is transferred back to the control room or a cooldown is initiated.
This LCO is not applicable in MODE 4, 5, or 6.
In these MODES, the facility is already subcritical and in a condition of reduced RCS energy.
Under these conditions, considerable time is available to restore necessary instrument control functions if control room instruments or controls become unavailable.
ACTIONS Action a.
Action a.
addresses the situation where one or more required Functions (instrument or control) of the Remote Shutdown Instrumentation and Controls are inoperable.
This includes any Function listed in Table 3.3-9, as well as the control and transfer switches.
The Required Action (Action a.) is to restore the required Function to OPERABLE status within 30 days.
The Allowed Outage Time (AOT) is based on operating experience and the low probability of an event that would require evacuation of the control room.
If the Required Action and associated AOT of Action a is not met, the unit must be brought to a MODE in which the LCO does not apply.
To achieve this status, the unit must be brought to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The AOTs are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
Action b.
Action b. excludes the MODE change restriction of TS 3.0.4.
This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a unit shutdown.
This exception is acceptable due to the low probability of an event requiring remote shutdown and because the equipment can generally be repaired during operation without significant risk of spurious trip.
~s
Action c.
Action c.
has been added to the ACTIONS to clarify the application of AOT rules.
Separate Condition entry is allowed for each Function listed on Table 3'.3-9.
The AOT(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.
SURVEILLANCE REQUIREMENTS SR 4.3.3.5.
1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred.
A CHANNEL CHECK is a comparison of the parameter indicated on one channel to a similar parameter on other channels.
It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious.
CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff based on a
combination of the channel instrument uncertainties, including indication and readability.
If the channels are within the match criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when Surveillance is required, the CHANNEL CHECK will verify only that they are off scale in the same direction.
Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.
The Frequency of 31 days is based upon operating experience which demonstrates that channel failure is rare.
CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.
The test verifies that the channel responds to measured parameters with the necessary range and accuracy.
The Frequency of 18 months is based upon operating experience and consistency with the typical industry refueling cycle.
SR 4.3.3.5.2 SR 4.3.3.5.2 verifies each required Remote Shutdown Instrumentation and Controls control circuit and transfer switch performs the intended function.
This verification is performed from the hot shutdown panel and at other locations for certain control transfer switches, as appropriate.
This will ensure that if the control room becomes inaccessible, the unit can be placed and maintained in NODE 3 from the hot shutdown panel and the local control stations.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
(However, this Surveillance is not required to be performed only during a unit outage.)
Operating experience demonstrates that remote shutdown control channels usually pass the Surveillance test when performed at the 18 month Frequency.
C 0
C gI
NOTE:
A surveillance of the reactor trip breaker OPERABILITY is not required as part of the SURVEILLANCE RE(UIREMENT for 4.3.3.5.2 since a TRIP ACTUATING DEVICE OPERATIONAL TEST of the reactor trip breakers is performed as part of the SURVEILLANCE RE(UIREMENT for TS 3/4.3. 1 (See Table 4.3-1 Item 21 and Note 10).
REFERENCES l.
10 CFR 50, Appendix A, GDC 19.
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