ML20072S600
| ML20072S600 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 09/02/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072S593 | List: |
| References | |
| NUDOCS 9409140145 | |
| Download: ML20072S600 (8) | |
Text
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UNITED STATES
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WASHINGTON. D.C. 20665-0001 p
49.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 94 TO FACILITY OPERATING LICENSE NO DPR-80 AND AMENDMENT NO.
93 TO FACILITY OPERATING LICENSE N0. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By letter of May 7,1993, Pacific Gas and Electric Company (or the licensee) submitted a request for changes to the Technical Specifications (TS) for Diablo Canyon Power Plant (DCPP) Units 1 and 2.
The proposed amendments would revise 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 Part 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:
1.
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.
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Survedlance 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 t'1at each required control circuit and transfer switch is capable of performing the intended function at I
least or.ce every 18 months.
2.
Action Statement (a) is revised to increase the Allowed Outage Time (A0T) from 7 days to 30 days.
3.
Action Statement (c) is added to clarify that separate condition entry is allowed for each function listed in Table 3.3-9.
4.
The associated TS Basis is expanded to be consistent with NUREG-1431.
2.0 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 outside 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 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 addition to being available in the DCPP control room, the primary l
instrumentation and control functions required to establish and maintain Mode 3 are located at the hot shutdown panel (HSP), with the ekception of the:
(1)
Reactor trip indication, which is located at the reactor trip switchgear (2)
Emergency diesel generator (EDG) local start control, which is located at each EDG control panel (3)
Reactor coolant system (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 10 CFR Part'50, Appendix A, General Design Criteria (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
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required to be operable per the current DCPP TS 3.3.3.5 (Table 3.3-9) are as follows:
(1)
Reactor Trip Breaker Indication (2)
Pressurizer Pressure Indication l
(3)
Pressurizer Level Indication l
(4)
SG Pressure Indication (5)
SG Wide Range Water Level Indication l
(6)
Condensate Storage Tank Water Level Indication (7)
AFW Flow Indication i
(8)
Emergency Borate Flow Indication (9)
Charging Flow Indication l
All of the above instrumentation, except for the reactor trip breaker l
indication, is located at the HSP at DCPP.
Reactor trip breaker indication is i
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 and 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.
3.0 EVALUATION The licensee 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
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(3)
CCW Pump Control l
(4)
ASW Pump Control l
(5)
EDG Local Start Control (6)
RCS Loop 1 Hot and Cold Leg Temperature Indication Section 7 of the DCPP Units 1 and 2 Updated Final Safety Analysis Report I
1
l (UFSAR) includes the above control functions as part of the remote safe shutdown systems.
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 DCPP UFSAR.
Although emergency borate flow indication is currently included in the TS, the licensee has determined that emergency borate flow is not required to maintain and establish Mode 3.
As such, the licensee proposes to delete emergency borate flow indication from TS Table 3.3-9 and TS Table 4.3-6.
The deletion i
l of the emergency borate flow control is consistent with TS 3.3.4 of the STS located in NUREG-1431.
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 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 i
additional restrictions to assure that it is available to establish and maintain the unit in Mode 3.
l 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 i
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.
7eactor 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 I
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.
4 Decay Heat Removal via the AFW System and the SG Safety Valves
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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 l
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) i i
, 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.
i In order to assure that SG level remains within its expected range, the AFW pumps and level control valves must be operable from the HSP. Upon initiation of a reactor trip, SG 1evel 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, i
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 l
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 1
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 4
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 4
pump controls to TS 3.3.3.5 is consistent with TS 3.3.4 of the STS located in NUREG-1431.
4 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 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 STS, but provides additional assurance that the ESF equipment can be powered.
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.
DCPP UFSAR Section 7.4 identifies those accidents which would result in the most severe consequences during a remote shutdown.
Based on the licensee's 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 0.0.0.5.
The above evaluation demonstrates that with the equipment previously discussed, the reactor can be maintaincd in a safe condition. Additional equipment is provided at the HSP, but is not required to be available to establish and maintain Mode 3.
_y-Jncreased Allowed Outaae Time Control of several components located at the HSP that a e required for safe shutdown can be accomplished at other locatms (e
., at the 4 kV switchgear, a motor control center, a local control r
., 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 range SG level.
Therefore, other means exist to monitor and control reactor conditions besides the indications and control functions located at the HSP.
Based on the above, an allowed outage time of 30 days for remote shutdown monitoring and control functions is acceptable.
In addition, this allowed outage time is consistent with NUREG-1431.
Seoarate Entry into Action Statement and Enhanced Bases The licensee's 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, therefore, are acceptable.
Based on the above, the staff finds the proposed changes acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released l
)
3 offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involvn no significant hazards considera-tion, and there has been no public comment on such finding (58 FR 39057).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defer.se and security or to the health and safety of the public.
Principal Contributor: S. Peterson Date:
September 2, 1994