ML20246D363
| ML20246D363 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/21/1989 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20246D360 | List: |
| References | |
| 1661, NUDOCS 8908280023 | |
| Download: ML20246D363 (19) | |
Text
. Dockat Nu'aber 50-346 License'Jumber NPF-3
- . Serial '. lumber 1661 Enclosurt
.Page 1.
APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Attached are requested changes to the Davis-Besse Nuclear Pover Station, Unit Number 1 Facility Operating License Number NPF-3. Also included are the Technical Description and Significant Hazards Consideration.
The'propcsed changes.(submitted under cover letter Serial Number 1661) concern:
Section 3/4.3.3.6, Post-Accident Instrumentation, Table 3.3-10, Post-Accident Monitoring Instrumentation and Table 4.3-10, Post-Accident Monitoring i Instrumentation Surveillance Requirements.
Bases Section 3/4.3.3.6, Post-Accident Instrumentation. t t
Section 3/4.6.4.1, Combustible Gas Control - Hydrogen Analyzers.
Bases Section 3/4.6.4, Combustible Gas Control.
By: /~\ !
D. C. Shelton, Vice President Nuclear 8908280023 890821 !
i Sworn and Subscribed before me this 21st day of August, 1989.
1shab Notary Publi4,' State of Ohio
' EVELYN L DRESS I NOTARYPUBUC,STATEOFOHIO 14 Commissioner 4iresJuly23,1954
51l1 ' Docket Numbar 50-346-
[f -License Number NPF-3~
4P : . Serial Number.1661 Enclosure Page.2' L
l The following'information is provided to support issuance of the requested I changes zo'the Davis-Besse Nuclear Power Station, Unit. Number 1 Operating License Number NPF-3,. Appendix A, Technical Specifications, Section 3/4.3.3.6, L JTables 3.3-10'and 4.3-10; Section 3/4.6.4.1; Bases section 3/4.3.3.6; and Bases Se'etion 3/4.6.4.
A. Time Required to Implement: This change is to be implemented within 45 days after the NRC issuance of the License Amendment.
B. Reason for Change (License Amendment Request Number 88-0017):
The term' " status as used in Technical Specification 3/4.3.3.6 is considered ambiguous since the exact definition as it is used in'the two tables cannot be identified.- Removal of the associated line items vill eliminate confusion regarding the ambiguous term. These changes are consistent with the Babcock and W.ilcox Standard Technical Specifications (NUREG 0103).
The. removal of the line item on containment vessel hydrogen from Specification 3/4.3.3.6:and the addition of a monthly. Channel Check to Specification 3/4.6.4.1 vill eliminate an inconsistency in mode applicability.
C.--Technical
Description:
See attached Technical Description (Attachment 1).
D. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment 2).
i
_-____._,__--,---_---._.-..a_--__-_----_-_----
E* - . Docket Number'50-346 3'
M License Number NPF-3.
' v .1 Serial Number 1661-Attachment 1 Page.1
-TECHNICAL DESCRIPTION Description of Proposed Technical Specification Change The purpose of this Technical Description is to reviav a proposed change-to the Davis-Besse Nuclear Power Station Technical Specification (TS) 3/4.3.3.6 (Post-Accident Instrumentation), Table 3.3-10 (Post-Accident Monitoring Instrumentation), Table 4.3-10 (Post-Accident Monitoring Instrumentation Surveillance Requirements), Technical Specification 3/4.6.4.1 (Combustible Gas Control - Hydrogen Analyzers), Bases 3/4.3.3.6 (Post-Accident Instrumentation) and Bases 3/4.6.4 (Combustible Gas Control). This change proposes the removal of all-line items using the term." status" in Tables 3.3-10 and 4.3-10 and.
~ removal of the line item on containment vessel hydrogen from the two tables.
In addition, several administrative changes have been made for clarification purposes. This request is a result of Toledo Edison's Technical Specification Verification Program - Phase'II commitment to the NRC to perform a review of its Technical Specifications and propose appropriate changes.
Systems Affected Post-Accident Monitoring System (no hardware changes) e Hydrogen Analyrers (no hardware changes)
Safety Function of System Affected The purpose of the Post-Accident Monitoring System'(PAMS) is to follow the-course of.an accident with instrumentation which provide the operators the essential safety status information needed to return the plant to a maintained, safe shutdown condition.-
-Effect on Safety / Proposed Changes Technical Specification Tables 3.3-10 and 4.3-10 have 14 of 34 items listed which utilize the ambiguous term " status". These items are:
Auxiliary Feedwater Status (item 6)
Containment Vessel Isolation Status (item 9)
SFAS Status (item 10)
Safety Features Equipment Status (item 11)
RPS Status (item 12)
SFRCS Status'(item 13)
HPI System Pump and Valve Status (item 16)
LPI System Pump and Valve Status (item 17)
Containment Spray Pump and Valve Status (item 18)
Core Flood Valve Status (item 19)
BVST Valve Status (item 20)
Containment Emergency Sump Valve Status (item 21)
Containment Air Cooling Fan Status (item 23)
EVS Fan and Damper Status (item 24)
Docket Number 50-346 License Numbar NPF-3 S Serial Number 1661 Attachment 1 Page 2 The term " status" is considered ambiguous since the exact definition as it is used in the two tables cannot be identified. In the current revision of the NRC's B&W Standard Technical Specification (STS) [NUREG-0103], Revision 4, September 13, 1980 the corresponding standard specification is also 3/4.3.3.6, Accident Monitoring Instrumentation. Although the Limiting Condition for l Operation (LCO) and Surveillance Requirement (SR) wording is similar and the Mode applicability is the same, the STS does not list any of these 14 " status" l line items in the associated tables. The term ' status' is not used in the STS 3/4.3.3.6. Similarly, Technical Specifications reviewed from a sample of other B&W and Non-B&W plants did not list any item using the term " status" in
- its accident monitoring Technical Specification or associated table.
Regulatory Guide (RG) 1.97, Revision 3, Instrumentation for Li g ht Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Faring and
! Following an Accident, utilizes the term ' status'. Five type of variables for ;
l either accident monitoring, monitoring of the operation of systems important to safety and monitoring radioactive effluent releases are discussed. Among these five types of variables, Type D variables are those that provide information to indicate the operation of individual safety systems and other systems important to safety. Information should be indicated by display to l ascertain the operating status of each individual safety system and other l systems important to safety to the extent necessary to determine if each system is operating or can be placed in operation to assist in mitigating the consequences of an accident. Regulatory Guide 1.97, Table 3 provides several examples of Type D variables with the use of the term ' status'. For exampla, i electric current for pressurizer heater status is monitored 'to determine .
l~ operating status'; emergency ventilation damper position is monitored as to !
'open-closed status' for the purpose of 'to determine damper status'; standby power is monitored 'to determine damper status' via status indication.
Although RG 1.97 discusses the term " status" it does not provide a strict definition of the term as used in LCO 3.3.3.6 and the associated tables, but allows a general definition to be developed. In the most general of terms, the " status" of a system or equipment can be defined as the condition / state of the system or equipment.
Since the term " status" is not strictly defined, the acceptance criteria for these same line items is subject to interpretation in the Surveillance l Requirement Table 4.3-10. The Minimum Channel Operable requirement listed for each of these items in Table 3.3-10 is also subject to interpretation in most cases. Definitions of each of these items in terms of the line items they are associated rith are difficult to determine.
For example, 'SFAS Status' has a Minimum Channel Operable requirement of
'l/ channel'. This is subject to interpretation as there are no 'SFAS Status' channels and is further confused by the lack of a strict definition of the term " status". While these items address systems whose status would be important before, during, and after an accident, the line items are not actual plant instruments or parameters. This does not appear to be within the scope of the Technical Specification (Post-Accident Instrumentation) or the information provided by the bases.
i l
__--______-_____-___-__-_w
1
.. ' Docket' Number 50 346 JO License Number NPF-3 M '. Serial Number 1661 Attachment 1 Page 3~
. In order to determine the. availability and condition of various plant systems, the operator has available monitoring capability, which includes indicators, recorders, lights, annunciators, CRT dinplays and the station computer, for the following:
- 1. Containment Vessel (CV) Environment
- 3. Reactor 4.- . Control Rod Drive Control System (trip portion) ,
Auxiliary Feedvater System 5.
- 6. Auxiliary Shutdown Panel j 7.. Engineered Safety Features (ESP)
- 9. . ' Anticipatory Reactor Trip System (ARTS)
- 10. Post Accident Monitoring System (PAMS)
- 11. Steam and Feedvater Rupture Control System (SFRCS) ,
Sufficient information is provided to enable the operator.to maintain the station in a safe condition following both anticipated operational occurrences and accident.and post-accident conditions.
'I
- For the convenience of the operating personnel, engineered safety features (ESP) systems are presented graphically on the main control panels designated for the safety features status display. Instruments are either located on
- process mimic lines or are shown connected to certain systems by' influence ;
lines.'.The Containment Vessel, Reactor, Core Flood Tanks, pumps, heat exchangers,'etc.,'are shown schematically in the systems. Pumps,' fans, and valves are in most cases represented by their respective control switches and status indicating lights. All mimic process lines and equipment are in color.
Additionally, ESF components which are automatically initiated to satisfy.
safety functions are provided with a Safety Actuation Monitoring (SAM) amber- l indicating light /svitch. These SAM lights are located on the safety features ~ J
. status display and panels.
The safety surveillance instrumentation and the SAM lights provide operations personnel with a means of readily determining the status / condition of certain )
1
- plant-systems.
The Surveillance Requirement for the PAMS Technical Specification requires- !
that a Channel Check and Channel Calibration be performed at frequencies listed for each line item in Table 4.3-10. All 14 of these " status" items require a Channel Check once every 31 days and do not require a channel calibration. The operability of these systems is verified at least once per 31 days by other individual Technical Specifications. Most of the systems have additional surveillance that are performed on frequencies greater than 31 days which require major testing to prove continued operability of the i systems.
As:an additional note, the line items being proposed to be deleted are not present in the same or similar tables or text in the NRC B&V Standard Technical Specifications or other similar B&W plant Technical Specifications.
l 1
i a
I
.1
, Docket Number 50-346 License'Nu'mber NPF-3~ ,
- *: Y ' Serial Number 1661 Attachment 1-Fage 4
.The following paragraphs provide a discussion of each of the line items to be deleted. Each paragraph discusses the information available to the operator to determine.the status of each of the systems and the surveillance testing 4 performed'to prove operability of the system.
P Item 6 - Auxiliary Feedvater Status The auxiliary feedvater system is designed to provide feedvater to the steam generators when the turbine-driven main feedvater pumps are not available or
- p following a loss of normal and reserve electric power. The Steam and Feedvater Rupture Control System (SFRCS) automatically initiates AFV as required.
Existing line item 25 (which would become line item 9 following this amendment) of Tables 3.3-10 and-4.3-10 requires that Auxiliary Feedvater (AFV) i.
Flow Rate instrumentation to be one of the post-accident instruments which provides important information on the status of the AFV system. In addition, other operating instrumentation is available to monitor performance of this system including pump discharge pressure, pump bearing oil and turbine bearing metal temperatures, steam generator startup range level (item 5 of TS Tables 3.3-10 and 4.3-10), and steam generator outlet pressure (item 1 of TS Tables 3.3-10~and 4.3-10). Station annunciators which are available to provide l
indication of AFW system trouble include AFV pump trouble, AFW pump lov suction pressure, AFW pump inlet strainer differential pressure, AFW pump room L vater level high, AFW pump turbine overspeed trip, and main steam pressure to
.AFW pump. turbine lov.
The non-operating status and verification of operability of the AFV system is accomplished in Modes 1-3 once per 31 days by the Surveillance Requirements of Technical Specifications 3.7.1.2.. These SRs verify:
l 1. The steam turbine driven AFU pumps develop a differential pressure of >_
1070 psid on recirculation flow when the secondary steam supply pressure is greater than~800 psig.
- 2. That each power operated or automatic valve in the flow path is in its
. correct position.
L
- 3. That all manual valves in the pump suction and discharge lines that affect the system's capacity to deliver Vater to the S/G are locked in proper
. position.
- 5. Operability of the Auxiliary Feed Fump Turbine Steam Generator Level Control System by a Channel Functional Test.
- 6. Operability of the Auxiliary Feed Fump Suction Fressure Interlocks by a Channel Functional Test.
g 4, l : Docket Number.50-346-g . ' . License Number NPF-3
' Serial Number.1661-Attachment 1 Page.5
- 7. ~ Operability of the Auxiliary' Feed Pump Turbine Iniet. Steam Pressure-Interlocks by a Channel Functional Test.
'Other surveillance testing is performed on an 18 month frequency which further ensures operability of the'AFV system.
.Therefore, there is' sufficient information available to provide the operator with-the. status and operability of the AFV system both in an accident and non-accident' condition.
Item 9 - Containment Vessel Isolation Status
-Isolation. valves are employed to maintain and/or re-establish the containment system integrity by the automatic isolation of all containment vessel fluid and gaseous penetrations, notLclosed by valves, blind flanges, or deactivated
-automatic valves, thereby. eliminating potential leakage paths.
Utilizing.the following available information the operator vould be able to determine if a containment isolation has or should have occurred and whether the containment isolation has been accomplished is required.
- ' SFAS annunciators
- Containment pressure indication
- Containment Radiation level indication Position indication for each containment isolation valve
- SAM lights for each containment isolation valve Technica1' Specification 3.6.1.1-(Containment Integrity), 3.6.1.2 (Containment-Leakage),.3'.6.1.3 (Containment' Air Locks),i3.6.1.4 (Internal Pressure),
3.6.1.6 (Containment Vessel. Structural Integrity), 3.6.1.7 (Containment Ventilation System), and 3.6.3.1 (Containment Isolation Valves) all contain SRs of a frequency C 31 days which verify containment integrity in Modes 1 through 4. While the Minimum Channels Operable entry'of Table 3.3-10 (1/ valve)' implies this item only applies.to containment isolation valves, there are surveillance which provide additional confidence that containment vill be properly isolated, but do not pertain to valves. The various SRs (both valves and non-valves) which are performed verify:
- 1. (4.6.1.1.a) Once per 31 days that penetrations not capable of being closed by an operable automatic containment isolation valve are closed by valves, blind-flanges, or deactivated automatic valves secured in position.
- 2. (4.6.1.1.b)_0nce per 31 days that the containment air lock is operable.
l '3. '(4.6.1.3.a) After each containment air lock opening (or once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for multiple entries) that no leakage exists by pressure testing.
a
- 4. (4.6.3.1.1) After maintenance, repair, or replacement is performed on a containment isolation-valve or it-associated operator or control power that the containment isolation valves is operable.
l u
ikf>*'
.. Docket 1 Number 50-346 A
LS: License Number NPF-3 M' Serial Number 1661
$ 'Att'achment 1 P Page 6
, other--surveillance testing is performed on a 18 month frequency which further- !
demonstrates. operability of the containment isolation system. I Therefore,_there is sufficient information available to provide the operator with the status and operability of the containment isolation system both in an accident and non-accident condition.
-Item 10 - SFAS Status The SFAS is required to sense unsafe conditions and actuate engineered safety' features-(ESF) equipment. The-design goal of the SFAS is to automatically prevent or. limit-fission product and energy release from.the core, to isolate the containment' vessel (CV) and to initiate the operation of the ESF equipment in the' event of a loss-of-coolant accident (LOCA).- The initiating circuits of the SFAS are the sensing circuits monitoring CV' radiation level,.CV pressure,
.RC pressure and BVST level. The SFAS monitors these parameters to-sense loss of coolant and activate protective action systems.
.The operator has'available station annunciators, station computer output, containment spray flov: indication, Containment High Range Radiation Monitors, Containment Vide Range Pressure Monitors, RCS pressure instrumentation, Containment Vessel Isolation, BVST level instrumentation, Containment Air Coolers operation, ECCS' flow, Emergency Diesel Generator operation, Service
. Vater operation, and Component Cooling Vater operation to determine the status of SFAS.
Technical. Specification 3.3.2.1 [ Safety Features Actuation System (SFAS)
. Instrumentation] provides Surveillance Requirements which verify operability of SFAS in the Modes appropriate for particular instrumentation. The instrumentation"and testing requirements [ Channel Check (CC) and Channel Functional ~ Test (CFT)] within a 31 day period are:
Functional Unit test (31 days or.less) Modes Instrument Strings Containment Radiation - High CC, CFT 1,2,3,4,6 Containment Pressure - High CC, CFT 1,2,3 Containment Pressure - High-!!igh CC, CFT 1,2,3 RCS Pressure - Lov CC, CFT 1,2,3 Pressure - Low-L~. CC, CFT 1,2,3 BVST Level - Low-Lov CC, CFT 1,2,3 Output Logic
. Incident Level #1: Containment Isolation CC, CFT 1,2,3,4 Incident Level #2: High Pressure CC, CFT 1,2,3,4 and Starting Diesel Generators Incident Level #3: Lov Pressure Injection CC, CFT 1,2,3,4 Incident Level #4: Containment Spray CC, CFT 1,2,3,4 Incident Level #5: Containment Sump CC, CFT 1,2,3,4
-Recirculation Permissive
~ Docket Numb 5r 50-346L k[@h*
Ji Licznsa Numbsr NPF-3'.
-Serial Number.1661 jy- ' Attachment'1 jn Page 7 Man'ual A'tuation:
e n -.SFAS, except Containment Spray and CFT l',2,3,4,6 Emergency Sump' Recirculation containment Spray l CFT 1,2,3
. Sequence Logic Channels' Sequence Logic Channels'- CC, CFT 1,2,3,4 Interlock Channels Decay' Heat Isolation Valve CC 1,2,3,4,5 Pressurizer Heater CC 3,4,5 CC'= Channel Check once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CFT - Channel Functional Test once per 31 days >
Oth'er surveillance testing is performed on a-18 month frequency which further demonstrates operability'of.the SFAS.
Therefore,.there is sufficient.information available to provide the operator with'the status and operability of the SFAS both in an accident and non-accident condition.
Item 11 - Safety Features Equipment Status The>SFAS actuates both engineered safety features equipment and essential auxiliary equipment which supports:the operation of safety features equipment.
The. equipment actuated by SFAS is listed in TS Table 3.3-5 and the systems are summarized as follows:
-Emergency Ventilation System Containment Purge and Sample Valve Isolation System HighJPressure Injection' System Containment Air Cooling System Component Cooling System Service Water System Containment Spray Emergency Diesel Generators Low Pressure' Injection System Emergency Sump Recirculation System Containment Vessel Isolation System Control Room Isolation System The operator has available annunciators, station computer output, instrumentation and. SAM' lights to determine the status of this equipment and systems. As discussed in item 10 above, channel checks and channel functional tests'are performed.for SFAS under TS 3.3.2.1. Each system which is actuated by.SFAS is itself covered by a separate Technical Specification which has Surveillance. Requirements which verify the operability of the system. 'A summary of these' systems and SRs follows:
i
Docket Number 50-346 ;
License Number NPF-3 l Serial Number 1661 l Attachment 1 1 Page 8
)
Emergency Ventilation System - TS 3.6.5.1, Modes 1 - 4, once per 30 days initiating (from the control room) flow through the HEPA filters and charcoal adsorbers and verifying the system operates for 15 minutes.
Containment Purge and Sample Valve Isolation System - TS 3.6.3.1, Modes 1 - 4, these valves are containment isolation valves and are part of the previous discussion of item 9.
L High Pressure Injection System - TS 3.5.2, Modes 1 - 3, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Containment Air Cooling System - TS 3.6.2.2, Modes 1 - 3, once per 31 days starting each unit and verify operation for 15 minutes.
Component Cooling System - TS 3.7.3.1, Modes 1 - 4, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured '
in position, is in its correct position.
Service Vater System - TS 3.7.4.1, Modes 1 - 4, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Control Room Emergency Ventilation System - TS 3.7.6.1, Modes 1 - 4, once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the control room emergency ventilation system is operating verifying the control room air temperature is < 110*F. Once per 31 days initiating (from the control room) flow througE the HEPA filters and charcoal adsorbers and verifying the system operates for 15 minutes.
Containment Spray System - TS 3.6.2.1, Modes 1 - 4, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Emergency Diesel Generators - TS 3.8.1.1, Modes 1 - 4, once per 31 days verify the fuel level in day fuel and fuel storage tanks, the fuel transfer pump can perform its function, the diesel starts and accelerates to 900 RPM, the i generator operates for 60 minutes loaded to 1000 KV and synchronized, the generator is aligned to provide standby power to the associated essential buses, and the automatic load sequence timer is operable.
Low Pressure Injection System - TS 3.5.2, Modes 1 - 3, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Emergency Sump Recirculation System - TS 3.5.2, Modes 1 - 3, once per 31 days verifying each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
CV Isolation System - see previous discussion of item 9.
?
Docket Numbar 50-346 .j License Number NPF-3 W Serial Number 1661 Attachment 1 l' Page 9 I Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the Safety Features Equipment. ;
Therefore, there is sufficient information available to provide the operator 3 vith the status and operability of the Safety Features Equipment both in an accident and non-accident condition.
Item 12 - RPS Status The purpose of the RPS is to initiate a reactor trip when a sensed parameter (or group of parameters) exceeds a setpoint value indiccting the approach of an unsafe condition. In this manner, the reactor core ir protected from exceeding design limits and the Reactor Coolant (RC) System is protected from overpressurization.
Each RPS channel contains meters and indicators mounted in ti e system cabinets which display each input analog signal and visual ind! cation af the state of each trip logic element. Total power (in percent pover) and power imbalance (in percent pover) are available to the operator on the control console. A strip chart recorder is available for continuously recording average auctioneered total power (in percent power). Each RPS channel montains an alarm panel which is visible at all times and indicates channel trip, cabinet fan failure, trip of any of the other three channels, and manual bypass.
The station annunciator and computer gPie the operator visual indication of the RPS trip status. The station computer system monitors all analog input signals, all channel power supplies, and all trip modules. The station computer system vill alarm if there is a power supply fault, a fan fails, or a cabinet door is open. The station annunciator indicates that an RPS channel trip has occurred, a power range detector power supply fault has occurred, shutdown bypass has been initiated in a channel, or a channel has been bypassed.
Technical Specification 3.3.1.1 [ Reactor Protection System (RPS)
Instrumentation] provides Surveillance Requirements which verify operability of the RPS in the Modes appropriate for particular instrumentation. The instrumentation and testing requirements within a 31 days period are:
Instrumentation test (31 days or less) Modes Manual Reactor Trip CFT* N/A High Flux CC, CFT 1 and 2 RC High Temperature CC, CFT 1 and 2 Flux-aFlux-Flow CC, CFT 1 and 2 RC Lov Pressure CC, CFT 1 and 2 RC Pressure-Temperature CC, CFT 1 and 2 High Flux / Number of RCPs on CC, CFT 1 and 2 Containment High Pressure CC, CFT 1 and 2 Intermediate Range, Neutron Flux and Rate CC, CFT* 1, 2 and #
Source Range, Neutron Flux and Rate CC, CFT** 2,3,4,5 Control Rod Drive Trip Breakers CFT** 1, 2, and #
L Docket Number 50-346 License Number NPF-3 Serial Number 1661 Attachment 1 Page 10 Reactor Trip Module Logic CFT 1, 2, and #
L Shutdown Bypass High Pressure CC, CFT 2, 3, 4, 5 ##
CC = Channel Check once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CFT = Channel Functional Test once per 31 days CFT* = Channel Functional Test within 7 days of each startup CFT** = Channel Functional Test within 7 days of each startup and once per 31 days
- Vith any control rod trip breaker closed
- only applies when. shutdown bypass is actuated Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the RPS.
Therefore, these is sufficient information available to provide the operator with the status and operability of the RPS both in an accident and non-accident condition.
Item 13 - SFRCS Status The design goal of the SFRCS is to prevent release of high energy steam, to autcmatica}1y ctart the Aarilicry Feedvater System in the event of a main steem line or main feedvater line rupture, to automatically start the Auxiliary Feedvater System on a loss of feedvater or the loss of all four RC pumps, and to prevent steam generator overfill and cubsequent spillover of water into the main steam lines. The SFRCS also provides a trip signal to the Anticipatory Reactor Trip System (ARTS). The initiating circuits of the SFRCS are the sensing circuits monitoring main steam line pressure, main feedvater/ steam generator differential pressure, steam generator level, and RC pump status.
The operator has available annunciators, station computer output, steam line pressure instrumentation and steam generator level instrumentation to determine the status of this equipment and systems.
Technical Specification 3.3.2.2 (Steam and Feedvater Rupture Control System Instrumentation) provides Surveillance Requirements which verify operability of the SFRCS in the Modes 1 - 3. The following instrumentation receives a !
Channel Check once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and a Channel Functional Test once per 31 days in Modes 1, 2, and 3.
Steam Line Pressure - Low Steam Generator Level - Lov Steam Generator - Feedvater Differential Pressure - High Loss of Reactor Coolant Pumps Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the SFRCS.
Therefore, there is sufficient information available to provide the operator with the status and operability of the SFRCS both in an accident and non-accident condition.
Docket Number 50-346 License Number NPF-3 Serial Number 1661 Attachment 1 Page 11 -
Item 16 - HPI System Pump and Valve Status The High Pressure Injection (HPI) system is part of the Emergency Core Cooling
. System. The HPI system will prevent uncovering of the core in the event that small ecolant piping leaks occur when RC system pressure is maintained above 600 psig and belov 1500 psig and vill delay uncovering of the core when intermediate-size leaks occur.
The operator has available annunciators, station computer output, HPI flow instrumentation, valve position indication and SAM lights to determine the status of the HPI system. Existing line item 14 of Table 3.3-10 requires that High Pressure Injection (HPI) Flov instrumentation be one of the post-accident instruments.
The non-operating status and verification of operability of the HPI system is accomplished in Modes 1 - 3 once per 31 days by the requirements of Technical Specification Surveillance Requirements 4.5.2.a. This SR verifies once per 31 days that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Other surveillance testing is performed on a 18 month frequency which further demanctratcc operchility of the UPI system.
Therefore, there is sufficient information available to provide the operator with the status and operability of the HPI system both in an accident and non-accident condition.
Item 17 - LPI System Pump and Valve Status The Low Pressure Injection (LPI) and Core Flood (see item 19) systems vill inject borated water into the core at intermediate to lov RC system pressures and vill ensure adequate core cooling for break sizes ranging from intermediate to the double-ended rupture of the RC piping in either the hot or cold leg.
The operator has available annunciators, station computer output, LPI flow instrumentation, valve position indication and SAM lights to determine the status of the HPI system.
The non-operating status and verification of operability of the LPI system is accomplished in Modes 1 - 3 once per 31 days by the requirements of Technical Specification Surveillance Requirement 4.5.2.a. This SR verifies once per 31 days that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Cther surveillance testing is performed on a 18 month frequency which further demonstrates operability of the LPI system.
Therefore, there is sufficient information available to provide the operator with the status and operability af the LPI system both in an accident and non-accident condition.
Docket Number 50-346 License Number NPF-3
. Serial Number 1661 Attachment 1 Page 12 Item 18 - Containment Spray Pump and Valve Status The Containment Spray System is an engineered safety feature which has the dual function of removing heat and fission product iodine from the post-accident containment atmosphere.
The operator has svailable annunciators, station computer output, Containment Spray flov instrumentation, valve position indication and SAM lights to determine the status of the Containment Spray system.
The non. operating status and verification of operability of the Containment Spray system is accomplished in Modes 1 - 4 once per 31 days by the requirements of Technical Specification Surveillance Requirement 4.6.2.1.a.
This SR verifies once per 31 days that each valve in the flov path that is not locked, sealed, or otherwise secured in position, is in its correct porition.
Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the Containment Spray system.
Therefore, these is sufficient information available to provide the operator with the status and operability of the Containment Spray system both in an accident end non-ac:ident ccndition.
Item 19 - Core Flood Valve Status The Core Flood (CF) and Lov Pressure Injection (see item 17) systems vill inject borated water into the core at intermediate to lov RC system presuures and vill ensure adequate core cooling for break sizes ranging from intermediate to the double-ended rupture of the RC piping in either the hot or cold leg. The CF system vill begin supplying vater to the reactor when the Reactor Coolant system pressure falls below the CF Tank pressure.
The operator has available annunciators, station computer output, valve position indication and SAM lights to determine the status of the Core Flood system.
The Core Flood system is addressed in Technical Specification 3.5.1. The operability of the Core Flood system is verified by the Surveillance Requirements of TS 4.5.1. The SRs that are performed once per 31 days or more frequent are:
- 1. (4.5.1.a.1) At least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify the contained borated water volume and nitrogen cover-pressure in the tanks.
- 2. (4.5.1.a.2) At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that each tank isolation valve is open.
- 3. (4.5.1.b) At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of each solution volume increase of > 80 gallons verify the boron concentration of the tank solution.
'Mhb PN Docket Number l50-346i f$h6 License Number.NPF-3 M4 Serial Number 1661.
f I Attachment 1 Page'13
, 4. (4.5.1.c) -At least once per 31 days verify that pover'to the isolation valve operator is disconnected by the breakers being locked in the open position.
Other surveillance' testing is performed on a.18 month frequency which further demonstrates operability of the Core Flood system.
Therefore, there is sufficient information available to provide the operator with the status and operability of the Core Flood system both in an accident and non-accident condition.
Item 20 - BVST Valve Status The Borated Vater Storage Tank (BVST) contains a minimum of 1800 ppm boron in
, solution and is used both.for. emergency core injection and filling the
- refueling canal during ' refueling. The BVST supplies borated water for emergency cooling.to the. containment spray system, decay neat removal (LPI)
- system, and high pressure injection system. It also supplies makeup water to the spent fuel pool-cooling system,and can serve as source for the makeup pumps.
The operator has available annunciators, station. computer output, valve-position indication and SAM lights to determine the status of the BVST.
.The BVST flow paths are addressed in Technical Specifications 3.1.2.1, 3.1.2.2 and 3.5.2. The operability of the flow path is verified by the SRs of the three specifications. SRs 4.1.2.1.a and 4.1.2.2.a require the pipe temperature of the flow path be verified > 105'F once per 7 days. SRs 4.1.2.1.b, 4.1.2.2.b, and 4.5.2a verify, once per 31 days, that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is
'in its correct. position.
s Other surveillance-testing is performed on a 18 month frequency which further demonstrates operability of the BVST.
Therefore, there is sufficient information available to provide the operator with the status and operability of the BVST both'in an accident and non-accident condition.
Item 21 - Containment Emergency Sump Valve Status Following a LOCA, after the BVST has been exhausted, the Containment Vessel Emergency Sump. vill serve continuous injection of the reactor coolant, through the low prca.7ure injection / decay heat' pump, into the ' eactor A Coolant , System.
This will maintain long-term core cooling by recirculating the spilled reactor coolant back to.the reactor vessel and/or through the containment spray pump,
'into the containment vessel atmosphere to remove the heat and decrease the pressure and temperature in the containment vessel.
The operator has available annunciators, station computer output, valve position indication and SAM lights to determine the status of the Emergency Sump.
' Docket Number 50-346 License Number NPF-3
- ' Serial Number 1661 Attachment 1 Page 14 The containment emergency sump is addressed in Technical Specification 3.5.2.
The operability of the containment emergency sump flow path is verified once per 31 days by SR 4.5.2.a. This SR verifies, once per 31 days, that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
Other surveillance testing is performed on an 18 month frequency which further demonstrates operability of the containment emergency sump.
Therefore, there is sufficient information available to provide the operator with the status and operability of the containment emergency sump both in an accident and non-accident condition.
Item 23 - Containment Air Cooling Fan Status The Containment Air Cooling system is composed of three air cooler units located within the containment vessel. These units are used for both normal and emergency cooling. The system is designed to control the containment vessel ambient air temperature to a maximum of 120*F vith two of the three units operating. The ductwork distribution system is designed to distribute air over and around all equipment which produces or releases heat.
The operator has available annunciators, station computer output, service water valve position indication, fan on/off indication, containment pressure instrumentation, containment air coolers inlet and outlet temperatures and SAM lights to determine the status of the containment air coolers.
The containment air coolers are addressed in Technical Specification 3.6.2.2.
The operability of the containment air coolers is verified by the Surveillance Requirements 4.6.2.2.a.1 and 4.6.2.2.a.2. These SRs verify, once per 31 days, that each cooling unit starts and operates for 15 minutes.
Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the containment air coolers.
Therefore, there is sufficient information available to provide the operator with the status and operability of the containment air coolers both in an accident and non-accident condition.
Item 24 - EVS Fan and Damper Status The function of the Emergency Ventilation System (EVS) is to collect and process potential leakage from the containment vessel to minimize environmental activity levels resulting from all sources of containment leakage following a loss-of-coolant-accident. The EVS is designed to provide a negative pressure within the annular space betveen the shield building and the containment vessel and in the penetration rooms following a loss-of-coolant-accident and to reduce airborne fission product leakage to the environment by filtration prior to release of air through the station vent.
Docket Number.50-346 C
License Number NPF-3
- Serial Number 1661
-Attachment 1 Page 15 The operator has available annunciators, station computer output, damper position indication, fan on/off indication, containment / shield building.
(, differential pressure instrumentation and SAM lights to determine the status of the EVS.
The EVS is addressed in Technical Specification 3.6.5.1. The operability of the' emergency ventilation system is' verified by SR 4.6.5.1.a. This SR verifies operability, once per 31 days, by initiating (from the control room) flow through'the HEPA filters and charcoal adsorbers and verifying the' system operates for 15 minutes'once per 31 days.
Other surveillance testing is performed on a 18 month frequency which further demonstrates operability of the EVS.
Therefore, there is sufficient information available to provide the operator with the' status and operability of the EVS both in an accident and non-acc1 dent condition.
~
Item 7 - Containment Vessel Hydrogen The Combustible Gas Control System is designed to control the concentration of hydrogen which may be released within the containment vessel atmosphere following a LOCA. The system is composed of the Containment Hydrogen Dilution (containment atmosphere dilution) System, the Hydrogen Purge System, and the Containment Recirculation System.
While this line item does not contain the term " status", a change is proposed to eliminate an inconsistency in mode applicability. There are three individual Technical Specifications for monitoring and controlling containment vessel hydrogen. Technical Specification 3/4.6.4.1, Hydrogen Analyzers, is applicable in Modes 1 and 2 and requires two independent containment hydrogen monitors to be operable. The Surveillance Requirement performs a channel calibration once per 92 days on a staggered test basis. The hydrogen monitors provide the control room with continuous indication of containment hydrogen concentrations. The range of measurement capability is zero to 10% hydrogen concentration under both positive and negative ambient pressure. Technical Specifications 3/4.6.4.3, Containment Hydrogen Dilution System, is applicable in Modes 1 and 2 and requires two independent containment hydrogen dilution.
systems to be operable. The Surveillance Requirements verify the system can be started from the control room and operates for 15 minutes once per 92 days.
Technical Specification 3/4.6.4.4, Hydrogen Purge System, is applicable in Modes 1 and 2 and requires two independent containment hydrogen purge systems to be operable. The Surveillance Requirements are primarily performed every 18 months except in the event of major maintenance or modifications of the system components or following painting or chemical releases in the systems ventilation zone. The SRs verify proper system flow and efficiency. Since there is no 31 day Surveillance Requirement under TS 3/4.6.4.1, Hydrogen Analyzers, to verify operability of the bydrogen monitors, addition of a 31 day Channel Check is proposed to be added to TS 3/4.6.4.1 by this amendment i request. Addition of this SR vill ensure that the operability of the i
analyzers will be verified once per 31 days. Vith the addition of this 31 day channel check to TS 3/4.6.4.1, line item 7 may then be deleted from Tables 3.3-10 and 4.3-10.
Docket: Number 50-346
- License Number NPF-3 6 Serial' Number 1661.
Attachment 1 Page 16 Although TS 3.3.3.6'is applicable in Modes 1 - 3, TS 3.0.4 does not apply
. thereby allowing reliance on TS 3.3.3.6 Action station "a" to. enter Modes 3, 2,cor 1.- However, this allowance is inconsistent with TS 3.6.4.1 which not allow entry into. Modes 1 or 2 with an inoperable hydrogen analyzer. If, under TS 3.3.3.'6, the monitor was not restored to operable status within 30 days, then the plant would be required to 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 />. It is unlikely for the plant to remain in Hot Standby (Mode 3) for
. anywhere near 30 days. Adding the 31 day channel check to TS 3.4.6.1 vill allow the monitor to.be inoperable in Mode 3, however~, the action statements for. Modes 1 and 2 are similar to that of TS 3.3.3.6 except that TS 3.0.4 is applicable (therefore, no change into Modes 1 or 2 could be made with the monitor inoperable). . Based on the inoperability allowances for entering Mode 3.by.TS 3.3.3.6, the normally short duration in Mode 3, and.the applicability of.TS 3.0.4;to TS 3.6.4.1 there is no degradation in safety by transferring this Surveillance' Requirement for the hydrogen monitors from TS 3.3.3.6 to TS 3.6.4.1. This proposed change vill eliminate the mode inconsistency and the.
unnecessary: inclusion of the monitors within the PAMS Technical Specification.
Administrative Changes Following deletion of the proposed line items, the remaining line items vill be renumbered sequentially in each table to improve readability and prevent.-
confusion. Additionally, the title of TS 3.3.3.6 is proposed to be changed to Post-Accident Monitoring Instrumentation. The addition of the word
" Monitoring" is for clarity and consistency with the LC0 statement. Likewise, editorial changes are proposed to action statement "a" of the LCO to provide similar clarity and consistency.
Line items.13 and 16 of Tables 3.3-10 and 4.3-10 are also proposed to be revised for consistency and clarity reasons.
The note in Table 4.3-10 which allowed a one time extension of the 18 month
' channel. calibration of the steam generator outlet stnta pressure (line item 1)
. from May 17, 1983 to September 17. 1983 is also being deleted. Also, the
" asterisk" at Table 4.3-10 line item 1 which denotes that this note applies to the steam generator outlet steam pressure channel calibration is being removed. This note was added by Amendment Number 59 and is no longer applicable since the time period has passed.
Technical Specification Bases 3/4.3.3.6 changes are proposed to reference the hydrogen analyzers as post-accident monitors which have their operability requirements in TS 3/4.6.4.1 and provide editorial clarification and consistency. Technical Specification Bases 3/4.6.4 changes are proposed-to reference'the hydrogen analyzers as post-accident monitors which are covered by TS 3/4.6.4.1.
Unreviewed Safety Questions / Evaluation The proposed changes would not increase the probability of occurrence of an accident previously evaluated in the USAR because the accident conditions and
- assumptions are not affected since no hardwale changes are being made nor is testing being degraded (10CFR50.59(a)(2)(i)).
___m.___ ._____-_________mm_-__ ____._______.m_.m.-_______._m_m ...._ . ._
- Docket: Number 50-346 '
License Number NPF-3
- f4' Serial' Number 1661
'~
Attachment.1 Page 17 The proposed changes would not. increase the consequences of an accident previously evaluated in the USAR because the accident conditions and assumptions'are not affected,since no hardware changes are being made. . Other available information will provide the equivalent information provided by the line items being deleted (10CFR50.59(a)(2)(i)).
The' proposed changes vould not increase the probability of occurrence of a.
malfunction of equipment important to safety previously evaluated in the USAR because the changes do not involve a test or experiment and no station j.
j . equipment is being' modified (10CFR50.59(a)(2)(i)).
The proposed changes would not increase the' consequences of a malfunction of equipment important to safety previously' evaluated'in the USAR because the-changes do not involve a test or experiment'and no station equipment is being modified. Other available information vill provide the equivalent information provided by the line items being deleted (10CFR50.59(a)(2)(1)).
l The proposed changes vould not create the possibility for an accident of'a-different type'than any evaluated previously in'the USAR becau'se the accident conditions and assumptions are not affected since no hardware changes are
! being made. On matters related to nuclear safety, all accidents are bounded l by previous analysis and no new malfunctions are involved (10CFR50.=59(a)(2)(i)).
The proposed changes would not create the possibility for a malfunction of a dif ferer t type. than any evaluated previously in the USAR because .the changes do notavolve a test or experiment and no station equipment is being modified (10CFR50.59(a)(2)(ii)).
The proposed changes vould not reduce the margin of safety as defined in the basis for any Technical Specification since the types of information provided by the line items being deleted is obtained by available instrumentation and annunciation (10CFR50.59(a)(2)(fii)).
Based on the above, it is concluded that the proposed Technical Specification and Fases changes do not constitute an unreviewed safety question.
l i
. _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _