ML011630370
| ML011630370 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick |
| Issue date: | 04/07/1995 |
| From: | Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR, JAFP-01-0133 | |
| Download: ML011630370 (174) | |
Text
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
DISCUSSION OF CHANGES (DOCs) TO THE CTS NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
FOR LESS RESTRICTIVE CHANGES MARKUP OF NUREG-1433, REVISION 1, SPECIFICATION JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1 MARKUP OF NUREG-1433, REVISION 1, BASES JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1, BASES RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
INSERT NEW SPECIFICATION 3.4.7 Insert new Specification 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown" as shown in the JAFNPP Improved Technical Specifications.
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown DISCUSSION OF CHANGES (DOCs) TO THE
.CTS "
DISCUSSION OF CHANGES ITS: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN ADMINISTRATIVE CHANGES None TECHNICAL CHANGES - MORE RESTRICTIVE M1 A Specification (ITS 3.4.7) is being added requiring two RHR shutdown cooling subsystems to be Operable in MODE 3 with reactor steam dome pressure less than the shutdown cooling permissive pressure.
In MODE 3, the RHR shutdown cooling subsystems are not required to mitigate any events or accidents in the safety analyses.
The RHR shutdown cooling subsystems were identified as important contributors to risk reduction and, therefore, included in the JAFNPP ITS in accordance with Criterion 4 of 10 CFR 50.36(c)(2)(ii).
Appropriate Actions and a Surveillance Requirement are also being added.
The addition of the new Specification is a more restrictive change necessary to ensure residual heat removal capability is available.
TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
None TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
None TECHNICAL CHANGES - RELOCATIONS None Page 1 of 1 JAFNPP Revi si on A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC) FOR LESS RESTRICTIVE CHANGES
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
There are no plant specific less restrictive changes identified for this Specification.
Page 1 of 1 Revi si on A JAFNPP
RHR Shutdown Cooling System-Hot Shutdown 3.4p 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4,$
Shutdown Cooling System-Hot Shutdown LCO 3.4.4)
Two RHR shutdown cooling subsystems shall be OPERABL aa
/w~ith nou pecruIFE4111on pang i a, 'Itio.,
o-,:....
t One RHR shutdown cooling subsystem may be inoperable frup to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.
NT#--------------------------
)
APPLIABILTY:
MODE 3, with reactor steam dome pressure *the RiIR cut in permissive pressurSei ACT IONS NOTES
- 1.
LCO 3.0.4 is not applicab
- 2.
Separate Condition entry subsystem.
- -NOTES---
le.
is allowed for each RHR shutdown cooling CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Initiate action to Immediately shutdown cooling restore RHR shutdown subsystems inoperable, cooling subsystem(s) to OPERABLE status.
AN(
(continued) 3.4-18 Tt, P.
A%1 e0je T W e 07"V Amemlwel't
RiR Shutdown Cooling System-Hot Shutdown 3.4ýý At?1'flU
gy I LUI.J CONDITION REQUIRED ACTION COIMPLETION TIME A.
(continued)
A.2 Verify an alternate I hour method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.
A.3 Be in MODE 4.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. No RHR shutdown cooling subsystem in operation.
AND No recirculatio pump in operation.
Initiate action to restore one RHR shutdown cooling reiruato pumpAto recircul ation pumpt operation.
AND B.2 Verify reactor coolant circu tion by an altern e method.
AND B.3 Mon tor reactor co ant temperature a pressure.
I diately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of no reactor coolant circulation AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter Once per hour Rev 1, 04/07/95 BWR/4 STS
/B1 I
I 3.4-19
RHR Shutdown Cooling System-Hot Shutdown R
Rev 1, 04/07/95 3.4-20 BWR/4 STS
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown MARKUP OF NUREG-1433, REVISION 1 SPECIFICATION
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 The requirement in ISTS LCO 3.4.8 (ITS LCO 3.4.7) associated with a recirculation pump or RHR shutdown cooling subsystem being in operation has been deleted.
The requirement that two RHR shutdown cooling subsystems are Operable is considered acceptable.
Requirements for RHR shutdown cooling subsystem operations are adequately controlled by JAFNPP plant operating procedures and policies.
There are no explicit requirements in the CTS for RHR shutdown cooling subsystem operability or that a recirculation pump should be in operation in Hot Shutdown.
- However, ITS SR 3.4.9.1 will require the reactor coolant temperature to be monitored during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing while SR 3.4.9.2 will require these readings to be taken during a planned startup where the reactor will achieve criticality.
These SRs will necessitate placing either the recirculation pump or a RHR shutdown cooling subsystem in service to ensure circulation for temperature monitoring or cooling.
The RHR shutdown cooling subsystems remove decay heat to reduce temperature of the reactor coolant to < 212°F in preparation for performing Refueling or Cold Shutdown maintenance operations, or for maintaining the reactor stable in the Hot Shutdown conditions.
Therefore, an RHR shutdown cooling subsystem will normally be in operation when a cooldown is in progress or to maintain reactor coolant temperature.
Without it an alternate will be necessary to achieve the plant objective.
During a heatup, the recirculation pumps will normally be in operation to ensure circul ation and temperature monitoring.
If plant conditions are maintained operations will consider the situation but normally either the recirculation pump will be in service or an RHR shutdown cooling subsystem will in operation to ensure adequate temperature monitoring.
This will ensure the plant objective is being met and to ensure the core is in a safe condition.
JAFNPP is operated in Hot Shutdown, in a safe manner to ensure decay heat is removed so that no core damage could result, therefore the system is normally continuously operated during these conditions.
The requirements in ISTS LCO 3.4.8 are not needed since the Surveillances in ITS 3.4.9 will clearly require temperature monitoring capability and this is accomplished with either any RHR shutdown cooling subsystem or recirculation pump in operation.
Furthermore, to be consistent with this modification, the allowances in ISTS LCO 3.4.7 Note 1 that both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the requirements in ISTS 3.4.8 ACTION B for no RHR shutdown cooling subsystem and no recirculation pump in operation have been deleted.
ISTS SR 3.4.8.1 (ITS SR 3.4.7.1) requires the verification that one RHR shutdown cooling subsystem or recirculation pump is in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Since the requirement to be in Page 1 of 2 JAFNPP Revision A
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 (continued) operation has been deleted this SR has been revised to verify each required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position every 31 days.
The Frequency is consistent with similar Surveillances in other LCOs and is considered adequate for this condition.
In Hot Shutdown, the RHR shutdown cooling system is of prime focus of plant operations and since the controls for this system are in the control room the Frequency is considered adequate.
This Frequency is consistent with the same type of Surveillance in other LCOs ( e.g.,
ITS 3.5.1).
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5, "RCS Pressure Isolation Valve (PIV)
Leakage", is not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PA2 Editorial change to maintain consistency with Restructured Technical Specifications.
PA3 The brackets have been removed and the proper nomenclature has been provided.
the Writer's Guide for the pl ant speci fic P1 ANT SPFCTFTCi flTFFFRENCF TN DFST(N AR flFSTN None DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
None Page 2 of 2 RAqTq (fnR PLN SPCII DIFREC IN DEIG OR*
DESIGN*
mIV gl*V
@WV.
V*..
U*
JAFNPP Revision A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown MARKUP OF NUREG-1433, REVISION 1, BASES
RiR Shutdown Cooling System-Hot Shutdown B 3.4*
B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.41 Residual Heat Removal (RHR)
Shutdown Cooling System-Hot Shutdown BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.
This decay heat must be
~L~.j~mped to reduce the temperature of the reactor coolant to Q*
T F
s in preparation for pe rming fuelini maintenanlceoperations5, or W
~
the actor n thHo ut condition.Al The two redundant, manually controlled shutdown cooling foo subsystem-51of the RHR System provide decay heat removal.
Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.
Both loo s have a comon suction from the sam recirculation oap.
Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via
)
t e assocea recirculation loop.
The RHR heat exchangers transfer heat to the RHR Service Water System (LCO 3.7.1,
'Resdual Heat Removal Service Water (RHRSW) System").
APPLICABLE SAFETY ANALYSES LCO Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses Decay heat removal is, however, an important safety fIn'm must be accomplished or core damage could result./&
Ahe RHR shutdown cooling subsystem mee Two RHR shutdown cooling subsystems are required to be OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves.
The two subsystems have a common suction source and are allowed to have a common heat exchanger and (continued) 4e.
A ll B 3.4-37 4~JP (c 0tnud
RHR Shutdown Cooling System-Hot Shutdoi B 3.4
('7, 7
LL ý,
ýfz -seý-rJe "W(
4-.r
ý-4Ai M1 BASES comon dscharge piping. Thus, to meet the LCO, both pumpsf in one loopi'or onr~e' -pu n each of the two loopsmust be OPERABLE.
Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems.
Each shutdown cooling subsyste is considered OPERABLE if it can be manually aligned jtEiWNjL 161M) in the shutdown cooling mode for removal of decay heat.
In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy.
Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
Note I permits th RHR shutdown cooling s ystems to t0 shutdownfoa neried of 2 hnuri in an
.our -eriodi4'tlote wallows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests.
These tests may be on the affected RHR System or on some oelant system or component that necessitates placing tliRHRIM Lm in an inoperable status during the performance.
This is permitted because the core heat generation can be low enough and the heatup rate/l ow enough a,
0 st or^?L P..r 1::,
to _al ow some cnangg'ees to te Ilsubsystems or/ttherw operations requfring loss of redundancy.
I r
,e p,;..r I
MJ In MODE 3 with reactor steam/domepressure below
-- Min permissive pressuremj( i.e., the actual pre
-which thelinterlock s----)ethe RHRSystemXUT"Nw in the shutdown cooling e to remove decay heal or maintain coolant temperature.
Otherwise, a r pump It-ný In MODES 1 and 2, and in MODE 3 with rea or steam dome pressure-greater than or equal to fthe RHR cutf pressures; this LCO is not applicable.
Operation of System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.
Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing (continued)
Rev 1, 04/07/95 BWR/4 STS 8 3.4-38 7*. tav A-.
/
0 h
l'
RHR Shutdown Cooling System-Hot Shutdown B 3. 4 BASES APPLICABILITY (continued)
ACTIONS the steam in the main condenser.
Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)
(LCO 3.5.1, "ECCS-Operating') do not allow placing the RHR shutdown cooling subsystem into operation.
Ph The requirements forecay heat removal in NODES 4 and 5 are discussed in LCO 3.4 w,'Residual Heat Removal (R R) 1 Shutdown Cooling System-Cold Shutdown'; LCO 3.9.';
"Residual Heat Removal (RHR)-High Water Level'; and LCO 3.9.9, Residual Heat Removal (RHR)-Low Water Level.'
A Note to the ACTIONS excludes the MODE change restriction of LCO 3.0.4.
This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown.
This exception is acceptable due to the redundancy of the OPERABLE subsystems, the low pressure at which the plant is operating, the low probability of an event occurring during operation in this condition, and the availability of alternate methods of decay heat removal capability.
A second Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems.
Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems.
As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
A.1. A.2. and A.3 With one required RHR shutdown cooling subsystem inoperable for decay heat removal, except as permitted by1CO Note 0, the inoperable subsystem must be restored to OPERABLE status (continued)
Rev 1, 04/07/95 BWR/4 STS B 3.4-39
RHR Shutdown Cooling System-Hot Shutdown B 3.4r BASES ACTIONS A.I. A.2. and A.3 (continued) without delay.
In this condition, the remaining OPERABLE subsystem can provide the necessary decay heat removal.
The overall reliability is reduced, however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability.
Therefore, an alternate method of decay heat removal must be provided.
With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heatl removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability.
This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities.
The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration)
S 1its capability to maintain or reduce temperature.
Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability.
Alternate methods that can be used include (but are not limited to)
- ;*(0,**o*
However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required cwfl P*J to reduce the reactor coolant temperature to the point where sew dr(.. 45A NODE 4 is entered.
e 6.V
-,1.
B.2-_and__. 3 tret~
iwef s
va With no RHR shutdown *ooltng subsystem and no re ~rculation\\
pump in operation, ecept as permitted by LCO Ite 1,
reactor coolant ci lation by the RHR shutdo cooling subsystem or recti tlation pump must be restored without UtlRHR or r cltion pump operation reetablished, an alternate 14hod of reactor coolant cir Ulaton must be placed into s ice.
This will provide t e necessary circvulation monitoring coolant temper ture.
The
(
hour Completion TrMe vis based on the coolant circulation function (continued)
Rev 1, 04/07/95 BWR/4 STS 8 3.4-40
RHR Shutdown Cooling System-Hot Shutdo V
B 3.
BASES ACTIONS
- 2. a (continued) and is modifi such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is app icable separately fo each occurrence involving a ass of coolant circulation.
Furthermore, verification oa the functioning of the alt ate method must be reconfi d every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafte.
This will provide assuran *of continued temperat e monitoring capability.
During he period when the reactor c olant is being circul ted by an alternate method (
her than by the requt RHR shutdown cooling subs tem or recirculat* n p
, the reactor coolant temper ure and pressure st be per odically monitored to ensure roper function of he al ernate method.
The once per our Completion Time is deemed appropri ate.
SURVEILLANCE REQUIREMENTS This Surveillance erifies that one RHR sh adown cooling subsystem or rec ulation pump is in ope ation and
/,1 circulating rea r coolant.
The requir d flow rate is determined by e flow rate necessary t provide suffi ent decay heat val capability.
The F quency of 12 urs is sufficient i view of other visual an audible indic tions St
( ~?
available t the operator for monito ing the RHR su system in the con 1 room.
This Surveillance is modified by a Note allowing su icient L
time to Q
RHR S shutdown cooling after clearing the pressure interlock that isolates the syst TWTDM yers A100 The Note takes exception to the requirements of the Surveillance being met (L.e.,
o not required for this initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period), which also Sallows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met' at the time of entry into the Applicability.
BREFERENTS
ýý c-es B 3.441.
Rev 1, 4/0/
B 3.4-41 Rev 1, 04/07/95 BWRI4 STS
INSERT SR 3.4.7.1 LIi Verifying the correct alignment for manual, power operated, and automatic valves in the RHR shutdown cooling flow path provides assurance that the proper flow paths will exist for RHR operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing.
A valve that can be manually (from the control room or locally) aligned is allowed to be in a non-RHR shutdown cooling position provided the valve can be repositioned.
This SR does not require any testing or valve manipulation: rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.
The Frequency of 31 days is further justified because the valves are operated under procedural control.
This Frequency has been shown to be acceptable through operating experience.
Insert Page B 3.4-41
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1, BASES
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1
-ITS BASES: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 The requirement in ISTS LCO 3.4.8 (ITS LCO 3.4.7) associated with a recirculation pump or RHR shutdown cooling subsystem being in operation has been deleted.
The requirement that two RHR shutdown cooling subsystems are Operable is considered acceptable.
Requirements for RHR shutdown cooling subsystem operations are adequately controlled by JAFNPP plant operating procedures and policies.
There are no explicit requirements in the CTS for RHR shutdown cooling subsystem operability or that a recirculation pump should be in operation in Hot Shutdown.
However, ITS SR 3.4.9.1 will require the reactor coolant temperature to be monitored during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing while SR 3.4.9.2 will require these readings to be taken during a planned startup where the reactor will achieve criticality. These SRs will necessitate placing either the recirculation pump or a RHR shutdown cooling subsystem in service to ensure circulation for temperature monitoring or cooling.
The RHR shutdown cooling subsystems remove decay heat to reduce temperature of the reactor coolant to < 212°F in preparation for performing Refueling or Cold Shutdown maintenance operations, or for maintaining the reactor stable in the Hot Shutdown conditions.
Therefore, an RHR shutdown cooling subsystem will normally be in operation when a cooldown is in progress or to maintain reactor coolant temperature.
Without it an alternate will be necessary to achieve the plant objective.
During a heatup. the recirculation pumps will normally be in operation to ensure circulation and temperature monitoring.
If plant conditions are maintained operations will consider the situation but normally either the recirculation pump will be in service or an RHR shutdown cooling subsystem will in operation to ensure adequate temperature monitoring.
This will ensure the plant objective is being met and to ensure the core is in a safe condition.
JAFNPP is operated in Hot Shutdown, in a safe manner to ensure decay heat is removed so that no core damage could result, therefore the system is normally continuously operated during these conditions.
The requirements in ISTS LCO 3.4.8 are not needed since the Surveillances in ITS 3.4.9 will clearly require temperature monitoring capability and this is accomplished with either any RHR shutdown cooling subsystem or recirculation pump in operation.
Furthermore, to be consistent with this modification, the allowances in ISTS LCO 3.4.7 Note 1 that both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the requirements in ISTS 3.4.8 ACTION B for no RHR shutdown cooling subsystem and no recirculation pump in operation have been deleted.
ISTS SR 3.4.8.1 (ITS SR 3.4.7.1) requires the verification that one RHR shutdown cooling subsystem or recirculation pump is in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Since the requirement to be in Page 1 of 3 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 (continued) operation has been deleted this SR has been revised to verify each required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position every 31 days.
The Frequency is consistent with similar Surveillances in other LCOs and is considered adequate for this condition.
In Hot Shutdown, the RHR shutdown cooling system is of prime focus of plant operations and since the controls for this system are in the control room the Frequency is considered adequate.
This Frequency is consistent with the same type of Surveillance in other LCOs ( e.g.,
ITS 3.5.1).
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5. "RCS Pressure Isolation Valve (PIV)
Leakage," is not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PA2 The Bases have been revised to be consistent with changes made to the Specifications.
PA3 Editorial changes have been made for enhanced clarification, correction, or improvement with no change in intent.
PA4 The Bases have been modified to reflect plant specific terminology.
PA5 The correct LCO number has been included.
PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 The Bases have been modified to reflect JAFNPP specific References.
DB2 The Bases have been revised to reflect the JAFNPP specific design.
DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None Page 2 of 3 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.4.7 - RHR SHUTDOWN COOLING SYSTEM-HOT SHUTDOWN DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
X1 NUREG-1433, Revision 1. Bases references to "the NRC Policy Statement" has been replaced with 10 CFR 50.36(c)(2)(ii), in accordance with 60 FR 36953 effective August 18, 1995.
Page 3 of 3 Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
RHR Shutdown Cooling System-Hot Shutdown 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown LCO
3.4.7 APPLICABILITY
Two RHR shutdown cooling subsystems shall be OPERABLE.
NOTE ---------------------------
One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.
MODE 3, with reactor steam dome pressure less than the RHR cut in permissive pressure.
ACTIONS
..-.--- NOTES -----------------------------------
- 1.
LCO 3.0.4 is not applicable.
- 2.
Separate Condition entry is allowed for each RHR shutdown cooling subsystem.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Initiate action to Immediately shutdown cooling restore RHR shutdown subsystems inoperable, cooling subsystem(s) to OPERABLE status.
AND A.2 Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.
AND A.3 Be in MODE 4.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Amendment 3.4-14 JAFNPP
RHR Shutdown Cooling System-Hot Shutdown 3.4.7 SR 3.4.7.1 I
SURVEILLANCE
---.NOTE -------------------
Not required to be met until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reactor steam dome pressure is less than the RHR cut in permissive pressure.
Verify each required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position.
Amendment CIIDUIPTI I AMIM DrnI ITDOMPINTq FREQUENCY 31 days 3.4-15 JAFNPP
RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown BASES BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.
This decay heat must be removed to reduce the temperature of the reactor coolant to s 212°F in preparation for performing Refueling or Cold Shutdown maintenance operations, or the decay heat must be removed for maintaining the reactor in the Hot Shutdown condition.
The two redundant, manually controlled shutdown cooling subsystems (loops) of the RHR System provide decay heat removal.
Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.
Both loops have a common suction from the same reactor water recirculation loop.
Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the associated reactor water recirculation loop.
The RHR heat exchangers transfer heat to the RHR Service Water System (LCO 3.7.1, "Residual Heat Removal Service Water (RHRSW) System").
Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses (Ref. 1).
Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.
The RHR shutdown cooling subsystem meets Criterion 4 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).
Two RHR shutdown cooling subsystems are required to be OPERABLE.
An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves.
The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.
Thus, to meet the LCO, both RHR pumps (and two RHR service water pumps) in one loop or one RHR pump (and one RHR service water pump) in (continued)
Revision 0 B 3.4-35 JAFNPP
RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES LCO (continued) each of the two loops must be OPERABLE.
Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems.
Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (from the control room or locally) in the shutdown cooling mode for removal of decay heat.
In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy.
Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
The Note allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests.
These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR shutdown cooling subsystems in an inoperable status during the performance.
This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR shutdown cooling subsystems or other operations requiring loss of redundancy.
APPLICABILITY In MODE 3 with reactor steam dome pressure below the RHR cut-in permissive pressure (i.e., the actual pressure at which the shutdown cooling suction valve isolation logic interlock resets (Function 6.a of LCO 3.3.6.1. Primary Containment Isolation Instrumentation)) the RHR System is required to be OPERABLE so that it may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant temperature.
Otherwise, a recirculation pump is normally in operation to circulate coolant to provide for temperature monitoring.
In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut-in permissive pressure, this LCO is not applicable.
Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.
Decay heat removal (continued)
Revision 0 B 3.4-36 JAFNPP
RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES APPLICABILITY at reactor pressures greater than or equal to the RHR cut in (continued) permissive pressure is typically accomplished by condensing the steam in the main condenser.
Additionally, in MODE 2 below this pressure. the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)
"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.
The requirements for decay heat removal in MODES 4 and 5 are discussed in LCO 3.4.8, "Residual Heat Removal (RHR)
Shutdown Cooling System-Cold Shutdown": LCO 3.9.7.
"Residual Heat Removal (RHR)-High Water Level": and LCO 3.9.8, "Residual Heat Removal (RHR)-Low Water Level."
ACTIONS A Note to the ACTIONS excludes the MODE change restriction of LCO 3.0.4.
This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown.
This exception is acceptable due to the redundancy of the OPERABLE subsystems, the low pressure at which the plant is operating, the low probability of an event occurring during operation in this condition, and the availability of alternate methods of decay heat removal capability.
A second Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems.
Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems.
As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
(continued)
Revision 0 JAFNPP B 3.4-37
RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES ACTIONS (continued)
A.1, A.2. and A.3 With one required RHR shutdown cooling subsystem inoperable for decay heat removal, except as permitted by the LCO Note, the inoperable subsystem must be restored to OPERABLE status without delay.
In this condition, the remaining OPERABLE subsystem can provide the necessary decay heat removal.
The overall reliability is reduced, however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability. Therefore, an alternate method of decay heat removal must be provided.
With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability.
This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities.
The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature.
Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability.
Alternate methods that can be used include (but are not limited to) the Condensate and Main Steam Systems, Reactor Water Cleanup System (by itself or using feed and bleed in combination with the Control Rod Drive System or Condensate System), or a combination of an RHR pump and safety/relief valve(s).
However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where MODE 4 is entered.
SURVEILLANCE SR 3.4.7.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR shutdown cooling flow path provides assurance that the proper flow paths will exist for RHR operation.
This SR does not apply to valves that are (continued)
Revision 0 B 3.4-38 JAFNPP
RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES SURVEILLANCE SR 3.4.7.1 (continued)
REQUIREMENTS locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing.
A valve that can be manually (from the control room or locally) aligned is allowed to be in a non-RHR shutdown cooling position provided the valve can be repositioned.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.
The Frequency of 31 days is further justified because the valves are operated under procedural control.
This Frequency has been shown to be acceptable through operating experience.
This Surveillance is modified by a Note allowing sufficient time to verify RHR shutdown cooling subsystem OPERABILITY after clearing the pressure interlock that isolates the system.
The Note takes exception to the requirements of the Surveillance being met (i.e., valves are aligned or can be aligned is not required for this initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period),
which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
REFERENCES
- 1.
UFSAR. Chapter-14.
- 2.
Revision 0 JAFNPP B 3.4-39
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
DISCUSSION OF CHANGES (DOCs) TO THE CTS NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
FOR LESS RESTRICTIVE CHANGES MARKUP OF NUREG-1433, REVISION 1, SPECIFICATION JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION I MARKUP OF NUREG-1433, REVISION 1, BASES JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1, BASES RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
New Specification 3.4.8 Insert new ITS Specification 3.4.8, "Residual Heat Removal (RHR)
Shutdown Cooling System-Cold Shutdown" as shown in the JAFNPP Improved Technical Specifications.
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown DISCUSSION OF CHANGES (DOGs) TO THE CTS 1.
DISCUSSION OF CHANGES ITS: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN ADMINISTRATIVE CHANGES None TECHNICAL CHANGES - MORE RESTRICTIVE M1 A Specification (ITS 3.4.8) is being added requiring two RHR shutdown cooling subsystems to be Operable in MODE 4.
In MODE 4. the RHR shutdown cooling subsystems are not required to mitigate any events or accidents in the safety analyses.
The RHR shutdown cooling subsystems were identified as important contributors to risk reduction and, therefore, included in the JAFNPP ITS in accordance with Criterion 4 of 10 CFR 50.36(c)(2)(ii).
Appropriate Actions and a Surveillance Requirement have also been added.
The addition of the new Specification is a more restrictive change necessary to ensure residual heat removal capability is available.
TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
None TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
None TECHNICAL CHANGES - RELOCATIONS None Page 1 of 1 JAFNPP Revi si on A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC) FOR LESS RESTRICTIVE CHANGES
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
There are no plant specific less restrictive changes identified for this Speci fi cati on.
Page 1 of 1 Revi si on A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown MARKUP OF NUREG-1433, REVISION 1 SPECIFICATION
RHR Shutdown Cooling System-Cold Shutdown 3.4$
3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.§ sidual eat Removal (RHR)
Shutdown Cooling System-Cold Shutdown LCO 3.4Two RHR shutdown cooling subsystems shall be OPERABLE a
Ttur-no recirculptlon pump rn0per atL R
Vqhutdown cooliA subsystem sWll bein oerat' W "
Le @
coollng smusysr ved from operat;
__*.o_*0ne RHR shutdown cooling subsystem may be inoperable e1for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.
APPLICABILITY:
MODE 4.
ACTIONS NOTE Seonarate Condition entry is allowed for each shutdown cooling subsystem.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown cooling method of decay heat subsystems inoperable, removal is available ANM for each inoperable RHR shutdown cooling Once per subsystem.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)
Alt 3.4-21 vwT W
euc(wetzl
System-Cold Shutdown 3.4jý L ACTIONS continued
.R&nrTTAu LýIU MI auiý 1~~~
I I/
__________________________*q. ---------
B.
No RHR shutdown cooling subsystem operation.
SAND No recirculation in operation.
/
in ump B.1 Verify reactor coolant circulating by an alternate B.2 Monlt cooli
- d.
- or reactor knt temperatur C
LETION TIME
/hour from iscovery of no reactor coolant circulation Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter Once per hour Rev 1, 04/07/95 BWR/4 STS ca5I metho RNR Shutdown cooling REQUIRED ACTION WWlUIUI 3.4-22
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN RETENTION OF EXISTING REQUIREMENt (CLB)
CLB1 The requirement in ISTS LCO 3.4.9 (ITS LCO 3.4.8) associated with a recirculation pump or RHR shutdown cooling subsystem being in operation has been deleted.
The requirement that two RHR shutdown cooling subsystems are Operable is considered acceptable.
Requirements for RHR shutdown cooling subsystem operations are adequately controlled by JAFNPP plant operating procedures and policies.
There are no explicit requirements in the CTS for RHR shutdown cooling subsystem operability or that a recirculation pump should be in operation in Cold Shutdown.
- However, ITS SR 3.4.9.1 will require the reactor coolant temperature to be monitored during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing while SR 3.4.9.2 will require these readings to be taken during a planned startup where the reactor will achieve criticality.
These Surveillance will necessitate placing either the recirculation pump or a RHR shutdown cooling subsystem in service to ensure circulation for temperature monitoring or cooling.
The RHR shutdown cooling subsystems remove decay heat to reduce temperature of the reactor coolant to < 212°F in preparation for performing Refueling or Cold Shutdown maintenance operations, or for maintaining the reactor stable in the Cold Shutdown conditions.
Therefore, an RHR shutdown cooling subsystem will normally be in operation when a cooldown is in progress or to maintain reactor coolant temperature.
Without it an alternate will be necessary to achieve the plant objective.
During a heatup, the recirculation pumps will normally be in operation to ensure circulation and temperature monitoring.
If the plant is maintaining plant conditions the plant will consider the situation but normally either the recirculation pump will be in service or an RHR shutdown cooling subsystem will in operation to ensure adequate temperature monitoring.
This will ensure the plant objective is being met and to ensure the core is in a safe condition.
JAFNPP is operated in Cold Shutdown, in a safe manner to ensure decay heat is removed so that no core damage could result, therefore the system is normally continuously operated during these conditions.
The requirements in ISTS LCO 3.4.9 are not needed since the Surveillances in ITS 3.4.9 will clearly require temperature monitoring capability and this is accomplished with either one RHR shutdown cooling subsystem or one recirculation pump in operation.
Furthermore, to be consistent with this modification, the allowances in ISTS LCO 3.4.9 Note 1 that both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the requirements in ISTS 3.4.9 ACTION B for no RHR shutdown cooling subsystem and no recirculation pump in operation have been deleted.
ISTS SR 3.4.9.1 (ITS SR 3.4.8.1) requires the verification that one RHR shutdown cooling subsystem or recirculation pump is in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Since the requirement to be in Page 1 of 2 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 (continued) operation has been deleted this SR has been revised to verify each required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position every 31 days.
The Frequency is consistent with similar Surveillances in other LCOs and is considered adequate for this condition.
In Cold Shutdown, the RHR shutdown cooling system is of prime focus of plant operations and since the controls for this system are in the control room the Frequency is considered adequate.
This Frequency is consistent with the same type of Surveillance in other LCOs
( e.g., ITS 3.5.1).
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5, "RCS Pressure Isolation Valve (PIV)
Leakage," is not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
None DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None flTFFFRFNCF BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
None Page 2 of 2 DIFFERENCE Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown MARKUP OF NUREG-1433, REVISION 1, BASES
RHR Shutdown Cooling System-Cold Shutdow n B 34 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4&) Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown BACKGROUtmperrdatefurel of the sudw reactor coolr.Ti eca heneratmust bhea ved to maintain the temperature of the reactor cool an Z%
FI 11in preparation for "Dpe armin reulnoeains, or 4M
~~ the reactor-in the Cold Shutdown condition.
The two redundant, manually controlled shutdown cooling sub ss ems of the RPIR System provide decay heat removal.
PAX Each loop consists of two motor driven pumps, a heat I
exchanger, and associated piping and valves.
Both loops have a common suction from the same recirculation lop Each pump discharges the reactor coolant, after circulation gh the respective heat exchanger, to the reactor via irecirculaton loop.
The RHR heat exchangers
"s Service Water System.
4,~~
rl NýtII APPLICABLE SAFETY ANALYSES Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigatoi of any event or accident evaluated in the safety analyses.
Decay heat removal is, however, an important safety funcJn-that mist be accomplished or core damage could result./. ]
LCO Two RHR shutdown cooling subsystems are required to be E
OPERABLE RHR shutdown cooling subsystem consists of one p~(p,d*OERABLE RHR pump, one _he`tUFr
.qr`ýand the associated.
3 piping and valves.
The two subsystems have a common suction 6.44-4; *source and are allowed to have a common heat exchanger and Sh*exý ev 0
As common discharge piping.
Thus, to meet the LCO, both pumps jala44(continued)
B 3.4-42 zw 5
oi 14 V.
All) pole T eL4 dýý 4&D
RHR Shutdown Cooling System-Cold Shutdown 8.44 '
BASES LCO in one lool~or oneipumpin each of the two loops must be (continued)
Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems.
In MODE 4, the RHR cross tie v e (
1I-V may be opened to allow pumps in one loop vdiJ
\\ies 0 Mo0\\ -20 6
ischarge through the opposite recirculation loop to make
/
-Oat) a complete subsystem.
Additionally, each shutdown cooling subs st is considered OPERABLE if it can be manually a gned e
in the shutdown cooling mode for removal of decay heat.
In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide Cc -oL roo"*
/
redundancy.
Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
OY IOC.4"(Ij However, to ensure adequate core flow to allow for accurate S-average reactor coolant temperature monitoring, nearly continuous operation is required.
ote I permIttboth RHR s utdown coo n
sso he
[sut down fat1-mb -A of* 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tia I hou~rpr*I Note<allows one RJR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests.
These tests may be on the affected RHR System or on some other plant system or c onen k*,
necessitates placing the RHR n an noperable status during the performance.
This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RH subsystems or other operations requiring loss-of redundancy.
(-LBI);
APPLICABILITY In NODE 4, the RHR-,.--e*:*.y in the shutdown cooling mode to remove maintain coolant te erature below0 recirculation pu-s--
__OM In NODES I and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable.
Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.
Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing (continued)
W V c-01 Rev 1, 04/07/95 B 3.4-43 BWR14 STS
RHR Shutdown Cooling System-Cold Shutdown B 3.
BASES APPLICABILITY the steam in the main condenser.
Additionally, in MODE 2 (continued) below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)
(LCO 3.5.1, "ECCS-Operating') do not allow placing the RHR shutdown cooling subsystem into operation.
The requirements for decay heat removal in MODE 3 below the cut in ermissive pressure and in MODE 5 are discussed in "Residual Heat Removal (RHR)
Shutdown Cooling System-Hot Shutdown'; LCO 3.9.D, §Residual Heat Removal (RHR)-High Water Level"; and (LO 3.9A 'Residual Heat Removal (RHR)-Low Water Level.
ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems.
Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems.
As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by LCO Note 0, the remaining subsystem is capable of providing the required decay heat removal.
However, the overall reliability is reduced.
Therefore, an alternate method of decay heat removal must be provided.
With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability.
This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (continued)
Rev 1, 04/07/95 B 3.4-44 B'dR/4 STS
RHR Shutdown Cooling System-Cold Shutdown BASES ACTIONS Ad.
(continued)
Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities.
Furthermore, verification of the functional availability of these alternate method(s) must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
This will provide assurance of continued heat removal capability.
The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature.
Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability.
Alternate methods that can be used include (but are not limited to) ooWplv-1-oo I nsolystem ano atJReaca ae, 4o~jyA& L OI=IUC*
With no RI4R shutdop cooling subsystem and no re irculation o'
pump in operation except as permitted by CO te 1, and ck ) ountil RHR or re rculation pump operation tisc established, alternate mohed of reactor coolant circulation must be PV TL u
placed into s ice.
This will provide the ecessary 5
rcirculation f r monitoring coolant tempera re. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> V-'
(Completion T
is based on the coolant c ulation function and is aodit ed such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is plicable separately or each occurrence involving a loss of coolant circulattoV. Furthermore, verification f the functioning of the al broate method must be reconfi every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaft This will provide assuranc of continued tempera re monitoring capability.
Durin the period when the reactor co ant is being circu ated by an alternate method (o er than by the requ red RHR $hutdownCooling Syst or recirculatio pump),
the eactor coolant temperature an pressure must pe odically monitored to ensure oper function-o the a ernate method.
The once per h ur Completion T is emed appropriate.
(continued)
Rev 1, 04/07/95 8 3.4-45 BWR/4 STS
RHR Shutdown Cooling System--Cold Shutdown Arl B3.44.
BASES (conti nued)
SURVEILLANCE MM REQUIREMENTS i
This Surveillance veife-s--that-one PPRHR sht~own cool ing*
ij*
subsystem or rectji lation pump is an opr n and
\\
~~circulating reacto/ coolant.
The requiredfo rue s
determined by th flow rate necessary to 4rovtde sufficient decay heat reso 1l capability.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is
- 'J suffficient in yew of other visual and audible indications available to *e operator for mmonittoria the RHR subsysteV
- _the contq ro of' REFERENCES Rev 1, 04/07/95 B 3.4-46 BWR/4 STS
C1 Insert SR 3.4.8.1 Verifying the correct alignment for manual, power operated, and automatic valves in the RHR shutdown cooling flow path provides assurance that the proper flow paths will exist for RHR operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing.
A valve that can be manually (from the control room or locally) aligned is allowed to be in a non-RHR shutdown cooling position provided the valve can be repositioned.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.
The Frequency of 31 days is further justified because the valves are operated under procedural control.
This Frequency has been shown to be acceptable through operating experience.
Insert Page B 3.4-46
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1, BASES
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433.
REVISION 1 ITS BASES: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 The requirement in ISTS LCO 3.4.9 (ITS LCO 3.4.8) associated with a recirculation pump or RHR shutdown cooling subsystem being in operation has been deleted.
The requirement that two RHR shutdown cooling subsystems are Operable is considered acceptable.
Requirements for RHR shutdown cooling subsystem operations are adequately controlled by JAFNPP plant operating procedures and policies.
There are no explicit requirements in the CTS for RHR shutdown cooling subsystem operability or that a recirculation pump should be in operation in Cold Shutdown.
- However, ITS SR 3.4.9.1 will require the reactor coolant temperature to be monitored during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing while SR 3.4.9.2 will require these readings to be taken during a planned startup where the reactor will achieve criticality.
These Surveillance will necessitate placing either the recirculation pump or a RHR shutdown cooling subsystem in service to ensure circulation for temperature monitoring or cooling.
The RHR shutdown cooling subsystems remove decay heat to reduce temperature of the reactor coolant to ' 212°F in preparation for performing Refueling or Cold Shutdown maintenance operations, or for maintaining the reactor stable in the Cold Shutdown conditions.
Therefore, an RHR shutdown cooling subsystem will normally be in operation when a cooldown is in progress or to maintain reactor coolant temperature.
Without it an alternate will be necessary to achieve the plant objective.
During a heatup, the recirculation pumps will normally be in operation to ensure circulation and temperature monitoring.
If the plant is maintaining plant conditions the plant will consider the situation but normally either the recirculation pump will be in service or an RHR shutdown cooling subsystem will in operation to ensure adequate temperature monitoring.
This will ensure the plant objective is being met and to ensure the core is in a safe condition.
JAFNPP is operated in Cold Shutdown, in a safe manner to ensure decay heat is removed so that no core damage could result, therefore the system is normally continuously operated during these conditions.
The requirements in ISTS LCO 3.4.9 are not needed since the Surveillances in ITS 3.4.9 will clearly require temperature monitoring capability and this is accomplished with either one RHR shutdown cooling subsystem or one recirculation pump in operation.
Furthermore, to be consistent with this modification, the allowances in ISTS LCO 3.4.9 Note 1 that both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the requirements in ISTS 3.4.9 ACTION B for no RHR shutdown cooling subsystem and no recirculation pump in operation have been deleted.
ISTS SR 3.4.9.1 (ITS SR 3.4.8.1) requires the verification that one RHR shutdown cooling subsystem or recirculation pump is in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Since the requirement to be in JAFNPP Revision A Page 1 of 3
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 (continued) operation has been deleted this SR has been revised to verify each required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position every 31 days.
The Frequency is consistent with similar Surveillances in other LCOs and is considered adequate for this condition.
In Cold Shutdown, the RHR shutdown cooling system is of prime focus of plant operations and since the controls for this system are in the control room the Frequency is considered adequate.
This Frequency is consistent with the same type of Surveillance in other LCOs
( e.g., ITS 3.5.1).
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5, "RCS Pressure Isolation Valve (PIV)
Leakage," not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PA2 The Bases have been revised to be consistent with changes made to the Specifications.
PA3 Editorial changes have been made for enhanced clarification, correction, or improvement with no change in intent.
PA4 The Bases have been modified to reflect plant specific terminology.
PA5 The correct LCO number has been included.
PA6 Editorial change made to ensure the allowances in the LCO Bases do not conflict with the description in the Background.
PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 The Bases have been revised to reflect the JAFNPP specific design.
In addition, a clarification regarding support systems necessary for RHR shutdown cooling system operability has been added since the ITS does not include a specific RHR Service Water System Specification for MODE 4.
DB2 The Bases have been modified to reflect JAFNPP specific references.
Page 2 of 3 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.4.8 - RHR SHUTDOWN COOLING SYSTEM-COLD SHUTDOWN DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
X1 NUREG-1433, Revision 1. Bases reference to "the NRC Policy Statement" has been replaced with 10 CFR 50.36(c)(2)(ii). in accordance with 60 FR 36953 effective August 18, 1995.
Page 3 of 3 Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown "RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
RHR Shutdown Cooling System -Cold Shutdown 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE.
NOTE O
TE-----------
One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.
APPLICABILITY:
MODE 4.
ACTIONS
NOTE ------------------------------------
Separate Condition entry is allowed for each shutdown cooling subsystem.
°..
CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown cooling method of decay heat subsystems inoperable, removal is available AND for each inoperable RHR shutdown cooling Once per subsystem.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Amendment JAFNPP 3.4-16
RHR Shutdown Cooling System-Cold Shutdown 3.4.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify each RHR shutdown cooling subsystem 31 days manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position.
Amendment 3.4-17 JAFNPP
RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown BASES BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.
This decay heat must be removed to maintain the temperature of the reactor coolant s 212°F in preparation for performing refueling operations, or the decay heat must be removed for maintaining the reactor in the Cold Shutdown condition.
The two redundant, manually controlled shutdown cooling subsystems (loops) of the RHR System provide decay heat removal.
Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.
Both loops have a common suction from the same reactor water recirculation loop.
Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via a reactor water recirculation loop.
The RHR heat exchangers transfer heat to the RHR Service Water System.
Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses (Ref. 1).
Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.
The RHR Shutdown Cooling System meets Criterion 4 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).
Two RHR shutdown cooling subsystems are required to be OPERABLE.
An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, one or two RHR service water pumps providing water to the heat exchanger, as required for temperature control, and the associated piping and valves.
The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.
Thus, to meet the LCO. both RHR pumps in one loop (and two RHR service water (continued)
Revision 0 JAFNPP B 3.4-40
RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 BASES LCO (continued) pumps) or one RHR pump in each of the two loops must be OPERABLE.
Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems.
In MODE 4, the RHR cross tie valves (1OMOV-20 and 1ORHR-09) may be opened to allow pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem.
Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (from the control room or locally) in the shutdown cooling mode for removal of decay heat.
In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy.
Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring.
nearly continuous operation is required.
The Note allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests.
These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR shutdown cooling subsystems in an inoperable status during the performance.
This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR shutdown cooling subsystems or other operations requiring loss of redundancy.
APPLICABILITY In MODE 4, the RHR System is required to be OPERABLE so that it may be operated in the shutdown cooling mode to remove decay heat to maintain coolant temperature below 212°F.
Otherwise, a recirculation pump is normally in operation to circulate coolant to provide for temperature monitoring.
In MODES 1 and 2. and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable.
Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.
Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser.
Additionally, in MODE 2 (continued)
Revision 0 JAFNPP B 3.4-41
RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 BASES APPLICABILITY below this pressure. the OPERABILITY requirements for the (continued)
Emergency Core Cooling Systems (ECCS)
(LCO 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.
The requirements for decay heat removal in MODE 3 below the cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.7, "Residual Heat Removal (RHR)
Shutdown Cooling System-Hot Shutdown": LCO 3.9.7, "Residual Heat Removal (RHR)-High Water Level"; and LCO 3.9.8. "Residual Heat Removal (RHR)-Low Water Level."
ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems.
Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems.
As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
A.1 With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by the LCO Note, the remaining subsystem is capable of providing the required decay heat removal.
However, the overall reliability is reduced.
Therefore, an alternate method of decay heat removal must be provided.
With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability.
This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the decay heat removal function (continued)
Revision 0 JAFNPP B 3.4-42
RHR Shutdown Cooling System -Cold Shutdown B 3.4.8 BASES ACTIONS A.1 (continued) and the probability of a loss of the available decay heat removal capabilities.
Furthermore, verification of the functional availability of these alternate method(s) must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
This will provide assurance of continued heat removal capability.
The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature.
Decay heat removal by ambient losses can be considered as. or contributing to, the alternate method capability.
Alternate methods that can be used include (but are not limited to) the Condensate and Main Steam Systems, Reactor Water Cleanup System (by itself or using feed and bleed in combination with the Control Rod Drive System or Condensate System), or a combination of an RHR pump and safety/relief valve(s).
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR shutdown cooling flow path provides assurance that the proper flow paths will exist for RHR operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing.
A valve that can be manually (from the control room or locally) aligned is allowed to be in a non-RHR shutdown cooling position provided the valve can be repositioned.
This SR does not require any testing or valve manipulation: rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.
The Frequency of 31 days is further justified because the valves are operated under procedural control.
This Frequency has been shown to be acceptable through operating experience.
(continued)
JAFNPP B 3.4-43 Revision 0
RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 BASES (continued)
REFERENCES
- 1.
UFSAR, Chapter 14.
- 2.
Revi si on 0 B 3.4-44 JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
DISCUSSION OF CHANGES (DOCs) TO THE CTS NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
FOR LESS RESTRICTIVE CHANGES MARKUP OF NUREG-1433, REVISION 1, SPECIFICATION JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION I MARKUP OF NUREG-1433, REVISION 1, BASES JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1, BASES RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
To determine the conditioof the Reactor Coolant Systei operation of the safetwodevices related to it.
&DL 3.tf~q 1.
CSr
-sui~ and L
mit Ln Lý
ý fqE r~
3'4 I n
lLif WhenTL in thWodcnitoterato eslha
[.44 I
(l-16 The reactor vessel head bolting studs shall not be under tension unless the temperatures of the reactor vessel flange and the reactor head flange are at least 90*F.
5 3.q4.7
[S qq.S During in-service hydrostatic or leak testinq the Re9L Coolant S stem pressure and temperatur h
be or the ho ureAshown iniFgur or the flange and the beltline region, and n
to ae h curve A for the non-beltline regions, and Seni of urve A,, for the bottom head region. The maximum temperature change during any one hour period shall bbe:
When in the cold condition, the reactor vessel head flange and the reactor vessel flange temperatures shall be
- a.
Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor vessel head flange is r120OF and the studs are
- b.
Every 30 minutes when the reactor vessel head flange is r 100OF and the studs are tensioned.
F3.~q*]c.
-Vr-Rhjn--u-rrmin-u-Ties
=prior to and every 30 minutes
£ p 9, 1 J
during tensioning o reactor vessel head bolting L NvE studs.
quf'4 WAiu
- 2.
In-Service Hydrostatic and Leak Tests
[s*R3..44,1 During hydrostatic and leak testing the Reacto lan ystem pressure and temperature shall beu e
L,.JIOWr J30 minutes w consecu iv per ure readingi aew in5 cF of e otr.
.,aew n5 o
dchý otherIq
)
Amendment No. 4,T4.r4W 49 0. 258 136 Pcqe,
ý 4-
,3 1
$n 3Xo<(_ 17 II
JAFMPP
.5 *.*a.
S20OF when to the left of curve C.
- b.
- 1OO F when on or to the"right of curve C.
I
- 3. -
[Lco 34Uq3q During heatup by non-nusfeer mans ImaschanicaIl, &
40]Owring beatup by Non-Nuclea means. cOOldowP following cooldown fdowing nuclear shutdown and low power In'*..
I.3t shutdown and low power physics tests. the hysics tests the Reactor Cooler.t Sstem prw*
s sur dLe tWW jcoolant system pressure and temperature shl u
-tamnhealtan al nh e rlie in M to t
ole eih curve S evfy 30 mltlw Iuwt two rVSOU1l90 TgMVWo rnal change during any one hour sha" be tLCO D3,f,93 During all modes of operation with a critical core lexcept ILfor low plower physics lestsl the~e co ol.svte/
/*
pl~roi-soxro an temlperature shal be*
lrt the rt f/
ýtph curv oC shown In Figur.6-e.* *r 2&.
/
11 maximnum temperature change duri any one how shall "ES j-1 be 9100F.
1(
ing Itpower phyic tets I tl temperature shall bel f withdrawa of control 130 minutes during hei LS5 t q-Z Amendment No. 2, 43, ;;;a. ;gg....
26.2 267 137 4ý low I
(
- -C-)ý-Cafiayx Fly-l
- 4.
M-*-
.1 ;-0,-,*
Paz OV, Z Isk 3ý,qj I
JAFNPP A44d CONDI)TIO) 14 "A)0o'rr J 36 (cont'd) 4.6(+1'1)
- 5.
With 7a oteo I imitso of
.1.
5 exceeded.
m MobE 1 Z1,or.5"wAV
- a.
restore the temperature end or pressure to within the limis within 30 n~xtes,/ 4 an
-evqkMion Io -deli a"
eeltof le o
f.i-,
\\cooalkm I
and deteminhlr thme reactor coolant system remains eabe for continued operations or 117Zi24.
.l
- b.
be In at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and c)In COLD SHUTDOW~IN within rtie"4 5r?- 3iojJ N
&S2 3
-137 an idle rec*rcti*-kon loop shll not be started unlemss
- a.
The temperature differential between the reacto coolant system and the reactor vessel bottom he*
drain lne Is < 145"F, and
- b.
When both loops ae Idle, the temperature dlf-90en between the reactor cool"m system and the ki%
loop to be started Is < 50"F. or
- c.
When only one loop Is Idle, the temnpeatur difference between the Idle loop dj q gi ý is < SO.F" Amendmken No. /3, 179 J.4.1 Not Used I
i-prior to startup of an Idle loop:
r
- a.
The dlfflerehl temperaltur between fth reactor
- C5R3, 4.9-3]
coolant system end the reactor vessel bottom head a ~.
- b.
When both lop we
- Idle, the ciffermntld
- Isek 3,1,1,51 temnperaure betee the reector coolant sysem 1and thedMe loop to bestarted shall or r0i~c When only one loop Is Idle, the temperature ernlbetween the Idle loop and Ahe I
6 lgshall be
~A Ae 57r 3
.t 138e P[yce-3 o (s-14,1
Li 400 200 0
0 1600
/
50 100
/
150
.7 200 250 300 350 Minimum Reactor Vessel Metal Temperature ("F)
FigurepE tlIReactor Vessel Pressure-Temperature Limits Through 24 EFPY Amendment No. 443.468, 258 163 JAFNPP VALID TO 24 EFPY A - System Hydrotest Limit with Fuel in Vessel B - Non-Nuclear Heating Umit C - Nuclear (Core Critical) Limit ABH ANB BBH A
B C
ABH - System Hydrotest Limit with Fuel in Vessel - Bottom Head ANB - System Hydrotest Limit with Fuel in Vessel - Non Beltline BAH - Non-Nuclear Heating Limit Bottom Head BELTLINE SART = 950F (6885S0)
(120,600)
NONBELTINER RTNT-30 *F 1400 1200 U)
"CL
- 0.
'U
- 0.
l U)
En C) 1000 800 600 (195.312.5) 71 SATURATION (120.312.5)/
JAFNPP 1600 "VALID TO 32 EFPY A - System Hydrotest Limit with Fuel In Vessel B - Non-Nuclear Heating Limit 1400 C - Nuclear (Core Critical) Limit 4 AH - System Hydrotest Limit with Fuel in Vessel - Bottom Head ANG - System Hydrotest Limit with Fuel in Vessel - Non Beliline 1200 Bq,*- Non-Nuclear Heating Limit
- Bottom Head CL 0
ART 1000 0m Cr E
0 0
0 C-NON-BELTUNE i
RTNT 3 09 *F*
400 120312.5.1 200 SATURATION 0
0 50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature (OF)
Figur r
Reactor Vessel Pressure-Temperature Limits 4
Through 32 EFPY Amendment No. 4.18.
258 163a
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits DISCUSSION OF CHANGES (DOCs) TO THE CTS
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ADMINISTRATIVE CHANGES Al In the conversion of the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes.
Editorial changes, reformatting, and revised numbering are adopted to make ITS consistent with the conventions in NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4",
Revision 1 (i.e., Improved Standard Technical Specifications (ISTS)).
A2 CTS 3.6.A does not state any Applicability requirements.
ITS 3.4.9 is Applicable "At all times".
Because the CTS does not specifically state Applicability requirements, and the limitations imposed apply at all times, it can be implied that the Specification is also Applicable "At all times."
Since no technical requirements are altered, this change is administrative and has no adverse impact on safety.
A3 CTS 3.6.A.5.a is revised by adding a NOTE (ITS 3.4.9 Condition A Note) which requires that a determination be made whether the RCS is acceptable for continued operation whenever the Condition is entered, regardless of whether compliance with the LCO is restored.
This change only provides clarification, because CTS 3.6.A.5.a already contains this requirement.
Since no technical requirements are altered, this change is administrative and has no adverse impact on safety.
A4 CTS 3.6.A.5 provides actions appropriate for placing the facility in a condition outside the MODE(S) of Applicability when the Applicability is MODES 1, 2, and 3.
Since certain PT limits apply even when not in MODES 1, 2, and 3, Action C was added (refer to DOC M4).
To clarify the use and application of applying the appropriate action depending on the MODE of operation, the specific clarification "in MODES 1, 2, and 3" is added.
No technical requirements are altered, this change is administrative and has no adverse impact on safety.
A5 Not used.
Ia0 Page 1 of 7 Revi sion E JAFNPP
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ADMINISTRATIVE CHANGES A6 The requirement to record the results in CTS 4.6.A.1, 4.6.A.2, 4.6.A.3, 4.6.A.4, 4.6.A.6.a, 4.6.A.6.b, and 4.6.A.6.c (ITS SRs 3.4.9.1, 3.4.9.2, I
3.4.9.3, 3.4.9.5, 3.4.9.6, 3.4.9.7 and 3.4.9.8) is proposed to be deleted and the requirement will be to verify the associated parameters are within the specified limits.
This requirement duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance Records):
maintain records of activities affecting quality, including the results of tests (i.e., Technical Specification Surveillances).
Compliance with 10 CFR 50 Appendix B is required by the JAFNPP Operating License.
The details of the regulations within the Technical Specifications are repetitious and unnecessary.
Therefore, retaining the requirement to perform the associated Surveillances (verifying the specified limits are met) and eliminating the details from Technical Specifications that are found in 10 CFR 50 Appendix B is considered a presentation preference, which is administrative.
A7 Thermal stresses on vessel components are dependent upon the temperature difference between the idle loop coolant and the RPV coolant.
ITS SR 3.4.9.5 ensures the temperature difference between the idle loop and the RPV coolant is acceptable.
The requirements to monitor the temperature difference between an idle loop and an operating loop (CTS 3/4.6.A.6.c) are unnecessary and are deleted since they are redundant to the loop-to-coolant requirement of ITS SR 3.4.9.5.
However, in accordance with procedures and as discussed in the Bases for ITS SR 3.4.9.4, the loop-to-coolant temperature check may use the operating loop temperature as representative of "coolant temperature".
A8 A Note has been added to CTS 4.6.A.1.a and 4.6.A.1.b (Note to ITS SR 3.4.9.8 and 3.4.9.7, respectively) which clarifies that the Surveillances are not required.to be performed until 30 minutes after RCS temperature
- 100'F or 120°F, respectively.
These requirements are consistent with the CTS requirements.
The Frequency of the CTS requirements are every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (CTS 4.6.A.1.a) or 30 minutes (CTS 4.6.A.1.b) when the reactor vessel head flange falls below the prescribed limit. Therefore, the first required Surveillance is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or 30 minutes after the specified temperature is reached.
These requirements are consistent with the proposed Note therefore this change is considered administrative.
This change is consistent with NUREG-1433, Revision 1.
A9 Not used.
Page 2 of 7 Revision E JAFNPP
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - MORE RESTRICTIVE M1 CTS 4.6.A.4 requires that RCS pressure and temperature be recorded within 30 minutes prior to withdrawal of control rods to bring the reactor critical.
ITS SR 3.4.9.2 requires this verification to be performed within 15 minutes prior to control rod withdrawal for the purpose of achieving critical ity. This Frequency is closer to when the control rods will actually be withdrawn and will help ensure the specified are limits are met.
Since the time is limited, this change is considered more restrictive but necessary to ensure the specified parameters are within limits prior to control rod withdrawal where the reactor has a potential of becoming critical.
This change is consistent with NUREG-1433, Revision 1.
M2 CTS 3.6.A.5 requires that, in the event the RCS pressure and temperature limits are exceeded, it be determined that the RCS remains acceptable for continued operation.
There is no Completion Time associated with this requirement.
ITS 3.4.9 Required Action A.2 Completion Time requires that this determination be made in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
This change imposes a time constraint where one does not exist, and is therefore more restrictive but necessary to ensure prompt action is taken to verify the RCPB is acceptable for continuous Operation.
M3 CTS 4.6.A.6 requires that certain RCS differential temperature measurements be recorded within 30 minutes prior to startup of an idle recirculation loop.
ITS SRs 3.4.9.3 and 3.4.9.5 require that these differential temperature measurements be verified within 15 minutes prior to startup of an idle recirculation loop.
This Frequency is closer to when the pumps will actually be started and therefore will help ensure the specified limits are met prior to pump startup.
Since the time is limited, this change is considered more restrictive but necessary to ensure the temperatures are within limits prior to a startup of an idle pump.
This change is consistent with NUREG-1433, Revision 1.
M4 CTS 3.6.A is revised by adding a new action (ITS 3.4.9 ACTION C), which requires that action be initiated immediately to restore the parameters to within limits, and a determination be made as to whether the RCS is acceptable for continued operation prior to entering MODE 2 or 3.
ITS 3.4.9 ACTION C is Applicable at all times other than in MODES 1, 2, and
- 3.
This change imposes additional requirements and is considered more restrictive but necessary for protection of the RCPB.
Page 3 of 7 JAFNPP Revision E
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
LA1 The requirement in CTS 4.6.A.2, 4.6.A.3, and 4.6.A.4 that specifies the criteria for ending the surveillances (performed until two consecutive temperature readings are within 50F of each other) is proposed to be relocated to the Bases.
Requirements of SR 3.4.9.1 provide adequate assurance that heatup and cooldown of the RCS will be monitored and maintained within limits.
As a result, the manner in which JAFNPP determines that a heatup or cooldown has been terminated is not necessary for ensuring limits are met.
Therefore, the relocated criteria for determining when a heatup or cooldown has terminated is not required to be in the ITS to provide adequate protection of the public health and safety.
Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.
LA2 The details in CTS 3.6.A.5.a to perform an "engineering evaluation" to determine the effects of the out-of-limit condition on the structural integrity of the RCS are proposed to be relocated to the Bases.
The requirement in ITS 3.4.9 Required Actions A.2 and C.2 to determine whether the RCS is acceptable for continued operation is adequate to ensure the proper analysis is performed.
Therefore, the relocated details are not required to be in the ITS to provide adequate protection of the public health and safety.
Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.
LA3 The method defined in CTS 4.6.A.6 to evaluate the temperature differential using the temperature at the reactor vessel bottom head "drain line" is proposed to be relocated to the Bases.
The requirement in ITS SR 3.4.9.3 to verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is within the limits is adequate.
As such, these details are not required to be in the ITS to provide adequate protection of public health and safety.
Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.
LA4 Not used.
LA5 The specific requirements in CTS 3.6.A.2, 3.6.A.3, and 3.6.A.4 that operation be on or to the right of the curves of Figure 3.6-1 Part 1 or 2 are proposed to be relocated to the Bases.
These details are not necessary to ensure that P/T limits are met.
The requirements to maintain the P/T limits in accordance with the Figures are still maintained in ITS 3.4.9 and SR 3.4.9.1.
Therefore, the relocated requirements are not required to be in the ITS to provide adequate Revision E JAFNPP Page 4 of 7
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
LA5 (continued) protection of the public health and safety.
Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.
TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
Li CTS 4.6.A.6 requires verification that the temperature differential between the RCS and the reactor vessel bottom head, and between the RCS and an idle recirculation loop, are within limits prior to startup of the idle recirculation loop.
CTS 3.6.A.6 specifies that this is only to be met when Reactor Coolant System temperature is > 1400F.
These requirements are modified by a Note which states that these P/T verifications are only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup (Note to SRs 3.4.9.3 and 3.4.9.5).
Since the overall Applicability of the Specification is reduced (Surveillance is no longer required in MODE 5 with temperatures > 1400F), this change is a relaxation of requirements and is less restrictive.
This change is acceptable because in MODE 5, the recirculation pumps are rarely placed in operation, and the overall stress on limiting components is lower.
Therefore, the differential temperature limits are not required.
This change is consistent with NUREG-1433, Revision 1.
L2 CTS 3.6.A.6.a requires the temperature differential between the reactor coolant system and the reactor vessel bottom head drain line be < 1450F during a recirculation pump startup (CTS 3.6.A.6).
ITS 3.4.9 provides the option to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes.
ITS SR 3.4.9.4 has been added which provides this allowance.
A Note 2 has been added to the requirements of CTS 3.6.A.6.a (ITS SR 3.4.9.3) which provides the option to perform ITS SR 3.4.9.4.
Similarly, a Note 2 has been added to proposed ITS SR 3.4.9.4 which provides the allowance to evaluate the temperature differential in SR 3.4.9.3.
This change is necessary to avoid an unnecessary plant shutdown to restart an idle recirculation pump when the bottom head drain line temperature indicating channel is inoperable, the drain line is plugged, or if the drain flow is low.
The requirement to ensure the differential temperature between the bottom head drain line and the reactor coolant is within limits has been established to assure avoidance of a thermal over stress condition to the Control Rod Drive (CRD) stub tubes and in core housing welds by sweeping hot water across these relatively cooler JAFNPP Page 5 of 7 Revision E
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L2 (continued) vessel structures and associated components.
The temperature in the bottom head region is usually measured by monitoring the temperature of flow being drawn out from the bottom head drain line.
In the past, JAFNPP has experienced the problem of the bottom head drain line being plugged with debris.
In order to have a good temperature reading, it is necessary to have sufficient flow through the bottom head drain line.
General Electric has determined an alternate method to the verification of the differential temperature between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature prior to starting a recirculation pump. This alternative is to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes prior to starting the recirculation pump (GE-NE-208-04-1292, Evaluation of Idle Recirculation loop Restart without Vessel Bottom Temperature Indication for JAFNPP Nuclear Power Plant). The GE alternative is based on an evaluation that collected data of startup testing at a BWR/3 and BWR/4 plant. The results from operating BWR plant provides the basis that if the above restart conditions are met stratification in the lower plenum region will be avoided since there will either be sufficient mixing in the lower plenum or there wasn't time for the condition to develop.
When the active recirculation pump flow exceeds 40% of its rated pump flow under one pump operating condition, the evidence shows, analytically and experimentally, that there is sufficient mixing to prevent the thermal stratification in the lower plenum region. In order, to achieve this flow rate reactor power must be above 25% RTP to clear the feedwater flow interlock at approximately 20% flow.
At 25% RTP and 40% recirculation pump flow, a GE steady state hydraulic computer code F'
predicts the core flow for JAFNPP to be at 35% of rated.
Essentially all of this flow is predicted to be coming down the jet pump diffusers on the active loop into the lower plenum to provide a good mixing effect.
Test data concerning lower plenum temperature was obtained during natural circulation startup test of a BWR/4 to confirm that this core flow is sufficient to ensure that adequate lower plenum mixing takes place under one-loop operation.
A stratified condition existed for very low power levels but it was swept out when core flow reached about 20%
of rated (corresponding to natural circulation at 5% RTP).
This indicated that 20% core flow was enough to sweep out the stratified cold water.
The 35% core flow, as predicted for JAFNPP, is well above the JAFNPP Page 6 of 7 Revision E
ý I
DISCUSSION OF CHANGES ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L2 (continued) 20% threshold, and it should clearly avoid stratification.
In addition, during a hot restart at a BWR/3 with both recirculation pumps at 20%
speed and reactor power at < 1% RTP, it was confirmed that a temperature difference of < 110°F was maintained between the bottom head drain line and the saturation temperature of the vessel. The core flow at this condition was estimated to-be 20%. The results at the BWR/3 is also considered to be conservative for the BWR/4 design at JAFNPP.
Test data from the BWR/3 plant also indicated that stratification does not occur immediately upon low flow conditions.
This test data showed that stratification did not occur until an hour after a main turbine trip and recirculation pump trip.
On this basis, it is believed that restoring the 40% recirculation pump flow condition within 30 minutes is an acceptable criteria for avoiding lower plenum stratification and allowing subsequent restart of the second loop.
SJAFNPP has reviewed GE report to evaluate whether the proposed criteria for JAFNPP would be acceptable.
The-original JAFNPP startup test data for a single recirculation pump operation was examined to determine how far away the plant was from stratification at various flow rates. The results indicate that with a recirculation pump flow rate of 27.5%
(resulting core flow of < 20 Mlb/hr) and power levels between 25% and 44% of RTP, the resultin-g.temperature difference was < 577F.
In addition, startup test data was reviewed for three te7sts in which both recirculation pumps were tripped.
A maximum differential temperature of 44°F was achieved and this}was observed after more than two hours after the pump trip. The results of the JAFNPP review is included in JAF RPT-RWR-02076, Rev 0 (Verification of Alternative Operation Conditions for Idle Recirculation Loop Restart without Vessel Bottom Temperature Indication).
There the proposed alternatives are considered acceptable for JAFNPP.
TECHNICAL CHANGES - RELOCATIONS None Page 7 of 7 JAFNPP Revision E
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC) FOR LESS RESTRICTIVE CHANGES
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
Li CHANGE New York Power Authority has evaluated the proposed Technical Specification change and has concluded that it does not involve a significant hazards consideration.
Our conclusion is in accordance with the criteria set forth in 10 CFR 50.92.
The bases for the conclusion that the proposed change does not involve a significant hazards consideration are discussed below.
- 1.
Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change does not involve any physical alteration of plant systems, structures or components, changes in parameters governing normal plant operation, or methods of operation.
The proposed change relaxes the Applicability for verification that the temperature differential between the RCS and the reactor vessel bottom head, and between the RCS and an idle recirculation loop, are within limits prior to startup of the idle recirculation loop when Reactor Coolant System temperature is > 1407F (MODES 1, 2, 3, 4 and 5), to MODES 1, 2, 3. and
- 4.
These differential temperatures are not assumed to initiate any accident.
Therefore, this change will not increase the probability of an accident previously evaluated.
In MODE 5, the recirculation pumps are rarely placed in operation and the resulting overall stress on limiting components is lower, and therefore, the differential temperature limits are not required.
Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of plant systems, structures or components, changes in parameters governing normal plant operation, or methods of operation.
The proposed change relaxes the Applicability for verification of temperature differential prior to startup of an idle recirculation loop.
Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
Page 1 of 4 JAFNPP Revision A
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
Li CHANGE
- 3.
Does this change involve a significant reduction in a margin of safety?
The proposed change relaxes the Applicability for verification of temperature differential prior to startup of an idle recirculation loop.
The Specifications will continue to require that RCS pressure and temperature be maintained within analysis limits.
Therefore, this change does not involve a significant reduction in a margin of safety.
Page 2 of 4 Revi si on A JAFNPP
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L2 CHANGE The Licensee has evaluated the proposed Technical Specification change and has concluded that it does not involve a significant hazards consideration.
Our conclusion is in accordance with the criteria set forth in 10 CFR 50.92.
The bases for the conclusion that the proposed change does not involve a significant hazards consideration are discussed below.
- 1.
Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change provides an alternate method to ensure the avoidance of thermal over stress condition to the Control Rod Drive (CRD) stub tubes and in-core housing welds during a recirculation pump startup.
The current method is to ensure the temperature differential between the reactor coolant system and the reactor vessel bottom head drain line be
< 145 0F prior to the recirculation pump startup.
ITS 3.4.9 provides the option to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40%
rated flow for a period no longer than 30 minutes.
The method of evaluating the temperature difference does not influence any assumptions of a design bases accident.
Therefore, this change will not significantly increase the probability of any accident previously evaluated.
General Electric has provided an alternative to the verification of the differential temperature between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature prior to starting a recirculation pump. This alternative is to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes prior to starting the recirculation pump (GE-NE-208-04-1292, Evaluation of Idle Recirculation loop Restart without Vessel Bottom Temperature Indication for JAFNPP Nuclear Power Plant). The GE alternative is based on an evaluation that collected data of startup testing at various BWR plants.
The results from operating BWR plant provides the basis that if the above restart conditions are met stratification in the lower plenum region will be avoided.
JAFNPP has reviewed this analysis and has evaluated similar data specific to JAFNPP.
JAF-RPT-RWR-02076, Rev 0 confirms that the conclusions reached in the GE study applies to JAFNPP. Therefore the consequences of any previously evaluated accident will be the same using the existing method.
Therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated.
Page 3 of 4 JAFNPP Revision E
NO SIGNIFICANT HAZARDS CONSIDERATION ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L2 CHANGE
- 2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
This change provides an alternate method to ensure the avoidance of thermal over stress condition to the Control Rod Drive (CRD) stub tubes and in-core housing welds during a recirculation pump startup.
The proposed change does not involve any physical alteration of plant systems, structures or components, changes in parameters governing normal plant operation, or methods of operation.
Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
- 3.
Does this change involve a significant reduction in a margin of safety?
This change provides an alternate method to ensure the avoidance of thermal over stress condition to the Control Rod Drive (CRD) stub tubes and in-core housing welds during a recirculation pump startup.
The current method is to ensure the temperature differential between the reactor coolant system and the reactor vessel bottom head drain line be
< 145 0F prior to the recirculation pump startup.
ITS 3.4.9 provides the option to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40%
rated flow for a period no longer than 30 minutes.
General Electric has provided an alternative to the verification of the differential temperature between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature prior to starting a recirculation pump.
This alternative is to verify the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes prior to starting the recirculation pump (GE NE-208-04-1292, Evaluation of Idle Recirculation loop Restart without Vessel Bottom Temperature Indication for JAFNPP Nuclear Power Plant).
The GE alternative is based on an evaluation that collected data of startup testing at various BWR plants.
The results from operating BWR plant provides the basis that if the above restart conditions are met stratification in the lower plenum region will be avoided.
JAFNPP has reviewed this analysis and has evaluated similar data specific to JAFNPP.
JAF-RPT-RWR-02076, Rev 0 confirms that the conclusions reached in the GE study applies to JAFNPP.
The alternative methods accomplishes the same function as the current method in avoidance of thermal over stress condition to the Control Rod Drive (CRD) stub tubes and in-core housing welds during a recirculation pump startup.
Therefore, this change does not involve a significant reduction in a margin of safety.
JAFNPP Page 4 of 4 Revision E
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits MARKUP OF NUREG-1433, REVISION 1 SPECIFICATION
RCS P/T Limits 3.4p 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4f(RCS Pressure and Temperature (P/T) Limits RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within M limits Magj= 7S APPLICABILITY:
At all times.
CONDITION REQUIRED ACTION CONPLETION TIME A. ---------NOTE------
Required Action A.2 shall be completed if this Condition is entered.
Requirements of the LCO not met in A. I Restore parameter(s) to within limits.
Determine RCS is acceptable for continued operation.
A.2 30 minutes 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> p1q2 B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued) 3.4-23 LCO 3.4.
[3-. A. - 11 Li.A 1-J 116t Ail AIrTTflUC lj 4A, 5, b]
mevi
- ýS
RCS P/T Limits 3.4.0 ArTTn*M*
(inntinsImedl CONDITION REQUIRED ACTION COMPLETION TIME C. --------NOTE --------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed if to within limits.
this Condition is entered.
C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation, or 3.
than MODES 1, 2, and 3.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I
--- - -------------------,NOTE---------------
Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
eu*-n:
- e:':-----
fciS 'r le s su r e, temperatureR~i C
Qnn mw a
are within the limits specified n
l pressure and 0 temperature are criticality limits specified in
[4.4A-2II1 SR3.&
4, A-23]
9,(#. A.
SR 3.4.04 30 minutes 9
(continued)
Rev 1, 04/07/95
[M43 4
I 2.4%
NJ 00- -
=M I"
4.og,.Me
,'(.,
(
3.4-24 BWR/4 STS
SL Insert SR-1
- b.
RCS temperature change averaged over a one hour period is:
- 1.
< 100OF when the RCS pressure and temperature are on or to Ehe right of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing;
- 2.
< 20°F when the RCS pressure of curve C of Figure 3.4.9-1 applicable, during inservice and and temperature are to the left or Figure 3.4.9-2, as leak and hydrostatic testing;
- 3.
< IO0°F during other heatup and cooldown operations.
Insert Page 3.4-24 I
Revision E
RCS P/T Limit 3.4&
Verify the differenc-e-ibetween the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature istwilf*n the TiimT9-Snecif~eo in Me FIL
-NOTE Only required to be met in MODES 1, 2, 3, and 4o Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant NOTE--------------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Once within 15 minutes prior to each startup of a recirculation Pump Once within 15 minutes prior to each startup of a recircul ation pump 30 minutes (continued)
Rev 1, 04/07/95
[-LA L.0A C4 q*
A. I A-
[4- ý, A. ý. 44 13-- 4. A.(.ý 3.4-25 BWR/4 STS
SP Insert Note 2
- 2.
Note required to be performed if SR 3.4.9.4 is satisfied.
9 Insert SR 3.4.9.4 NOTES --------------------
- 1.
Only required to be met in MODES 1, 2.
3, and 4 during recirculation pump startup.
- 2.
Not required to be met if SR 3.4.9.3 is satisfied.
Verify the active recirculation pump flow exceeds 40X of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes.
Once within 15 minutes prior to each startup of a reci rcul ati on pump Insert Page B 3.4-25
&aavý SR 3.4.9.4 1*
RCS P/T Limits I3.4.f I
I ZJ 4--ZT 3
- CI1S, II Rev 1, 04/07/95 Al
3.4-26 BWR/4 STS
TA4'e*4 fPQrc
- Ase 1600 1400 1200 -
VALID TO 24 EFPY A - System Hydrotest Limit with Fuel in Vessel B - Non-Nuclear Heating Limit C - Nuclear (Core Critical) Limit ASH - System Hydrotest Limit with Fuel in Vessel - Bottom Head ANe - System Hydrotest Limit with Fuel in Vessel - Non Bettline B8H - Non-Nuclear Heating Limit Bottom Head CL 0
0 z
- 0.
0 E:3 co (a
RCS P/T Limits 3.4.9 0
50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature (OF)
Figure 3.4.9-1 (page 1 of 1)
Reactor Coolant System Pressure and Temperature Limits through 24 Effective Full Power Years (EFPY)
Amendment (Rev. E)
J~ee~riph-q.~
q-U ABH ANB BaH A
B C
L
/IELTLINE____
ART = 95OF
- i//
'...NON-BELTL NE V88 42D)
RTmoT=30 OF (g0.3ý12-=
ý195,312.5)
(103 1000 800 600 400 200 0
///
v I
/
1600 1400 1200 C)
S1000
- 0.
0 800 c
0 4- 00 4O RCS P/T Limits 3.4.9 3, /4ge 3..?-2.
VALID TO 32 EFPY A - System Hydrotest Umitr with Fuel in Vessel B - Non-Nuclear Heating Limit C - Nuclear (Core Critical) Limit eH ANS B8 -
A sH-System Hydrotest Limit with Fuel in Vessel - Bottom Head AN- - System Hydrotest Limit with Fuel in Vessel - Non Beltline BBH - Non-Nuclear Heating Limit
- Bottom Head/
BEI. FLPINE SA R 'I 109 OF NON-BELTLINE RTNoT = 30 0F (120,312.5)
(
.32 (209,312.5)
SATURAMON 10 50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature CF)
Figure 3.4.9-2 (page 1 of 1)
Reactor Coolant System Pressure and Temperature Limits through 32 Effective Full Power Years (EFPY)
Amendment (Rev. E) rAise*-+ /?e3.f-24
B C
//
200 0
0 JAFNPP
/
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 JAFNPP has not developed the "Reactor Coolant System (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)".
References to limits in the PTLR are replaced with current requirements.
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5, "RCS Pressure Isolation Valve (PIV)
Leakage", is not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PA2 Editorial changes have been made to achieve consistency with the Writer's Guide for the Restructured Technical Specifications.
PA3 Editorial changes have been made with no change in intent.
PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 The bracketed allowance has been deleted since it does not apply to JAFNPP.
DIFFERENCE BASED ON APPROVED TRAVELER (TA)
TA1 The changes presented in Technical Specification Task Force (TSTF)
Technical Specification Change Traveler Number 35, Revision 0, have been incorporated.
DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE MX)
X1 ITS SR 3.4.9.4 has been added to the requirements of ISTS 3.4.10 (ITS 3.4.9) to allow an alternative to the requirements of ITS SR 3.4.9.3.
This Surveillance has been added to the CTS in accordance with Li.
A Note 2 was added to SR 3.4.9.3 which allows the option to perform SR 3.4.9.4.
In addition, subsequent Surveillances have been renumbered, as required.
Page 1 of 1 JAFNPP Revision A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits MARKUP OF NUREG-1433, REVISION 1, BASES
B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.
RCS Pressure and Temperature (P/T) Limits BASES RCS P/T Limits B 3.4
'3zPA BACKGROUND (continued)
B 3.4-47 hp.
6-oJ P.
All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.
These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
1U contains P/T limit curvesu tup, cooldon,rCb inservice leakage and hydrostatic testing,)and the maximum rate of change of reactor coolant temperature.29>
ieppcrye-provides linmts T~r both heftpad*-
Each P/T l curve defines an acceptable region for normal operation. Qhe suaV*s-o1Lhe curves operational guidance during heatup or cooidown maneuveringED*)
_pressure and temperature (nTh fs) are monitored and ompared to the applicable curve to~ieT~ that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Dl reactor coolant pressure boundary (RCPB).
The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref.
1), requires the establishment I".
of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference I requires an adequate margin to brittle failure during normal operation, f
operationalf6fiEERTfrs, and system y rostatic s*1n,.
t fondates the use of the ASNE Code,Section III, PAZ Appendix G (Ref. 2
- The actualih t in the RTW, of the vessel material (
Ul).$l~j*J*(lperiodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTN E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4).
The operating P/T limit curves (2
]
- adjusted,
Insert BKGD The nil-ductility transition (NDT) temperature, RT
, is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner.
The RTNDT increases as a function of neutron exposure at integrated neutron exposures greater than approximately 1017 nvt with neutron energy in excess of 1 MeV.
Insert Page B 3.4-47
RCS P/T Limits BASES BASES BACKGROUND (continued)
Al Si as necessary, based on the evaluation findings and the recommendations of Reference 5.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.
At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.
Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most retitv c-ýý?
The heatup curve represents a different set of restrictions than the cooldown curve because the directionsof the thermal gradients through the vessel wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. _
The)criticality limits include the Reference 1 requirement that they be at least 40"F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB'components.
ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
PPLICABLE The P/T limits are not derived from Design Basis Accident APETY ANALYSES (DBA) analyses.
They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCP9, a condition that is unanalyzed.
Reference 7 establishes the methodology for determining the P/T limits.
Since the P/T limits are not derived from any DBA, thererare no accetance GV Aks SýOeec.'Pcraa.
e -
Pec1 U(continued)
Rev 1, 04/07/95 BWR/4 STSB B 3.4-48
RCS PIT Limits BASES APPLICABLE SAFETY ANALYSES (continued) limits related to the P/T limits.
Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condit RC$ P/T limits satisfy Criterion ion.
2 of Ft-LCO The eloment9s LCO are:
- a.
RCS presesreO h temperature, are within the limits sDS(
- b.
The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is alin in 1MIT5 f It Lduring
_Mcirculation pump startuI* -nd durtng~incea ItHE4MAL POWER or loep owhle operatngat
ý
- c.
The temperature difference between the rea in the respective recirculation loop and ii reactor vessel I---
r h n
a 11
- d.
RCS pressure and temperature are within the e
-L2 Slimits specified in ME I
prior tke2*
achieving criticality; and
- e.
The reactor vessel flange and the head flange temperatures arefitZin A 1et ij ofthe Fwhen tensioning the reactor vessel head bolting studs S4 These limits define allowable operating regions and permit a The 0 +fC ridi large number of operating cycles while also providing a wide C4 kit
'e
\\
margin to nonductile failure.
t S k. t, Pty~
ra. arature liscontro] the thermal j
gradien through the vessel wall MV~ are used as inputs for calcu4 1lating the heatp cooldown,7 an isrice leakage and Ci~ /CS
~
740 ) hydrostatic testing P/T limit curves. Thus, the LCD for the IT PV
- PPV (continued)
Rev 1, 04/07/95 "ný
- - I 0?e B 3.4-49 BWR/4 STS
6ý9 3
Insert LCO-1 In addition, RCS temperature change averaged over a one hour period is:
< 100°F when the RCS pressure and temperature are on or to the right of curve
ý of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing; < 20°F when the RCS pressure and temperature are to the left of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing; and < 100°F during other heatup and cooldown operations; Insert Page B 3.4-49
RCS P/T Limits B3.
BASES LCO (continued) rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.
XpýSeA4 Violation of the limits places the reactor vessel outside of Q
the bounds of the stress analyses and can increase stresses in other RCS components.
The consequences depend on several
- Cfactors, as follows:
I a.
The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;
- b.
The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c.
The existence size& and orientatior@of flaws in the vessel material.
APPLICABILITY ACTIONS The potential for violating a P/T limit exists at all times.
For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup.
Therefore, this LCO is applicable even when fuel is not loaded in the core.
Operation outside the P/T limits while in MODM i, 2, must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range.
violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Most PT liw+
Besides restorin a
within limits, an evaluatlo.is S
required to determine RCs operation can continue.X
-ý (continued)
I Rev 1, 04/07/95 B 3.4-50 BtdR/4 STS
9 3
Insert LCO-2 P/T limit curves are provided for plant operations through 24 EFPY (Figure 3.4.9-1) and 32 EFPY (Figure 3.4.9-2).
Curves A, A. (bottom head), and A.
(non-beltline) establish the minimum temperature for hydrostatic and leak testing, Curves B and B. (bottom head) establish limits for plant heatup and cooldown when the reactor is not critical or during low power physics tests, and Curve C establishes the limits when the reactor is critical. In addition, ART is the adjusted reference temperature.
Insert Page B 3.4-50 Revision E
RCS P/T Limi B 3.4 BASES ACTIONS A,. and A.2 (continued) evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
ASNE Code,Section XI, Appendix E (Ref.
6), may be.used toP support the evaluation.
However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable to accomplish the evaluation of a mild violatton.
re severe violations may requ re special, event specific stress analyses or inspections.
A favorable evaluation must be completed if S* fcontinued operation is desired.
S,~.
Condition A is modified by a Note requiring Required "Action A.2 be completed whenever the Condition is entered.
The Note emphasizes the need to perform the evaluation of 6LL k a",
the effects of the excursion outside the allowable limits.
/
bd Restoration alone per Required Action A.1 is insufficient 2.. olA.ok because higher than analyzed stresses may have occurred and w 4 may have affected the RCPB'integrity.
&4 ej*7A If a Required Action and associated Completion Tim of 7vt I ut A *-
Condition A are not met, the plant must be placed in a lower NODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable ccu-'re-e*
e region.
Etth etnbs i*)i1%
- indicates a need for more careful examinat event, best accomplished wihthe RCS at reduced pressure and temperature.
With the reduced dFI2.
pressure and temperature conditions, the of propagation of undetected flaws is decreas 1.( W*,k00 Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
Rev 1, 04/07/95 BWR/4 STS B 3.4-51 B 3.4-51 BWR/4 STS
RCS P/T Limits B3.4 BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS I'
Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed.
This evaluation must verify that the RCPB integrity is acceptable and must be comlete before approaching criticality or heating up to >
methods may be used, including comparison with r~e-aaliyzed transients, new analyses, or inspection of the components.
ASNE Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.
Verification that operation is withintfuR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.
Thi Frequency j
is considered reasonable in view of the control room indication available to monitor RCS status.
Also, since temperature rate of change limits are specified in hourly d-,4rrd¢IPV.
increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.;
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevantplant procedure for ending the activty a t.~ias~id This SRý modified*w a Note that requires this r-v Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
(continued)
Rev 1, 04/07/95 BdR/4 STS B 3.4-52 F
B 3.4-52 BWR/4 STS
Insert ACTION C Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered.
The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
Insert SR 3.4.9.1 Unlike steady-state operation, these intentional operational transients may be characterized by large pressure and temperature changes, and performance of this SR provides assurance that RCS pressure and temperature remain within acceptable regions of the P/T limit curves as well as within RCS temperature change limits.
Insert Page B 3.4-52
RCS P/T Limit 1___A BASES SURVEILLANCE REQUIREMENTS (continued)
A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.,
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod ithdrawal.
Difretl tmeatrswithin lh) 3U1mIimits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref.,*Are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start."
P An acceptable means of demonstrating compliance Xh the temperature differential requirement in SR 3
.4
-is to compare the temperatures of the operating recirculation loop uvW and the idle loop.
I him~
Ps' 9Aw' I4 CU'tz P
pr" P SR 3_4WL modified by a Note, that requires the
)Q
&'4ý Surveill ance to be prformed only in MODES 1, 2, 3, and 4 stowEWES ar a s
In MODE 5, thel overall stress on limiting components is ower.
Therefore, o /limits aentrqie.
S~~SR A..-
SRd S...i-a"R 3.4-*(
(~/zt Limits on the reactor ves~se flange and head ange temperatures are generally bounded by the other P/T limits (continued)
BWR/4 SIS B 3.4-53 Rev 1, 04/07/95 Rev 1, 04/07195 B 3.4-53 BW/4 STS
a R
Insert SR 3.4.9.3-1 Compliance with the temperature differential requirement in SR 3.4.9.3 is demonstrated by comparing the bottom head drain temperature to the reactor vessel steam dome saturation temperature.
SR 3.4.9.4 requires the verification that the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes.
As specified in Reference 11 and 12, the alternative verification of SR 3.4.9.4 will ensure the temperature differential of SR 3.4.9.3 is met.
S Insert Note 2 SR 3.4.9.3 is modified by a second Note, which clarifies that the SR does not have to be performed if SR 3.4.9.4 is satisfied.
This is acceptable since References 10 and 11 demonstrate that SR 3.4.9.4 is an acceptable alternative.
In addition, SR 3.4.9.4 is modified by a second Note, which clarifies that the SR does not have to be performed if SR 3.4.9.3 is satisfied.
This is acceptable since SR 3.4.9.3 directly ensures there is no stratification.
Insert Page B 3.4-53 1
I kEVISO'i (r
RCS P/T Limits L,
SURVELLANE SR3.
L. 12 nd S 3-0 (continued)Q REQUIRENENTS RU----T during system-eatup and cooldown.
However, operations temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO The flange temperatures mus be verified to be above the W"+kn TiitMOmintes before an while tensioning the vessel head bolting studs to ensure t once the head is tensioned OZ the limits ar satisfied.
When with RCS temperature e
"'F, 30 minute checks of the flange t
Laures a qutred because of the reduced mar
- to e
imits.
When i*-
M*-vtth RCS temperature s g*
C monitoring of the flange temperature is required every j,'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within **
limits
?.s D e c -1r~ d -1fn T Hn E
i e.
. 1
ý ý The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
REFERENCES
- 1.
- 2.
ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
- 3.
ASTM E 185-82, July 1982.
- 4.
10 CFR SO, Appendix H.
le sevf Vesrel Met";dr S.
Regulatory Guide 1.99,. Revision 2IMay 1988.
Rev 1, 04/07/95 BWR/4 STS B 3.4-54
Insert SR 3.4.9.6 V-w-SR 3.4.9.6 is modified by a Note which requires the SR to be performed only when tensioning the reactor vessel head bolting studs.
SR 3.4.9.7 is modified by a Note which states that the SR is not required to be performed until 30 minutes after RCS temperature is s 100OF in MODE 4.
SR 3.4.9.8 is modified by a Note which states that the SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is s 120°F in MODE 4.
These Notes are necessary to specify when the reactor vessel flange and head flange temperatures are required to be within specified limits.
Insert Ref-i1
- 8.
Letter from Guy Vissing (NRC)
Issuance of Amendment No. 258 Nuclear Power Plant, November to James Knubel (NYPA),
to James A. FitzPatrick 29,1999.
- 9.
Insert Ref-2
- 11.
GE-NE-208-04-1292, Evaluation of Idle Recirculation Loop Restart Without Vessel Bottom Temperature Indication for FitzPatrick Nuclear Power Plant, December 1992.
- 12.
JAF-RPT-RWR-02076, Verification of Alternative Conditions for Idle Recirculation Loop Restart Vessel Bottom Temperature Indication, June 25.
Operating without 1995.
Insert Page B 3.4-54 lb4I Revision E
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9 RCS Pressure and Temperature (P/T) Limits JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1, BASES
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 JAFNPP has not developed the "Reactor Coolant System (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLRY).
References to limits in the PTLR are replaced with current requirements.
CLB2 The details of CTS Surveillance Requirement 4.6.A.3 and 4.6.A.4 allowance, for discontinuing the Surveillance when 2 consecutive readings are within 50F of each other, is being retained (see LA1).
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
PAl NUREG-1433 Specification 3.4.5, "RCS Pressure Isolation Valve (PIV)
Leakage," is not incorporated in ITS.
Subsequent ITS Specifications and Bases have been renumbered accordingly.
PA2 Editorial changes have been made for enhanced clarification, correction.
or improvement with no change in intent.
PA3 The Bases have been modified to reflect plant specific nomenclature.
PA4 Editorial changes have been made to maintain consistency with Specification and/or other Bases.
PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 The JAFNPP RCS P/T limits do not include specific requirements for increases in power or flow while operating at low power or low flow.
The JAFNPP P/T limits (associated with temperature differences between the reactor vessel bottom head coolant and the reactor pressure vessel coolant, and temperature differences between reactor coolant in the respective recirculation loop and in the reactor vessel) apply only during a recirculation pump startup.
Therefore, the Bases are revised to reflect the limitations of the Specifications.
DB2 The Bases have been revised to reflect the plant specific References.
Subsequent References have been renumbered, as applicable.
DB3 The brackets have been removed and the plant specific Reference included.
Page 1 of 2 JAFNPP Revi sion A
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433.
REVISION 1 ITS BASES: 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS DIFFERENCE BASED ON APPROVED TRAVELER (TA)
TA1 The changes presented in Technical Specification Task Force (TSTF)
Technical Specification Change Traveler Number 35, Revision 0, have been incorporated.
DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
X1 NUREG-1433, Revision 1, Bases reference to "the NRC Policy Statement" has been replaced with 10 CFR 50.36(c)(2)(ii), in accordance with 60 FR 36953 effective August 18, 1995.
X2 ITS SR 3.4.9.4 has been added to the requirements of ISTS 3.4.10 (ITS 3.4.9) to allow an alternative to the requirements of ITS SR 3.4.9.3.
This Surveillance has been added to the CTS in accordance with LI.
A Note 2 was added to SR 3.4.9.3 which allows the option to perform SR 3.4.9.4.
In addition, subsequent Surveillances have been renumbered.
as required. Modifications have been made to the Bases to reflect the changes made to the Specification and to justify the allowance.
Page 2 of 2 Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION ITS: 3.4.9
-RCS Pressure and Temperature (PIT) Limits RETYPED PROPOSED IMPROVED TECHNICAL SPECIFICATIONS (ITS) AND BASES
RCS P/T Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE --------- A.1 Restore parameter(s) 30 minutes Required Action A.2 to within limits.
shall be completed if this Condition is AND entered.
A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> acceptable for Requirements of the continued operation.
LCO not met in MODE 1.
2, or 3.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment (Rev. E)
JAFNPP 3.4-18
RCS P/T Limits 3.4.9 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. -........
NOTE --------- C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed if to within limits.
this Condition is entered.
AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation, or 3 than MODES 1, 2, and 3.
Amendment (Rev. E)
I?-
JAFNPP 3.4-19
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.9.1 NOTE....................
Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
S........
°.............
Verify:
- a.
RCS pressure and temperature are within the limits specified in Figure 3.4.9-1 or Figure 3.4.9-2, as applicable.
- b.
RCS temperature change averaged over a one hour period is:
- 1. < 100OF when the RCS pressure and lemperature are on or to the right of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing;
- 2. < 20°F when the RCS pressure and lemperature are to the left of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing; and
- 3. < 100°F during other cooldown operations.
heatup and FREQUENCY 4
30 minutes 1
1.
(continued)
Amendment (Rev. E)
I JAFNPP 3.4-20
RCS P/T Limits 3.4.9 SURVE ILLANCE REOUIREMENTS (continued)
Amendment (Rev. E)
SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and temperature are Once within within the criticality limits specified in 15 minutes Figure 3.4.9-1 or Figure 3.4.9-2, as prior to applicable, control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3 NOTES -----------------
- 1.
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
- 2.
Not required to be performed if SR 3.4.9.4 is satisfied.
Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is < 1450F.
startup of a reci rcul ati on pump SR 3.4.9.4 NOTES -------------------
- 1.
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
- 2.
Not required to be met if SR 3.4.9.3 is satisfied.
Verify the active recirculation pump flow Once within exceeds 40% of rated pump flow or the 15 minutes active recirculation pump has been prior to each operating below 40% rated flow for a period startup of a no longer than 30 minutes.
recirculation pump (continued) ilk JAFNPP 3.4-21
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and temperature are Once within within the criticality limits specified in 15 minutes Figure 3.4.9-1 or Figure 3.4.9-2, as prior to applicable, control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3
NOTES -----------------
- 1.
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
- 2.
Not required to be performed if SR 3.4.9.4 is satisfied.
Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is : 145°F.
startup of a recirculation pump SR 3.4.9.4 NOTES -------------------
- 1.
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
- 2.
Not required to be met if SR 3.4.9.3 is satisfied.
Verify the active recirculation drive flow Once within exceeds 40% of rated drive flow or the 15 minutes active loop has been operating below 40%
prior to each rated flow for a period no longer than 30 startup of a minutes.
recirculation pump (continued)
Amendment (Rev. E)
JAFNPP 3.4-21
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued:
SURVEILLANCE SR 3.4.9.5
NOTE -------------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature r 50 0F.
FREQUENCY Once within 15 minutes prior to each startup of a recirculation pump SR 3.4.9.6
NOTE -------------------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor vessel flange and head 30 minutes flange temperatures are > 900F.
NOTE -------------------
Not required to be performed until 30 minutes after RCS temperature < lO0F with any reactor vessel stud tensioned.
Verify reactor vessel flange and head 30 minutes flange temperatures are k 900F.
NOTE ------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature : 120°F with any reactor vesselstud tensioned.
Verify reactor vessel flange and head 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> flange temperatures are k 90 0F.
Amendment (Rev. E)
V.A JAFNPP 3.4-22
RCS P/T Limits 3.4.9 1600-I / /
I `J/N B
C
/"BEL'F'TLINE____
ART = 950F
___ 5 50" NON-BELTL NE RTNDT-30 OF SATURATION 50 100 150 200 250 300 350 Temperature Minimum Reactor Vessel Metal Temperature (OF)
Figure 3.4.9-1 (page 1 of 1)
Reactor Coolant System Pressure and Limits through 24 Effective Full Power Years (EFPY)
Amendment (Rev. E)
VALID TO 24 EFPY A - System Hydrotest Limit with Fuel in Vessel B - Non-Nuclear Heating Limit C - Nuclear (Core Critical) Limit ABH ANB BBH A
ASH - System Hydrotest Limit with Fuel in Vessel - Bottom Head AN8 - System Hydrotest Limit with Fuel in Vessel - Non Beltline BBH - Non-Nuclear Heating Limit Bottom Head 1400 4 1200 4 0r) 10 S1000 800 600 0
0 IM_
E 40 0D
.J 4
E 0
400 200 0
0
/.
JAFNPP 3.4-23
RCS P/T Limits 3.4.9 1600 1400 1200 C.
S1000 0
0 080 t
r:
600 400 VALID TO 32 EFPY A - System Hydrotest Limit with Fuel in Vessel B - Non-Nuclear Heating Limit C - Nuclear (Core Critical) Limit ASH-System Hydrotest Limit with Fuel in Vessel - Bottom Head ANB*- System Hydrotest Limit with Fuel in Vessel - Non Beltline BBH - Non-Nuclear Heating Limit
- Bottom Head B
/
J C
iZYI _I/_
// //
/
/ /
/
BEqrLINE ARI 109 OF (12M.550)
NON-BELTLINE RTNDT = 3 O (68,420)3D (120.312.5)
(1192,312.5)
(209.312.5) i /
RIQ 50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature (OF)
Figure 3.4.9-2 (page 1 of 1)
Reactor Coolant System Pressure and Temperature Limits through 32 Effective Full Power Years (EFPY)
Amendment (Rev. E) kBH ANS B81 T77f 200 0
0 A
3.4-24 JAFNPP
B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.9 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.
These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
This Specification contains P/T cooldown, inservice leakage and criticality and also limits the reactor coolant temperature.
limit curves for heatup, hydrostatic testing, and maximum rate of change of Each P/T limit curve defines an acceptable region for normal operation.
The curves are used for operational guidance during heatup or cooldown maneuvering.
Pressure and temperature are monitored and compared to the applicable curve to ensure that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).
The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference 1 requires an adequate margin to brittle failure during normal operation, abnormal operational transients, and system inservice leakage and hydrostatic tests.
It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).
The nil-ductility transition (NDT) temperature, RTT. is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner.
The RTWT increases as a function of neutron exposure at integrated neutron exposures greater than approximately 10'7 nvt with neutron energy in excess of 1 MeV.
(continued)
Revision E JAFNPP B 3.4-45
RCS P/T Limits B 3.4.9 BASES BACKGROUND The actual shift in the RT*T of the vessel material is (continued) determined periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4).
The operating P/T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.
At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.
Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive locations.
The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.
However, the P/T limit curves reflect the most restrictive of the heatup and cooldown curves.
The P/T criticality limits include the Reference 1 requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
ASME Code,Section XI. Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
(continued)
JAFNPP B 3.4-46 Revision E
RCS P/T Limits B 3.4.9 BASES (continued)
APPLICABLE SAFETY ANALYSES The P/T limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.
Reference 7 establishes the methodology for determining the P/T limits.
Reference 8 approved the curves and limits required by this Specification.
Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits.
Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii)
(Ref. 9).
The elements of this LCO are:
- a.
RCS pressure and temperature are within the limits specified in Figure 3.4.9-1 or Figure 3.4.9-2, as applicable.
In addition, RCS temperature change averaged over a one hour period is:
- 100°F when the RCS pressure and temperature are on or to the right of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing: <
20°F when the RCS pressure and temperature are to the left of curve C of Figure 3.4.9-1 or Figure 3.4.9-2, as applicable, during inservice leak and hydrostatic testing; and < 100°F during other heatup and cooldown operations;
- b.
The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is K 145 0F during recirculation pump startup;
- c.
The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is & 50°F during recirculation pump startup;
- d.
RCS pressure and temperature are within the limits specified in Figure 3.4.9-1 or Figure 3.4.9-2. as applicable, prior to achieving criticality; and (continued)
Revision E LCO I
I I
JAFNPP B 3.4-47
RCS P/T Limits B 3.4.9 BASES LCO
- e.
The reactor vessel flange and the head flange (continued) temperatures are a 90°F when tensioning the reactor vessel head bolting studs and when any stud is tensioned.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
The limits on the rate of change of RCS temperature, influenced by RCS flow and RCS stratification, control the thermal gradient through the vessel wall.
For this reason, both RCS temperature and RPV metal temperatures are used as inputs for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves.
Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.
P/T limit curves are provided for plant operations through 24 EFPY (Figure 3.4.9-1) and 32 EFPY (Figure 3.4.9-2).
Curves A, ANH (bottom head), and At (non-beltline) establish the minimum temperature for hydrostatic and leak testing, Curves B and BH (bottom head) establish limits for plant heatup and cooldown when the reactor is not critical or during low power physics tests, and Curve C establishes the limits when the reactor is critical. In addition. ART is the adjusted reference temperature.
Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components.
The consequences depend on several factors, as follows:
- a.
The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature:
- b.
The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced): and
- c.
The existence, size, and orientation of flaws in the vessel material.
(continued)
Revision E JAFNPP B 3.4-48
RCS P/T Limits B 3.4.9 BASES APPLICABILITY The potential for violating a P/T limit exists at all times.
For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup.
Therefore, this LCO is applicable even when fuel is not loaded in the core.
ACTIONS A.1 and A.2 Operation outside the P/T limits while in MODE 1, 2, or 3 I
must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range.
Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation can continue.
This evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation.
However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable to accomplish the engineering evaluation of a mild violation.
A mild violation is one which is technically acceptable because it is bounded by an existing evaluation or one which reasonably can be expected to be found acceptable following evaluation.
More severe violations may require special, event specific stress analyses or inspections.
A favorable evaluation must be completed if continued operation is desired.
Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered.
The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
(continued)
Revision E JAFNPP B 3.4-49
RCS P/T Limits B 3.4.9 BASES ACTIONS B.1 and B.2 (continued)
If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region.
Either occurrence indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature.
With the reduced pressure and temperature conditions, the likelihood of propagation of undetected flaws is decreased.
Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed.
This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 2120F.
Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.
ASME Code,Section XI, Appendix E (Ref. 6). may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.
Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered.
The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
Restoration alone per Required Action C.1 is insufficient (continued)
Revision E B 3.4-50 JAFNPP
RCS P/T Limits B 3.4.9 BASES ACTIONS C.1 and C.2 (continued) because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within RCS pressure and temperature limits as well as within RCS temperature change limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.
This is accomplished by monitoring the bottom head drain.
recirculation loop, and RPV metal temperatures.
This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.
- Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
The limits of Figures 3.4.9-1 and 3.4.9-2 are met when operation is on or to the right of the applicable curve.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
In general, if two consecutive temperature readings taken : 30 minutes apart are within 50F of each other the activity can be considered complete.
This SR is modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
Unlike steady-state operation, these intentional operational transients may be characterized by large pressure and temperature changes, and performance of this SR provides assurance that RCS pressure and temperature remain within acceptable regions of the P/T limit curves as well as within RCS temperature change limits.
SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before (continued)
Revision E JAFNPP B 3.4-51
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.2 (continued)
REQUIREMENTS withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
SR 3.4.9.3. SR 3.4.9.4, and SR 3.4.9.5 Differential temperatures within the specified limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 10) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
Compliance with the temperature differential requirement in SR 3.4.9.3 is demonstrated by comparing the bottom head drain temperature to the reactor vessel steam dome saturation temperature.
SR 3.4.9.4 requires the verification that the active recirculation pump flow exceeds 40% of rated pump flow or the active recirculation pump has been operating below 40% rated flow for a period no longer than 30 minutes.
As specified in Reference 11 and 12, the alternative verification of SR 3.4.9.4 will ensure the temperature differential of SR 3.4.9.3 is met.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.5 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3, SR 3.4.9.4 and SR 3.4.9.5 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4 during a recirculation pump startup since this is when the stresses occur.
In MODE 5, the (continued)
Revision E JAFNPP B 3.4-52
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3, SR 3.4.9.4 and SR 3.4.9.5 (continued)
REQUIREMENTS overall stress on limiting components is lower.
Therefore, AT limits are not required.
SR 3.4.9.3 is modified by a second Note, which clarifies that the SR does not have to be performed if SR 3.4.9.4 is satisfied.
This is acceptable since References 10 and 11 demonstrate that SR 3.4.9.4 is an acceptable alternative.
In addition, SR 3.4.9.4 is modified by a second Note, which clarifies that the SR does not have to be performed if SR 3.4.9.3 is satisfied.
This is acceptable since SR 3.4.9.3 directly ensures there is no stratification.
SR 3.4.9.6, SR 3.4.9.7, and SR 3.4.9.8 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown.
However, operations when any reactor vessel stud is tensioned with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
The flange temperatures must be verified to be above the limits within 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied.
When any reactor vessel stud is tensioned with RCS temperature r eqi0rF 30 minute checks of the flange temperatures are required because of the reduced margin to the limits.
When any reactor vessel stud is tensioned with RCS temperature K 120 0F, monitoring l
of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within specified limits.
The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
SR 3.4.9.6 is modified by a Note which requires the SR to be performed only when tensioning the reactor vessel head Iting studs.
SR 3.4.9.7 is modified by a Note which states that the SR is not required to be performed until 30 minutes after RCS temperature is & 100°F in MODE 4.
SR 3.4.9.8 is modified by a Note which states that the SR is (continued)
Revision E JAFNPP B 3.4-53
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.6, SR 3.4.9.7, and SR 3.4.9.8 (continued)
REQUIREMENTS not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is g 120OF in MODE 4.
These Notes are necessary to specify when the reactor vessel flange and head flange temperatures are required to be within specified limits.
REFERENCES
- 1.
- 2.
ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
- 3.
ASTM E 185-82, July 1982.
- 4.
- 5.
Regulatory Guide 1.99, Revision 2, Radiation Embrittlement Of Reactor Vessel Materials, May 1988.
- 6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
- 7.
NEDO-21778-A, Transient Pressure Rises Affecting Fracture Toughness Requirements For Boiling Water Reactors, December 1978.
- 8.
Letter from Guy Vissing (NRC) to James Knubel (NYPA),
Issuance of Amendment No. 258 to James A. FitzPatrick Nuclear Power Plant, November 29, 1999.
- 9.
- 10.
UFSAR, Section 14.5.
- 11.
GE-NE-208-04-1292, Evaluation of Idle Recirculation Loop Restart Without Vessel Bottom Temperature Indication for FitzPatrick Nuclear Power Plant, December 1992.
- 12.
JAF-RPT-RWR-02076, Verification of Alternate Operating Conditions for Idle Recirculation Loop Restart Without Vessel Bottom Temperature Indication, June 25, 1995.
Revision E JAFNPP B 3.4-54
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.5 RCS Pressure Isolation Valve (PIV) Leakage THIS SPECIFICATION IS DELETED.
THERE ARE NO REQUIREMENTS FOR THIS SPECIFICATION AT JAFNPP; THEREFORE THIS MARKUP PACKAGE CONTAINS ONLY THE FOLLOWING SECTIONS:
MARKUP OF NUREG-1433, REVISION 1, SPECIFICATION JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION I MARKUP OF NUREG-1433, REVISION 1, BASES JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1, BASES
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION N3.4.5 RCS Pressure Isolation Valve (PIV) Leakage MARKUP OF NUREG-1433, REVISION 1 SPECIFICATION NUREG:
RCS PIV Leakage 3.4.5 3.4 REACTOR 3.4.5 RCS Pressure LCO 3.4.5 from each RCS PIV shall be within 1 APPLICABILITY:
operation.
in the residual heat removal (RHR) flow path when in, or during the from, the shutdown cooling mode of ition entry is allowed for each fli le Conditions and Required Actions
ýPIVs.
A.
One or more flow paths with leakage from one or more RCS PI~s not within limit.
Ea valve used to satisfy Requ ced Action A.1 and Requir'm Action A.2 must have been ef* fed to meet SR 3.4.5.
and be in the reactor cooý nt pressure boundary [or e high pressure portio of the system].
(continued)
BWR/4 STS I
.4-9
A*TT fN*
X NIINREQUIRED ACT COMPLETION TIME A. (continued A.1 Isolate the high 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> pressure portion \\of the affected system from the low pressure portion by use of one closed manual, de-acti vated automatic, or check A.
ae the high 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> p e u eportion of t e a ced system from th ow pressure potoc
- o~ssecond cb
- anual, dc-acti vated automatic, or c c
valve.
B. sReqirted Acto a~
nd 9.1 Be in MODE 3.
12 ors Time not met.
ktN
.E A.2 IB ae th e high 4.23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> Rev 1, 04/07/95 BWR/4 STS
'3.4-10
REOUIREMENTS SR 3.4.5.1 Verify equivalent leakagL is g 0.5 gpu per nominal up to a maximum of 5 gpm, pressure k [
I and : [
] ps.
In accordance with the Inservice Testing Program or
[18] months Rev 1, 04/07/95 m
n A4-11 BWR/4 STS
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION N3.4.5 RCS Pressure Isolation Valve (PIV) Leakage JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1 NUREG:
JUSTIFICATION FOR DIFFERENCES
-NUREG: 3.4.5
- RCS PRESSURE ISOLATION VALVE (PIV)
LEAKAGE RETENTION OF EXISTING REOUIREMENT (CLB)
CLB1 NUREG-1433 Specification 3.4.5 sets forth Limiting Conditions for Operation and Surveillance Requirements for Reactor Coolant System (RCS) pressure isolation valve (PIV) leakage.
PIVs are defined as any two valves in series within the reactor coolant pressure boundary (RCPB) which separate the high pressure RCS from an attached low pressure system.
These valves are normally closed during power operation.
The Reactor Safety Study (WASH-1400) identified the potential intersystem loss of coolant accident (Event V) in a PWR as a significant contributor to the risk of core melt.
In this scenario, check valves fail in the injection lines of the RHR or low pressure injection systems, allowing high pressure reactor coolant to enter low pressure piping outside containment.
Subsequent failure of this low pressure piping would result in loss of reactor coolant outside containment and subsequent core meltdown.
Similar scenarios were also determined to be possible in BWRs.
All plants licensed since 1979 have PIVs listed in their Technical Specifications, along with testing intervals, acceptance criteria, and limiting conditions for operation.
Certain older plants were required to periodically leak test, on an individual basis, only those PIVs which were listed in an Order dated April 20, 1981 (Event V Order).
That Order was sent to 32 operating PWRs and 2 operating BWRs.
Other older plants have had no specific requirements imposed to individually leak test any of their PIVs.
A number of events have occurred involving leakage past PIVs, failures of the valves, inadvertent actuation of the valves, or mispositioning of the valves.
As a result, the NRC issued Generic Letter 87-06, "Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves."
which requested that licensees submit (1) a list of all PIVs in their plant: and (2) a description of the periodic tests or other measures performed to assure the integrity of the valve as an independent barrier of the RCPB, along with the acceptance criteria for leakage, operational limits, and frequency of test performance.
NYPA responded to Generic Letter 87-06 by letter JPN-87-034, dated June 11, 1987.
All PIVs are tested in accordance with 10 CFR 50. Appendix J, Type B test requirements, except for certain testable check valves, which are cyclically pressure tested in accordance with the JAFNPP Inservice Testing Program.
Page 1 of 2 JAFNPP Revision A
NUREG: 3.4.5 JUSTIFICATION FOR DIFFERENCES
- RCS PRESSURE ISOLATION VALVE (PIV)
LEAKAGE RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 (continued)
JAFNPP was licensed prior to 1979, and was not a recipient of the Event V Order to perform periodic leak tests of PIVs.
Therefore, the requirements of NUREG-1433 Specification 3.4.5 do not apply to JAFNPP. and are not incorporated in the ITS.
Subsequent Specifications are renumbered accordingly.
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
None PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
None flTFFFPPJrP RAqZRf Aid ADDDAV*fl T1DA1I:1 0D (TA)
None DIFFERENCE BASED ON PENDING TRAVELER (TP)
None DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
None Page 2 of 2 ln~il l.lr
.ll.rmr i.#--*mln nm~l /-1 D
\\D l611 ln'vlrn.
JAFNPP Revision A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.5 RCS Pressure Isolation Valve (PIV) Leakage MARKUP OF NUREG-1433, REVISION 1, BASES
SB8 3.4.5 B 3..5 RCS Pressure Isolation Valv (PIY) Leakage BASES N
/
Rev 1, 04/07/95 BWR/4 STS B 3.4-22 B 3.4-22 BACKGROUND The function of RCS PIVS is, separate the high press'ure RCS from an attached low press e system.
This protects the RCS pressure boundary described 10 CFR 50.2, CFR 50.55a(c), and GDC 55 of I CFR 50, Appendix A (R s. 1, 2, and 3).
RCS PIVs are fined as any two no ily closed valves in series with the reactor coolant press boundary (RCPB).
PIVs are des ned to meet the require nts of Reference 4.
During the lives, these valves ca produce varying amounts of reac r coolant leakage th ugh either normal operational we or mechanical deteri orati o The RCS PIV LCO llows RCS high pressure operation n
leakage through t se valves exists In amounts that do not compromise safety.
The PIV leakage limit applies to each individual valve.
L kage through these valves is not included in any allowa e LEAKAGE specified in LCO 3.4.4, "RCS Operational LEAKAG Although this specification rovides a limit on allowable IV leakage rate, its main pu ose is to prevent o rpressure failure of the 1 ressure portions of con ecting systems.
The leakage inmit is an indication that the Vs between the RCS and the c necting systems are degrad or degrading.
PIV leakage uld lead to overpres re of the low pressure pipi or components.
Failure c equences could be a loss of oolant accident (LOCA) outs e of containment, an unanal ed event that could degrade he ability for low pressure njection.
A study (Ref. 5) valuated various PTV config ations to determine the prob ility of intersystem LOCAs.
This study concluded that per ic leakage testing of the Ps can substantially reduce tersystem LOCA probability.
PIVs are provided to iso te the RCS from the followi typically connected syst
- a.
System; Core Spray System; (continued)
RCS PIV Leakage B 3.4.5 (conine d) d. Reactor Core lation Cooling System.
the effect on the probability of in rsyste LOCAs.
This "A
Ctudy concluded that periodic leakagetesting of the PI~s stantially reduce the probabili on intersysteT DIV T~a so. is not considered in any Design is Accident analyses.W This Specification provides for moiring the condition the RCPB to detect PIV degradation t*nt has the potential to ause a LOCA outside of containment.
RCS PlY leakage satis s Criterion 2 of the NRC Policy Statement.
LCO RCS PIV leakage is le age into closed systems connected to the RCS.
Isolation va leakage is usually on the order of
&a
- .41.
drops per minute.
LeaKa that increases sly",
canmmim suggests that something is rationally wrong and corrective action must be ta n.
Violation of this LCO could result in continued degr ation of a PIV, which could lead to overpressurization of a pressure system and the "loss of the integrity of a fissio roduct barrier.
Th LCO PlY leakage limit is 0.5 gpm r nominal inch of val size with a maximum limit of 5 g (Ref. 4).
Referen 7 permits leakage testing at a lo r pressure different 1 than between the specified maxi RCS pressure and the no I pressure of the connected system uring RCS operation (thmaximum pressure differential)..T observed rate may be adsted to the maximum pressure differ tial by assuming leakage is directly proportional to the pressure differential to t one-half power.
(continued)
BWR/4 STS B 3.4-23 Rev 1, 04/07/95 0
RCS PIV Leakage B3.4.5 nb~ required to meet the requirements of this 04*
when in, or *Jring transition to or from, the RHR shutdo cooling mode AfP eration A L L In MODES1 and 5, leakage limits are not provided becau the lower eactor coolant pressure results in a reduced potential f i
eakage and for a LOCA outside the o
ring,÷,42 nn t
ordinalv. the ootrntial for the consequences o actor coolant leakage is far lower during these MODES.
AC NS The ACTIONS are modified two Notes.
Note 1 has been provided to modify the ACT related to RCS PIV flow paths.
Section 1.3, Complet n Times, specifies once a Condition has been entered, su equent divisions, subsystems, components, or varia es expressed in the Condition discovered to be 1nopera 1. or not within limits will not result in separate entry I the Condition.
Section 1.3 also specifies Required ions of the Condition ontinue to apply for each dditional f lure, with C letion Times based on initial entry
- o the Condition.
er, the Required Actions for the Condi on of RCS PIV leaka limits exceeded provide appropriate c ensatory measure for separate affected RCS PIV flow pat As such, a Note ha been provided that allows separate Con tion entry for ch affected RCS PIV flow path. Note 2 ires an evaluatio of affected systems if a PIV is inoperable.
The leakage ma have affected system OPERABILITY, or isolation a
aking flow path with an alternate valve may have degraded the bility of the interconnected system to perform its safety unction.
As a result, the applicable Conditions and Requi d Actions for systems made inoperable by PIVs must be entere This ensures appropriate remedial actions are taken, if ne ssary, for the affected systems.
A.1 and A-2 If leakage from one or more RCSP s is not within limit, the flow path must be isolated by a least one closed manual, deactivated automatic, or ch k valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(continued)
BWR/4 STS B 3.4-24 Rev 1, 04/07/95 I
" _' 's
ý AC1 SR 3.4.5.)
Performance of I kage testing on each RCS PlV isre uired to verify that Is age is below the specified limit and to identify each leaki valve.
The leakage limit of 0.5 gpm per inch of nominal lve diameter up to 5 gpm maximum applies to each valve.
Leakage testing requires a stable pressure condition.
Fo the two PIVs in series, the leakage (continued) I Rev 1, 04/07/95 CL8I 1
8 3.4-25 S
4 TIONS A1.andLA.2 (continued)
Required Action A.1 and Requi d Action A.2 are modified by a Note stating that the valves ed for isolation must meet the same leakage requirements as he PIVs and must be on the CPB [or the high pressure portion f the system].
Fo hours provides time to reduce Is age in excess of the all able limit and to isolate the fl ath if leakage canno be reduced while corrective actlo to reseat the leakin PIVs are taken.
The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> all time for these actions nd restricts the time of operation ith leaking valves.
Required Act n A.2 specifies that the double is ation barrier of t valves be restored by closing anoth valve qualified for i latlon or restoring one leaking PI The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completi Time considers the time required to complete the actio the low probability of a second va e
failing during this ime period, and the low probability of a pressure boundary r ture of the low pressure ECCS piping when overpressurized t reactor pressure (Ref. 7).
B.1esnsur If leakage cannot be reduced the system Isolated, the plant must be brought to a MOD In which the LCO does not pply.
To achieve this status, e plant must be brought to E 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 ithin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
This ac on may reduce the leakage and alo reduces the potential for LOCA outside the containment.
e Completion Times are asonable, based on operating expe jence, to achieve the re ired plant conditions from full wer conditions in an orde manner and without challenging ant systems.
k BR4 STq
BASES \\
SURVEILLANCE SRL.4.5A.1. (continued)
REQUIREMENTS
\\requirement applies to each val individually and not to e combined leakage across both alves.
If the PIVs are n~t individually leakage tested, oa valve may have failed coletely and not be detected if tl other valve in series meet the leakage requirement.
In th s situation, the protec on provided by redundant valve would be lost.
The1lB ma h Frequency required by the In Nitce Testing Program is Ithin the ASNE Code,Section XI Frequency requirement kd is based on the need to perf*,
this Surveillance ri
- ng an outage and the potentia *for an unplanned trasn f the Surveillance were pe frued with the reactor at po 5r.
This SR is modifiedoy a Note that states the leakage Surveillance is not rulred to be performed in NODE 3.
Entry Into MODE 3 is pe itted for leakage testing at high differential pressures w h stable conditions not possible i n the lower NODES.
\\
REFERENCES
- 1.
10 CFR SO.SSa(c).
- 3.
- 4.
E, Boiler and Pres S.
NUR 0677, Nay 1980.
B6.
VR4 StBon
[ ].
- 7.
NED)C-3133 *November BW/
ST B 3.4-26 i,
sure Vessel Co,Section XI.
1986.
Rev 1, 04/07/95 C-LI 4
I
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION N3.4.5 RCS Pressure Isolation Valve (PIV) Leakage JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1, BASES NUREG:
JUSTIFICATION FOR DIFFERENCES NUREG BASES: 3.4.5 - RCS PRESSURE ISOLATION VALVE (PIV)
LEAKAGE RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 The Bases for NUREG-1433 Specification 3.4.5 are deleted.
NUREG-1433 Specification 3.4.5 does not apply to JAFNPP, and is not incorporated in the ITS.
Subsequent Specifications and Bases are renumbered accordingly.
PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
None PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
None nTFFPPFIJC BASED ON APPROVED TRAVELER (TA)
None DIFFERENCE None DIFFERENCE FOR OTHER REASONS THAN ABOVE MX)
None Page 1 of 1 i.Wrl pm.rlm*1r.i-BASED ON PENDING TRAVELER (TP)
Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.11 Reactor Steam Dome Pressure THIS SPECIFICATION IS DELETED.
THERE ARE NO REQUIREMENTS FOR THIS SPECIFICATION AT JAFNPP; THEREFORE THIS MARKUP PACKAGE CONTAINS ONLY THE FOLLOWING SECTIONS:
MARKUP OF NUREG-1433, REVISION 1, SPECIFICATION JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1 MARKUP OF NUREG-1433, REVISION I, BASES JUSTIFICATION FOR DIFFERENCES (JFDs) FROM NUREG-1433, REVISION 1, BASES
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.11 Reactor Steam Dome Pressure MARKUP OF NUREG-1433, REVISION 1 SPECIFICATION
3 1
Reactor Steam Dome Pressure 3.4.11 3.4 REACTOR COOLANT SYS RS LCO 3.4.1 1
heSreactor steam dome pressure shall be s t1020 p
s 3.4P11 Reactor SteSm e Pressure CONDITION REQUIRED INCOMPLETION TIME A. Reactor steam dome A.P store reactor steam 15 minutes nrLuO nnt within
/dome pressure to I
limit.
within limit.
B. Required Action a 8.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Comp ion L Time not met./
SU ZILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 Vertfy reactor ste idome pressure is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> S[(1020] psig.
Rev 1, 04/07/95 BWR/4 STS 3.4-27
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.11 Reactor Steam Dome Pressure JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 NUREG: 3.4.11 - REACTOR STEAM DOME PRESSURE RETENTION OF EXISTING REQUIREMENT (CLB)
None PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
None PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 NUREG-1433 Specification 3.4.11, "Reactor Steam Dome Pressure." is not incorporated in the ITS.
This NUREG Specification is required to help ensure the vessel overpressure protection analysis can be met by ensuring the initial conditions of the event are preserved.
The JAFNPP site specific overpressure protection analysis is analyzed with an initial condition equivalent to the analytical limit (1094 psig) associated with the Reactor Protection System Instrumentation Reactor Pressure - High Function in ITS 3.3.1.1. Reactor Protection System Instrumentation.
A CHANNEL CHECK of the associated instrumentation is required to be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during operations in MODES 1 and 2 which is consistent with the Surveillance Frequency in ISTS SR 3.4.11.1.
The CHANNEL CHECK Surveillance will ensure reactor pressure is below the Allowable Value (1080 psig) which is lower than the analytical limit associated with this Function as well as below the initial condition of reactor pressure assumed in the overpressure protection analysis.
In fact, since reactor pressure is normally at or below 1040 psig during reactor operations, action will be taken at a much lower pressure to restore reactor pressure or to restore inoperable channels than what is required by ISTS 3.4.11.
The requirements of ITS 3.3.1.1. Reactor Protection System Instrumentation, are adequate to ensure that reactor pressure remains below the analytical limit assumed in the vessel overpressure protection analysis and that the Reactor Pressure - High Function channels remain Operable. Therefore, ISTS 3.4.11 is not required to be included in the JAFNPP ITS.
DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None DIFFERENCE BASED ON PENDING TRAVELER (TP)
None Page 1 of 2 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 NUREG: 3.4.11 - REACTOR STEAM DOME PRESSURE DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
None Page 2 of 2 JAFNPP Revision A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG: N3.4.11 Reactor Steam Dome Pressure MARKUP OF NUREG-1433, REVISION 1, BASES
F
/
~Reactor Steaml Dome Presqsure B 3.4 REACTOR COOLANT SYSTEM4 (RC B 3.4.11 Reactor Steam Dome essure BASES BACGRUN hrafetyorelief amv dom rsurigte imiting prssuredination trsoan asimedt Tauihe dtermination of cp liance wt h
crveria.~e tri sdpnen nteIita eco
- LIAL, h eatrsteam dome pessure tfrfoe th 11imit on tis prssr ensurANALYS ES i t
titheossum of the overpressure protection analysi ron Reference alys assumes an initial mxmmreactor steam dom pressurefrte andealyssaftesig btsi safdents/ele and transientsuegoeemn the limitigpe suriftor trcaddieng. inetegrmi ty (eBase ofcorpiac LO 3..2 tNhe CvrprTsCAL POERi RAiO depRenden on claddingtiplasticstorai (se Bae o CO321 AVERAGE PLANAR LINEAR HEAT
/
e~~~~ENERATION RATE (APLHGR)').
Rsteam dome pressuresifies the requtresntr Critrio 2 o-th-NRCPolcy Sateent ICO~~
Teseiidratorsemdmhrse limit Poteto suin ns sponse mre severe than APPICAILTY"**ensures*,*,
analy (en tial (se B s fr 3.Rev 1, 04/07/95 BWR/4 STS
APPLICABILITY
- MODES, ae reactor my be generating significant tem and (continued) the d gn basis accidents and transients are unding.
In DES 3, 4, and 5, the limit is not app cable because tt reactor is shutdown.
In these MODE the reactor ssure is well below the required Ii
, ana no anticipated events will challenge the verpressure limits.
&I With the reactor steam dome ressure greater than the limit, prompt action should be t en to reduce pressure to below the limit and return th actor to operation within the bounds of the analyses The 15 minute Completion Time is reasonable consideri the importance of maintaining the pressure within li
- s.
This Completion Time also ensures that the probabil y of an accident occurring while pressure is greater than e limit is minimized.
If the operator is unable to rest the reactor steam dome pressure to below the limit, t n the reactor should be placed in MODE 3 to be operating hin the assumptions of the transient analyses.
If reactor stem dome pressure cannot be restored to b in the limit within the associated Completion,Time, the
- ~~
~
~
-+.L~
II iant must be brougnt to a IMJME in wnii Lne Lthe uu* mes apply.
To achieve this status, the plant must be broug to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The allowed Coplet n Tim of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power cond ons in an orderly manner and without challenging plant s ems.
ii EILELA MCE SR 3 4 1.
- UIREMEKTS Verification that reactor steam pressure is is
[1020] psig ensures that th nitial conditions of the design basis accidents and ansients are met. Operating experience has shown the hour Frequency to be sufficient for identifying trends d verifying operation within safety analyses assumptions.
PMM/4 STS B 3.4-56 Rev 1, 04/07/95 7ACTIONS
/
i,
BWR/4 STS B 3.4-57 Rev 1, 04/07/95
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION NUREG:
N3.4.11 Reactor Steam Dome Pressure JUSTIFICATION FOR DIFFERENCES (JFDs)
FROM NUREG-1433, REVISION 1, BASES
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433. REVISION 1 NUREG BASES:
3.4.11 - REACTOR STEAM DOME PRESSURE RETENTION OF EXISTING REQUIREMENT (CLB)
None PLANT SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)
None PLANT SPECIFIC DIFFERENCE IN DESIGN OR DESIGN BASIS (DB)
DB1 NUREG-1433 Specification 3.4.11, "Reactor Steam Dome Pressure," is not incorporated in the ITS.
This NUREG Specification is required to help ensure the vessel overpressure protection analysis can be met by ensuring the initial conditions of the event are preserved.
The JAFNPP site specific overpressure protection analysis is analyzed with an initial condition equivalent to the analytical limit (1094 psig) associated with the Reactor Protection System Instrumentation Reactor Pressure - High Function in ITS 3.3.1.1. Reactor Protection System Instrumentation.
A CHANNEL CHECK of the associated instrumentation is required to be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during operations in MODES 1 and 2 which is consistent with the Surveillance Frequency in ISTS SR 3.4.11.1.
The CHANNEL CHECK Surveillance will ensure reactor pressure is below the Allowable Value (1080 psig) which is lower than the analytical limit associated with this Function as well as below the initial condition of reactor pressure assumed in the overpressure protection analysis.
In fact, since reactor pressure is normally at or below 1040 psig during reactor operations, action will be taken at a much lower pressure to restore reactor pressure or to restore inoperable channels than what is required by ISTS 3.4.11.
The requirements of ITS 3.3.1.1, Reactor Protection System Instrumentation, are adequate to ensure that reactor pressure remains below the analytical limit assumed in the vessel overpressure protection analysis and that the Reactor Pressure - High Function channels remain Operable. Therefore, ISTS 3.4.11 is not required to be included in the JAFNPP ITS.
DIFFERENCE BASED ON APPROVED TRAVELER (TA)
None DIFFERENCE BASED ON PENDING TRAVELER (TP)
None Page 1 of 2 Revision A JAFNPP
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 NUREG BASES:
3.4.11 - REACTOR STEAM DOME PRESSURE DIFFERENCE FOR OTHER REASONS THAN ABOVE (X)
None Page 2 of 2 JAFNPP Revi si on A
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION 314.6.F Structural Integrity THIS SPECIFICATION IS Relocated.
MARKUP OF CURRENT TECHNICAL SPECIFICATIONS (CTS)
DISCUSSION OF CHANGES (DOCs) TO THE CTS NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
FOR LESS RESTICTIVE CHANGES CTS:
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION 314.6.F Structural Integrity MARKUP OF CURRENT SPECIFICATIONS TECHNICAL (CTS)
CTS:
1f Cu rreo/
ý2ek~f JAFNPP 3.6 (cont'd)
F.
Structural Inteaity The structural integrin f the Reactor Coolant S maintained at the le required by the original a standards throug the Ife of the Plant.
Whenever the reactor is in the startup/hot standi modes, all jet pumps shall be operable. If it Is de a jet pump is inoperable, the reactor shall be plac condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
/
ystem shall be 1
No structive inspections shall be performed on.he cceptance AS E Boiler and Pressure Vessel Code Class 1, ' and 3 c
ponents and supports in accordance withkae quirements of the weld and support inser ie inspection program. This inservice inspection progr is based on an NRC approved edition of, and adde to. Section Xl of the ASME Boiler and Pressure Vess Code which is in effect 12 months or less prior to the ginning of thl inspection interval.
- 2.
An augmented inservice inspec on program is required for those high stressed circu erential piping joints in the main steam and feedwaterf *es larger than 4 inches in diameter, where no restra t against pipe whip is provided. The augment in-service inspection progra shall consist of 100 perent inspection of these wel per inspection interv
- 3.
An Inservice Insl ction Program for piping ide lified in the NRC Goner' Letter 88-0 1 shall be implyne n
accordance wi h NRC staff positin ons
- dules, methods, pe/onnel, n apeepansio included in this Generi Letter, or in accordance wi lternate measues,.
poved byteNCsaf by or run Whenever there is recirculation flow with the reactor in the itermined that startup/hot.standby or run modes, jet pump operability shall be d in a coldocuheckedsmydaily by veriying that the following conditions do not Amendment No. P4. 1.A0.
203 144 I
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION 314.6.F Structural Integrity DISCUSSION OF CHANGES (DOCs) TO THE CTS CTS:
DISCUSSION OF CHANGES CTS: 3/4.6.F - STRUCTURAL INTEGRITY ADMINISTRATIVE CHANGES None CWAWMF MAPF PF~qT1PTTTVF I
r..n,1 I-J..
L1
, I I"
J I
\\-
I.
,I.'-,
0V;lD r
None TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
None TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
None TECHNICAL CHANGES - RELOCATIONS R1 The structural integrity inspections are provided to prevent long term component degradation and provide long term maintenance of acceptable structural conditions of the system.
The associated inspections are not required to ensure immediate OPERABILITY of the system.
Therefore, the requirements specified in CTS 3.4.F did not satisfy the NRC Policy Statement Technical Specification screening criteria as documented in the Application of Selection Criteria to the JAFNPP Technical Specifications and have been relocated to the Technical Requirements Manual controlled in accordance with 10 CFR 50.59.
Page 1 of 1 TEPhMTPAI Revision A JAFNPP
JAFNPP IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS) CONVERSION CTS:
314.6.F Structural Integrity NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)
FOR LESS RESTICTIVE CHANGES
NO SIGNIFICANT HAZARDS CONSIDERATION CTS: 3/4.6.F - STRUCTURAL INTEGRITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
There are no plant specific less restrictive changes identified for this Speci fi cati on.
Page 1 of 1 Revi si on A JAFNPP
MODIFIED RAI RESPONSES FOR ITS SECTION 3.4
Revision E Changes to Section 3.4 RAI Responses RAIs Generic Terminology for jet pump loop flow, jet pump flow, recirculation loop, recirculation pump loops, and recirculation drive flow have been changed, interchanged, etc.
Only change the nomenclature that is plant specific.
Other changes are generic and have to be changed through the established change process, e.g., the TSTF.
This refers to all PA changes.
JAFNPP Response:
- 1.
The Authority agrees that some of the changes made were not necessary to ensure understanding of the terms used while some of the changes are necessary to provide consistent terminology.
- 2.
A summary of the changes needed in ITS 3.4.1 and 3.4.2 to achieve consistent use of terms is provided below:
- a.
in ITS 3.4.1, "Insert ACTIONS A and B" - change "Jet pump loop flow..." to "Recirculation loop jet pump flow..." in Condition B to make it consistent with ITS SR 3.4.1.2.a discussed in b below,
[Revised Response provided with Revision E Package - delete part 2.a since RAI 3.4.1-01 deletes Condition B]
- b.
in ITS SR 3.4.1.2.a - retain "...recirculation loop jet pump flow..." as stated in the NUREG,
- c.
in ITS SR 3.4.1.2 Bases (first paragraph, last sentence)
- retain "I...recirculation loop jet pump loop flow..." as stated in the
- d.
in ITS SR 3.4.2.1.a - retain "Recirculation pump flow..." as stated in the NUREG,
- e.
in ITS SR 3.4.2.1.a - change "...jet pump loop flow..." to
"..-recirculation loop jet pump flow..." to make it consistent with NUREG SR 3.4.1.1 (ITS SR 3.4.1.2),
- f.
in NUREG (and ITS) SR 3.4.2.1 Bases (first paragraph, 20th line) change "...jet pump loop flow.." to "...recirculation loop jet pump flow..." to make consistent with NUREG SR 3.4.1.1 and NUREG SR 3.4.1.1 Bases,
- g.
in NUREG (and ITS) SR 3.4.2.1 Bases (first paragraph, 20th line) change "...recirculation loop flow..." to "...recirculation pump flow..." to make consistent with NUREG (and ITS) SR 3.4.2.1.a, Page 1
Revision E Changes to Section 3.4 RAI Responses
- h.
- change
"..(recirculation versus..." to make 3.4.1.2) and NUREG SR 3.4.2.1 Bases (second paragraph, first
"...(pump flow and loop flow versus..." to pump flow and recirculation loop jet pump flow consistent with NUREG SR 3.4.1.1 (ITS SR (and ITS) SR 3.4.2.1.a, and
- i.
in NUREG (and ITS) SR 3.4.2.1 Bases (second paragraph, third sentence)
- change "...pump flow and loop flow versus..." to
"...recirculation pump flow and recirculation loop jet pump flow versus..." to make consistent withNUREG SR 3.4.1.1 (ITS SR 3.4.1.2) and NUREG (and ITS) SR 3.4.2.1.a.
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Revision E Changes to Section 3.4 RAI Responses 3.4.9-04 CTS 3.6.A.5 DOC A4 CTS 3.6.A.5 indicates that with any of the limits 3.6.1 through 3.6.A.4 exceeded....
3.6.A.4 specifies "during all Modes of operation.
Would this not imply that 3.6.A.5 then should be the same. CTS 3.6.A does not specify Applicability.
DOC A2 concluded that because there was not a stated Applicability in CTS 3.6.A. it implies that CTS 3.6.A is applicable at all times.
DOC A2 logic conflicts with DOC A4.
DOC A4 concludes that because CTS 3.6.C does not include an Applicability statement then the Applicability can be determined from the actions required when the LCO cannot be met.
DOC A4 states "Since this Specification requires that, if the Required Actions and Completion Times are not met, the reactor be placed in Cold Shutdown (MODE 4).
it can be implied that the Specification is Applicable in MODES 1. 2 and 3."
A similar difference in logic exists between DOC L3 of ITS 3.4.6 and DOC A2 of ITS 3.4.9.
Comment: Provide discussion regarding the above apparent conflict in the discussions.
JAFNPP Response:
- 1.
The FitzPatrick ITS conversion has noted in a number of DOCs that Applicability of a particular CTS LCO is implied based on CTS Required Action that stipulates an "end state" that is presumed to place the plant in a Mode or specified condition that is outside the (unstated)
Applicability for the particular LCO.
This "logic" for determining the Applicability of CTS 3.6.A.5 was (in error) used in ITS 3.4.9, DOC A4 and is (as stated above by the NRC reviewer) in conflict with ITS 3.4.9, DOC A2.
- 2.
The Authority will revise ITS 3.4.9, DOC A2 and DOC A4 as well as ITS 3.4.6, DOC L3 as necessary to eliminate the conflicts.
[Revised Response provided with-Revision E Package]
The Licensee will revise ITS, DOC A2 and DOC A4.
However, there does not appear to be a conflict with ITS 3.4.6, DOC L3; thus it will not be revised.
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Revision E Changes to Section 3.4 RAI Responses 3.4.9-06 CTS 3.6.A.2..3,
.4 Figure 3.6-1 Bases CTS 3.6.A.2, 3.6.A.3 and 3.6.A.4 specify being to the right of CTS Figure 3.6 1 curves A, B, and C respectively, which makes clear the safe area for operation.
By implication the same applies (being to the right) of the curves on ITS Figure 3.4.9-1.
ITS 3.4.9 including ITS Figure 3.4.9-1. which is exactly the same as CTS figure 3.6-1 Part 3, and ITS 3.4.9 Bases do not specify anywhere that the safe area relative to curve A. B, or C is to the right.
ITS 3.4.9 simply requires maintaining pressure and temperature within limits.
Comment: State where clarification in ITS 3.4.9-1 curves A. B, in the LCOthe limits are found.
Additionally, provide 3.4.9 Bases where the safe area relative to ITS Figure and C is located.
JAFNPP Response:
- 1.
As noted in response RAI 3.4.9-01,NUREG-1433, Revision 1, does not state where pressure and temperature limits are found beyond making reference to an external report.
In contrast, FitzPatrick limits are incorporated directly into ITS 3.4.9, with specific limits identified or referenced as applicable.
Each specific limit is identified in its respective surveillance.
(See response RAI 3.4.9-01.)
- 2.
A note will be added to ITS Figure 3.4.9-1 specifying that safe operation is on or to the right of curve A, B. or C, as appropriate.
[Revised Response provided with Revision E The requirement that operation be on or to be relocated to the Bases,.consistent with submittals.
Package]
the right of the Figures will recently approved BWR ITS Page 4