ML20005E002
| ML20005E002 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/26/1989 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005E003 | List: |
| References | |
| NUDOCS 9001030094 | |
| Download: ML20005E002 (24) | |
Text
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',j UNITED STATES NUCLEAR REGULATORY COMMISSION-o r,,
l WASHINGTON, D. C. 20555 '
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 148 License No. DPR-59 1
-1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Power Authority of the State of New York (the licensee) dated May 31, 1989, amended by letter dated July 18. 1989 and amplified by letter dated November 20, 1989, a
complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regu-lations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities-authorized.
by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:
9001030094 891226 DR ADOCK 0500 3
.t
-(2) Technical Specifications The Technical Specifications contained in Appendices A and B,L as revised through Araendment No.148. are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
-t 3
This license amendment is effective as of the date of its issuance.
to be implemented within 30 days.
FOR TliE NUCLEAR REGULATORY COMMISSION b
"f 7
y Robert A.-Capra, Ofrector i
Project Directorate I-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation 1 -kttachment:
Changes to the Technical Spacific6tions 1
' Date of' Issuance: December 26, 1989
0,..
- \\
s ATTACHMENT TO LICENSE NiENDMENT NO.148 FACILITY OPERATING LICENSE fiO. OPR-59 DOCKET f!0.'50-333
- Revise Appendix A as.follows:.
k_emove Pages Insert Pages 89 89 106' 106 109 109 114 114 115 115 115a 115a 116.
116 118 118 i
120 120 4
121a 121o I
125 125 126 126 127 127 129-129 132 132
'145c-145c 156 156 183 183 218 218 239 239 241 241
7 T
JAFNPP 1
3.3 (cont'd) 4.3 (cont'd) l a.
Control rods which cannot be moved with control a.
Each partially or fully withdrawn operable control rod I
rod drive pressure shall be considered inoperabic. If shall be exercised one notch at least once each a partially or fully withdrawn control rod drive cannot week when operating above 30 percent power. In be moved with drive or scram pressure, the reactor the event power operation is contin'. sing with three or sha!! be brought to the Cold Shutdown condition more inoperable control rods, this test shali be l
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and shall not be started unless (1) performed at least once each day, when operating l
investigation has shown that the cause of the failure above 30 percent power.
is not a failed control rod drive mechanism collet b.
The scram discharge volume drain and vent valves l
housing, and (2) adequate shutdown margin has been demonstrated as required by Specification shall be verified open at least once per 31 days (these valves may be closed intermittently for testing under administrative control).
l l
If investigation shows that the cause of control rod A SM li d pata Mi vai4 t%
l failure is a cracked collet housing, or if this possibility cannot be ruled out, the reactor shall not conformance to Specification 3.3.A.2.d before a rod be started until the affected control rod dnve has may be bypassed in the Rod Sequence Control been replaced or repaired.
System.
d.
Once per week check status of pressure and level alarms for each accumulator.
l Amendment No. /,fg 148 89
=
,y b
JAFNPP 4.4 (cont'd) pump solution in the recirculation path.
Explode one of three primer assemblies' manufactured in same batch to verify proper function. Then install the two remaining primer assemblies of the same batch in the explosive valves.
Demineralized water shall be injected into the reactor vessel to test that valves (except explosive valves) not checked by the recirculation test are not clogged.
Test that the setting of the system pressure relief valves is between 1,400 and 1,490 psig.
3.
Disassemble and inspect one explosive valve so that it can -
be established that the valve is not clogged. Both valves shall be inspected in the course of two operating cycles.
B.
Operation with inoperable Components B.
Operation with Inoperable Components When a component becomes inoperabic its redundant From and after the date that a redundant component is made or component shall be verified to be operable immediately and g'
found to be inoperable, Specification 3.4.A shall be considered daily thereafter.
fulfilled, and continued operation permitted, provided that:
1.
The component is returned to an operable condition within 7 days.
Amendment No. 5, % 148 106
JAFNPP l
ATWS requirements are satisfied at all concentrations above 10 The relief valves in the Standby Uquid Control System protect l
weight percent for a minimum enrichment of 34.7 atom percent the system piping and positive displacement pumps, which are of B-10.
nominally designed for 1,500 psig, from overpressure. The Figure 3.4-1 shows the permissible region of operation on a pressure relief valves discharge back to the standby liquid i
sodium pentaborate solution volume versus concentration control pump suction line.
graph. This curve was developed for 34.7% enriched B-10 and B.
Operation with inoperable Components a pumping rate of 50 gpm. Each point on this curve provides a minimum of 660 ppm of equivalent natural boron in the reactor Only one of two standby liquid control pumping circuits is vessel upon injection of SLC solution. At a solution volume of needed for operation. If one circuit is inoperable, there is no immediate threat to shutdown capability, and reactor operation 2200 gallons, a weight concentration of 13% sodium pentaborate, enriched to 34.7% boron-10 is needed to meet may continue during repairs. Assurance that the remaining shutdown requirements. The maximum storage volume of the system will perform its function is obtained by verifying pump solution is 4780 gallons which is the net overflow volume.n the o erability in the operable circuit at least daily.
SLC tank.
C.
Sodium Pentaborate Solution Boron concentration, isotopic enrichment of boron-10, solution To guard against precipitation, the solution, including that in the temperature, and volume are checked on a frequency pump suction piping, is kept at least 10*F above saturation adequate to assure a high reliability of operation of the system temperature. Figure 3.4-2 shows the saturation temperature should it every be required. Experience with pump operability including 10*F margin as a function of sodium pentaborate indicates that monthly testing is adequate to detect if failures solution concentration. Tank heater and heat tracmg system have occurred.
are provided to assure compliance with this requirement. The sd pnts fa h WWc acWon d the M h W l
The only practical time to test the Standby Uquid Control System is during a refueling outage and by initiation from local at Wng system am estaNM W on h Wu Mn@nk U n-NPerate W W M darms is stations. Components of the system are checked periodically systs annundate m the wnW room. % p@y I
as described above and make a functional test of the entire is I
system on a frequency of more than once cach refueling cheded on a fmquency to asse a Ngh daNW of opeah syshn sM he hqh
)
outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the charges are satisfactory. A continuous check of the firing circuit continuity is provided by pilot lights in the control room.
1 Amendment No. ;MI, id 148 109
=
- a
l JAFNPP 3.5 (cont'd) 4.5 (cont'd) 2.
From and after the date that one of the Core Spray.
2.
When it is determined that one Core Spray System is Systems is made or found inoperable for any reason, inoperable, the operable Core Spray. System, and both l
continued reactor operation is permissible during the LPCI subsystems, shall be venfied to be operable' l l
succeeding 7 days unless the system is made operable immediately. The remaining Core Spray System shaR be L
earlier, provided that during the 7 days all active verified to be operable daily thereafter.
.l 1 -
components of the other Core Spray System and the LPCI System shall be operable.
3.
LPCI System testing shall be as specified in 4.5.A.1a, b, c, l
l 3.
Both LPCI subsystems of the RHR System shall be d, f and g except that each RHR pump shall deliver at least operable whenever irradiated fuel is in the reactor and prior 9,900 gpm against a system head corresponding to a to reactor startup from a cold condition, except as reactor vessel to primary containment differernial pressure specified below.
of greater than or equal to 20 psid.
I a.
From the time that one of the LPCI subsystems is a.
When it is determined that one LPCI subsystem is made or found to be inoperable for any reason, inoperable, the operable LPCI subsystem and both continued reactor operation is permissible during the Core Spray Systems shall be verified to be operable succeeding 7 days unless that subsystem is made immediately and daily thereafter.
operable earlier provided that during these 7 days the operable LPCI subsystem and both Core Spray Systems shall be operable.
i l
s Amendment No. [f6'l96'% 148 114
.x m.,,
m JAFNPP 3.5 (cont'd) 4.5 (cont'd)'
l b.
When the reactor water temperature is greater than.
b.
The power source disconnect and chasn lock to
[
2127, the rnotor operator for the RHR cross-tie valve motor operatr4 RHR cross-tie valve, and lock on (MOV20) shall be maintained discunsected from its manually ocerated gate valve shall be inspected electric power source. It shall be maintained once each opaaik g cycle to venfy that both valves chain-locked in the closed position. The manually are closed andlocked.
operated gate valve (10-RHR-09) in the cross-tie line, in series with the motor operated valve, shall be maintained locked in the closed position.
4.
a.
The reactor shall not be started up with the RHR System supplying cooling to the fuel pool.
b.
The RHR System shall not supply cooling to the spent fuel pool when the reactor coolant tempetature is above 212 F.
Amendment No.Ig 148 115
t o
l l
t JAFNPP 3.5 (cont'd) 4.5 (cont'd) 5.
All recirculation pump discharg6 valves shall be operable 5.
All recirculation pump discharge valves shall be tested for -
prior to reactor startup (or closed if permitted elsewhere in operability any time the reactor is in the cold condition these specifications).
exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been 6.
If the requirements of 3.5.A cannot be met, the reactor ng tWW 31 days.
shall be placed in the cold condition within 24 hrs.
B.
MWnM Mm hystem M (m RHR Systan)
B.
Containment Cooling Subsystem Mode (of the RHR System) 1.
Subsystems of the containment cooling mode are tested 1.
Both subsystems of the containment cooling mode, each including two RHR, one ESW pump and two RHRSW in conjunction with the test performed on the LPCI pumps shall be operable whenever there is irradiated fuel subsystems and given in 4.5.A.1.a. b, c, and d. Residual l
in the reactor vessel, prior to startup from a cold condition, heat removal service water pumps, each loop cOiwining of and reactor coolant temperature two pumps opwaung in paraW, d M M M testy, specified below.
->212"F except as
~
supplying 8,000 gpm. The Emergency Service Water System, each loop of which consists of a sogle operating emergency service water pump will 'be _ tested in accordance with Section 4.11D.
During each five-year period, an air test shall be pertaviried on the containment spray headers and nozzles.
2.
Continued reactor operation is permissible for 30 days with 2.
When it is determined that one RHR pump and/or one one spray loop inoperable and with reactor water RHRSW pump of the components required in 3.5.8.1 temperature greater than 212 F.
above are inoperable, the remaining redundant active components of the containment cooling mode subsystems shall be verified to be operable immediately and daily l
thereafter.
Amendment No.Mg % g 148 115a
,.. ~
JAFNPP 3.5 (cont'd) 4.5 (cont'd) 3.
Should one RHR pump and/or one RHRSW pump of the 3.
When one contamment coolog subsystem loop becomes components required in 3.5.8.1 above be made or found inoperable, the operable loop shall be verdied to be l
l' inoperable, continued reactor operation is permissible only operable immediately and daily thereafter.
during the succeeding 30 days provided that dunng such l
30 days. all remaining active. components of tte l
containment cooling mode are operable.
4.
Should one of the containment coohng _ subsystems become inoperable, continued reactor operation is permissible for a period not to exceed 7 days, urness such subsystem is sooner made operable provided that during such 7 days all active curgients of - the other containment cooling subsystem are operable.
5.
If the requirements of 3.5.B cannot be met, tfm reactor shall be placed in a cold condition within 24 hr.
6.
Low power physics testing and reactor operator training shall be permitted with reactor. coolant tempei ::ure
<212 F with an inoperable component (s) as specified in 3.5.8 above.
Amendment No. XS!f148 116 t
^
3 JAFNPP 3.5 (cont'd) 4.5 (cont'd) a.
From and after the date that the HPCI System is a.
When it is determined that the HPCI subsystem is made or found to be inoperable for any reason, inoperable the RCIC, the LPCI subsy&, both core -
continued reactor operation is permissible only spray subsystems, and the ' ADS stbsystem during the succeeding 7 days unless such system is actuation logic, shall be.venfied to be operable l.
sooner made operable, provided that during such 7 immediately. The RCIC system and ADS subsystem.
days all active components of the Automatic logic shall be venfied to be operable daily thereafter.
l Depressurization System, the Core Spray System, LPCI System, and Reactor Core isolation Cooling System are operable.
b.
If the requirements of 3.5.C.1 cannot be met, the reactor shall be placed in the cold condition and pressure less than 150 psig within 24 hrs.
2.
Low power physics testing and reactor operator training shall be permitted with reactor coolant temperature
<212 F with an inoperable component (s) as specified in 3.5.C.1 above.
i Amendment No M MM 148 118-
+-
JAFNPP 3.5 (cont'd) 4.5 (cont'd) during such time, the HPCI System is operable.
2.
If the requirements of 3.5.D.1 cannot be met, the reactor 2.
Alogic system functional test.
shall be placed in the cold condition and pressure less than 100 psig, within 24 hr.
a.
When it is dets6nwas that one valve of the ADS is inoperable, the ADS subsystem actuabon logic for the operable ADS valves and the HPCI subsystem shall be verified to be operable immediately and at l
least weekly thereafter.
b.
When it is determined that more than one relief / safety valve of the ADS is inoperable, the HPCI '
System shall be verified to be operable immediately.
l 3.
Low power physics testing and reactor operator training shall be permitted with inoperable components as specified in 3.5.1.a and 3.5.1.b above, provided that reactor coolant temperature is <212 F and the reactor vessel is vented or reactor vessel head is removed.
Amendment No. gt 48 120
r
.e JAFNPP' l
l 3.5 (cont'd) 4.5 (cont'd).
The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor pressure of 1,120 psig to 150 psig.
. 2.
When it is determined that the RCIC System is 'moperable at a time when it is required to be operable, the HPCI
~
l System shall be verified to be operable immediately and l-daily thereafter.
i l~
l l
t I
(
l Amendment No.g 148 121a
. ~
=
=
-a
=
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- r O :-
JAFNPP 3.5 BASES a,.
Core Spray System and Low Pressure Coolant.. injection (LPCI)
The LPCI mode of the RHR System is designed to provide Mode of the RHR System emergency cooling to the core by flooding in the event of a loss-of-coolant accident. These subsystems are completely This specificatiori assures that adequate emergency cooling independent of the Core Spray System; however, they function capability is available whenever irradiated fuel is in the reactor in comtynation with the Core Spray System to prevent vessel.
excessive fuel clad temperature. The LPCI mode of The loss-of-coolant analysis is referenced and described in General Electric Topical Report NEDE-24011-P-A.
The limiting conditions of operation in Specifications 3.5.A.1 through 35.A.6 specify the combinations of operab!c subsystems to assure the availability of the minimum cooling systems. No single failure of ECCS equipment occurring during a loss-of-coolant accident under these limiting conditions of operation will result in inadequate cooling of the reactor core.
Core spray distribution has been shown, in full scale tests of systems similar in design to that of the FitzPatrick_ Plant, to exceed the minimum requirements by at least 25 percent. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the internal pressure has fallen to 113 psig.
Amendment No.ff 148 125
JAFNPP 3.5 BASES (cont'd) the RHR System in conjunction with the Core Spray System Should one Core Spray System become inoperable, the provides adequate cooiing for break areas of approximately 02 remaining Core Spray and the entire LPCI System are avalable sq. ft. up to and including the double-ended reactor recirculation should the need for core cooling arise. To assure that the line break without assistance from the high pressure Emergency remaining Core Spray and LPCI Systems are available, they are verified operable immediately. This verification mciudes the Core Cooling Systems.
pumps W Mated v&es. W on j@ d h The allowable repair times are established so that the average rdM f r
dng systans, i.e, tM %e 5ay W risk rate for repair would be no greater than the basic risk rate.
, a seven-day repair pan was Wa,M W, M The method and concept are described in Reference 8. Using one shystan h inoph, h rm the results developed in this reference, the repair period is found s@systan W tM Ne 5ay Systan are avne to pe to be less than 1/2 the test interval. This assumes that the Core cooling.
l Spray and LPCI Systems constitute 1-out-of-2 systems; however, the combined effect of the two systems to limit excessive clad temperatures must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 30 days, and this specification is within this period. For multiple failures, a shorter interval is specified and to improve the assurance that the remaining systems will function, a daily test is called for.
Although it is recognized that the information given in Reference 8 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgement.
Amendment No.p2( 148 126
JAFNPP 3.5 BASES (cont'd) containment cooieng subsystem beco Ties inoperable only one system remams, a seven day repair penod was spacJ,ed.
l Low power physics testing and reactor operator trameng with B.
Containment Cooling Subsystem Mode (of the RHR System) inoperable cuir onenB will be conducted only when the v
The containment heat removal portion of the LPCl/ containment "9
the
~
spray mode is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the Calculations have been made to determme the effects of the flow specified, the containment long-term pressure is limited to design basis LOCA whole conductog low power physics testeg less than 8 psig and, therefore, is more than ample to provide or operator training at or below 212"F. The results of these the required heat removal capability.
conservative calculations show that the suppresseon pool water The containment cooling mode (of the RHR System) consists of two sets of two RHR Pumps, two RHR service water pumps, ystans d not M WW MM h h postN one ESW Pump, and one heat exchanger.. Either set of equipment is capable of performing the containment cooling function. Loss of one RHR service water pump does not seriously jeopardize the containment cooling capability as any two of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left, a thirty-day repair period is adequate. Loss of one subsystem of the containment cooling mode leaves one remaining system to perform the containment cooling function. The operable
{
system is verified to be operable each day wilen the above condition occurs.
Based on the fact that when one Amendment No.148 127
O JAFAPP I
l 3.5 BASES (cont'd) vessel head off the LPCI and Core Spray Systems will perform F.
Minimum Emergency Core and Containment Cooling System their designed safety function without the help of ADS.
Availability E.
Reactor Core isolation Cooling (RCIC) System The purpose of Specification 45.D is to assure a minimum d The RCIC is designed to provide makeup to the Reactor segmcy me Wng W is ave at d times. N, Coolant System as a planned operation for periods when the fa exa@, one me pay wwe M d sawce W h normal heat sink is unavailable. The RCIC also serves as awgmcy M M powwM the We me pay wwe redundant makeup system on total loss of all offsite power in wt y sawce, @ two RHR W M M avm.
the event that HPCI is unavailable. In all other postulated Ukewise, if two RHR pumps were out of service and two RHR on te @ wme also M d serwce, no conth accidents and transients, the ADS provides redundancy for the avaHh M,s dunng rW Mages M HPCI. Based on this and judgements on the reliability of the ng i
HPCI system, an allowable repair time of 7 days is specified.
m ja mance is @M W W M h M d l
Immediate and daily verifications of HPCI operability during I w pswe me Mng systems may be M d sawce. M RCIC outage is considered adequate based on judgement and Wcation pmwh that M 2 ocm nc M W W.
practicality' perf rmed on the Reactor Coolant System wfwch could lead to draining the vessel. This work would include work on certain Low power physics testing and reactor operator training with control rod drive components and Reactor Recirculation inoperable components will be conducted only when the RCIC System. Thus, the specification precludes the events wtuch System is not required, (reactor coolant temperature <212 F could require core cooling. Specification 3.9 must also be and coolant pressure <150 psig). If the plant parameters are consulted to determine other below the point where the RCIC System is required, physics testing and operator training will not place the plant in an unsafe condition.
Operability of the RCIC System is required only when reactor pressure is greater than 150 psig and reactor coolant temperature is greater than 212 F because core spray and low pressure coolant injection can protect the core for any size pipe break at low pressure.
t l
Amendment No.)4k 148 129
JAFNPP 4.5 BASES The testing interval for the Core and Containment Cooling With components or subsystems out-of-service, overall core Systems is based on a quantitative reliability analysis, industry and containment coohng reliabikty is mamtamed by ventying practice, judgement, and practicality. The Emergency Core the operability of the remainirw] cooling equipment Consistent Cooling Systems have not been designed to be fully testable with the definition of operable in Section 4.0.C, demonstrate means conduct a test - to show; venfy means that the during operation. For example, the core spray final admission valves do not open until reactor pressure has fallen to 450 psig-associated surveillance activities have been satisfactorily.
thus, during operation even if high drywell pressure were performed within the speciTW time intervd.
simulated, the final valves would not open. In the case of the h weilh WMs to me M N hp HPCI, automatic initiation during power operation would result i ing of the core spray, LPCI mode of the RHR, HPCI, and in pumping cold water into the reactor vessel which is not RCIC Systems are filled provides for a' visual observation that
&straNo.
water flows from a high point vent. This ensures that The systems will be automatically actuated during a refueling outage. In the case of the Core Spray System, condensate storage tank water will be pumped to the vessel to verify the operability of the core spray header. To increase the availability of the individual components of the Core and Containment Cooling Systems the components which make up the system i.e., instrumentatien, pumps, valve operators, etc., are tested more frequently. The instrumentation is functionally tested each month. Ukewise, the pumps and motor-operated valves are also tested each month to assure their operability. The combination automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems.
Amendment No. [148 132
o w
G JAFNPP.
3.6 (cont'd) 4.6 (cont'd) which are required to be operable in these modes, The snubbers may be categorized irWo two groups: Those completeoneof thefollowing:
accessible and those inaccessible during reactor operation.
Each group may be inspected independently in accordance with.
a.
replace or restore the inoperable snubber (s) to the above schedule.
operable status or, indications of damage or.UI I
b.
declare the supported system inoperable and follow impaired OPERABILITY, (2) the appropriate limiting condition for operation a
s to N fee a M he m statement for that system or, secure, and (3) in those locations where snubber I
c.
perform an engineering evaluation to show the movements can be manually induced without inoperable snubber is unnecessary to assure disconnecting the snubber, that the snubber has freedom operability of the system or to meet the design of movement and is not frozen up. Snubbers which criteria of the system, and remove the snubber from appear inoperable as a result of visual inspections may be the system.
determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause 3.
With one or more snubbers found inoperable, within 72 of the rejection is clearly established and remedied for that hours perform a visual inspection of the supported particular snubber and for other snubbers that may be component (s) associated with the inoperable snubber (s) generically susceptible; and (2) the affected snubber is and document the results. For alt modes of operation functionally tested in the as found condition and except Cold Shutdcwn and Refueling, within 14 days determined OPERABLE per Specifications 4.6.1.7 or 4.6.I.8, complete an engineering evaluation as per Specification as applicable. Hydraulic snubbers which have lost 4.6.l.6 to ensure that the inoperable snubber (s) has not sufficient fluid to potentially cause uncovering of the fluid adversely affected the supported component (s). For Cold reservoir-to-snubber valve assembly port or bottoming of Shutdown or Refueling mode, this evaluation shall be the fluid reservoir piston with the snubber completed within 30 days.
Amendment No.MJCW 148 145c
,i ii.i.. i...
JAFNPP 3.6 and 4.6 BASES (cont'd) 72 % a vid We M be #mned on b H'
(DELETED) supported ceivwient(s) associated with the inoperable snubber (s) and the results shall be documented. For all modes 1.
Shock S@ pressors d opsh e4 W Wh and % within M Snubbers are designed to prevent unrestrained pipe motion days an engineenng whion shaN be M to answe under dynamic loads as meght occur during an earthquake or that the inoperable snubber (s) hss not adversely anected the severe transeent, while allowing normal thermal mobon dunng suppated wnweWs). 6 Cold Shuh a %
startup and shutdown. The consequence of an inoperable m de, M wh M be wuA
- 2 days. A snubber is an increase in the probability of structural damage penod d 7 days has been WW rp a W of to piping as a result of a seismic or other event initiating
& snopwaW W dunng W N a cW dynamec loads. It is therefore required that all strJrt;ers mode d opwah becuase in these modes the rh required to protect the primary coolant systens or any other pr babelity of Wuctwal dari. age to the piping @.w would safety system or componert be gerable during reactor be laws due to W values of total Wesses on the piping operation. Snubbers excluded from this inspection program systems. In case a Wh is r@ed, N h d 36 are those installed on nor' safety related syCm and then only Ws to re a cold shutdown Me M W an adely if their failure or failure of the system r=
wtuch they are shutdown consistent with standard operating procedures.
installed would have no adverse effect on any safety-related system. Because the snubber protection is required only during low probability events, a period d 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (for normal operation) or 7 days (for cold shutdown or refueling mode of operation) is allowed for repairs or replacar, eat of the snubber prior to taking any other action. Following the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (or 7 day) period, the supported system must be declared inoperable and the Umiting Condition of Operation statement for the supported system followed. As an altemative to snubber repair or replacerrent an engineering evaluation may l
be performed: to show that the inoperable snubber is unnecessary to assure operability of the system or to meet the design criteria of the system; and, to remove the snubber from the system. With one or more snubbers found snoperable, Amendment No.M[gi48
^
156
JAFNPP 3.7 (cont'd) 4.7 (cont'd) e.
At least once per operating cycle, manual operabihty l
of the bypass valve for felter coohng shall be j
A. 0 6,aled.
l f.
Standby Gas Treatment System Instrumentation l
Cahbration:
differential Once/ operating j
pressure cycie switches 2.
From and after the date that one circuit of the standby Gas Treatment System is made or found to be inoperable for 2.
When one circuit of the Standby Gas Treatment System j
any reason, the following would apply-becomes inoperable, the operable circuit shall be venlied l
to be operable immediately and dai*f thereafter.
a.
If in Start-up/ Hot Standby, Run or Hot Shutdown l
mode, reactor operation or irradiated fuel handling is l
permissible only during the succeeding 7 days l
unless such circuit is sooner made operable, i
provided that during such 7 days all active components of the other Standby Gas Treatment Circuit shall be operable.
b.
If in Refuel or Cold Shutdown mode, reactor operation or irradiated fuel handling is permissible only during the succeeding 31 days unless such circuit is sooner made operable, provided that l
during such 31 days all active components of the other Standby Gas Treatment Circuit shall be operable.
3.
If Specifications 3.7.B.1 and 3.7.82 are not met, the l.
reactor shall be piaa,cd in the cold condition and irradiated fuel handling operations and operations that could reduce the shutdown margin shall be prohibited.
Amendment No. gJiidf 148 183
a 4.
JAFNPP 3S Continued 4S Continued 6.
Once within one hour and at least once per cight fours thereafter, while the reactor is being operated in accordance with Specifications 3S.B.1, 3S.B.3 and 3S.B.4, the availability of off-site power shall be assured by verifying correct breaker alignment and by verifying that the associated off-site electrical line is energized.
C.
Diese! Fuel C.
Diesel Fuel There will be a minimum of 64,000 gal. of diesel fuel on site for Once a month the quantity of diesel fuel available in each each operable pair of diesel generators.
storage tank shall be manually measured and compared to the reading of the local level indicators to ensure the proper operation thereof.
1.
From and after the time that fuel oil storage tank Icvel 1.
Once a month a sample of the diesel fuel in each storage instrumentation is made or found to be inoperable for any tank shall be checked for quality as per the following-reason continued reactor operation is permissible Flash Point - T 1257 min.
indefinitely, provided that the level in the affected storage P
Point -T 1W m tank is manually measured at least once/ day.
Water & Sedinent 0.50% max.
Ash 0.5% max.
Distillation 90% Point 540 min.
Viscosity (SSU) at 1007 40 max.
Sulfur 1% max.
Copper Strip Corrosxm No. 3 max.
i Cetane #
35 min.
l l
l l
l Amendment No. [g 148 218
o
.~
JAFNPP 3.11 (cont'd) 4.11 (cont'd)
B.
Crescent AreaVentilation B.
Crescent AreaVentilation Crescent area ventilation and cooling equipment shall be 1.
Unit coolers serving ECCS components shall be operable on a continuous basis whenever specification 3.5.A' 3.5.B, and 3.5.C are required to be satisfied.
demonstrated operable once/3 months.
l 2.
Temperature indicator controllers shal1 be calibrated 1.
From and after the date that more than one unit cooler once/ operating cycle.
serving ECCS compartments in the same half of the crescent area are made or found to be inoperable, all ECCS components in that half of the crescent area shall be considered to be inoperable for purposes of specification 3.5.A,3.5.8, and 3.5.C.
2.
If 3.11.B.1 cannot be met, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
Battery Room Ventilation C.
Battery Room Ventilation Battery room ventilation equipment shall be demonstrated l
Battery room ventilation shall be operable on a continuous basis operabic once/ week.
whenever specification 3.9 E is required to be satisfied.
1.
When it is determined that one battery room ventilation 1.
From and after the date that one of the battery room system is inoperable, the remaining ventilation system ventilation systems is made or found to be inoperable, its shall be verified operable and daily thereafter.
l associated battery shall be considered to be inoperable for 2.
Temperature transmitters and differential pressure purposes of specification 3.9.E.
Switches shall be calibrated once/ operating cycle.
Amendment No. 38'JM(%)d 148 239
o JAFNPP 3.11 (cont'd) 4.11 (cont'd) e.
ESW Once/ day instrumentation-Once/3 months check calibrate test Once/3 months f.
Logc System Once/each FunctionalTest operating cycle 2.
ESW will not be supplied to RBCLC system during testeg.
2.
From and after the time that one Emergency Service Water System is made or found to be inoperable for any reason continued reactor operation is permissible for a penod not to exceed 7 days total for any calendar month, provided that:
the operable Emergency Diesel Generator System is demonstrated to be operable immediately and daily thereafter; and, all Emergency Diesel Generator System emergency loads are verified operable immediately and daily thereafter.
3.
If specification 3.11.G.2 cannot be met an orderly shut down shall be initiated and the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amendment No. 148 241
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