ML20116G918
| ML20116G918 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/26/1985 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20116G904 | List: |
| References | |
| NUDOCS 8505020017 | |
| Download: ML20116G918 (12) | |
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ATTACHMENT I TO JPN-85-36 Proposed Chances to the Technical Specifications Regardino Reduction of Emeroency Diesel Generator Tests (PTS-84-23)
New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333
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8505020017 850426 DR ADOCK 05000 33 L
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- se JAFNPP Q
3.5 (cont'd) 4.5 (cont'd) 2.
From and after the date that 2.
When it is determined that one, one of the Core Spray Systems is made or found inoperable for Core Spray System is inoperable, -
any reason, continued reactor the operable Core Spray System, and the LPCI System. shall be demon-operation is permissible during the succeeding 7 days unless strated to be operable immediately.
g the system is made operable The remaining Core Spray System shall earlier, provided that during be demonstrated to be operable daily the 7 days all active components thereafter, of the other Core Spray System and the LPCI System shall be operable.
3.
The LPCI mode of the RHR System 3.
LPCI System testing shall be as shall be operable whenever ir-radiated fuel is in the reactor specified in 4.5.A.la, b,
c,d, and prior to reactor startup f and g except that three RHR pumps from a cold condition, except shall deliver at least 23,100 gpm as specified below, against a system head corresponding to a reactor vessel pressure of 20 psig.
a.
From the time that one a.
When it is determined that of the RHR pumps is made or found to be inoperable one of the RHR pumps is for any reason, continued inoperable, the remaining reactor operation is per-active components of the missible during the LPCI, containment spray succeeding 7 days unless subsystem and both Core Spray the pump is made operable Systemsrequiredforoperationl shall be demonstrated to be earlier provided that dur-ing such 7 days the remain-operable immediately, and the ing active components of remaining RHR pumps shall be the LPCI, containment demonstrated to be operable spray mode, and all activh daily thereafter.
components of both Core Spray Systems are operable.
Amendment No. )(, A6 k
114
-s,
JAFNPP 3.5 (Cont'd) 4.5' (Cont'd) b.
From the time that the LPCI mode is made or
.b.
When it is determined that the LPCI mode is found to be inoperable for any reason, inoperable, both Core Spray Systems, and the continued reactor operation is permissible containment spray subsystem shall be demonstrated to during the succeeding 7 days unless the LPCI be operable immediately and daily thereafter, mode is made operable earlier provided that during these 7 days all active components of both Core Spray Systems and the containment l
spray subsystem (including two RHR pumps) shall be operable.
c.
When the reactor water temperature is greater c.
The power source disconnect and chain lock to motor than 211*F, the motor operator for the RHR operated RHR cross-tie valve, and lock on manually cross-tie valve (MOV20) shall be maintained operated gate valve shall be inspected once each disconnected from its electric power source.
. operating cycle to. verify that both valves are It shall be maintained chain-locked in the closed and locked, closed position. The manually 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 temperature is above 212*F.
Amendment No.
26 115
m-
+
i JAFhPP si W
^
3 /5 (Cont'd) 4.5 (Cont'd)#
?
s.
k 5.
All recirculation pump discharge valves and 5.
All recirculation pur:p discharge and bypass valves bypass valves shall be operable prior to shall be tested for operability any time the reactor reactor startup (or closed if permitted ic.in the cold condition exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if elsewhere in these specifications).
operability tests have not been performed during the preceding 31 days.
\\
6.
If the requirements of 3.5.A cannot be met, the reactor shall be placed in the cold condition within 24 hrs.
B.
CONTAINMENT C0OLING SUBSYSTEM MODE (OF B.
CONTAINMENT COOLING SUBSYSTEM MODE (OF THE RHR SYSTEM)
THE RHR SIETEM)
~
J 1.
Both subsystems of the containment cooling 1.
Subsystcas of;the containment cooling mode are mode, each includint two RHR, one ESW pump and
LPCI System and~given in 4.5.A.1.a. b, c, and ~d.
there is irradiated fuel in the reactor vessel, Residual heat removal service water pumps, each loop prior to startup from a cold condition,.and' consisting of two pumps operating in parallel, will reactor coolant temperature > 212*F except as specified below:
he included in testing, supplying 8,000 gpm.
The Emergency Service Water System, each loop of which consists of a single operating emergency, cervic,e water pump of 3,700 gpm will be tested in accordance with Section 4.11D.
During~eacnfive-yearperiod,an$1htest shall be performed on the containment spray headers and nozzles.
2.
Continued reactor operation is permissible for 2.
When it is determined that one RHR pump and/or one 30 days with one spray loop inoperable ~and with RHRSW pump of the components required in 3.5.B.1 reactor water temperature greater than 212*F.
above are inoperable, the remaining redundant active components of the containment cooling mode subsystems shall he demonstrated to be operable immediately and daily.thereafter.
Amendtent No.,36' 115a
JAFNPP 3.5 (Cont'd).
4.5 (Cont'd) 3.
Should'one RHR pump and/or one RHRSW pump of 3.
When one containment cooling subsystem loop becomes the components required in 3.5.B.1 above be inoperable, the operable loop shall be demonstrated made or found inoperable,' continued reactor
.to be operable immediately and daily thereafter.
operation is permissible only during the succeeding 30 days provided that during such 30 days all remaining active components of the centainment cooling mode are operable.
4 Should one of the containment cooling subsystems become inoperable, continued reactor operation is Permissible for a period not.t o exceed 7 days, unless such subsystem is sooner made operable proviced that during such 7 days all active components of the other containment cooling subsystem are operable.
'5.
If the requirements of 3.5.B cannot be met, the 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 temperature < 212*F with an inoperable component (s) as specified in 3.5.B above.
e e Amendment No. )I 116
JJiFNPP 3.5 (cont'd) 4.5 (cont'd)
DELETED I
C.
HIGH PRESSURE COOLANT INJECTION C.
HIGH PRESSURE COOLANT INJECTION (HPCI SYSTEM)
(HPCI SYSTEM)
-Surveillance of HPCI System shall be performed as follows provided reactor steam supply is available.
If steam is not available at the time the surveillance test is scheduled to be performed, the test shall be performed within 10 days of continuous operation from the time steam becomes available.
1.
The HPCI System shall be operable 1.
HPCI System testing shall be as specified whenever the reactor pressure is in 4.5.A.l.a, b, c, d, f, and g except greater than 150 psig and irradiated that the HPCI pump shall deliver at least fuel is in the reactor vessel and 4,250 gpm against a system head correspon-prior to reactor startup from a ding to a reactor vessel pressure of 1120 cold condition, except as specified psig to 150 psig..
below:
Amendment No. g 117
- ~3 m nm ~-~.- --
O JAFNPP 3.5 BASES (cont'd) the RHR System in conjunction with given in Reference 8 provides a the Core Spray System provides quantitative method to estimate adequate cooling for break areas of allowable repair times, the lack of approximately 0.2 sq. ft. up to and operating data to support the including the double-ended reactor analytical approach prevents recirculation line break without complete acceptance of this method assistance from the high pressure at this time.
Therefore, the times Emergency Core Cooling Systems.
stated in the specific items were established with due regard to 4
i The allowable repair times are judgement.
established so that the average risk i
rate for repair would be no greater Should one Core Spray System become than the basic risk rate.
The inoperable, the remaining Core Spray j
method and concept are described in and the entire LPCI System are i
Reference 8.
Using the'results available should the need for core developed in this reference, the cooling arise.
To assure that the i
repair period is found to be less remaining Core Spray and LPCI
}
than 1/2 the test interval.
This Systems are available, they are I
assumes that.the Core Spray and LPCI demonstrated to be operable Systems constitute a 1-out-of-2 immediately.
This demonstration i
systems; however, the combined effect includes a manual initiation of the of tne two systems to limit excessive pumps and associated valves.
clad temperatures must also be con-Based on judgements of the reli-j sidered.
The test interval specified ability of the remaining systems, in Specification 4.5 was 3 months.
i.e.,
the Core Spray and LPCI.
4 l
Therefore, an allowable repair period seven-day repair period was obtained.
a 1
which maintains the basic risk considering single ailures should be 1
less than 30 days, and this speci-Should the loss of one RHR pump occur, a nearly full complement of
]
fication is within this period.
core and containment cooling For multiple failures, a shorter equipment are available.
Three RHR interval is specified and to improve pumps in conjunction with the Core the assurance that the remaining Spray System will perform the core systems will function, a daily test is cooling function.
Because of the called for.
Although it is availability of the majority of the recognized that the information Amendment No.
126 i
JAFNPP 3.9 (cont'd) 3.
From and after the time that one 3.
The emergency diesel generator
.of the Emergency Diesel Generator
' system instrumentation shall be Systems is made or found to be inoperable, continued reactor checked during the' monthly generator test.
operation is permissible for a period not to exceed 7 days provided that the two incoming power sources are available and connected to the emergency bus associated with the A
inoperable Emergency Diesel Generator System and that the remaining Diesel Generator System is operable.
At the end of the 7-day period, 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 />, unless one or both diesel generator systems are made operable sooner.
4.
When both Emergency Diesel Generator 4.
Once each operating cycle, the Systems are made or found to be conditions under which the inoperable, a reactor shutdown shall be initiated within two hours Emergency Diesel Generator System is required will be simulated to and the reactor 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 /> after demonstrate that the pair of diesel initiation of shutdown.
generators will start, accelerate, force parallel, and accept the emergency loads in the prescribed sequence.
5.
Once within one hour and at least once per twenty-four hours there-l after while the reactor is being operated in accordance with Specifications 3.9.B.1, 3.9.B.2, and 3.9.B.3, the availability of the operable 1:mergency nienel Generators shall be desix> in t a a ted by manual nLarting and force paralleling where applicable.
Amendment tio. Jef
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7 ATTACHMENT II TO JPN-85-36 Safety Evaluation For Proposed Chances To The Technical Specifications Recarding Reduction Of Emercency Diesel Generator Tests (PTS-84-23)
New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 I
e
Section I Description of the Chances The proposed changes to the Technical Specifications are shown in Attachment I to the Application for Amendment.
This proposed Amendment revises Sections 3.5.A.2, 3.5.A.3(a), 4.5.A.2 and 4.5.A.3(a) (page 114), 3.5.A.3(b) and 4.5.B.3(b) (page 115), 3.5.B.4 and 4.5.B.3 (page 116), 4.9.B.5 (page 217) and the Bases on page 126.
The changes on these pages delete the requirement to demonstrate the operability of the emergency diesel generators when subsystems of the Emergency Core Cooling or Containment Cooling System are declared inoperable.
The proposed changes also remove the diesel generators from the Limiting Conditions for Operation (LCO) for these systems and reduces the surveillance test requirements.
The proposed Amendment also includes administrative changes on pages llSa, 116 and 117 of the Technical Specifications.
Section II -
Purpose of the Chances The proposed changes would delete the requirement that "the emergency diesel generators shall be operable", from the Limiting Conditions for Operation (LCO) which are applicable when the following systems are declared inoperable:
Core Spray (CS); Low Pressure Coolant Injection (LPCI) mode of Residual Heat Removal (RHR); and Containment Cooling.
This deletion has been proposed, because the LCO for the emergency diesel generators are'specified in Section 3.9.B of the Technical Specifications under Emergency A-C Power System.
This is also in accordance with the Standard Technical Specifications.
In addition, Section 3.0.E of the FitzPatrick Technical Specifications states that reactor operation is governed by the time limits of the Action Statement of the LCO for the emergency power source; and, not by the Action Statement of the individual system that is determined to be inoperable due to the inoperability of its emergency power source.
The proposed changes would also delete the surveillance test requirements for the emergency diesel generators when the above mentioned systems are declared inoperable. Two of the FitzPatrick diesels have a reliability factor of 1.0 and the other two have a reliability factor of 0.99.
These reliability factors have been determined in accordance with Regulatory Guide 1.108 ' Periodic Testing of Diesel Generator Units used as On-Site Electric Power Systems of Nuclear Power Plants'.
In addition, the NRC staff has concluded that excessive testing of diesel generators results in premature degradation of diesel engines (Reference 1).
These changes have been proposed because unnecessary testing would reduce their reliability.
In the current FitzPatrick Technical Specifications, the diesels are required to be tested every eight hours when reserve power is not available from either one or both off-site sources or when one of the diesel generators is declared inoperable.
The proposed amendment would also change this requirement for testing from eight hours to every twenty-four hours and thereby reduce the number of diesel surveillance tests.
11-1
)
Section III - Impact of the Chances The proposed changes would reduce the number of emergency diesel generator tests by approximately 42%.
This estimated reduction is based on the surveillance test data from 1981 to 1984.
The proposed changes do not change any system or subsystem and will not alter the conclusions of either the FSAR or SER accident analyses.
Based on the discussions in Section II, operation of the FitzPatrick plant in accordance with the proposed amendment would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated because:
a) the FitzPatrick diesel generators have a high degree of reliability; b) a reduction in the number of surveillance tests would avoid premature degradation of the diesel generators thereby maintaining their high level of reliability; c) the operation of safety-related equipment is not affected by the proposed changes; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, because the diesel generators are a part of the emergency A-C power system which is used as a back-up system and no new mode of failure is l
introduced; or (3) involve a significant reduction in margin of safety because a l
reduction of unnecessary surveillance tests would prevent l
premature degradation of the diesel generators thereby maintaining their high level of reliability.
Section IV Implementation of the Chances Implementation of the changes, as proposed, will not impact the ALARA or fire protection programs at FitzPatrick, nor will the changes impact the environment.
Section V Conclusion The incorporation of these changes:
a)
Will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety' Analysis Report; I
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b)
Will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; c)
Will not reduce the margin of safety as defined in the basis for any Technical Specifications; d)
Does not constitute an unreviewed safety question as defined in 10 CFR 50.59; and e)
Involves no significant hazards considerations, as defined in 10 CFR 50.92.
~
Section VI References 1)
NRC Generic Letter 84-15 " Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability".
2)
NYPA letter, C. A. McNeill, Jr. to D. G. Eisenhut, dated January 17, 1985 (JPN-85-04).
3)
James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev 2, July 1984, Section 8.6 4)
James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER) i 4
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f II-3 I
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