ML19282B844
| ML19282B844 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/09/1979 |
| From: | Schmieder J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19282B841 | List: |
| References | |
| NUDOCS 7903160250 | |
| Download: ML19282B844 (121) | |
Text
. .
BEFORE THE UNITED STATES (g) NUCLEAR REGULATORY COMMISSION In the Matter of )
)
POWER AUTHORITY OF THE STATE OF NEW YORK ) Docket No. 50-333
)
(James A. FitzPatrick Nuclear Power Plant) )
APPLICATION FOR AMENDMENT TO OPERATING LICENSE The Power Authority of the State of New York, Licensee in the above-captioned docket, hereby files an Application for Amendment to Operating License DPR-59, which would make certain changes to the Technical Specifications as set forth in Appendix A.
With this Application for Amendment, Licensee hereby transmits documents entitled " Proposed Changes to Technical Specifications" (Attachment A) and " James A. FitzPatrick Nuclear Power Plant Pump and Valve Inservice Testing Program", Revision 1, January 6, 1979 (Attachment B). The proposed changes would provide for revisions to Sections 1.0, 4.5, 4.7 and 4.11 of the Specifications. The pro-posed changes would delete explicit requirements for operability testing of certain pumps and valves from the above Specification Sections 4.5, 4.7 and 4.11, and instead apply the Pump and Valve Operability Test Program for FitzPatrick (Attachment B) to the Specifications.
790316 0750
The proposed changes would not authorize any change in the types or any increase in the amounts of effluents or any increase in the authorized power level of the facility.
Wile RE FOl<E , Applicant respectfully requests that Appendix A to Facility Operating License No. DPR-59 be amended in the form attached hereto as Attachment A.
POWER AUTIIORITY OF Tile STATE OF NEW YORK j!l l O,,(?
lj / / , !,
U By / (l'If6 4')1bt.16 t Jo, soph R. Schmieder Clief Engineer J
Subscribed and sworn to '
before me this I/ day of March, 1979.
/
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NotaryPubl/c'g/
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ATTACHMENT A Power Authority of the State of New York License No. DPR-59 Docket No. 50-333 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS
JAFNPP Q. Refueling Outage - Refueling outage 1.0 (cont'd) is the period of time between the opened to perform necessary shutdown of the unit prior to a operational activities. refueling and the startup of the Plant subsequent to that refueling.
- 2. At least one door in each air-lock is closed and sealed. R. Safety Lbmits - The safety Ibnits are limits within which the
- 3. All automatic containment iso- reasonable maintenance of the fuel lation valves are operable or cladding integrity and the reactor de-activated in the isolated coolant system integrity are position. assured. Violation of such a limit is cause for unit shutdown and
- 4. All blind flanges and manways review by the Nuclear Regulatory are closed. Commission before resumption of unit operation. Operation beyond such a N. Rated Power - Rated power refers to limit may not in itself result in operation at a reactor power of serious consequences but it indi-2,436 MWt. This is also termed 100 cates an operational deficiency sub-percent power and is the maximum ject to regulatory review.
power level authorized by the operating license. Rated steam S. Secondary Containment Integrity -
flow, rated coolant flow, rated Secondary containment integrity nuclear system pressure, refer to means that the reactor building is the values of these parameters when intact and the following conditions the reactor is at rated power. are met:
O. Reactor Power Operation - Reactor 1. A!: least one door in each power operation is any operation access opening is closed, with the Mode Switch in the Startup/ Hot Standby or Run position 2. Ihe Standby Gas Treatment with the reactor critical and above System is operable.
I percent rated thermal power.
- 3. All automatic Jentilation P. Reactor Vessel Pressure - Unless system isolation valves are otherwise indicated, reactor vessel operable or secured in the pressures listed in the Technical isolated position.
Specifications are those measured by the reactor vessel steam space sensor.
Amendment No. 5 ,
JAFNPP 1.0 (Cont'd) and applicable Addenda shall be applied as follows in these T. Surveillance Frequency - Periodic sur- Technical Specifications:
aillance tests, checks, calibrations, and examinations shall be performed ASME Boiler and Required frequencies within the specified surveillance Pressure Vessel for performing inservice intervals. These intervals may be Code and appli- inspection and testing adjusted + 25 percent. The interval as cable Addenda activities.
pertaining to instrument and electrical terminology for surveillance shall never exceed one op- inservice inspec-erating cycle. In cases where the elapsed tion and testing interval has exceeded 100 percent of the activities.
specified interval, the next surveillance interval shall commence at the end of the original specified interval. Weekly At least once per 7 days U. Surveillance Requirements - Inservice Monthly At least once per inspection and testing of ASME Code Class 31 days 1, 2 and 3 pumps and valves shall be applicable Quarterly or At least once per as follows: every 3 months 92 days Semiannually or At least once per
- 1. Inservice inspection of ASME Code every 6 months 184 days Class 1, 2 and 3 pumps and valves Yearly or annually At least once per shall be performed in accordance 366 days with Section XI of the ASME Boiler and Pressure Vessel Code and Applicable 3. The provisions of Specification 1.0.T, Addenda as required by 10 CFR 50, the definition of Surveillance Fre-Section 50.55a(g), except where speci- quency, are applicable to the above fic written relief has been granted by required frequencies for performing the Commission pursuant to 10 CFR 50, inservice inspection and testing Section 50.55a(g) (6) (1). activities.
- 2. Surveillance intervals specified in 4. Performance of the above inservice Section XI of the ASME Boiler and inspection and testing activities Pressure Vessel Code and applicable shall be in addition to other speci-Addenda for the inservice inspection fled Surveillance Requirementa.
and testing activities required by the ASME Boiler and Pressure Vessel Code 5. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
Amendment No. 6
JAFNPP 1.0 (cont'd) W. Electrically Disarmed Control Rod V. Thermal Parameters To disarm a rod drive electrically, the
- 1. four amphenol type plug connectors Minimum critical power ratio (MCPR) - are removed from the drive insert and Ratio of that power in a fuel assembly which is calculated to withdrawal solenoids rendering the rod incapabic of withdrawal. This pro-cause some point in that fuel assembly to experience boiling cedure is equivalent to valving out the drive and is preferred. Elec-transition to the actual assembly trical disarming does not eliminate operating power as calculated by position indication.
application of the GEXL correlation (Reference NEDE-10958) . X. High Pressure Water Fire Protection System
- 2. Fraction of Limiting Power Density -
The ratio of the linear heat genera- The High Pressure Water Fire Protec-tion rate (LHGR) existing at a given tion System consists of: a water location to the design LHGR for that source and pumps; and distribution bundle type. Design LHGR's are 18.5 system piping with associated post KW/ft for 7x7 bundles and 13.4 KW/ft indicator valves (isolation valves).
for 8x8 and 8x8R buadles. Such valves include the yard hydrant
- 3. Maximum Fraction of Limiting Power curb valves and the first valve Density - The Maximum Fraction of ahead of the water flow alann de-Limiting Power Density (MFLPD) is vice on each sprinkler or water spray subsystem.
the highest value existing in the core of the Fraction of Limiting Y. Staggered Test Basis Power Density (FLPD).
A Staggered Test Basis shall consist
- 4. Transition Boiling - Transition of:
boiling means the boiling region between nucleate and film boiling, a. A test schedule for n systems, Transition boiling is the region in subsystems, trains or other desig-which both nucleate and film boil- nated components obtained by ing occur intermittently with dividing the specified test in-neither type being completely stable. terval into n equal subintervals.
- b. The testing of one system, subsystem, train or other designated compon-ent at the beginning of each sub-int e rval .
Amendment No. pd,}d,gd 6a
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
- b. Flow Rate Test - Once/3 months Core spray pumps shall deliver at least 4,625 gpm against a system head corresponding to a total pump developed head of 2 113 psig
- c. Logic System Once/each opera-Functional Test ting cycle
- d. Core Spray IIeader lip Instrumentation Check Once/ day Calibrate once/3 months Test once/3 months Amendment No. jb 113
JAFNPP 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 Core Spray System is inoperable, is made or found inoperable for the operable Core Spray System, any reason, continued reacter the LPCI System, and the emergency operation is permissible during diesel generators shall be demon-the succeeding 7 days unless strated to be operable immediately.
the system is made operable The remaining Core Spray System earlier, provided that during shall be demonstrated to be operable the 7 days all active compo- daily thereafter.
nents of the other Core Spray System and the LPCI System 3. LPCI System testing shall be as and the emergency diesel l specified in 4.5.A.l.a, b and c, generators shall be operable, except that three RHR pumps shall deliver at least 23,100 gpm against
- 3. The LPCI mode of the RHR System a system head corresponding to a shall be operable whenever ir- reactor vessel pressure of 20 psig.
radiated fuel is in the reactor and prior to reactor startup a. When it is determined that one from a cold condition, except of the RHR pumps is inoperable, as specified below, the remaining active components of the LPCI, containment spray
- a. From the time that one of subsystem, both Core Spray the RHR pumps is made or Systems, and the emergency found to be inoperable for diesel generators required for any reason, continued operation shall be demonstrated reactor operation is to be operable immediately and permissible during the the remaining RHR pumps shall succeeding 7 days unless be demonstrated to be operable the pump is made operable daily thereafter.
earlier provided that during such 7 days the remaining active com-ponents of the LPCI, containment spray mode, all active components of both Core Spray Systems, and the emergency diesel generators are operable.
Amendment No. d, 4d j 114
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
- 5. All recirculation pump discharge 5. All recirculation pump discharge valves and bypass valves shall be and bypass valves shall be tested operable prior to reactor startup for operability any t0me the reactor (or closed if permitted elsewhere is in the cold condition exceeding in these specifications). 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 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 hr.
Containment Cooling Subsystem Mode (of B. Containment Cooling Subsystem Mode (of B.
the RHR System) the RilR System)
- 1. Both subsystems of the containment 1. Subsystems of the containment cooling cooling mode, each including two RRR, mode are tested in conjunction with one ESW pump and two RHRSW pumps shall the test performed on the LPCI System be operable whenever there is irradiated and given in 4.5.A.1.a and b.
fuel in the reactor Residual heat removal Amendment No. 115a
}d
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
- 4. Should one of the ccntainment cooling sub-systems become inoperable, continued reactor operation is permissible for a period not to exceed 7 days, unless such subsysten is sooner made operable provided that during such 7 days all active components of the other containment cooling subsystem, including its associated diesel generator, are operable.
- 5. If the requirements of 3.5.B cannot be met, the reactor shall be placed in a cold condition within 24 hrs.
- 6. Low power physics testing and reactor operator training shall be penmitted with reactor cool-ant temperature 6 212 F with an inoperable com-ponent(s) as specified in 3.5.B above.
C. High Pressure Coolant Injection (HPCI) System C. High Pressure Coolant Injection (HPCI) System
- 1. Ihe HPCI System shall be operable whenever the Surveillance of HPCI System shall be performed reactor pressure is greater than 150 psig and as follows provided a reactor steam supply is irradiated fuel is in the reactor vessel and available. If steam is not available at the prior to reactor startup from a cold condition, time the surveillance test is scheduled to be except as specified below: performed, the test shall be performed within ten days of continuous operation irem the time steam becomes available.
- 1. HPCI System testing shall be as specified in 4.5.A.l.a. b and c except that the l HPCI pump shall deliver at least 4,250 gpm against a system head corresponding to a reactor vessel pressure of 1,120 psig to 150 psig.
Amendment No. gb 117
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
E. Reactor Core Isolation Cooling E. Reactor Core Isolation Cooling (RCIC) System (RCIC) System
- 1. The RCIC System shall be operable 1. RCIC System testing shall be performed whenever there is irradiated fuel as follows provided a reactor steam in the reactor vessel and the reactor supply is available. If steam is not pressure is greater than 150 psig and available at the time the surveillance prior to a reactor startup from a test is scheduled to be performed, the cold condition, except from the time test shall be performed within ten days that the RCIC System is made or found of continuous operation from the time to be inoperable for any reason, con- steam becomes available.
tinued reactor power operation is per-missible during the succeeding 7 days Item Frequency unless the system is made operable earlier provided that during these 7 a. Simulated Once/ operating days the HPCI Systen is operable. Automatic cycle Actuation
- 2. If the requirements of 3.5.E cannot Test be met, the reactor shall be placed in the cold condition and pressure less b. Flow Rate Once/3 months than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The RCIC pump shall deliver at least
- 3. Low power physics testing and reactor 400 gpm for a system head correspond-operator training shall be permitted ing to a reactor pressure of 1,120 with inoperable components as speci- psig to 150 psig, fied in 3.5.E.2 above, provided that reactor coolant temperature is 5 212 F.
Amendment No. 4d j 121
JAFNFP 4.5 BASES Test Program for FitzPatrick. The combi-Thm testing interval for the Core and nation of automatic actuation test and'. pump and valve operability tests is deemed to con-Containment Cooling Systems is based on a quantitative reliability analysis, stitute adequate testing of these systems, industry p'ra ctice , judgement, and With components or subsystems out-of-practicality. The Emergency Core service, overall core and containment Cooling Systems have not been designed cooling reliability is maintained by to be fully testable during operation. demonstrating the operability of the For example, the core spray final remaining cooling equipment. The degree admission valves do not open until of operability to be demonstrated reactor pressure has fallen to 450 psig; depends on the nature of the reason for thus, during operation even if high the out-of-service equipment. For drywell pressure were simulated, the routine o.1t-of-service periods caused by final valves would not open. In the preventative maintenance, etc. , the pump case of the HPCI, automatic initiation and valve operability checks will be during power operation would result in performed to demonstrate operability of pumping cold water into the reactor the remaining components. Ilowever, if a vessel which is not desirable. failure, design deficiency, etc., caused the out-of-service period, then the The systems will be automatically demonstration of operability should be actuated during a refueling outage. In thorough enough to assure that a similar the case of the core Spray System, problem does not exist on the remaining condensate storage tank water will be components. For example, if an pumped to the vessel to verify the out-of-service period were caused by operability of the core spray header. failure of a pump to deliver rated To increase the availability of the capacity due to a design deficiency, the individual camponents of the Core and other pumps of this type might be sub-Containment Cooling Systems. the jected to a flow rate test in addition components which make up the system to the operability checks.
1.e., instrumentation, pumps, valve operators , etc., are tested more The surveillance requirements to ensure frequently. The instrumentation is that the discharge piping of the core functionally tested each month. spray, LPCI mode of the RHR, HPCI, and i Pumps and motor-operated valves are tested RCIC Systems are filled provides for a to assure their operability in accordance visual observation that water flows from with the ASME Section XI Pump and Valve a high point vent. This ensures that Amendment No.
/. 5 BASES (cont'd) JAFNFP the line is in a full condition.
Between the monthly intervals at which the lines are vented, instrumentation has been provided in the Core Spray System and LPCI System to monitor the presence of water in the discharge piping. This instrumentation will be calibrated on the same frequency as the safety system instrumentation. This period of periodic testing ensures that during the interval between the monthly checks the status of the discharge piping is monitored on a continuous basis.
133
t JAFNFP 3.7 (cont'd) 4.7 (cont'd)
- c. Secondary containment capability to maintain a 1/4 in. of water vacuum under calm wind conditions with a filter train flow rate of not more than 6,000 cfm, shall be deuonstrated at each refueling outage prior to re fue ling.
D. Primary Containment Isolation Valves D. Primary Containment Isolation Valves
- 1. During reactor power operating 1. The prLmary containment conditions, all isolation valves isolation valves surveillance listed in Table 3.7-1 and all shall be perfonned as follows:
instrument line flow check valves shall be of ereble, except a. At least once per operating as specified in 3.7.D.2. cycle, the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.
- b. At least once per operating cycle, the instrument line excess flow check valves shall be tested for proper operation.
Amendment No. 185
JAFNPP 3.7 (cont'd) 4.7 (cont'd)
- 2. In the event any isolation valve 2. Whenever an isolation valve specified in Table 3.7-1 listed in Table 3.7-1 is becomes inoperable, reactor inoperable, the position of at power operation may continue, least one other valve in each provided at least one valve in line having an inoperable valve each line having an inoperable shall be recorded daily, valve is in the mode corresponding to the isolated condition.
- 3. If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold ccndition within 24 hr.
Amendment No. 186
4.7 BASES (cont'd) JAFNPP by in-place testing with DOP as In order to assure that the doses testing medium. that may result from a steam line break do not exceed the 10CFR100 The test interval for filter guidelines, it is necessary that no efficiency was selected to minimize fuel rod perforation resulting from plugging of the filters. In the accident occur prior to closure addition, retention capacity in of the main steam line isolation terms of milligrams of iodine per valves. Analyses indicate that fuel gram of charcoal will be rod cladding perforations would be demonstrated. This will be done by avoided for main steam valve closure testing the charcoal once a year, times, including instrument delay, unless filter efficiency seriously as long as 10.5 sec.
deteriorates. Since shelf lives greater than 5 yr. have been For Reactor Coolant System demonstrated, the test interval is temperatures less than 2120F, the reasonable. containment could not become pressurized due to a loss-of-coolant D. Primary Obntainment Isolation valves accident. The 2120F lLmit is based on preventing pressurization of the The large pipes cortprising a portion reactor building and rupture of the of the Reactor Coolant System, whose blowout panels.
failure could result in uncovering the reactor core, are supplied with The primary containment isolation automatic isolation valves (except valves are highly reliable, have low those lines needed for Emergency service requirement, and most are Core Cooling Systems operation or normally closed. The initiating containment cooling) . The closure sensors and associated trip channels times specified herein are adequate are also checked to demonstrate the to prevent loss of more coolant from capability for automatic isolation.
the circumferential rupture of any The test interval of once per of these lines ,
- aide the operating cycle for automatic containment then frr- .eam line initiation results in a failure rupture. Theref're, isolation probability of 1.1 X 10 7 that a valve closure time ' icient to line will not isolate.
prevent uncovering _ s 196 Amendment No.
JAFNPP 4.7 BASES (cont'd)
The main steam line isolation valves and all other normally open power-operated isolation valves are func-tionally tested in accordance with the ASME Section XI Pump and Valve Test Program for FitzPatrick.
The primary containment is penetrated by several small diameter instrument lines connected to the reactor cool-ant system. Each instrument line contains a 0.25 in. restricting orifice inside the primary con-tainment and an excess flow check valve outside the primary contain-ment.
Amendment No. 197
JAFNPP 3.11 (Cont'd) 4.11 (cont'd)
D. Emergency Service Water System D. Emergency Service Water System 1 To ensure the adequate '
- l. Surveillance of the ESW system equipment and area cooling, shall be performed as follows:
both Esw systems shall Le operable when the requirements Item Frequency of specification 3.5.A and 3.5.B must be satisfied, a. Simulated Each except as specified below in Automatic operating specification 3.11.D.2. Actuation cycle Test
- b. Flow Rate Once/
Test - ESW 3 months pumps shall deliver at least 3,700 gpm against a system head corresponding to a total pump head of 2 80 psi, as determined from the pump certifi-cation curve by measuring the pump shutoff hee.d whi.ch shall be 2120 psi.
- c. Instrumenta- Once/
tion test 3 months 240
JAFNPP 3.11 (Cont'd) 4.11 (Cont'd) .
- d. Instrumentation once/each Calibration operating cycle
- e. Logic System Once/esch Functional operating Test cycle
- 2. From and after the time that 2.
one Emergency Service Water ESW will not be supplied to System is made or found to be RBCLC system during testing.
inoperable for any reason continued reactor operation is permi,ssible for a period not to exceed 7 days total for any calendar month, provided that:
the operable Emergency Diesel Generator System and all its emergency loads be demonstrated to be operable immediately and daily thereafter.
- 3. If specification 3.11.D.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 />.
241
JAFNPP 3.11 5 4.11 BASES conducted in accordance with manuf acturers' reconrnendations.
A. Main Control Room Ventilation System One snain control room emergency ven-B. _ Crescent Area Ventilation tilation air supply fan provides Engineering analyses indicate that adequate ventilation flow under the temperature rise in safeguards accident conditions. Should one compartments without adequate emergency ventilation air supply fan ventilation flow or cooling is such and/or fresh air filter train be out that continued operation of the of service during reactor operation, safeguards equipment or associated the a,llowable repair time of 1 month auxiliary equipment cannot be is justified, based on the 3 month assured. .
test interval.
C. Battery Room Ventilation The 3 month test interval for the main control room emergency Engineering analyses indicate that ventilation air supply fan and the temperature rise and hydrogen dampers is sufficient since two buildup in the battery, and battery redundant trains are provided and charger compartments without neither is normally in operation, adequate ventilation in such that continuous operation of equipnent in A pressure drop test across each these compartments cannot be filter and across the filter system assured.
is a measure of filter system condition. DOP injection measures D. Emergency Service Water System particulate removal efficiency of the high efficiency particulate The ESWS has two 100 percent cooling filters. A Freon-112 test of the capacity pumps, each powered from a charcoal filters is essentially a separate standby power supply. The leakage test. Since the filters ESWS utilizes lake water to the have charcoal of known efficiency cooling system of the emergency and holding capacity for elemental diesel generators. The system will iodine and/or methyl iodine, the also supply water to those test also gives an indication of the components of the RBCLCS which are relative efficiency of the installed required for emergency conditions system. Laboratory analysis of a during a loss of power condition.
sample of the charcoal filters These include ECCS pumps and area positively demonstrates halogen unit coolers. Pump and valve functional removal efficiency. These tests are testing is performed in accordance with the ASME Section XI Pump and Valve Operability Test Program for FitzPatrick.
Amendment No. 243