ML18038A776
| ML18038A776 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/12/1993 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17056C347 | List: |
| References | |
| NUDOCS 9304200342 | |
| Download: ML18038A776 (60) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NIAGARA MOHAWK POWER CORPORATION DDKKKT DD.
KD 225 NINE MILE POINT NUCLEAR STATION UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
140 License No. DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated November 20,
- 1990, as superseded February 7,
- 1992, as supplemented June 22,
- 1992, January 29,
- 1993, February 18,
- 1993, and March 29,
- 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations 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 (i) 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 5
E.
The issuanc'e of this amendment is in accordance with 10 CFR Part 51 of the Commission'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-63 is hereby amended to read as follows:
.- 9304200342 9304l2 PDR ADQCK 05000220 P
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 140, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance to be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
Apr>y y2, y993 Robert A. Capra, Director Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
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ATTACHMENT TO 'LICENSE AMENDMENT AMENDMENT NO.
1 TO FACILITY OPERATING LICENSE NO.
DPR-63 DOCKET NO. 50-220 Revise Appendix A as follows:
~ll P
Insert Pa es 118 118a 119 120 120b 135 136 137 138 139 140 141 142 143 144 146 147 148 149 118 118a 119 119a 119b 120 120b 135 136 137 138 139 140 140a 140b 140c 141 142 143 143a 144 146 147 148 148a 148b 149 (added page)
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UMITINOCONDITIONS FOR OPERATION Table 3.2.7 REACTOR COOLANT SYSTEM ISOLATIONVALVES Une or System Location Relative No. of Valves to Primary IEach Une)
Containment Normal Position Iw/lax)mum Initiating Signal IAII Oper. Time Action on Valves have Remote Motive Powers
{Sec)
Initiating Signal Manual'Backup)
Main Stsami I
ITwo Lines)
Inside Outside Open Open AC Motor 10 Pn/DC Solenoid 10 Close Close Reactor water level low-low or low reactor pressure, (with mode switch in run) or main steam line high radiation, or main steam line high flow,.or low-low-low condenser vacuum, or high temperature in the steam tunnel Feedwatsrl1
)
ITwo Lines)
Emer enc Coolin Steam Lesvin Reactori
)
ITwo Uncs)
Condensate Return to Resctorl
)
ITwo Lines)
Outside Outside Outside Outside Inside Outside Open Open Open Open Closed Closed AC Motor Self Act. Ck.
AC Motor DC Motor Self Act. Ck.
Pn/DC Solenoid 60 38 38 60 60 Close Close Close Open Remote Manual
(
High emergency cooling system flow High emergency cooling system fl Reactor water level low-lowor high reactor pressure AMENDMENTNO.P6, 9+, 1P6, 140 118
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UMITINQCONDITIONS FOR OPERATION Tsb'le 3.2.7 >>Continued)
REACTOR COOLANTSYSTEM ISOLATION VALVES Reactor Clssnu Une or System No. of Valves
>>Each Une)
Location Relative to Primary Normal Containment Position Maximum Initlatlng Signal >>AII Oper. Time Action on Valves Have Remote Motive Power~
>>Sec)
Initiating Signal Manual Backup) 1 Water Lesvin Reactor>>
)
>>One Line)
Water Return o Reactor>>
)
>>One Line)
Inside Outside Inside Outside Open Open Open Open AC Motor DC Motor AC Motor Self Act. Ck.
18 18 18 Close Close Close Reactor water level low-low or high area temperature or liquid poison initiation Shutdown Coolin Water Lesvin Reactor>>
)
>>One Line)
Water Return o Reactor>>
)
>>One Line)
Inside Outside Inside Outside Closed Closed Closed Closed AC Motor DC Motor AC Motor Self Act. Ck.
40 40 40 Close Close Close Reactor water level low-low, or high area temperature AMENDMENTNO. 1, 140 118a
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UMITINQCONDITIONS FOR OPERATION Table 3.2.7 (Continued)
REACTOR COOLANT SYSTEM ISOLATION VALVES Une or System No. of Valves (Each Une)
Location Re'lative To Primary Containment Normal Position Maximum Initiating Signal tAll Oper. Time Action on Valves Have Remote Motive Powers (Sec)
Initiating Signal Manual Backup)
~Li uid Poison (One Line)
Control Rod Drive H draulic(2)
(One Une)
Scram Discher e Volume{ )
~Sstem Vent'One Line)
Inside Outside Inside Outside Outside Closed Closed Open Open Open Self Act. Ck.
Self Act. Ck.
Self Act. Ck.
Self Act. Ck.
Pn/AC Solenoid 10 Close Automatic or manual reactor scram Scram Dischar e Volume
~Sstem Drain' (One Line)
~DareS ra Core S re In'ectio I
)
{Two Uncs)
Core S ra Hi h Point Ven
{Two Uncs)
Outside Inside Outside Inside Outside Open Closed Open Closed Closed Pn/AC Solenoid AC Motor AC Motor AC Motor Pn/DC Solenoid 10 22.6 22.6 27 27 Close Open Open Close Close Reactor water level low-low or high drywell pressure coincident with reactor vessel pressure less than 386 psig E
Reactor water level low-low or high drywell pressure Core S ra Condensate Su I { )
(Keep Fill)
(Two Lines)
Core S ra S stem Vslvesl
)
(Two Lines)
Core S ra Pum Discher e{")
(Two Test Lines to Suppression Chamber)
Outside Outside Outside Open Self Act. Ck.
Closed AC Motor Closed Self Act. Ck.
27 Close Reactor water level low-low or high drywell pressure AMENDMENTNO... 1, 140 119
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LIMITINGCONDITIONS FOR OPERATION Table 3.2.7 (Continued)
REACTOR COOLANTSYSTEM ISOLATION VALVES Une or System Location Relative No. of Valves To Primary (Each Une)
Containment Max(mum initiating Signal (All Normal Oper. Time Action on Valves Have Remote Position Motive Powers (Seo)
Initiating Signal Manual Backup)
Post Acciden Reactor Sam lin (One Line) sector Recirculation S stem Sem lin ( )
(One Line)
Outside Outside inside Outside Open Closed Closed Closed Self Act. Flow Fuse Pn/DC Soleno>d AC Motor DC Motor 30 20 20 Close Close Close Reactor water level low-low or main steam line high radiation or low-low-lowcondenser vacuum or reactor low pressure, (with mode switch in run) or high temperature in the steam tunnel or main steam line high flow AMENDMENTNC).
140 119a)
Notes:
'n - Pneumatically Operated
" Section 3.1.1e for LCO Requirements (1)
These valves do not have to be vented during the Type A test.
However, Type C leakage from these valves is added to the Type A test results, if not vented.
(2)
These valves have flow through them during and following an accident (a water seal) and receive a water leak rate test in accordance with the IST Program.
(3)
The inside core spray injection isolation valves are water sealed during and after an accident.
These valves are leak rate tested with water in accordance with the IST Program.
The outside core spray injection isolation valves are open with their breakers locked in the off position. Therefore, the outside core spray injection isolation valves do not have to be tested under the IST or Appendix J Leakage Program.
(4)
These valves are provided with a water seal.
Valves shall be tested during each refuel outage not to exceed two years consistent with Appendix J water seal testing requirements.
Leakage rates shall be limited to 0.5 gpm per nominal inch'of valve diameter up to a maximum of 5 gpm.
(5)
These valves are tested in accordance with Section 4.2.7.1a.
(6)
The self actuating flow fuse is tested in accordance with Section 4.3.4c.
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AMENDMENTNO. i40
<<Bb I
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BASES FOR 3.2.7 AND 4.2.7 REACTOR COOLANT SYSTEM ISOLATIONVALVES Double isolation valves are provided in lines which connect to the reactor coolant system to assure isolation and minimize reactor coolant loss in the event of a line rupture.
The specified valve requirements assure that isolation is already accomplished with one valve shut or provide redundancy in an open line with two operative valves.
Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation and the closure times presented in Table 3.2.7.
These closure times were selected to minimize coolant losses in the event of the specific line rupturing.
Using the longest closure time on the main-steam-line valves following a main-steam-line break (Section XV C.1.0) ", the core is still covered by the time the valves close.
Following a specific system line break, the cleanup and shutdown cooling closing times will upon initiation from a low-low level signal limitcoolant loss such that the core is not uncovered.
Feedwater flow would quickly restore coolant levels to prevent clad damage.
Closure times are discussed in Section VI-D.1.0( ).
I The valve operability test intervals are based on periods not likely to significantly affect operations, and are consistent with testing of other systems.
Results obtained during closure testing are not expected to differ appreciably from closure times under accident conditions as in most cases, flow helps to seal the valve.
The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 (Fifth Supplement;
- p. 115) that a line willnot isolate.
More frequent testing for valve operability results in a more reliable system.
(1) UFSAR (2) FSAR AMENDMENTNO.
140 120
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TABLE.2.7.1 PRIMARY COOLANT SYSTEM PRES URE ISOLATIONVALVE
~Ss em 1.
Core Spray System Valve No.
40-03 40-13 IIaximum(al Allowable Leaka e
c5.0 gpm a5.0 gpm 2.
Condensate Supply to Core Spray (Keep Fill System) 40-20 40-21 40-22 40-23
<5.0 gpm a5.0 gpm S5.0 gpm c5.0 gpm
~Fo oo e:
(a) 1.
2.
3.
Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.
Test differential pressure shall not be less than 150 psid.
The observed leakage at test differential pressure shall be adjusted to the functional maximum pressure differential.
AMENDMENTNO. 140 120b
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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT
.3.3 ~*
~)lica~bili y:
Applies to the allowable leakage rate of the primary containment system.
Applies to the primary containment system leakage rate.
~e~iv:
gt~e~iv:
To assure the capability of the containment in limiting radiation exposure to the public from exceeding values specified in 10CFR100 in the event of a loss-of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a metal-water reaction.
To verify that the leakage from the primary containment system is maintained within specified values.
To assure that periodic surveillances of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the-service life of the containment, and systems and components penetrating primary containment.
Whenever the reactor coolant system temperature is above 215F the primary containment leakage rate shall be within the limits of 4.3.3.b.
a.
In ra e Primar n
inm n L aka Ra I
(1)
Integrated leak rate tests shall be performed at the test pressure (Pt) of 22 psig.
I AMENDMENTNO, 140 135
l l I
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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREIVIENT Containment pressure shall not be permitted to decrease more than one (1) psi below Pt.
(2) Type B and C tests should be completed prior to each Type A test.
Type B and C leakages (penalties) not accounted for in the Type A test shall be incorporated as minimum pathway additions to the Upper Confidence Limit (UCL) to determine the overall as left integrated leakage rate.
(3)
If the leakage rate exceeds the acceptance criterion, corrective action shall be required.
If, during the performance of a Type A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory completion of the test, these leakage paths may be isolated and the Type A re-test continued until completion.
The Type A test shall be considered a failed test.
A local leakage test shall be performed at Pt before and after the repair of each isolated leakage path.
The sum of the post repaired local leakage rates and the UCL shall be less than 75 percent of the maximum allowable leakage rate, L, (22).
Local leakage rates shall not be subtracted from the Type A test results to determine the acceptability of a test.
The as found and as left leakage data values of excessive leakage areas beyond acceptance criteria shall be provided to the NRC.
AMENDMENTNO.
~ >40 136
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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT (4)
Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves.
(5) A Type A test shall last a minimum of eight (8) hours with leakage rates calculated based on "Total Time" method.
If a twenty-four (24) hour test is performed the "Mass Point" method will be used to calculate leakage rates.
A verification test shall be performed following each Type A test.
The verification test provides a method for assuring that systematic error or bias is given adequate consideration.
During the
- verification test, containment pressure may not decrease more than one (1) psi below Pt.
b.
A e
an ri ri A T tt) The maximum aliawable leakage rate Lt (22) shall not exceed 1.19 weight percent of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at-the test pressure of 22 psig (Pt).
I (2) The maximum allowable operational leakage,
)
rate Lto (22) which shall be met prior to power operation following a Type A test AMENDMENTNO.
140 137
t g
I r
LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT (either as measured or following repairs and retest) shall not exceed 0.75 L (22) (0.892 weight percent per day).
(3)
When adding the leakage rate measured during a Type C test to the results of a Type A test, the leakage rate shall be determined using minimum pathway analysis.
c.
~Fr unpen c (1)
Three Type A tests shall be conducted during each ten year service interval at approximately equal intervals.
The third test will be conducted when the plant is shutdown for the 10 year inservice inspections.
(2)
Retesting (a)
If a Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), a Corrective Action Plan that focuses attention on the cause of the problem shall be developed and implemented.
A Type A test that meets the requirements of AMENDMENTNO.; 140 138
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LIMITINGCONOITION FOR OPERATION SURVEILLANCEREQUIREMENT 4.3.3.a.(3) and 4.3.3.b.(2) is required prior to plant start-up.
A report of the Corrective Action following the failed Type A shall be submitted to the NRC for review and approval with the Containment Leak Test Report.
(b)
Ifany periodic Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), the test schedule for subsequent Type A tests will be reviewed and approved by the NRC.
(c)
If two consecutive periodic Type A tests (not including an immediate retest under (a)) fail to meet the acceptance criteria of 4.3.3.a.(3),
4.3.3.b.(1) or 4.3.3.b.(2), not-withstanding the periodic retest schedule of 4.3.3.c.(1), a Type A test must be performed at each refueling outage or every 18 months, which-ever occurs first, unless alternative leak test requirements are accepted by the NRC by means of specific exemption from Appendix J per 10CFR50.12.
This testing shall be performed until two AMENDMENTNO. 140 139
f
UMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT consecutive periodic Type A tests (not including an immediate retest under (a)) meet the acceptance criteria of 4.3.3.a. (3), 4.3.3.b. (1) and 4.3.3.b.(2), then the retest schedule specified in 4.3.3.c.(1) should be resumed.
d.
Local Leak Ra
-T B an T
Te (1)
Primary containment testable penetrations and isolation valves required to be Type B or Type C tested by regulatory requirements, shall be tested at a pressure of 35.0 psig (Pa) each major refueling outage, not to exceed two years, except as provided in (a) and (b) below.
(a)
Bolted double gasketed seals which shall be tested whenever the seal is closed after being opened and at least at each refueling outage not to exceed a two year interval.
(b)
Type B tests for primary containment penetrations employing a continuous leakage monitoring system shall be conducted at intervals not to exceed three years.
AMENDMENTNO. 140 140
E I
LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT (2)
When system pressure (Psys) on the opposite side of the isolation valve under test cannot be reduced to atmospheric pressure, then the test pressure shall not be less than Pa + Psys.
(3)
Personnel airlocks shall be leak tested in accordance with the following:
(a)
The airlocks shall be tested at a test pressure of 35 psig following a refueling outage or maintenance outage requiring drywall access prior to primary containment integrity being required.
(b)
Airlocks opened during periods when primary containment integrity is required shall be tested within three days after being opened.
For airlock doors opened more frequently than once every three days, the airlocks shall be tested at least once every three days.
(c)
The airlocks shall be tested every six months at a test pressure of 35 psig.
(d)
Leakage rate for airlocks shall not exceed 0.05La at 35 psig.
AMENDMENTNO.
140 140a I
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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT (4)
Primary containment penetrations and isolation valves that are not defined as Type B or Type C test components (e.g., seal welded cold instrument lines, CRD lines, drywell to wetwell connections, etc.) shall not be individually tested.
The penetrations will be considered as integral parts of the Type A test.
e.
A anc ri eri -T B
n T
e CTe The combined leakage rate for penetrations and valves subject to Type B and C tests determined by maximum pathway analysis shall be less than 0.60 La. If this value is exceeded, repairs and retests shall be performed to correct the condition.
f.
nin u
L akRa Mni r
(1)
When the primary containment is inerted, the containment shall be monitored for gross leakage by a weekly review of the inerting system makeup requirements.
(2) This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as these activities are completed.
AMENDMENTNO.
140
<40b l
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LIIVlITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT g.
~lns )~egin The accessible interior surfaces of the primary containment shall be visually inspected each operating cycle for evidence of deterioration.
'MENDMENT NO. i40 140c
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BASES FOR 3.3.3 AND 4.3.3 LEAKAGERATE The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident.
The peak drywell pressure would be 35 psig which would rapidly reduce to 22 psig within 100 seconds following the pipe break.
The total time the drywell pressure would be above 22 psig is calculated to be about 10 seconds.
Following the pipe break, the suppression chamber pressure rises to 22 psig within 10 seconds, equalizes with drywall pressure and thereafter rapidly decays with the drywell pressure decay.
I I The design pressures of the drywell and suppression chamber are 62 psig and 35 psig, respectively.
As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 10 seconds.
Based on the calculated primary containment pressure response discussed above and the suppression chamber design pressure; primary containment preoperational test pressures were chosen.
Also, based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than testing the individual components separately.
The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.9%/day at 35 psig. The analysis showed that with this leak rate and a standby gas treatment system filterefficiency of 90 percent for halogens, 95 percent for particulates, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about 6.0 rem and the maximum total thyroid dose is about 150 rem at the site boundary considering fumigation conditions over an exposure duration of two hours. The resultant doses that would occur for the duration of the accident at the low population distance of 4 miles are lower than those stated due to the variability of meteorological conditions that would be expected to occur over a 30-day period.
Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident.
These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs.
Therefore, the specified primary containment leak rate and filter efficiency (Specification 4.4.4) are conservative and provide margin between expected off-site doses and 10 CFR 100 guideline limits.
AMENDMENTNO. 140 141
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BASES FOR 3.3.3 AND 4.3.3 LEAKAGERATE The maximum allowable leakage rate (La) is 1.5%%d/day at a pressure of 35 psig (Pa).
This value for the test condition was derived from the
)
maximum allowable accident leak rate of about 1.9%/day when corrected for the effects of containment environment under accident and test conditions.
In the accident case, the containment atmosphere initiallywould be composed of steam and hot air depleted of oxygen whereas under test conditions the test medium would be air or nitrogen at ambient conditions.
Considering the differences in mixture composition and temperatures, the appropriate correction factor applied was 0.8 and determined from the guide on containment testing.
Although the dose calculations suggest that the allowable test leak rate could be allowed to increase to about 3.0%/day before the guideline thyroid dose limitgiven in 10 CFR 100 would be exceeded, establishing the limitat 1.5%%d/day provides an adequate margin of safety'to assure ~
the health and safety of the general public.
It is further considered that the allowable leak rate should not deviate significantly from the ~
containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate. The operational limitis derived by multiplying the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
A reduced pressure test program is used for the integrated test. The test pressures are based on loss-of-coolant accident conditions. The peak primary containment pressure following a loss-of-coolant accident would be 35 psig.
This would rapidly reduce to 22 psig.
The total time drywell pressure would be above 22 psig would be about 10 seconds.
Preoperational integrated leak tests were performed at test pressures at 35 psig and 22 psig.
Subsequent integrated tests are performed at a test pressure of 22 psig.
Closure of the containment isolation valves for the purpose of the test is accomplished by the means provided for normal operation ofthe valves.
The reactor is vented to the containment atmosphere during testing.
The acceptance criteria states that the maximum allowable leakage rate (L ) shall not exceed 1.19 weight percent of the contained air in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 22 psig (Pt). This corresponds to the maximum allowable leakage rate (La) of 1.5 weight percent at 35 psig (Pa).
The maximum allowable test leak rate L (at 22 psig) shall not exceed the 1.5%/day times the square root of the ratio of the pressures Pt (at 22 psig) and Pa (at 35 psig), respectively since the ratio of measured leakages for Nine Mile Point Unit 1 is 0.735. The allowable operational leakage rate, L o (at 22 psig) shall not exceed 75 percent of L (at 22 psig) and shall be met prior to resumption of power operation following a test.
AMENDMENTNO.
140 142
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BASES FOR 3.3.3 AND 4.3.3 LEAKAGERATE The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains withinthe specification.
The leak rate test frequency is based on 10 CFR 50 Appendix J.
I The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends.
Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.
It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other pa'rts of the facilitycould occur.
Such leakage paths that may affect significantly the consequences of accidents are to be minimized. Ifthe leakage rates of the double-gasketed seal penetrations, testable penetration isolation valves, containment air purge inlets and outlets and the vacuum relief valves are at the maximum specified, they willtotal 90 percent of the allowed leak rate.
Hence, 10 percent margin is left for leakage through walls and untested components.
Leakage from airtocks is measured under accident pressures in accordance with 10 CFR 50 Appendix J.
Monitoring the nitrogen make-up requirements of the inerting system provides a method of observing leak rate trends.
This instrumentation
)
equipment must be periodically removed from service for test and maintenance, but this out-of-service time willbe kept to a practical minimum.
The test program follows the guidelines stated in the Bechtel Topical Report.
This program provides adequate assurance that the test results realistically estimates the deIIree of containment leakage following a ioss of coolant accident.
The containment leakage rate is calculated using the Absolute Methodology.
Containment leakage results are presented in the test report as calculated using the Total Time and Mass Point techniques.
The results of local leak rate tests, including, "as-found" and "as-left" leakages, are also included in the containment leak test report.
AMENDMENTNO.
140 143
BASES FOR 3.3.3 AND 4.3.3 LEAKAGERATE The specific treatment of selective valve arrangements including the acceptability of the interpretations of 10 CFR 50 Appendix J requirements are given in References 5, 6, and 7. They serve as the bases for alternative test configurations (e.g., reverse accident, multi-valve, water leakage flow tests) as well as relaxations from previous leakage limits or constraints.
References:
(1)
FSAR, Volume II, Appendix E (2)
UFSAR, Section Vl B,2.1 (3)
TID-20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Determinations
~
'4)
BN-TOP-1 "Testing Criteria for integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants,".Revision 1, Bechtel Corporation, November 1, 1972 (5)
NRC Safety Evaluation Report dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."
(6)
Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications 5 Conformance with 10 CFR 50 Appendix J SER."
(7)
NRC Letter dated November 9, 1988, "Review of the July 28, 1988 Letter on Appendix J Containment Leakage Rate Testing at Nine Mile Point Unit 1."
(8)
ANSI/ANS - 56.8 - 1987, "Containment System Leakage Testing Requirements."
AMENDMENTNO.
140 143a I
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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT 3.3.4 PRIMARY NTAINMENTI OLATIONVALVE 4.3.4 PRIMARY NTAINMENTI LATI N VALVES
~)~li I~ill y:
Applies to the operating status of the system of isolation valves on lines open to the free space of the primary containment.
~Obe ~iv:
To assure that potential leakage paths from the primary containment in the event of a loss-of-coolant accident are minimized.
Applies to the periodic testing requirements of the.
primary containment isolation valve system.
~be ~iv:
,To assure the operability of the primary containment isolation valves to limit potential leakage paths from the containment in the event of a loss-of-coolant accident.
a.
Whenever the reactor coolant system temperature is greater than 215F, all containment isolation valves on lines open to the free space of the primary containment shall be operable except as specified in 3.3 4b below.
The primary containment isolation valves surveillance shall be performed as indicated (See Table 3.3.4) a.
At least once per operating cycle the operable isolation valves that are power operated and automatically initiated shall be tested for automatic initiation and closure times.
b.
In the event any isolation valve becomes inoperable the system shall be considered operable provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.
b.
At least once per quarter all normally open power operated isolation valves shall be fully closed and reopened.
AMENDMENTNO.
140 144
'I
LIMITINOCONDITION FOR OPERATION Table 3.3A PRIMARYCONTAINMENTISOLATIONVALVES UNES ENTEAINO FREE SPACE OF THE CONTAINMENT Une or System D
eil Vent & Pur e
No. of Valves
{Each Une)
Location Relative To Primary Containment Normal Position Maximum Initfatlng Signal {All Oper. Time Action on Valves Have. Remote Motive Power'Seo)
Initiating Signal Manual Backup)
N> Connection
{One Line) ir Connection
{One Line)
Su ression Chamber Ven & Pur e
N~ Connection
{One Line) ir Connection
{One Line)
Outside Outside Outside Outside Outside-Outside Outside Outside Closed Closed Closed Closed Closed Closed Closed Ctosed Pn/DC Solenoid AC Motor Pn/DC Solenoid AC Motor Pn/DC Sotenoid AC Motor Pn/DC Solenoid AC Motor 16 30 16 30 16 30 16 30 Close Close Close Close Close Close Close Close Reactor water level low-lowor high dryweil pressure or high radiation at stack monitoring Reactor water level low-lowor. high drywall pressure or high radiation at stack monitoring Qriiwall Nu~Makeu
{One Line)
Su reaaiau Chamber Mk~Mekeu
{One Line) ell ui ment Drain Line{ )
{One Line) eil Floor Drain Line{i)
{One Line)
Outside Outside Inside Outside Inside Outside Closed Pn/DC Solenoid Closed Pn/DC Solenoid Open AC Motor Open Pn/DC Solenoid Open AC Motor Open Pn/DC Solenoid eo eo eo eo 60 60 Close Close Close Close Close Close Reactor water level low-lowor drywall high pressure Reactor water level low-lowor drywall high pressure Reactor water level tow-Iow or drywall high pressure Vacuum Relief Atmosphere to Pressure Suppression System
{Three Lines) eactor Cleanu S stem Re'lief Valve{ )
Outside Outside Closed Closed Pn/DC Solenoid Self Act. Ck.
Open Negative pressure relative to atmosphere
~iechar e
{One Line to Suppression Chamber)
Outside Closed Self Act. Ck.
AMENDMENTNQ, 1 4p
~4O 146
I
UMITINQCONDITIONS FOR OPERATION Table 3.3e4 {Continued)
PRIMARYCONTAINMENTISOLATIONVALVES LINES ENTERINQ FREE SPACE OF THE CONTAINMENT Vne or System No. of Valves
{Each Une)
Location Relative To Primary Normal Containment Position Maximum Initiating Signal {AII Oper. Time Action on Valves Have Remote MotIve Powers
{Sec) initiating Signal Manual Backup)
I Drrwe~ll Su ~l
{Two Lines)
Outside Open Pn/DC Solenoid 60 Close Su ression Chamber Su I
{One Line)
Drrrell Return Outside Outside Open Open Pn/DC Solenoid Pn/OC Solenoid eo 60 Close Close Reactor Water level low-low or high drywell pressure
{One Line)
Su ression Chamber Return
{One Line)
Outside Open Pn/DC Solenoid 60 Close jt¹O¹~¹t2 Sem tin Orwell~Su )~1{1)
{Three Unss)
Su ression Chamber Su I { )
{One Line)
Drrwell~Retur {1)
{One Line)
Outside Outside Outside Open Open Open Pn/DC Solenoid Pn/OC Solenoid Self Act. Ck.
eo 60 Close Close Reactor water level low-low or high drywell pressure Su ression Chamber Return{ )
IOne Line)
Outside Open Self Act. Ck.
AMENDMENTNO.
1 6, 140 147
LIMITWQCONDITION FOR OPERATiON Table 3.3.4 (oonthwed)
PRIMARYCONTAWMENT ISOLATIONVALVES W S Wo FR SPAC 0 THE CON AWM N Une or System ggre S~ra (
Looathn Refathra No. of Valves To Primary (Each Line)
Conte@ment Normal Poaldon Maxknum Infdatkig Signal (AN Oper. T)me Actbn on Valves Have Remote Motive Power (Seo)
Inltlatlng Signal Mamal Backup)
~um Suc io
~
~
(Four Uncs From Suppression Chamber)
(Two Teat Lines to Suppression'Chamber)
(Keep Fill)
(Two Lines)
Core S ra Hi h Pain Vent(
)
(Two Uneal Outside Outside Outside Outside inside Open Closed Open Closed Closed AC Motor AC Motor Self Act. Ck.
Pn/DC Solenoid AC Motor 90 21 27 27 Close Close Close Remote Manual Reactor water level Iow-lowor high drywall pressure Reactor water level low-Iowor high drywall pressure Co el IIi S rasa n Ch ber(2) u I
Outside Open Pn/DC Solenoid Remote Manual (Four Linea)
~illBronc IFour Lines)
Su rassion hamber Branc
( I (One Branch for Each System) um Suction om Su ressio (Four Lines)
Con ai e
ra es I e 0
0 20 ~
Outside Closed Selt Act. Ck.
Outside Outside Open AC Motor Outside Closed AC Motor Closed Self Act. Ck.
70 60 Remote Manual Remote Manual lOne Une) er Iin Va o
rua tOne Line)
Outside Closed AC Motor Remote Manual AMENDMENTNO.
1 2, 140 148
8, ~
LIMITINOCONDITIONS FOR OPERATION Table 3.3.4 (Continued)
PRIMARYCONTAINMENTISOLATIONVALVES LINES ENTERINO FREE SPACE OF THE CONTAINMENT Une or System No. of Valves (Each Une)
Location Relative To Primary Normal Containment Position Maximum Initiating Signal (All Oper. Time Action on Valves Have Remote Motive Powers (Seo)
Initiating Signal Manual Backup)
Containment Atmos here Monitorin Su I
Line
{One Line)
Outside Open Pn/DC Solenoid eo Close Reactor water. level low-lowor high drywell pressure Containment Post LOCA Vent (Two Lines) 2 Pur e-TIP Indexers>>
)
Outside Outside Closed Pn/DC Solenoid BO Closed Self Act. Ck.
Close Reactor water level low-lowor high drywell pressure (One Line)
Traversin Incore Probe{
)
(Four Uncs)
Breathin Air Connectio
{One Une)
Service Water Connection(
)
(One Line)
Outside Inside Outside Inside Outside Closed Closed Closed Closed Closed AC Motor eo Close Reactor water level Iow-lowor high drywell pressure Local Manual Recirculation Pum Coolin Water{ )
LINES WITH A CLOSED LOOP INSIDE CONTAINMEN VESSELS Supply Line Return Line Outside Outside Open Open Self Act. Ck.
DC Motor eo Remote Manual D
ell Cooler Water>>
)
Supply Line Return Une Outside Outside Open Open Self Act. Ck.
DC Motor eo Remote Manual I
~
I AMENDMENTNO.
140
>aSB[
Notes:
Pn - Pneumatically Operated One valve in each separate line and one valve in each common line.
(1)
These valves do not have to be vented during the Type A test.
However, Type C leakage from these valves is added to the Type A test results, if not vented.
(2)
These valves are provided with a water seal capability.
No Appendix J or IST leakage rate testing is required.
(3)
These valves are water leak rate tested and acceptance criteria are established in accordance with the IST Program.
(4)
These valves are provided with a water seal.
Valves shall be tested during each refuel outage not to exceed two years consistent with Appendix J water seal testing requirements.
Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.
(5)
These valves do not meet the requirements of 10CFR50 Appendix J, Section II-H. No testing required.
~
AMENDMENTNO.
140 1aeb I
I
BASES FOR 3.3.3 AND 4.3.4 PRIMARY CONTAINMENTISOLATIONVALVES Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.
Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Except where check valves are used as one or both of a set of double isolation valves, the isolation closure times are presented in Table 3.3.4.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a toss-of-coolant accident.
Details of the isolation valves are discussed in Section VI-D.(1>> For allowable leakage rate specification, see Section 3.3.3/4.3.3.
For the design basis loss-of-coolant accident fuel rod perforation would not occur until the fuel temperature reached 1700F which occurs in approximately 100 seconds.(2>>
A required closing time of 60 seconds for all primary containment isolation valves will be adequate'to prevent fission product release through lines connecting to the primary containment.
For reactor coolant system temperatures less than 215F, the containment could not become pressurized due to a loss-of-coolant accident.
The 215F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels.
The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 that a line willnot isolate (Fifth Supplement, p. 115).
More frequent testing for valve operability results in a more reliable system.
I ln addition to routine surveillance as outlined in Section VI-D.1.0 each instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument.
The line will be purged by isolating the flow check valve, opening the bypass valves, and opening the drain valve to the equipment drain tank. When purging is sufficient to clear the line of non-condensibles and crud, the flow-check valve will be cut into service and the bypass valve closed.
The main valve willagain be opened and the flow-check valve allowed to close.
The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing part of the system.
Instruments will be cut into service after closing the bypass valve.
Repressurizing of the individual instruments assures that flow-check valves have reset to the open position.
An in-depth review of the NMP-1 design and operation relative to Appendix J requirements has evaluated the various system/valving configurations. "
The results of the evaluation and subsequent clarifications>> are reflected in this specification and its bases.
(1)
(2)
(3)
(4)
(5)
UFSAR Nine Mile Point Nuclear Generation Station Unit 1 Safer/Corecool/GESTR-LOCA Loss of Coolant Accident Analysis, NEDC-31446P, Supplement 3, September, 1990.
FSAR NRC Safety Evaluation Report, dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."
Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications & Conformance with 10CFR50 Appendix J SER."
AMENDMENTNO.
140 149
)
~