ML19345H255
| ML19345H255 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/20/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19345H256 | List: |
| References | |
| NUDOCS 8105010437 | |
| Download: ML19345H255 (10) | |
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UNITED STATES
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.,E NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D. C. 20555
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COMMONWEALTH EDISON COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-254 s
QUAD CITIES NUCLEAR POWER STATION UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 1.icense No. DPR-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated March 29,1979, 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 i
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 t'he common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 l
of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is anended by changes to the Technical Speciff-l cations as indicated in the attachment to this license amendment and l
paragraph 3.B of Facility Operating License No. DPR-29 is hereby amended to read as follows:
3.B Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 68, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8105010 hTl l
. 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Thomas #(Ippolito, Chief A
Operating Reactors Branch #2 Division of Licensing
Attachment:
l Changes to the Technical Specifications Date of Issuance: April 20,1981
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ATTACHMENT TO LICENSE AMENDMENT NO. 68 FACILITY OPERATING LICENSE NO. DPR-29' DOCKET NO. 50-254 Revise the Appendix "A" Technical Specifications as follows:
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QUAD-CITIES DPR-29 tion valves which shall be med at a pressure of 25 psig, each operating cycle. Bolted double gasketed seals sha!! be easted whelevar the seal is closed aAer being opened sad at least as each operating N
- 2) Personnel air lock door seals shall be tested at a pressure of 10 peig each operating cycle.
L Acceptance Criteris and Correc-tive Action for LLRT If the total leakage rates listed below are exceeded repsirs and reests shall be performed to cor.
rues the condition.
- 1) Double gasketed seals 10f.L.
(48)
- 2) a) Testable penetrations and isolation valves 30%L,(48) b) Any one penetration or isolation valve except main steamline isolation valves 5%L,(48) c) Any one main steamline isolation valve 11.5 scf/hr at 25 psig.
- 3. Pressure Suppression Chamber.
- 3. Pressure Suppression Chamber '
Reactor Building Vacuum Breakers Rasetoe Building Vacuum Breakers
- s. Except as specified in Specification a.
The pressure supp ession cham.
l 3.7.A.3.b below, two pressure sup-ber-reactor build.ng vacuum
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l Amendment No. 68 3.7f4.7 4 i
t QUAD-CITIES DPR-29 I
multiplying the maximum allowable 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.
The primary containment leak rate test frequency is based on mentaining adequate assurance that the leak rate remains within the specincation. Allowing the test intervals to be extended up to 8 months permits sorne Aexibility needed to have the tests coincide with scheduled or unscheduled shutdown Periods.
The data reduction methods of ANSI N45.41972 will be applied for integrated leak rate tests.
l The pnetration and air purge piping leskage test frequency, along with the containment !cak rate tests, is adequate to allow detecdon of leakage trends. Whenever a doubic.gasketed penetration (primary containment head equipment hatches and the suppression che: ber 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 48 psig is consistent with the accident analyses and the maximum preoperational leak rate tes'. pressure. It is expected that the majority of the leakage from valves, penetrations, and seats would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur.
Sus!. leakage paths that raay afect signiAcantly the consequences of accidents are to be minimized. The-l personnel alt lock is tested at 10 psig because the inboard door is nct des:gned to shut in the opposite darsenon.
The results of the loss-of. coolant accident analysis referenced in Section 5.2.4.3 of the SAR indicate that
(
Sasion products would not be released directly to the environs because ofleakage from the main steamline isolation valves due to holdup in the steam system complex. Although this effect would indicate that an i
l adequate margin exists with regard to the release of fission products, a program will be undertaken to thrther reduce the potential for such Icakage to bypass the standby Jtas t.eatment system.
Surveillance of the reactor building pressure suppression chamber vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leaktightnt:s tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition.
As a result, a testing frequency of 3 months for operability is considered justifed for this equipmer:t.
Inspections and calibrations are performed during refuelin5 outages, this frequency being based on -
experience and judgment.
Pressure suppression chamber-drywell vacuum breakers monthly operability tests are perfortred to check the capability of the disks to open and close and to verify that the position indication and al.trm circuits fhnction properly. The disks must open dunng accident conditions and during transient additions of energy through relief valves. This periodic operation of the disks and the quality ofequipmentjustify the frequency of operability tests of this equipment.
Following each quarterly operability test, a ditferential pressure decay rate test is performed to verify that leakage from the drywc!! to the suppression (hamber is within specified limits.
Measurement of force to open. calibration of position switches, inspection of equipment, and functional testing are performed during eacn refuelint outage. This frequency is based on equipmer.: quality, experience, and judgment. Also, a more string ni diferential pressure decay rate test is performed during refueling outages than is performed monthly. This test is performed to verify that total !cakage paths between the drywel: and suppression chamber are not in excess of the equivalent to a 1-inch orifice.
'lliis small leakage path is only a smal: fraction of the allowable, thus integrity of the containment system is asstired prior to startup following each reft eling outage (Reference 1).
Amendment No. 68 3.7/4.7-16 P00RORBE
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UNITED STATES E%
NUCLEAR liEGULATORY COMMISSION o
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WASHINGTON, D. C. 20665
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COMMONWEALTH EDISON COMPANY _
AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY _
DOCKET NO. 50-265.
QUAD CITIES NUCLEAR POWER STATION UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 62 License No. DPR-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Commonwealth Edison Company l
(the licensee) dated March 29, 1979, complies I
with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regu-lations set forth in 10 CFR Chapter I; 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 I
and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l
D.
The issuance of this amendment will not be inimical to the common l
defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amendeo by changes to the Technical Specifi-2.
cations as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No. DPR-30 is hereby amended to read as follows:
3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendraent No. 62, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
2-3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/I
/
p 1
, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Srecifications Date of Issuance:
April 20,1981 l
l l
l
i ATTACHMENT TO LICENSE AMENDMENT NO. 62 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 l
Revise the Appendix "A" Technical Specifications as follows:
t Remove Replace 3.7/4.7-4 3.7/4.7-4 3.7/4.7-16 3.7/4.7-16
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QUAD-CITIES DPR-30 tion valves which sha!! be tested at a pressure or25 psig.
sach operating cycle. Bolted double gasketed seals shall be tested whenever the seal is closed ener being opened and at least at each operating cycle.
- 2) Personnel air lock door seals shall be tested at a pressure of 10 psig each operating cycle.
L Acceptance Criteria and Correc-tive Action for LLRT If the total !'akage rates listed e
below are exceeded, repairs and retests shall be performed to cor-rect the condition.
- 1) Double gasketed seals 10%L, (48)
- 2) a) Testable penetrations and isolation valves 30%L, (48) b) Any one penetration or isolation valve except main steamline isolation valves 5%L,(48) c) Any one main steamline isolation valve i 1.5 scf/hr at 25 psig.
i
- 3. Pressere Suppression Chamber-
- 3. Pressure Suppression Chamber.
Reactor Building Vacuum Breakers Reactor Building Vacuum Breakers a.
Except as speciAedin Speci5 cation
- a. The pressure suppression cham-3.7.A.3.b below, two pressure sup-ber reactor build ng vacuum 3.7/47-4 Amendment iio. 62
QUAD-CITIES DPR-30 multiplying the maximum a!!owable leak rate by 0.75, thereby providing a 25% margin to a!!ow for leakage deterioration which may occur during the period between leak rate tests.
The peirnary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specincation. A!!owing the test intervals to be extended up to 8 months permies some Sexibility needed to have the tests coincide with scheduled or unscheduled shutdown l
Periods.
l l
The data reduction methods of ANSI N45.41972 will be applied for integrated Icak rate tests.
j 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 l
oipatainment 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 48 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 parts of the facility could occur.
Such leakage paths that may afect significantly the consequences of accidents are to be minimized. The i
personnel air lock is tested at 10 psig because the inboard door is not designed to shut in the opposite direcuen.
The results of the loss-of coolant accident analysis refexenced in Section 5.2.4.3 of the SAR indicate that Assaon products would not be released directly to the environs because ofleakage from the main steamline isolation valves due to holdup in the steam system complex. Although tnis efest would indicate that an edequate margin exists with regard to the release of fission products, a program will be undertaken to 4.her reduce the potential for such feakage to bypass the standby gae treatment system.
Surveillance of the reactor building. pressure suppression chamber vacuum breakers consists of operability checks and leakage tests (condueted as part of the containment leaktightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition.
As a result, a testing frequency of 3 months for operability is considered justified for this equipment.
Inspections and calibra: ions are performed during refueling outages, this frequency being based on experience and judgment.
Pressure suppression chamber-drywell vacuum breakers monthly operability tests are performed to check the capability of ti e disks to open and close and to verify that the position indication and alarm circuits function properly. The disks must open during accident conditions and during transient additions of energy through relie.* valves. This ;;eriodit operation of the disks arid the quality of equipment justify the frequency of operability tests of this equipment.
Following each quarterly operability test, a diferential pressure decay rate test is performed to verify that leakage from the drywell to the suppression chamber is within specified limits.
Measurement of force to open, calibration of position switches inspection of equipment, and functional testing are performed during each refueling outage. This frequeng " base 9 on equipment quality, a
experience, and judgment. Also, a more stringent diferential pressure decay rate test is performed durms refheling outages than is performed monthly. This test is performed to verify that total leakage pa:hs between the drywell and suppression chamber are not in excess of the equivalent to a 1-inch orifice.
This smailleakaga path is only a small fraction of the a!!om able, thus integrity of the containment system is assured prior to startup following each refue!ing outage (Reference 1).
3.7/ 47-16 UIllustW s
Amendment No. 62
._