ML20096C156

Amend 106 to License DPR-49,revising Tech Specs to Incorporate Changes Resulting from Facility Compliance W/ Requirements of 10CFR50,App J

ML20096C156
Person / Time
Site:	Duane Arnold
Issue date:	08/24/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co
Shared Package
ML20096C160	List: ML20096C156 ML20096C162
References
DPR-49-A-106 NUDOCS 8409050120
Download: ML20096C156 (21)

Text

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9-UN!TED STATES

['jg,g,g NUCLEAR REGULATORY COMMISSION M

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IOWA-ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARN0LD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.106 License No. DPR-49 5

1.

The Nuclear Regulatory Comission (the Commission) has found that:

A '.

The application for amendment by Iowa Electric Light & Power Company, et al, dated March 16, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 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.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:

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-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.106, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: August 24, 1984 l

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ATTACHMENT TO LICENSE AMENDMENT N0. 106 FACILITY OPERATING LICENSE N0. DPR-49 DOCKET NO. 50-331 Revise the Appendix A Technical Specifications by' removing the current pages and inserting the revised pages listed below. The revis'ed areas are identified by vertical. lines.

LIST OF AFFECTED PAGES 3.7-3 3.7-4 3.7-5 3.7-6 3.7-6a 3.7-7 3.7-8 3.7-9 3.7-20 3.7-21 3.7-22 3.7-23 3.7-23a*

3.7-24 3.7-37 3.7-38 3.7-39 3.7-49

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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 2)

Closure of containment isolation valves for the Type A test shall be accomolished by normal-mode of actuation and without any preliminary exercisina or adiustments.

3)

The containment test pressure shall be' allowed to stabilize for a period of about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of a leakaae rate test.

4)

The reactor coolant pressure boundary shall be vented to the containment atmosphere prior to the test and remain open during the test.

5)

Test methods are to comply with ANSI N45.4-1972.

6Property "ANSI code" (as page type) with input value "ANSI N45.4-1972.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.)

The accuracy of the Type A test shall be verified by a supplemental test.

An acceptable method is described in Appendix C of ANSI N45.4-1972.

3Property "ANSI code" (as page type) with input value "ANSI N45.4-1972.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..7-3 Amendment No.106

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DAEC-1' LIMITING CONDITION FOR OPERATION

' SURVEILLANCE REQUIREMENT 7)

Periodic Leakaae Rate Tests Periodic leakage rate tests shall be performed at peak pressure (Pa).

8)

Acceptance Criteria The maximum allowable ' leakage.

rate (Lam) is 0.75 La where La is defined as the design basis accident -leakage rate of 2.0 weight percent of contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 54 psig.

9)

Additional Requirements If any periodic Type A test fails to meet the applicable acceptance criteria the test schedule aoplicable to subsequent Type A tests will be reviewed and approved by the Commission.

If two consecutive periodic Type A tests fail to meet the acceptance criteria of 4.7.A.2.(a)(9) a Type A test shall be performed at each Diant shutdown for major refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the subiect acceptance criteria after which time the retest schedule of 4.7.A.2.(d) may be resumed.

Amendment No, 106 3 7-4

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT b.

Type B Tests Type 8 tests refer to penetrations with casketed seals, expansion bellows or other type of resilient seals as shown in Table 3.7-1.

1)

Test Pressure All Type B tests shall be l

oerformed by local pneumatic oressurization of the cor,tainment penetrations, eith'er individually or in groups, at a pressure not less than Pa.

l 2)

Acceptance Criteria The combined leakage rate of all non9trations subiect to Type 8 and C tests shall be less than 0.60 La.

c.

Tyoe C Tests 1)

Type C tests shall be performed as specified in Table 3.7-2.

Each valve to be tested shall be closed by normal operation and withoub any preliminary exercisino or adiustments.

2)

Acceptance criteria - The l

combined leakage rate for all penetrations subject to Type B l

and C tests shall be less than 0.60 La.

3)

The leakage from any one main steam isolation valve shall not exceed 11.5 scf/hr at an initial test pressure of 24 psig.

4)

The leakage rate from any containment isolation valve whose seating surface remains water covered post-LOCA, and which is hydrostatically Type C tested, shall be included in the Type C test total.

These valves are identified in Table 3.7-2 of this Technical Specification.

3.7-5 Amendment No. 106

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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT-d.

Periodic Retest Schedule 1)

-Type A Test

. After the preoperational leakage. rate tasts, a set of three Type A tests shall be.

oerformed. at approximately equal intervals durino each 10-year service - oarind.

(These intervals may be extended up to 1

eight months if necessary to coincide with refueling-outages.) The third test of each set shall be conducted when the plant is shut down for the 10-year plant in-service insoections.

The performance of Type A tests shall be limited to oeriods when the olant facility is nonoperational and secured in the shutdown condition under administrative control and in accordance with the plant safety procedures.

2)

Type B Tests a)

Penetrations and seals of this type (except air locks) shall be leak tested at Pa (54 osig) durinq each reactor shutdown for ma,ior fueling or other convenient interval but in no case at intervals areater than two years, b)

The personnel airlock shall be cressurized to Pa (54 psia) and leak tested at least once every six (6) months.

This test interval may be extended to the next refueling outage (uo to a maximum interval between Pa tests of 24 months) provided there have been no airlock openings since the last successful test at P -

a Amendment No.106 3.7-6*

' DAEC-1 LIMITING CONDITION FOR OPERATION SURVEII. LANCE RE0VIREMENT c)

Within three (3) days after securing the airlock when containment inteority is required, the airlock gaskets shall be leak tested at a pressure of P -

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3.7-6a Amendment No. 106 A

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DAEC-1

' LIMITING CONDITION FOR OPERATION ^

SURVEILLANCE REOUIREMENT

3). Type C Tests

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Type C tests shall.be performed during each reactor shutdown for major refueling or other convenient interval but in no case at-intervals greater.than two years.

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Containment Modification-Any major modification,

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replacement of a c'omponent which is part of'the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test'shall be followed by either a Type A, Type.8, or Tyoe C test, as apolicable, for the area affected by the modification.

The measured leakage from this test shall be included in the test report.

The acceptance criteria as appropriate, shall be met.

Minor modifications, replacements, or resealing of seal-we4ded doors, performed directly prior to the conduct of a scheduled Type A test do not require a separate test.

f.

Reporting-Periodic tests shall be the sub.iect of a summary technical report submitted to the Commission approximately 3 months after the conduct of each test.

The report will be titled " Reactor Containment Integrated Leakage Rate Test."

l 3.7-7 Amendment'No. 106

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_ LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT.

The report shall. include a schematic arrangement or description of the leakage rate measurement system, the instrumentation used, the supplemental test method, the test program selected, and all subseouent periodic tests.

The report shall contain an analysis and interpretation of the leakage rate test data for the-Tyoe A test results to the extent necessary to demonstrate the acceptability'of the containment's leakage rate in meetina the acceptance criteria.

For each periodic test, leakaae test results from Type A, B, and C tests shall be reported.

The report-shall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last Type A test.

Leakage test results from Type A, B, and C tests that f ailed to meet the acceptance criteria shall be reported in a separate accompanying summary report.

The Type A test summary report shall include an analysis and interpretation of the test data, the least-squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the 3.7-8 Amendment No, 106 c

e DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE RE0VIREMENT acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.

The Tyoe B and C tests summary report shall include an analysis and interpretation of the data and the condition of the components which contributed to any failure in meeting the acceptance criteria.

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3.7-9 Amendment No. 106

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-TABLE 3.7-1 CONTAINMENT PENETRATIONS SUBJECT TO TYPE B TEST REQUIREMENTS Penetration #

Type -

Description 2

Personnel Lock Equipment Door l

1 Testable Gaskets l

1 Personnel Lock 2 Personnel Lock Doors and Phnetrations 2

Testable Gaskets Equipment Access 4-Testable Gaskets Head Access 6

Testable Gaskets CRD Removal Hatch 35A-D Testable Gaskets TIP Drives (4) l 53 Testable Gaskets Soare Testable Gaskets Drywell Head Flange 58 A-H Testable Gaskets Stabilizer Access Ports (8) 200A-B Testable Gaskets Torus Access Hatches (2) l 1008,C,E,F,G Electrical Canister (8,C,E,F) Neutron Monitorina, (G) RPV Vibration Monitorino 101A,C Electrical Canister (C)' (A) Recirc Pump Power 103 Electrical Canister Thermocouples 104A-D Electrical Canister CR0 Rod Position Indicator 1058,0 Electrical Canister (8,D) Power & Control 106A,C Electrical Canister (A,C) Power & Control 2308 Electrical Canister Vacuum Breakers Electrical Cables 3.7-20 Amendment No,106

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DAEC-1 TABLE 3.7-1;(Continued)

CONTAINMENT-PENETRATIONS SUBJECT!TO TYPE B TEST REQUIREMENTS Penetration #

Type-Description 7A-0 Exoansion Bellows Steam to Turbine 9A,B Expansion Bellows' RPV Feedwater 10 Expansion Bellows Steam to RCIC Turbine

'll Expansion Bellows Steam to HPCI Turbine i'

12 Expansion Bellows Shutdown Pump Supply RHR 13A,8 Expansion Bellows RHR Pump Discharge 15 Expansion Bellows RWCU Supply l

16A.B Expansion Bellows Core Spray Pump Discharae 17 Exoansion Bellows RPV Head Soray 201A-H Expansion Bellows Vent Lines l

25 Flange "0" Rinos!

Orywell Purge Outlet CV-4302 26,220 Flange "0" Rings Drywell & Torus Purge Supoly, CV-4307, CV-4308 205 Flange "0" Rings Torus Purge Outlet, CV-4300 213A, 8 Flange "0" Rings Torus Orain Lines l

231 Flange "0" Rings Torus Vacuum Breakers, CV-4304, CV-4305 ITest inboard flange of designated valves.

2 Testing to be in accordance with Technical Specification Section 4.7.A.2.d.2.

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Amendment No. 106 3.7-21

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TABLE 3.7-2 CONTAINMENT ISOLATION VALVES SU6 JECT TO TYPE C TEST REOUIREMENTS-

' PENETRATION #

' SYSTEM BOUNDARY VALVES l

7A' Main Steam Line

_CV-4412", 4413 7B Main Steam Line CV-44154, 4416-7C.

Main Steam Line CV-44184, 4419 70 Main Steam Line CV-4420", 4421 8

Main Steam Line Drain MO-4424 9A Feedwater & HPCI Feed V-14-3 9A2 Feedwater & HPCI Feed M0-4441, MO-2312 98 Feedwater V-14-1 982 Feedwater & RCIC Feed & RWCU Return M0-2740, MO 4442, M0-2512 10 RCIC Condensate Return CV-2411 10 Steam to RCIC Turbine MO-2401 11 Steam to HPCI Turbine M0-2239 11 HPCI Condensate Return CV-2212 15 RWCU Supply

' M0-2700, MO-2701 16A Core Spray Pump Discharge M0-2115, MO-2117 168 Core Spray Pump Discharge M0-2135, M0-2137 4

19 Orywell Floor Drain Discharge CV-3704, CV-3705 20 Demineralized Water Supply V-09-65, V-09-111 l

21 Service Air Supply V-30-287, Blind F1ance l

22, 229 Containment Compressor Discharge CV-4371A, CV-4371C, V-43-214 23A, B Well Cooling Water Supply CV-5718A, CV-57188, l

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CV-5719A, CV-57198, 3

3 Well Cooling Water Return CV-5704A, CV-57048, 24A,8 CV-5703A, CV-57038, 25 Drywell Purge Outlet CV-43024, CV-4303, CV-431,0 26, 220 Orywell and Torus Purge Supply CV-4306, CV-4307", CV-4308" 26, 220 Orywell and Torus Nitrogen Makeup CV-4311, CV-4312, CV-4313 3.7-22 Amendment tio.106

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TABLE 3.7-2 (Continued)

CONTAINMENT ISOLATION VALVES SUBJECT TO TYPE C TEST REQUIREMENTS PENETRATION 4 SYSTEM B0UNDARY VALVES l

320 Containment Comoressor Suction CV-4378A, CV-43788 12E Recirc Pump "A" Seal Purge V-17-96, CV-18048 32F Recirc Pump "B" Seal Purge V-17-83, CV-180dA 35A,B,C,0 T.I.P Drives T.I.P. Ball Valves and Check Valve on X-35A 361 CR0 Return V-17-53. V-17-52 39A Containment Soray/ CAD Supoly SV 4332A, SV 4332B 398 Containment Spray / CAD Supply SV-4331A, SV-43313 400 Post-Accident Sampling / Jet Pump Sample SV-4594A, SV-45948 41 Recirc Loop Sample CV-46394, CV-4640 42 Standby Liquid Control V-26-8, V-26-9 46E O Analyzer SV-81058, SV-8106B g

48 Drywell Equipment Drain Discharge CV-3728. CV-3729 508 0 Analyzer SV-8101A, SV-8102A, 2

50E 0 Analyzer SV-8103A, SV-8104A, 2

500 O Analyzer SV-8105A, SV-8106A g

543 Reactor Building Closed Coolina Water Return M0-4841A 553 Reactor Building Closed Cooling Water Supply M0-4841B 56C 0 Analyzer SV-81018, SV-8102B, 2

56D O Analyzer SV-8103B, SV-8104B g

205 Torus Purge Outlet CV-43004, CV 4301 CV-4309 211A Torus Spray /CAO Supply SV-4333A, SV-43338 211B Torus Spray /CAO Supply SV-4334A, SV-43348 2121 RCIC Turbine Exhaust V-24-84, V-24-23 V-24-46, V-24-47 2141 HPCI Turbine Exhaust V-22-16, V-22-17*

V-22-63, V-22-64 3.7-23 Amendment No. 106

DAEC-1 TABLE 3.7-2 (Continued)

CONTAINMENT ISOLATION VALVES SUBJECT TO TYPE C TEST REQUIREMENTS.

PENETRATION #

SYSTEM BOUNDARY VALVES 219 HPCI/RCIC Exhaust Vacuum Breaker M0-2290A, MO-22908 l'

2221 HPCI Condensate

~V-22-21, V-22-22" 2298 0 Analyzer SV-8107A, SV-8108A, 2

229C 0 Analyzer SV-8109A, SV-8110A, 2

229G 0 Analyzer-SV-81078, SV-810AB, 2

229F 0 Analyzer SV-81098, SV-8110B 2

l 229H Post-Accident Sampling System SV-8772A, SV-87728

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Liquid Sample Return i

231 Torus Vacuum Breakers CV-4304", V-43-169 231 Torus Vacuum Breakers CV-4305", V-43-168 l

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3.7-23a l

Amendment flo. 106 l

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DAEC-1 NOTES TO TABLE 3.7-2 I Test volume is filled with domineralized water then pressurized to-54 psig with air'or nitrogen for test.

For all other penetrations (except 7A-0), test volumes are pressurized to 54 psia with air or nitrogen for test.

2M0.4441, MO-4442 will be remote manually closed.

3 In accordance with 10 CFR 50, Appendix A, General Design Criterion 57, the redundant barriers are a single isolation valve outside containment and a closed system inside.

Testing of the sinale isolation valve only is required.

• Tested in reverse direction.

l 3.7-24 Amendmer t No.106

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DAEC with this -leak rate and a standby gas treatment system filter efficiency of 90% for halogens, 90% for particulate iodine, and assumino the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about 2 rem and the maximum thyroid dose is about 32 rem at the site boundary over an exposure duration of two hours.

The resultant thyroid dose that would occur over the course of the accident is 98 rem at the boundary of the low population zone (LPZ).

Thus, these doses _ 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 is conservative and provides additional margin between expected offsite doses and 10 CFR 100 guidelines.

The design basis accident leak rate (L ) at the peak accident l

a oressure of 54 psig (P ) is 2.0 weight percent per day.

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a allow a margin for possible leakaae deterioration durina the interval between Type A tests, the maximum allowable containment operational leak rate (Lam), is 0.75 L,.

l 3.7-37 Amendment No. 106

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r Type B and Type C tests are performed on testable penetrations and isolation valves during the interim period between Type A tests.

This provides assurance that components most likely to undergo degradation between Type A tests maintain leaktight integrity.

The containment leakaae testing program is based on NRC quidelines l

for. development of leak rate testing and surveillance schedules for reactor containment vessels, (Reference 4).

I 5.

Drywell Interior I

The interiors of the drywell and suppression chamber are coated to prevent corrosion and for ease of decontamination.

The inspection of the coating during each ma.for refueling outage, 1

3.7-38 Amendment No. 106 i

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DAEC-1 assures the coating is intact.

Experience with this type of coating at fossil fueled generatino stations indicates that the inspection interval is adequate.

6.

Post LOCA Atmosphere Dilution In order to ensure that the containnent atmosphere remains inerted, i.e., the oxygen-hydrogen mixture below the flamnable limit, the capability to in,iect nitrocen into the containment after a LOCA is provided.

The CAD system serves as the post-LOCA Containment Atmosphere Oilution System.

By maintaining a minimum of 50,000 scf of liquid N in the storage bank it is 2

assured that a seven-day supply of N for post-LOCA containment l

2 inerting is available.

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The Post-LOCA Containment Atmosphere Oilution System design basis and description are presented in the response to Question l

G.7.3 and question G.7.4 of the FSAR.

In summary, the limiting criteria, based on the assumotions of Safety Guide No. 7 are:

1.

Maintain oxygen concentration in the containment during I

post-LOCA conditions to less than 4 Volume %.

Amendment No. 106 3.7-39

p-3 DAEC-1 3.7.A & 4.7 A REFERENCES

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Section 14.6 of the FSAR.

ti 2fI ~ ASME Boiler and Pressure Vessel Code, Nuclear Vessels, Section !!!,

maximum allowable internal pressure is 62 psig.

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3.

Staff Safety Evaluation of DAEC, USAEC, Directorate of Licensing,

. January 23, 1973.

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10 CFR Part-50, Aooendix J. Reactor Containment Testing Requirements,

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. Federal Register, April 19, 1976.

5.

DAEC Short-Term Program Plant Unique Analysis, NUTECH Ooc. No.

10W-01-065,~ August 1976.

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Supplement to,0AEC Short-Term Program Plant Unique Analysis, NUTECH Dod,..No. IOW-01-071~, October 1976.

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I Amendment No. 106 7,49

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