ML20211B123

Amend 97 to License DPR-40,transferring Recirculation Heat Removal Sys Surveillance Requirements from Containment Test Part of Tech Specs to New Tech Spec Section

ML20211B123
Person / Time
Site:	Fort Calhoun
Issue date:	06/03/1986
From:	Thadani A Office of Nuclear Reactor Regulation
To:
Shared Package
ML20211B101	List: ML20211B095 ML20211B123 ML20211B135
References
TAC-60597, NUDOCS 8606110483
Download: ML20211B123 (14)

Text

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UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 h

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 97 License No. DPR-40

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Omaha Public Power District (thelicensee)datedJanuary 26, 1983, 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 safetyinofcompliance the public,ithand (ii) that such activities will be conducted w 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 CFI, Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

8606110483 860603 PDR ADOCK 0500028D P PDR

2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.

DPR-40 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 97 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective within 30 days of date of its issuance.

FOR THE NUCLEAR REGULATORY COMISSION

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Ashok . Thadani, Director PWR Pr ject Directorate #8 Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: June 3, 1986

ATTACHMENT TO LICENSE AMENDMENT NO. 97 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages ii ii 3-44 3-44 3-45 3-45 3-46 3-46 3-47 3-47

< 3-48 3-48 3-49 3-49 3-51 3-51 3-52 3-52 3-84 3-85 i

e

. TABLE OF CONTENTS (Continued)

Page 2.12 Con trol Room Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 5 9 2.13 Nuclea r Detector Cooli ng Sy3 tem. . . . . . . . . . . . . . . . . . . . . . . . . . 2-60 2.14 Engineered Safety Features System Initiation

)

Ins trumenta ti on Settings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-61 2.15 Ins trumentati on and Control Sys ters . . . . . . . . . . . . . . . . . . . . . .

2-65 '

2.16 River Level.............................................. 2-71 l 2.17 Miscellaneous Radioactive Haterial Sources . . . . . . . . . . . . . . . 2-72 2.18 Shock Suppressors (Snubbers ) . . . . . . . . . . . . . . . . , . . . . . . . . . . . . 2-73  !

2.19 Fi re Protecti on Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-89 )

2.20 Steam Generator Coolan t Radioactivi ty. . . . . . . . . . . . . . . . . . . . 2-96 2.21 Pos t-Accident Moni toring Instrumentation. . . . . . . . . . . . . . . . . 2-97 2.22 Toxi c Ga s Moni to rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-99 1

3.0 SU RVE I LL AN C E REQU I REMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 3.1 Instrumentation and Control.............................. 3-1 3.2 Eq uipmen t a nd S ampl i ng Tes ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-17 3.3 Reactor Coolant System, Steam Generator Tubes, and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testin Surveillance..............................g ......... ..... 3-21 3.4 Reactor Coolant Sys tem Integri ty Testing. . . . . . . . . . . . . . . . t . 3-36

' 3.5 Containment Test.................w....................... 3-37 3.6 Safety Injection and Containment Cooling Systens Tests..,. 3-54 3.7 Emergency Power Sys tem Pe riodi c Tes ts . . . . . . . . . '. . . . . . . . . . . 3-58 3.8 Mai n Steam Is olation Va1ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-61 3.9 Auxilia ry Feedwater Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-62 3.10 Reac to r Co re Pa rame te rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 6 3 ~

3.11 Radiological Environmental Monitoring Programs. .. ........ 3-64 3.12 Radiological Was te Samplins and Monitoring. . . . . . . . . . . . . . . 3-69 3.12.1 Liqui d and Gaseous Effl uents . . . . . . . . . . . . . . . . . . . . . 3-69 3.12.2 Solid Radioactive Waste.......................... 3-71a 3.13 Radioactive Material Sources Surveillance................ 3-76 3.14 Shock Suppressors (Snubbers ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 7 3.15 Fi re Protecti on Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-80 3.16 Recirculation Heat Removal System Integrity Testing...... 3-84 l

4.0 DESIGN FEATURES................................................ 4-1 4.1 4.2 Site.....................................................

Contai nment Des i gn Fea ture's . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4-1 4-1 4.2.1 Containmint Structure............................ 4-1 4.2.2 4.2.3 Penetrations..................................... 4-1 Containnent Structure Cooli ng Sys tems . . . . . . . . . . . . 4-2 11 Amendment No. 38,43,#6,54,60,SA,86,93, 97

. 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

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 failed to meet the applicable acceptance criteria shall be reported in a separate summary report approximately 3 months after the conduct of these tests. The Type A test report shall include an analysis and interpretation of the test data, the least-squares fit analysis of the test data (Type A tests only), the instrumenta-tion error analysis (Type A tests only), and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the 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.

I 3-44 Amendment No. 77,97

3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

(7) Surveillance for Prestressing System l

a. Surveillance Requirements Two hundred ten dome tendons and 616 wall tendons shall be periodically inspected for symptoms of material deterioration or force reduction. Inspections will be performed on three dome tendons, one from each layer, and on three wall tendons of each orientation.

The surveillance tendons shall be inspected as follows:

(i) Lift-off readings shall be taken on each of the tendons selected to determine the load existing in the tendon at the time of inspection. At each surveillance period, readings may also be taken on the load cells of the special instrumented tendons.

Force reductions on the surveillance tendons and on the instrumented tendons will be compared. If good correlation exists between these two groups of tendons through several surveillance periods, consideration will be given to eliminating some lift-off readings and monitoring of the load cells as an alternative.

Each selected tendon shall be completely detensioned and examined for broken wires and any evidence of damage or deterioration of anchorage hardware.

(ii) One wire from each of three helical tendons and one wire of a dome tendon shall be removed. Each removed wire shall be carefully examined over its entire length for evidence of corrosion or other deleterious effects. Tensile tests shall be made on at least three samples cut from each of the four wires, removed, one at each end and one at midlength, the samples being of a maximum leagth practical for testing. In spe::ial cases, the use of fatigue tests and accelerated corrosion tests may be considered.

(iii) Comparisons shall be made between the quality control records and each of the surveillance inspection records for each of the surveillance tendons.

1 After completion of the tendon surveillance the individual detensioned tendons shall be retensioned to a force commen-surate with the average wire stress indicated by the last l lift-off reading for that tendon.

, 3-45 Amendment No.95,97

3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

b. Acceptance Criteria (i) The tendon force determined by the lift-off test shall be considered adequate if it is not less than the force shown on the appropriate lower limit curve of USAR l Figure 5.10-4, as adjusted for wire removal, for the elapsed time between the original prestressing and the particular surveillance period. These lower limit curves have been generated by calculating the difference between the anticipated tendon force at end of plant life and the minimum tendon force to meet the design requirements. One half of this difference has been added to the anticipated total loss of prestress at the end of plant life and the curves have been drawn to meet this limit. Since the lock-off force on individual tendons is varied to compensate for elastic shortening of the structure, the tendon force at 70% of ultimate strength, rather than the actual lock-off force shall be taken as the initial prestress force. An allowable limit of not more than one defective tendon out of the total sample population is acceptable, provided an adjacent tendon on each side of the defective tendon is tested and is found to meet criteria. Should one of the adjacent tendons be also found defective, the Commission shall be notified in accordance with Regula-tory Guide 1.16, " Reporting of Operating Information".

(ii) No unexpected change in corrosion conditions or grease properties.

(iii) All three tensile tests on any one wire indicate an ultimate strength at least equal to ths specified minimum ultimate strength of the wire. If a single test on any one wire shows an ultimate strength less than the specified minimum, the Commission will be i notified in accordance with Regulatory Guide 1.16, l

" Reporting of Operating Information". '

c. Corrective Action i

If the above acceptance criteria are not met, an imediate J investigation shall be made to determine the cause(s) for the non-conformance to the criteria, and results will be reported to the Commission within 90 days. ,

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d. Test Frequency '

The tendons in the prestressing system shall be inspected once every 5 years. .

l 3-46 Ancndtcat ;10. H, 97

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3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (continued) t 4

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. 3-47 Amendment No. 75, 97 I l

. 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (continued) l I

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I 3-48 Amendment No. 35, 97

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3.0 SURVEILLANCE REQUIREMENTS '

3.5 Containment Tests (Continued) l 1

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I l l Basis The containment is designed for an accident pressure of 60 psig.(2) While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 120 F. With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 288 F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at l

60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

l Safety analyses have been performed on the basis of a leakage rate of 0.1%

! of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, a reactor power level of 1500 MWt, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure

would be well bel theticalaccident.f3}0CFRPart100valuesintheeventofthemaximumhypo-The performance of a periodic integrated leakage rate l

3-49 Amendment No. d$' 97

l 3.0 SURVEILLANCE REQUIREMENTS 3.5 Con 91..;nent Tests (continued)

A reduction in prestressing force and changes in physical conditions are expected for the prestressing system. Allowances have been made in the reactor building design for the reduction and changes. The inspection results for each tendon shall be recorded on the forms provided for that purpose and comparison will be made with the previous test results and the initial quality control records.

Force-time trend lines will also be established and maintained for each of the surveillance tendons.

If the force-time trend line, as extrapolated, falls below the predicted force-time curve for one or more surveillance tendons, then before the next scheduled surveillance inspection, an inves-tigation shall be made to determine whether the rate of force reduction is indeed occurring for other tendons. If the rate of reduction is confirmed, the investigation shall be extended so as to identify the cause of the rate of force reduction. The exten-sion of the investigation shall determine the needed changes in the surveillance inspection schedule and the criteria and initial planning for corrective action. If the force-time trend lines of the surveillance tendons at any time exceed the upper bound curve of the band on the force-time graph, an investigation shall be made to determine the cause.

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3-51 Amendment No. 58, 97

3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

If the comparison of the corrosion conditions, including chemical tests of the corrosion protection material, indicates larger than expected change in the conditions from the time of installation or last surveillance inspection, an inves+igetion shall be made to detect and correct the causes.

The prestressing system is a necessary strength element of the plant safeguards and it is considered desirable to confirm that the allowances are not being exceeded. The technique chosen for surveillance is based upon the rate of change e' force and physical conditions so that the surveillance can either confirm that the allowances are sufficient or require maintenan:e before minimum levels of force or physical conditions are reached. The end anchorage concrete is needed to maintain the prestressing forces.

The design investigations have concluded that the design is adequate and this has been confirmed by tests. The prestressing sequence has shown that the end anchorage concrete can withstand I loads in excess of those which result when the tendons are anchored. :

Further, the containment building was pressure tested to 1.15 times j the maximum design pressure.

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3-52 Amendment No.97 1

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3.0 SURVEILLANCE REQUIREMENTS 3.16 Recirculation Heat Removal System Integrity Testing Applicability Applies to determination of the integrity of the shutdown cooling system and associated components.

Objective To verify that the leakage from the recirculation heat removal system components is within acceptable limits.

Specifications (1) a. The portion of the shutdown cooling system that is outside the containment shall be tested at 250 psig at each refueling outage, or other convenient intervals, but in n, case at intervals greater than 2 years.

b. Piping from valves HCV-383-3 and HCV-383-4 to the dis-charge isolation valves of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig at the testing fre-quency specified in (1)a. above.

c. Visual inspection of the system's components shall be performed at the frequency specified in (1)a. above to uncover any significant leakage. The Jeakage shall be measured by collection and weighing or by any other equivalent method.

(2) a. The maximum allowable leakage from the recirculation neat renoval system's components (which include valve stems, flanges, and pu:,p seals) shall not exceed one j gallon per minute, under the normal hydrostatic head from the SIRJ tar.k.

b. Repairs shall be msde as required to maintain leakage within the acceptable limits.

Basis The limiting leakage rates from the shutdown cooling system are judgment values based primarily on assuring that the components could operate without rechanical failure for a period on the order of 200 days after a design basis accident. The test pressure (250 psig) achieved either by normal system operation or by hydrostatic testing gives an adequate margin over the highest pressure within the system after a design basis accident.(1) Similarly, the hydrostatic test pressure for the return lines from the contain-ment to the shutdown cooling system (100 psig) gives an adequate margin ever the highest pressure within the lines after a design basis accident.

3-84 Amendment No.97

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3.0 SURVEILLANCE P.EQUIREMENTS 3.16 Recirculation Heat Removal System Integrity Testing (Continued)

A shutdown cooling system leakage of one gpm will limit off-site exposures due to leakaFe to insignificant levels relative to those calculated for direct leakage from the containment in the design basis accident. The safety injection system pump rooms are equipped with individual charcoal filters which are placed into operation by means of switches in the control room. The radiation detectors in the auxiliary building exhaust duct are used to detect high radiation level. The one gpm leak rate is suffi-ciently high to permit prompt detection and to allow for reason-able leakage through the pump seals end valve packings, and yet small enough to be readily handled by the pumps and radioactive waste system. Leakage to the safety injection system sumpswillbereturnedtothespentregeneranttanks.(gumproom ) Addi-tional makeup water to the containment sump inventory can be readily accommodated via the charging pumps from either the SIRW tank or the concentrated boric acid storage tanks.

In case of failure to meet the acceptance criteria for leakage frcm the shutdown cooling system or the associated components, it may be possible to effect repairs within a short time. If so, it is considered unnecessary and unjustified to shutdown the reactor.

The times allowed for repairs are consistent with the times developed for other engineered safeguards components.

References (1) USAR, Section 9.3 (2) USAR, Section 6.2 3-85 Amendment No. 97

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