ML20248C077
| ML20248C077 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/02/1989 |
| From: | Bournia A Office of Nuclear Reactor Regulation |
| To: | Morris K OMAHA PUBLIC POWER DISTRICT |
| References | |
| TAC-72703, NUDOCS 8906090252 | |
| Download: ML20248C077 (2) | |
Text
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June 2, 1989 Docket No. 50-285 f, Mr. Kenneth J. Morris Division Manager - Nuclear Operations l
Omaha Public Power District
-1623 Harney Street Omaha, Nebraska 68102
Dear Mr. Morris:
SUBJECT:
FORT CALHOUN STATION, UNIT NO.1 - AMENDMENT NO.122 TO FACILITY OPERATING LICENSE NO. DPR-40 (TAC NO. 72703)
The Commission has issued the enclosed Amendment No.122 to facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. This amendment consists of changes to the Technical Specifications in response to your application dated January 6,1989 as supplemented on February 28, 1989.
The amendment modifies the Technical Specifications to: (1)extendthesurveillance interval by 25 percent, but the total interval for three consecutive intervals shall not exceed 3.25 times the specific interval, (2) define the regular surveillance intervals, (3) include the 25 percent extension to all applicable codes and standards referenced within, (4) delay an action statement for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowabic outage time limit of the action requirement is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and (5) eliminate the need to perform surveillance on inoperable equipment.
A copy of our related Safety Evaluation is enclosed. The notice of issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/s/
Anthony Bournia, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amendment No.122 to DPR-40
- 2. Safety Evaluation l I
cc w/ enclosures:
See next page DISTRIBUTION:
- Docket File BGrimes NRC PDR TMeek (4)
Local PDR Wanda Jones PD4 Reading JCalvo PNoonan ACRS(10) ABournia(2) GPA/PA FHcbdon OGC-Rockville DHagan I ARM /LFMB EJordan Plant File DOCUMENT NAME: FC AMENDMENT
- See previous concurrences: [8!
PD4/LA* PD4/FM* OTSB* OGC-Rockville* PD4/D)1 PNocoan ABournia:bj JCalvo SHLewis FHebdon' f 05/23/89 05/23/89 05/24/89 05/31/89 0E'/ 2-/89 ,
($9060kOL& Y ,ea <-
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d Mr. Kenneth J. Morris Fort Calhoun Station Omahe Public Power District Unit No. I cc:
Harry H.. Voigt , Esq.
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampsh rt ' venue, NW Washir.gton, D.C. 20036 nr. .h ek Jensen, Chairman Vr.eu-;1on County Board cf Supervisors Blair, Nebraska 68008 Mr. Phillip Harrell, Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 309 Fort Calhoun, Nebraska 68023 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, MD 20852 Regional Administrator, Region IV .
U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Harold Borchert, Director Division of Radiological Health Department of Health 301 Centennial Mall, South .
P.O. Box 95007 Lincoln, Nebraska 68509 -
W. G. Gates, Manager Fort Calhoun Station P. O. Box 399 Fort Calhoun, Nebraska 68023 s
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fp* %q'c, UNITED STATES E .ec() NUCLE AR REGULATORY COMMISSION 7 W ASHINGTON D. C. 20555
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OMAHA PUBLIC POL'ER DISTRICT DOCKET NO. 50-285 F0 fit CALHOUN STATION, UNIT NO. I
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AMENDMENT TO FACILITY OPERATING LICENSE Arnendment No.122 License No. DPR-40
- 1. Thc Nuclear Regulatory Commission (the Corrission) has found that:
A. The application for acendment by the Omaha Public Power District (the licensee) dated January 6,1989 as supplemented on February 28, 1989, complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the rules
, and the Cornission's Atoraic end regulatters set forth in 10 CFR Chapter I; B. The facility will operate in conformity with' the appi-ication, as amended, the provisions of the Act, and the rules and regulations j of the Comissior; 1 C. There is reasonable assurance: (1)thattheactivitiesauthorized by this arendment can be conducted without endangering the health and safety cf the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuarce of this license anendment will not be inimical to the l common defense and security or to the health and safety of the public; .
4 and I
E. The issuance of this amendment is in accordance with 10 CFR Part 51 !
of the Corr 1ssion's regulations and all applicable requirements have ,
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been satisfied.
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- 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.122 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. Thelicenseamendmentiseffectiveafter90daysfromthedateofissuanie.
FOR THE NUCLEAR REGULATORY COMMISSION
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Ab:. . h Frederick J. He on, Director Project Directorate - IV Division of Re. actor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 2, 1989
ATTACHMENT TO LICENSE AMENDMENT NO.122 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 ifnes indicating the area of change..
Remove Pages Insert Pages 11 11 3-0a 3-Ob 3-Oc 3-1 3-1
- 3-6 3-6 3-12a 3-12a 3-16c 3-16c 3-17 3-17 3-63 3-63 3-69 3-69 3-70 3-70 3-71 3-71* .
3-72 3-72 3-73 3-73 3-76 3-76 3-80 3-80 3-84 3-84
l TABLEOFCONTENTS(Continued)
Pag i' 2.12 Control Room Systems..................................... 2-59 l 2.13 Nuclear Detector Cooling System.......................... 2-60 2.14 Engineered Safety Features System Initiation Instrumentation Settings............................... 2-61 2.15 Instrumentation and Control Systems.... ................. 2-65 2.16 River Leve1.............................................. 2-71 2.17 Miscellaneous Radioactive Material Sources............... 2-72 2.18 Shoc k Su ppress ors (Snubbers). . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-73 2.19 Fire Protection System................................... 2-89 2.20 Steam Generator Coolant Radioactive ty. . . . . . . . . . . . . . . . . . . . 2-96 2.21 Post-Accident Monitoring Instrumentation................. 2-97 2.22 Toxic Gas Monitors....................................... 2-99
. 3.0 S URVEI L L ANC E REQ UI REMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-0a- . l 3.1
-3.2 Instrumentation and Contro1.............................. 3-1 Eq u i pme n t a n d S a mp l i n g Te s t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-17 3.3 Reactor Cociant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance........ 3-21 3.4 Rea ctor Coolant System Integrity Testing. . . . . . . . . . . . . . . . . 3-36 3.5 Containment Test......................................... '3-37 3.6 Safety Injection and Containment Cooling -
Systems Tests.......................................... 3-54 3.7 Emergen cy Powe r System Periodic Tests. . . . . . . . . . . . . . . . . . . . 3-58 3.8 Ma in Steam I so la tion Va 1ves. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-61 3.9 Aux i l ia ry Fe edwa te r Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-62 3.10 Rea c t o r Co re Pa ra me ters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-63 3.11 Radiological Environmental Monitoring Programs........... 3-64 3.12 Radiological Waste Sampling and Monitoring............... 3-69 3.12.1 L iquid a nd Ga seous Effluents. . . . . . .'. . . . . . . . . . . . . 3-69 3.12.2 Solid Radioactive Waste......................... 3-71a 3.13 Radioactive Material Sources Surve111ance................ 3-76 3.14 Shoc k Suppres sors (Snubbers ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-77 3.15 Fire Protection System................................... 3-80 3.16 Recirculation Heat Removal System Integrity Testin 3-84 3.17 Steam Generator Tubes.............................g...... ....... 3-86 4.0 DESIGN FEATURES................................................ 4-1 4.1 Site..................................................... 4-1 4.2 Containment Design Features.............................. 4-1 4.2.1 Containment Structure........................... 4-1 4.2.2 Penetrations.................................... 4-1 4.2.3 Containment Structure Cooling Systems. . . . . . . .. . . 4-2 11 Amendrent No. 5#,86,77.19(, 122
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. i 3.0 SURVEILLANCE RE0VIREMENTS BASIS-Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Surveillance Requirements. These requirements are uirements stated in the Code of Federal based on the Surveillance Reo(3):
Regulations,10CFR50.35(c)
" Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting condition of operation will be met."
Specification 3.0.1 established the conditions under which the specified time interval for Surveillance Requirements say be extended. Item a.
permits er allowable extension of the nomal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient cerditions or other ongoing surveillance or maintenance activities. Item b. limits the use of the provisions of item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified. The limits of Specification 3.0.1 are based on engineering judgement and the recognition that the sost probable result of any particular surveillance; being performed is the verification of conforreance with the Surveillance Requirements. These provisions are sufficient to ensure that the reliability demonstrated through actual surveillance activities is not significantly degraded
, beyond that obtained from the specified surveillance interval.
The provisions of $ specification 3.0.2 define the surveillance intervals for use in the Technical Specifications. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications. A few surveillance requirements have uncemon intervals, for example, Table 3-9 requires sampling of fish once per season. In such a case the surveillance interval shall be perfortned as defined by the individual specifications.
Specification 3.0.3 extends the testing interval required by codes and standards referenced by the Technical Specifications. This clarification is provided to remove any ambiguities relative to the frequencies for perfoming the required inservice inspection and l testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the codes and standards referenced therein.
Specification 3.0.4 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, as defined by the provisions of Specifications 3.0.1 and 3.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for the corresponding Limiting Condition for Operation. Under the provisions 3-Ob bendment No. 122
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9 3.0 SURVEILLANCE REQUIREMENTS Basis (continued) of this specification, systems and components are assumed to be OPERABLE when Surveillance Requirements have been satisfactorily performed within the specified time interval. However, nothing in.
~ this provision is to be construed as implying that systems or components are OPERAELE when they are found or known to be inoperable
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even though the Surveillance Requirements are met. This specifica-tion also clarifies that.the ACTION requirements are applicable when Surveillance Requirements have not been completed within the allowed surveillance interval and that the time limits of the ACTION requirements apply from the point in time it is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance Requirement within the allowable outage time limits of
, the ACTION requirements restores compliance with the requirements of Specification 3.0.4. However,'this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification of 3.0.1, was a violation of the OPERABILITY requirements of a Limiting Condition for Operation that is subject to enforcement action.
Further, the failure to perform a surveillance within the provisions of Specification of 3.0.1 is a violation of a Technical Specification requirement and is, therefore, a reportable ev.ent under the require-ments of 10 CFR 50.73(a)(2)(1)(B) because it is a condition prohibli:ed by the plant's Technical Specifications.
If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or a shutdown is required to cosply with the ACTION requirements, e.g., Specification 2.0.1, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance is provided to permit a delay in implementing the ACTION requirements.
This provides an adequate time limit to complete Surveillance Require-ments that have not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown is required to comply with ACTION requirements or before other remedial measures would be required that may preclude completion of a surveillance.
The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in completing the required surveillance. If a surveillance is not completed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance, the time limits of the ACTION requirements are applicable at this time. When a surveillance is performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance and the Surveillance Requirements are not met, the time limits of the ACTION requirements are applicable at the time that the surveillance is terminated.
Surveillance Requirements do not have to be perfonned on inoperable equipment because the ACTION requirements define the remedial measures that apply. However, the Surveillance Requirements must be met to demonstrate that inoperable equipment has been restored to operable ,
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3-Oc Amendment No.122
3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control Applicability Applies to the reactor protective system and other critical instru- I mentation and controls.
Objective To specify the minimum frequency and type of surveillance to be applied to critical plant instrumentation and controls.
Specifications Calibration, testing and checking of instrument channels, reactor
' protective syster ard engineered safeguards system logic channels and miscellaneous instrument systems and controls shall be performed as specified in Tables 3-1 to 3-3.
Basis Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in " upscale" or "downscale" indir.a-tion can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such' failures.are, in many cases, revealed by alarm or annunciator action and a check supple-ments this type of built-in surveillance.
Based on the District's experience in operation of conventional power plants and on reported nuclear plant experience, a checking frequency of once-per-shif t is deered adequate for reactor and(
steam system instrumentation. Calibrations are performed to ensure the presentation and acquisition of accurate information.
The power range safety channels are calibrated daily against a calorimetric balance standard to acccunt for errors induced by l changing rod patterns and core physics parameters.
I Other channels, subject only to the " drift" errors, can be expected to remain within acceptable tolerances if recalibration is performed at each refueling shutdown interval.
l 3-1 Amendment No. 9,122
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3.0 SURVEILLANCE REQUIRE!!ENTS 3.2 Eccipment and Samplino Tests Applicability Applies to plant equipment and conditions related to safety.
Objective To specify the minimum frequency er.d type of surveillance to be applied to critical plant equipment and conditions.
Specifications Equipment and sampling tests shall be conducted as specified in Tables 3-4 and 3-5.
Basis The equipment testing and system sampling frequencies specified in Tables 3-4 and 3-5 are considered adequate, based upon experience, to maintain the status of the equipment and systems so as to assure safe operation. Thus, those systems where changes might occur relatively rapidly are sampled frequently and those static systems not subject to changes are sarpled less frequently. , !
The control room air treatment system consists of high efficiency particulate air filters (HEPA) and the charcoal adsorbers. HEPA filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. Tht charcoal adsorbers are installed to reduce the potential intake of iodine to the control room. The in-place test results will confirm system integrity and performance. The laboratory carbon sample tests results should indicate methyl iodide removal efficiency of at least 90 percent for expected accident conditions.
The spent fuel storage-decontamination areas air treatment system is designed to filter the building atmosphere to the auxiliary building vent during refueling operations. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. In-place testing is performed to confirm the integrity of the filter system. The charcoal adsorbers are periodically sampled to insure capability for the removal of radioactive iodine.
3-17 Amendment No. 15, 57, 122
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4 3.0 SURVEILLANCE REQUIREMENTS , 3.12 F.adiolocical Weste Sampling and Monitoring '
3.12.1 Liquid and Gaseous Effluents Applicability !
Applies to the sampling, monitoring, and testing used for liquid and gasecus effluents.
Objective To ensure that radioactive liquid and gaseous releases from the facility are maintained as low as reasonably achievable and within the limits'specified by Specifications 2.9.1(1) and 2.9.1(2).
Specifications (1) Liquid Effluer.ts
- a. Radioactive liquid waste sampling and activity analyses shall be performed in accordance with Table 3-11. The results of these analyses shall be used with the calculational methods in the ODCH to assure that the concentration at the point of releaseislimitedtothevaluesinSpecification2.9.1(1)a.
- b. Prior to release of each batch of li. quid effluent, the batch shall be mixed, sampled, and analyzed for principal gasuna emitters. When operational or other limitations preclude specific gama radionuclides analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive materials released in the batch, and a weekly sample composited from proportional aliquets from each batch released during the week shall be analyzed for the principal gama-emitting radionuclides.
- c. The overboard header radiation monitor shall have a:
(1) Source check prior to any release of radioactive materials from the monitor or the hotel waste tanks.
(ii) Quarterly channel functional test.
(iii) Channel calibration at refueling frequency.
- d. The steam generator blowdown radiation monitors shall have:
(i) Daily channel checks.
(ii) Monthly source checks.
3-69 Amendment h'o. 28, Sf,122
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3.0 SURVEILLANCE REQUIREMENTS 3.10 Reactor Core Parameters Applicability
' pplies A to reactor core parameters that affect shutdown margin, MTC, linear heat rate and DNB margin.
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Objective To require evaluation of reactor core parameters.
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Specification (1) Shutdown Margin
- a. The shutdown margin shall be determined:
- 1. By verifying that the CEA group withdrawal is above the Transient Insertion Limits of Specification 2.10.2 when-ever the reactor is in hot standby or power operation conditions at least once per shift, or
- 2. By considering the following factors whenever the re-actor-is in hot or cold shutdown at least once per day.
(i) Reactor coolant system boron con' centration; (ii) CEA position; (iii) Reactor coolant system temperature; (iv) Fuel burnup; (v) Xenon concentration; and (vi) Samarium concentration.
- b. The overall core reactivity balance shall be compared to pre-dicted values to demonstrate agreement with i 1.0% Ak/k at least once per 31 EFPD. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core con- !
ditions prior to exceeding a cycle burnup of 2000 MWD /MTU '
after each refueling.
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3-63 Amendment No. If, 26, 32, 122 ,
't 3.0 SUPNEILLANCE REQUIREMENTS 3.12 Radiological Waste Sampling and Monitoring (Continued) 3.12.2 Licuid and Gaseous Effluents (Continued)
(iii) Quarterly channel functional tests.
(iv) Channel calibration at refueling frequency. l
- e. The steam generator blowdown effluent flow rate will be calibrated at refueling frequency and visually determined I operable daily.
- f. Records shall te reintained of the radioactive concentrations and volume before dilution of each batch of liquid effluent released and of the average dilution flow and length of time over which each discharge occurred. Analytical results shall l be. submitted to the Commission in accordance with Section
. 5.9.4.a of these specifications.
(2) Gaseous Effluents
- a. Radioactive gaseous waste sampling and activity analyses shall
. be performed in accordance with Table 3-12. The results of these analyses shall be used with the calculational methods in the ODCM to assure that the concentration of. radioactive materials in unrestricted areas is limited to the values in Specification 2.9.1(2)a. *
- b. (1) A ventilation stack radiation monitor shall have a source check prior to any release of radioactive materials from a gas decay tank or the containment. A monthly source check will be performed during refueling outages if a purge or gas decay tank release is not done during that month.
(ii) Each ventilation stack monitor shall have a quarterly channel functional test.
(iii) Each ventilation stack monitor shall be calibrated at refueling frequency. l (iv) The ventilation stack flow rate will be calibrated and functionally tested at refueling frequenqy. The stack l-radiation monitor flow rate will be calibrated and functionally tested at refueling frequency. Both will I be determined operable by visual inspection daily.
- c. The condenser air ejector monitor shall have a:
(1) Daily channel check.
(ii) Monthly source check.
3-70 Amendment No. Si,122 a- - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _. ___ _
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'2 ' O SURVEILLANCE REQUIREMENTS 3.12 FediologicalWastesemplingandMonitoring(Continued) 3.12.3 Liquia anc Gaseous Effluents (Continued)
(iii) Quarterly channel functional test.
(iv) Channel calibration at refueling frequency. l
(i) Daily channel check (when in service).
(ii) Honthly cross comparison with a grab sample.
(iii) Quarterly channel calibration using gas mixtures with concentrations in the range of interest.
- e. Records shall be maintained and reports of the sampling and results of analyses shall be submitted to the Comission in accordance with Section 5.9.4.a of these specifications.
Basis The surveillance requirements given under Specification 3.12.1(2) provide assurance that radioactive gaseous effluents from the station are properly controlled and monitored over the life of the station in conformance with the requirements of General Design Criteria 60,and 64 of 10 CFR Part 50, Appendix A. These surveillance requirements provide the data for the licensee and the Comission to evaluate the performance of the station relative to radioactive gaseous wastes released to the environment. The existing minimum sensitivity of airborne effluent monitor RM-062 is SE-06 mci /cc/100 cpm and this minfrum sensitivity shall be maintained if the monitor is replaced. Reports on the quantities of the radioactive materielt released in gaseous effluents shall be furnished to the Comis-sion on the basis of Section 5.9.4.a of these Technical Specifications.
On the basis of such reports and any additional information the Comission may obtain from the licensee or others, the Comission may from time to time require the licensee to take such action as the Comission deems appropriate.
The surveillance requirements given under Specification 3.12.1(1) provide assurance that liquid wastes are properly controlled and monitored in conformance with the requirements of General Design Criteria 60 and 64 of 10 CFR Part 50, Appendix A, during any planned release of radioactive materials in liquid effluents. These surveillance requirements provide the data for the licensee and the Comission to evaluate the station's performance relative to radioactive liquid wastes released to the environ-ment. Reports on the quantities of radioactive materials released in liquid effluents shall be furnished to the Comission on the basis of Section 5.9.4.a of these Technical Specifications. On the basis of such reports and any additioral information the Commission may obtain from the licensee or others, the Comission may from time to time require the licensee to take such action as the Comission deems appropriate.
3-71 Amendment No. Sf, 195,122
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, TABLE 3-11 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS Ai Monitor & Hotel Waste Tanks Releases Lower Limit of Type of Detection (LLD)
Sampling Frequency Activity Analysis (4) (pCi/ml)
Each Batch Principal Gamma Emitters (2)(5) 5.0 E-07 I-131(2) 1.0 E-06 Monthly From One Dissolved Noble Gases (2) 1.0 E-05 Batch (Gamma Emitters)
Monthly Composite II) H-3 1.0 E-05 t
! aross a 1.0 E-07 Quarterly Composite (1) Sr-89, Sr-90 5.0 E-08 B.* Steam Generator Blowdown tower Limit of Type of Detection (LLD)
Sampling Frequency Activity Analysis (4) (pCi/ml) eekly CompositeII) Principal Gamma Emitters (5) 5.0 E-07 I-131(6) 1.0 E-06 Weekly Dose Equivalent I-131 1.0 E-06 (Gamma Emitters)
Nonthly Dissolved Noble Gases 1.0 E-05 Monthly Composite (1) H-3 1. 0 E-05 Gross a 1.0 E-07 Quarterly Composite (1) St-B9, Sr-90 5.0 E-08 NOTES:
(1) To be representative of the average quantities and concentrations of radioactive materials in liquid effluents, samples should be collected in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite should be mixed in order for the composite sample to be representative of the average effluent release.
3-72 Amendment No. 28, 86, 122
1 TABLE 3-11 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS (Continued)
NOTES:
(2) Or gross radioactivity as described in Specification 3.12.1(1)b.
(3) When steam generator iodine activity exceeds 50 percent of limits in Specification 2.20, the sampling and analysis frequency shall be increased to a m:inimum of five times per week. When the steam generator iodine activity exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.
(4) The lower limit of detection (LLD) is defined in the ODCM based on MUREG 0472, Rev. 3.
(5) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144.
- (6) A weekly grab sample and analyses program including gamma isotopic identification will be initiated for the turbine building sump effluent when the stear generator blowdown water composite analysis indicates the I-131 concentration is greater than 1.0 E-06 sicrocurie/ milliliter.
l 3-73 Amendment No. 28,56,122
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3.0 SURVEI! LAliCE REQUIREMENTS 2.12 RADIOACTIVE MATERIAL SOURCES SURVEILLANCE Applicability Applies to leakage testing of byproduct, source, and special nuclear radioactive material sources.
l Objective To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.
L Specification Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the NRC or an agreement State, as follows:
- 1. Each sealed source, except startup sources subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at intervals of six months. l
- 2. The periodic leak test required does not ' apply to sealed sources that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or l transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
- 3. Startup sources shall be leak tested prior to and following any repair or maintenance and before being subjected to core flux.
3-76 Amendment No. II,122
3.0 SURVEIt. LANCE REQUIREMENTS 1
3.15 Fire Protection' System !
Applicability Applies to fire' detection and fire extinguishing subsystems in nuclear safety related areas and other areas which may impact on safety related systems.
l Objective 1 To' ensure related systems. the operability of the fire protection system in nuclear safety j
Specifications (1) Each fire detector be demonstrated listed in Table 2-7 and in containment shall operable:
- a. At least once per 6 months by performance of a channel functicnal test and a test of the supervision circuitry.
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- b. Testing interval for fire detectors which are inaccessible due to high radiation or require an equipment alignment not used in power operation may be extended until such time as the detectors become accessible for a minimum of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. However, the shuttiown need not be extended solely for the purpose of this testing. Such detectors shall be functionally tested at a maximum interval of once per refueling cycle.
(2) The fire suppression water system shall be demonstrated operable:
a.
At least once per month by starting each pump and operating it for at least 15 minutes.
- b. At least once per month by verifying that each valve in the flow path is in its correct position.
- c. At least once per 12 months by cycling each testable valve (those which can be cycled without endangering the safety of equipment) in the flow path through at least one complete cycle of full travel.
- d. At least once per 18 months by performing a system func-tional test which includes:
- 1. Verifying that each pump develops at least 1800 gpm at a system head of 260 feet.
- 2. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and 3-80 Amendrent No. (p, 122
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- _# f* **G%je. UNITED STATES j .. .s< i NUCLEAR REGULATORY COMMISSION s WASHINGTON, D. C. 20555
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SAFETY EVALUATICK PY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.122 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOL't.' STATION, UNIT NO.1 DOCKET NO. 50-285 1.0 ' INTRODUCTION By letters dated January 6,1989 and as supplemented on February 28, 1989, Oraha Public Power District (OPPD) submitted an application for an amendeent Station,UnitNo.I,TechnicalSpecifications(TS)to:tofacilityOperatingLicenseNo.DPR-40thatw (1 extend the surveil-lance interval by 25 percent, but the total interval for three consecutive intervals shall not exceed 3.25 times the specific interval, (2) define the regular surveillance intervals, (3) include the 25 percent extension applicable to all codes and standards referenced within, (4) delay an action statement for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limit of the action requirement is less than 24 hou'rs, and (5) eliminate the need to perfom surveillance on inoperable equipment.
2.0 DISCUSSION The NRC staff issced the Generic Letter (GL) 87-09, dated June 4, 1987,
" Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) on the appifcability of Limiting Conditions for Operation and Surveillance Require-ments," to provide guidance for .;hort-term imprever4nts to, resolve iMEdiate concerns that have been identified in investigations of TS problems. In GL E7-09, the NRC staff encourage the utilities to incorporate these changes; -
however, these changes would only be on a voluntary basis. .
Consequently, OPPD incorporated scoe of these modifications in the Fort Calhoun Station TS to bring them closer in line to the Standard TS with respect to the applicability of surveillance requirements.
All the changes requested by OPPD are changes that are addressed by GL 87-09, and therefore, the staff finds these changes acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
The amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has detemined that the amendment involves no significant increase in the amounts, and ne significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual er 3
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cumulative cceupational radiation exposures. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public connent on such finding' Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forthin10CFRSection51.22(c)(9).Pursuantto10CFR51.22(b),noenvironmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
4.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will nct be endangered by operation in the proposed manner, ar.d such (2) public
, ectivities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:. June 2, 1989 Pr'incipal Contributors: A. Bournia e
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