ML20064L290

Revised Proposed Changes to Tech Specs Lowering Operating Limit for Radioactive Iodine in RCS to 1.0 Uci/Gm

ML20064L290
Person / Time
Site:	Fort Calhoun
Issue date:	02/03/1983
From:	OMAHA PUBLIC POWER DISTRICT
To:
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ML20064L279	List: ML20064L277 ML20064L290
References
NUDOCS 8302150191
Download: ML20064L290 (15)

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TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page DEFINITIONS ............................................................ 1 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM ...................... 1-1 1.1 Safety Limits - Reactor Core ........................... 1-1 1.2 Safety Limit, Reactor Coolant System Pressure .......... 1-4 1.3 Limiting Safety Systam Settings, Reactor Protective System ................................................. 1-6 2.0 LIMITING CONDITIONS OF OPERATION .............................. 2-0 2.0.1 G e n e ra l Re q u i reme nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-0 2.1 Reacto r Cool ant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.1 Operable Components ............................ 2-1 2.1.2 Heatup and Cool down Rate . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.1.3 Reactor Cool ant Radi oacti vity . . . . . . . . . . . . . . . . . . 2-8 l 2.1.4 Reactor Coolant System Leakage Limits .......... 2-1 2.1.5 Maximum Reactor Coolant Oxygen and Halogens Concentrations.................................. 2-13 2.1.6 Pressurizer and Steam System Safety Valves ..... 2-15 2.1.7 Pressuri zer Operabil ity . . . . . . . . . . . . . . . . . . . . . . . . 2-16a 2.2 Chemical and Vol ume Control System . . . . . . . . . . . . . . . . . . . . . 2-17 2.3 Emergency Core Cooling System .......................... 2-20 2.4 Containment Cooling .................................... 2-24 2.5 Steam and Feedwater Systems ............................ 2-28 2.6 containment System ..................................... 2-30 2.7 E l ec t ri c al Sy s t em s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-32 2.8 Refuel i n g Ope rat i ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-37 2.9 Radioactive Materi al s Release . . . . . . . . . . . . . . . . . . . . . . . . . . 2-40 2.10 Reactor Core ........................................... 2-48 2.10.1 Minimum Conditions for Criticality ............. 2-48 2.10.2 Reactivity Control System and Core Physics Parameter Limits................................ 2-50 2.10.3 In-Core Instrumentation ........................ 2-54 2.10.4 Power Di st ri bution Limits . . . . . . . . . . . . . . . . . . . . . . 2-56 2.11 Containment Building and Fuel Storage Building Crane ... 2- 58 2.12 Con t rol Room Sy st ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-59 2.13 Nucl ear Det ector Cool ing 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 Level ............................................ 2-71 2.17 Miscellaneous Radioactive Material Sources . . . . . . . . . . . . . 2-72 2.18 Shock Suppressors (Snubbers)............................ 2-73 2.19 F i re P ro t e c t i o n Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-89 2.20 Steam Generator Coolant Radioactivity................... 2-96 l

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Amendment No. 32,38 52,54,57 i Attachment A

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8302150191 830203 PDRADOCK05000g8 P

DEFINITIONS Azimuthal Power Tilt - To Azimuthal Power Tilt shall be the maximum difference between the power generated li, any core quandrant (upper or lower) and the average power of all quandrants in that axial half (upper or lower) of the core divided by the average power of all quandrants in that axial half (upper or lower) or the core.

Unrodded Planar Radial Peaking Factor - Fry The unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power censity of the individual ruel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, Tq .

Unrodded Integrated Radial Peaking Factor - Fn The unrodded Integrated Radial Peaking Factor is the ratio of the peak pin power to the average pin power in an unrodded core, excluding azimuthal tilt, T q.

Fire Suppression Water System The fire suppression water system consists of fire pumps and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, and the first valve ahead of the water flow alann device on each sprinkler, hose standpipe or spray system riser.

Dose Equivalent I-131 That concentration of I-131 (pCi/gm) which alone would produce the same

thyroid dose as the quantity <ind isotopic mixture of I-131, I-132, I-133,

,tendment No. 32',38

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1-134 and I-135 actually present. In other words, Dose Equivalent I-131 (uCi/gm) = pCi/gm of I-131

+ 0.0361 x pCi/gm of I-132

+ 0.270 x pCi/gm of I-133

+ 0.0169 x pCi/gm of I-134 i + 0.0838 x pCi/gm of I-135 li-AverageDisintegrationEnergy IIistheaverage(weightedinproportiontotheconcentrationofeach radionuclide in the reactor coolant at the time of sampling) of the sum ef the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.3 Reactor Coolant Radioactivity Applicability Applies to the radioactivity of the reactor coolant.

Objective To ensure that the reactor coolant radioactivity is maintained at a level commensurate with the occupational and public safety.

Specification (1) The radioactivity of the reactor coolant shall be limited to:

a. I 1.0 pCi/gm DOSE EQUIVALENT I-131, and

b. I 100/E pCi/gm (2) With the radioactivity of the reactor coolant >1.0 pCi/gm DOSE EQUIVALENT I-131 but < 60 pCi/gm, operation, to include restart if shutdown, may con-tinue for up to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> cumulative operating time under these circum-stances, not to exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecutive 12 month period.

(3) With the radioactivity of the reactor coolant >1.0 pCi/gm DOSE EQUIVALENT I-131 for more than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during one continuous time interval or exceeding 60 pCi/gm, be in at least H0T SHUTDOWN with Tavg <536 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) With the radioactivity of the reactor coolant > 100/E pCi/gm, be in at least HOT SHUTDOWN with Tavg < 536 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(5) With the radioactivity of the reactor coolant >1.0 pCi/gm DOSE EQUIVALENT I-131, perform the sampling and analysis requirements of items 1.(a)(2)(ii) and 1.(b)(2)(1) of Table 3-4 until the radioactivity of the reactor coolant is restored to within its limits. A REPORTABLE OCCURRENCE, pursuant to Specification 5.9.2, shall be submitted to the Commission. This report shall contain the results of the radioactivity analyses together with the following information:

a. Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded.

Amendment 28 2-8 ,

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.3 Reactor Coolant Radioactivity (Continued)

b. Purification System flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded.

c. The time duration when the radioactivity of the reactor coolant exceeded 1.0 uCi/gm DOSE EQUIVALENT I-131.

Basis

.The limitations on the radioactivity of the reactor coolant ensure that the resulting 2-hour doses at the site boundary will be well within the limits of 10 CFR Part 100 following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite power.

"ermitting power operation to continue for limited time periods with the reactor coolant's radioactivity levels >1.0 pCi/gm DOSE EQUIVALENT I-131, but < 60 pCi/gm, accommodates possible iodine spiking phenomenon which may occur following changes in thermal power.

Reducing Tavg to < 536 F prevents the release of radioactivity should d steam generator tube rupture, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive radioactivity levels in the reactor coolant will be detected in sufficient time to take appropriate corrective action (s).

References (1) USAR, Section 11.11.3 (2) USAR, Section 14.14 Ainendment 60 2-9

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DELETE I

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Amendment 28 2-10

2.0 LIMITING CONDITIONS FOR OPERATION 2.20 Steam Generator Coolant Radioactivity Applicability Applies to the radioactivity of the steam generator coolant.

Objective To ensure that the steam generator coolant radioactivity .s maintained at a level commensurate with the occupational and public stfety.

Specificaticn, (1) The radicactivity of the steam generator coolant shall be < 0.10

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pCi/gm DOSE EQUIVALENT I-131.

(2) With the radioactivity of the steam generator coolant > 0.10 pCi/gm DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Basis The limitations on the steam generator coolant's radioactivity ensure that the resultant off-site doses will be well within the limits of 10 CFR Part 100 in the event of a steam line break. This dose also includes the effects of a coincident 1.0 GPM primary-to-secondary tube leak in the steam generator of the affected steam line and a concurrent loss of off-site power.

References (1) USAR Section 14.12.

2-96

3.0 SURVEILLANCE REQUIREMEliTS 3.2 Equipent and Sampling Tests Applicability Applies to plant equipment and conditions related to safety.

Objective To specify the minimum frequency and 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. The specified intervals may be adjusted to accommodate normal test schedules except that the interval shall not exceed 1.25 times the specified interval.

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 sampled less frequently.

l 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. The charcoal adsorbers are installed to reduce the potential inttke 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 envi ronr.ent. 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 radioactivity iodine.

Amendment No. 15 3-17

3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (Continued) j The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts.

This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drop across the combined HEPA filters and charcoal adsorbers, for each of the air treatment systems, of less than 6 inches of water will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability and remove excessive moisture build-up on the adsorbers.

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorber: could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

L References USAR, Section 9.10 Amendment No. 15 3-17a

TABLE 3-4 MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frequency

1. Reactor Coolant (a) Power Operation (1) Gross Radioactivity 1 per 3 days (2) Isotopic Analysis for (i) 1 per 14 days DOSE EQUIVALENT I-131 (ii) 1 per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (l) whenever the radio-activity exceeds 1.0 pCi/gm DOSE EQUIVALENT I-131.

(iii) 1 sample between 2-8 hours following a thermal power change exceeding 15%

of the rated thermal power within a 1-hour period.

(3) li Determination 1 per 6 months (2)

(4) Dissolved oxygen 1 per 3 days and chloride (b) Hot Standby (1) Gross Radioactivity 1 per 3 days Hot Shutdown (2) Isotopic Analysis for (i) 1 per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (l)

DOSE EQUIVALENT I-131 whenever the radio-activity exceeds 1.0 pCi/gm DOSE EQUIVALENT I-131.

(ii) 1 sample between 2-8 hours following a thermal power change exceeding 15% of the rated thermal power within a 1-hour period.

(3) Dissolved oxygen 1 per 3 days and chloride Amendment No. 28 3-18

i TABLE 3-4 (Continued)

MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frequency

1. Reactor Coolant (Continued)

(c) Cold Shutdown (1) Chloride 1 per 3 days (d) Refueling (1) Chloride 1 per 3 days Operation (2) Boron Concentration 1 per 3 days

2. Steam Generator Isotopic Analysis for DOSE 1 per 7 days Coolant EQUIVALENT I-131

3. SIRW Tank Boron Concentration 1 per 31 days

4. Concentrated Boric Boron Concentration 1 per 31 days Acid Tanks

5. SI Tanks Baron Concentration 1 per 31 days

6. Spent Fuel Pool Boron Concentration 1 per 31 days (1) Until the radioactivity of the reactor coolant is restored to j:,1 pCi/gm DOSE EQUIVALENT I-131.

(2) Sample to be taken after a minimt:m of 2 EFPD and 20 days of power operation have elapsed since reactor was subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

t Amendment No. 28 3-19

DELETE Amendment No. 28 3-19a

DISCUSSION OF PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed revisions to the Fort Calhoun Station Unit No. 1 Technical Specifications are intended to provide the following functions:

1. Respond to the Commission's letter dated July 22, 1980, and

2. Incorporate Limiting Conditions for Operation during or following a power transient for which Section 2.1.3 of the present Technical Specifications does not have explicit provisions.

The proposed Technical Specifications provide reasonable assurance that following a steam generator tube rupture incident or a main steamline break in conjunction with as assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite power, the resulting doses at the site boundary will be well within the exposure guidelines of 10 CFR Part 100. At the same time, these Technical Specifications permit the operating flexibility, compatibility with con-siderations of health and safety of the public, under unusual conditions of operation, on a temporary basis. Conversely, these Technical Specifi-cations provide compliance with the limit specified in 10 CFR Part 20 under normal reactor operation.

It is concluded that based on the following reasons the proposed Technical Specifications da not involve an unreviewed safety question as per 10 CFR Part 50, Paragraph 50.59 (a)(2):

1. The proposed changes do not increase the probability or consequences of accidents or malfunction of safety-related equipment previously considered,

2. There is a reasonable assurance that the health and safety of the public will not be endangered under the proposed changes,

3. The possibility for an accident or malfunction of, a different type than previously evaluated is not created, and

4. The margin of safety as defined in the applicable Technical Specifications is not reduced.

A comparison of standard Technical Specifications and the proposed Technical Specifications attached to the Commission's letter dated July 22, 1980, is presented on the next page.

Attachment B

COMPARIS0N OF STANDARD TECHNICAL SPECIFICATIONS (STS)

AND THE PROPOSED TECHNICAL SPECIFICATIONS FOR FORT CALHOUN STATION UNIT N0. 1 Section or Section or Subsection Subsection of of Proposed STS Tech. Specifications Remarks 1.0 Definitions Appropriate / applicable definitions have been incorporated.

I. REACTOR COOLANT SYSTEM 3.4.9.a 2.1.3(1)a Incorporated.

3.4.9.b 2.1.3(1)b Incorporated.

Action a 2.1.3(2) 1. The proposed Technical Specification is considered conservative since the upper limit for DOSE EQUIVALENT I-131 during power transients (iodine spiking) is not allowed to exceed 60 pCi/gm.

Figure 3.4-1 of STS allows the radio-activity to exceed 60 pC1/gm whenever the reactor thermal power is less than 80%.

2. The specified time limit for reactor operation during iodine spiking is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> instead of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This 100-hour time limit has been obtained after reviewing the past 7 years operating history. It was determined that it takes approximately 100 to 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to restore the radioactivity within the acceptable values.

Section or Section or Subsection Subsection of of Proposed STS Tech. Specifications Remarks Action b 2.1.3(3) Incorporated Action c 2.1.3(4) Incorporated Action d 2.1.3(5) Incorporated Table 4.4-4 Item 1 Table 3-4, Items Incorporated 1(a)(1) and 1(b)(1)

Item 2 Table 3-4, Item Incorporated 1(a)(2)(i)

Item 3 Table 3-4, Item Incorporated 1(a)(3)

Item 4(a) Table 3-4, Items Based on the operating history of the 1(a)(2)(ii) and plant, sampling requirements once per 1(b)(2)(i) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are considered appropriate.

Based on the operating history of Item 4(b) Table 3-4, Items the plant and especially during iodine 1(a)(2)(iii),and spiking phenomenon, the sampling 1(b)(2)(ii) requirement of one sample between 2-8 hours is considered appropriate.

II. SECONDARY COOLANT SYSTEM 3.7.1.4 Propose.1 new Incorporated Specification 2.20(1)

Action 2.20(2) Incorporated Table 4.7-1 Iten 1 -

Not considered appropriate due to its implication / interaction with Item 2 of STS. Also, the determination of gross radioactivity does not have any bearing on the safety considerations following a main steam line break.

Item 2 Table 3-4, Item 2 The proposed sampling requirements are considered more limiting.

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