ML19339C482

Tech Specs Table of Contents,Definitions for Azimuthal Power Tilt,Unrodded Planar Radial Peaking Factor,Unrodded Integrated Radial Peaking Factor,Fire Suppression Water Sys & Dose Equivalent I-131 & 2.0 for Limiting Conditions

ML19339C482
Person / Time
Site:	Fort Calhoun
Issue date:	11/14/1980
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19339C473	List: ML19339C472 ML19339C476 ML19339C482
References
NUDOCS 8011180461
Download: ML19339C482 (16)

Text

__ .. _ - .-. .

i f

.e . -en.

,, ... .-n.n

- ,. ._ .. .~. .

n.,.Cn

. .. . .~. r I

TABLE OF CONTENTS i Pace D .v +.,.

. .r e. .-c..Le

. ............................................................. 2 1.0 SAFE'"Y LIMITS AND LIMITINO SAFETY SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 ' Safety Li=its - Reactor Core ............................... 1-1 1.2 Safety Li=it, Reacter Coolant Syste: Pressure .............. 1L j 1.3 Limiting Safety Systes Settings, Reactor Protection Syste . 1-6 2.0 LIMITING CONDITIONS FCR OPERATION ................................. 2-0 2.0.1 General Require =ents .............................. 2-0 2.1 Reactor Ccolant Syste: ..................................... 2-1 2.1.1 Cperable Co=ponents ............................... 2-1 2.1.2 Heatup and Cooldevn Rate .......................... 2-3 2.1.3 Reactor Coolant Radicactivity...................... 2-8 l l 2.1.L Reactor Coolant Syste= Leakage Li=its . . . . . . . . . . . . . 2-11 2.1.5 Maxi =u= Reactor Ccclant Oxygen and Halogens Concentrations ................................. 2-13 2.1.6 Pressurizer and Stes: Syste= Safety Valves ........ 2-15 ,

l 2. l'. T E NB M argin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16a 4

2.2 Chemical and Volu=e Control Syste .......................... 2 2.3 E=ergency Core Cooling Syste ............................... 2-20'

)

2.h Contain=ent Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 k 2.5 Stea= and Feedvater Syste=s ................................ 2-28 2.6 Containment Syste .......................................... 2-30 2.7 Electrical Systems ......................................... 2-32 2.S' Refueling Operations ....................................... 2-37 2.9 Radioactivs Materials Release .............................. 2 h0 2.10 Reactor Core ............................................... 2 h8

, 2.10.1 Mini === Ccnditions for Criticality ................ 2 h8 2.10.2 CIA and Fever Distribution Li=its . . . . . . . . . . . . . . . . . 2-50

, 2.10 3 In-Core Instrumentation ........................... 2-5h

, 2.10.k Moderator Te=perature Coefficient of Reactivity ... 2-56 2.11 Contair.=ent Building and Fuel Storage Building Crane ....... 2-58 2.12 Cont ro l Roc = Syst e=s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-59 I

2.13- Nuclear Detector Cooling.Syste ............................. 2-60 2.1L -Engineered Safety Features Syste= Initiatien Inst ru=entation Settings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-61 2.15  !=stru=entation and Control Syste=s ........................ 2-65 2.16 River Level ................................................ 2-71 2.17 Miscellaneous Radicactive Material Sources .................. 2-72 l 2.15 Shock Suppresscrs ( Snubbers ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-73 2.19 Fire Protecticn Syste ..................................... 2-8o Stea= Generator Coolant Radioactivity. . . . . . . . . . . . . . . . . . . . . . . 2-9k

~

2.20 j m

-Anen::ent No. ,[, J ,I)d, 52 i ATTACEMENT A

$01118g M l

DEFINITIONS Azimuthal' Power Tilt - To Azimuthal Power Tilt shall be the maximun difference between the power generated in any core quandrant (upper or lover) and the average power of l

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) of the core.

Unrodded Planar Radial Peaking Factor - Fry

• The unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, T . q Unrodded Integrated Radial Peaking Factor - FR The unrodded Integrated Radial Peaking Factor is the ratio of the peak ,

pin power to the average pin power in an unrodded core, excluding azi-muthal 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 alaen device on each sprinkler, hose standpipe or spray system riser.

Dose Eouivalent I-131 That concentration of I-131 (uci/gn) which alone would produce the same i

thyroid dose as. the quantity and isotopic mixture of I-131, I-132, I-133, f

4 A=endmentNo.)F[,38 7 1

i

\

} ..

_ . _ . - - . . . . -. . ~ . - . ._

i i 8 I

i

< I-13h and I-135 actually present. In other words, Dose Equivalent I-!31 (uCi/gs) = pCi/gm of I-131

+ 0.~0361 x uCi/gm of I-132

+ 0.270 x pCi/gm of I-133

+ 0.0169 x pCi/gm of I-134

+ 0.0838 x pCi/gm of I-135 i

l 5 - Average Disintegration Energy l 5 is the average (weighted in proportion to the concentration of each  !

I radionuclide in.the reactor coolant at the time of sampling) of the sum i i

j of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes 8 making up at least 95% of the total non-iodine radioactivity in the ,

i coolant.

References i

(1) FSAR, Section 7.2' (2) FSAR, Section T.3 c

I s

4 4

i i

i I

t 1

' Amendment No. , - 38 8 v , e w - , w - ~ e ~ v e .-v-,-g,+ +--,ew ,, ,a-- ----- e-- ., . ~--w

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.

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

! a. < 2.0 uCi/gm DOSE EQUIVALENT I-131, and

b. < 100/E uCi/gm (2) With the radioactivity of the reactor coolant > 2.0 pCi/gm DOSE EQUIVALENT I-131 but 5 60 uCi/gs, operation =ay continue for up to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during one continuous time interval.

(3) With the radioactivity of the reactor coolant > 2.0 uCi/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 uCi/gs, be in at least HOT SHUTDOWN with Tavg < 536 F vithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(h) With the radioactivity of the reactor coolant > 100/5 uCi/gs, be in at 3. east HOT SHUTDOWN with Tavg < 536 F vithin 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 > 2.0 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 h 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 h8 hours prior to the first samples in which the limit was exceeded.

l l

I Amendment 28 2-8 t l

l

. .- . ._ . ~

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 h8 hours prior to the first sample in which the limit was exceeded.

c. The time duration when the radioactivity of the reactor coolant exceeded 2.0 uCi/gn 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.

Permitting pcVer operation to continue for linited time periods with the reactor coolant's radioactivity levels > 2.. pCi/gn DOSE EQUIVALENT I-131, but 5 60 uCi/gs, acccmmodates possible iodine spiking phenomenon which may occur following changes e in thermal power.

Reducing Tayg to < 536 F prevents the release of radioactivity should a 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 radio-activity levels in the reactor coolant will be detected in sufficient time to take appropriate corrective action (s).

References (1) FSAR, Section 11.11.3 (2) FSAR, Section 14.14 Amendment 28 2-9 i

e e

J DELETE Amendment 28 2-10

2.0' LIMITING CONDITIONS FOR OPERATION 1

2.20 Steam Generator Coolant Radioactivity i Applicability i .

Applies to the radioactivity of the steam generator coolant.

Objective I

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

/

. Specification (1) The radioactivity of the steam generator coolant shall be

< 0.10 uCi/gm DOSE EQUIVALENT I-131.

i (2) With the radioactivity of the steam generator coolant l > 0.10 uCi/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 a

The limitations on the steam generator coolant's radioactivity ensure-that the resultant off-site doses vill 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) FSAR Section 14.12 3

'i 2-94

3.0 SURVEILLANCE REQUIRE E TS 3.2 Eauipment and Sampling Tests Applicability Applies to plant equipment and conditions related to safety.

Objective To specify the minimum frequency and type of surveillance-to j be applied to critical plant equipment and conditions.

Specifications Equipment and sampling tests shall be conducted as specified in Tables 3 h and 3-5 The specified intervals may be adjusted to accc:::modate 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 syste:rs so as to assure safe operation. Thus, those systems where 4 changes might occur relatively rapidly are sampled frequently

and those static systems not subject to changes are sampled 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. The charcoal adsorbera 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 radiciodine 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.

I i

Amendnent No. 15 3-lT i

3.0 SURVEILLANCE REQUIREMENTS 3.2 Equiument and Sampling Tests (Continued)

-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 4

of radiciodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers vill assure system integrity and performance.

i Pressure drop across the combined HEPA filters and charcoal j adsorbers, for each of the air treatment systems, of less than 6 inches of water vill indicate that the filters and edsorbers 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 vill demonst eate 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 adsorbers could become con-taminated from the fumes, chemicals or foreign materials, testing vill be performed to confirm system performance.

Demonstration of the automatic and/or ratnual initiation capability

< vill assure the system's availability. ,

References FSAR, Section 9.10 i.

r i

Amendment No. 15 3-lTa 4

. . - - ~ , - _ - , - - - . . _ ,, ..--, , ,..vy

TABLE 3 h 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 lh 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 /> whenever the radio-setivity exceeds 2.0 pCi/gs DOSE EQUIVALENT I-131.

(iii) 1 sample within 2h hours following a thermal power change cxceeding 15%

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

(3) E Determination 1 per 6 months (

(h) Dissolved oxygen 1 per 3 days and chloride (b) Hot Standby (1) Gross Radioactivity 1 per 3 days Hot Shutdown (2) Isotopic Analysis for (1) 1 per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> DOSE EQUIVALENT I-131 whenever the radio-activity exceeds 2.0 pCi/g= DOSE EQUIVALENT I-131.

(ii) 1 sample within 2k I

I 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

e 1

TABLE 3 k_(Continued)

MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frequency i 1. Reactor Coolant (Continued)

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

2. Steam Generator Isotopic Analysis for DOSE 1 per 7 days.

Coolant EQUIVALENT I-131

3. SIRW Tarl Boron Concentration 1 per 31 days

h. Concentrated Boric Boron Concentration- 1 per 31 days Acid Tanks e

5 SI Tanks Boron Concentration 1 per 31 days

, 6. Spent Fuel Pool Boron Concentration 1 per 31 days l

4 (1)Until the radioactivity of the reactor coolant is restored to < 212C1/gm DOSE EQUIVALENT I-131 (2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

1 Amendment No. 28 19

~ - - , .-y . - .

4 DET.m2 l

l l

A: lend =ent To. 28 3-19a

r ATTACHMENT 3 DISCUSSION OF PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS I

The proposed revisions to the Fort Calhoun Station Unit No. 1 Technical Specifications are intended to provide the following functions:

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

2. Adds Limiting Conditions for Operation during or following a power transient for which Section 2.1.3 of the present Technical Specifi-cations 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 an 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, compatibi] tty 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 norral reactor operation.

o It is concluded that based cn the following reasons the proposed Technical Specifications do 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 accinents 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 safety of margin as defined in the applicable Technical Specifica-tions 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.

1

')

i COMPARISON OF STANDARD TECICTICAL SPECIFICATIONS (STS)

AND THE PROPOSED TECHNICAL SPECIFICATIONS FOR FORT CALHOUN STATION UNIT NO. 1 Section or Section or Subsection Subsection of of Proposed STS Tech. Specifications Remarks 1.0 Definitions Appropriate / applicable definitions have been incorporated, i

I. REACTOR COOLANT SYSTEM 3.h.9.a 2.1.3(1)a The radioactivity of the reactor coolant for DOSE EQUIVAISIT I-131 is 2.0 pCi/gm instead of 1.0 pCi/gm, as per STS. The 2.0 pCi/gm limit is based on approximately 1% failed fuel as referenced in Table

, 11.1.5of the FSAR and is based on the methodology presented in NUREG-0017 Also, the thyroid doses under accident conditions using 2.0 pCi/gm without iodine spiking are less than 1% of 10 CFR Part 100 value. Fort Calhoun Station has operated ~

for approximately 7 years without any undue hazard to the public as per 10 CFR Part 20 eld Appendix I to 10 CFR Part. 50.

3.h.9.b. 2.1. 3( 2 )b -----

Action a 2.1.3(2) 1. The proposed Technical Specification is considered conservative since the upper limit for DOSE EQUIVAISIT I-131 during power transiet.'3 (iodine spiking) is not allowed to exceed 60 pCi/gm.

Figure 31h-1 of STS allows the radio-activity to exceed 60 pCi/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 h8 hours. This 100-hour time limit has been obtained after reviewing the past 7 years operating history. It was determined i that it takes approximately 100 to 150

' hours to restore the radioactivity within acceptable values.

2

.. - ~_ - . . .~ ~ ~. -

J

}

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 i Action d 2.1.3(5) Incorporated Tabie h.h h Item 1 Table 3 h, Items Incorporated 1(a)(1) and'1(b)(1)

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

Item 3 Table 3 h, Item Incorporated 1(a)(3)

Item h(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)(1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are considered sppropriate.

Based on the operating history of the plant and especially during iodine spiking phenomenon, the sampling require-

, ment of one sample within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is

coneidered appropriate.

l II. SECONDARY COOLANT SYSTEM 3 7 1.4 Proposed new Incorporated p Specification 2.20(1)

Action 2.20(2) Incorporated Table h.7-1 Item 1 -

Not considered appropriate due to its implication / interaction with Item 2 cf STS. Also, the determination of-gross radioactivity does not have any bearing

on the safety considerations following a main steam line break.

~

l Item 2 Table 3 h, Item 2 The proposed sampling requirements are cornidered more limiting.

l 3

l-

- - . . .,. , ._- -.~ -

. - . . ~ , -

JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be Class III within the meaning of 10 CFR 170.22 because its acceptability has been identified by Comission positions. The Comission 4

identified the need and format for the proposed amendment by letter dated July 22, 1980.

i i

s l

ATTACHMENT C

- . . ., ,. .. --.