ML20133C274

Amend 179 to License DPR-40,revising TSs to Increase Amount of TSP Dodecahydrate Located in Containment Sump Storage Baskets

ML20133C274
Person / Time
Site:	Fort Calhoun
Issue date:	12/30/1996
From:	Wharton L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20133C261	List: ML20133C258 ML20133C274 ML20133C277
References
NUDOCS 9701070102
Download: ML20133C274 (10)

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pV84 gj 44 UNITED STATES g

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001

\...../ OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 179 .

l License No. DPR-40 '

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1. The Nuclear Regulatory Comission (the Comission) has found that: l A. The application for amendment by the Omaha Public Power District (the licensee) dated May 31, 1996, complies with the standards and l requirements of the Atomic Energy Act of 1954, as amended (the  ;

Act), and the Comission's rules and regulations set forth in 10 i CFR Chapter I; B. The facility will operate in conformity with the application, the 4 provisions of the Act, and the rules and regulations of the  !

Comission; C. There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be 3 i

conducted in compliance with the Comission's regulations- '

i D. The issuance of this license amendment will not be inimical to the 4 comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements  !

have been satisfied.

9701070102 961230 PDR ADOCK 05000285 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. 179, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION L. Raynard Wharton, Project Manager Project Directorate IV-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 30, 1996

8IJACHMENT TO LICENSE AMENDMENT NO. 179 FACILITY OPERATING LICENSE NO. DPR-40

, DOCKET NO. 50-285 s

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 2-22 2-22 2-23 2-23 2-23a 2-23a 2-23b 2-23b 3-54a 3-54a 3-57 3-57 i 3-57a 3-57a l l

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. 2.0 LIMITING CONDITIONS FOR OPERATION ,

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2.3 Emergency Core Cooline System (Continued) l l (3) Protection Against Low Temperature Overoressurization The following limiting conditions shall be applied during scheduled heatups and coo: downs. Disabling of the HPSI pumps need not be required if the RCS is vented through at least a 0.94 square inch or larger vent.

Whenever the reactor coolant system cold leg temperature is below 385*F, at least one (1) HPSI pump shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 320*F, at least two  ;

(2) HPSI pumps shall be disabled.

l Whenever the reactor coolant system cold leg temperature is below 270*F, all three (3) HPSI pumps shall be disabled.

l In the event that no charging pumps are operable when the reactor coolant system cold l leg temperature is below 270 F, a single HPSI pump may be made operable and j utilized for boric acid injection to the core, with flow rate restricted to no greater than l 120 gpm.

(4) Trisodium Phosphate frSP) Dodecahydrate l During operating Modes 1 and 2, the TSP baskets shall contain 2: 100 ft' of active l TSP.

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a. With the above TSP requirements not within limits, the TSP shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. l

( b. With Specification 2.3(4)a required action and completion time not met, the plant shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within .

the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.  !

Halui The normal procedure for starting the reactor is to first heat the reactor coolant to near  !

operating temperature by running the reactor coolant pumps. The reactor is then made j critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode i of start-up, the energy stored in the reactor coolant during the approach to criticality is i substantially equal to that during power operation and therefore all engineered safety i

features and auxiliary cooling systems are required to be fully operable.

2-22 Amendment No. 4-7739,43,47,64,74,77, 100,103,133,141,157,161, 179

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2.0 LIMITING CONDITIONS FOR OPERATION  !

. 2.3 Emernency Core Coolina System (Continued) f l  ;

f; The SIRW tank contains a minimum of 283,000 gallons of usable water containing a boron  !

, . concentration of at least the refueling boron concentration. This is sufficient boron l concentration to provide a shutdown margin of 5 %, including allowances for uncertainties, i

. with all control rods withdrawn and a new core at a temperature of 60'F.*  ;

I The limits for the safety injection tank pressure and volume assure the required amount i

, of water injection during an accident and are based on values used for the accident  !

analyses. The minimum 116.2 inch level corresponds to a volume of 825 ff and the l maximum 128.1 inch level corresponds to a volume of 895.5 ft'. Prior to the time the _

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reactor is brought critical, the valving of the safety injection system must be checked for ,

correct alignment and appropriate valves locked. Since the system is used for shutdown  !

cooling, the valving will be changed and must be properly aligned prior to start-up of the i

reactor, j

! The operable status' of the various systems and components is to be demonstrated by  !

I 1-periodic tests. A large fraction of these tests will be performed while the reactor is  !

operating in the power range.  !

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If a component is found to be inoperable, it will be possible in most cases to effect repairs ,

j and restore the system to full operability within a relatively short time. For a single I

! component to be inoperable does not negate the ability of the system to perform its j

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function. If it develops that the inoperable component is not repaired within the specified  ;

allowable time period, or a second component in the same or related system is found to I

be inoperable, the reactor will initially be put in the hot shutdown condition to provide for i reduction of cooling requirements after a postulated loss-of-coolant accident. .This will

j. also permit improved access for repairs in some cases. After a limited time in hot

shutdown, if the malfunction (s) is not corrected, the reactor will be placed in the cold

shutdown condition utilizing normal shutdown and cooldown procedures. In the cold shutdown condition, release of fission products or damage of the fuel elements is not 4

considered possible.

. The plant operating procedures will require immediate action to effect repairs of en inoperable component and therefore in most cases repairs will be completed in less than the specified allowable repair times. The limiting times to repair are intended to assure that operability of the component will be restored promptly and yet allow sufficient time to effect repairs using safe and proper procedures.

The requirement for core cooling in case of postulated loss-of-coolant accident while in )

the hot shutdown condition is significantly reduced below the requirements for a postulated l loss-of-coolant accident during power operation. Putting the reactor in the hot shutdown  !

condition reduces the consequences of a loss-of-coolant accident and also allows more free j access to some of the engineered safeguards components in order to effect repairs. l l

i Failure to complete repairs.within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is i considered indicative of a requirement for major maintenance and, therefore, in such a case, the reactor is to be put into the' cold shutdown condition.

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.' 2-23 Amendment No. 32,39,47,49,74,179 l

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2.0 LIMITING CONDITIONS FOR OPERATION  !

-' 2.3 Emernency Core Cooline System (Continued) g  :

i' i ; With respect to the core cooling function, there is functional redundancy over most of the  !

j range of break sizes." l l -

l The LOCA analysis confirms adequate core cooling for the break spectrum up to and j

]- including the 32 inch double-ended break assuming the safety injection capability which ,

most adversely affects accident consequences and are defined as follows. The entire  :

contents of all four safety injection tanks are assumed to be available for emergency core  !

cooling, but the contents of one of the tanks is assumed to be lost through the reactor j coolant system. In addition, of the three high-pressure safety injection pumps and the two i

low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one low pressure operate while only one of each type is assumed to operate l i in the small break analysis *; and also that 25% of their combined discharge rate is lost i i from the reactor coolant system out of the break. The transient hot spot fuel clad ,

L temperatures for the break sizes considered are shown in USAR Section 14. j

! The restriction (. , * /SI pump operability at low temperatures, in combination with the -  ;

{- PORV setpoints ensure that the reactor vessel pressure-temperature limits would not be j 4

exceeded in the case of an inadvertent actuation of the operable HPSI and charging pumps.  !

l i Removal of the reactor vessel head, one pressurizer safety valve, or one PORV provides i sufficient expansion volume to limit any of the design basis pressure transients. Thus, no i additional relief capacity is required. l l .

Technical Specification 2.2(1) specifies that,'when fuel is in the reactor, at least one flow  !

! path shall be provided for boric acid injection to the core. Should boric acid injection i

become necessary, and no charging pumps are operable, operation of a single HPSI pump l t would provide the required flow path. The HPSI pump flow rate must be restricted to that l l of three charging pumps in order to minimize the consequences of a mass addition  ;

, transient while at low temperatures.

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i Trisodium Phosphate (TSP) dodecahydrate is required to adjust the pH of the recirculation

water to 2 7.0 after a loss of coolant accident (LOCA). This pH value is necessary to prevent significant amounts of iodine, released from fuel failures and dissolved in the recirculation water, from converting to a volatile form and evolving into the containment j atmosphere. Higher levels of airborne iodine in containment may increase the releases of i- radionuclides and the consequences of the accident. A pH of 2 7.0 is also necessary to l prevent stress corrosion cracking (SCC) of austenitic stainless steel components in containment. SCC increases the probability of failure of components.

i j The required amount of TSP is represented in a volume quantity converted from the

Reference 7 mass quantity using the manufactured density. Verification of this amount

! during surveillance testing utilizes the measured volume.

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t i 2-23a Amendment No. 39,47,P,74,77,100,179 l w,

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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued) i l References

, (1) USAR, Section 14.15.1 (2) USAR, Section 6.2.3.1 (3) USAR, Section 14.15.3 (4) USAR, Appendix K (5) Omaha Public Power District's Submittal, December 1,1976 (6) Technical Specification 2.1.2, Figure 2-1B (7) USAR, Section 4.4.3 l

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2-23b Amendment No. 47,5' ,7< ,179

. 3.0 SURVEILLANCE REOUIREMENTS

.- 3.'6 Safety Iniection and Conminment Cooling Systems Tests (continued)

, , (i) Verifying that the TSP baskets contain 2: 110 ft' of granular trisodium phosphate dodecahydrate. l (ii) Verifying that a sample from the TSP baskets provides adequate pH upward adjustment of the recirculation water.

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3-54a Amendment No. 44,179

. 3.0 SURVEILLANCE REQUIREMENTS l i .- 3.'6 Safety Iniection and Containment Cooling Systems Tests (Continued)  !

that the spray piping and nozzles are open will be made initially by a smoke test or other suitably sensitive method, and at appropriate intervals thereafter. A single containment i spray header flow rate of 3155 gpm of atomized spray is required to provide the. I

containment response

• specified in Section 2.4 of the Technical Specification: To achieve the 3155 gpm flow rate, no greater than ten (10) spray nozzles may be ,

inoperable of which no more than one may be missing. Since the material is all stainless l l steel, normally in a dry condition, with no plugging mechanism available, retesting at

, appropriate intervals is considered to be more than adequate. l

i Other systems that are also important to the emergency cooling function are the Si tanks, I the component cooling system, the raw water system and the containment air coolers.

The Si tanks are a passive safeguard. In accordance with the specifications, the water ,

i volume and pressure in the SI tanks are checked periodically eThe other systems l mentioned operate when the reactor is in operation and are continuously monitored for i  ;

L satisfactory performance. i The in-containment air treatment system is designed to filter the containment building atmosphere during accident conditions. Both in-containment air treatment systems are  :

o designed to automatically start upon accident signals. Should one system fail to start, the j redundant system is designed to start automatically. Each of the two systems has 100 l percent capacity."'

High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiolodine to the environment. The laboratory carbon l sample test results should indicate a radioactive methyl iodide removal efficiency of at i least 85 percent. If the efficiencies of the HEPA filters and charcoal adsorbers are as  !

specified, the resulting doses will be less than the 10 CFR Part 100 guidelines for the accidents analyzed. )

l Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.

If significant painting, fire or chemical release occurs in a ventilation zone communicating with the system that could lead to the degradation of charcoal adsorbers  :

or HEPA filters, testing will be performed to assure system integrity and performance.  !

3-57 Amendment No. 15,121,17F,179 i

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., 3.0 M'RVEILLANCE REQUIREMENTS 3.6 Safety Iniection and Containnant Cooling Systems Tests (Continued) i Operation of the system for 10. hours every month will demonstrate operability of the filters and adsorbers system and remove excessive moisture build-up on the adsorbers.

Demonstration of the automatic initiation capability will assure system availability.

Periodic determination of the volume of TSP in containment must be performed due to l the possibility of leaking valves and components in the containment building that could cause dissolution of the TSP during normal operation. A refueling frequency shall be utilized to visually determine that 2 110 ft' of TSP is contained in the TSP. baskets.

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This requirement ensures that there is an adequate quantity of TSP to adjust the pH of j the post-LOCA sump solution to a value 2 7.0. i l

The periodic verification is required on a refueling frequency. Operating experience has  !

shown this surveillance frequency acceptable due to margin in the volume of TSP placed  ;

in the containment building. i Testing must be performed to ensure the solubility and buffering ability of he TSP after exposure to the containment environment. A representative sample of 1.80 - 1.83 grams j of TSP from one of the baskets in containment is submerged in 0.99 - 1.01 liters of ,

water at a boron concentration of 2445 - 2465 ppm. At a standard temperature of 115 - l 125 F, without agitation, the solution should be left to stand for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The liquid is l then decanted and mixed, the temperature adjusted to 75 - 79 F and the pH measured. l At this point the pH must be 2 7.0. The representative sample weight is based on the minimum required TSP weight of 5,788 lbs, which at a manufactured density of at least l 53.0 lb./ft' corresponds to the minimum volume of 110 ft', and maximum possible post- j LOCA sump volume of 375,143 gallons, normalized to buffer a 1.0 liter sample. The i boron concentration of the test water is representative of the maximum possible boron j concentration corresponding to the maximum possible post-LOCA sump volume. The i post-LOCA sump volume originates from the Reactor Coolant System /RCSh the Safety l Injection Refueling Water Tank (SIRWT), the Safety Injection Tank.t (Sris; and the  :

Boric Acid Storage Tanks (BASTS). The maximum post-LOCA m1p boron j concentration is based on a cumulative boron concentration in the RCS, SIRWT, SITS 1 and BASTS of 2445 ppm. Agitation of the test solution is proh'bited, since an adequate

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standard for the agitation intensity cannot be specified. TI<e t st time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is '

necessary to allow time for the dissolved TSP to naturally diffuse through the sample solution. In the post-LOCA containment sump, rapid mixing would occur, significantly decreasing the actual amount of time before the required pH is achieved. This would ensure achieving a pH 2 7.0 by the onset of recirculation after a LOCA.

References (1) USAR, Section 6.2 (2) USAR, Section 6.3 (3) USAR, Section 14.16 (4) USAR, Section 6.4 3-57a Amendment No. +M,179