ML20209E912

Amend 53 to License DPR-45,revising Tech Specs to Add Operational Requirements on Containment Bldg Ventilation Inlet & Exhaust Dampers & Correcting Typo in Amend 50

ML20209E912
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	09/05/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20209E891	List: ML20209E887 ML20209E912 ML20209E924 ML20209E942
References
NUDOCS 8609110385
Download: ML20209E912 (7)

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/ 'o,, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION 3 a wAsHmorow, o. c. 20sss g .....

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DAIRYLAND POWER COOPERATIVE DOCKET NO. 50-409 LA CROSSE BOILING WATER REACTOR AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.53 License No. DPR-45

1. TheNuclearRegulatoryCommission(theCommission)hasfoundthat:

A. The application for amendment by Dairyland Power Cooperative (the

_ licensee) dated July 11, 1984 as amended by letter dated September 17 1985 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter It B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the pubitc; and (ii) that such activities will be

, conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8609110305 860905 DR ADOCK 0500 9 l

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Provisional Operating License No. DPR-45 is hereby amended to read as follows:

r (2) Technical fpecifications The Technic.i Specifications containec' in Appendix A issued October 31, 1969, with Authorization No. DPRA-6, as revised through Amendment No. 53, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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~y 3. This license amendment is effective as of the date of its issuance.

f FOR THE NUCLEAR REGpAT @ COMISSION Y( -

Joh A. Zwolinski, Director BWR roject Directorate #1 Division of BWR Licensing 1

Attachment:

Changes to the Technical

, Specifications i

f . Date of Issuance: September 5, 1986 i

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i ATTACHMENTTOLiCENSEAMENDMENTNO.53

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PROVISIONAL OPERATING LICENSE NO. DPR-45 I

DOCKET NO. 50-409 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT

, 28 28*

r 29 29*

'e 29a 29a r 29b 29b i

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• Pagination change only.

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'i 4.1 GENERAL 4.1.3 During periods when the reactor is in Condition 3, 4 or 5, either Channel 1 or 2 of the Nuclear Instrumentation System shall be in operation and shall be monitored by the operator.

4.1.4 Whenever the reactor contains one or more fuel elements, any operations '

from points outside the control room of equipment which may affect the reactor shall be conducted under the direction, or with the knowledge, of the control room operator.

4.1.5 If the plant is operational during a tornado warning, the Shift Super-visor on duty shall keep informed of the actual tornado activity which may i - approach the plant. In the event that reports indicate an inmiinent tornado strike at or near the LACBWR plant, the Shift Supervisor shall reduce reactor

,- power to a level which permits prompt reduction of power generation to station load. However, the Shift Supervisor shall be instructed to discontinue plant

,_ operation if, in his judgment, this action is required to ensure plant safety.

~ 4.1.6 If the plant is in CONDITION 1, 2, or 3 and the Mississippi River level adjacent to the plant reaches 639.2' and is predicted to exceed 640', commence reactor shutdown and be in CONDITION 4 prior to the river level exceeding 640'.

4.2 OPERATING LIMITS i ,

4.2.1 Reactor Building i 4.2.1.1 CONTAINMENT INTEGRITY ahall be maintained in Conditions 1, 2, 3 and during:

(a) CORE ALTERATIONS,

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I (b) handling of irradiated fuel, or (c) there is fuel in the reactor and any control rod is withdrawn.

l 4.2.1.2 Gasketed closures and ventilation system closures which have been sub-jected to maintenance, repair or other operations which might affect their per-formance shall, before any subsequent operation for which containment integrity is required, be tested for leak tightness using the soap-bubble technique (or other method of equivalent sensitivity). This test shall be performed using a pressure differential no less than 0.5 psi and the results shall be used as a guide in evaluating leakage.

i Amendment No. A/e, M ,53 i

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I 4.2.1.3 The number of containment,sessel electrical penetrations may be

{ varied, as may the containment vessel piping penetrations, provided the new penetrations are equivalent in design to the existing penetrations. New penetratio6s are defined as conduit or piping requiring attachment to the

containment vessel shell. After any such penetration changes, an integrated reactor building leak test shall be performed at approximately 52 psig. This test must demonstrate that the containment vessel meets the leak rate specified in Sec. 5.2.1. .

4.2.1.4 Existing containment vessel penetrations may be removed from or placed in service, provided the containment vessel is not affected and the closures are equivalent in strength and tightness to those previously

' installed. Af ter any such change the penetration, exclusive of the containment shell connection, shall be tested for leak tightness at a pressure no less than 52 psig using the soap-bubble technique (or other method of equivalent sensitivity) or by determining the rate of pressure loss of a test

][ chamber. The penetration leakage rate shall not exceed 1.0 percent of the containment vessel leak rate Lp specified in Sec. 5.2.1.

'r 4.2.1.5 The reactor building vacuum breakers shall be set to relieve at a A' differential (external-over-internal) pressure not exceeding 0.5 psi.

4.2.1.6 The reactor building ventilation system isolation dampers shall

, - _ close upon loss of control air or loss of electrical power supply, and they shall be automatically closed by abnormal conditions as specified in Table 1.

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4.2.1.7 The main steam line, the reactor building vent header, and the decay heat system blowdown line shall be isolated automatically by abnormal

! conditions as specified in Table 1.

e 4.2.1.8 The containment building steel shell temperature shall be greater than 0*F during reactor operation or during any integrated leak rate test

! performed with a pressure exceeding 10.4 psig.

I! 4.2.1.9 The containment building shall be isolated whenever the spent fuel storage well contains irradiated fuel which has decayed less than 43* days j after exposure in a critical reactor and a shipping cask'.for irradiated fuel i is being moved by the crane on the 701 foot level or located within one cask

! length of the top of the spent fuel storage well or is within the spent fuel

} storage well. During cask movement near or at the FESW the water level in the FESW sust be at least 16 ft. above the top of the fuel storage rack (no more than 7 feet below the top of the FESW).

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• 43 days for off loading less than one half of the core, i.e. less than 36 j

fuel elements. 51 days for off loading more than 36 fuel elements.

'I Amendment No. 39, AI,53 1

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REACTOR BUILDING

'f ' CONTAINMENT VENTILATION DAMPERS ,.

i LIMITING CONDITION FOR OPERATION t.

4.2.1.10 The Containment ventilation inlet and outlet dampers shall be OPERABLE with isolation times of less than or equal to 10 seconds.

APPLICABILITY: Whenever CONTAINMENT INTEGRITY (Specification 4.2.1.1) .-

is required.

ACTION:

With one or more of the above ventilation damper (s) inoperable:

A. Operation may continue provided that at least one damper in each affected penetration is maintained OPERABLE and either:

1. The inoperable damper (s) is restored to OPERABLE status within 4

., hours or Yr

'., 2. Each affected penetration is isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one automatic damper secured in the isolation position, or a blank flange; OR

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B. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless the affected penetration is isolated.

C. The provisions of Specification 3.0.4 are not applicable if the affected

, penetration is isolated.

y SURVEILLANCE REQUIMMENTS __

i - 5.2.1.10 1 The ventilation dampers shall be demonstrated OPERABLE prior to

){ returning the das,>er to service after maintenance, repair or replacement work j.

is performed on the damper or its associated actuator, control, or power l circuit by performance of a cycling test, and verification of isolation time.

5.2.1.10.2 The isolation time of each above damper shall be determined to be within its limit when tested pursuant to specification 5.2.2.

{ 5.2.1.10.3 The seat rings of the ventilation inlet and outlet dampers shall be replaced at least once per 5 years.

! BASES 1 The Nuclear Regulatory Commission requested that steilar Technical j

5pecifications per Generic Ites B-24 and NUREG 0737 Ites II.E.4.2 be submitted i, to help assure operabilty of containment ventilation despers. DPC had i;

previously committed to replace the containment ventilation inlet and outlet

{> dampers' resilient sealing material at least once each 5 years until such time

. as additional in-situ data can be accumulated to justify a longer interval.

1 j' If in-situ data is accumulated which supports a longer seal replacement intervals, a change to Specification 5.2.1.10.3 may be requested.

j Specification 3.0.4 is not applicable if the affected penetration is isolated ll since the safety function of the dampers is to close.

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{I Amendment No. 53 ii A

4.2.2 Reactor Vessel, Coolant, and Auxiliary Systems 4.2.2.1 Additional penetrations to the systems containing reactor coolant shall be designed, manufactured, and tested according to the provisions of the ASME Boiler and Pressure Vessel Code and the ASA Code for Pressure Piping applicable as of June 1962. These additional penetrations shall be limited to instrument connections and piping connections, the latter being no larger than 1-in. inside diameter.

4.2.2.2 The reactor coolant shall be light water and shall conform to the following requirements.

CONDITION 1 Normal Limit Maximum Limit Chloride concentration .2 ppe .5 ppe pH 5.3 - 8.6 NA Conductivity 3 paho/cm 10 paho/cm The time spent above 3 paho/cm at 70*F - 80*F and .2 p not exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per incident nor 2 weeks per year.peIf chloride should l either time limit is exceeded, an orderly shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unless returned to within the limits. When the maximum conductivity or chloride limits are exceeded an orderly shutdown should be initiated immediately. If the pH is outside the limits for a period of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> an orderly shutdown shall be initiated.

CONDITION 2 & 3 Normal Limit Maximum Limit Chloride concentration .1 ppa NA pH 5.3 - 8.6 NA Conductivity 5 paho/cm NA i

The time above 5 phso/cm at 70*F to 80*F and .1 ppa chloride concentra-tion is restricted to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for any single occurrence during Condi-tion 2.

i When this time limit in condition 2 is exceeded the reactor shall be brought to the hot shutdown Condition (Condition 3) until the limits i are restored. If the limits can not be restored in an additional 7 days, the reactor shall be taken to the cold shutdown condition (Condition 4).

CONDITION 4 & 5 Normal Limit Chloride concentration .5 ppa pH 5.3 - 8.6 i

Conductivity 10 paho/cm k The primary system chemistry parameters defined in this section shall be determined at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in Condition 1, 2, and 3 and at least once every 7 days in Condition 4 and 5.

l 4.2.2.3 Deleted

- 29b - Amendment No. M, 80, 53 i t