ML20214K388

Amend 6 to License NPF-36,revising Tech Specs to Reflect Util Plan to Enrich Boron in Standby Liquid Control Sys to 85 Atom Percent B-10 Per 10CFR50.62(c)(4)

ML20214K388
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	05/18/1987
From:	Butler W Office of Nuclear Reactor Regulation
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ML20214K333	List: ML20214K331 ML20214K388 ML20214K396
References
NUDOCS 8705280425
Download: ML20214K388 (11)

Text

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UNITED STATES y g NUCLEAR REGULATORY COMMISSION 3  : E WASHINGTON, D. C. 20555 LONG ISLAND LIGHTING COMPANY DOCKET NO. 50-322 SHOREHAM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE -

Amendment No. 6 License No. NPF-36

1. The Nuclear Regulatory Comission (the Comission) has found that A. The application for amendment by Lono Island Lighting Company ,

(the licensee), dated February 4,1987 and supplemented by letter dated April 10,1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comis-sion's rules and regulations set forth in 10 CFR Chapter I; -

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 public, and (ii) that such activities will be-conducted )

in compliance with the Comission's regulations; D. The issuance of this amendment will not be ininical to the comon defense and security or to the health and safety of the oublic; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable reouirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications I as indicated in the attachment to this license amendment, and paragraph 2.C.(2) I of Facility Operating License No. NPF-36 is hereby amended to read as follows: l 1

Technical Sp?cifications l The Tachnir.al Specifications contained in Appendix A and the Environniental Protection Plan contained in Appendix B, as revised through Amendment No. 6 are hereby incorporated into this license. .

Long Island Lighting Company shall operate the facility in accordance with the Technical Speci'ications and the Environmental-Protectinn Plan, s2%h 6 P P

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

FOR THE NUCLEAR REGULATORY COMMISSION

/s/

Walter R. Butler, Director J

Pro.iect Directorate I-2 Division of Reactor Projects I/II

Attachment:

Changes to the Technical Specifications Date of Issuance: May 18, 1987 f

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3. This license amendirent is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION .

Walter R. Butler, Director Pro.iect Directorate I-2 Division of Reactor Pro.iects I/II .

Attachment:

Changes to the Technical . .

Specifications c Date of Issuance: May 18, 1987 t

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ATTACHMENT TO LICENSE AMEN 0 MENT N0. 6 FACILITY OPERATING LICENSE NO. NPF-36 l

DOCKET NO. 50-322 .

-*__e Replace the followino pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages provided to maintain dacument completeness.*

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Remove Insert 3/4 1-19 3/4 1-19 . .

3/4 1-20 3/4 1-20 <

3/4 1-21 3/4 1-21*

3/4 1-22 3/4 1-22 B 3/4 1-3 8 3/4 1-3*

B 3/4 1-4 8 3/4 1-4 8 3/4 1 da B 3/4 1-4a O

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REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE.

. -e APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*

ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

With one pump and/or one explosive valve inoperable, restore the

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inoperable pump and/or explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. With the standby liquid control system otherwise inoperable, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. In OPERATIONAL CONDITION 5*:

1. With one pump and/or one explosive valve inoperable, restore the inoperable pump and/or explosive valve to OPERABLE status within 30 days or insert all insertable control rods within the next hour.

2. With the standby liquid control system otherwise inoperable, insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:

1. The available volume and temperature of the sodium pentaborate solution are within the limits of Figures 3.1.5-1 and 3.1.5-2.

2. The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 65'F.

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• With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

SHOREHAM - UNIT 1 3/4 1-19 Amanthent No. 6

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge. , _

2. Determining that the available net weight of sodium pentaborate .

is greater than or equal to 1171 lbs and the concentration of l boron in solution is within the limits of Figure 3.1.5-2 by chemical analysis.*

3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured .

in position, is in its correct position.

4. Starting both pumps and recirculating demineralized water to the test tank. *

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1220 psig is met. l '

d. At least once per 18 months during shutdown by:

1. Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve-shall be from the same manufactured-batch as the one fired or from another batch which has been certified by having one of that batch successfully fired. Both injection loops shall be tested in 36 months. No replacement l

squib valve shall be used that will exceed its shelf life or l useful life as applicable during its inservice period.

2. Verifying that the relief valve does not actuate during recirculation to the test tank.

3. Demonstrating that all heat traced piping is unblocked by pumping  ;

from the storage tank to the test tank and then draining and  !

flushing the piping with domineralized water.** l

4. Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.

e. At each refueling outage by:

1. Verifying that a minimum Boron-10 enrichment of 85 atom percent exists by analyzing a sample taken from the tank.

• This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limit of Figure 3.1.5-1.

• • This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, over-lapping or total flow path steps such that the entire flow path is included.

SHOREHAM - UNIT 1 3/4 1-20 amendment too. 6

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CONCENTRATION REQUIREMENTS . '

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NOTE :

MIN. BORON 8 80 ISOTOPE F IGURE 3.I.5-2

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l l REACTIVfTY CONTROL SYSTEMS BASES CONTROL R005 (Continued)

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature -a provides the only positive means of determining that a rod is properly cou' pled ,

and therefore this checs must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have af fected the control rod coupling integrity. The subseque.nt check is performed as a backup to the initial demon-stration.

In order to ensure that the control rod patterns can be followed and there -

fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control  ?

rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage ,

to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velncity limiter, could result in a peak i enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

Theanalysisoftheroddropaccident[spresentedinSection15.1.33of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.

The RBM is designed to automatically prevent fuel damage in the event o't erroneous rod withdrawal from locations of high power density during high power operation. Two channel', are provided. Tripping one of the channelf will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdraw.il of control ruds.

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SHOREHAM - UNIT 1 B 3/4 1-3 '

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I REACTIVITY CONTROL SYSTEMS j BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM i

The Standby Liquid Control System (SLC) provides a backup capability to~ '*--*

bring the reactor from full power to a cold, xenon-free shutdown, assuming that <

the withdrawn control rods remain fixed in the rated power pattern. The generic system design bases requires the injection of a quantity of sodium pentaborate analytically equivalent to a concentration of 660 ppm of natural boron into the reactor core. The amount of sodium pentaborate available for injection into the reactor is the amount required to provide 660 ppm of natural baron in the 70*F moderator at water level 8, including the recirculation loops, and with the RHR shutdown cooling subsystem in operation. There is also an additional allowance equivalent to 165 ppm of natural baron in the moderator to account for imperfect * ,

mixing and leakage, resulting in a total required equivalence of 825 ppm. c With the use of enriched baron, the required boron concentration is reduced in proportion to the enrichment ratio. The ratio is expressed in terms of atom percent of boron-10 since only the B-10 isotone controls the reactor reactivity. -

Ratio = minimum B-10 isotopic enrichment in atom perecnt B-10 isotopic fraction in natural boron in atom percent The fraction of B-10 isotope naturally occurring in boron is 19.78 atom percent. At an 85 atom percent enrichment level,193 ppm of enriched boron is equivalent to the generic shutdown requirement of 825 ppe of natural boron.

The generic design basis of the SLC System provides a specified cold shut-down boron concentration in the reactor core. The SLC System was typically designed to inject the cold shutdown boron concentration in 90 to 120 minutes.

• The time requirement was selected to override the reactivity insertion rate due to cooldown following the xenon poison peak and the required pumping rate is 41.2 gpm.

The minimum storage volume of the solution is established to include the generic shutdown requirement and to allow for the portion below the pump suction nozzle that cannot be inserted. An additional allowance in the SLC storage volume is provided to account for storage tank instrument inaccuracy and drif t.

Even with the maximum specified instrument inaccuracy and drift, the required quantity of sodium pentaborate solution is always available for injection.

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A normal quantity of 1,225 gallons of sodium pentaborate solution having a 10.9 percent concentration and 85 percent B-10 enrichment is required to meet the shutdown requirements. The temperature requirement for the sodium  :

pentaborate solution and the pump suction piping is necessary to ensure that the sodium pentaborate remains in solution.

1 With redundant pumps and explosive injection valves and with a highly I reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable, i

SHOREHAM - UNIT 1 B 3/4 '-4 hamnement no. 6

REACTIVITY CONTROL SYSTEMS (continued)

BASES 3/4.1.5 STANOBY LIQUID CONTROL SYSTEM Surveillance requirements are established on a frequency that assures a --$

high reliability of the system. Once the solution is established, no deterio- 1 ration of the boron-10 enrichment level is expected during system standby operation. The boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals -

will assure that these valves.will not fail because of deterioration of the

, charges. Analyses have been performed and tests conducted that show when l highly enriched boron is used in the SLC System, there will be sufficient *

• mixing of the reduced volume, even at Emergency Procedure Guideline water <

1evels between the top of the active fuel (TAF) and the Minimum Steam Cooling RPV Water Level to avoid core power oscillations.

The ATWS Rule (10CFR50.62) requires the addition of a new design require-ment to the generic SLC System design basis. Changes to flow rate, solution concentration or boron enrichment, to meet the ATWS Rule must not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution." (natural boron enrichment)

The described minimum system parameters (41.2 gpm, 9.8% concentration and 85 atom percent baron-10 enrichment) will ensure an equivalent injection capa-city that is 200% of the ATWS Rule requirement for Shoreham. This enhanced mitigation capability increases the probability that the SLC System will be initiated in a timely manner, result in less severe primary containment transients and reduce the risk of offsite radiological consequences in excess of the 10CFR100 limits, due to the unlikely occurrence of an ATWS event.

1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for large BWR's", G. E. Topical Report NE00-10527, March 1972

2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement.1 to NE00-10527, July 1972.

3. J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2. " Exposed Cores",

Supplement 2 to NE00-10527, January 1973.

SHOREHAM - UNIT 1 8 3/4 1-4a Amendment No. 6