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Amend 22 to License NPF-39,revising Tech Specs Re Standby Liquid Control Sys to Ensure Compliance W/Paragraph (c)(4) of ATWS Rule (10CFR50.62) & to Simplify & Improve Tech Spec Requirements for Sys
	ML20245D029
Person / Time
Site:	Limerick
Issue date:	06/08/1989
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
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ML20245D030	List: ML20245D029; ML20245D044;
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	NUDOCS 8906260370
	Download: ML20245D029 (13)
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UNITED STATES NUCLEAR REGULATORY COMMISSION n

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PHILADELPHIA ELECTRIC COPPANY i

DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 22 License No. NPF-39 1.

The Nuclear Regulotory Comission (the Ccwnission) has found that A.

The application for amendment by P'111adelphia Electric Company (the licensee) dated February 22, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set i

forth in 10 CFR Chapter I; l

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

Thereisreasonableassurance(1)'thattheactivitiesauthorizedby j

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted t

in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon l

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.

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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) l of Facility Operating License No. NPF-39 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 22 are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

8906260370 890608 PDR ADOCK 05000352 i

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3.

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

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FOR THE NUCLEAR REGULATORY COMMISSION

/S/

Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II

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Attachment:

Changes to the Technical Specifications l

Date of Issuance:

June 8, 1989 0

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This license amendment is effective as of its date of. issuance, a

FOR THE NUCLEAR REGULATORY COWISSION L ()

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Walter R. Butler, Director Project Directorate.I.2 Division of Reactor Projects I/II

Attachment:

Changes to'the Technical Specifications Date of Issuance:

June 8, 1989

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A!

L ATTACHMENT TO LICENSE ~ AMENDMENT NO. 22 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352

-Replace the following pages of the Appendix A Technical Specifications with the attached page.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.*

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3/4 1-19 3/4 1-19 3/4 1-20 3/4 1-20 3/4 1-21 3/4'l-21, 3/4 1-22 3/4 1-22 B 3/4 1-3 8 3/4 1-3*

B 3/4 1-4 BJ,/4 1-4 i

B'3/4 1-5 1

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3-INDEX LIMITINGCONDITI0dFOROPERATIONANDSURVEILLANCERE0VIREMENTS I

SECTION PAGE

'3/4.0 APPLICABILITY...............................

3/4 0-1 I

3/4.1 REACTIVITY CONTROL SYSTEMS

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.3/4.1.1 SHUTDOWN MARGIN...................................

3/4 1-1 l

3/4.1.2 REACTIVITY AN0MALIES.......................................

3/4 1-2 3/4.1.3 CONTROL R005 i

Control. Rod Operability..................................

3/4 1-3 Control Rod Maximum Scram Insertion Times................

3/4 1-6 Control Rod Average Scram Insertion Times................

3/4 1-7 Four Control Rod Group Scram Insertion Times.............

3/4 1-8 r

Control Rod Scram Accumulators...........................

3/4 1-9 f o

Control Rod Drive Coupling...............................

3/4 1-11 Control Rod Position Indication.........................

3/4 1-13 Control Rod Drive Housing Support........................

3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer......................................

3/4 1-16 Rod Block Monitor........................................

3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................

3/4 1-19

_ Figure 3.1.5-1 Sodium Pentaborate Solution h

Temperature / Concentration

%f Requirements........................

3/4 1-21 L

Figure 3.1.5-2 -Deleted (LEFT BLANK INTENTIONALLY)..

3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 1

l 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...............

3/4 2-1 L

L Figure 3.2.1-1 Maximum Average Planar Linear Heat I

Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB278............

3/4 2-2 l~

LIMERICK - UNIT 1 v

Amendment No. 22 n

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION pAGE t

POWER DISTRIBUTION LIMITS (Continued)

Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Average Planar Exposure Initial t

Core f uel Types 98CI8248...........

3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus l'

Average Planar Exposure Initial Core Fuel Types PSCIB163...........

3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types 98CIB094...........

3/4 2-5 Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB071...........

3/4 2-6 Figure 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure For Fuel Type BC320A (GE8X8EB)..............

3/4 2-6a Figure 3.2.1-7 Maximum Average Planar linear Heat Generation Rate (MAPLHGR)

Versus Average Planar Exposure For Fuel Type 8C318A (GE8X8EB).....

3/4 2-Eb Figure 3.2.1-8 Maximus Average Planar Linear Heat Generation Rate (MAPLGHR)

Versus Average Planar Exposure For Fuel Type BC322A (GE8X8EB).....

3/4 2-6c 3/4 2.2 APRM SETP0!NTS..........................................

3/4 2-7 3/4 2.3 MINIISM CRITICAL POWER RATI0............................

3/4 2-8 Table 3.2.3-1 Deleted 7 Figure 3.2.3-la Minimus Critical Power Ratio (MCPR)

.n Versus I (P8X8R/BP8X8R Fuel) BOC to EOC-2000 MWD /ST........................

3/4 2-10 Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR)

Versus t (P8X8R/BP8X8R Fuel) EOC-2000 MWD /ST to E0C..........................

3/4 2-10a LIMERICK UNIT - 1 vi Amendment No.V.19 APR 2 41989 e

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' REACT 1V1TY CONTROL SYSTEMS

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3/4.1.5 STANDBY 1,1 QUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*

- ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

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

b.

In OPERATIONAL CONDITION 5*:

1.

With only one pump and corresponding explosive valve.0PERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 30 days or insert all insertable E

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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

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

1.

The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2.>l The available volume of sodium pentaborate solution is at least 7 4537 gallons.

3.

The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 70'F.

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

l LIMERICK - UNIT 1 3/4 1-19 Amendment No. 22

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1 REACTIVITY CONTROL SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by:

1.'

Verifying the continuity of the explosive charge.

2.

Determining by chemical analysis and calculation

that the available weight of sodium pentaborate is greater than or equal to 5389 lbs; the concentration of sodium pentaborate in solution is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:

C x

Q 13% wt.

86 gpm

->1 where C = Sodium pentaborate solution (% by weight)

Q = Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

3.

Veritying 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.

. Demonstrating that, when tested pursuant to Specification 4.0.5, the*"

c.

minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1190 psig is met.

d.

At least once per 18 months during shutdown by:

1.

Initiating at least 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 success-fully fired.

All injection loops shall be tested in 3 operating cycles.

2.

- Demonstrating that all heat traced piping is unblocked by pumping

" from the storage tank to the test tank and then draining and u flushing the piping with demineralized water.

3. # 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.

This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below 70*F.

- This test shall clso be performed whenever all three heat tracing circuits have

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been found to be inoperable and may be performed by any series of sequential, 4

overlapping or total flow path steps such that the eritire flow path is included.

LIMERICK - UNIT 1 3/4 1-20 Amendment No. 22

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l 50 10 11 12 13 14 15 CONCENTRATION, ',' SY WEIGHT S0DIUM PENTABORATE SOLUTION TEMPERATURE / CONCENTRATION REQUIREMENTS FIGURE 3.1.5-1 1

LIMERICK - UNIT 1 3/4 1-21 Amendment No. 22

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LIMERICK - UNIT 1 3/4 1-22 Amendment No. 22 n

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REACTIVITY CONTROL SYSTEMS:

BASES 9

. CONTROL R005 (Continued)

Contro1 r'od coupling integrity is required to ensure compliance with the.

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analysis of the rod drop accident in the FSAR.

The overtravel position-feature i

provides the only positive means of ' determining that a rod is properly coupled and therefore this check must be performed prior to acnieving criticality after completing. CORE ALTERATIONS that could have affected the control rod coupling-integrity. The subsequent 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 OPERA 8LE.

u 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 withdrawai isL 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 R00 PROGRAM CONTROLS Control rod witM rawal 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 10% of RATED THERMAL POWER, there is no possible. rod worth which, if I

dropped at the design rate of the velocity limiter, could result in a peak enthalpy-of 280 cal /gs. Thus requiring the RWM to be OPERA 8LE when THERMAL l

POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate L

control.

The M provides automatic supervision to assure that out-of-l sequence rods will not be withdrawn or inserted.

1 The analysis of the rod drop accident is presented in Section 15.4.0 of the FSAR and the techniques of the analysis are presented in a topical report, j

3 Reference 1, and two supplements, References 2 and 3.

Additional pertinent i'i analysis is also contained in Amendment 17 to the Reference 4 topical report.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.

Two channels are provided.

Tripping one of the channels will block erroneous rod withdrawal to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

LIMERICK - UNIT 1 8 3/4 1-3 Amendment No. 17 MAR 2 ': 1989 i

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i REACTIVIT CONTROL SYSTEMS 1

BASES 7

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.

To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel.

To allow for potential leakage and improper mixing, this concentration is increased by 25%.

The required concentration is achieved by having available a minimum quantity of 4,537 gallons of sodium pentaborate solution containing a minimum of 5,389 lbs of sodium pentaborate.

This quan-tity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permis-sible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay.

The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly 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.

The SLCS system consists of three separate and independer pumps and explosive valves.

Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

l The SLCS'must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants).

This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

L LIMERICK - UNIT 1 B 3/4 1-4 Amendment No. 22

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REACTIVITY CONTROL SYSTEMS 1 BASES sy

' STANDBY. LIQUID CONTROL SYSTEM (Continued)

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Surveillance requirements are established on a frequency that assires a high reliability.of the system.

Once the solution is established, boron con-centration will not vary unless more boron'or water is added, thus a checkton the temperature and~ volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

1.

C. J..Peone, 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 NED0-10527, July 1972.

3.

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

Supplement 2 to NEDO-10527, January 1973.

4.

Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, l

" General Electric Standard Application for Reactor Fuel".

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LIMERICK - UNIT 1 B 3/4 1-5 Amendment No. 22 L_____._____.______.______._____________.

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ML20245D029

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