ML20011F337

Forwards Corrected Pages to Amend 171 to License DPR-62 Issued on 900206.Amend Changes Tech Specs to Add Footnote to Tables 2.2.1-1 & 3.2.2-2 for Adjustment of Main Steam Line Radiation Monitors When Water Chemistry Sys in Svc

ML20011F337
Person / Time
Site:	Brunswick
Issue date:	02/23/1990
From:	Loflin L CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS-90-052, NLS-90-52, TAC-75137, NUDOCS 9003050154
Download: ML20011F337 (6)

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s Coronne Power & Lhht Company SERIAL: NLS-90-052 897SB12 f FEB 2 31990 i  :

United States Nuclear Regulatory Commiesion ATTENTION: Document Control Desk Washington, DC 20555

' BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 r l DOCKET NO. 50-324/ LICENSE NO. DPR-62 '

CORRECTIONS TO RECENTLY ISSUED LICENSE AMENDMENT NO. 171

[' (TAC NO. 75137) -

r Gentlemen:

CP&L has received Amendment No. 171, dated February 6, 1990, for BSEP Unit 2 and id:ntified several corrections which are required.

Amendment No.171 changes the Technical Specification (TS) to add a footnoto to Tables 2.2.1-1 and 3 3 2-2 for the adjuritment of the main steam line radiation monitors trip setpoints to compensate for the increased background j

. radiation levels while the hydrogen water chemistry system is in servioo.

• Table 2.2.1-1 (pages 2 4 and 2-5) inadvertently failed to incorporate changes previously approved by Amendment No.168. Also, Bases Section 2.2 (pages BP.6,

• 32-7, and 62-8). inadvertently failed to inocrporate changes previously apprev0d ,

by. Amendment No. 166. i Tne Company regrets any inconvenience these oversights may have caused the i Staff and hereby submits corrected TS pages for reissuance. Please refer any ,

quest-lons regarding this submittal to Mr. M. R. Oates at (919) 546-6063.

Yours very truly, l f

L. .holin Manas r +

Nuclear Licens .ng Section DJK/eoo (625ECC)

Enclosure oo: Mr. S. D. Ebneter Mr. W. H. Ruland Mr. E. G. Tourigny

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9003050154 900223 h f/ 8

[DR ADOCK 05000324 PDC

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1 lll 411 Faye 1teville Street

• P. O Box 1$51

• Raleign. N C. 27602 '

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e E E ELE 2.2.1-1 E -

E REACTOR PROTECTION SYSTEM IV">TAUMENTATION SETPOINTS E

M ALLOWABLE

' FUNCTIONAL UNIT TRIP SETPOINT VALUES E

Q 1. Intermediate Range Monitor, Neutron Flux - Pinh(a) $ 120 divisions of full scale - $ 120 divisions' u of full scale

2. Average Powet Ruige Monitor

a. Neutron Flux - High, 15I(b) $ 15% of RATED THERMAL POWER $ 15% of RATED THERMAL POWER b.

Flow g gd Simulated Thercal Power - 1 (0.66 W + 641) with a $ (0.66 W + 67%) with High maximum i 113.5% of RATED a maximum $ 115I THERMAL POWER of RATED THERMAL u POWER e

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c. Fixed Neutron Flur - High(d) $ 120% of RATr.D THERMAL POWER $ 120% of RATED THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High $ 1045 psig $ 1045 psig 4 I Reactor Vessel Water Level - Low, Ipvel J 3 +162.5 inches E) 3 +162.5 inches (8)

5. Main Steam Line Isolation W 1ve - ClorureI '} $ 101 closed $ 10I closed

6. Main Steam Line Radiation - High IN $3x full power background $ 3.5 x full power l

> background i  ! ~ ~

E 7. Drywell Pressure - High 3 2 psig $ 2 psig

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E 8. Scram Discharge Volume Water Level - High < 109 gallons

$ 109 gallons z

Turbine Stop Valve-Closure II)

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. 9. $ 10% closed $ 10% closed

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Z 10. Turbine Control Valve Fa g Closure, Control Oil Pressure-Low 3.500.psig 3 500 psig

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1 TABLE 2.2.1-1 (Continued )

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS NOTES (a) The Intermediate Rar,e Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and the Average Power Range Monitors are on scale.

(b) This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the Run position.

(c) The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of the fraction of rated recirculation loop flow (W) in percent.

(d) The APRM flow-biased simulated thermal power signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds .ilrectly to instantaneous neutron flux.

(e) The Main Steam Line Isolation Valve-Closure scram f unction is automatically bypassed when the reactor mode switch is in other than the Run position.

(f ) These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER as measured by turbine first stage pressure.

(g) Vessel water levels refer to REFERENCE LEVEL ZEPO. '

(b) The Hydrogen Water Chr.mistry (HWC) system thall not be placed in service until reactor power reaches 20% of RATED THERMAL POWER. After re6ching 20% of RATED THERHAL POWER, the nc,rmal full power bacaground radiation level and associated trip setpoints may be increased to compensate for increased radiation levels as a result of full power operation with hydrogen injection. Prior to decreasing power below 20% of RATED THERMAL POWER and af ter the HWC system has been shut of f, the background level and associated setpoint shall be returned to the normal full power values. Control rod motion shall be suspended, when the reactor power is below 20% of RATED THERMAL POWER, until the necessary adjustment is made (except for scram or other emergency action).

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BRUNWICK - UNIT 2 2-5 Amendment No. 171

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2.2 LIMIT!WC SAFE 17 SYSTEM SETT!WCS ,

BASES (Continued)

4. Reactor Vessel Water Level-Low. Level #1 The reactor water level trip point was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the f uel to assure that there is adequate water to account for evaporation losses and displacement of cooling following the most severe transients. This setting T was also used to develop the thermal-hydraulic limits of power versus flow.

5. Main Steam Line Isolation Valve-Closure [

The low pressure isolation of the main steam line trip was provided to -

give protection against' rapid depressurization and resulting cooldown of the reactor vessel. Advantage was taken of the shutdown feature in the run mode which occurs when the main steam line isolation valves are closed, to provide  !

for reactor shutdown so that high power operation at low pressures does not occur. Thus, the combination of the low pressure isolation and isolation valve closure reactor trip with the mode switch in the Run position assures ,

the availability of neutron flux protection over the entire range of the Safety Limits. In addition, the isolation valve closure trip with the mode ,

switch in the Run position anticipates the pressure and flus t ransients which occur during normal or inadvertent isolation valve closure.

6. Main Steam Line Radiation - lligh The Main Steem Line Radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a scram is ireitiated to reduce the continued failure of fuel cladding. At the same time, the Vain Steam Line Isolation Valves are closed to limit the release of fission products. The trip setting is high enough above background radiation levelJ to preveat spurious sciams, yet low enough to promptly detect grces failures in the fuel c19dding.

The Main Steam Line Radiation detectors setpnints may be adjusted prior to placing the hydrogen water chemistry (WHC) system in service. If the setpoints are adjusted, the llWC system shall be placed in service or the retpoints shall be returned to the normal full power values within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the llWC system is not placed in service and the setpoints are not readjusted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, control rod motion shall be suspended (except for scram or other emergency action) until the necessary adjustments are made.  !

Hydrogen injection may cause the radiation levels in the main steam lines to increase. After shutting off the HWC system or decreasing powar, the setpoints shall be returned to the normal full power values.

P The Technical Specification wording was derived using the EPRI

" Guidelines for Permanent BWR Hydrogen Water Chemistry Installations, 1987 Revision".

BRUNWICK - UNIT 2 B 2-6 Amendment No. 171 l

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. , s LIMITING SAFETY SYSTEM SETTlWC BASES (Continued)

7. Drywell Pressure-High High pressure in the drywell could indicate a break in the nuclear process systems. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.

8. Scram Discharge Volume Water Level-High The scram discharge tank receives the water displaced by the motion of ,

the control rod drive pistons during a reactor scram. Should this tank fill up to a point where there is insuf ficient volume to accept the displaced water, control rod movement would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water f rom the movement of the rods when they are tripped.

9. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pre 6sure, neutron flux, and heat flux increases that would result from closure of the stop valves. *dith a trip eetting of 10% of valve closure f rom full open, the resultant increase in heat flux 1A such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass vailves remain closed. This scram is bypassed when the turbine steam flow is below that. corresponding to 30% of RATED THERMAL POWER, as measured by the turbine first-stage pressure.

10. Turbine Cont rol ulve rast Closure2 Centrol Oil Pressuro - Low Low turbine centrol valve hydraulic oressure will initiate the Select Rod Insert function and the preselected group of control rods will be f o lya inserted. Select Rod Insert is an operational aid designed to insert a predetermined group of control rods immediately following either a generator load rejection, loss *of turbine control valve hydraulic press'are, or by manual operator action using a switch on the R-T-C board. The assignment of control rods to the Select Rod Insert function is based on the start up and fuel warranty service associated with each control rod pattern, on RCS considerations, and on a dynamic function of both time and core patterns.

Approximately ten percent of the control rods in the reactor will be h assigned to the Select Rod Insert function by the operator. This selection will be accomplished by moving the rod scram test switch for those rods from the Normal position to the Select Rod Insert position.

BRUNWICK - UNIT 2 b P-7 Ame ndme nt No. 171 l

i LIMITING S AFETY SYSTEM SETTINGS BASES (Continued) c

10. Turbine Control Valve rast closure, cont rol Oil Pressure - Low (Continued)

Any rod selected for Select Rod Insert shall also have other rods in its notch group selected to ensure that the RSCS criteria of plus-minus one notch position equality is met when the rod pattern is greater than 50% ROD DENSITY and THERKAL POWER $ 20% of RATED THERMAL POWER. It is possible that a rod pattern within these limits may occur after the Select Rod insert function operates.

in order to reduce the number of reactor scrams, a 200 millisecond time delay, ref erenced from the low turbine control valve hydraulic pressure and Select Rod Insert signals, was incorporated to determine turbine bypass valve status via limit switches prior to initiating a reactor scram. If the turbine bypass valves opened in < 200 milliseconds, the reactor scram was bypassed.

It was found that during certain reload cycles the MCPR penalties involved with this time delay were more penalizing than the number of scrams savedt therefore, CP&L requested and received NRC approval to set this time at "0" in Amendment No. 14. With the timer set at "0", Select Rod Insert and RP3 trip will be initiated simultaneously.

The control valve closure time is approvinately twice as long as that for the stop valver which means that resulting transients, while similar, are less severe than for stop valve closure. No fuel damage occurs, and reactor system pressure does not exceed the safety relief valve setpoint. This is an anticipatory scram and results in reactor shutdown before any significant increase in pressure or neutron flux occurs. This scram is bypassed when turbine steam flow is below that corresponding to 30 percent of RATED THERMAL POWER, as measured by turbine first-stage pressure. )

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4 BRUNWICK - UNIT 2 B 2-8 Amendment No. 171

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