ML19309G347

Revised Pages to Tech Specs Re 800303 Application for Amend to License DPR-44 Correcting Typographical Errors & Discrepancy Re Scram Delay Times

ML19309G347
Person / Time
Site:	Peach Bottom
Issue date:	03/03/1980
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19309G344	List: ML19309G343 ML19309G347
References
NUDOCS 8005060202
Download: ML19309G347 (7)

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s 005 0 6 0$D PBAPS Unit 2 SAFETY LIMIT ____ LIMITING SAFETY SYSTEM SETTING  !

_B. Core Thermal Power Limit, B. APRM Rod Block __ Trip Setting I (Reactor Pressure < 800 psia) l When the reactor pressure is SRB f 0.66W + 42%

f 800 psia or core flow is where:

O less than 10% of rated, the core thermal power shall nct SRB = Rod block setting in exceed 25% of rated thermal percent of rated thermal power. power (3293 MWt)

W = Loop recirculation flow rate in percent of design W is 100 for core flow of 102.5 million Ib/hr or greater.

In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

SRB f (0.66 W + 42%) ( .FRP)

MFLPD where:

FRP =

fraction of rated thermal power (3293 MWt).

MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for all 8X8 fuel.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating -

value is less than the design value of 1.0, in which case the actual operating value C. Whenever the reactor is in the will be used.

shutdown condition with irradiated fuel in the reactor C. Scram and isolation-->538 in, above vessel, the water level shall reactor low water ~ vessel zero not be less than 17.1 in. above level (0" on level the top of the normal active instruments) fuel zone.

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PBAPS UNIT 2 2.1.A BASES (Cont'd.)

The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5-decades are covered by the IRM by means of a range switch and the 5-decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be a 120 divisions for that range; likewise, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

In order to assure that the IRM provided adequate protection I against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. This conditjon exists at quarter rod density. Additional conservatism was ti,<en in this analyses by assuming that the IRM channel clos < . s to the withdrawn rod is bypassed. The results of this dna.ysis show that the reactor is scramed and peak power limited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in-sequence and provides backup protection for the APRM.

B. APRM Rod Block Trip Setting

• I The APRM system provides a control rod block to avoid conditions which would result in an APRM scram trip if allowed to proceed.

l The APRM rod block trip setting, like the APRM scram trip setting, is automatically varied with recirculation loop flow rate. The flow variable APRM rod block trip setting provides margin to the APRM scram trip setting over the entire  !

recirculation flow range. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gain.

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[ PBAPS UNIT 2 4

j 3.3 and 4.3 BASES (Cont'd)

C. Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e.,

to prevent the MCPR from becoming less than the fuel cladding integrity safety limit. Analysis of the limiting power ,

transients shows that the negative reactivity rates resulting from the scram vith the average response of all drises as given in the above Specification, provides the required protection.

The numerical values assigned to the specified scram performance are based on the analysis of data from other BWR's with control rod drives the same as those on Peach Bottom.

4 The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod drives especially if the number of drives exhibiting such scram times exceeds one control rod of a (5x5) twenty-five control array.

4 In the analytical treatment of the transients, which are assumed l to scram on high neutron flux, 340 milliseconds are allowed ,

between a neutron sensor reaching the scram point and the start 1 of negative reactivity insertion, the 340 milliseconds used in l the analyses consists of 140 milliseconds for sensor and circuit delay and 200 millisecond to start of control rod motion. The 200 millisecondu are included in the allowable scram insertion times specified in Specification 3.3.C. In addition the control rod drop accident has been analyzed in NEDO-10527 and its supplements 1 & 2 for the scram times given in Specification 3.3.C.

I j Surveillance requirement 4.3.C was originally written and used as

) a diagnostic surveillance technique during pre-operational and startup testing of Dresden 2 & 3 for the early discovery and i iaentification of significant changes in drive scram performance following major changes in plant operation. The reason for the l application of this surveillance was the unpredicatable and . 1 degraded scram performance of drives at Dresden 2. The cause of j the slower scram performances has been conclusively ,

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PBAPS UhIT 2 Table 3.5-2 OPERATING LIMIT MCPR VALUES AS DETERMINED FROM INDICATED TRANSIENTS FOR VARIOUS CORE EXPOSURES MCPR Operating Limit Fuel Type For Incremental Cycle 5 Core Avarage Exposure BOC to 1000 MWD /t 1000 MWD /t before EOC Before EOC To EOC 8x8 1.28 1.31 8x8R & LTA 1.28 1.31 P8x8R 1.30 1.33 i

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PBAPS UNIT 2 TABLE 3.5-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS PLANT PARAMETERS:

Core Thermal Power 3440 MWt which corresponds to 105% of rated steam flow Vessel Steam Output 14.05 x 106 lbm/h which correspcnds to 105% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area For Large Breaks -

Discharge 1.9 fta (DBA)

Suction 4.1 fta Assumed Number of Drilled Bundles 360 l

FUEL PARAMETERS: Peak Technical Initial l Specification Design Minimum l Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geometry (KW/ f t) Factor Ratio 7x7, Type 2 7x7 18.5 1.5 1.2 7x7, Type 3 7x7 18.5 1.5 1.2 8x8, Type H 8x8 13.4 1.4 1.2 8x8, Type L 8x8 13.4 1.4 1.2 8x8R/LTA 8x8 13.4 1.4 1.2 P 8x8R 8x8 13.4 1.4 1.2 Type i

P8DRB284H i

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P 8x8R 8x8 13.4 1.4 1.2  !

Type P8DRB285 i

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0 PBAPS Unit 2 LIMITING CONDITIONS FOR OPERATION ' SURVEILLANCE REQUIREMENTS 3.6.A Thermal and Pressurization 4.6.A. Thermal and Pressurization Limitations (Cont'd) Limitations (Cont ' d)

Figures 3. 6.1,3. 6. 2 and Selected neutron flux 3.6.3 will be updated to specimens shall be account for radiation removed

• damage prior to 9 effective full power and tested to years of operation. experimentally verify or adjust the calculated values of inegrated neutron flux that are used to determine the RT for Figure 3.6.4 NDT

3. The reactor vessel head bolting 3. When the reactor vessel head studs shall not be under bolting studs are tensioned tension unless the temperature and the reactor is in a cold ,

of the vessel head flange Condition, the reactor

. and the head is greater vessel shell temperature than 1000F. immediately below the head flange shall be permanently recorded.

4. The pump in an idle recircu- 4. Prior to and during startup lation loop shall not be of an idle recirculation started unless the tempera- loop, the temperature of the tures of the coolant within reactor coolant in the the idle and operating recir- operating and idle loops culation loops are within shall be permanently logged.

500F of each other.

5. The reactor recirculation 5. Prior to starting a recir-pumps shall not be started culation pump, the reactor unless the coolant tempera- coolant temperatures in the tures between the dome and dome and in the bottom head -

the bottom head drain are drain shall be compared and within 1450F. permanently logged.

• Specimen 1 7-9 EFPY 2 15-18 EFPY 3 Standby

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l PBAPS UNIT 2

5. 0 MAJOR DESIGN FEATURES 5.1 SITE FEATURES The site is located partly in Peach Bottom Township, York County, partly in Drumore Township, Lancaster County, and partly in Fulton Township, Lancaster County, in southeastern Pennsylvania on the westerly shore of Conowingo Pond at the mouth of Rock Run Creek. It is about 38 miles north-northeast of Baltimore, Maryland, and 63 miles west-southwest of Philadelphia, Pennsylvania. Figures 2.2.1 through 2.2.4 of the FSAR show the site location with respect to surrounding communities.

5. 2 REACTOR A. The core shall consist of not more than 764 8X8 fuel assemblies. 8 x 8 fuel assemblies shall contain 62 or 63 fuel rods.

B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B4C) compacted to approximately 70% of the theoretical densityg except as described in Section 5.2.C below.

C. Two test control rods (maximum) with up to 12 boron carbide (B4C) pins per control rod replaced with solid hafnium metal control pins may be substituted for two B4p control rods (Section 5.2.B above) .

5. 3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR. The applicable design codes shall be as described in Table 4.2.1 of the FSAR.

5. 4 CONTAINMENT A. The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR. The applicable -

design codes shall be as described in Appendix M of the FSAR.

B. The secondary containment shall be as described in Section 5.3 of the FSAR.

C. Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.

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