ML19294C289

Application to Amend OL DPR-44 to Permit Prepressurized Retrofit 8x8 Fuel Design.Includes Changes to Identify Cycle 5 Operating Limits for All Fuel Types & Modify Average Power Range Monitor & rod-block Monitor Setpoint Equations

ML19294C289
Person / Time
Site:	Peach Bottom
Issue date:	03/03/1980
From:	Bauer E, Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19284A871	List: ML19284A870 ML19290E315 ML19290E318 ML19290E322 ML19294C288 ML19294C289
References
NUDOCS 8003100211
Download: ML19294C289 (53)

Text

4 4

BEFORE Til E UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of Docket Nos. 50-277 Pl!I L AD EL Plil A ELECTRIC C O I1P A N Y APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE DPR-44 Edward G.

Bauer, Jr.

Eugene J.

Bradley 2301 Market Street Philadelphia, Pennsylvania 19101 Attorneys for Philadelphia Electric Company 8003100 2'11

4 BEFORE T !! E UNITED STATES NUC' CAR R F.G U L ATO RY C O MtII S S IO !!

In the !!a t t e r of Docket No. 50-277 PiiI L AD EL PIII A ELECTRIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE DPR-44 Philadelphia Electric Company, Licensee under Facility Operating License DPR-44 for Peach Bottom Atomic Power S ta t i on Unit No.

2, hereby requests that the Technical Specifications incorporited in Appendix A of the Operating License be amended by revising ce rt ain s ect i ons as indicated by a vertical bar in the margin of attached pages iv, 1,

3, 4,

7, 10, 11, 15, 18, 19, 20, 21, 31, 33, 35, 37, 40, 54, 73, 74, 91, 92, 108, 111, 115, 119, 120, 122, 133a, 133c, 140, 140a, 140b, 140c, 140d, 140e, 142b, 142e, ;44, 152a, 157, 241, and by adding pages 142g and 142h.

s s

Page 242 is included because of the re-distribution of natt cial on the revised pages.

The changes to the Technical Specifications are being requested to: 1) accomnodate the f ou rt h refueling of the Peach Bottom Unit 2 reactor, (2) identify the operating limits for all fuel types, including reload 4 pre-pressurized retrofit 8X8 fuel (P 8X8R), for cycle 5 operation (3) modify the APRM and RBM se t p oi nt equations, (4) delete the fuel densification p owe r spiking penalty for 8X8 fuel (5) d e le t e the reactor vessel pressure operating limit, (6) permit rea ct or operation with up to two control rods containing hafnium control pins, (7) increase the capability of the Standby Liquid Control System (8) delete reference to the results of a specific General Electric reload safety evaluation, and (9) accommodate extended e xp os u re Lead Test Assemblies (LTA).

An analysis of the safety c on s i de r a t i on s involved in the reactor refueling and the cycle 5 operating limits for all fuel type are set f ort h in a document entitled " Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit 2 Reload 4"

(NEDO 24237), which is filed herewith and I

in c orp ora t ed herein by reference.

The changes requested herein would p er mi t the IcAding of the p rep res su riz ed retrofit 8X8 fuel design.

A safety evaluation report concerning the use of the prepressurized retrofit 8X8 fuel is provided by General Electric document entitled " Generic Reload Fuel Application" ( N ED E-2 4 0 l l-P- A),

s August, 1978.

This docunent was reviewed and approved (Ar-il, s

1979) by the NRC for reference in connection with P 8XdR reload licensing applications.

The Licensee p rop os es that the APRM and RBM setpoint equations shown on pages 10, 11. 37, and 73 be modified.

A safety evaluation covering the APRM and RBM se tp oin t equation modifications has been provided in Section 3.5 of the Nuclear Regulatory Commission's Safety Eva lu a t i on Report supporting Amendments 35, 32 and 9to Licensee Numbers DPR-33, 52, and 68 for Browns Ferry Nuclear Plant Units 1,

2 and 3 respectively.

The safety a na ly sis provided in NEDO-24237 includes the effects of densification p ow e r spiking on allowable linear heat generation rate ( LilG R).

This provides the bases for de le t ing the f uel densification p ow e r spiking penalty factor for all 8X8 fuel.

The amendment request p rop os es deletion of the ma xi mu m operating dome p res su re linit of 1020 psig.

The operating limit w ou l d therefore be limited by the react or vessel high pressure scram s e t p oin t of 1055 psig.

The results of a sensitivity study performed by Cencral Electric s h ow that the peak vessel p ressu re increase is only 13 psi (1299 psig to 1312 psig) f or an increase in the assumed initial pressure from 1020 psig to 1055 psig.

The design bases over pressure transient analysis provided in !!?DO-24237 demonstrates a margin of 76 psi to the vessel code limit.

This margin is far in excess of a possible 13 psi additional rine resulting f rom an initial dome p res su re increase over 1020 paig.

There is added conservatism by virtue of the fact that the trend s

s is for the resulting pressure peak increase to be much l e n t, than directly p r op or t i on a l to the increase in initial dome p re s s u re.

The changes requested herein would permit reactor op e ra t i on s with a maxinun of two hafnium control rods installed.

The hafnium control rod will contain up to twelve solid hafnium control pins replacing the boron carbide control pins.

The p u rp os e of the hafnium control rod installation and operation is to obtain inf orma ti on regarding p e r f ormance of hafnium when e xp os ed to the boiling water reactor environment.

It is intended that the haf nium c ont rol rod be installed du ring the Peach Bottom Unit No. 2 refueling outage scheduled for tiarch, 1980.

An analysis of the safety considerations involved with the installation, operation, and inspection of the hafnium control rod are set forth in a document entitled "P r op os e d Peach Bot t om At omic Power Station Unit 2 Alternate Absorber Control Blade Test Program" (NEDO 24231), January 1980, which is filed herewith and incorp orat ed herein by reference.

The Technical Specification bases 3.4.A state that the Standby Liquid Control System (SLCS) has the capability of bringing the react or 3 0% A k subcritical.

The reload licensing a na ly s i s for Peach Bot t on 2 ti e l oa d 4 indicated that the S L C t.

with its present capability (600 ppm boron) would only bring the reactor 2.6 % A k suberitical.

Therefore, the Licensee propcacs increasing the SLCS capability from 600 ppm to 660 ppm boron.

At this increased boron concentration, the safety analysis provided i

l I

-4

s in NEDO-24237 demonstrates that the SLCS will bring the co:e te s

at least 3.7 % A k subcritical.

The changes requested herein delete references in the Technical Specifications to the General Electric reload safety evaluation document for a specific fuel cycle, and the results of that analysis for a specific fuel cycle, in order to permit the acceptability of future fuel cycles to be determined in accordance with the provisions of 10 CFR 50.59 The changes requested herein provide for the extended operation of the 4 Lead Test Assenblies (LTA).

The extended op e ra ti on is necessary to obtain inf orma tion for a p rogram which is directed toward i mp rovi ng uranium utilization.

Since the p rop os ed changes to the Technical S p e ci f i ca t i ons do not involve a significant hazards consideration, pu rs u an t to 10 CFA 170.22, Philadelphia Electric Company, for fee p u rp os es, p rop os es that the Application for Amendment be considered a Class III Amendment.

The Plaat Operation Review Committee and the Operation and Safety Review Conmittee have r e vi ew e d these p r op os e d changes to the Technical Specifications, and have concluded that they do not involve an unreviewed safety q u es t i on or a significant hazard consideration; and will not endanger the health and safety of the public.

Respectfully submitted, PIIIL ADELPilI A ELEC,TRIC COMP NY

((h(k.<IEL

./

..i By,_

~Vice' President [

-S-

s COMMONWCALTil 0F PENNSYLVANIA ss.

COUNTY OF P ilI L AD EL Pili A S.

L.

Daltroff, being first duly sworn, deposes and says:

That he is Vice President of Philadelphia Electric Company, the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating Licenses and knows the contents t here of ; and that the statements and matters set forth therein are true and c or re ct to the best of his knowledge, information and belief.

f /, ~/

t

_. +,

Subscribed and sworn to thisc[

day before me l

O.

of I

b t

N ta y Public r,-.~.,.,r-efj c.

tL4,.

J.C'O 1

l

s Cf.RTIFICATE OF SERVICE I certify that service of the f oregoing Application was made upon the Board of Supervisors, Peach Bot t om Towns hip, York County, Pennsylvania, by mailing a copy the re of, via f i r s t. -c l a s s mail, to Albert R.

Steele, Chairman of the Board of Supervisors, R.

D.

No.

1, Delta, Pennsylvania 17314. upon the Board of Supervisors, Fulton Township, Lancaster County, Pennsylvania, by mailing a copy there of, via first-class mail, to George K.

Brint on, Chairman of the Board of Super,isort, Peach Bottom, Pennsylvania 17563; and upon the Board of Supervisors, Drumore Township, Lancaster County, Pennsy'.vania, by n.a i li n g a copy thereof, via f i rs t -c la s s raa t ]

.o Wilmer P.

Bolt on, Chairman of the Board of Superviscrs, R.

D.

No.

1, ll ol t w o o d, Pennsylvania 17532; all this Third day of

March, 1980.

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[Eugen'e J.

Bradley

't Attorney for Philadelphia Electric Company I

PBAPS Unit 2 LIST OF FIGURFS Fiqure Title Page 1.1 - 1 APRM Flow Bias Scram Relationship To 16 Normal Operating Conditions 4.1.1 Instrument Test Interval Determination 55 Curves 4.2.2 Probability of System Unavailebility 98 Vs. Test Interval

3. 4.1 Re.luired Volume and Concentration of 122 Standby Liquid Control System Solution

3. 4. 2 Required Temperature vs. Concentration 123 for Standby Liquid Cont.ol System Solution

• Figures 3.5.1 A and 3.5.1 B (7x7 Fuel) deleted.(PB2 Cycle 5-all 8x8 core) 3.5.1.C MAPLHGR Vs. Planar Average Exposure, 142b Unit 2, 8x8 Fuel, Type H 80 mil 6 100 mil channels 3.5.1.D MAPLHGR Vs. Planar Average Exposure, 142c Unit 2, 8x8 Fuel, Type L, 100 mil channels 3.5.1.E Kf Factor Vs. Core Flow 142d 3.5.1.F MAPLHGR Vs. Planar Average Expocure, 142e Unit 2, 8x8 LTA Fuel, 100 mil channels 3.5.1.G MAPLHGR Vs. Planar Average Exposure, 142f Unit 2, 8x8R Fuel, Type 8DBB284, 100 mil channels

3. 5.1. H MAPLHGR Vs. Planar Average Exposure, 142g Unit 2, P 8X8R Fuel, Type P8DRB285, 100 mil channels 3.5.1.I MAPLHGR vs. Planar Average Exposure, 1h2h Unit 2, P 8x8R Fuel, Type P8DR8284 H, 80 mil 6 100 mil channel & 120 mil channels 3.6.1 Minimum Temperature for Pressure Tests 164 such as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup 164a or Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation 164b (Criticality) 3.6.4 Transition Temperature Shif t vs. Fluence 164c 6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant Operation 245

-iv-

PBAPS Uti1T 2 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications reay be achieved.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud.

Normal control rod movement with the control drive hydraulic system is not defined as a core alteration.

Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a core alteration.

Channel a channel is an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic.

A channel terminates and loses its identity where individual channel outputs are combined in logic.

Cold Condition - Reactor coolant temperature equal tO or less than 2120F.

Cold shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 2120F, and the reactor I

vessel is vented to atmosphere.

Critical Power Ratio (CPR)

- The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the GEXL correlation. (Ref erence NEDO-10958)

Engineered safequard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in I

response to accidents.

Fraction of Limiting Power Density (FLPD) - The ratio of the linear heat generation rate (LHGR) existing at a given location to the design LilGR for that bundle type.

Functional Tests - A functional test is the manual operation or initiation or a system, subsystem, or component to verify that it l

functions within design tolerances (e.g.,

the manual start of a 1

PDAPS UNIT 2 1.0 DEFINITIONS (Cont'd) the automatic protective action at a level such that the saf ety limits will not be exceeded.

The region between the safety limit and these settings represent margin with normal operation lying below these settings.

The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

Logic - A logic is an arrangement of relays, contacts and other components that produce a decision output.

(a) Initiating - A logic that receives signals from channels and produces decision outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action.

Logic System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent.

Where practicable, action will go to completion; i.e.,

pumps will be started and valves operated.

Maximum Fraction of Limiting Power Density (MFLPD) - The Maximum Eraction of Limiting Power Density (MFLPD) is tbc highest value existing in the core of the Fraction of Limiting Power Density (FLPD).

Minimum Critical Power Iatio (MC PR) - The minimum in-core critical power ratio corresponding to the most limiting fuel auuembly in the core.

Mode of Operation - A reactor mode switch selects the proper interlocks for the operational status of the unit.

The following are the modes and interlocks provided:

Refuel Mode, Run Mode, Shutdown Mode, Startup/ Hot Standby Mode.

Operable - A system or component shall be considered operable when it in capable of periorming its intended function.in its required manner.

PBAPS UNIT 2 1.0 DEFINITIONS (Cont' d)

Operating - Operating means that a system or component is parforming its intended functions in its required manner.

Operating Cycle - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1.

All non-automatic containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

These valves may be opened to perform necessary operational activities.

2.

At least one door in cach airlock is closed and sealed.

3.

All automatic containment isolatior. valves are operable or deactivated in the isolated position.

4.

All blind flanges and manways are closed.

l Protective Action

- An action initiated by the protection system when a limit is reached.

A protective action can be at a channel or system level.

l Protective Function - A system protective action which results I

from the protective action of the channels monitoring a I

particular plant condition.

g Rated Power - Rated power refers to operation at a reacter power of 3,293 MWt; this is also terned 100 percent power and is the maximum power level authorized by the operating license.

Rated l

steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.

l l

-u.

l

PDAPU UN !'I 2

1. 0 DEFINITIONS (Cont'd) operable or are tripped, then

t. hey shall be per formed prior to returning the system to an operable status.

Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling.

Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function.

A trip system may require one or more instrument channel trip signals related to one or more plant parameteru in order to initiate trip system action.

Initiation or protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

l l

l 1

4 PBAPS Unit 2 SAFETY LIMIT LIMITING SAFETY SY ST E51 SETTING 2.1.A (Cont ' d)

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

S < (0.66 W + 54 %) ( FRP )

MFLPD

where, FRP = fraction of rated thermal power (3293 MWt)

MFLPD = maximum fraction of l

limiting power density where the limiting power density is 13.4 KW/f t f or 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 will be used.

2.

APRM--When the reactor mode switch is in the STARTUP position, the APRM scram shall be set at less than or equal to 15 p*arcent of rated power.

3.

IRM--The IRM scram shall be set at less than or equal to 120/125 of full scale.

4.

When the reactor mode switch is in the STARTUP or RUN position, the reactor shall not be operated in the natural circulation flow mode.

I PBAPS Unit 2 SAFETY LIMIT LIMITING S AF ETY SY ST EM SETTINC B.

Co re Thermal Power Limit B.

APRM Rod Block Tri p Set tir.3 JR ea ct or_P res sure < 800 osia)

SRB i 0.66W + 42%

When the reactor pressure is

< 800 psia or core flow is where:

less than 10% ot rated, the core thermal power shall not SRB =

Rod block setting in exceed 25% of rated thermal percent of rated thermal power.

power (3 293 MWt) rateinpercentofdesigln Loop recirculation flow W

=

W is 100 for core flow of 102. 5 million lb/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 $ (0.66 W + 4 2%) ( FRP)

MFLPD where:

FRP = fraction of rated thermal power (3 29 3 MWt).

MFLPD = maximum fraction of limiting oower density where the limiting power density is and 13.4 KW/ft for all 8X8 fuel The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operatina value is less thar. 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--1538 in, above ve ss el, 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) f uel zone.

4 PBAPS U:41 r 2

,1.1.C BASES (Cont'd.)

However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design.

The concept of not approaching a Safety Limit, provided scrum signals are operable, is supported by the extensive plant safety analysis.

The computer provided with Peach Bottom Unit 2 has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.

This program also indicates when the scram setpoint is cleared.

This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.

Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied upon to determine if a safety Limit has been violated.

D.

Reactor Water Level (Shutdown Condition)

During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat.

If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced.

This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.

The core can be cooled sufficiently should the water level be reduced to two-thirds the core height.

Establishment of the safety liait at 17.7 inches above the top of the fuel provides adequate margin.

This level will be continuously monitored.

E.

References 1.

General Electric BWR Thermal Analysis Basis (GETAB) :

Data, Correlation and Design Application, January 1977 (NEDO-1095 8-A).

2.

Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systerm Department, June 1974 (NEDO-2034 0) 3.

" General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-24011-P-A.

PBAPS Ut3IT 2 2.1 BASES (Co n t ' d. )

For analyses of the thermal consequences of the transients a MCPE equal to or greater than the operating limit MCPR given in Specification 3.5.K is conservatively assumed to exist prior to initiation of the limiting transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Steady state operation without forced recirculation will not be permitted.

The analysis to support operation at various power I

and flow relationships has considered operation with either one or tuo recirculating pumpu.

l In summary:

1.

The abnormal operational transients were analyzed to a power i

level of 3440 MWt (104.5% rated power) to determine operating limit MCPR's.

ii. The licensed maximum power level is 3293 MWt.

iii. Analyses of transients employ adequately conservative values

'of the controlling reactor parameters.

iv.

The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for the parameters.

The bases for individual trip settings are discussed in the following paragraphs.

A.

Neutron Flux Scram The Average Power Range Monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (3293 MWt).

i I

because fission chambers provide the basic input signals, the l

APRM system responds directly to average neutron flux.

During transie nts, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, d uring abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrate that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage.

Therefore, the use of flow referenced scran trip provides even additional margin.

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

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached.

The APRM scram trip setting was determined by an analysis of nargins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious scrams which have an adverse ef f ect on reactor saf ety because of the resulting thermal stresses.

Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted to assure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power.

The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the MFLPD is greater than the fraction of rated power' (FRP).

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than the fuel cladding integrity safety limit when the transient is initiated from MCPR greater than the operating limit given in Specification 3.5.K.

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated porer provides adequate thermal margin between the setpoint and the Safety Limit, 25 percent of rated.

The margin is adequate to accommodate anticipated maneuvers associated with power plaat startup.

Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coe f ficientt are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the Rod Worth Minimizer and Rod Sequence Control System.

Worth of individual rods is very low in a uniform rod pattern.

T hus,

of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power is very slow.

Generally, the neat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the Safety Limit.

The 15 percent APRM scram remains active until the mode switch is placed in the RUN position.

This switch occurs when the reactor pressure is greater than 850 psig.

PBAPS UfHT 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 che instrument were on range 1, the scram s etting 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.

Thas, 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 IM4 scram would result in a reactor shutdown well before any Safety Limit is exceeded.

In order to assure that the IRM provided adequate protection 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 condition exists at quarter rod density.

Additional conservatism was taken in this analyses by assuming that the IRM channel closest to the withdrawn rod is bypassed.

The results of this analysis show that the reactor is scramed and peak power limited to one percent of rated power, thus twintaining MCPR above the fuel cladding integrity safety limit.

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

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 purgin 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.

PBAPS UNIT 2 2.1 BASES (Cont ' d. )

C.

Reactor Water Low Level Scram and Tsolation (Except Main Steamlines)

The set point for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease.

The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the fuel cladding integrity safety limit in all cases, and system pressure does not reach the safety valve settings.

The scram setting is approximately 31 in. below the normal operating range and is thus adequate to avoid spurious scrams.

D.

Turbine Stop Valve Closure Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.

With a scram trip setting of less than or equal to 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed.

This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressure.

E.

Turbine Control Valve Scram The turbine control valve f ast closure scram anticipates the pressure, neutron flux and heat flux increase that could result from fast closure of the turbine control valves due to a load rejection exceeding the capacity of the bypass valves or a failure in the hydraulic control system which results in a loss 1

of oil pressure.

This scram is initiated from pressure switches in the hydraulic control system which sense loss of oil pressure due to the opening of the faut acting solenoid valves or a failure in the hydraulic control system piping. Two turbine first stage pressure switches for each trip system initiate automatic bypass of the turbine control valve fast closure scram when the J

first stage pressure is below tnat required to produce 30% of rated power.

Contol valve closure time is approximately twice as long as that for stop valve closure.

l PbAPS UN 1'1 2

1. 2 BAS ES The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products.

It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1325 psig as measured by the vessel steam space pressure indicator assures not exceeding 1375 psig at the lowest elevation of the reactor coolant l

system.

The 1375 psig value is derived from the design pressures of the reactor pressure vessel (1250 psig at 575 degrees F) and coolant system piping (suction piping:

114P psig at 562 degrees F; discharge piping:

1326 psig at 562 degrees F).

The pressare safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:

ASME Boiler and Pressure Vessel Code,Section III for the pressure vessel and ANSI B31.1.0 for the reactor coolant system piping.

The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% X 1250 = 1375 psig), and the ANSI Code permits pressure transients up to 20% over the design pressure (120%

X 1148 = 1378 psig; 120% X 1326 = 1591 psig).

A safety limit is applied to the Residual Heat Removal System (RilRS) when it is operating in the shutdown cooling mode.

At this time it is included in the reactor coolant system.

l l

PBAPS UNIT 2 2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Peach Bottom Atomic Power Station has been sized to meet two design bases.

First, the total capacity of the safety / relief valves and safety valves has been established to meet the overpressure protection criteria of the ASME Code.

Second, the distribution of this l

required capacity between safety valves and relief valves has been set to meet design basis 4.4.4.1 of subsection 4.4 of the FSAR which states that the nuclear system safety / relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.

The detaile of the analysis which shows compliance with the ASME Code reqL.rements are presented in subsection 4.4 of the FSAR and l

the Reactor Vessel Overpressure Protection Summary Technical l

Report submitted in Appendix K.

Eleven safety / relief valves and two safety valves have been installed on Peach Bottom Unit 2.

The analysis of the worst overpressure transient, is provided in the Supplemental Reload Licensing Submittal and demonstrates margin to the code allowable overpressure limit of 1375 psig.

The analysis of the plant isolation transient is provided in the Supplemental Reload Licensing submittal Safety Evaluation and demonstrates that the safety valves will not open.

The safety / relief valve settings satisfy the Code requirements that the lowest valve set point be at or below the vessel design pressure of 1250 psig.

These settings are also sufficiently above the normal operating pressure range to prevent unnecessary cycling caused by minor transients.

The design pressure of the shutdown cooling piping of the Residual Heat Removal System is not exceeded with the reactor vessel steam dome less than 75 psig. l

PBAPS Unit 2 LI_gTIN"> CONDITIONS FOR DPERSTION SURVEILLANCE REQUIR D1ENT 9 3.1 REACTOR PROPECTION SYSTEM 3.1 REACTOR PROTECTION SYS TEM Applica bility:

Aoplicability:

Applies to the instrumenta-Applies to the surveillance tion and associated devices of the instrumentation and which initiate a reactor associated devices which sc ra m.

initiate reactor scram.

Objective:

Obiective:

To assure the operability To specify the type and of the reactor protection frequency of surveillance system.

to be applied to tne pro-tection instrumentation.

Specification:

Soecification:

The setpoint, minimum A.

Instrumentation systems number of trip systems, shall be functionally and minimum number of tested and calibrated instrument channels that as indicated in Tables must be operable for each 4.1.1 and 4.1.2 position of the reactor respectively.

mode switch shall be as given in Table 3.1.1.

The B.

Daily during reactor designed system resoonse power operation, the I

times from the opening of maximum fraction of the sensor contact up to limiting power density and including the opening shall be checked and ot the trip actuator the SCRAM and APRM Rod contacts shall not exceed Block settings given 100 milli-seconds by equations in Specifi-cation 2.1.A.1 and 2.1.B shall be calculated if maximum fraction of limiting power density exceeds the fraction of rated power.

. l

TABLE 3.1.1 R EACTOR PROTECTIOtl SYSTEM (SCRAM) IllSTRUMEtiTATIOti PEQUIREME!!T Mininun !1o.

Modes in which flumber of of Operable Function Must be Instrument Instrument Trip Level Operable Channels Action Channels Trip Function Settinq

_ _ _ _ Refuel _____

Provided (1) per Trip Startup Run by Design Systen (1)

(7) 1 Mode Switch In X

X X

1 Made Switch A

Sh u t ' lown (4 Sections) 1 Manuai Scran X

X X

2 Instrument A

Channels 3

IRM friqh Flux 5120/125 of Full X

X

( 5) 8 Instranent A

Scale Channels 3

IRM Inonerative X

X (5) 8 Instrument A

u channels 2

APR" I!igh Flux

(. 66W+5 4) FRP/MFLPD X

6 Instrument A or B (12) ( 13)

Channels 2

APPM I nope ra t ive (11)

X X

X 6 Instrument A or B channels 2

APRM Downscale 22.5 Indicatel (10) 6 Instrument A or B on Scale Channels 2

A P R.'-1

!!igh Flux 5154 Power X

X 6 Instrument A

in Startup Channels 2

liigh Reactor i1055 psig X (9)

X X

4 Instrument A

e Pressure Channels d

H 2

Iligh Drywell 52 psig X (8)

X (8)

X 4 Instrument A

N Pressu re Ch ara.e i s 2

Reactor Iow 20 in. Indicate.1 X

X X

4 Instrument A

Water IAvel Level N'

Channels

9 PBAPS UNIT 2 l

NOTES FOR TABLE 3.1.1 (Cont'd)

10. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high.

11. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement.

12. This equation will be used in the event of operation with a maximum fraction of limiting power density (MPLPD) greater than the fraction of rated power (FRP), where:

FRP = fraction of rated thermal power (3293MWt).

MFLPD = maximum fraction of limiting power density where the limiting power density is 12.4 KW/ft for all 8x8 fuel.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless tne actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

W = Loop Recirculation flow in percent of I

design.

W is 100 for core flow of l

102.5 million lb/hr or greater.

I Trip level setting is in percent of rated power (3293 MWt).

13. See Section 2.1.A.1.

1 l

l l l

PBAPS UllIT 2 l

4.1 BAS ES (Cont'd)

Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.

For thone devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4% month; i.e., in the period of a month a maximum drift of 0.4% could occur, thus providing for adequate margin.

For the APRM systems, drif t of electronic aparatus is not the only consideration in determining a calibration frequency.

Change in power distribution and loss of chamber sensitivty dictate a calibration every seven days.

Calibration on this frequency assures plant operation at or below thermal limits.

A comparison of Tables 4.1.2 and 4.1.3 indicates that two instrument channels have not been included in the latter tables.

These are:

mode switch in shutdown and manual scram.

All of the devices or sensors associated with these scram functions are simple on-off switches, and, hence, calibration during operation is not applicable.

B.

The MFLPD is checked once per day to determine if the APRM scram requires adjustment.

Only a small number of control rods are moved daily and thus the MFLPD is not expected to change significantly.

Therefore, a daily check of the MFLPD is adequate.

i 1

l The sensitivity of LPRM detectors decreases with exposure to l

neutron flux at a slow and approximately constant rate.

This is compensated for in the APRM system by calibrating twice a week using heat balance data and by calibrating individual LPRM's every 6 weeks, using TIP traverse data.

TABLE 3 2.0 INSTRUME:.'TATION THAT IITITIATES CONTROL ROD BLOCKS Mini := No.

Of Operable Nurber of Instrument Instr =ent Inst'rument Trip Level Setting Channels Provided Acticn Channels Per by Design Trip System 2

APPli Upscale (Flow f(0.66W+h2)x FRP 6 Inst. Channels (1) 2)

Biased)

MFLPD 2

APPJi Upscale (Startup 3,12?S 6 Inst. Channels (1)

Mode) 2 APPli I)ownscale 12.5 indicated on 6 Inst. Channels (1) scale O

~(0.66W+L1)x FRP 1 (7)

Rod Block Monitor u

4.

(FlowBiased)

MFLPD(2) 2 Inst. Channels (1) 1 (7)

Rod Block Monitor L2.5 indicated on 2 Inst. Channels (1)

Downscale scale 3

IFli Downscale (3) 2,2.5 indicated on 8 Inst. Channels (1) scale 3

IPli Detector not in (8) 8 Inst. Channels (1)

Startup Position 3

IRM Upscale 1108 indicated on 8 Inst. Channels (1) scale z.

2 ($)

SRM Detector not in (h) h Inst. Channels (1)

Q Startup Position p

2 (i;) (6)

SF; Upscale 3,10 counts /sec.

h Inst. Chdnnels (1) s e

F e

e

PBAPS UNIT 2 NOTES FOR TABLE 3.2.C 1.

For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.

The SRM and IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode.

If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped.

If the first column cannot be met for both trip systems, the systems shall be tripped.

2.

This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP) where:

FRP = fraction of rated thermal power (3293 MWt) 1 MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for all 8x8 fuel.

l 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 will be used.

W = Loop Recirculation flow in percent of design.

W is 100 for core flow of 102.5 million lb/hr or greater Trip level setting is in percent of rated power (3293 MWt).

3.

IRM downscale is bypassed when it is on its lowest range.

(

4.

This f unction is bypassed when the count rate is 2 100 cps.

I S.

One of the four SRM inputs may be bypassed.

6.

This SRM function is bypassed when the IRM range switches are on range 8 or above.

7.

The trip is bypassed when the reactor power is 5 30%.

8.

This function is bypassed when the mode switch is placed in Run. l

PBAPS UNI: 2 3.2 BAS ES (Cont'd)

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 850 psig.

The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not in the RUN Mode is less severe than the loss of feedwater analyzed in section 14.5 of the FSAR, therefore, closure of the Main Steam Isolation valves for thermal transient protection when not in RUN mode is not required.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.

Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1 out of 2 logic.

Temperature is monitored at four (4) locations with foar (4) temperature sensors at each location.

Two (2) sensors at each location are powered by "A" DC control bus and two (2) by "B"

DC control bus.

Each pair of sensors, e.g.,

"A" or "B" at each location are physically separated and the tripping of either "A" or "B" bus sensor will actuate HPCI isolation valves.

The trip settings of $300% of design flow for high flow and 2000F for high I

temperature are such that core uncovery is prevented and fission l

product release is within limits.

l The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI.

The trip setting of $300% for high flow and 2000F for temperature are based on the same criteria as the HPCI.

The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar to that for the HPCI.

The trip settings are such that core uncovery is prevented and fission product release is is within limits.

The instrumentation which initiates CSCS action is arranged in a dual bus system.

As for other vital instrumentation arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

l The control rod block functions are provided to present excessive control rod withdrawal so that MCPR does not decrease to the fuel cladding integrity safety limit.

The trip logic for this function is 1 out of n:

e.g.,

any trip on one of 6 APRM's, 8 IRM's, or 4 SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration.

This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

PBAPS Uti1T 2 a.2 BASES (Cont ' d)

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow.

The APRM provides gross core protection:

i.e.,

limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.

The trips are set so that MCPR is maintained greater than the f uel cladding integrity safety limit.

The RBM rod block function provides local protection of the core; i.e., the prevention of hoiling transition in the local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

The IRM rod block function provides local as well as gross core protection.

The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in the control rod motion and thus, control rod motion is prevented.

The downscale trips are set at 2.5 indicated on scale.

The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety.

The flow comparator must be bypassed when operating with one recirculation water pump.

The refueling interlocks alsa operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.

The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.

The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.

The trip settings given in the specification are adequate to assure the above criteria are met.

The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e.,

only one instrument channel out of service.

Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off gas line.

Isolation is initiated when both instruments reach their high trip point when one has an upscale. l

PBAPS UNIT 2 a.3 and 4.3 BASES (Cont ' d. )

B.

Control Rods 1.

Control rod dropout accidents as discussed in the FSAk can lead to significant core damage.

If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.

The overtravel position feature provides a positive check as only uncoupled drives may reach this position.

Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

Absence of such rest or.se to drive movement could indicate an uncoupled condition.

Rod position indication is required for proper function of. the rod sequence control system and the rod worth minimizer (RWM).

2.

The control rod housing support restricts the outward movement of a control rod to less then 3 inches in the extremely remote event of a housing failure.

The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system.

The design basis is given in subsection 3. 5. 2 of the FSAR and the safety evaluation is given in subsection 3. 5.4.

This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

Additionally, the support is not required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the strongest control rod.

3.

The Rod Worth Minimizer (RWM) and sequence mode of the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to prespecified sequences.

The group notch mode of the RSCS restricts movement of rods assigned to each notch group to notch withdrawal and insertion.

All patterns associated with these restrictions have the characteristic that, assuming the worst single deviation from the restrictions, the drop of any control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in the peak enthalpy of any pellet exceeding 280 calories per gram.

An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e.,

425 calories per gram).

Primary system damago in this accident is not possible unless a significant amount of fuel is rapidly dispersed.

Ref. Sections 3.6.6, 14.6.2 and 7.16.3.3 of the FSAR, NEDO-10527 and supplements thereto, and NEDE-24011-P-A.

-108-

PBAPS U: 2T 2 Unit 3 l

3.3 and 4.3 BASES (Cont ' d)

C.

Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate f ast 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 with the average response of all drives as given I

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.

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 l

exceeds one control rod of a (5x5) twenty-five control array.

In the analytical treatment of the transients, which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a nettron sensor reaching the scram point and the start of negative reactivity insertion.

This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results.

The 290 milliseconds used in the analyses consists of 90 milliseconds for sensor and circuit delay and 200 millisecond to start of control rod motion.

times specified in Specification 3.3.C.

In addition the control rod drop accident has been analyzed in NEDO-10527 and its supplements 1 6 2 for the scram times given in Specification 3.3.C.

Surveillance requirement 4.3.C was originally written and used as a diagnostic surveillance technique during pre-operational and startup testing of Dresden 2 6 3 for the early discovery and identification of significant changes in drive scram performance following major changes in plant operation.

The reason for the application of this surveillance was the unpredicatable and degraded scram performance of drives at Dresden 2.

The cause of the slower scram performances has been conclusively

-111-

LIMITING CONDITIONS FOR OPEFATTON S UR VEILL Alt - F EO; ' ' EM 3.4 STANDBY LIQUID CONTROL 44 STANDBY _LJ2HID_PCNTROf.

SY ST EM fj ST EM Anplicability:

Aoplicabigicly:

Applies to the operating Applies to the surveillance status of the Standby requirements of the Liquid Control System Standby Liquid Control System Obiective Obiective To assure the availability To verif y the operability ot a system with the of the Standby Liquid capability to shut down the Control System.

reactor and maintain the shutdown condition without the use of control rods.

Specification Specification A.

Normal System Availa bility A.

Normal System Availabilitt 1.

During periods when tuel is The operability of the in the reactor and prior to Standby Liquid Control startup from a Cold Condi-System the performance of tion, the Standby Liquid the following tests:

Control System shall be operable, except as specified 1.

At least once per month in 3.4.B below. This system each pump loop shall be need not be operable functionally tested by when the reactor is i n recirculating demineralized the Cold Condition and all water to the test tank, control rods are fully inserted and Specification 3.3.A is met.

2.

At least once during each operating cycle:

a.

Check that the setting of the system relief valves is 1400<P<1680 psig.

b.

Manually initicte the system, except explosive valves.

J Pump boron solution through the recirculation path and back to the Standby Liquid Ccatrol Solution Tank.

Minimum pump flow rate of 43 gpm against a system head of 122 5 psig shall be verified.

Af ter pumping boron solution the system will be flushed with demineralized water.

-115 -

PBAPS UNIT 2 3.4 BAS ES STANDEY Ig UID CONTROL SYSTEM A.

The conditions under which the Standby Liquid Control System must provide shutdown capability are identified via the Plant l

Nuclear Safety Operational Analysis (Appendix G).

If no more l

than one operable control rod is withdrawn, the basic shutdown reactivity requirement for the core is satisfied and the Standby Liquid Control system is not required.

Thus, the basic reactivity requirement for the core is the primary determinant of when the liquid control system is required.

The purpose of the liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown condition assuming that none of the withdrawn control rods can be inserted.

To meet this objective, the liquid control system is designed to inject a quantity of boron that produces a concentration of 660 ppm of boron in the reactor core in less than 125 minutes.

The 660 ppm concentration in the reactor core will bring the reactor from full power to at least a 3.0%

k subcritical condition, l

considering the hot to cold reactivity difference, xenon poisoning, etc.

The time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poision peak.

The minimum limitation on the relief valve setting is intended to prevent the recycling of liquid control solution via the lifting of a relief valve at too low a pressure.

The upper limit on the relief valve settings provides system protection from overpressure.

3.

Only one of the two standby liquid control pumping loops is needed for operating the system.

One inoperable pumping circuit does not immediately threaten shutdown capability, and reactor operation can continue while the circuit is being l

repaired.

Assurance that the remaining system will perform l

its intended function and that the long term average availability of the system is not reduced is obtained for a one out of two system by an allowable equipment out of service time of one third of the normal surveillance frequency.

This method determines an equipment out of service time of ten days.

Additional conservatism is introduced by reducing the allowable out of service time to seven days, and by increased testing of the operable redundant component.

-119-

PUhPS UluT 2 3.4 BAS ES (Cont'd.)

C.

Level indication and alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration change.

The test interval has been established in consideration of these f actors.

Temperature and liquid level alarms for the system are annunciated in the control room.

The solution is kept at least 100F above the saturation temperature to guard against boron precipitation.

The margin is included in Figure 3.4.2.

The volume versus concentration requirement of the solution is such that, should evaporation occur from any point within the curve, a low level alarm will annunciate before the temperature versus concentration requirements are exceeded.

The quantity of stored boron includes an additional margin (25 percent) beyond the amount needed to shut down the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water.

A minimum quantity of 3080 gallons of solution having a 19.3 percent sodium pentaborate concentration, or the equivalent as shown in Figure 3.4.1, is required to meet this shutdown requirement.

The minimum required pumping rate is based on the injection of the maximum net storage volume within 125 minutes.

t C/

-120-

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o Figure 3.4.1 i

d

-122-

PRAPS Uatt 2 LIMITING CONDITIONS FOR OPERATIOff S URVEILLANCE REQUIREMENTS 3.5.I Average Planar LHGR

4. 3. I Averaoe Planar LHGR During power operation, the APLHGR The APLHGR for each type of fuel for each type of fuel as a function as a function of average planar I

of average planar exposure shall not exposure shall be checked daily exceed the limiting value shown in during reactor operation at Figu re 3. 5.1. C, D, F,3, H S I 225% rated thermal power.

as applicable.

If at any time during operation it is determined by normal surveillance that the limiting value of APLHGR is being exceeded, action shall be initiated within one (1) hour to restore APLHGR to within pre-scribed limits.

If the APLHGR is not returned to within prescribed limits within five (5) hours reactor power l

shall be decreased at a rate which l

would bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless APLHGR is returned to within l

limits during this period.

S urveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits.

3.5.J Lo cal LHGR 4.5.J Local LHGR During power operation, the linear The LHGR as a function of core heat gene ra tion rate (LUGR) or height shall be checked daily any rod in any fuel assembly at during reactor operation at any axial location shall not exceed 225% rated thermal power.

the design LHGR.

LHGR$ LHGRd LHGRd = Design LHGR 13.4 kW/ft for all 8x8 fuel

~l

-133a-

PBAPS Uti1I 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 Average Exposure BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To EOC p

8x8 1.24 1.30 PTA SP 8X8R 1.27 1.32 8x8R 1.27 1.30

-133c-

PLAPS Uhl_ 2 S. 5 BAS ES (Cont 8d.)

H.

Engineering Safeguards Compartments Co_oling and Ventilation One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling.

Engineering analyses indicate that the temperature rise in safeguards compartments l

without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or associated auxiliary equipment cannot be assured.

Ventilation associated with the High Pressure Service Water Pumps is also associated with the Emergency Service Water pumps, and is specified in specification 3.9.

I.

Average Planar LHGR This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR Part 50, l

Appendix K.

l I

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat I

generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an assembly.

The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR.

This LHGR times 1.02 is used in the heat-up code along with the exposure dependent l

steady state gap conductance and rod-to-rod local peaking factors.

The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor.

The limiting value for APLHGR is shown in Figure 3.5.1.C, D, F, G,

H, and I.

The calculational proceaure used to establish the APLHGR shown on Figures 3. 5.1. C, D,

F, G,

II, and I is based on a loss-of-coolant accident analysis.

The analysis was performed using General l

Electric (GE) calculational models which are consistent with the l

requirements of Appendix K to 10 CFR Part 50.

A complete j

discussion of each code employed in the analysis is presented in Reference 4.

Input and model changes in the Peach Bottom loss-of-coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in Reference 8.

These changes to the analysis include: (1) consideration of the counter current flow limiting (CCFL) effect, (2) corrected code inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate.

-140-

l l

PuAPS UIC " 2 a.5.I BASES (Cont' d. )

A list of the significant plant parameters to the loss-cf-coolant accident analysis is prer nted in Table 3.5-1.

J.

Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation.

The maximum LHGR shall be checked daily during reactor operation at 225% power to determine if fuel burnup, or control rod movement has caused changes in power distribution.

For LHGR to be at the design LHGR below 25% rated thermal power, the peak local LHGR must be a factor of approximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible control rod pattern.

K. Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR The required operating limit MCPR's at steady state operating j

conditions as specified in Specification 3.5.K a; 3 derived from the established fuel cladding integrity Safety Limit MCPR of the fuel cladding integrity safety limit, and analyses of the I

abnormal operational transients presented in the Supplemental l

Reload Licensing Submittal and Reference 7.

For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.

To assure that the fuel cladding integrity Safety Limit ia not exceeded during any anticipated abnormal operational transient, H

the most limiting transients have been analy_.ed to determine l

which result in the largest reduction in critical power ratio (CPR).

The type of transients evaluated are as described in reference 7.

'l 1

l I

l

-140a-

PBAPS Uli 2 2

3.5.K BASES (Cont ' d. )

The limiting transients which determine the required steady state MCPR limits are given in Table 3.5-2.

These transients yield the largest CPR for each class of fuel.

When added to the safety limit MCPR of the fuel cladding integrity safety limit, the required minimum operating limit MCPR's of specification 3. 5.K are obtained.

Two codes are used to analyze the rod withdrawal error transient.

The first code simulates the three dimensional BWR core nuclear and thermal-hydraulic characteristics.

Using this code a limiting control rod pattern is determined; the following assumptions are included in this determination:

(1) The core is operating at full power in the xenon-free condition.

(2) The highest worth control rod is assumed to be fully inserted.

(3) The analysis is performed for the most reactive point in the cycle.

(4) The control rods are assumed to be the worst possible pattern without exceeding thermal limits.

(5) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the maximum allowable linear heat generation rate.

(6) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the minimum allowable critical power ratio.

The three-dimenaional BWR code then simulates the core response to the control rod withdrawal error.

The second code calculates the Rod Block Monitor response to the rod withdrawal error.

This code simulates the Rod Block Monitor under selected failure conditions (LPRM) for the core response (calculated by the 3-dimensional BWR simulation code) for the control rod withdrawal.

The analysis of the rod withdrawal error for Peach Bottom Unit 3 considers the continuous withdrawal of the maximum worth control rod at its maximum drive speed from the reactor which is l

operating with the limiting control rod pattern as discussed above.

I

-140b-

PBAPS Umi 2 3.5.K BAS ES (Cont ' d. )

l A brief summary of the analytica] nethod used to determine the nuclear characteristics is given in Section 3 of Reference 7.

l Analysis of the abnormal operational transients is presented in Section 5.2 of Reference 7.

Input data and operating conditions used in this analysis are shown in Table 5-8 of Reference 7 and in the Supplemental Reload Licensing Submittal.

L.

Average Planar LHGR ( APLHGR), Local LHGR, and Minimum Critical Power Ratio (MCPR)

In the event that the calculated value of APLHGR, LHGR or MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed limits.

The status of all indicated limiting fuel bundles is reviewed as well as input data associated with the limiting values such as power distribution, instrumentation data (Traversing In-core Probe-TIP, Local Power Range Monitor -

l LPRM, and reactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated l

values are valid.

In the event that the review indicates that the calculated value i

j exceeding limits is valid, corrective action is immediately undertaken to restore the value to within prescribed limits.

Following corrective action, which may involve alterations to the control rod configuration and consequently changes to the core power distribution, r viewa instrumentation data, including e

changes to the relatt,.ve neutrea flux distribution for up to 43 incore locations is Obtained and the power distribution, APLHGR, LHGR and MCPR calculated.

Corrective action is initiated within one hour of an indicated value exceeding limits and verification that che indicated s alue is within prescribed limits is obtained within five hours of the initial indication.

In the event that t'le calculated value of APLHGR, LHGR or MCPR exceeding its limit ing value is not valid, i.e., due to an erroneous instrumentation indication etc., corrective action is initiated within one hour of an indicated value exceeding limits.

Verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication.

Such an invalid indication would not be a violation of the limiting condition for oper6 tion and therefore would not constitute a reportable occurrer ce.

I 1

-140c-

PEAPS Util ? 2 Unit 3 3.5.L BAS ES (Cont ' d. )

Operating experience has demonstrated that a calculated value of APLHGR, LHGR or MCPR exceeding its limiting value predominately occurs due to this latter cause.

This experience coupled with the extremely unlikely occurrence of concurrent operation exceeding APLHGR, LHGR or MCPR and a Loss of Coolant Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M.

References 1.

" Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7, and 8 NEDM-10735, August 1973.

2.

Supplement i to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (Regulatory Staf f).

3.

Communication:

V. A. Moore to I.

S.

Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974.

4.

General Electric Company Analytical Model for Loss-of-Coolant Analy31s in Accordance with 10 CFR 50, Appendix K, NEDE-20566 (Draf t), August 1974.

5.

General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G.

L.

Gyorey to Victor Stello, Jr., dated December 20, 1974.

6.

- DELETED 7.

General Electric Boiling Water Reactor Generic Reload Fuel Application.

HEDO-?4011-P-A.

8.

Loss-of-Coolant Accident Analysis For Peach Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977.

-140d-

PBAP9 i!t:IT T A PLF 3. ':-1 SIGMIPirAUT [ M P ilT F /. i A!9 TEE: TO TliE LOST-OF-Cn0LAN"' ACCIDENT ANALYSIS PLAf1T PARAMETEPS:

Core Thermal Power 3440 MWL which corresponds to 1051 of rated steam flow Vessel Steam Output 14.06 x 10 lbm/h which corresnonds to 1057 of rated steam flow Vessel Steam Dome Presnure 1055 psia Pecirculation Line Preak Area For Large Preaks -

Discharge 1.9 P L' (DBA)

Suction 4.1 r d' Assumed Number of Drilled Pundles 360 FIIEL PARAMETFRS:

Peak Technical Initial Specification Design Minimum Linear Feat Axial Critical Fuel Pundle Generation Rate Peaking Power Fuel Tvnn Gnometrv (VW/ft)

Paetor Rntio 7x7, Tyne 2 7x7 18 5 1.5 1.2 7x7, Type 3 7x7 18.5 1.5 1.2 8x8, Tyne H 8 x8 13.4 1.4 1.2 3x8, Type L A x8 13 4 1.4 1.2 8x89/LTA 8x8 13.4 1.4 1.2 P 8x8R 8x8 13,4 1.4 1.2 Type P8DRB284H P

8xBR 8x8 13.4 1.4 1.2 Type P8DRP285

-140e-

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- 1h2b -

l

I PEACH BOTTOM UNIT 2 FUEL TYPE: 8XS LTA 260 (Applicable to 100 mil channels)

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4 PEACH BOTTOM UNIT 2 P8X8R FFl TYPE PSDRB285 (Applicable to 100 mil channels)

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igure 3.5.1.I F.arim m Average Planar Liner Heat Generation Rate Versus Planar Average Exposure

PBAPS U:u t 2 LIMITING CONDITIONS FOR OPERATION S UR VEI LL ANE 11EQUj k i.'41 N 1.,

3.6.A Thermal and Pressurization 4.6.A. Thermal and Pressurization Limitations (Cont' d)

Limitations (Cent' d)

Figures 3. 6.a,3.6. 2 and Selected neutron flux

3. 6. 3 will be updated to specimens shall be account for radiation removed *during the damage prior to 9 third refueling outage ef fective full power and tested to years of operation.

experimentally verify or l

adjust the calculated I

values of inegrated neutron flux that are l

used to determine the RT f or Fiture 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 shall 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 la tion loop shall not be of an idle recirculation started unless the tempera-loop, the temperature of the tures of the coolant within reactor cool 7t in the the idle and operating recir-operating an

'.dle loops culation loops are within shall be permanently logged.

500F of eac h 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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s.

PBAPS Ufe

'2

3. 6. A & 4. 6. A. Bases (Cont'd)

Tho' vessel pressurization temperatures at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Figure 3.6.1, 3.6.2, or 3.6.3 in conjunction with Figurs 3.6.4.

Note:

Figure 3.6.3 includes an additional 400F margin required by 10 CFR 50 Appendix G.

Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level.

The wires and samples will be removed and tested to experimentally verify the values used for Figure 3.6.4.

As described in paragraph 4.2.5 of the Safety Analysia report, detailed stress analyses have been made on the reactor vessel for both steady state and transient conditions with respect to material fatigue.

The results of these transients are compared to allowable stress limits.

Requiring the coolant temperature in an idle recirculation loop to be within 500F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.

The plant safety analyses (Re f: NEDE-24011-P-A) states that all MSIV valve closure - Flux scram is the event which satisfies the ASME Boiler and Pressure Code requirements for protection from the consequences of pressure in excess of the vessel design pressure.

The reactor vescel pressure code limit of 1375 psig, given in Subsection 4.2 of the FSAR, is well above the peak pressure prod.uced by the above overpressure event.

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c J

PBAPS UtJ :T2 3.6.D & 4.6.D BASES Safety and Relief Valves The safety / relief and safety valves are required to be operable above the pressure (122 psig) at which the core spray system is not designed to deliver full flow.

The pressure relief system for each unit at the Peach Bottom APS has been sized to meet two design bases.

First, the total capacity of the safety / relief and the safety valves has been established to meet the overpressure protection criteria of the ASME code.

Second, the distribution of this required capacity between safety / relief valves and safety valves has been set to meet design basis 4.4.4.1 of subsection 4.4 of the FSAR which states that the nuclear system saf ety/ relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.

The details of the analysis which shows compliance with the ASME code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical Report presented in Appendix K of the FGAR.

Eleven safety / relief valves and two safety valves have been installed on Peach Bottom Unit 2 sith a total capacity of 79.51%

of rated steam flow.

The analysis of the worst overpressure transient demonstrates margin to the code allowable overpressure limit of 13 75 psig.

To meet the power generation design basis, the total pressure relief system capacity of 79.51% has been divided into 65. 96%

safety / relief (11 valves) and 13.55% safety (2 valves).

The analysis of the plant isolation transient shows that the 11 safety / relief valves limit pressure at the safety valves below the setting of the safety valves.

Therefore, the safety valves will not open.

Experience in safety / relief and safety valve operation shows that a testing of 50 per cent of the valves per year is adequate to detect fallure or deteriorations.

The safety / relief and safety valves are benchtested every second I

l

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4 6

1 PDAPS Uf)1T 2 50 MAJOR DESIGN FEATURES 5.1 SITE FEATURES The site is located partly in Peach Ecttom 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, j

Pennsylvania.

Figures 2.2.1 through 2.2.4 of the FSAR show the site location with respect to surrounding communit.es.

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.

D.

The reactor core shall contain 185 cruciform-shaped control rods.

The control material shall be boron carbide powder (B C) compacted to approximately 70% of the theoretical I

density, except as described in Section 5.2.C below.

C.

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

e

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T, s

4~

PDAPS U:4IT 2

5. 5 PUEL STORAGE A.

The new fuel storage facility shall be such that the Keff dry is less than 0.90 and flooeded is less than 0.95.

B.

The Kef f of the spent fuel storage pool shall be less than or equal to 0.95.

C.

Spent f uel shall only be stored in the spent fuel pool in a vertical orientation in approved storage racks.

D.

The average fuel assembly loading shall not exceed 17.3 grams U-235 per axial centimeter of total active fuel height of the assembly.

5.6 SEISMIC DESIGN The station Class I structures and systems have been designed for ground accelerations of 0.05g (design earthquake) and 0.1;g (maximum credible earthquake).

1

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