ML20062C040

Amend 48 to License DPR-44,revising Tech Specs to Permit Plant Operation W/New 8x8R Reload Fuel for Cycle 4

ML20062C040
Person / Time
Site:	Peach Bottom
Issue date:	10/16/1978
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062C037	List: ML20062C035 ML20062C040 ML20062C044 ML20062C048
References
NUDOCS 7811060125
Download: ML20062C040 (36)

Text

i ,4.. . ,

ye h UNITED STATES y NUCLEAR REGULATORY COMMISslON

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. nAsHINGTON, D. C. 20555 g

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PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY 4 DOCKET NO. 50-277

} PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILTTY OPERATING LICENSE i .

Amendment No. 48 4

License No. DPR-44 1

1. The Nuclear Regulatory Commission (the Commission) has found that: .

A. The application for amendment by Philadelphia Electric Company, et al (the licensee), dated July 28, 1978, as supplemented September 5, 26, and October 4,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR

, Chapter I;

, B. The facility will operate in conformity with the

( application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 1

D.

The issuance of this amendment will not be inimical to the l common defense and security or to the health and safety I of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2).of Facility Operating License No. DPR-44 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through knendment No. 48, are t hereby incorporated in the license. The licensee shall operate the facility in accordance with the

(' Technical Specifications.

3.

This issuance. license amendment is effective as of the date of its FOR THE NUCLEAR REGULATORY COMMISSION

~ -

Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors

Attachment:

Changes to the Technical Specifications

(

Date of Issuance: October 16,1978 4

• . m en = s . eO w e. =

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ATTACHMENT TO LICENSE AMENDMENT NO. 48 FACILITY OPERATING LICENSE NO. DPR-44

! DOCKET NO. 50-277 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Replace iv iv i y v vi vi

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( 10 10 11 11 12* 12*

14 14 15 15 15a deleted 15b deleted 18 18 19 19 20 20 21 '

33 21 33 g-35 35 40 40 73 73 74 74 91

' 91 92 92 103 103 108 108 111 111 133a 133a 133b 133b 140 133c (added) 140 140a 140a 140b 1406 140c 140c 140d 140d l 140e 140e 141b 141 b 157 142f (added) 157

• No change on this page.

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~ PBAPS -Unit 2 l k

LIST OF FIGURES Ficure Title Pace 1.1-1 APRM Flow Bias Scram Relationship To 16

• Normal Operating Conditicns 4.1.1 Instrument Test Interval Determination 55 Curves 4.2.2 Probability of System Unavailability 98 Vs. Test Interval '

3.4.1 Required Volume and Concentration of 122 Etandby Liquid control System Solution

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3.4.2 Required Temperature vs. Concentration 123 for Standby Liquid Control System Solution l 3.5.1.A MAPLHGR Vs.' Plana: Average Exposure, . 142

• Unit 2, 7x7 Fuel, Type 3 .

3. 5.1. B MAPLEGR Vs. P1anar Average Exposure, 142a Unit 2, 7x7 Fuel, Type 2 i

3. 5.1. C MAPLHGR Vs. Jlanar Average Exposure, 142b t Unit 2, 8x8 Fuel, Type E - 80 mil, 100 all  ;

and 120 mil channels l

{ i 3.5.1.D MAPLHGR Vs. Planar Average Exposure, 142c Unit 2, 8x8 Fuel, Type L "

'3.5.1.I Kf Factor Vs. Core Flow 142d 3.5.1.F MAPLHGR Vs. Planar Average Exposure, 142e  :

Urtit 2, 8xS LTA Fuel

• 3.5.1.G MAPLHGR Vs. Planar Average Exposure, 142f Unit 2, 8x8R Fuel l I  !

i 3.6.1 Minimum Temperature for Pressure Tests such 164 i as required by Section XI . t 3.6.2 Minimum Temperature for Mechanical Heatup or 164a i Cooldcyn following Nuclear Shutdown I 3.6.3 Minimum Temnerature for Core Operation (Criticality) 164b 3.6.4 Transition Temperature Shift vs. Fluence 164c 6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant 245 Operations <

Amendment No.

(( 48 -iv- ,

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,- PSAPS Unit 2 i

a LIST OF TABLES i  ?

. Table , Title Page i

3.1.1 I e Reactor Protection System (Scram) 37 l Instrumentation Requirement t 4.1.1 Reactor Protection System (Scram) 41 l Instrument Functional Tests 4.1.2 Reactor Protection System (Scram) 44 ,

Instrument Calibration l

J 3.2.A Instrumentation That Initiates Primary 61 Containment Isolation ,

3.2.B Instrumentation That Initiates or controls 64 the core and containment cooling Systems t

3.2 C Instrumentation That Initiates Control 73 Rod Blocks 3.2.D Radiation Monitoring Systems That Initiate 75 and/or Isolates Systems 3.2.E Instrumentation That Monitcrs Drywell Leak 76 ,

Detection

. 3.2. F Surveillance Instrumentation 77

' 3. 2. G Instrumentation That Initiates Recirculation 79 Pump-Trip  ;

4. 2. A Minimum Test and Calibration Frequency 80 i for PCIS '

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Amendment No. Jd, 48 -v-

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.JBAPS ... .._ . . . . . . . Unit 2 .

LIST OF TABLES Table Title Pace

4. 2. B Minimum Test and Calibration Frequency 81 for CSCS -

4.2.C Minimum Test and Calib' ration Frequency 83 for Control Rod Blocks Actuation 4.2 D Minimum Test and Calibraticn Frequency 84 for Radiation Monitoring Systems

4. 2. I Minimum Test and Calibratica Frequency 85 for Drywell Leak Detection i

4. 2. ? Minimum Test and Calibration Frequency . 86 for Surveillance Instrumentation -

' 4. 2. G Minimum -Test and Calibration Trequency 88 for Recirculation Pump Trip 3.5-2 operating Limit'MCPR Values as Determined 133c l from Indicatad Transients for various l Core Exposures

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3.5-1 Significant Input Parameters To The 140e Loss-of-Coolant Accident Analysis

4.6.1 In-Service Inspection Program for Peach 150 Bottom Units 2 and 3

,3.7.1 Primary containment Isolation Valves 179 3.7.2 Testable Penetrations With Double 184 0-Ring Seals 3.7.3 Testable Benetrations Wi.tb Testable 184 Bellows 3.7.4 Primary containment Testable Isolation 185 Valves  ;

4. 8.1 Radioactive Liquid waste Sampling 210  !

and Analysis

• 4.8.2- Radioactive Gaseous Waste Sampling 211 and Analysis e 3.11.D.1 Safety Related Shock Suppressors (Snubbers) 234d 3.14.C.1 Fire Detectors .

d240k Amendment No. 23, 48 -vi.

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PBAPS Unit 2 l .

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1

1.1 FUEL CLADDING IN"'EGRITY 2.1 FUEL CLADDING INTEGRITY i

i I' Aeolicability: Acolicability:

The Safety Limits established The Limiting Safety System Settings to preserve the fuel cladding apply to trip settings of the instru-integrity apply to those ments and devices which are provided variables which monitor the to prevent the fuel cladding integrity fuel thermal behavior. Safety Limits from being exceeded.

M eetives: cbiectives:.

The objective of the Safety The objective of the Limiting Safety Limits is to establish limits System Settings is to define the level

' hich assure the integrity of of the process variables at which auto-the fuel cladding. matic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Speci*icatien: Specification:

The limiting safety system settings shall be as specified below:

A. Reactor cressure 2800 osia A. Neutron Flux Scram anc Core Flow 210% of Rated The existance of a ms.nimum 1. APRM Flux Scram Trio Setting critical power ratio MCPR less (Run Model

.than 1.07 shall constitute

[

' violation of the fuel cladding When the Mode Switch is in the a integrity safety limit. RUN position, the APRM flux scram trip setting shall be:

To ensure that this safety limit is not exceeded, neutron S $ 0.66W +54%

flux shall not be above the I scram setting established in where:

specification 2.1.A for longer than 1.15 seccmds as S = Setting in percent of indicated by the process com- rated thermal power puter. When the process ccm- (32 93 MWt) puter is out of service this safety limit shall be assumed W = Loop recirculating flow

-to be exceeded if the neutron rate in percent of rated flux exceeds its scram setting (rated 1 cop recircula-and a control rod sczam does tion flow rate equals not occur.

34.2 x 10* lb/hr).

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Amendment No. - 22, 25, 52 , 48 .

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-PBAPS Unit 2 a[ lI l SAFETY LIMIT LIMITING SAFETY SYSTEM srn ING i

i 2.1.A (Cont'd)

} In the event of operation with a maximum total peaking factor (MTPM greater than the design value of A, the setting shall be modified to the more

' limiting (lower) of the 3 (

values determined by the following:

a. S$ (0. 66W+545) 2.62 _ 1 MTPF for 7x7 fuel 1 s b. S$ (0. 66W+545) _ 2.44 MTPF for 8x8 fuel

c. S$(0. 66W+54%) 2.51 (

,MEPF for 8x8R f uel {

,i j MEPF = The v'alue of the

' existing maximum

total peaking factor For no combination of loop i recirculation flow rate and (

core thermal power shall the

' APRM flux scram trip setting

• be allowed to exceed 120% of i

rated thermal power.

' Design value of A = 2.62 for (

7x7 fuel, 2.44 for 8xS fuel, I and 2.51 for 8x8R fuel. l APRM-When the reactor mode

2. I i,

switch is in the STARTUP position, the APRM scram shall be set at less than or equal 4 to 15 percent of rated power. -

1 4

3. IRM--The IRM scram shall be

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set at less than or equal to 1

i 120/125 of f ull scale.

' 4. When the reactor mode switch I i

is in the STARTUP or RUN position, the reactor anall l

not be operated in the natural circulaticn flow mode.

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I Amendment No. I J ,[,)(I, 48- 1 0-4 4 a

PBAPS Unit 2 o

SAFETY LIMIT LIMITING SAFETY SYSTD1 SecrrING t l

?. Core Thermal Power Limit B. APRM Rod Block Trio Setting (Reactor Pressure s 800 osia)

When the reactor pressure is SRB 5 0.66W + 42%

5 800 psia or core flow is 1 less than 10% of rated, the where:

core thermal power shall not exceed 25% of rated thermal SRB = Rod block setting in power. percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2 '

x 10* lb/hr) .

In the event of operation with a maximum total peaking factor (MTPF) greater than the design value of A, the setting shall be modified to the more limiting (lower) of the 3 i values determined by the ~

following:

1. SRBS (O. 66W+ 42 %) 2.62 E PF for 7x7 fuel

2. SRB 5 (0. 66W+425) 2.a4 EFF for 8x8 fuel 3.

SRas (0. 66W+ 42%) 2.51 MEFF for 8x8R fuel MIPF = The value of the existing j maximum total peaking factor i

Design value of A = 2. 62 fcr 7x7 (

fuel, 2.44 for 8x8 fuel, and i 2.51 for 8x8R f uel. (

C. Whenever the reactor is in the C.

Scram and isolatien--2:538 in. above chutdown condition with reactor icw water irradiated fuel in the reactor level vessel zero (On on level vessel, the water laml aball -

not be less than 17.1 in . above instruments) the top of the normal active fuel zone.

Amendment No. 23, 3A, A2, 48 _

. PBAPS Unit 2 SAFETY LIMIT LIMITING SAFETY SYSTEM SEirING 2.1 (Cont' d)

D. Scram- turbine stop 510 percent valve closure

2. Scram-- turbine cen* col fast closure on loss of control oil pressure. -

500<P<85 0 psig.

F. Scram--low 223 inches condenser vacuum Hg vaccum G. Scram--main steam $10%

(. line isolation. valve i

closure t

H. Main steam 2850 psig isolation valve closure--nuclear system low pressure I. Core Spray & LPCI 2378 in.

actuation-reactor above vessel low water level zero (-159.5 in. indicated level)

J. HPCI & RCIC 2490 in.

actuation-reactor above vessel

' low water level zero (-49.5 in. indica'ad level)

K. Main steam 2490 in isolation valve above vessel closure--reactor zero (-4 9.5 low water level in. indicated level) 4 Amendment No.15 ,

PBAPS Unit 2 1.1.A R&gI,g (Cont'd) '

The required input to the statistical model are the uncertainties listed in Table 5-1 of Reference 3, the nominal values of the l  !

core parameters listed in Table 5-2 of Reference 3, and the I relative assembly power distribution shown in Figure 5-1a of l .

Reference 3. l 5

The basis for the uncertainties 'in'the core para .eters is giyen l  !

in Reference 2 and the basis fcr the uncertainty in the GEZL i correlation is given in Reference 1. The power distribution is based on a typical 764 assembly core in which the rod pattern was  ;

arbitrarily chosen to produce a skewed power distribution having j the greatest number of assemblies at the highest power levels.  !

The worst distributim in Peach Bottom Atomic Power Station Unit 2 during any fuel cycle would not ce as severe as the (

distribution used in the analysis, i j

B. Core Thermal Power Limit (Re' actor Pressure < 800 csia on Core Flow < 10% of Rated) I l

t The use of the GEIL correlation is not valid for the critical I power calculaticns at pressures below 800 psia or core flows. less than 10% of rated. Therefore, the fuel cladding integrity safety limit is established by other means. This is done by establishing a limiting condition of core thermal power operation  ;

with the fallowing basis. 3 t

Since the pressure drop in the bypass region is essentially all i elevation head which is 4.56 psi the core pressure drop at low -  !

j power and all flows will always be greater than 4.56 psi. l Analyses show that with a flow of 28 x 103 1bs/hr bundle flow,  ;

bundle pressure drop is nearly independent of bundle power and l has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi l driving head will be greater than 28 x 10a.1bs/hr irrespective of i total core flow and independent of bundle power for the range of bundle pcwers of ecmcern. r Full scale ATI.AS test data taken at  !

i pressures from 14.7 paia to 800 psia indicate that the fuel i assembly critical power at this flow is approximately 3.35 MWt. I  ;

With the design peaking factor this bundle power corresponds to a i  !

core thermal power of more than 50%. Therefore a care thermal pcwer limit of 25% for reactor pressures below 800 psia or core  !

flew less than 105 is conservative.

C,. Egi1E Transient t

' Plant safety analyses have shown that the scr=ms caused by exceeding any safety setting will assure that the Safety Limit of Specificaticn 1.1. A or 1.1.B will not be exceeded. Scram times l i are checked periodically to assure the insertion times are .l adequate. The thermal power transient resulting wnen a scram is accomplished other than by the expected scram signal -(e.g. , scram from neutrcn flux following closure of the main turbine .st=p valves) does not necessarily cause fuel damage. ~

Amendment No. 22, 24, 48 ,

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.-- TBAPs -

-Unit 2 l; l

1.1. C REZE (Cont' d.) >l

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However, fcr 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 scram signals are operable, is supported by the extensive plant saf ety analysis.

The comp 2ter 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, et=. occur. This program also indicates when the scram setpoint is cleared. This will previde information on how  ;

long a scram conditim exists and thus provide some measure of the energy added during a transient. Thus, computer infarmation normally will be available for analyzing scrams; however, if the i computer inf armatim abould 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 Leva,1 (Shutdown Condition)

During periods wnen the reactor is shutdown, consideraticn 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 l 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 limit at f 17.7 inches above the top of the fuel provides adequate margin. .

This level will be centinuously monitored.  !

E. References  :

l

1. General Elect:ic Thermal Analysis Basis (GETAB) : Data, i Correlation and Design Application, January 1977 (NEDO- l 10958-A) i

2. Process Ccaputer Performance Evaluation Accurac.

• General Electric Company BWR Systems Department, June 1S e 4 (NEDO- [

20340)  ;

3. " General Electric Boiling Water Reactor Generic Reload Fuel l Applicaticn*, NEDE-24011-P-3, March 1978. l l

i l

Amendment No. 23, JA, 48 i

,aea mus.d .-.e ae=NNNN**"8

PBAPS thit 2 l J

2.1 3AEg (Cont'd.)

For analyses of the thermal consequences of the transients a MCPR equal to or greater than the operating limit MCPR given in j specification 3.5.K is conservatively assumed to exist prior to 1 initiation of the limiting transients. This choice of using i  !

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 operaticn without forced recirculation will not be permitted, except during startup testing. The analysis to  ;

support operation at various power and flow relationships has i considered oseration with either one or two recirculating pumps.

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In summary, i i

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The abnormal operational transients were analyzed to a power level of 3440 MWt.  !

ii. The licensed maximum power level is 3293 MWt. '

iii. Analyses of transients employ adequately conservative values of the controlling reactor parameters.

h iv.

The analytical procedures now used result in a more logical i answer than the alternative method of assuming a higher ,

starting power in conjunction with the expected values for the parameters.

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l The bases for individual trip settings are discussed in the  !

fc11cwing paragraphs. l i

"A. Neutren Flux Scram ,

The Average Power Range Monitoring (APRM) system, which is r I

calibrated using heat balance data taken during steady state l conditiens, reads in percent of rated thermal power (3293 MWt) . i Because fissicn chambers provide the basic input signals, the  :

APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel  !

(reactor thermal power) is less than the instantaneous neutren flux due to the time constant of the fuel. Therefore, during abnormal operaticnal transients, the thermal power of the fuel will setting.

be less than that indicated by the neutron flux at the scram Analy m demonstrate that with a 120 percent scram trip l setting, none of the abnormal operational transients analyzed .

violate frem fuelthe fuel safety I.imit and there is a substantial margin damage. Tnerefore, the use of flow referenced scram l trip provides even additional margin. '

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~ _. j Amendment No. 2J, 2A, 48 '

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PBAPS Unit 2 2.1.A E,3Xg (cont'd.)  !

An increase in the APRM scrds trip setting would decrease the margin reached.

present before the fuel cladding integrity safety Limit is 1he APRM scram trip setting was determined by an analysis of margins 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 effect al reactor safety' because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because

it provides adequate margin for the fuel cladding integrity saf ety Limit yet allcws operating margin that reduces the ,

l possibility cf unnecessary scrams. l' i

The scram trip setting must be adjusted to assure that the LEGR  !

transient peak is not increased for any combination of MTPF and  !

reactor core thermal power. The scram setting is adjusted in

(, accordance with the formula in Specification 2.1. A.1, when the i "4""!m total peaking factor is greater than the design value of i A for each class of fuel.  !

Analyses of the limiting transients show that no scram adjustment  ;

is required to assure MCPR greater than 1.07 when the transient i  !

is initiated from MCPR greater than the operating limit given in Specification 3.5.K.  !

i For operation in the startup mode while the reactor is at low t pressure, the APRM scram setting of 15 percent of rated power l i

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 plant  !

startup. Effects of increasing pressure at zero or low void  ;

coratent are minor, cold water from sources available during '

startup is not much colder than that already in the system, i temperature coefficients are small, and control rod patterns are l constrained to'be uniform by operating procedures backed up by j

the Rod Worth Minimizer and Rod sequence control System. Worth of individual rods is very low in a uniform rod pattern. Thus, ,

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 change of power is very slow. I Generally, the beat 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 exce=d the safety Limit. The 15 percent APRM scram resins active until the mode switch is placed in the RUN position. This switch occurs when the reactor pressure is greater than 850 psig.

I Amendment No. JA, 48  !

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PBAPS Unit 2  !

l 2.1.A BASES (Con t' d. )

The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade j 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 rarse i of the IRM. For example, if the instrument were on range 1, the j

scram setting would be at 120 divisions for that range; likewise,  ;

if the instrument were on range 5, the scram would be 120 l divisions en 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 f- change of power is slow enough due to the physical limitation of l

l 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. ,

r In order to assure that the IRM provided adequate protection l 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 ,

t

was takan in this analysis by assuming that the IRM channel ]

i closest to the. withdrawn rod is bypassed. The results of this analysis show that the reactor is scrarrined and peak power limited to one percent of rated power, thus maintaining MCPR above 1.07. " ll (

Based on the above analysis, the IRM provides protection against  !

-local control ved withdrawal errors and continuous withdrawal of I control rods in-sequence and provides backup protecticn for the  !

APRM. I B. AEEi Egg 21gc_h IE12 settina s i

The APRM system provides a control rod block to avoid conditions  !

which wculd 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 recirculatias loop flow l rate. The flow variable APRM rod block trip setting provides margin to the APRM scram trip setting over the entire i recirculation ficw range. As with the APR." 9 cram trip setting, the APRM rod block trip setting is adjusted downward if the Maximum Total Peaking Facter (MTPF) exceeds the design value A J for each fuel type.

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l l Amend m nt No. 22, 24, 48 i

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PBAPS . Unit 2 2.1 R&ggg (Cont'd.) '

C. Reactor Water Ifigg Level gLam gnd Isolation (Except gain i i

Steamlinest 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 l i

FSAR subsection 14.5 show that s' cram and isolation of all process '

lines (except main steam) at this level adequately protects the i fuel and the pressure barrier, because MCPR is greater than 1.07 1 (

in all cases, and system pressure does not reach the safety valve l settings. The scram setting is approximately 31 in. below the I normal operating range and is thus adequate to avoid spurious scrams.  ;

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D.. Turbine gji2g Valve Closure g,gggjg She turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase tnat 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 i such that MCPR remains above 1.07 even during the worst case { l transient that assumes the turbine bypass is closed. This scraa is bypassed when turhine steam flow is below 30% of rated, as  ;

measured by turbine first stage pressure.

E. Turbine Centrol Valve Scram t

The turbine control valve fast closure scram anticipates the (

pressure, neutron flux and heat flux increase that could result l frem fast closure of the turbine control valves due to a load

' rejectim exceeding the capacity of the bypass valves or a

- failure in the hydraulic control system wnich results in a loss ,

of oil pressure. This scram is initiated from pressure switches i i

in the hydraulic ccntrol system which sense loss of oil pressure due to the opening of the fast acting solenoid valves or a i failure in the hydraulic centrol system piping. Two turbine first t l

stage pressure switches for each trip system initiate automatic  :

bypass of the turbine control valve fast closure scram when the '

first stage pressure is below that required to produce 30% of j rated pcwer. Cental valve closure time is approximately twice as long as that for stop valve closure.  !

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. t Amendment No. 2d , 48  :

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2.2 BASES

REACTOR COOLANT SYSTEM INTEGRITN The pressure relief system for each unit at the Peach Bottom  !

i Atomic Power Station has been sized to meet two design bases.

First, the total capacity of the safety / relief valves 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 bas been set to meet design basis 4.4.4.1 of subsection 4.4 which states that the nuclear system safety / relief valves j shall prevent isolations andopening of the safety valves during normal plant load rejections.  !

I n

The details of the analysis which shows compliance with the ASME  !

Code requirements are presented in subsection 4.4 of the FSAR and }

the Reactor Vessel Overpressure Protection Summary Technical Report submitted in Appendix K.

Eleven safety / relief valves and two safety valves have been l installed on the Peach Bottom Units. The analysis of the worst  ;

overpressure transient, (3-second closure of all main steamline isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1295 psig for l l

Peach Botton Unit 2 if a neutron fluz scram is assumed. This  !

results limit of in a 80psig.

1375 psig margin to the code allowable overpressure l  !

The analysis of the plant isolation transient (load rejection I with bypass valve failure to open and Recirculation Pump Drive Motor Trip) assuming a turbine trip scram is presented in NEDO- l }

24132, Revision 1 for Peach Botton Unit 2. This analysis shows l r l

that the 11 safety / relief valves limit pressure at the safety l i

. valves to 25 psig below the setting of the safety valves. i Therefore, the safety valves will not open. l The safety / relief valve settings satisfy the Code requirements '

that the lowest pressure of 1250valve psig. setThese point be at or below the vessel design

settings are also sufficiently above cycling caused by minor transients. range to present unnecessary the normal operating pressure I'

The results of postulated transients where inherent safety / relief I valve actuation is required are given in Section 14.0 of the Final Safety Analysis Report. l i

i The design pressure of the shutdown cooling piping of the I Res~idual Heat Removal System is not exceeded with the reactor vessel steam dome less than 75 psig. )

l l

l 1

i l

l l

Amendment No. 23, 28, 48 September 1978

-- - L _ m~ :*: _- -_ m

^~

. - - _ . _._ T l__^ _  ?

~

- -- -- ~ ~ - ~ ~

~

' ~

PBAPS Unit 2 i EQEg Egg M W (Cont'd)

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

T1. An APRM will be considered, operable if there are at least 2 LPR:* inputs per level and at least 14 LPRM inputs of the ,

nor:nal complement.

12. W is the recirculation loop flow in percent of design. W is  !

equal to 100 for core flow of 102.5 million pounds / hour or '

greater. Trip level setting is in percent of rated power (3293 MWt) . A = 2. 62 for 7x7 fuel, 2.44 for 8x8 fuel, and (

2. 51 for 8x8R fuel. MTPF is the value of the existing l maximum total peaking factor.

,m 13. See section 2.1.A.1.

(. f i

~

i l

i i

i

. b I

i r

Amandant No. 23, 48 - 4 0-

_._.r. . .._ _ _.__ , , - - - _ . . . ,

l TABLE 3.2.C 3 INSTRUMENTAiION TilAT INITIATES CONTROL ROD BLOCKS *

=

Minimum Ho. - _ - - - - - ~ . - - -.- _

of Op3rable Inctrument- Ins trumen t Number'of Instrument Channals Per Trip Level Setting Channels Provided Action  ;

By Eesign '

IElE.EH19.m__ __

2 APRM Upscale (Flow 5 (0.66 We 42) x L (2) 6 Inst. Channels (1) .t Diased) HTPF l j 2 APRM Upscale (Startup 5125 i 6 Inst. Channels (1) '

Hode) 2 APRM Downscale 22.5 indicated on 6 Inst. Channels (t) scale l5 1 (7) Rod Block Honitor {

$( 0.66 W+ 41)' x __8__ (2) 2 Inst. Channels. (1)

(Flow Blased) HTPF ,

8 1 (7) R'od Block Monitor 22.5 indicated on ' 2 Inst. Channels Downscale scale (1) f 3

IRN Downscale (3) 22.5 indicated on 8 In u t. Channels (1) ,

scale 3 IRM Detector not ih (8) 8 inst. Channels Startup Position (1) 3 IBM Upscale  !?

5108 indicated on 8 Inst. Channels (1) p -

scale y  !

2 (5 ) SHM Detector not in (4) 4 Inst. Channels Startup Position (1) [

i 2 (5) (6) SDH Upscale 5105 coun ts/s ec. 4 Inst. Channels I

(1) i

)

. 1, Amendment No. 23, JA, 24, A2, 48 .

O

. _ _ _ _ _ _ . . , _ _ . _- - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ' ' ' ' ~ ' ~ ' ' ~ ' " " " ~ ' " ' ~ ~ '

~ ~ ~ ~ '

., .;.-.--- u - - - - - - - - - - ~

PBAPS Unit 2 NOTES EQE Ijugd M '

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems f cr each function. The SRM and IRM blocks need not be operable in "Runa mode, and the APRM and RBM rod blocks need not be operable in "Startupa mode. If the first column cannot be met for one of the two trip systems, this conditica 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. W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MWt) .

Refer to Limiting Safety Settings for variation with peaking i f actors, A = 2. 62 for 7x 7 fuel, 2. 44 f or 8x8 f uel , and 2. 51 i for 8x8R fuel. MTPF is the value of the existing maximum total peaking facter. I 3.

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

4.

This functicn is bypassed when the count rate is 2 100 cps.

5. One of the four SRM inputs may be bypassed.

6.

This SRM functicn is typrased 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 functicn is bypassed when the mode switch is placed in f

Run. -

i l

Amendment No. 2J , 48 _ _ _ _ _ _ _ .. _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _

- . t

- - PBAPS ~

Snit 2 i,

3.2 BASES (Cont' d)

Pressure instrumentation is provided to close the main steam  :

l 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 wnen 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 Isolatim valves for thermal transient  ;

protection when not in RUN mode is nct required. I t

The HPCI high flow and temperature instrumentation are provided i to detect a break in the EPCI steam piping. Tripping of this (

in3trumentation results in actuation of HPCI iscL.stion valves.

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

Temperature is monitored at four (4) locations with four (4) i (3 temperature sensors at each location. Two (2) sensors at each l j

location are powered by "A" DC ccaatrol 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 sencor will actuate HPCI isolation valves. The trip l settings of $300% of design flow for high flow and 2000F for high i temperature are such that core uncovery is prevented and fission product release is within limits. l I

The RCIC high flow and temperature instrumentation are arranged I the same as that for the HPCI. The trip setting of $3005 for

, high flow and 2000F for temperature are based on the same i criteria as the EPCI. ,

{

The Reactor Mter Cleanup System high flow and temperature t instrumentatica r i s arranged similar to that for the HPCI. The

{

l trip settings ar- such that core uncovery is prevented and fission product rJ. ease is within limits. I

[

, t i

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in l this fashion, the Specification preserves the effectiveness of  !

i the system even during periods when maintenance or testing is [

being performed. An exception to this is when logic functional testing is being performed. I The control rod block functicus are provided to prevent excessive l  !

centrol rod withdrawal so that MCPR does not decrease to 1.07. t l The trip logic for this function is 1 out of n: e.g., any trip  !

on ene of 6 APRM's, 8 IRM's, or 4 SRM's will result in a rod i  !

bicc k. t The minimum instrument channel requirements assure sufficient  !

instrumentation to a ssure the single failure criteria is met.  ?

The minimum instrument channel requirements for the R3M may be reduced by one f cr maintenance, testing, or calibration. This t time period is only 35 of the operating time in a month and does not significantly increase the risk of preventing an inadvertent t centrol rod withdrawal.

Amendment No. X8, 48 i

s l i PEAPS Unit 2 3.2 BA_ GEE (Cont'd)

The APRM rod block function is flow biased and prevents a

~

significant reduction in MCPR, especially during operaticn at reduced fim. The APRM provides gross core protection: 1. 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 1.07.

l The i.e.,RBM red block function provides local protection of the core; the preventien of boiling transition in the local region of the core, control rod pattern.

for a single red withdrawal error from a limiting The protection. IRM rod block function provides local as well as gross core The scaling arrangement is such that trip setting is

.m less than a factor of 10 aknve the indicated level.

U A downscale indicatim 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 caparator and scram discharge volume high level components safety. have only one logic channel and are raot required for one recirculation The flowwater caparator pump.must be bypassed v;nen operating with The refueling interlocks also operate one logic channel, and are required position. for safety mly when the mode switch is in the refueling For effective emergency core cooling for small pipe breaks, the

~HPCI system must function since reactor pressure does not decrease operate in rapidly time. enough to allow either core spray or LPCI to I The automatic pressure relief function is provided operate. as a backup to the HPCI in the event the HPCI does not Se arrangement of the tripping contacts is such as to provide this operation. function when necessary and minimize spurious Se 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 ma.ntenance, of. inadvertent tasting, or calibration, and also minimizes the risk operation; of service. i.e., only one instrument channel cut Two' air ejector off gas monitors are provided and when their trip point is reached, cause an isolaticn of the air ejector eff-gas line.

high tripIsolation point when isone initiated has anwhen upscale. both instruments reach their 1

Amendment No. JE, 48 . . . , . _

a PBAPS Unit 2 LIMITING CONDITIONS FOR OPERATION SURVEILIANCE RECUITEMENTS 3.3.B Control Rods (Cont'd) 4.3.B Centrol Rods (Cont'd)

4. Control rods shall not be 4. Prior to control rod with-withdrawn for startup or drawal for startup or during refueling unless at least refueling, verify that at two source range channels least two scurce range channels have an cbserved count have an observed count rate '

rate equal to or greater of at least three counts per than three counts per second.

second.

. 5. During operation with 5. When a limiting control rod limiting control rod pat- pattern exists, an instru- -

,m terns, as determined by the ment functional test of the .

(

) designated qualified person- RBM shall be performed nel, either: prior to withdrawal of the designated rod (s),

a. Both RMB channels aball be operable, or

b. Control rod withdrawal shall be blocked, or

c. The operating power level shall be limited so that the MCPR will remain above 1.07 assuming a single error l that results in complete $

withdrawal of a single operable control rod.

Sc am Insertion Times C. Scram Insertion Times i

' 1. The average. scram inser- 1.

tion time, based on the After each refueling outage l all operable fully withdrawn deenergization of the scram insequence rods shall be scram pilot valve solenoids as time tested during operational time zero, of all operable hydrostatic testing or during I control rods in the reactor startup from the fully with-power operation condition drawn position with th'e nuclear shall be no greater than:

system pressure above 800 psig.

Above 950 csig This testing shall be completed prior to synchronizing the main turbine generater initially 5 Inserted from . Avg. Scram Inser-following restart of the plant.

Smv Withdrawn tien Times (s ec) 5 C.375 20 0.90 50 2.0 90 5.0 l

l Amendment No. 23, 48 -1 03-

_ = -. . .- N. .- . -- A p a .

y .,. .

PBAPS Unit 2 i 3.3 and 4.3 BASES (Cont ' d.) l t

B. Control Rods

1. Control rod dropout accidents as discusced in the FSAR can 1ead to significant core damage. If coupling integrity is {

maintained, the possibility of a rod dropout accident is j eliminated. The overtravel posit. ion feature provides a positive l check as only uncoupled drives may reach this position. Neutron i instrumentation response to rod movement provides a verification  !

that the rod is following its drive. Absence of such response to drive movement could indicate an uncoupled condition. Rod i position indicatica 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 l movement of a control rod to less then 3 inches in the extremely I 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 i not required if the reactor coolant system is at atmospheric 4 pressure since there would then be no driving force to rapidly e]ect a drive housing. Additionally, the support is not required E

i 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 Systen (RSCS) restrict withdrawals and insertions of control rods to prespecified sequences. The group i notch mode of the RSCS restricts movement of rods assigned to i each notch group to notch withdrawal and insertion. All patterns  !

associated with these restrictions bave the characteristic that, j 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 '

i sustain a power excursion resulting in the peak enthalpy cf any  !

' pellet exceeding 280 calories per gram. An enthalpy of 280 i calories per gram is well below the ' level at which rapid fuel dispersal could occur (i.e., 425 calories per gram) . i Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly discarsed. Ref. Sec-dons 3.6.6, 14.6.2 and 7.16.3.3 of the FSAR, NIDO-10527 and i supplements thereto, and NIDO-24132, Revision 1. ,

l Amendment No. J7, 28, 48 -

108-

___'.--_____-___+wwm-=-..i--,---+<

,r..---- ..-y.- - ~ ---,.-,---# , _. - - - - . _ _ _ -

._ .. . . . - - -- - - - - - - ~ ~ ~ - - ^ ~~' ^ ~

l

• ' ~

• .. I

~ '

' PBAPS Unit 2 3.3 and 4.3 BASES (Cont'd.) .

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

to prevent the MCPR from becoming less than 1.07. Analysis of l i

the limiting power transients shows that the negative reactivity l t

rates resulting from the scram (Ref. NEDO-24132, Revision 1) with '

the average response of all drives as. given in the above {

specification, provide the required protection, and the MCPR remains greater than 1.07.

j i t

The numerical values assigned to the specified scram performance f are based on the analysis of data from other BWR's with control  !

s rod drives the same as those on Peach Bottom. I The occurrence of scram times within the limits, but signifiedetly 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 i l

exceeds one control rod of a (Sx5) twenty-five control rod array.  !

In the analytical treatment of the transients, 390 milliseconds I are allowed between a neutron sensor reaching the scram point and the start of negative reactivity insertion. This is adequate and i i

conservative when compared to the typically observed time delay of about 270 milliseconds. Approximately 70 milliseconds after j neutron flux reaches the trip point, the pilot scram valve  !

solenoid power supply voltage goes to zero and approximately 200 milliseconds later, control rod motion begins. The 200 [

milliseconds specified are included in3.3.c.

'm specification the allowable scram insertion times  ;

In addition the control rod i drop acci.'.ent has been analyzed in NEDO-10527 and its supplements 1 & 2 for the scram times given in specification 3.3.C. l surveillance requirement 4.3.c was originally written and used au  !

a diagnostic surveillance technique during pre-operational and  !

startup testing of Dresden 2 8 3 for the early discovery and identification of significant changes in drive scram performance i following major changes in plant operation. The reason for the i 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 i t

l i

Amendment No. 2A, Ja, 48 -111-

F . .

~

' ~~ ,

' Unit 2 PBAPS i

l e LIMI'"ING CONDITIONS FOR OPERATION SURVIILIANCE REOUIREMENTS l 3.5.I Averace Planar LEGR 4.5.I Averace Planar LEGR During power opration, the APLEGR The APLHGR for each type of fuel for each type of fuel as a function as a function of average planar f

of average planar exposure shall not exposure shall be checked daily I

exceed the limiting value show:.. in during reactor operation at Figure 3. 5.1. A, B, C, D, F, & G, as applicable. If at any time during 2255 rated thermal power.

- l operation it is determined by normal trarveillance that the limiting value  :

of APLHGR is being exceeded, action  ;

shall be initiated within one (1) hour to restore APLEGR to within pre-  !

scribed limits. If the APLEGR is not  !

returned to within prescribed limits within five (5) hours reactor power  ;

shall be decreased at a rate which 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 limits during this period. Surveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits. .

l  ;

3.5.J Local LEGR 4.5.J Local LEGR During power operation, the linear The LHGR as a function of core heat generation rate (LHGR) o; height shall be checked daily any rod in any fuel assenhly at during reactor operatien at any axial location shall not exceed 2255 rated thermal power.

the maximum allcwable LaGR as calcu-

~ 1ated by the following equaticm:

LHGR$LEGRd [ 1-( AP/P) max (I/LT) ]

LEGRd = Design LHGR

=

18.5 kW/ft for 7x7 fuel t

13.4 kW/ft for 8x8, 8x8R, 6 ,

and 8x8 LTA fuel l .

l (AP/P) max = Maximum pcwer spiking penalty

= 0.026 for 7x7 fuel

= 0.0 22 for 8x8, 8x 8R, and 8x3 LTA fuel l i LT == Total core length I

=

12 ft for 7x7 6 8x8 fuel 12.5 f t for 8x8R S 8.vS LTA fuel I L = Axial posicion abcve bottom of l core Amendment No. #7. 48 -133a-

_ . ............- -. - - - - - - - --- -- ~ " ~ ^ ~ ~ ~ ~

i

• PRM .

Udt 2 - .

+

LIMITING CONDITIONS FOR OPERATION ' SURVE.ILLANCE REOUIREMEttrS 3.5.J Local LBGR (Cont'd)

If at any time during osaration it -

is determined by normal surveillance that limiting value ,

for LEGR is being exceeded, action i

shall be initiated within one (1) hour to restore LPGR to within -

prescribed limits. If the LEGR is not returned to within prescribed limits within five (5) bours, i reactor power shall be decreased at a rate which 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 k .' haring this period.~ .HGR

. is returned to within limits Surveillance and corresponding action shall i continue until reactor operatien is within the prescribed limits.

3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio IMCPR) Ratio (MCPR)

During power operation, the MCPR MCPR shall be checked daily fcr the applicable incremental I cycle core average exposure and during reactor power operation l for each type of fuel shall be at 2255 rated thermal power. I cqual to or greater than the value i given in Table 3. 5-2 times kf, j waere kf is as shown in Figure I 3.5.1.E. If at any time during 1

/

peration it is determined by 1 normal surveillance that the i limiting value for MCPR is being ' l exceeded, action shall be 1 initiated within one (1) bour to restore MCPR to within prescribed limits. If the MCPR is not returned to within prescribed limits within five (5) bours, reactor power shall be decreased at a rate which 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 MCPR is returned to within limits during this period. Surveillance ,

end correspending action shall centinue until reactor cperation is within the prescribed limits.

I s _

Amendment No. JA. 48 -133b-

._ _ _ . . _ _ . _ . ~ . - -

e PBAPS

-Unit 2 - -

i Table 3.5-2 ' '

OPERATING LIMIT MCPR VAIDES AS DETERMINED FPOM f INDICATED TPANSIENTS FOR VARIOUS CORE EXPOSURES

(

[

Fuel Tvee MCPR Cperating Limit For Incremental Cycle 3 Core Averace Exoosupee BOC to 1000 MWD /t 1000 MWD /t before EOC  ;

Before EOC To ECC 7x7 1.31 (RWE)

/m. 8x8 1.31 (RWE)

' 1. 26 (RWE) 1. 28 (LR) 8x8R/LTA 1. 25 (LR) 1. 28 (LR)

RWE - Rod Withdrawal Error LR -

Load Rejection with failure of bypass valves to open l

l

=

(

i f

8

-133c-

i , ,

j, .

• PBAPS Unit 2 3.5 EA_EEE (Cont'd.)

H. Encineerina Safecuards Cemcartments Cooline and Ventilation one unit cooler in each pump compartment is capable of providing adequate ventilaticn flow and cooling. Engineering analyses indicate that the temperature rise in safeguards ccmpartments ,

without adequate ventilation flow or cooling is such that continued operaticn 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 Specificaticn 3.9.

I. Averace Planar IgGR O' f This ollowingspecification assures that the postulated the peak design cladding basis temperatureaccident loss-of-coolant will not exceed the limit specificd in the 10 CFR Part 50, Appendix K.

The peak cladding temperature (PCT) following a postulated loss-cf-coolant accident is primarily a function of the average heat 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 distributicn within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered red which is equal. to or less than the design LEGR. This LEGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The Technical Specification APLEGR is this LEGR of the highest powered rod divided by its local peaking fac+wr. The limiting value for APLEGR is shcwn in Figure 3.5.1-A, B, C, D, F 1

. a nd G. l The calculatiorial procedure used to establish the APLHGR shcwn on Figures 3. 5.1. A, B, C, D, F and G is based on a less-cf-coolant I accident analysia. The analysis was performd using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. A complete discussion of each code employed in the analysis is presented in Reference 4. Input and model changes in the Peach Ecttom loss-of-coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in Ref erence 8. These changes to the analysis include: (1) consideration of the counter current flow limiting (CCFL) effec:,

(2) ~ corrected code inputs, and (3) the effect of dril'ing alternate flow paths in the bundle lower tie plate.

Amendment No. 22, JA, AA, 48 -140-

~

~~PBAPS Unit 2

, 3.5.I R&ggg (Cont'd.)

A list of the significant plant input parameters to the loss-of- {

coolant accident analysis is presented in Table 3.5-1. ,

i J. Loca1 LBGR This specifica : ion assure ; that the linear heat generation rate in any rod is less than the design linear heat generation if fuel s pellet densification is postulated. The power spike penalty specified is based on the analysis presented in section 3.2.1 of Reference 1 and References 2 and 3, and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with a 955 confidence, that no more than one fuel rod i exceeds the design linear heat generation rate due to power

^

spiking. The LHGR as a function of core height shall be checked daily during reactor operatica at 255 power or greater to determine if fuel buznup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value '

below 255 rated thermal power, the MPPF would have to be greater than 10 which is precluded by a consideracle margin when employing any permissible control rod pattern.

Densification analyses for 8x8 fuel are presented in section 5.2.3 of Reference 7. I K. Minimum Critical Power Ratio (MCPR)

The required oprating limit MCPR's at steady state operating ccnditions as specified in specification 3.5.K are derived from the established fuel cladding integrity safety Limit MCPR of 1.07, and analyses of the abnormal operational transients 1 presented in References 6 and 7. For any abnormal operating b- l

~ transient analysis evaluation with the. initial condition of the

• reactor being a' the steady state operating limit it is required that the result 2 Ty MCPR does not decrease below the Saf ety Limit MCPR at any time *:ing the transient assuming instrument trip setting given ir tecification 2.1.

To assure that tu ~uel cladding integrity safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). The type of transients evaluated were loss of ficw, increase in pressure and power, positive reactivity insertion, and coolan temperature decrease.

l l

l Amendment No. 22, JJ, AB . 4E - 140a-

i PBAPS Unit 2 . . .

3.5.K E gig (Cont'd.)

I The limiting transients which determine the required steady state I r

MCPR limits are given in Table 3.5-2. These transients yield the  ;

?

largest ACPR for each class of fuel. When added to the safety -

l l limit MCPR of 1. 07, the required minimum operating limit MCPR's l j of specification 3.5.K are obtained.

t i

Two codas 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 f allowing assumpticns are included in this determination:

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

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

(3) The cycle.

analysis is performed for the most reactive point in the (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 wcrth control rod is assumed to be operating at the minimum allowable critical poter ratio.

~ The three-diensional BWR code then simclates the core response to the control rod withdrawal error. The second code calculates the Rod Block M'onitor response to the rod withdrawal error. This code simulates the Rod Block Monitor under selected failure ccnditions (LPRM) for the core response (calculated by the 3-dimensicnal BWR simulation code) for the control rod withdrawal.

l The analysis of the rod withdrawal error for Peach Bottom Unit 2 considers the continuous withdrawal of the maximum worth control rod at its maximum drive speed from the reactor which is operating with the limiting control rod pattern as discussed abcve.

~

Amendment No. 22, JA, 48 .-140b l

i_ , . ~

_ . PBAPS . . Unit 2 ..

3.5.K 3&ggg(Cont *d.)

. i A brief summary of the analytical method used to determine the nuclear characteristics is given in section 3 of Reference 7. ,

I Analysis of the abnormal operational transients is presented in i

section 5. 2 of Reference 6. Input data and operating conditions l used in this analysis are abown in Tables 5-6 and 5-8 of I Reference 7 and in Reference 6. , I i

L.  !

4verage Planar LgGR fAPLEGR). Lgiggi LHGR. M Minimum  ;

Critical Egg g M iMCPR) l In the event that the calculated value of APLEGR, LEGR or MCPR exceeds its limiting value, a determiution is made to ascertain ,

the cause and initiate corrective action to restore the value to }

7 within prescribed limits.

The status of all indicated limiting Q 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 Moni*wr -

LPRM, and reactor heat balance instrumentation) , control rod  :

configuration, etc., in order to datermine whether the calculated I values are valid. i In the event that the review indicates that the calculated value exceeding limits is valid, corrective a'ction is immediately '

undertaken to restore the value to within prescribed limits.

Fo11 cuing corrective action, which may involve alterations to the  :

?

control rod configuration and consequently changes to the core power distribution, revised instrumentation data, including changes to the relative neutron flux distribution for up to 43 ,

incore locations is obtained and the power distribution, APLEGR, LHGR and MCPR calculated. Corrective action is initiated within

( one hour of an indicated value emceeding limits and verification

' that the indicated value is within prescribed limits is cbtained '

within five hours of the initial indication.

In the event that the calculated value of APLEGR, LEGR or MCPR exceeding its limita.ng value is not valid, i.e. , due to an ';

erroneous instrumentation indication etc., corrective action is initiated within ene 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 wculd not be a violation of the limiting conditicn for operation and therefore would not constitute a reportable occurrence.

I i

Amendment No. 27, Ja, JJ 48 - 140c-i 1

j ..

PBAPS Unit 2 l 3.5.L BASES (Cont'd. ) i operating experience has demonstrated that a calculated value of .{

APLHGR, LEGR or MCPR exceeding its limiting value predominately occurs due to this latter cause. This experience coupled with l the extremely unlikely occurrence of concurrent operation i exceeding APLEGR, LEGR or MCPR and a Loss of Coolant Accident or '

applicable Abnormal Operational Transients demonstrates that the ,

times required to initiate correc'tive 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 APLEGR, IRGR 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. I M. References

1. " Fuel Densification Effects on General Electr#.: Boiling  :

Water Reactor Fuel", Supplements 6, 7, and C NEDM-10735, Q August 1973. i i

2. Supplement 1 to Technical Report on Densifications of  !

General Electric Reactor Fuels, December 14, 1974 (Regulatory Staff) .

3. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model 1974.

for Fuel Densification", Docket.50-321, March 27, 4

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendiz K, NEDE-20566 (Draft), 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.

Supplemental Reload Licensing Submittal for Peach Bottom l Atomic Power Station Unit 2 Reload No. 2, NEDO-24132, Revision 1, September 1978. l l

7.

General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P-3, March 1978. l l l 8.

Loss-of-Coolant Accident Analysis For Peach Bottom Atomic ,

Power Station Unit 2, NEDO-24081, December 1977. I I

Amendment No. 25, 28, AS. 48 -140d-

l.

l

-PBAPS - -

-43 nit 2 [

TABLE 3.5-1  !

SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS I l

PLANT PARAMETERS: ,

Core Thermal Power 3440 MWt which corresponds t

to 1055 of rated st.eam flow j i

vessel Steam output 14.05 x 10* 1hn/h which corresponds to 1055 of i rated steam flow t (n Vessel Steam Dome Pressure 1055 paia Recirculation Line Break Area For Iarge Breaks -

Discharge 1. 9 f ts (DBA)

\ suction 4.1 fta Assumed Number of Drilled Bundles 360  ;

FUEL PARAMETERS: Peak Technical Ld tial  !

Specification Design Minimum '

Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Tvre _ Ge onetrv (KW/ft) Factor Ratio .

7x7, Type 2 7x7 18.5 1.5 1. 2 t

i 7x7, Type 3 7.x 7 18.5 1.5 1.2 8x8, Type H Bz8 13.4 1.4 1.2 8xS, Type L 8x8 13.4 1.4 1.2 8x8R/LTA 8x8 13.4 1.4 1.2 l A,more detailed list of input to each model and its source is presented in Secticn II of Reference 5.

l w

Amendment Nc. 27, #8 ,48 - 14 0e-

PB&PS ' Unit 2

~ . . . . ..

4.5.L R&qZg (Cont'd) adjusted until the MCPR was slightly above the Safety Limit.

Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the Kf. l For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow.

The Kf factors shown in Figure 3.5.1-E, are acceptable for Peach Bottom operation because the operating limit MCPR is greater than i

the original 1.20 operating limit MCPR used for the generic derivatic . of Kf.

I i

1 i I l

I i

i i

l i

i l

i I

I i

Amendment No. JK, 48 - 141b-

I-i .

PBAPS Unit 2 .

f m

t

-a

-r- += -

_ ,_ ": :- ^

i ,

e- ,

s a i

e

/-

• c

_  : o

,- ..3 1

• _-:.. o .u i -a e

' * ,j -- e u

-m e e c

e

.- 0

e 48 e

-o. ==

= ':

=> e

, - :o%

mo uu es

, ( ' .Nz .es z

eo a,

e au M

' u u s ee a -

o a cm ee E.

~- ':

a c. .eu w

z -

- au a. e c

2 .o a

-e >

e<

e-o e -a e e n o eu 8 ra. a ue i a u oe S 2 e >c o e > aC a=1 m

e u < a.

E

= u o as u me es

<  : =c -e Cn3 s.

'., on nu

- 2.e ge 4 .

.~ m>

1 O

i * ,

~ )

1 .

= c

. 2 .

3 m c

o u

. 3

= .

a.

YW-r .

m =,

n .m a s==t *=9 e=4 eg Maximum Average Planar Linear Heat Generation Rate (KW/FT) 9

-142f-4

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