Text

Amend 15 to License NPF-57,permitting Operation During Cycle 2 W/New Fuel Assemblies Loaded During First Refueling Outage & Use of Extended Load Line Limit Operations & Increased Core Flow Operations
	ML20148C039
Person / Time
Site:	Hope Creek
Issue date:	03/13/1988
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20148C042	List: ML20148C039; ML20148C057;
References
	NUDOCS 8803220249
	Download: ML20148C039 (29)
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UNITED STATES

[

NUCLEAR REGULATORY COMMISSION j

j W ASHING TO N, 0. C. 20555 e

PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY 00CKET N0. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY CPERATING LICENSE Amendment No.15 License No. NPF-57 1.

The Nuclear Regulatory Comission (the Comission or the NRC) las found that:

A.

The application for amendment filed by the Public Service Electric &

Gas Company (PSE&G) dated December 14, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

There is reasonable assurance: (i) 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 Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and srfety 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.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.15, and tne Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

8803220249 880315 PDR ADOCK 05000354 P

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3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/s/

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

Attachment:

Changes to the Technical Specifications Date of Issuance:

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3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 15, 1988

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ATTACHMENT TO LICENSE AMENOMENT NO.

15 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Overleaf pages provided to maintain document completeness.*

Remove Insert 2-1 2-1 2-2*

2-2*

B 2-1 R 2-1 8 2-2*

B 2-2*

B 2-3 R 2-3 8 2-4*

B 2-4*

3/4 2-l*

3/4 2-1*

3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4*

3/d 2-4*

3/4 2-7*

l 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 l

3/4 2-11 3/4 2-11 3/4 2-12 3/4 2-12 3/4 3-59 3/4 3-59 3/4 3-60*

3/4 3-60*

3/4 4-1 3/4 4-1 3/4 4-2*

3/4 4-2*

3/4 4-2a 3/4 d-2a l

3/4 4-2b*

B 3/4 2-3*

B 3/4 2-4 R 3/4 2-4 8 3/4 2-5 B 3/4 2-5

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2.0 SAFETY LfMITS AND LIMITING SAFETY SYSTEM SETTXNGS 2.1 SAFETY LIMITS THERHAL POWER, low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with two recirculation loop operation and shall not be less than 1.08 with single recirculation loop operation, in both cases with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 with two recirculation loop operation or less than 1.08 with single recirculation loop operation and in both cases with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10%

of rated flow, be in at least HOT SHUT 00WN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1. 3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

HOPE CREEK 2-1 Amendment No.15

SAr!Tv L w!?S AND t:wi :NG SAFETY SYSTEM SETTINGS Sar!** L:"I'5 (Continued)

G!aC*04 k!5SEL WATER '.EVEL l

2. '. 4 The reactor vessel water level shall be above the top of the j

activa irradiatec fuel.

A P D'. ! C A!! L I T Y : OPERATICNAL CCN0!TIONS 3, 4 and 5 AC* ION:

With the reactor vessel water level at or below the toD of the active l

irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required. Comply with the requirements of Specification 6.7.1.

l l

l I

l HOPE CREEK 22

2.1 SAFETY LIMITS BASES

2. 0 INTRODUCTION l

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation.

MCPR greater than 1.07 for two re-circulation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a signi-ficant departure from the condition intended by design for pla.1ned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means.

This is done by establishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially 311 elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

HOPE CREEK B 2-1 Amendment No.15

SACUY L M *5 BASE 5 2.1. 2 TWERMAL DOWER, W qn Dressure and High Flow The fuel claccing integrity Safety Limit is set such that no fuel camage is calculatec to occur if tne limit is not violated.

Since the parameters wnich result in fuel damage are not cirectly observable during reactor opera: :-

the thermal and hydraulic conditions resulting in a departure from nucleate toiling have been used to mark the beginning of the region where fuel damage could occur.

Althougn it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a conven'ert limit. However, the uncertainties in monitoring the Core operating state anc in the p-ocedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Sa'ety Limit is define as the CPR in the limiting fuel assemely for which more than 99.9% of the fuel rocs in the core are expected to avoid boiling transition consicering tne power cistribution witnin the core and all uncertair-ties.

The Safety 1.imit MCPR is determined using the General Electric Thermal 8

Analysis Basis, GETAB, which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calcula;e critical power. The probability of the oce rrence of boiling transition is determinec using the General Electric critical Quality (x) Boiling Lengtn (L),

(GExt), correlation.

The GEXL correlation is valid over the range of conditions used in the tests of the cata used to develop tne correlation.

The reouired input to the statistical model are the uncertainties listec in Bases Table 82.1.2-1 and tne nominal values of the cors parameters listeo in Bases Table B2.1.2 2.

The bas.es fo-the uncertainties in the core parameters are given in D

NE00-20340 and the basis for the uncertair,ty in the GEXL correlation is given in NE00-10958-A*. 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 the greatest numM r of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis, "ceneral Electric BWit thermal Analysis Bases (GETAB) Data, Correlation anc a.

Design Applicatien," NE00-10958 A.

b.

General Electric "Process Computer Performance Evaluation Accuracy" i

NE00-20340 and Amencment 1, NE00-20340-1 dated June 1974 and December l

1974, respectively.

I HOPE CRE[K B22 l

l l

ga,ses Table 82.1.2-1 U"rERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

Standard Deviation Quantity

(% of Point)

Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop Operation 2.5 Single Recirculation Loop Operation 6.0 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings Two Recirculation loop Operation 8.7 Single Recirculation Loop Operation 9.1 R Factor 1.6 l

Critical Power 3.6 t

"The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core. The values herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.

t l

HOPE CREEK B 2-3 Amendment No.15 l

. ~ _ _ - - - -. _

Bases Table 92.1.2 2 N0u!NAL VALUES OF DARAW[T[Q$ U$[0 IN Tdi 5'A5'::aL'ANa'.v5IS Or UEL CLA00!NG INT!3RI'v SAFE *'V:'

TwERWAL PCnER 5323 MW Core Flew 108.5 M1b/hr Co e Pressure 1010.4 psig 2

Channel Flow Area 0.1089 ft R Fact:r High enrichment - 1.043 Medium enrichment - 1.039 Low enrichment - 1.030 r

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s:: ::N rec CD!RA'!ON 3 - '. Ali A',ERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHG#s) fo ea: t,:e e' fue as a fun: tion of AVERA3E PLANAR EXPOSURE shall not exceec tme 1 s'ts s. - i-Figs es 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, anc 3.2.1-5.

Tne lie. ts e

cf Fig -es 3. 2.1-1, 3. 2.1-2, 3. 2.1-3, 3. 2.1-4 and 3. 2.1-5 shall be recu:e: t:

I a vai.e cf 0.85 times the two recirculation loop operation limit when ir s: ;'e

-e:ir: 'ation icop operation.

A::.::12:.:~': 00ERATIONAL CON:ITION 1, when THERMAL P0w!R is greate-t a-e:.a' t: 25% of RATED THERMAL POWER.

A: : N.

. t a-A LH3: esceeding the limits of Figurg 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1 A, er 3.2.1 5, initiate corre:tive a: tion witnin 15 minutes an: rest:-e A:.-3: t:.itrir the reovire: limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or recu:e THERMAL PD.ER ::

le.. tra-25'i c' RA*EC THERMA. POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURv!:'LAN:! REOUl#EWEN'S A? '. A11 APLw3:s shall ce verified to be eaual to or less than the limits

etea.'ne: from Figu es 3.2.1 1, 3.2.1-2, 3.2.1 3, 3.2.1-4 and 3.2.1-5:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

c.

The provisions of Specification 4.0.4 are not applicable.

HODE CREEK 3/4 2*1 Amendment hc. 3 APA 7 iso 7 1

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10 15 20 25 30 35 40 AVERAElf PLAMAll EXP050RE. GWd/st MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE TUEL TYPE P8CIB163 Figure 3.2.1-3

aE: ::**::!.*:0N.:w:~5 34 2: A :" SET:CIN*5

.:v:~:N CON:'*!CN 20 C:!:A*:CN 3 2.2 Tee APRM flo-Diased simulate: thermal po er upscale scram trip set:: --

(5) aa: '::. :'asec neutrer flui upscale control ro block trip setpoint (5.."

Small te esta:lishe: a::cading tc the fo11 ewing relationships:

TRID SEID0!NT ALLOWABLE VAtVE

$ < ( 0. 66 ( w-a.)" + 51*.)T 5 < (0.66(w-aw)" - 54*.)i S j 1 (0.66(w-a=)" + 45*.)*

S j 1 (0. 6E(w-a.)"

42*.)i p

p w.eee:

5 an: 5,n are in cercent of RATED THERMAL POWER, Le:; re:ir:ulation flow as a percentage of the loep recirculat.:-

W = fle= which produ:es a rated core flow of 100 million Ibs/hr.

T = 60.est value of the ratio of FRACTION OF RATED THERMAL POWES (FR*:) civice: ty the CORE MAxlMJM FRACTION 0F LIMITING P0.!:

DEN 5:TY (CMrLCD). T is applied only if less than or ecual t:

A:h::AS:.:in OPERATIONAL CON 0! TION 1, when THERMAL POWER is greater tna er e:.a 1: 2 P2 c' RATE 0 TrERMAL POWER.

A"':0N:

W4tr t*e AO:* 'lom biase: sieulated the* ai powe* upscale scrat t'ip set::' :

a-:/c tre 'le. t'ase: ne.t-:- flux-upscale contrcl roc Dio:k trip set:ci-t less cerse vat ve tnan tne value sho=m in the Allowable Value column for 5 or 5...

as i

a: se cete- 'ae:. 'r'tiate cer e:tive actio within 15 minutes an a just ! ar:

c-5.. t ce c<ns' stent witn the Trip Setooint values" within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or re:w:e THER 3. P;mER 6e less than 25*. cf RATED THERMAL P0nER within the next a nou s.

0 SL:.E:wLANCE REQUIREMEN'S 2.2 ine FRTD anc the CMFLDC shall be determined, the value of T calculate:.

a: t e e:st re:ent a:tuai ASRW 'lo eiasec simulated thermal power u:sca'e s: car an: fic. ciase: neutron flux upscale control rod block trip setpoints vec,'iec 10 te within the abeve limits or adjusted, as required:

a At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , t.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER inersase cf at least 15% cf RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operatin; c.

with CHFLPD greater than or equal to FRTP.

c.

The provisions of Specification 4.0.4 are not applicable.

"=it-CMFLDD greater than the FRTP, rather than adjusting the APRM setDoints, t e l

APRM gain may be adjusted such that the APRM readings are greater than or ecua' to 100*. times CHFLP0 provided that the adjusted APRM reading does not excee:

100*. of RATED THERMAL POWER ano a notice of adjustment is posted on the rea :-

conteel panel.

"The Average Power Range Monitor Scram function varies as a function of recia:u-latice loop drive flow (w).

Sw is defined as the difference in indicatec c*ise fic. (in percent of drive flow which produces rated core flow) between two IC::

anc single loop operation at the same core flow. Aw = 0 for two recirculation loop operation. Aw = "To be aetermined at a later date" for single retirculatica loep operation.

HODE CRE!A 3/4 2 7 Amendment No. 3 APR 7 IW

_____________________u

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figure 3.2.3-1 or Figure 3.2.3-2, as applicable, times the K shown in Figure 3.2.3-3, with:

7 (Iave IB) l I

~I A

B where:

t = 0.86 seconds, control rod average scram insertion A

time limit to notch 39 per Specification 3.1.3.3, l h g = 0.688 + 1.65[

J (0.052),

t N

I i

i=1 n

I t

Nt ave =

i=1 9g n

I NI i=1 n = number of surveillance tests performed to date in cycle, th Ng = number of active control rods measured in the i surveillance tes'.;,

4 = average scram time to notch 39 of all rods measured 1

th in the i surveillance test, and Ny = 4.1.3.2.a. total number of active rods measured in Specification APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25%

l of RATED THERMAL POWER.

1 HOPE CREEK 3/4 2-8 Amendment No. 15

POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR ODc"ATION ACTION:

a With the end-of-cycle recirculation pump trip system inoperable per Spe-cification 3.3.4.2, operation may continue and the provisions of Speci-fication 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the MCPR limit as a function of the average scram time shown in Figure 3.2.3-1 or Figure 3.2.3-2, as applicable, EOC-RPT inoperable curve, times the K shown in Figure 3.2.3-3.

f b.

With MCPR less than the applicable MCPR limit shown in cigures 3.2.3-1 and 3.2.3-2, times the K shown in Figure 3.2.3-3, initiate corrective l

f action within 15 minutes and restore MCPR to within the required limit withir. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERKAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

l a.

t = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or b.

I as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit shown in determined f rom Figure 3.2.3-1 or Figure 3.2.3-2, times the Kf Figure 3.2.3-3:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a LIMITING CONTROL R00 PATTEPH for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

HOPE CREEK 3/4 2-9 Amendment No.15

)

i

1.4 1.39 1.38 1.37 --

I'M j EOC-ITi IHOPERABLE j 4

1.35

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34 7

g 1.33 _ _____

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EOC-fFT #0 MAIN TWBIE s

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BYPASS OPOUSLE 1.23 I

1.22 i

1.21

1. 2 ----

0.2 0.4 0.6 0.8 1

i Tay MIN? HUM CRITICAL POWER RATIO (MCPR)

VERSUS TAU AT HATED FLOW BOC 'O E0C-2000 MWO/ST l

Figure 3.2.3-1 HOPt CREEK 3/4 2-10 Amendment No.15

1.4 1.39 1.38 1.37 l EOC-RPT INOPERABLE l 1.36 1

\\

s s.n 1.34

,sedd p

1.32 -

i.31 #

,,.m __

M 1.3 Y

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1.as 1.27 1.26 EOC-fFT AM) MAIN TipilE BYPASS OIERABLE 3,3 1.24 1.23 1.22 1.21 1.2 0

0.2 0.4 0.6 0.8 1

Tau MINIMUM CRITICAL POWER RATIO (MCPR)

\\ERSUS TAU AT RATED FLOW EOC-2000 MWD /ST TO EOC

(

Figure 3.2.3-2 HOPE CREEK 3/4 2-11 Amendment No. 15

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HOPE CEEK 3/42-12 Amendment No.15 l

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENIATION SETPOINTS IRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE g

A 1.

ROD BLOCK MONITOR n

a.

Upscale N

1.

Flow Biased

< 0.66 (w-Aw) + 40%*

< 0.66 (w-Aw) + 43%*

L ii. High Flow Clamped i106%

5109%

b.

Inoperatire NA NA Downscale

> 5% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER 2.

APRN a.

Flow Biased Neutron Flux -

i Upscale

$ 0.66(w-Aw) + 4 N*

< 0.66(w-Aw) + 45%"

b.

Inoperative NA NA c.

Downscale

> 4% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER d.

Neutron Flux - Upscale, Startup 312%ofRATEDTHERMALPOWER jI4%ofRATEDTHERMALPOWER 3.

SOURCE RANGE MONITORS

-~

a Detector not full in NA NA 0

5 R

b.

Upscale

< 1.0 x 10 cps

< 1.6 x 10 cp3 c.

Inoperative NA NA

{

d.

Downscale

> 3 cps

> 1.8 cps 4.

IHTERMEDIATE RANGE MONITORS o

Detector not f ul! in MA NA 3.

Upscale 5 108/125 divisions of

$ 110/125 divisions of full scale full scale c.

Inoperative NA NA d.

Downscale

> 5/125 divisions of

> 3/125 divisions of fu11 scale full scale 5.

SCRAM DISCHARGE VOLUME a.

Water Level-High (Float Switch) 109'1" (North Volume) 109'3" (North Volume) 6.

REr.CTOR COOLANT SYSTEM RECIRCULATION FLOW

{

a.

Upscale

< 108% of rated flow

< 111% of rated flow t.

Inoperative NA NA

=

S c.

Comparator

< 10% flow deviation

< 11% flow deviation I

7.

REACTOR MODE SWITCH SHt'TDOWN POSITION NA NA 5

The rod block function is varied as a function of recirculetion loop flow (w) and Aw which is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two loop and single loop operation at the same core flow.

The trip sett.ing of the Average Power Hange Monitor Rod Block function must be maintained in accordance with Specification 3.2.2.

I ABi f 4. 3. 6___1 M

CONIROL ROD RIOCK IN51RllMINIAll0N SURVilll ANCf RIQUIRtMINIS 4

CHANN(l l

')

OPIRAi10NAL CHANN(l IllNCIID'4AL CHANNII CONDIIIONS IOR WHICH l

TRIP IUNCTION g

Clll CK li st CAllBRAll0N,)

r SifRVEIllAleCl RIQllRIO I.

ROD Bl0CK SONITOR l

a.

Upscale NA 5/U

,N 5A l'

i b.

Inoperative NA 5/U N

NA c.

Downscale NA 5/Ug ),Mg3 l'

SA l*

2.

APilM a.

Flow Blased Neutro. Ilum -

. I I

Upscale NA 5/U

,M 5A I

I

)

b.

Inoperative NA 5/U

,M NA 1, 2, 5 c.

Downscate NA 5/U M

5A I

1 d.

Neutron FIux - Upscale, Startup NA

$/Ug,M SA 2, 5 3.

SOURCE RANCE fGNITORS, a.

Detector not full in NA 5/UI '} W NA J,

b.

Upscale NA 5/U( I,W 2, 5 SA 2, 5 c.

Inoperative NA 5/U

,W NA 2, 5

{

d.

Downscale NA 5/U

,W SA Z, 5 1

4.

INTEREDIATE RAsIGE MONITORS 1

a.

Detector not full in M4 5/UI ',W i4A b.

Upscale NA g

2, 5 5/U 5A 2, 5 5/ll(p,),W q

c.

Inoperative NA 4

d.

Downstale NA 5 /'8 h)*W SA 2, 5 j

5.

SCRAM DISCNARGE V0lupE Water level-High (Float % itch)

NA M

R 1, 2, 5**

a.

6.

Rf ACTOR C00 TANT SY51lM RECIRCUI All0N I10W j

a.

Upscale II'I NA 5/tl 5/u 'I M SA 1

b.

Inoperative II NA 4

II'I M MA I

c.

Comparator NA 5/II

.M 5A I

4 7.

Rf ACTOR fEDE SWIICH SHUIDOWN i

NA R

NA I, 4 I

4 i

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

a.

Total core flow greater than or equal to 45% of rated cora flow, or b.

THERMAL POWER leii chan or equal to the limit specified in Figure 3.4.1.1 1.

APPLICABILITY: OPERATIONAL CONDITIONS la and 2*.

ACTION:

a.

Witn one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

Place the recirculation flow control system in the Local Manual mode, and b)

Reduce THERMAL POWER to 1 70% of RATED THERMAL POWER, and c)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and l

d)

Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.86 times the two recirculation loop limit per Specification 3.2.1, and e)

Reduce the Average Power Range Monitor (APRM) Scram and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1, 3.2.2 and 3.3.6, and f)

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and g)

Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is < 30% ** of RATED THERMAL POWER c:t the recirculation loop flow in the operating loop is 1 50% ** of rated loop flow.

l 2.

The provisions of Specification 3.0.4 are not applicable.

3.

Otherwise be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

See Special Test Exception 3.10.4.

- Initial values.

Final values to be determined during Startup Testing based i

upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

l l

l HOPE CREEK 3/4 4-1 Amendment No. 15 l

1 l

REAC700 MOLAN' SYSTEu SURVIILL4NCE REQU:#E9ENTS With no reactor coolant system recirculation loops in operation, b.

immediately initiate 1ction to reduce THERMAL POWER to less tha9 equal to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months anc initiate measures to Deice the unit in at least STARTUP within 6 me., s anc in hcl $ HUT 00WN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i With one or two reactor coolant system recirculation loops in opera-c.

tion and total core flow less than 45% but greater than 39V of rated core flow and THERMAL POWER greater than the limit specif*e:

i-Figure 3.4.1.1 1:

1.

Determine the APRM and '.PRM' noise levels ($urveillance

4. 4.1.1. 4 ):

a)

At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and b)

Vithin 30 minutes af te the comoletion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

2.

With the APRM or LPRM' neutron flux noise levels greater than t9ree times their established baseline noise levels, witnin 15 minutes initiate corrective action to restore the noise levels to within the require li. tits.

'n 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow to greater than 45% of rat.

are flow or by reducing THER-MAL POWER to less than or equal - the limit specified in Fig-ure 3.4.1.1-1.

d.

With one or two reactor coolant system recirculation loops in ope *ation and total core flow less than or equal to 39%s and THERMAL POWER gaette*

than the limit specified in Figure 3.4.1.1-1, within 15 :inutes int ate corrective action to reduce THERMAL POWER to less than or ecual to the limit specified in Figure 3.4.1.1-1 or increase core flow to greate*

than 39%# within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.4.1.1.1 With one reactor coolant systes recirculation loop not in operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that:

Reactor THERMAL POWER is 5 70% of RATED THERMAL POWER, and a.

1 The recirculation flow control systee'is in the Local Manual mode, b.

and The speed of the operating recirculation pump is less than or equal c.

to 90% of rated pump speed, and Core flow is greater than 39V when THERMAL POWER is greater than d.

the limit specified in Figure 3.4.1.1-1.

"Detector levels A and C of one LPRM string per core octant plus detectors A l

and C of one LPRM string in the center of the core should be monitored.

lnitial values.

Final values to be determined during Startup Testing (core flow with both recirculation pumps at a minimum pump speed).

HOPE CREEK 3/4 4-2 Amendment No. 3 APR 7 W

, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.2 With one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recir-culation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30%# of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is 5 50%# of rated loop flow:

a.

5 145*F between reactor vessel steam space coolant and bottom head drain line coolant, and b.

< 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.

1 50 F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements or Specifications 4.4.1 1.2b and 4.4.1.1.2c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

4.4.1.1.3 Each pump HG set scoop tube mecnanical and electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 109% and 107%, respectively, of rated core flow, at least once per 18 months.

4. 4.1.1. 4 Establish a baseline APRM and LPRM* neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of enterirg the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.

"Detector levels A and C of one LPRM string per core octan

plus detectors A and C of one LPRM string in the center of the core should be monitored.

Initial values.

Final values to bn determined curing Startup Testing based upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

HOPE CREEK 3/4 4-2a Amendment No. 15

THIS PAGE INTENTIONALLY LEFT BLANK HOPE CREEK 3/4 4-2b Amendment No. 3 gh T

Eases 'ab e B 3.2.1-1 l

513s!r!:AN' INDU' PARAWETERS TO THE LCi!.:r.:o;;As* A CIDENT ANALYSIS cia-; c a eters:

a C:re Ta!RVAL 80=ER 3430 Mwt* which coraesec-es to 105% of ratec steam flo.

Vesse' Stear Oct: t 14.87 x IOC lbm/hr whien corresponds to 105% of ratec steam flow vesse Steat 00 e Dressure.

1055 psia Cesig-Basis Recirculation Line Betas Area for:

a.

Large B*eaks 4.1 ft2 Sea'1 BeeaAs 0.09 fts, Eue'. Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRt (kw/ft)

FACTOR RATIO Ir.itial Core 8x6 13.4 1.4 1.20**

A mere cetailec listing of input of each model and it; source is presentec in Section !! of Reference 1 and subsection 6.3.3 of the FSAR.

This pc.er level meets the Appendix K requirement of 102%. The core.

heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

- For single recirculation loop operation, loss of nucleate boiling is assumec at 0.1 secones af ter LOCA regardless of initial MCPR.

MOPE CREEK 8 3/4 2-3 Amenoment No. 3 APR 7 567

POSER OfSTRfBUT10N LXMfTS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel l

cladding integrity Safety Limit MCPR, and an analysis of cbnormal operational i

transients.

For any abnormal operating transient analysis evaluation with the I

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 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-3 that are input to a GE-core dynamic behavior transient computer program. The code used to tvaluate pressurization events is described in NE00-24154(3) and the program used in non pressurization events is described in NEDO-10802(2)

The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE-25149(4)

The principal result of this evaluation is the reducticn in MCPR caused by the transient.

The purpose of the K factor of Figure 3.2.3-3 is to define operating l

limitsatotherthanratebcoreflowconditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the K, factor.

The K 7 factors assure that the Safety Limit MCPR will not be violated during a flow incresse transient resulting from a motor generator speed control failure.

The K factors may be applied to both manual and automatic flow control modes.

7 The K factors values shown in Figure 3.2.3-3 were developed generically I

andareaphlicabletoallBWR/2,BWR/3andBWR/4 reactors.

The K factors were f

derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow.

For the manual flow control mode, the K ^ actors were calculated such that 9

for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows.

The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the K.

f HOPE CREEK B 3/4 2-4 Amendment No. 15

POWER OfSTRXBUTXON LIMXTS BASES

$NIMUMCRITICALPOWERRATIO(Continued)

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 THERMAL POWER and rated thermal flow.

The K factors shown in Figure 3.2.3-3 are conservative for the General l

ElectricpfantoperationbecausetheoperatinglimitMCPRsofSpecification 3.2.3 is the same as the original 1.20 operating limit MCPR used for the generic derivation of K.

7 At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting !!CPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

(

References:

1.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

2.

R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NEDO-10802, February 1973.

l 3.

Qualification of the One Dimensional Core Transient Model for l

Boiling Water Reactors, NEDO-24154, October 1978.

4.

TASC 01-A Computer Program for the Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.

HOPE CREEK B 3/4 2-5 Amendment No. 15 t

i l

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