ML20237C160

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Proposed Tech Specs Revising Sections 3/4.2.1,3/4.3.6 & 3/4.4.1 Re Average Planar Linear Power Ratio,Control Rod Block Instrumentation & Recirculation Sys,Respectively
ML20237C160
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 12/14/1987
From:
Public Service Enterprise Group
To:
Shared Package
ML20237C153 List:
References
NUDOCS 8712210082
Download: ML20237C160 (21)
v • d • e


Text

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1 8712210082 DR 071214 p ADDCK 05000354 PDR

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SCT_1,1NGS \

v 2.la SU?ETY LIMITS e

~yERMAL 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 cor gflow less tha-10% of ran d-flow.

t APPLICAdLITY: OPERATIONAL CONDITIONS 1 and 2.

i 1 ACTION:

( ( --+

g With THERMAL POWR exceMin'q-25% of RATED THERMAL POWER vnd the reactor vessel

~c steam dome pressure less than 785 psig or core flow less than 30% of rated 11o.,.

be in at least H01'5HUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of '-

Specification 6.7.1.

3

<-4 7 THERMAL POWER, High Pressure and High Flow '

/

/ 2.1.2 The MINIMUM CRITICAL KNER RATIO (MCPR) shall not be les with two recirculation loop operaadniand shall not be less thanEds ?.han with single s.

I l.07 / recirculation. loop operation,.in botF cases with the reactor vessel steam dome W/ g ". pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CGNDITIONS I and 2.

AC'10N:

L. _ v With CPF/lessMiiHMl with two recirculation loop operation or less than a with single recirculation loop operation and in both cases with the reactor l

vessel steer l come pressure greater than 785 psig and core flow greater than 10%

of rated fiow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and co7. ply with the requirements of'Specificat!on 6.7.1.

t REACTOR COOLANT SYSTEM PRES 5URE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel l' C steam dome, shall not exceed 1325 psig.

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

Ji i ACTION:

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7

{ With the reactor coolant system pressure, as mea?ured in the reactor vessel I steam dome, above 1325 psig, be in at least 107 [HUTDOWN with reactor coolant l system pressure less than or equal to 1325 pt.ig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with l the requirements of Specification 6.7.1. ~

l HOPE CREEK 2-1 Amendment No. 3 1

~

s E- ---- - e

2.1 SAFET) LIMITS BASES

2.0 INTRODUCTION

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 approact) is used to establish a Safety Limit such that IM, the MCPR is not less tharydt-e6I for two recirculation loop operation ands i for single recirculation loop peration. M W greater tnan M 'for twd re-circulation loop operation and 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 N' 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 gress rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce i onset of transition boiling, MCPR of 1.0. These conditions represent a signi-l ficant departure from the condition intended by design for planned operation, i 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 all elevation head, the core pressure drop at low power and flowswillalwagsbegreaterthan4.5 psi. Analyses show that with a bundle flow of 28 x 10 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 I 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. 3

i I

Bases Table B2.1.2-1

]

UNCERTAINTIES USED IN THE DETERMINATION l 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 Il 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 D l 8.7 J Single Recirculation Loop Operation NW M R Factor D= [ .6l Critical Power 3. 6

  • 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.

HOPE CREEK B 2-3 Amendment No. 3

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3/42-3  !

M M6uks 3.2 3 2, POWER DISTRIBUTION LIMITS 1 1 2 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION i

3.2.3 yThe MINIMUM CRITICAL POWER RATIO (MCPR) shall b equal to or greater thanith: :c c': the MCPR limit shown in Figure 3.2.3-11;';; in: :::utt:r i lN::t!n:  :::::::'; clu:te:nt :t;:r "- T:::1: 2.2.3-11 times the K shown in P Figuren.J.; Z;, with: f l3.2 3 3 l T

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(t,y,. g)

T -I A B where:

T A = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, T

B = 0.688 + 1.65[ 1 ]b(0.052),

" N I i i=1 n

I t,y, , j,7 Nj tj ,

n I N 4

i=1 n = number of surveillance tests performed to date in cycle, Nj = number of active control rods measured in the ith surveillance test, tg = average scram time to notch 39 of all rods measured in the i th surveillance test, and N

y = 4.1.3.2.a. total number of active rods measured in Specification APPLICABILITY: I l

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

of RATED THERMAL POWER.

HOPE CREEK 3/4 2-8  !

1 i

POWER DISTRIBUTION LIMITS i

MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 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-

" " E *b fication 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, MCPR is '

As APPUCABLE w _ determined to be areater than or eaual to the MCPR limit as a function of 1 the average scram time shown in Figure 3.2.3-14, EOC-PPT inoperable curve, iclu: th: f::1 :ter 5::tinc :::::ity :c e:te:r.t ci :r "- ::::: 3. z. R, {

times the Kf shown in Figurej:.e. aw

/S.2 3-31 '

b.

With MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1 P MN and 3.2.3-2,Vplu: the ft:d::ter 50:tinc c! : fty 21tur+ nt ci'/cr

/U:bl:

d-I

      1. 3.2.3-idinitiate corrective action within 15 minutes and restore Fore 12. s-3, MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

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 l within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, p ,g,7 shall be determined to be equal to or greater than the applicable MCPR limit i

J determined from Figures 3.2.3-nc .; .:. c. .: a. j TINES THs Kf 'T 4 i Sl*3wM N a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, )

R6aRs 3.2 3 3: i

b.  ;

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.
d. The provisions of Specification 4.0.4 are not applicable.

HOPE CREEK 3/4 2-9 Amendment No. 1

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t 250 F 0.06 DE(ETE AMD RER. ACE WITH Amps 3.2 3-3 kg FACrcR 1

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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 core flow, or
b. THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:
1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the recirculation flow control system in the Local Manual mode, and b) Reduce THERMAL POWER to 5 70% of RATED THERMAL POWER, and l1'c6'i c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety I Limit by 0.01 t per Specification 2.1.2, and 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 I 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 or the recirculation loop i flos in the operating loop is 150% ** of rated loop flow. l

2. The provisions of Specification 3.0.4 are not applicable. l
3. Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
  • See Special Test Exception 3.10.4.
    • Initial values. Final values to be determined during 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 4-1 Amendment No. 3

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 150%# of rated loop flow;

a. < 145*F between reactor vessel steam space coolant and bottom head Brain 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. 5 50 F between the reactor coolant within the loop not in operation .

and the operating loop. l 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.

l l09%

4.4.1.1.3 Each pump MG set scoop tube mechanical and electrical stop shall be I demonstrated OPERABLE with overspeed setpoints.less than or equal t 6 and We-MI, respectively, of rated core flow, at least once per 18 months. F Y

l lot % 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering 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 octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
  1. Initial values. Final values to be determined during 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. 3

~

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POWER DISTRIBUTION LIMITS BASES I

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 cladding integrity Safety Limit MCPR, and an analysis of abnormal operational ll transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any2.2.

Specification time during the transient assuming instrument trip :>etting given in 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. l:'

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 evaluate 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 reduction in MCPR caused by the transient.

JL /

The purpose of the fK factor of Figurel3.2.3-21is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the K factor. The K factorsassurethattheSafetyLimitMCPRwillnotbeviolafedduringaflo I#' Z'#' 3 I increase transient resulting from a motor generator speed control failure.

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

a The K  ;

andareaphlichbletoallBWR/2,BWR/3andBWR/4 reactors. factors values sh The K, factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. )

{

t For the manual flow control mode, the K g factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the {

I 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. 3

- ~

POWER DISTRIBUTION LIMITS k .

BASES MINIMUM CRITICAL POWER RATIO (Continued)

For operation in the automatic flow control mode, the same procedure was l 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. <

t /

g3,y j The K factors shown in Figure!?.2.3-2 rare conservative for the General Electric p, lant operation because the operating limit MCPRs of Specification 3.2.3 is the same as the original 1.20 operating limit MCPR used for the j generic derivation of K .

f j At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, I 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 MCPR value is in excess of requirements by a considerable margin. Ducing 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 Ac.cordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NE00-10802, February 1973.

t

3. 1 qualification of the One Dimensional Core Transient Model for Boiling Water Reactors, NED0-24154, October 1978.
4. TASC 01-A Computer Program for the Transient Analysis of a Single l' Channel, Technical Description, NEDE-25149, January 1980.

HOPE CREEK B 3/4 2-5 0 ,

- -  :

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