ML20141C162

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Proposed Tech Specs Revising Sections of TSs to Delete Reference to Rod Sequence Control Sys & to Reduce Rod Worth Minimizer Low Power Setpoint from 20% to 10%
ML20141C162
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 06/19/1997
From:
Public Service Enterprise Group
To:
Shared Package
ML20141C158 List:
References
NUDOCS 9706250018
Download: ML20141C162 (13)
v • d • e


Text

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b l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGs 3/4.0 APPLICABILITY................................................ 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS .

3/4.1.1 SHUTOOWN MARGIN........................................... 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES...................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability...................................

3/4 1-3 Control Rod Maximum. Scram Insertion Times.... r........... 3/4 1-6 Control Rod 6verage Scram I,nsertion Times................. 3/4 1-7 Four Control Rod Group Scram Insertion Times.............. 3/4 1-8 Control' Rod Scram Accumulators............................ 3/4 1-9 Control Rod Drive Coupling................................ 3/4 1-11

' Control Rod Position Indication............................' 3/4 1-13 Control Rod Drive Housing Support......................... 3/4 1-15 3/4.1.4 CONTROL R00 PROGRAM CONTROLS (De,le ted) a Rod Worth Minimizer............ .......................... 3/4 1-16 Rod Sequence Control System............................... 3/41-17 Rod Block Monitor......................................... 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................. 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution Volume /

Concentration Requirements............... 3/4 1-21 3/4:2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION- RATE. . . . . . . . . . . . . . . . 3/4 2-1 l

! 9706250010 970619 PDR ADOCK 05000354 P PENT ,

HOPE CREEK v , Amendment No. 34

i t

INDEX LIMITING CONDITIONS FCR OPERATION AND SURVEILLANCE REQUIREMENTS IIiM P.A91 3/4,9.11 RESIDUAL HEAT Rt.MOVAI. AND COOLANT CIRCULATION Mign Water Leve1........................................ 3/4 9 17 l Low Water Leve1..................................... ... 3/4 9-18 l 3/4.10 SPECIAL TEST EXCEPTIOtra WORTH MINIMtMR 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY. . . ....................... 3/4 10-1 3/4.10.2 ROD l!!^75MC5 O^MTE^i  !!!!:M............................. 3/4 10-2 3/4.10.3 SKU.*DOWN KARGIN DEMONSTRATIONS.......................... .

3/4 10-3 3/4.10.4 RECIROULATION LOOPS..................................... 3/4 10-4 3/4.10.5 OxycEN CONCENTRATION.................................... 3/4 10-5 3/4.10.6 TRAINING STARTUPS....................................... 3/4 10-6 l

i 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.......... 3/4 10-7 3/4.10.8 INSERVICE LEAR AND KYDROSTATIC TESTING................. 3/4 10-8 l l 3/4.11 RADIOACTIVE EFFLVTd[T1 3/4.11.1 LIQUID EFFLUENTS Concentration........................................... 3/4 11-1 Table 4.11.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program... 3/4 11-2 Dose.................................................... 3/4 11-5 4 l

Liquid Waste Treatment.................................. 3/4 11-6 Liquid Holdup Tanke..................................... 3/4 11-7 ,

1 3/4.11.2 GASEOUS EFFLUENTS I l

I

' Dose Rate............................................... 3/4 11-8 l l

l Table 4.11.2.1.2-1 Radioactive Gaseous Waste  !

Sampling and Analysis Program... 3/4 11-9 l l

Dose - Noble Gases...................................... 3/4 11-12 I . Dose - Iodine-131, Iodine-133, Tritium and Radionuclides in Particulate Form. . . . . . . . . . . . . . . . 3/4 11-13 i Gaseous Radwaste Treatment.............................. 3/4 11-14  ;

l Ventilation Exhaust Treatment System. . . . . . . . . . . . . . . . . . . . 3/4 11-15 i

HOPE CREEK xy Amendment No. 69 l l

i

I 1

INDEX 1 1

BASES PAGE SECTION 3/4.10 SPECIAL TEST EXCEPTIONS WORTH MIMIMI2ER 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY ............ B 3/4 10-1

/ B 3/4 10-1 3/4.10.2 ROD ""*"""" " - " e v e '""

3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS ........... . B 3/4 10 1 3/4.10.4 RECIRCULATION LOOPS . . . . . ...... ...... B 3/4 10-1 3/4.10.S OXYGEN CONCENTRATION . . . . . ............. B 3/4 10-1 3/4.10.6 TRAINING STARTUPS . . . . . . ..... ....... B 3/4 10-1 3 /4.10."1 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING

..... B 3/4 10-1 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING ......... B 3/4 10-2 3/4.11 RADIOACTIVE EFFLUENT _S

3. 4.11.1 LIQUID EFFLUENTS Concentration . . . . . . . . ..... ..... . B 3/4 11-1 Dose . . . . . . . . . . . . . ............. B 3/4 11-1 Liquid Radwaste Treatment System . ........ .. B 3/4 11-2 Liquid Holdup Tanks . . . . . ............. B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS Dose Rate . . . . . . . . . . ............. 3 3/4 11 2 Dose - Noble Games . . . . . . ............. B 3/4 11-3 I Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form .......... B 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation Exhaus,t Treatment Systems ........ B 3/4 11-4 Main Condenser . . . . . . . . ............. B 3/4 11-5 Venting or Purging . . . . . . ............. B 3/4 11-5 3/4.11.3 SOLID RADIOACTIVE WASTE TREATMENT ....... ... B 3/4 11-5 3/4.11.4 TOTAT., DOSE . . . . . . . . . . ............. B 3/4 11-5 l

HOPE CREEK xxi Amendreent No.91 l

i

i .

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to nssist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift q specifically allocated for each trip in the safety analyses.

l 1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120

divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM

, and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conserva-tism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure thres-i hold of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accorrmodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-4 ture coefficients are small and control rod patterns are constrained by the R:CS =d RWM. Of all the possible sources of reactivity input, uniform con-trol rod withdrawal is the most probable cause of significant power increase.

HOPE CREEK B 2-6

i. l
REAC'TIVITY CONTROL SYSTEMS  !

I 1

LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

2. If the inoperable control rod (s) is inserted, within one hour disarm the associated directional control valves ** either:

a) Electrically, or b) Hydraulically by closing the drive water and exhaust water

} isolation valves.

t j 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 />.

l \

3. The provisions of Specification 3.0.4 are not applicable.

! c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN  !

j within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, j

?

d. With one scram discharge volume vent valve and/or one scram discharge i

volume drain valve inoperable and open, restore the inoperable valve (s) i to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within i the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

! e. With any scram discharge volume vent valve (s) and/or any scram discharge volume drain valve (s) otherwise inoperable, restore the inoperable valve (s) l i

to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within l the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l 2 i

! SURVEILLANCE REQUIREMENTS i

)

> \

4.1.3.1.1 The scram discharge volume drain and vent valves shall be 4 i demonstrated OPERABLE by: l i

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 /> verifying each valve to be open,* and

, b. At least once per 31 days cycling each valve through at least one

} complete cycle of full travel.

i 4.1.3.1.2 When above the low power setpoint of the RWM rd RSC., all withdrawn ,

control rods not required to have their directional control valves disarmed i l

  • These valves may be c,losed intermittently for testing under administrative i controls.
**May be rearmed intermittently, under administrative control, to permit
testing associated with restoring the control rod to OPERABLE status.

4 i

  • HOPE CREEK 3/4 1-4 y

REACTIVITY CONTROL SYSTEMS CONTDCL ROD ORIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1. 3. 6 All control rods shall be coupled to their drive mechanisms.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a. In OPERATIONAL CONDITION 1 and 2 with one control rod not coupled to its associated drive mechanism, within 2 h  :
1. If permitted by the RWM crd RSSS, insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod, and:

a) Observing any indicated response of the nuclear instrumentation, and b) Demonstrating that the control rod will not go to the overtravel position.

2. If recoupling is not accompli permiltedbytheRWMcrRSCS,pdonthefirstattemptor,ifnot then until permitted by the RWM EG F, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves ** either:

a) Electrically, or l b) Hydraulically by closing the drive water and exhaust water isolation valves.

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 />,

b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1. Insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or
2. If recoupling is not accomplished, insert the control rod and disarm I the associated directional control valves ** either:

a) Electrically, or i b) Hydraulically by closing the drive water and exhaust water isolation valves.

c. The provisions of Specification 3.0.4 are not applicable.

^At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

    • May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

HOPE CREEK 3/4 1-11

' REACTIVITY CONTROL SYSTEMS CONTROL ROD POSITION INDICATION l LIMITING CONDITION FOR OPERATION l

$ 3.1.3.7 The control rod position indication system shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*,

ggg gg ACTION:

me had

a. In OPERATIONAL CONDITION 1 or 2 with one or more control rod position
indicaters inoperable, within 1 hour:

2

1. Determine the position of the control rod by#bti'i:ing the RSGS cub titute p0:itica-44: play withi precet pe':!cr level, or:

3 a) Moving the control rod, by single notch movement, to a position.

with an OPERABLE position indicator, 1

b) Returning the control rod, by single notch movement, to its original position, and c) Verifying no control rod drift alarm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or l

2. Move the control rod to a position with an OPERABLE position indicator, or 3.

When THERMAL POWER is: l a) Within the preset power level of the iNHNi, declare the control j rod inoperable.

RbVM b) Greater than the preset power level of the RSES, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves ** either:

1) Electrically, or
2) Hydraulically by closing the drive water and exhaust water isolation valves.

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 />,

b. In OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable, move the control rod to a position with an OPERABLE position indicator or insert the control rod.
c. The provisions of Specification 3.0.4 are not applicable.

"At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

    • May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

HOPE CREEK 3/4 1-13

1 j ,

REACTIVITY CONTROL SYSTEMS j 3/4.1.4 CONTROL ROD PROGRAM CONTROLS

} ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION i

3.1.4.1 i The rod worth minimizer (RWM) shall be OPERABLE.

j APPLICABILITY:

than er equal to OPERATIONAL 205 COND IONS 1 and 2*#, when THERMAL POWER is less i setpoint. f RATED THERMAL S0WER, the minimum allowable low power 10'/*

, ACTION: a ter the u 412. opero. tion May cont'inue prowded t%.t M VCf tbfd (edro rods afe_gutly wdhk j a.

With the RWlfinoperab e, ver'fy control rod movement and compliance i with the prescribed control rod pattern *by a second licensed operator

! or other technically qualified member of the unit technical staff

who is present at the reactor control console.4 i

Otherwise control gg i

the reactor mode switch in the Shutdown position. Arod move 1

} SURVEILLANCE REQUIREMENTS s *.?

4.1.4.1 f.

The RWM shall be demonstrated OPERABLE:

1 a.

i In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of

control rods for the purpose of making the reactor critical, and in l OPERATIONAL CONDITION 1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, prior to RWM automatic initia-i tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod..

i b.

j In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of i control rods for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod. -

c.

In OPERATIONAL CONDITION 1 within one hour after RWM automati initiation when reducing THERMAL POWER, by verifying the rod block functionrod.

control by demonstrating inability to withdraw an out-of-sequence d.

.By verifying that the control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer. .

x Entry into OPERATIONAL CONDITION 2 and , withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWH prior to withdrawal criticality. of control rods for the purpose of bringing the reactor to

  1. See Special Test Exception 3.10.2.

HOPE CREEK 3/4 1-16 Amendment No,' 19- * ' '

"~"*

1 NSERT8 s

REACTIVITY CONTROL SYSTEMS i

4 ROD SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION

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HOPE CREEK 3/4 1-17 d

^

j ,

SPECIAL TEST EXCEPTIONS WORTH MIMIMI2 Git

/ .

! 3/4.10.2 R0000:NC: CONTRCL SYCT "

4 LIMITING CONDITION FOR OPERATION 1

3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 and by the cd requence centro!

cytter (PSCS) per Speci'icatier 3.1.d.2- may be suspended by mean: ef byp;;;

switches for the following tests provided that control rod movement prescribed 4

' for this testing is verified by a second licensed operator or other technically  !

qualified member or the unit technical staff present at the reactor console: j 4

a. Shutdown margin demonstrations, Specification 4.1.1. '

1 t

4

b. Control rod scram, Specification 4.1.3.2.

i c. Control rod friction measurements.

i APPLICABILITY:

when TtlERM AL POWR 15

, OPERATIONAL CONDITIONS 1 and 2. [e ss .Lhaps or ec cod to 10 */*

a ACTION: of RATED THTRMAL PdMR l

With the requirements of the above specification not satisfied, verify that the RWM and/ r the RSCS is OPERABLE per Specificationt 3.1.4.1 and 3'.1.?.2, l rc:pec ti':c?y.

a SURVEILLANCE REQUIREMENTS 4

i

) 4.10.2 When the sequence constraints imposed by the PSCS and/cr RWM are bypassed, verify:

a. That movement of the control rods from 75% ROD DENSITY to the RvVM -RSGS-1 w power setp int is limited t the approved control rod withdrawal sequence during scram and friction tests,
b. That movement of control rods during shutdown margin demonstra-tions is limited to the prescribed sequence per Specification 3.10.3.

! c. Conformance with this specification and test procedurec by a second licensed operator or other technically qualified member of the unit technical staff.

J

HOPE CREEK 3/4 10-2 Amendment No. 35

s .

REACTIVITY CCNTROL SYSTEMS BA'SES I

3/4.1.4 CCNTROL ROD PROGRAM CCNTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER

[lh ,j)) is greater thad*991 of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 260 cal /gm. Thus requiring the ECCC n'. d / : RWM to be ,

CPERABLE when THERMAL POWER is less than or equal to ee% of RATED THERMAL POWER provides adequate control.

The-nCC cad RWM provide utomatic supervision to asaure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.4 IN SEIRT C The RBM is designed to automatically prevent fuel damage in the event of i erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.

t g;pg CREEK B 3/4 1-3 Amendment No. 98 j

- ~ _

4 REACTIVITY CONTROL SYSTEMS

, BASES rate, solution concentration or boron equivalent to meet the ATWS Rule must not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 j states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron con'.^ol equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution (natural boron

) enrichment)."

The described minimum system parameters (82.4 gpm, 13.6 percent concentra-tion and natural boron equivalent) will ensure an equivalent injection capability that exceeds the ATWS' Rule requirement. The stated minimum allowable pumping rate of 82.4 gallons per minute is met through the simultaneous operation of both pumps.

1.

C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's", G. E. Topical Report NED0-10527, March 1972

2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NED0-10527, July 1972
3. J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2. " Exposed Cores",

'-Supplement 2 to NEDO-10527, January 1973 ,

IM5GRT E HOPE CREEK B 3/4 1-5 Amendment No.ll

l j 3/4.19 SPECIAL TEST EXCEPTIONS l

i i

j BASES A

3/4.10.1 PRIMARY CONTAINMENT INTEGRITY i

i the period when open vessel tests are being performed d PHYSICS TESTS.

I g WORTH MNIMIEER 3/4.10.2 ROD ,[00:NCE CONTROL SYSTE" In order to perform the tests required in the technical specifications i it is necessary to bypass the sequence restraints on . control rod movement. The i additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded period when these tests are being performed and that individual rod worths doduring th t

not exceed the values assumed in the safety analysis, 3/4.10.3 SHUTOOWN MARGIN DEMONSTRATIONS i

i Performance of shutdown margin demonstrations during open vessel testing j requires monitored and additional controlle.. reetrictions in order to ensure that criticality is properly These additional restrictions are specified in this LCO.

i 3/4.10.4 RECIRCULATION LOOPS -

i #

i This special test exception permits reactor criticality under no flow j POWER levels.and is required to perform certain PHYSICS TESTS while at low THERMA cdnditions l

5 3/4.10.5 OXYGEN CONCENTRATION i

deleted with the issuance of Amendment No. 35Lp.e mater 4l originally cont However, to maintain the i historical left blank. reference to this specification, this section has been intentionally i

1 3/4.10.6 TRAINING STARTUPS 4

' This special test exception permits training startups to be performed with

the reactor vessel depressurized at low THERMAL POWER and temperature while 1 controlling RCS temperature with one JtHR subsystem aligned in the shutdown i cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.
3/4.10.7 i .

SPECIAL INSTRUMENTATION - INITIAL CORE LOADING

]

i The material originally contained in Bases Section 3/4.10.7 was deleted l

with the issuance of Amendment No.14. However, to maintain the historical

.l referenc,e to this section, Bases Section 3/4.10.7 is intentionally left blank.

HOPE CREEK B 3/4 10-1

{ Amendment No. 35 1 DEC 18198g i . . _ , _ _ _ . _ __ _

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