ML20211K074
| ML20211K074 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/30/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Public Service Electric & Gas Co, Atlantic City Electric Co |
| Shared Package | |
| ML20211K080 | List: |
| References | |
| NPF-57-A-105 NUDOCS 9710090132 | |
| Download: ML20211K074 (17) | |
Text
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lt UNITED STATES y&
NUCLEAR REZULATCRY 62MMISS10N WASHINGTON. O.C. soteHoot
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PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 HOPE CREEK GENERATING STAT 1.qN AMENDHENT TO FACILITY OPERATING LICENSE Amendment No.105 License No. NPF-57 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Public Service Electric
& Gas Company (PSE&G) dated June 19, 1997, as su)plemented by letters dated July 30 and 31, 1997, complies wit 1 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; B.
The facility will operate in conformity with the appli-ton, the provisions of the Act, and the rules and regulations of the Commission; 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 coriucted 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 comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in th: attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:
97i0090132 970930 DR ADOCK 050003 4
-t-(2) Technical Saecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, es revised through Amendment No.105, and the Environ;.antal Protection Plan conta'ned in Appendix B, are hereby incorporated irto the license.
PSE&G shall operate the facility in accot.4ance with the Tschnical Specifications and the Environmental Protection Plan.
I In addition, paragraph 2.C.(14) to Facility Operating License No. NPF-57 l
1s amended as follows:
(14) Additional Conditions The Additional Conditions contained in AppendiA C, as revised through Amendment No.105 are hereby incorporated into this license.
PublicServiceElectricarJGasCompany-shalloperatethe facility in accordance with the Additional Conditions.
3.
The license amer:imnt is effective as of its date of issuance and shall be implemented w..,
' 60 days.
FOR THE NUCLEAR REGULATOR,Y ColWISSION
. Stolz, Dir tor r ject Directora I-t vision of Reactor Projects - I/II office of Nuclear Reactor Regulation
Attachment:
1.
rage 1 of Ap>endix C of License
- No. NPF-57 2.
Changes to tto Technical Specifications Date of Issuance: September 30, 1997
- Page 1 of Appendix C is attached, for convenience, for the composite license to reflect this change.
ATTACHMENT TO LICENSE AMENDMENT N0.105 l
FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 1.
Insert Appendix C, Page 1 Replace the following pa 2.
with the attached pages.ges of the Appendix "A" Technical Specifications The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
4 Remove Insert y
y i
xV xy xxt xxi B 2-6 8 2-6 3/4 1-4 3/4 1-4 3/4 1-11 3/4 1-11 3/4 1-13 3/4 1-13 3/4 1-16 3/4 1-16 3/4 1-16a 3/4 1-17 3/4 1-17 3/4 10-2 3/4 10-2 B 3/4 1-3 8 3/4 1-3 B 3/4 1-5 B 3/4 1-5 8 3/4 10-1 B 3/4 10-1 i
d e
-v
APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-57 Public Service Electric and Gas Company and Atlantic City Electric Company shall comply with the following conditions on the schedules noted below:
Amendment Additional Condit vn Impleinentation Number Date 97 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from March 21, to the appropriate documents, as describad in
- 1997, the licensee's application dated January 11, 1996, as supplemented by letters dated 4
February 26, May 22, June 27, July 12, December 23, 1996, and March 17, 1997, and evaluated in the staff's safety evaluation attached to this amendment.
103 The licensee shall relocate the list The amendn.ent of " Motor Operated Valves - Thermal Overload shall be Protected (BYPASSED)" from the Technical implemented Specifications (Table 3.8.4.2-1) to the Updated within 60 days Final Safety Analysis Report, as described in from September 16, the licensee's application dated July 7, 1997, 1997 and evaluated in i.hc staff's safety evaluation atta:.hed to this amendment.
105 The licensee M'. use the Banked The amendment Pattern Witd r m l System or an improved shall be version such a tae raduced Notch Worth implemented Procedure as described in the licensee's wittin 60 days application dated Juns 19, 1997, and from September 30, evaluated in the staff's safety evaluation 1997.
atta: bed to this amendment.
Amendment No. 92, M3,105
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i j-g 4-
. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRUMENTS 7
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...........................................................,n..,.....
SECTION E&QI 3/4.0 APPLICABILITY..............................................
-3/401 m
j 3/4.1 REACTIVITY CONTROL SYSTEMS l
3/4.1.1 SHUTDOWN MARGIN........................................
3/4 1-1 i
3/4.1.2 REACTIVITY ANOMALIES...................................
3/4 1-2 t.
1 3/4.1.3 CONTROL RODS Control Rod Operability................................
3/4 1-3 i
Control Rod Maximum Scram Insertion Times..............
3/4 1-6 I
Control Rod Average Scram Insertion Times..............
3/4 1-7 Four Control Rod Group Scram Insertion Times...........
3/4 1-8 j
Control Rod Scram Accumulators.........................
3/4 1-il-Control Rod Drive Coupling.............................
3/4 1-11 Control Rod Position Indication........................
3/4 1-13 l
Control Rod Drive Housing Support......................
3/4 1-15 l
1 i
.3/4.1.4 CONTROL ROD PROGRAM CONTROLS 1
1-Rod' Worth Minimizer....................................
3/4 1-16 Rod Sequence Control System (Deleted)..................
3/4 1-17 f l
Rod Block Monitor.....................
3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..........................
3/4 1-19 i
l Figure 3.1.5-1 Sodium Pentaborate Solution Volume / Concentration Requirements.....
'3/4 1-21 1
4 3 /4. 2 POWER DISTRIBUTI'-
LIMITS j
3/4.2.1 AVEKAGE PLANAR LINEAR HEAT GENERATION RATE.............
3/4 2-1 s
l.
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HOPE CREEK v
Amendment No.105 y-a i
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.11 R2SIDUAL HE.TT REMOVAL AND COOLANT CIRCULATION High Water Level.......................................
3/4 S-17 Low Water Level........................................
3/4 9-1B 3/4.10 SPECIAL TEST EXCEPTIQ]ig 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY...........................
3/4 10-1
..?
3/4.10.2 ROD WORTH MINIMIZER......................................
3/4 10-2 l l-3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS..........................
3/4 10-3 i
3/4.10.4 RECIRCULATION LOOPS.....................................
3/4 10-4 1
3/4.10.5 OXYGEN CONCENTRATION....................................
3/4 10-5 3/4.10.6 'IRAINING STARTUPS.......................................
3/4 10-6 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING..........
3/4 10-7 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING..................
3/4 10-8 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Co.Teentration...........................................
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 Liquid Waste Treatment..................................
3/4 11-6 Liquid Holdup Tanks........
3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate...............................................
3/4 11-8 Table 4.11.2.1.2-1 Radioactive Gaseous Waste Sampling and Analysis Program....
3/4 11-9 Dose - Noble Gases.......................................
3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium and Radionuclides in Particulate Form...............................
3/4 11-13 Gaseous Radwaste Treatment...............................
3/4 11-14.
Ventilation Exhaust Treatment System.....................
3/4 11-15 HOPE CREEK xv Amendment No.105
INDEX BASES SECTION PAgg 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.
B 3/4 10 1 3/4.10.2' ROD WORTH MINIMIZER.
B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS B 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION B 3/4.10-1 3/4.10.6 TRAINING STARTUPS.
B 3/4 10-1 3/4.10.7 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 EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration...
B 3/4 11-1
. Dose B 3/4 11-1 Liquid Radwaste Treathnent System B 3/4 11-2 Liquid Holdup Tanks.
B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS Dose Rate.
B 3/4 11-2 Dose - Noble Gases B 3/4 11-3 Dose - Iodine-131, Iodine-133. Tritium, and Radionuclides in Particulate Form.
B 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation-Exhaust Treatment Systems.
B 3/4 11-4 Main Condenser B 3/4 11 6 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 TOTAL DOSE B 3/4 11-5 HOPE CREEK xxi Amendment No.105
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2.2 LIMITING SAFETY SYSTEM SETTINGS EASES-l 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS j
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 j
parameter. The Trip Setpointe have been selected to ensure that the reactor i
core and reactor coolant system are prevented from exceeding their Safety l
Limits during normal operation and design basis anticipated operational occurrences and to assist 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 i
between each Trip Setpoint and the Allowable Value'is an allowance for instrument drift specifically allocated for.each trip in the safety analyses.
l 1
1.
Intermediate Ranoe Monitor. Neutron Flux - Hioh
)
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 IRN 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.
e The most aignificant source of reactivity changes during the power increase is due to control 1 rod withdrawal.
In order to ensure that the IRM i
provides the required protection, a range of rod withdrawal accidents have l
been analyzed. The results of these analyses are in section 15.4 of the:FSAR.
The mont savare case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken i
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 i
shutdown and peak power is. limited to 21%.of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal /gm. Based on this analysis, the_IRM provides protection against. local control rod _erroru and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
2.
Averaos Power Ranoe 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 s the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure i
at sero or low void content are minor and cold water from sources available during startup is not much colder than that,already in the system.
4
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Temperature coefficients are small and control rod patterns are constrained by
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the RWM. -Of:all the possible sources of reactivity input, uniform control rod j
- withdrawal.is the most probable cause of significant power increase.
1' 1-i k
HOPE CREEK.
B 2-6 Amendment No. 105 4
9-
l REACTIVITY CONTROL SYSTEMS LIMITING OONDITION 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.
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 />.
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 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
d.
With one scram discharge volume vent valve and/or one scrks discharge volume drain valve inoperable and open, restore the inoperable valve (s) 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 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) 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 the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent va?ves shall be demonstrated OPERABLE by 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 cycliag each valve through at least one complete cycle of full travel.
l 4.1.3.1.2 When above the low power setpoint of the RWM, all withdrawn control rods not required to have their directional control valves disarmed
- These valves may be closed intermittently for testing under administrative controls.
- May be rearmed intermittently, under administrative' control, to permit testing associated with restoring the control rod to OPERABLE status.
HOPE CREEK 3/4 1-4 Amendment No.105
- - ~ - -.
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REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE 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*.
4 I
ARTIQN:
- a. In OPERATIONAL CONDITION 1 and 2 with one control rod not. cot. pled to r
its associated drive mechanism, within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
1.
If permitted by the RWM, insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod, ands 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 accomplished on the first attempt or, if not permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and disarm the
- associated directional control valves ** either:
4 a)
Electrically, or b)
Hydraulically by clos.ng 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 net go to the overtravel position, or 2.
If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves ** either a)
Electrically, or 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 1
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.
l HOPE CREEK 3/4 1-11 Amendment No.105 i
2
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REACTIVITY CONTROL SYSTEMS l
CONTROL ROD POSITION INDICATION LIMITING CONDITION-FOR OPERATION 3.1.3.7 The control rod position indication system shall be OPERABLE.
APPLICABILITY. OPERATIONAL CONDITIONS 1, 2 and 5*.
7 ACTION:
l' i
a.
In OPERATIONAL CONDITION 1 or 2 with one or more control rod position indicators inoperable, within 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.
e:
Cetermine the position of the control rod by using an alternate method, or:
. a) Moving the control rod, by single notch movement, to a j.
position with an OPERABLE position indicator, i-b) Returning the control rod, by single notch novament, to its original position, and j
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 2.
Move the control rod to a position with.an OPERABLE position indicator, or_.
3.
When THERMAL POWER ist-1 a) Within the preset power level of the RWM, declare the control rod inoperable, j-b) Greater than the preset power level of-the RWM,; declare the control rod inoperable, insert the control rod and disarm the associated directional control. valves ** either:
i-1)
Electrically, or
'2)
Hydraulically by closing the drive' water and exhaust water isolation valves.
i
- Otherwise, be in at least HOT SHUTDOWN within the next 12:
hours.
I' 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.
Theprovisionsofspecification3.0.4arenot;applidable.
1:
c.:
l'
- At least each withdrawn control rod.
Not applicable to control rods removed j
per specification 3.9.10.1 or 3.9.10.2.
5
> **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 Amendment No.105 i
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REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS l
ROD WORTH MINIMIEER LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod worth minimiser (RWM) shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2", when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER, minimum allowable low power setpoint.
ACTION:
a.
With the RWM inoperable after the first 12 control rods are fully withdrawn, operation may continue provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control
- Console, b.
With the RWM inoperable before the first twelve (12) control rods are fully withdrawn, one startup per calendar year may be performed provided that the control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console, c.
Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the shutdown position.
SURVEILLANCE REQUIREMENTS 4.1 4.1 The RWM shall be demonstrat'sd OPERABLE:
a.
In OPERATIONAL COND LION 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 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 initiation when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod, s
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods
.is permitted for the purpose of' determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to
-criticality.
- see-special Test Exception 3.10.2.
HOPE CREEK 3/4 1-16 Amendment No.l@
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REACTIVITY CONTROL SYSTEMS-3/4.1.4 CONTROL ROD PROGRAM CONTROLS KOD WORTH MINIMIZER SURVEILLANCE REQUIREMENTS (CONTINUED) b.
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, by verifying the rod block function by demonstrating inability to a
withdraw an out-of-sequence control rod.
c.
In OPERATIONAL CONDITION 1 within one hour after RWM automatic initiation when reducing-TNERMAL POWER, by verifying the rod block
- function by_ demonstrating inability to withdraw an out-of-sequence control rod.
_ 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.
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' NOPE CREEK 3/4 1-16a Amendment No.105 w w. e.e.9-o
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REACTIVITY CONTROL SYSTEMS ROD SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION The material originally contained in section 3/4.1.4.2 was deleted with the issuance of Amendment No.-
However, to maintain numerical continuity between the succeeding sections and existing station procedural references to those Technical Specification sections, 3/4.1.4.2 has been intentimally lef t
- blank.
f NOPE CREEK 3/4 1-17 Amendment No.105 h,_
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SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH MINIMIEER LIMITING CONDITION FOR OPERATION.
j i
3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimiser (MWM) per specification 3.1.4.1 may be suspended for the following j
tests provided that control rod movement prescribed for this testing is verified by a second licensed' operator or other technically qualified member of the unit technical staff present at the reactor console l
a.-
shutdown margin demonstrations, specification 4.1.1.
I b.
Control rod scram, specification 4.1.3.2.
1 c.
Control rod friction measurements.
3 M,5 1,pABILITY: OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than j
or equal to 10% of RATED THERMAL POWER.
{
mil!2H8 With the requirements of the above specification not satisfied, verify that.
i the RWM is OPERABLE per Specifications 3.1.4.1.
l f
SURVEILLANCE REQUIREMENTS 4.10 2 When the-sequence constraints imposed by the RWM are bypassed, verify:
i -
a That movement of the control rods from 75% ROD DENSITY to the RWM low power setpoint is limited to the approved control rod withdrawal l
sequence during scram and friction tests.
That mova, ment of control rods during shutdown margin demonstrations b.
3 is limited to the prescribed sequence per specification 3.10.3.
c.
Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit
-technical staff.
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HOPE CREEK 3/4 10-2 Amendment No. 105
=_
REACTIVITY CONTROL SYSTEMS BASES 3 /4.1. 4 CONTROL,10p__EROGRAM CONTROLS 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 sequentes are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth l
which, if dropped at the design rate of the velocity limiter, could result in a peak entha3py of 280 cal /gm. Thus requiring the RWM to be OPERABLE when THERMAL POWEP is less than or equal to 10% of RATED THERMAL POWER provides adequate control.
The RWM provides automatic supervision to assure that out-of-sequence rods l
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.
Additional pertinent analysis is also contained in Amendment 17 to the Reference 4 topical report.
The RBM is designed to automatically prevent fuel damage in the event of 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.
HOPE CREEK B 3/4 1-3 Amendment No.105
i i
REACTIVITY CONTROL SYSTEMS BASES 5
rate, solution concentration or boron equivalent.to meet the ATWS Rule must i
not invalidate the original system design basis. Paragraph (c) (4) of 10 CFR SC.62 Jtates thatt
- Each boiling water reactor must have a Standby Liquid Control System j
(SLCS) with a minimum flow capacity and boron control equivalent in control capacity to 86 gallons per minute of 13 weight i
percent sodium pentaborate solution (natural boron enrichment)."
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)
The described minimum system parameters (82.4 gpm, 13.6 percent 1
concentration and r.stural 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.
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- 1. C. J.
Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's", G. E. Topical Report NEDO-10527, March 1972
- 2. C. J..Paone, R. C. Stirn and R.'M. Young, Supplement 1 to NEDO-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
- 4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A,
" General Electric Standard Application for Reactor Fuel".
HOPE CREEK B 3/4 1-5 Amendment No.105 l
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3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRIU The requirement for PRIhARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.
3/4.10.2 ROD WORTH MINIMIZER In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints ou control rod movement. The cdditional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
3/4110.3 SHUTDOWN fMRGIN DEMONSTRATIONS Performance of shutdown margin demonstrations during open vessel testing requires additional restrictions in order to ensure that criticality is properly monitored and controlled. These additional restriccions are
.pecified in this LCO.
3/4.10.4 RECIRCULATION LOOPS This special test exception permfts reactor criticality under no flow conditions and is required to perform certain PHYSICS TSSTS while at low THERMAL POWER levels.
3/4.10.5 OXYGEN CONCENTRATION The material originally contained in thie Technical Specification war deleted with the issuance of Amendment No. 35.
However, to maintain the historical reference to this specification, this section has been intentionally left blank.
3/4.10.6 TRAINING STARTUPS This special test exceptica permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RER subsystem aligned in the shutdown cooling mode in order to minimite contaminated water discharge to the radioactive waste disposal system.
3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING The material originally contained in Bases Section 3/4.10.7 was deleted with the issuance of Amendment No. 14.
However, to maintain the historical reference to this section, Bases Section 3/4.10.7 is intentionally left blank.
HOPE CREEK B 3/4 10-1 Amendment No.105
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