ML20096F691
| ML20096F691 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/07/1992 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20096F696 | List: |
| References | |
| NUDOCS 9205210306 | |
| Download: ML20096F691 (15) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _.
/phhtGu 'o UNITED STATES NUCLE AR REGULATORY COMMISSION n
{
j WASHINGTON, D. C. 20656 EUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. NPF-57 1.
The Nuclear Regulatory Comission (the Commission or the NRC) has f ound that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company (PSE&G) dated February 25. 1992 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(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 Commission'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 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 the attachment to this-license amen (ment, and paragraph 2.C.(2)llows:of Fac_ility Operating License.No. NPF-57 is hereby amended to read as fo (2) -Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A. as revised
-through Amendment No. 50, and.the Environmental Protection Plan.
contained in Appendix B, are hereby incorporated into the. license.
PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
92052103069205g7 DR -- ADOCK : 0500 4
2-3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h h. 0\\
Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation-
Attachment:
Changes to the Technical Specifications Date of Issuance: May 7, 1992 3
____.__.___m.___
)
ATTACliMENT TO LICENSE AMENDMENT NO. 50 FACIllTY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354.-
Replace the following pages of the f.ppendix "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 xiii xiii xiv xiv*
3/4 7-13 3/4 7-13 3/4 7-14 3/4 7-14 3/4 7-17 3/4 7-17*
3/4 7-17a 3/4 7-17b 3/4 7-18 3/4 7-18*
B 3/4 7-1 B 3/4 7-l*
B 3/4 7-2 B 3/4 7-2 B 3/4 7-3 8 3/4 7-3 8 3/4 7-4 B 3/4 7-4*
l
8 1
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTI,0N PAGE 3/4.7.3 FLOOD PROTECTION..........................................
3/4 7-9 Table 3.7.3-1 Perimeter Flood Doors.....................
3/4 7-10 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM.....................
3/4 7-11 3/4.7.5 SNUBCERS..................................................
3/4 7-13 Table 4.7.5-1 Snubber Visual Inspection Interval........
3/4 7-17a Figure 4.7.5-1 Sample Plan 2) for Snubber Functional Test.........................................
3/4 7-18 3/4.7.6 SEALED SOURCE CONTAMINATION...............................
3/4 7-19 3.4.7.7 MAIN TURBINE BYPASS SYSTEM................................
3/4 7-21 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C.
Sources-Operating....................................
3/4 8-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule........
3/4 8-10 A.C. Sources-Shutdown.....................................
3/4 8-11 3/4.8.2 D.C. SOURCES D.C. Sources-Operating.............................
3/4 8-12 Table 4.8. 2.1-1 Battery Surveillance Requiremunts.......
3/4 8-16 D.C. Sources-Shutdown.....................................
3/4 8-17 3/4.8.3 ONSITE POWER DISTRIBUTION-SYSTEMS Distribution - Operating..................................
3/4 8-18 Distribution - Shutdown...................................
3/4 8-21 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overentrent Protective Devices......................................
3/4 8-24 Table 3.8.4.1-1 Primary Containment-Penetration Conductor Overcurrent Protective Devices..........
3/4 8-26 Motor Ope sted Valve Thermal Overload Protection (Bypasswo)..............................................
3/4 8-30 HOPE CREEK xiii Amendment No. 50
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACC Table 3.8.4.2-1 Motor Operated Valves Thermal Overload Protection (Bypassed)...................
3/4 8-31 Motor Operated Valve Thermal Overload Protection (Not Bypassed)...........................
3/4 8-38 Table 3.8.4.3-1 Motor Operated Valves-Thermal Overload Protection (Not Bypassed)...............
3/4 8-39 Reactor Protection System Electric Powar Monitoring.......
3/4 8-40 Class 1E Isolation-Breaker Overcurrent Protection Devices (Breaker Tripped.by LOCA Signal)................
3/4 8 41 Table 3.8.4.5-1 Class 1E Isolation Breaker Overcurrent Protective Devices (Breaker Tripped by a LOCA Signa 1)...........................
3/4 8-42 Power Range Neutron Monitoring System Electric Power Monitoring..............................................
3/4 8-44 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH.......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................
3/4 9-3 3/4.9.3 CONTROL ROD POSIT 10N......................................
3/4 9-5 3/4.9.4 DECAY TIME......................................,...
3/4 9-6 2/4.9.5 COMMUNICATIONS....................
3/4 9-7 3/4.9.6 REFUELIt4G. PLATFORM........................................
3/4 9-8 3/4.9.7 CRANE TRAVEL - APENT FUEL STORAGE.P00L....................
3/4 9-10 3/4.9.8 WATER ' LEVE L - RE ACTOR VESSE L..............................
3/4 9-11 3/4.9.9 WATER ' LEVEL - SPENT FUEL STORAGE P00L.....................
3/4 9-12 3/4.9.10 -CONTROL ROD REMOVAL Single Control Rod Remova1................................
3/4 9-13 Multiple Control Rod Remova1..............................
3/4 9-15
. HOPE CREEK xiv
PLANT SYSTEMS 3/* 7. 5, SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.5 All snubbers shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1,2, and 3.
OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS.
ACTION:
With one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.5.g on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
SURVEILLANCE REQUIREMENTS 4.7.5 Each snubber shall be demonstrated OPERABLE by performance of the i
following augmented inservice inspection program and the requirements of Specification 4.0.5.
a.
Inspection Types z
As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.
b.
Visual Inspecti ji Snubbers are categorized as inaccessible or accessible during reactor operation.
Each of these categories (inaccessinle and accessible) may be intpected independently according to the schedule determined by Table 4.7.5-1.
The. visual inspection interval for each type of snubber shall be determined based upon the criteria provided in -
Table 4.7.5-1 and the first inspection interval determined using this criteria chall be based upon the previous inspection interval as established by the requirements in effect before amendment 50.
3 HOPE CREEK 3/4-7-13 Amendment Noi S0-
4 o
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
(
c.
Visual Inspection Acceptance Criteria l
Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are secure.
Snubbers which appear inoperable as a resJ1t of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual' inspection period, providing that:
(1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type on thdt system that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifications 4.7.4.f.
A review and evaluation shall be Per-formedanddocumentedtojustify:continuedoperationwithanunacc6pt-able snubber.
If continued operation cannot be dustified, the snubber shall be declared inoperable and the ACTION requ<raments shall be met.
d.
Transient Event Inspection An inspection shall be performed of all snubbers-ettached to sections of systems that have experienced unexpected, potentially-damaging transients, as determined from a review of operational data or a visual-inspection or the systems, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'for accessible systems and 6 months for inaccessible sys. ems following this deter-mination.
In addition to satisfying the vicual inspection acceptance criteria. freedom-of-motion of mechanical snubbers shall be verified using at least one of the following-(1) ma,Jally induced snubber movement. or (2) evaluation of in-p hce snubber piston setting.
HOPE CREEK 3/4.7-14
-Amendment'No. 50
s PLANT SYSTEMS SURVEILLANCE REQUIREMENTS ' Continued)
OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
For the snubbers found inoperable, an engineering evaluation shall be performed on the components tJ which the inoperable snubbers are attached.
The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.
If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of-the same type subject to the same defect shall be functionally tested.
This testing requirement shall be independent of the requirenents stated in Specification 4.7.5.e. for snubbers not meeting the functional test acceptance criteria.
h.
Functional Testino of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced.
Replacement snubbers and snubbers which have repairs which might affect the functional test result shall be tested to meet the functional test criteria before installation in the unit.
Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
i.
Snubber Service Life Replacement Program The service life of all snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections.
The maximum expected service life for various seals, springs, and other critical parts shall be extended or shortened based on moni-tored test results and failure history.
Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE.
The parts replacements shall be documented and the documentation shall be retained in accordance with Specification 6.10.3.
HOPE CREEK 3/4 7-17 4
.i TABLE 4.7.5-1 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C or Category Extend Repeat Reduce Interval Interval Interval (Notes 1 (Notes 3 (Notes 4 (Notes 5 and 2) and 6) and 6) and 6) 1 0
0 1
80 0
0 2
100 0
1 4
150 0
3 8
200 2
5 13 300 5
12 25 400 8
18 36 500 12 24 48 750 20 40 78 1000 or 29 56 109 greater Note 1:
The next visual inspection interval for a snubber popu-14 tion or category size shall be determined cased upon the previous inspec' ion interval and the number of un-acceptable snubbers found during that interval.
Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible.
These categories may be examined separately or jointly.
How-ever, that decision shall be made and documented before any inspection and shall serve as the basis upon which the next inspection interval for that category is determined.
Note 2:
Interpolatioi, between population or category sizes and the number of unacceptable snubbers is permissible.
Use the next lower integer for the value of the limit for Columns A, B, or C if that integer includes a fractional value of unacceptable snubbers as 6stermined by interpolation.
Note 3:
If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.
(Continued)
HOPE CREEK 3/4 7-17a Amendment No. 50
i.
TABLE 4.7.5-1 (Continued)
SNUBBER VISUAL INSPECTION INTERVAL Note 4:
If the number of unacceptable snubbers is equal ~to or less than the number in Column B but greater than the number in Column A, the next inspection interval shall be the same as the previous interval.
Note 5:
If the number of unacceptable snubbers is equal to or_ greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval.
However, if the number of unaccept-able snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced propnt-tionally by interpolation, that is:
I
=
1 I
. 1/3. U - B.
t 0
O C-B where:
11 = next inspection inteaal
-IO = previous inspection interval U = number of unacceptable snubbers found t
during the previous inspection interval B = number in Column B-C = number in Column C Note 6:
The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.
HOPE CREEK 3/4'7-17b Amendment No. 50-.
1
10 9
8
/
REJECT
,/
e 01 /
og**# 0/
Cs s.o /
4
/
CONTINUE
/
TIESTING
/
/
/
A*/c1/
c.
ACCEPT 1
/
/
0 10 h0 30 40 50 40 70 80 90 100 N
SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST Figure 4,7.5-1 HOPE CREEK 3/4 7-18
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 SERVICE WATER SYSTEMS The OPERABILITY of the station service water and the safety auxiliaries cooling systems ensures that sufficient m oling capacity is available for con-tinued operation of the SACS and its associated safety-related equipment during normal and accident conditions.
The redundant cooling capacity of these sys-tems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM The OPERABILITY of the control room emergency filtration system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all design basis accident conditions.
Continuous operation of the system with the heaters and humidity control instruments OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radia-tion exposure to personnel occupying the control room to 5. rem or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",10 CFR Part 50.
3/4.7.3 FLOOD PROTECTION The requirement for flood protection ensures that facility flood protection features are in place in the event of flood conditions.
The limit of elevation 10,5' Mean Sea Level is based on the elevation at which facility flood protection features provide protection to safety related equipment.
3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the Emergency Core Cooling System equipment.
The RCIC system is conservatively required to be OPERABLE whenever reactor steam dome pressure exceeds 150 psig.
This pressure is substantially below that for shich the RCIC system can provide adequate core cooling for events requiring the RCIC system.
The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel steam dome pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.
With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCL system and justifies the specified-14 day out-of-service period.
HOPE CREEK B 3/4 7-1 i
7--
PLANT SYSTEMS BASES REACTOR CORE ! % ATION COOLING SYSTEM (Continued)
The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.
3/4.7.5 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.
Snub-bers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety related system.
Snubbers are classified and grouped by design and manufacturer but not by size.
For example, mechanical snubbers utilizing the same design features of the 2-kip,10 kip, and 100-kip capacity manufactured by Company "A" are of the same type.
The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.
A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50.
The accessibility of each snubber shall be determined and approved by the Plant Operations Review Committee.
Tha determination shall be based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g.,
temperature, atmosphere, location, etc.), and the recoraendations of Regulatory Guide 8.0 and 8.10.
The addition or deletion of any snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.
The visual inspection frequency is based upon maintaining a constant-level of snubber protection to each safety-related system.
Therefore, the required inspection interval is based on the number of unacceptable snubbers found c
during the previous inspection in proportion to the sizes of the various snubber populations or categories.
This inspect bn schedule is based on the guidance provided in Generic letter 90-09.-
In order to establish the inspec-tion frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is con-stant with time and that the failure of any snubber on that system could cause the system to be unprotected and to result in failure during an assumed HOPE CREEK B 3/4 7-2 Amendment No. 50
PLANT SYSTEMS BASES
$NUBBERS (Continued) initiating event.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.
- However, the results of such early inspections performed before the original required time interval has elasped (nominal time less 25%) may not be used to lengthen the required inspection interval.
Any inspection whose results required a shorter inspection interval will override the previous schedule.
The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers.
To provide assurance-of snubber-functional reliability one of three functional testing methods is used with the stated acceptance criteria:
1.
Functionally test 10% of a type of snubber with an additional 10%
tested for each functional testing failure, or 2.
Functionally test a sample size and determine sample acceptance or rejectionusingFigure4.7.5-1,or 3.
Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
Figure 4.7.5-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in Quality Control and Industrial Statistics" by Acheson J.'
Duncan.
. Permanent or other exemptions from the surveillance program for individual snubbers may he granted by.the Commission if a justifiable basis for exemption is presented ahd: if applicable, snubber life destructive. testing was performed to qualify the snubbers for tho applicable design conditions at either the com-pletion of their fabrication or at a subsequent date.
Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
The service life of a snubber is evaluated via manufacturer input and information through consideration of_the' snubber service conditions and asso-ciated installation and maintenance records (i.e., newly installed snubber, seal replaced, spring replaced, in high radiatica area, in high temperature area,etc.).
The requirement-to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.
These records will provide-statis -
tical bases for future consideration of snubber service life.
HOPE CREEK B 3/4 7-3 Amendment No. 50
PLANT SYSTEMS BASES 3/4.7.6 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.
1 Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of Jamage to a source in that group.
Those sources which are frequently hand 1ed are required to be tested more often than those which are not.
Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7.7 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system i; required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis for FSAR Chapter 15.
e HOPE CREEK B 3/4 7-4 1